IRDF2002 in metrology format cross sections (050112) 6000 0 0 3.00600E+3 5.96340E+0 0 0 34 10 325 1451 0.0 0.0 0 0 0 6 325 1451 1.00000E+0 2.00000E+7 0 0 10 2002 325 1451 3.00000E+2 0.0 1 0 237 3 325 1451 3-Li- 6 LANL EVAL-APR89 G.M.HALE, P.G.YOUNG 325 1451 DIST-Feb2004 325 1451 ----IRDF-2002 MATERIAL 325 325 1451 -----INCIDENT NEUTRON DATA 325 1451 ------ENDF-6 FORMAT 325 1451 ****************************************************************** 325 1451 3-LI - 6 LANL EVAL-APR89 G.M.HALE, P.G.YOUNG 325 1451 DIST-SEP91 REV1-JUL91 19910806 325 1451 ----ENDF/B-VI MATERIAL 325 REVISION 1 325 1451 ****************************************************************** 325 1451 3-LI - 6 LANL EVAL-APR89 G.M.HALE, P.G.YOUNG 325 1451 DIST-FEB91 910201 325 1451 ----ENDF/B-VI MATERIAL 325 325 1451 *****************EXTRACT FOR SPECIAL PURPOSE FILE***************** 325 1451 DOSIMETRY 325 1451 ****************************************************************** 325 1451 891221 325 1451 ----ENDF/B-VI MATERIAL 325 325 1451 325 1451 ****************************************************************** 325 1451 325 1451 ***************************************************************** 325 1451 ***************************************************************** 325 1451 325 1451 MOD1 OF ENDF/B-VI 325 1451 325 1451 The following revisions were made for MOD1 of ENDF/B-VI: 325 1451 325 1451 1. MF=1,MT=451 - Comments were added regarding estimated 325 1451 (expanded) covariance for the Standards Cross Sections. 325 1451 2. MF=3,MT=53 - LF flag and Q-value corrected. 325 1451 325 1451 ***************************************************************** 325 1451 ****************************************************************** 325 1451 325 1451 ENDF/VI EVALUATION 325 1451 G. M. Hale and P. G. Young 325 1451 325 1451 MAJOR CHANGES FROM VERSION V OF ENDF/B ARE: 325 1451 325 1451 1. Inclusion of the ENDF/B-VI standard (n,t) cross section 325 1451 from the simultaneous standards analysis (ca85) over the 325 1451 energy range thermal to 1 MeV. 325 1451 2. Replacement of all major cross sections and elastic angular 325 1451 distributions at energies between 10^-5 eV and 3 MeV with 325 1451 results from the R-matrix analysis performed in conjunction 325 1451 with the simultaneous standards analysis. 325 1451 3. Revision of the elastic cross sections and angular distri- 325 1451 butions at energies between 3 and 20 MeV to match recent 325 1451 experimental data, resulting in a general decrease of the 325 1451 elastic cross section in this energy range. 325 1451 4. Revision of the (n,n')d cross sections to account for 325 1451 recent measurements, resulting in a general increase in 325 1451 the total (n,n')d cross section that tends to offset the 325 1451 decrease in the elastic cross section and maintain about 325 1451 the same total cross section as before. 325 1451 325 1451 325 1451 ****************************************************************** 325 1451 325 1451 STANDARDS COVARIANCES 325 1451 325 1451 Phase 1 reviewers of the ENDF/B-VI standards cross sections have 325 1451 expressed the concern that the uncertainties resulting from the 325 1451 combination of R-matrix and simultaneous evaluations might have 325 1451 led to uncertainties that are too small. As a result, the 325 1451 Standards Subcommittee produced (at the May, 1990 CSEWG meeting) 325 1451 a set of expanded covariance estimates for the standard cross 325 1451 section reactions. These uncertainties are estimates such that 325 1451 if a modern day experiment were performed on a given standard 325 1451 cross section using the best techniques, approximately 2/3 of 325 1451 the results should fall within these expanded uncertainties. The 325 1451 expanded uncertainties for the Li-6(n,t) cross section are given 325 1451 in the following table and are compared to values from the 325 1451 combined output of the standards covariance analysis: 325 1451 325 1451 Energy Range Estimated Uncertainty Combined Analysis 325 1451 (keV) (percent) (percent) 325 1451 325 1451 1.0E-08 - 0.1 0.3 0.14 325 1451 0.1 - 1.0 0.5 325 1451 1.0 - 10. 0.7 0.14 325 1451 10. - 50. 0.9 325 1451 50. - 90. 1.1 0.25 325 1451 90. - 150 1.5 325 1451 150 - 450 2.0 0.29 325 1451 450 - 650 5.0 325 1451 650 - 800 2.0 0.36 325 1451 800 - 1000 5.0 325 1451 325 1451 **************************************************************** 325 1451 325 1451 325 1451 mf=2 --------- resonance parameters ---------------------------- 325 1451 325 1451 mt=151 effective scattering radius = 2.31175e-13 cm. 325 1451 325 1451 mf=3 --------- smooth cross sections --------------------------- 325 1451 325 1451 the 2200 m/s cross sections are as follows: 325 1451 mt=1 sigma = 941.6928 barns 325 1451 mt=2 sigma = 0.67157 barns 325 1451 mt=102 sigma = 0.03850 barns 325 1451 mt=105 sigma = 940.9827 barns 325 1451 325 1451 mt=105 (n,t) cross section 325 1451 below 3 mev, values are taken from the r-matrix analysis, 325 1451 which includes (n,t) measurements from re78, la78, br77, 325 1451 ov74, and ba75. between 3 and 5 mev, the values are 325 1451 based on ba75, and at higher energies are taken from the 325 1451 evaluation of pe64, extended to 20 mev considering the 325 1451 data of ke58. 325 1451 325 1451 325 1451 325 1451 mf=33----------cross section covariances------------------------- 325 1451 (to be added later) 325 1451 325 1451 the relative covariances for mt=1,2, and 105 below 4 mev are 325 1451 given in file 33. they are based on calculations using the co- 325 1451 variances of the r-matrix parameters in first-order error 325 1451 propogation. 325 1451 mt=1 total 325 1451 relative covariances entered as nc-type sub-subsection, 325 1451 implying that they are to be constructed from those for 325 1451 mt=2 and 105. they are not intended for use at energies 325 1451 above 4 mev. 325 1451 mt=2,105 elastic and (n,t) 325 1451 relative covariances among these two cross sections are 325 1451 entered explicitly as ni-type sub-subsections in the 325 1451 lb=5 (direct) representation at energies below 4 mev. 325 1451 although values for the 3.95-4.05 mev bin are repeated 325 1451 in a 4-20 mev bin, the covariances are not intended for 325 1451 use at energies above 4 mev. 325 1451 325 1451 325 1451 --------------------- references -------------------------------- 325 1451 325 1451 ab70 U.Abbondanno, Nuo.Cim. A166,139(1970). 325 1451 aj74 f.ajzenberg-selove and t.lauritsen, nucl. phys. a227,55 325 1451 (1974). 325 1451 ar64 a.h.armstrong, j.gammel, l.rosen, and g.m.frye, nucl. phys. 325 1451 52,505 (1964). 325 1451 as63 v.j.ashby et al, phys. rev. 129,1771 (1963). 325 1451 ba53 m.e.battat and f.l.ribe, phys.rev. 89,80 (1953). 325 1451 ba63 r.batchelor and j.h.towle, nucl. phys. 47,385 (1963). 325 1451 ba65 r.bass, c.bindhardt, and k.kruger, eandc(e)-57u (1965). 325 1451 ba75 c.m.bartle, proc. conf. on nuclear cross sections and 325 1451 technology, vol.2,688 (1975), and private communication 325 1451 (1976). see also nucl. phys. a330, 1 (1979). 325 1451 be75 besotosnyj et al., yk-19, 77 (1975). 325 1451 br77 r.e.brown,g.g.ohlsen,r.f.haglund, and n.jarmie, phys. rev. 325 1451 16c, 513 (1977). 325 1451 ca85 a.d.carlson,w.p.poenitz,g.m.hale, and r.w.peele, nuclear 325 1451 data for basic and applied science (santa fe, n.m.), 1429 325 1451 (1985). 325 1451 co67 j.a.cookson and d.dandy, nucl. phys. a91,273 (1967). 325 1451 co82 h.conde,t.andersson,l.nilsson, and c.nordborg, nuclear data 325 1451 for science and technology (antwerp, belgium), 447 (1982). 325 1451 de73 F.Demanins et al., infn/be-73 (1973). 325 1451 dr82 m.drosg,d.m.drake,r.a.hardekopf, and g.m.hale, la-9129-ms 325 1451 (1982). 325 1451 dr85 m.drosg et al., santa fe conf.1, 145(1985). 325 1451 fo71 d.g.foster and d.w.glasgow, phys. rev. c3,576 (1971). 325 1451 fr54 g.m.frye, phys. rev. 93,1086 (1954). 325 1451 go72 c.a.goulding and p.stoler, eandc(us)-176u,161 (1972). 325 1451 ha75 j.a.harvey and n.w.hill, nuclear cross sections and 325 1451 technology (washington, d.c.), 244 (1975). 325 1451 ha84 g.m.hale, nuclear standard reference data (geel,belgium) 325 1451 iaea tecdoc-335, 103 (1984). describes preliminary analysis. 325 1451 ho68 j.c.hopkins,d.m.drake, and h.conde, nucl. phys. a107,139 325 1451 (1968), and j.c.hopkins, d.m.drake, and h.conde, la-3765 325 1451 (1967). 325 1451 ho79 h.h.hogue et al., n.s.&e. 69, 22 (1979). 325 1451 ju73 e.t.jurney, lasl, private communication (1973). 325 1451 ke58 r.d.kern and w.e.kreger, phys. rev. 112, 926 (1958). 325 1451 ke79 j.d.kellie,g.p.lamaze, and r.b.schwartz, nuclear cross 325 1451 sections for technology (knoxville, tn.), 48 (1979). 325 1451 kn77 h.h.knitter,c.budtz-jorgensen,m.mailly, and r.vogt, eur- 325 1451 5726e (1977). 325 1451 kn79 h.d.knox,r.m.white, and r.o.lane, n.s.&e. 69, 223 (1979). 325 1451 kn83 h.h.knitter,c.budtz-jorgensen,d.l.smith, and d.marletta, 325 1451 n.s.&e. 83, 229(1983). 325 1451 la61 r.o.lane,a.s.langsdorf,j.e.monahan, and a.j.elwyn, ann. 325 1451 phys.12, 135 (1961). 325 1451 la78 g.p.lamaze,o.a.wasson,r.a.schrack, and a.d.carlson, n.s.&e. 325 1451 68, (1978). 325 1451 li80 p.w.lisowski et al., la-8342 (1980). 325 1451 ma69 d.s.mather and l.f.paine, awre-o-47/69 (1969). 325 1451 me65 f.merchez,n.v.sen,v.regis, and r.bouchez, compt. rend. 260, 325 1451 3922 (1965). 325 1451 ov74 j.c.overley,r.m.sealock, and d.h.ehlers, nucl. phys. a221, 325 1451 573 (1974). 325 1451 pe64 e.d.pendlebury, awre-o-60/64 (1964). 325 1451 pr69 g.presser et al., nuc.phys. a131, 679(1969). 325 1451 re78 c.renner,j.a.harvey,n.w.hill,g.l.morgan, and k.pusk, bull. 325 1451 am. phys. soc. 23, 526 (1978). 325 1451 sa82 e.t.sadowski,h.knox,d.a.resler, and r.o.lane, bap 27,624(c5) 325 1451 (1982). 325 1451 sm77 a.b.smith,p.guenther,d.havel, and j.f.whalen, anl/ndm-29 325 1451 (1977). 325 1451 sm82 a.b.smith,p.t.guenther, and j.f.whalen, nucl. phys. a373, 325 1451 305 (1982). 325 1451 wo62 c.wong,j.d.anderson, and j.w.mcclure, nucl. phys. 33,680 325 1451 (1962). 325 1451 ****************************************************************** 325 1451 325 1451 325 1451 325 1451 325 1451 ****************************************************************** 325 1451 325 1451 ***************** Program LINEAR (VERSION 2002-1) *************** 325 1451 For All Data Greater than 1.0000E-10 barns in Absolute Value 325 1451 Data Linearized to Within an Accuracy of .100000000 per-cent 325 1451 ***************** Program SIGMA1 (VERSION 2002-1) *************** 325 1451 Data Doppler Broadened to 300.000000 Kelvin 325 1451 for All Data Greater than 1.0000E-10 barns in Absolute Value 325 1451 Data Linearized to Within an Accuracy pf .100000000 per-cent 325 1451 ***************** Program FIXUP (Version 2002-1) **************** 325 1451 Corrected ZA/AWR in All Sections-----------------------------Yes 325 1451 Corrected Thresholds-----------------------------------------Yes 325 1451 Extended Cross Sections to 20 MeV----------------------------No 325 1451 Allow Cross Section Deletion---------------------------------No 325 1451 Allow Cross Section Reconstruction---------------------------No 325 1451 Make All Cross Sections Non-Negative-------------------------Yes 325 1451 Delete Energies Not in Ascending Order-----------------------Yes 325 1451 Deleted Duplicate Points-------------------------------------Yes 325 1451 Check for Ascending MAT/MF/MT Order--------------------------Yes 325 1451 Check for Legal MF/MT Numbers--------------------------------Yes 325 1451 Allow Creation of Missing Sections---------------------------No 325 1451 Allow Insertion of Energy Points-----------------------------No 325 1451 Create Uniform Energy Grid-----------------------------------No 325 1451 Delete Section if Cross Section =0 at All Energies-----------Yes 325 1451 ***************** Program GROUPIE (VERSION 2002-1) ************** 325 1451 Unshielded Group Averages Using 640 Groups 325 1451 Weighting Spectrum: Flat (Constant) Spectrum 325 1451 1 451 244 2 325 1451 3 105 217 1 325 1451 33 105 33 1 325 1451 325 1 0 325 0 0 3.00600E+3 5.96340E+0 0 0 0 0 325 3105 4.78380E+6 4.78380E+6 0 0 1 641 325 3105 641 1 325 3105 .000100000 14794.9461 .000105000 14449.9534 .000110000 14122.2061 325 3105 .000115000 13820.1942 .000120000 13467.1458 .000127500 13075.4530 325 3105 .000135000 12717.5721 .000142500 12388.6847 .000150000 12032.7367 325 3105 .000160000 11661.9312 .000170000 11324.5439 .000180000 11014.8459 325 3105 .000190000 10726.6978 .000200000 10463.6234 .000210000 10216.1135 325 3105 .000220000 9986.21324 .000230000 9772.62292 .000240000 9521.91859 325 3105 .000255000 9246.40272 .000270000 9033.79728 .000280000 8797.12696 325 3105 .000300000 8508.50325 .000320000 8247.16668 .000340000 8007.82467 325 3105 .000360000 7787.37574 .000380000 7586.13528 .000400000 7376.33068 325 3105 .000425000 7161.64277 .000450000 6965.50583 .000475000 6785.66231 325 3105 .000500000 6616.52235 .000525000 6461.94757 .000550000 6315.93048 325 3105 .000575000 6180.13284 .000600000 6040.94642 .000630000 5897.75751 325 3105 .000660000 5766.73697 .000690000 5641.05202 .000720000 5507.32050 325 3105 .000760000 5363.93249 .000800000 5230.80234 .000840000 5108.78635 325 3105 .000880000 4992.81623 .000920000 4886.50890 .000960000 4784.70637 325 3105 .001000000 4679.19964 .001050000 4569.04414 .001100000 4465.76191 325 3105 .001150000 4370.59678 .001200000 4258.33842 .001275000 4134.96155 325 3105 .001350000 4022.16216 .001425000 3916.88925 .001500000 3805.04153 325 3105 .001600000 3688.10030 .001700000 3581.34573 .001800000 3482.60854 325 3105 .001900000 3392.40441 .002000000 3308.77765 .002100000 3230.31157 325 3105 .002200000 3158.44480 .002300000 3089.72099 .002400000 3011.30697 325 3105 .002550000 2924.17510 .002700000 2856.02126 .002800000 2781.96182 325 3105 .003000000 2690.80549 .003200000 2607.85391 .003400000 2531.98785 325 3105 .003600000 2462.79638 .003800000 2398.91608 .004000000 2332.33692 325 3105 .004250000 2264.72689 .004500000 2203.02081 .004750000 2145.35559 325 3105 .005000000 2092.44576 .005250000 2043.40654 .005500000 1997.05482 325 3105 .005750000 1954.57647 .006000000 1909.96714 .006300000 1865.38391 325 3105 .006600000 1823.18871 .006900000 1784.04467 .007200000 1741.48065 325 3105 .007600000 1695.96373 .008000000 1654.34447 .008400000 1615.27928 325 3105 .008800000 1578.91854 .009200000 1545.13798 .009600000 1513.00017 325 3105 .010000000 1479.73644 .010500000 1444.57603 .011000000 1412.42313 325 3105 .011500000 1381.75051 .012000000 1346.59615 .012750000 1307.68584 325 3105 .013500000 1271.64181 .014250000 1238.56317 .015000000 1203.27730 325 3105 .016000001 1166.23862 .017000001 1132.28052 .017999999 1101.28202 325 3105 .018999999 1072.79317 .020000000 1046.06875 .021000000 1021.61496 325 3105 .022000000 998.572690 .023000000 977.099923 .024000000 951.867476 325 3105 .025500000 924.650334 .027000001 903.085714 .028000001 879.664218 325 3105 .029999999 850.841313 .032000002 824.584971 .034000002 800.612528 325 3105 .035999998 778.732778 .037999999 758.549586 .039999999 737.487875 325 3105 .042500000 716.110259 .045000002 696.571444 .047499999 678.366928 325 3105 .050000001 661.623345 .052499998 646.130877 .055000000 631.464013 325 3105 .057500001 618.035173 .059999999 603.916290 .063000001 589.802419 325 3105 .066000000 576.497152 .068999998 564.095301 .071999997 550.648650 325 3105 .075999998 536.246480 .079999998 523.089245 .083999999 510.719324 325 3105 .088000000 499.211911 .092000000 488.553833 .096000001 478.374129 325 3105 .100000001 467.864268 .104999997 456.740352 .109999999 446.578283 325 3105 .115000002 436.871661 .119999997 425.753899 .127499998 413.473559 325 3105 .135000005 402.085828 .142499998 391.626836 .150000006 380.471804 325 3105 .159999996 368.744950 .170000002 358.015375 .180000007 348.198886 325 3105 .189999998 339.184378 .200000003 330.720268 .209999993 322.982244 325 3105 .219999999 315.682189 .230000004 308.885854 .239999995 301.048031 325 3105 .254999995 292.298232 .270000011 285.625887 .280000001 278.184188 325 3105 .300000012 269.039086 .319999993 260.740769 .340000004 253.189079 325 3105 .360000014 246.269336 .379999995 239.820982 .400000006 233.210767 325 3105 .425000012 226.472440 .449999988 220.243253 .474999994 214.511408 325 3105 .500000000 209.256293 .524999976 204.282890 .550000012 199.729453 325 3105 .574999988 195.403926 .600000024 190.991063 .629999995 186.517846 325 3105 .660000026 182.279486 .689999998 178.400407 .720000029 174.092531 325 3105 .759999990 169.580921 .800000012 165.402550 .839999974 161.482057 325 3105 .879999995 157.896838 .920000017 154.457639 .959999979 151.313264 325 3105 1.00000000 147.924471 1.04999995 144.457738 1.10000002 141.215388 325 3105 1.14999998 138.163257 1.20000005 134.661002 1.27499998 130.732247 325 3105 1.35000002 127.134297 1.42499995 123.849352 1.50000000 120.293925 325 3105 1.60000002 116.585459 1.70000005 113.209752 1.79999995 110.125223 325 3105 1.89999998 107.244386 2.00000000 104.595170 2.09999990 102.145940 325 3105 2.20000005 99.8337043 2.29999995 97.7171215 2.40000010 95.2040034 325 3105 2.54999995 92.4334553 2.70000005 90.3148276 2.79999995 87.9381155 325 3105 3.00000000 85.0523601 3.20000005 82.4380221 3.40000010 80.0560785 325 3105 3.59999990 77.8616733 3.79999995 75.8730582 4.00000000 73.7205775 325 3105 4.25000000 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32533105 7.915241-7 9.299034-6 8.280879-6 5.098584-6 1.338568-6 1.841448-6 32533105 1.270541-5 1.262244-5 7.390554-6 4.714562-6 1.835564-5 1.767288-5 32533105 9.582569-6 3.027060-5 1.780456-5 3.434466-5 32533105 32533 0 325 0 0 0 0 0 5.00000E+3 1.07191E+1 -1 0 34 0 500 1451 0.0 0.0 0 0 0 6 500 1451 1.00000E+0 2.00000E+7 0 0 10 2002 500 1451 3.00000E+2 0.0 1 0 75 2 500 1451 5-B - 0 NDS IAEA-JAN04 500 1451 DIST-Feb2004 500 1451 ----IRDF-2002 500 1451 -----INCIDENT NEUTRON DATA 500 1451 ------ENDF-6 FORMAT 500 1451 ***************************************************************** 500 1451 *****************EXTRACT FOR SPECIAL PURPOSE FILE***************** 500 1451 DOSIMETRY Assembled at NDS from the ENDF/B-VI evaluations for 500 1451 5-B -10 and 5-B -11. The files were processed 500 1451 through the 2002 pre-processing codes LINEAR,RECENT, 500 1451 SIGMA1 and FIXUP to produce pointwise files prior to 500 1451 input to MIXER. 500 1451 500 1451 MIXER INPUT FILE 500 1451 500 1451 Boron DENSITY 2.34 G/CC. CONSTITUENTS: 10-19.8 11-80.2 500 1451 FIXUP.out 500 1451 MIXER.OUT 500 1451 5000 500 3 1 (ID FOR COMBINED REACTION) 500 1451 5010 1 0.429086 500 1451 5011 1 1.910914 500 1451 (BLANK LINE TERMINATES LIST) 500 1451 ****************************************************************** 500 1451 500 1451 5-B - 10 LANL EVAL-NOV89 G.M.HALE, P.G.YOUNG 500 1451 DIST-SEP91 REV1-JUL91 19910806 500 1451 ----ENDF/B-VI MATERIAL 525 REVISION 1 500 1451 ***************************************************************** 500 1451 5-B - 11 LANL EVAL-MAY89 P.G.YOUNG 500 1451 DIST-SEP 1 REV1- 20010926 500 1451 ----ENDF/B-VI MATERIAL 528 500 1451 ***************** Program MIXER (VERSION 2002-1) **************** 500 1451 Boron DENSITY 2.34 G/CC. CONSTITUENTS: 10-19.8 500 1451 11-80.2 500 1451 ---------------------------------------- 500 1451 Composition 500 1451 ---------------------------------------- 500 1451 Isotope MF MT Atom-Fract Grams/cc 500 1451 ---------------------------------------- 500 1451 5010 3 1 .198003533 .429086000 500 1451 5011 3 1 .801996467 1.91091400 500 1451 ---------------------------------------- 500 1451 5000 3 1 1.00000000 2.34000000 500 1451 ---------------------------------------- 500 1451 Composition 500 1451 ---------------------------------------- 500 1451 Isotope MF MT Atom-Fract Grams/cc 500 1451 ---------------------------------------- 500 1451 5010 3 101 .198003533 .429086000 500 1451 5011 3 101 .801996467 1.91091400 500 1451 ---------------------------------------- 500 1451 5000 3 101 1.00000000 2.34000000 500 1451 ---------------------------------------- 500 1451 Composition 500 1451 ---------------------------------------- 500 1451 Isotope MF MT Atom-Fract Grams/cc 500 1451 ---------------------------------------- 500 1451 5010 3 102 .198003533 .429086000 500 1451 5011 3 102 .801996467 1.91091400 500 1451 ---------------------------------------- 500 1451 5000 3 102 1.00000000 2.34000000 500 1451 ---------------------------------------- 500 1451 Composition 500 1451 ---------------------------------------- 500 1451 Isotope MF MT Atom-Fract Grams/cc 500 1451 ---------------------------------------- 500 1451 5010 3 107 .198003533 .429086000 500 1451 5011 3 107 .801996467 1.91091400 500 1451 ---------------------------------------- 500 1451 5000 3 107 1.00000000 2.34000000 500 1451 ---------------------------------------- 500 1451 ***************************************************************** 500 1451 ***************** Program GROUPIE (VERSION 2002-1) ************** 500 1451 Unshielded Group Averages Using 640 Groups 500 1451 Weighting Spectrum: Flat (Constant) Spectrum 500 1451 1 451 81 0 500 1451 3 1 217 0 500 1451 500 1 0 500 0 0 5.00000E+3 1.07191E+1 0 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1 12400000.0 1.42114992 12500000.0 1.42179096 12600000.0 1.42196161 500 3 1 12700000.0 1.42229522 12800000.0 1.42235987 12900000.0 1.42263263 500 3 1 13000000.0 1.42405226 13100000.0 1.42427635 13200000.0 1.42446247 500 3 1 13300000.0 1.42459770 13400000.0 1.42479286 13500000.0 1.42496373 500 3 1 13600000.0 1.42513722 13700000.0 1.42529220 13800000.0 1.42573183 500 3 1 13900000.0 1.42636889 14000000.0 1.42653165 14100000.0 1.42620005 500 3 1 14200000.0 1.42604859 14300000.0 1.42607695 14400000.0 1.42620803 500 3 1 14500000.0 1.42645059 14600000.0 1.42640948 14700000.0 1.42608295 500 3 1 14800000.0 1.42604421 14900000.0 1.42627763 15000000.0 1.42640597 500 3 1 15100000.0 1.42642142 15200000.0 1.42620980 15300000.0 1.42577383 500 3 1 15400000.0 1.42552047 15500000.0 1.42545867 15600000.0 1.42546625 500 3 1 15700000.0 1.42552674 15800000.0 1.42540050 15900000.0 1.42507008 500 3 1 16000000.0 1.42454069 16100000.0 1.42380209 16200000.0 1.42319399 500 3 1 16300000.0 1.42273229 16400000.0 1.42230819 16500000.0 1.42192518 500 3 1 16600000.0 1.42134787 16700000.0 1.42055453 16800000.0 1.41979771 500 3 1 16900000.0 1.41908886 17000000.0 1.41825331 17100000.0 1.41730518 500 3 1 17200000.0 1.41640109 17300000.0 1.41553330 17400000.0 1.41458477 500 3 1 17500000.0 1.41355380 17600000.0 1.41256255 17700000.0 1.41161949 500 3 1 17800000.0 1.41074469 17900000.0 1.40994157 18000000.0 1.40904997 500 3 1 18100000.0 1.40808103 18200000.0 1.40715562 18300000.0 1.40626368 500 3 1 18400000.0 1.40541552 18500000.0 1.40460736 18600000.0 1.40372433 500 3 1 18700000.0 1.40276326 18800000.0 1.40182117 18900000.0 1.40089943 500 3 1 19000000.0 1.39988249 19100000.0 1.39876550 19200000.0 1.39769893 500 3 1 19300000.0 1.39669575 19400000.0 1.39567395 19500000.0 1.39465125 500 3 1 19600000.0 1.39354180 19700000.0 1.39233869 19800000.0 1.39113602 500 3 1 19900000.0 1.38993562 20000000.0 0.0 500 3 1 500 3 0 500 0 0 0 0 0 5.01000E+3 9.92692E+0 0 0 34 10 525 1451 0.0 0.0 0 0 0 6 525 1451 1.00000E+0 2.00000E+7 0 0 10 2002 525 1451 3.00000E+2 0.0 1 0 256 3 525 1451 5-B - 10 LANL EVAL-NOV89 G.M.HALE, P.G.YOUNG 525 1451 DIST-Feb2004 525 1451 ----IRDF-2002 MATERIAL 525 525 1451 -----INCIDENT NEUTRON DATA 525 1451 ------ENDF-6 FORMAT 525 1451 ****************************************************************** 525 1451 5-B - 10 LANL EVAL-NOV89 G.M.HALE, P.G.YOUNG 525 1451 DIST-SEP91 REV1-JUL91 19910806 525 1451 ----ENDF/B-VI MATERIAL 525 REVISION 1 525 1451 ****************************************************************** 525 1451 5-B - 10 LANL EVAL-NOV89 G.M.HALE, P.G.YOUNG 525 1451 DIST-FEB91 910201 525 1451 ----IRDF-90 MATERIAL 525 525 1451 525 1451 *****************EXTRACT FOR SPECIAL PURPOSE FILE***************** 525 1451 DOSIMETRY 525 1451 ****************************************************************** 525 1451 525 1451 MOD1 OF ENDF/B-VI 525 1451 525 1451 The following revisions were made for MOD1 of ENDF/B-VI: 525 1451 525 1451 1. MF=1,MT=451 - Comments were added regarding estimated 525 1451 (expanded) covariance for the Standards Cross Sections. 525 1451 2. MF=3,MT=55,57,62,64,65,68,70,71,73,74,76-81,83,84 - LR flags 525 1451 and Q-values corrected. 525 1451 525 1451 ***************************************************************** 525 1451 ****************************************************************** 525 1451 525 1451 ENDF/VI EVALUATION 525 1451 G. M. Hale and P. G. Young 525 1451 525 1451 MAJOR CHANGES FROM VERSION V OF ENDF/B ARE: 525 1451 525 1451 1. Inclusion of the ENDF/B-VI standard (n,alpha) and (n,alpha1) 525 1451 results from the simultaneous standards analysis (Ca85) over 525 1451 the standard energy range thermal to 100 keV. 525 1451 2. Replacement of all major cross sections and elastic angular 525 1451 distributions from 10-5 eV to 1 MeV with results from the 525 1451 R-matrix analysis performed in conjunction with the 525 1451 simultaneous standards analysis. 525 1451 3. Replaced the total cross section 1-20 MeV with results 525 1451 from a covariance analysis of available data. 525 1451 4. Revised elastic and inelastic cross sections for low-lying 525 1451 levels incorporating new elastic, inelastic, and (n,xgamma) 525 1451 experimental data. We attempted to better reconcile the 525 1451 inelastic and gamma ray data. 525 1451 5. Refit all elastic angular distributions from 1-20 MeV with 525 1451 Legendre expansions and incorporated results from new 525 1451 measurements. 525 1451 6. Fit inelastic neutron angular distributions for first 5 525 1451 excited states of B10 with Legendre expansions. 525 1451 7. Incorporated new (n,t2alpha) cross section data into MT113 525 1451 and adjusted (n,alpha) cross sections above standard region 525 1451 for better consistency with data as well as other cross 525 1451 sections (esp. total and elastic) determined by data. 525 1451 525 1451 *****Note that covariance data will be added at a later date. 525 1451 525 1451 ****************************************************************** 525 1451 525 1451 STANDARDS COVARIANCES 525 1451 525 1451 Phase 1 reviewers of the ENDF/B-VI standards cross sections have 525 1451 expressed the concern that the uncertainties resulting from the 525 1451 combination of R-matrix and simultaneous evaluations might have 525 1451 led to uncertainties that are too small. As a result, the 525 1451 Standards Subcommittee produced (at the May, 1990 CSEWG meeting) 525 1451 a set of expanded covariance estimates for the standard cross 525 1451 section reactions. These uncertainties are estimates such that 525 1451 if a modern day experiment were performed on a given standard 525 1451 cross section using the best techniques, approximately 2/3 of 525 1451 the results should fall within these expanded uncertainties. The 525 1451 expanded uncertainties for the B-10(n,alpha0) and B-10(n,alpha1) 525 1451 cross sections are given in the following tables and are 525 1451 compared to values from the combined output of the standards 525 1451 covariance analysis: 525 1451 525 1451 525 1451 B-10(n,alpha0) Cross Section 525 1451 525 1451 Energy Range Estimated Uncertainty Combined Analysis 525 1451 (keV) (percent) (percent) 525 1451 525 1451 1.0E-08 - 0.1 0.5 0.21 525 1451 0.1 - 5.0 1.5 525 1451 5.0 - 30. 3.0 0.38 525 1451 30. - 90. 5.0 525 1451 90. - 150 10.0 0.86 525 1451 150 - 200 12.0 525 1451 200 - 250 15.0 0.79 525 1451 525 1451 525 1451 525 1451 B-10(n,alpha1) Cross Section 525 1451 525 1451 Energy Range Estimated Uncertainty Combined Analysis 525 1451 (keV) (percent) (percent) 525 1451 525 1451 1.0E-08 - 0.1 0.2 0.16 525 1451 0.1 - 5.0 0.4 525 1451 5.0 - 30. 0.6 0.20 525 1451 30. - 90. 1.0 525 1451 90. - 150 1.5 0.48 525 1451 150 - 200 2.0 525 1451 200 - 250 2.5 0.62 525 1451 525 1451 **************************************************************** 525 1451 525 1451 mf=2 --------- resonance parameters ---------------------------- 525 1451 525 1451 mt=151 effective scattering radius = 4.129038-13 cm 525 1451 525 1451 mf=3 --------- smooth cross sections --------------------------- 525 1451 525 1451 the 2200 m/s cross sections are as follows, 525 1451 mt=1 sigma = 3842.146 barns 525 1451 mt=2 sigma = 2.142435 barns 525 1451 mt=102 sigma = 0.5 barns 525 1451 mt=103 sigma = 0.000566 barns 525 1451 mt=107 sigma = 3839.496 barns 525 1451 mt=113 sigma = 0.0069993 barns 525 1451 mt=600 sigma = 0.000566 barns 525 1451 mt=800 sigma = 241.2677 barns 525 1451 mt=801 sigma = 3598.228 barns 525 1451 525 1451 mt=1 total cross section 525 1451 0 to 1 mev, calculated from r-matrix parameters obtained 525 1451 from simultaneous standards analysis (ca85) used to 525 1451 obtain the endf/b-vi standard cross sections. 525 1451 1 to 20 mev, covariance analysis of measurements of di67, 525 1451 ts62,fo61,co52,au79, and co54, constrained to match 525 1451 r-matrix fit at 1 mev. glucs covariance analysis code 525 1451 (he80) was used in the calculations. 525 1451 525 1451 525 1451 mt=107 (n,alpha) cross section 525 1451 0 to 20 mev, sum of mt=800,801. 525 1451 525 1451 mt=800 (n,alpha0) cross section 525 1451 0 to 1 mev, calculated from the r-matrix parameters 525 1451 described for mt=1. experimental (n,alpha0) data input 525 1451 to the fit were those of ma68 and da61. in addition, the 525 1451 angular distributions of va72 for the inverse reaction 525 1451 were included in the analysis. 525 1451 1 to 20 mev, based on da61 measurements, with smooth extra- 525 1451 polation from 8 to 20 mev using 14-MeV data of an69. The 525 1451 da61 data above approximately 2 mev were renormalized 525 1451 by a factor of approximately 1.4. Note that some of the 525 1451 structure seen in da61 was expanded to give consistent 525 1451 nonelastic, elastic, and total cross sections when 525 1451 compared with experimental data. 525 1451 525 1451 mt=801 (n,alpha1) cross section 525 1451 0 to 1 mev, calculated from the r-matrix parameters 525 1451 described for mt=1. experimental (n,alpha1) data in- 525 1451 cluded in the fit are those of sc76. in addition, the 525 1451 absolute differential cross-section measurements of 525 1451 se76 were included in the analysis. 525 1451 1 to 20 mev, smooth curve through measurements of da61 and 525 1451 ne70, with smooth extrapolation from 15 to 20 mev. the 525 1451 da61 data above approximately 2 mev were renormalized 525 1451 by a factor of approximately 1.4. Note that some of the 525 1451 structure seen in da61 was expanded to give consistent 525 1451 nonelastic, elastic, and total cross sections when 525 1451 compared with experimental data. 525 1451 525 1451 525 1451 525 1451 ----------------------- references ----------------------------- 525 1451 525 1451 aj75 f. ajzenberg-selove, nucl. phys. a248,6 (1975) 525 1451 aj88 f. ajzenberg-selove, nucl. phys. a490,1 (1988) 525 1451 an69 B. Antolkovic, Nuc.Phys.A139, 10 (1969). 525 1451 as70 a. asami and m.c. moxon, j.nucl.energy 24,85 (1970) 525 1451 au79 g.auchampaugh et al., nucl.sci.eng.69,30(1979) 525 1451 ba60 r.bardes and g.e. owen, phys.rev.120,1369 (1960) 525 1451 be56 r.l. becker and h.h. barschall, phys.rev.102,1384 (1956) 525 1451 bo51 c.k.bockelman et al., phys.rev. 84,69 (1951) 525 1451 bo69 d.bogart and l.l.nichols, nucl.phys.a125,463 (1969) 525 1451 ca85 a.carlson et al., nuc.data for basic & applied science, 525 1451 santa fe, nm (1985) p.1429. 525 1451 co52 j.h.coon et al., phys.rev. 88,562 (1952) 525 1451 co54 c.f.cook and t.w. bonner,phys.rev. 94,651 (1954) 525 1451 co67 s.a. cox and f.r. pontet, j.nucl.energy 21,271 (1967) 525 1451 co69 j.a. cookson and j.g.locke,nucl.phys.a146,417(1970) 525 1451 co73 m.s. coates et al., priv. comm. to l.stewart (1973) 525 1451 da56 r.b.day,phys.rev.102,767 (1956) 525 1451 da60 r.b. day and m.walt,phys.rev.117,1330 (1960) 525 1451 da61 e.a. davis et al., nucl.phys.27,448 (1961) 525 1451 di67 k.m. diment, aere-r-5224 (1967) 525 1451 di88 j.k.dickens, proc.conf. on nuc.data for sci.& tech.,mito, 525 1451 japan (1988) p.213. 525 1451 fo61 d.m. fossan et al., phys.rev. 123,209 (1961) 525 1451 fr56 g.m. frye and j.h. gammel,phys.rev. 103,328 (1956) 525 1451 gl82 s.glendinning, nuc.sci.eng.80,256(1982) 525 1451 ha73 s.l.hausladen, thesis, ohio univ. coo-1717-5 (1973) 525 1451 he80 d.hetrick & c.y.fu, ornl/tm-7341 (1980). 525 1451 hy69 m.hyakutake, eandc(j)-13 (1969) p.29 525 1451 ho69 j.c. hopkins, priv. comm. lasl (1969) 525 1451 ir67 d.c.irving, ornl-tm-1872 (1967) 525 1451 ka87 R.Kavanagh & R.Marcley, Phys.Rev.C36, 1194 (1987). 525 1451 la71 r.o. lane et al., phys.rev.c4,380 (1971) 525 1451 ma68 r.l.macklin and j.h.gibbons,phys.rev.165,1147 (1968) 525 1451 mo66 f.p.mooring et al.,nucl.phys.82,16 (1966) 525 1451 ne54 n.g.nereson,la-1655 (1954) 525 1451 ne70 d.o.nellis et al., phys.rev. c1,847 (1970) 525 1451 po70 d.porter et al., awre o 45/70 (1970) 525 1451 qa85 S.Qaim et al., Santa Fe Conf. (1985)p.97. 525 1451 qa88 S.Qaim et al., Mito Conf. (1988) p.225. 525 1451 sa88 E.T. Sadowski, Ph.D thesis, Ohio U., (Nov.,1988). 525 1451 sc76 r.a. schrack et al., proc.icinn(erda-conf-760715-p2),1345 525 1451 (1976) 525 1451 se76 r.m. sealock and j.c. overley, phys.rev.c13,2149 (1976) 525 1451 si65 r.h.siemssen et al., nucl.phys.69,209 (1965) 525 1451 sp73 r.r. spencer et al., eandc(e)147,al (1973) 525 1451 te62 k.tesch, nucl.phys.37,412 (1962) 525 1451 th67 g.e. thomas et al., nucl.instr.meth.56,325 (1967) 525 1451 ts63 k.tsukada and o.tanaka,j.phys.soc.japan 18,610 (1963) 525 1451 va65 v.valkovic et al., phys.rev. 139,331 (1965) 525 1451 va70 b.vaucher et al.,helv.phys.acta 43,237 (1970) 525 1451 va72 l.van der zwan and k.w.geiger, nucl.phys. a180,615 (1972) 525 1451 wi55 h.b. willard et al., phys.rev. 98,669(1955) 525 1451 wy58 m.e. wyman et al., phys.rev.112,1264 (1958) 525 1451 **************************************************************** 525 1451 525 1451 525 1451 525 1451 525 1451 ***************** Program LINEAR (VERSION 2002-1) *************** 525 1451 For All Data Greater than 1.0000E-10 barns in Absolute Value 525 1451 Data Linearized to Within an Accuracy of .100000000 per-cent 525 1451 ***************** Program SIGMA1 (VERSION 2002-1) *************** 525 1451 Data Doppler Broadened to 300.000000 Kelvin 525 1451 for All Data Greater than 1.0000E-10 barns in Absolute Value 525 1451 Data Linearized to Within an Accuracy pf .100000000 per-cent 525 1451 ***************** Program FIXUP (Version 2002-1) **************** 525 1451 Corrected ZA/AWR in All Sections-----------------------------Yes 525 1451 Corrected Thresholds-----------------------------------------Yes 525 1451 Extended Cross Sections to 20 MeV----------------------------No 525 1451 Allow Cross Section Deletion---------------------------------No 525 1451 Allow Cross Section Reconstruction---------------------------No 525 1451 Make All Cross Sections Non-Negative-------------------------Yes 525 1451 Delete Energies Not in Ascending Order-----------------------Yes 525 1451 Deleted Duplicate Points-------------------------------------Yes 525 1451 Check for Ascending MAT/MF/MT Order--------------------------Yes 525 1451 Check for Legal MF/MT Numbers--------------------------------Yes 525 1451 Allow Creation of Missing Sections---------------------------No 525 1451 Allow Insertion of Energy Points-----------------------------No 525 1451 Create Uniform Energy Grid-----------------------------------No 525 1451 Delete Section if Cross Section =0 at All Energies-----------Yes 525 1451 **************************************************************** 525 1451 ***************** Program GROUPIE (VERSION 2002-1) ************** 525 1451 Unshielded Group Averages Using 640 Groups 525 1451 Weighting Spectrum: Flat (Constant) Spectrum 525 1451 1 451 263 2 525 1451 3 107 217 1 525 1451 33 107 26 1 525 1451 525 1 0 525 0 0 5.01000E+3 9.92692E+0 0 0 0 0 525 3107 2.78950E+6 2.78950E+6 0 0 1 641 525 3107 641 1 525 3107 .000100000 60369.5672 .000105000 58961.8482 .000110000 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2.243903-6 2.000165-6 52533107 1.927335-6 1.633085-6 1.490809-6 1.647004-6 1.200942-6 1.128964-6 52533107 2.552806-6 2.624028-6 2.350393-6 2.167907-6 2.540208-6 1.939686-6 52533107 1.230224-6 2.113235-6 1.424748-6 1.662929-6 2.916571-6 2.927420-6 52533107 3.121191-6 3.670394-6 2.363825-6 1.299338-6 2.929283-6 1.881633-6 52533107 2.253798-6 4.641171-6 6.294455-6 5.607427-6 2.930722-6 2.917079-6 52533107 4.089449-6 2.707325-6 2.287332-6 1.083199-5 1.198476-5 7.899932-6 52533107 5.246804-6 7.085706-6 4.943829-6 5.243390-6 2.019061-5 1.881386-5 52533107 8.922377-6 9.271985-6 7.711829-6 9.574556-6 2.834902-5 2.180783-5 52533107 7.471345-6 8.265557-6 6.775191-6 3.334345-5 1.719968-5 8.391731-6 52533107 6.705476-6 4.365900-5 1.552481-5 1.322365-5 4.232958-5 2.020522-5 52533107 1.222438-4 52533107 52533 0 525 0 0 0 0 0 9.01900E+3 1.88352E+1 0 0 34 10 925 1451 0.0 0.0 0 0 0 6 925 1451 1.00000E+0 3.00001E+7 0 0 10 2002 925 1451 3.00000E+2 0.0 1 0 164 3 925 1451 9-F - 19 FEI EVAL-OCT01 K.I.Zolotarev 925 1451 DIST-Feb2004 925 1451 ----IRDF-2002 MATERIAL 925 925 1451 -----INCIDENT NEUTRON DATA 925 1451 ------ENDF-6 FORMAT 925 1451 ***************************************************************** 925 1451 9-F -19 FEI EVAL-OCT01 K.I.Zolotarev 925 1451 DIST-JAN02 20020125 925 1451 ----BROND-2 MATERIAL 925 925 1451 -----INCIDENT NEUTRON DATA 925 1451 ------ENDF-6 FORMAT 925 1451 ------Russian Reactor Dosimetry File RRDF-2002 925 1451 ***************************************************************** 925 1451 Authors of evaluation: K.I.Zolotarev and A.B.Pashchenko 925 1451 ***************************************************************** 925 1451 MF= 3 925 1451 MT= 16 -(n,2n) cross section data 925 1451 -------------------------------------- 925 1451 Excitation function for the F-19(n,2n)F-18 reaction in the 925 1451 energy region from threshold to 30 MeV was evaluated by means of 925 1451 statistical analysis of experimental cross section data [1-17] and 925 1451 data from GNASH [18,19] calculation. 925 1451 Experimental data [2-3],[5],[9-13],[15-17] were renormalized 925 1451 to the new standards for monitor reactions cross sections. Data 925 1451 of Mc Crary et al. [1], Rayburn et al.[2], Bormann et al.[4], and 925 1451 Shiokawa et al. [7] were renormalized to the results of precise 925 1451 measurements Vonach et al. [9] and Ikeda et al. [15] in the 14 MeV 925 1451 region. Data of Chatterjee et al. [8] were corrected to their pre- 925 1451 vious measurements at the 14.8 MeV point [6]. Results of measure- 925 1451 ments of Nagel [5] were taken only from SUBENT 005, EXFOR 20198. 925 1451 Experimental cross section data from ref. [20-31] were rejec- 925 1451 ted due to their big discrepancy with the main bulk of experimen- 925 1451 al data [1-17] and data from theoretical model calculation. 925 1451 The final procedure of evaluation F-19(n,2n)F-18 excitation 925 1451 function from threshold to 30 MeV has been carried out within the 925 1451 framework of generalized least squares method. Rational function 925 1451 was used as model function [32]. Calculations was performed by 925 1451 means of Pade-2 code [33]. 925 1451 U-235 thermal fission [34] and Cf-252 spontaneous fission 925 1451 neutron spectra [35] averaged cross-sections calculated from the 925 1451 evaluated F-19(n,2n)F-18 excitation function are the following: 925 1451 925 1451 ----------------------------------------------------------------- 925 1451 TYPE OF SPECTRUM I ,mb (calc.) I , mb (measured) 925 1451 ----------------------I-----------------I------------------------ 925 1451 U-235 neutron fission I 0.007299 I 0.007200+-0.001000 [36] 925 1451 I I 0.006509+-0.000300 [37] 925 1451 I I 0.008653+-0.000464 [38] 925 1451 ----------------------I-----------------I------------------------ 925 1451 CF-252 spont. fission I 0.01615 I 0.01628+-0.00054 [39] 925 1451 I I 0.01612+-0.00054 [40] 925 1451 ----------------------------------------------------------------- 925 1451 925 1451 MF=33 925 1451 MT= 16 -(n,2n) cross section cov. matrix 925 1451 --------------------------------------------- 925 1451 Uncertainties in the evaluated excitation function for the 925 1451 reaction F-19(n,2n)F-18 are given in the form of relative covari- 925 1451 ance matrix for the 26-neutron energy groups (LB=5). Covariance 925 1451 matrix of uncertainties was calculated simultaneously with 925 1451 recommended cross section data by means of PADE-2 code. 925 1451 Eigenvalues of the 6-th digits relative covariance matrix 925 1451 given in the 33-file are the following: 925 1451 925 1451 3.67622E-08 3.74071E-08 3.79051E-08 3.91111E-08 925 1451 4.11094E-08 4.32321E-08 4.68199E-08 4.98738E-08 925 1451 5.55426E-08 6.01322E-08 7.00100E-08 7.85485E-08 925 1451 9.57150E-08 1.06820E-07 1.28694E-07 1.47868E-07 925 1451 2.58858E-07 6.32301E-07 3.80384E-06 5.45023E-04 925 1451 1.36448E-03 3.28506E-03 8.26671E-03 1.39754E-02 925 1451 1.58981E-02 6.37025E-02 925 1451 925 1451 References : 925 1451 1. J.H.Mc Crary,I.L.Morgan Bull. American Phys. Soc., v.5, 925 1451 p.246, April 1960 ; 925 1451 J.H.Mc Crary,I.L.Morgan Report AFSWC-TR-60-30, 1960 925 1451 2. L.A.Rayburn Proc. of Conf. on Direct Interactions and Nuclear 925 1451 Reaction Mechanisms, Padua, 3-8 September 1962, Gordon and 925 1451 Breach, New York, 1963, p.322. ; 925 1451 L.A.Rayburn Bull. Am. Phys. Soc., v.7, p.335, April 1962 925 1451 3. M.Cevolani, S.Petralia Nuovo Cimento, v.26, p.1328, Dec.1962 925 1451 4. M.Bormann et al. Nucl. Phys., v.63, p.438, March 1965 925 1451 5. W.Nagel EXFOR 20198.005, 1966 925 1451 6. A.Chatterjee et al. Progress Report BARC-305, p.30, Nov. 1967 925 1451 7. T.Shiokawa et al. J. Inorg. Nucl. Chem., v.30, p.1, Jan. 1968 925 1451 8. A.Chatterjee e.a. Proc.of Symposium Nucl.Phys. and Solid State 925 1451 Phys., Roorke, India, 1969, v.2, p.117, December 1969 925 1451 9. H.K.Vonach et al. Proc. of 2-nd Conference on Nuclear Cross- 925 1451 Sections and Technology, Washington D.C., 4-7 March 1968, v.2, 925 1451 p.885 925 1451 10. R.C.Barrall et al. Report AFWL-TR-68-134, March 1969 925 1451 11. R.Mogharrab, H.Neuert Atomkernenergie, v.19, p.107, April 1972 925 1451 12. J.C.Robertson et al. J. Nucl. Energ., v.27, p.531, Aug. 1973 925 1451 13. R.A.Sigg Dissert. Abstr., sect. B, v.37, p.2237, Nov. 1976 925 1451 14. T.B.Ryves et al. J. of Physics, pt.G, v.4, n.11, p.1783, 1978 925 1451 15. Y.Ikeda et al. Report JAERI-1312, 1988 925 1451 16. I.Kimura, K.Kobayashi Nucl. Sci. Eng., v.106, p.332, 1990 925 1451 17. C.L.Hartmann, P.M.DeLuca Nucl. Sci. Eng., v.109, p.319, 1991 925 1451 18. P.G.Young, E.D.Arthur A Preequilibrium Statistical Nuclear 925 1451 Model Code for Calculation of Cross Section and Emission 925 1451 Spectra. Report LA-6947, Los Alamos, 1977 925 1451 19. E.L.Trykov, G.Ya.Tertychnyi Private communication, IPPE, 925 1451 Obninsk, May 1999 925 1451 20. E.B.Paul, R.L.Clarke Canadian J. Phys., v.31, p.267, 1953 925 1451 21. V.J.Ashby e.a. Phys. Rev., v.111, p.616, 1958 925 1451 22. L.A.Rayburn Phys. Rev., v.122, p.168, 1961 925 1451 23. O.D.Brill et al. Doklady Akademii Nauk, v.136,n.1,p.55, 1961; 925 1451 O.D.Brill et al. Soviet Physics-Doklady, v.6, p.24, 1961 925 1451 24. J.Picard, C.F.Williamson Nucl. Phys., v.63, p.673, April 1965 925 1451 25. J.E.Strain, W.J.Ross Report ORNL-3672, January 1965 925 1451 26. J.Csikai Report EANDC-50 S, paper 102, July 1965 925 1451 27. A.Pasquarelli Nucl. Phys. A, v.93, p.218, March 1967 925 1451 28. M.Bormann, I.Riehle Zeitschrift f.Physik, v.207, p.64, 925 1451 October 1967 925 1451 29. H.O.Menlove et al. Phys. Rev., v.163, p.1308, 1967 925 1451 30. D.Crumpton J.Inorg. Nucl. Chem., v.31, p.3727, December 1969 925 1451 31. J.Araminowicz, J.Dresler Report INR-1464, p.14, May 1973 925 1451 32. S.Badikov, N.Rabotnov, K.Zolotarev Proc. of NEANSC Speciali- 925 1451 st's Meeting on Evaluation and Processing of Covariance Data, 925 1451 Oak Ridge, USA, 7-9 September 1992, OECD, Paris, 1993, p.105 925 1451 33. S.A.Badikov et.al. Preprint FEI-1686, Obninsk, 1985 925 1451 34. L.W.Weston et al. Evaluated Neutron Data for U-235, ENDF/B-VI 925 1451 Library, MAT=9228, MF=5, MT=18, eval. April 1989 925 1451 35. W.Mannhart IAEA-TECDOC-410, p.158, IAEA, Vienna, 1987 925 1451 36. F.Nasyrov, B.D.Sciborskij Atomnaja Energija, v.25, no.5, 925 1451 p.437, November 1968 925 1451 37. M.Najzer, J.Rant Report IAEA-208, v.2, p.247, 1978 925 1451 38. W.Mannhart Progress Report INDC(Ger)-045, pp.40-43, 1999 925 1451 39. W.Mannhart Handbook on Nuclear Activation Data. IAEA Tech- 925 1451 nical Report series No.273, p.413, 1987 925 1451 40. W.Mannhart Validation of Differential Cross Sections with 925 1451 Integral Data , Report INDC(NDS)-435, pp.59-64, IAEA, Vienna, 925 1451 September 2002 925 1451 ***************************************************************** 925 1451 File 2 added to the pointwise file containing only the effective 925 1451 scattering radius with no resonance parameters given. 925 1451 Taken from ENDF/B-VI 925 1451 925 1451 ***************************************************************** 925 1451 ***************** Program LINEAR (VERSION 2002-1) *************** 925 1451 For All Data Greater than 1.0000E-10 barns in Absolute Value 925 1451 Data Linearized to Within an Accuracy of .100000000 per-cent 925 1451 ***************** Program SIGMA1 (VERSION 2002-1) *************** 925 1451 Data Doppler Broadened to 300.000000 Kelvin 925 1451 for All Data Greater than 1.0000E-10 barns in Absolute Value 925 1451 Data Linearized to Within an Accuracy pf .100000000 per-cent 925 1451 ***************** Program FIXUP (Version 2002-1) **************** 925 1451 Corrected ZA/AWR in All Sections-----------------------------Yes 925 1451 Corrected Thresholds-----------------------------------------Yes 925 1451 Extended Cross Sections to 20 MeV----------------------------No 925 1451 Allow Cross Section Deletion---------------------------------No 925 1451 Allow Cross Section Reconstruction---------------------------No 925 1451 Make All Cross Sections Non-Negative-------------------------Yes 925 1451 Delete Energies Not in Ascending Order-----------------------Yes 925 1451 Deleted Duplicate Points-------------------------------------Yes 925 1451 Check for Ascending MAT/MF/MT Order--------------------------Yes 925 1451 Check for Legal MF/MT Numbers--------------------------------Yes 925 1451 Allow Creation of Missing Sections---------------------------No 925 1451 Allow Insertion of Energy Points-----------------------------No 925 1451 Create Uniform Energy Grid-----------------------------------No 925 1451 Delete Section if Cross Section =0 at All Energies-----------Yes 925 1451 ***************** Program GROUPIE (VERSION 2002-1) ************** 925 1451 Unshielded Group Averages Using 640 Groups 925 1451 Weighting Spectrum: Flat (Constant) Spectrum 925 1451 1 451 171 1 925 1451 3 16 34 1 925 1451 33 16 71 1 925 1451 925 1 0 925 0 0 9.01900E+3 1.88352E+1 0 0 0 0 925 3 16 -1.04322E+7-1.04322E+7 0 0 1 92 925 3 16 92 1 925 3 16 10900000.0 2.02752E-6 11000000.0 5.63229E-5 11100000.0 .000137655 925 3 16 11200000.0 .000292275 11300000.0 .000548721 11400000.0 .000917543 925 3 16 11500000.0 .001402145 11600000.0 .002002095 11700000.0 .002714490 925 3 16 11800000.0 .003534705 11900000.0 .004456885 12000000.0 .005474310 925 3 16 12100000.0 .006579685 12200000.0 .007765400 12300000.0 .009023740 925 3 16 12400000.0 .010347050 12500000.0 .011727850 12600000.0 .013159000 925 3 16 12700000.0 .014633700 12800000.0 .016152650 12900000.0 .017695750 925 3 16 13000000.0 .019265950 13100000.0 .020863250 13200000.0 .022460550 925 3 16 13300000.0 .024071763 13400000.0 .025696888 13500000.0 .027322013 925 3 16 13600000.0 .028947138 13700000.0 .030572263 13800000.0 .032197388 925 3 16 13900000.0 .033822513 14000000.0 .035447638 14100000.0 .037049625 925 3 16 14200000.0 .038628475 14300000.0 .040207325 14400000.0 .041786175 925 3 16 14500000.0 .043331888 14600000.0 .044844463 14700000.0 .046357038 925 3 16 14800000.0 .047869613 14900000.0 .049339675 15000000.0 .050767225 925 3 16 15100000.0 .052194775 15200000.0 .053622325 15300000.0 .054999450 925 3 16 15400000.0 .056326150 15500000.0 .057652850 15600000.0 .058979550 925 3 16 15700000.0 .060248850 15800000.0 .061460750 15900000.0 .062672650 925 3 16 16000000.0 .063884550 16100000.0 .065033000 16200000.0 .066118000 925 3 16 16300000.0 .067203000 16400000.0 .068288000 16500000.0 .069304550 925 3 16 16600000.0 .070252650 16700000.0 .071200750 16800000.0 .072148850 925 3 16 16900000.0 .073024838 17000000.0 .073828713 17100000.0 .074632588 925 3 16 17200000.0 .075436463 17300000.0 .076166188 17400000.0 .076821763 925 3 16 17500000.0 .077477338 17600000.0 .078132913 17700000.0 .078714000 925 3 16 17800000.0 .079220600 17900000.0 .079727200 18000000.0 .080233800 925 3 16 18100000.0 .080667413 18200000.0 .081028038 18300000.0 .081388663 925 3 16 18400000.0 .081749288 18500000.0 .082040025 18600000.0 .082260875 925 3 16 18700000.0 .082481725 18800000.0 .082702575 18900000.0 .082858125 925 3 16 19000000.0 .082948375 19100000.0 .083038625 19200000.0 .083128875 925 3 16 19300000.0 .083180900 19400000.0 .083152770 19500000.0 .083082710 925 3 16 19600000.0 .083012650 19700000.0 .082942590 19800000.0 .082872530 925 3 16 19900000.0 .082740370 20000000.0 0.0 925 3 16 925 3 0 925 0 0 9.01900E+3 1.88352E+1 0 0 0 1 92533 16 0.000000+0 0.000000+0 0 16 0 1 92533 16 0.000000+0 0.000000+0 1 5 406 28 92533 16 1.000000-5 1.090000+7 1.200000+7 1.250000+7 1.300000+7 1.350000+7 92533 16 1.400000+7 1.450000+7 1.500000+7 1.550000+7 1.600000+7 1.650000+7 92533 16 1.700000+7 1.750000+7 1.800000+7 1.850000+7 1.900000+7 1.950000+7 92533 16 2.000000+7 2.100000+7 2.200000+7 2.300000+7 2.400000+7 2.500000+7 92533 16 2.700000+7 2.800000+7 2.900000+7 3.000000+7 0.000000+0 0.000000+0 92533 16 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0 92533 16 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0 92533 16 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0 92533 16 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0 92533 16 0.000000+0 7.844560-3 4.898350-3 3.027060-3 1.929170-3 1.321910-3 92533 16 9.926111-4 8.097050-4 6.990610-4 6.218640-4 5.596050-4 5.048120-4 92533 16 4.555230-4 4.121780-4 3.758470-4 3.473540-4 3.269110-4 3.140740-4 92533 16 3.073570-4 3.120960-4 3.231580-4 3.310920-4 3.298620-4 3.041000-4 92533 16 2.525070-4 2.039030-4 1.464150-4 3.648070-3 2.360540-3 1.449110-3 92533 16 9.104930-4 6.319380-4 5.059420-4 4.569550-4 4.388680-4 4.266530-4 92533 16 4.088340-4 3.819840-4 3.470900-4 3.071470-4 2.656870-4 2.259100-4 92533 16 1.902720-4 1.486420-4 1.141580-4 1.016320-4 1.039070-4 1.134800-4 92533 16 1.281520-4 1.350840-4 1.314670-4 1.212070-4 1.826270-3 1.354210-3 92533 16 9.782671-4 7.059900-4 5.267220-4 4.207220-4 3.665780-4 3.449020-4 92533 16 3.398810-4 3.397380-4 3.365240-4 3.256910-4 3.054490-4 2.760780-4 92533 16 2.392420-4 1.754770-4 8.900070-5 1.611170-5-3.205140-5-5.099350-5 92533 16 -2.365820-5 5.330250-5 1.291950-4 2.188860-4 1.233180-3 1.050650-3 92533 16 8.482810-4 6.625510-4 5.129570-4 4.047920-4 3.347390-4 2.952800-4 92533 16 2.776250-4 2.733860-4 2.755370-4 2.787900-4 2.796490-4 2.762050-4 92533 16 2.614270-4 2.288330-4 1.889520-4 1.513570-4 1.231830-4 1.085890-4 92533 16 1.236560-4 1.520640-4 1.922260-4 1.030240-3 9.309550-4 7.910340-4 92533 16 6.422140-4 5.056870-4 3.931520-4 3.093600-4 2.543630-4 2.252570-4 92533 16 2.174790-4 2.257800-4 2.449250-4 2.701550-4 3.105180-4 3.558030-4 92533 16 3.811660-4 3.836510-4 3.655150-4 3.091160-4 2.316670-4 1.739440-4 92533 16 1.144030-4 9.301140-4 8.611210-4 7.506000-4 6.229440-4 4.974770-4 92533 16 3.876710-4 3.015590-4 2.426370-4 2.107830-4 2.032840-4 2.157690-4 92533 16 2.430530-4 3.003160-4 3.817500-4 4.465680-4 4.803090-4 4.784630-4 92533 16 4.115300-4 2.889300-4 1.821310-4 6.207860-5 8.648120-4 8.135850-4 92533 16 7.261930-4 6.210140-4 5.137590-4 4.162000-4 3.360620-4 2.773020-4 92533 16 2.407400-4 2.248090-4 2.263440-4 2.532360-4 3.094430-4 3.658040-4 92533 16 4.035410-4 4.133410-4 3.692840-4 2.697090-4 1.753330-4 6.488820-5 92533 16 8.276580-4 7.994630-4 7.403910-4 6.622230-4 5.756870-4 4.895590-4 92533 16 4.102830-4 3.420170-4 2.868660-4 2.452630-4 2.075860-4 1.890180-4 92533 16 1.937990-4 2.070150-4 2.180840-4 2.167570-4 1.920930-4 1.606910-4 92533 16 1.194350-4 8.367870-4 8.381600-4 8.077220-4 7.513760-4 6.760250-4 92533 16 5.886680-4 4.958200-4 4.031020-4 3.150170-4 2.001340-4 8.278800-5 92533 16 1.074360-5-2.078540-5-2.009100-5 2.343720-5 9.478700-5 1.508320-4 92533 16 2.094530-4 9.030970-4 9.293760-4 9.163470-4 8.669560-4 7.869960-4 92533 16 6.839560-4 5.660230-4 4.412210-4 2.582200-4 4.586960-5-1.068070-4 92533 16 -1.912730-4-2.107930-4-1.358110-4 2.138250-5 1.612100-4 3.183170-4 92533 16 1.012600-3 1.049310-3 1.037800-3 9.812610-4 8.866920-4 7.633760-4 92533 16 6.215440-4 3.970180-4 1.170730-4-1.006010-4-2.340860-4-2.800390-4 92533 16 -1.974460-4 1.093280-5 2.077500-4 4.350440-4 1.136340-3 1.170610-3 92533 16 1.152160-3 1.086070-3 9.810220-4 8.476920-4 6.195350-4 3.158060-4 92533 16 6.329550-5-1.052330-4-1.794050-4-1.175480-4 9.273780-5 3.048660-4 92533 16 5.562250-4 1.254720-3 1.286290-3 1.267530-3 1.204830-3 1.107390-3 92533 16 9.184080-4 6.431870-4 3.946850-4 2.121830-4 1.124980-4 1.261560-4 92533 16 2.843660-4 4.638940-4 6.851270-4 1.376450-3 1.420770-3 1.422120-3 92533 16 1.386320-3 1.278440-3 1.085040-3 8.822720-4 7.106800-4 5.921990-4 92533 16 5.353610-4 5.914810-4 6.921020-4 8.298950-4 1.539830-3 1.622460-3 92533 16 1.669850-3 1.679580-3 1.616300-3 1.498950-3 1.363950-3 1.236560-3 92533 16 1.094900-3 1.009270-3 9.922760-4 1.001160-3 1.798240-3 1.944450-3 92533 16 2.099870-3 2.205700-3 2.207720-3 2.133310-3 2.009010-3 1.777510-3 92533 16 1.524740-3 1.362900-3 1.209860-3 2.198710-3 2.517890-3 2.820120-3 92533 16 2.967440-3 2.974120-3 2.866070-3 2.549010-3 2.119470-3 1.798830-3 92533 16 1.465550-3 3.093180-3 3.712880-3 4.106620-3 4.263970-3 4.206430-3 92533 16 3.790030-3 3.114020-3 2.560510-3 1.957740-3 4.741660-3 5.470710-3 92533 16 5.854090-3 5.903210-3 5.430360-3 4.514070-3 3.710880-3 2.809610-3 92533 16 6.497010-3 7.107000-3 7.299680-3 6.881850-3 5.882950-3 4.952680-3 92533 16 3.884000-3 7.920590-3 8.280630-3 8.031330-3 7.128890-3 6.218540-3 92533 16 5.143690-3 8.820440-3 8.840330-3 8.205440-3 7.459920-3 6.540140-3 92533 16 9.408330-3 9.443150-3 9.180490-3 8.759390-3 1.037760-2 1.080020-2 92533 16 1.110710-2 1.176100-2 1.264710-2 1.416170-2 92533 16 92533 0 925 0 0 0 0 0 1.10230E+4 2.27923E+1 0 0 34 101125 1451 0.0 0.0 0 0 0 61125 1451 1.00000E+0 2.00000E+7 0 0 10 20021125 1451 3.00000E+2 0.0 1 0 214 51125 1451 11-Na- 23 ORNL,TOH EVAL-DEC77 D. C. LARSON,S.IWASAKI 1125 1451 DIST-Feb2004 1125 1451 ----IRDF-2002 MATERIAL 1125 1125 1451 -----INCIDENT NEUTRON DATA 1125 1451 ------ENDF-6 FORMAT 1125 1451 *****************EXTRACT FOR SPECIAL PURPOSE FILE*****************1125 1451 DOSIMETRY 1125 1451 ******************************************************************1125 1451 11-NA- 23 ORNL EVAL-DEC77 D. C. LARSON 1125 1451 DIST-SEP 1 REV1-JUL91 20010926 1125 1451 ----ENDF/B-VI MATERIAL 1125 REVISION 1 1125 1451 1125 1451 **************************************************************** 1125 1451 11-NA- 23 TOH EVAL-AUG96 S.IWASAKI 1125 1451 DIST-JUL98 1125 1451 ----JENDL/D-99 MATERIAL 1125 1125 1451 -----INCIDENT NEUTRON DATA 1125 1451 ------ENDF-6 FORMAT 1125 1451 1125 1451 NA-23 (N,2N) NA-22 (HALF-LIFE = 2.6019Y) 1125 1451 **************************************************************** 1125 1451 ================================================================ 1125 1451 ************** Start of ENDF/B-VI Bibliography ************** 1125 1451 ================================================================ 1125 1451 ENDF/B-VI MOD 2 Revision, September 2000, S.C.Frankle, R.C.Reedy,1125 1451 P.G.Young (LANL) 1125 1451 1125 1451 The secondary gamma-ray spectrum for radiative capture (MF 12, 1125 1451 MT 102) has been updated using new experimental data. 1125 1451 The previous evaluation (MOD 1: 61 gamma rays) is replaced with 1125 1451 292 discrete gamma rays based on thermal neutron experimental 1125 1451 data. Thermal spectrum assumed to 20 MeV. 1125 1451 The Q-value for radiative capture was also updated in File 3. 1125 1451 1125 1451 Gas production added as MF=3, MT=203,207 1125 1451 1125 1451 Details of these changes are described in Frankle et al. [Fr01]. 1125 1451 1125 1451 **************************************************************** 1125 1451 1125 1451 ENDF/B-VI MOD 1 Revision, July 1991, V. McLane (NNDC) 1125 1451 1125 1451 REVISION 1 CORRECTION 1125 1451 1125 1451 Corrected number of energy ranges in File 32, MT = 151 to be 1. 1125 1451 1125 1451 **************************************************************** 1125 1451 1125 1451 ENDF/B-VI MOD 0, January 1990 (NNDC) 1125 1451 1125 1451 ENDF/B-V Material 1311 converted to ENDF-6 format by NNDC. 1125 1451 1125 1451 **************************************************************** 1125 1451 1125 1451 ENDF/B-V Evaluation, December 1977, D. C. Larson (ORNL) 1125 1451 1125 1451 MF=2 1125 1451 Resonance parameters are used from 600 eV to 500 keV. The 1125 1451 thermal capture cross section is given as 528 mb, in agreement 1125 1451 with the experimental results of 526.9+-4.5 mb [RY71], as well 1125 1451 as earlier results summarized in [RY71]. Using resonance 1125 1451 parameters E=2.81 keV, GN=376 eV, and GG=0.353 eV for the large 1125 1451 2.81 keV resonance gives the correct thermal capture value using 1125 1451 a Breit-Wigner shape, and this form is used to calculate the 1125 1451 capture crposs section from 1.0E-5 to 600 eV. The capture 1125 1451 width GG=0.353 eV is consistent with a recent result by Wilson 1125 1451 [WI77] where they find 0.24 1125 1451 NA-23 (N,2N) NA-22 (HALF-LIFE = 2.6019Y) 1125 1451 1125 1451 MF=1 GENERAL INFORMATION 1125 1451 MT=451 DESCRIPTIVE DATA AND DICTIONARY 1125 1451 1125 1451 MF=3 NEUTRON CROSS SECTIONS 1125 1451 MT=16 (N,2N) CROSS SECTION 1125 1451 EVALUATED USING SPLINE FITTING METHOD /6/TO A SET OF 1125 1451 SELECTED EXPERIMENTAL DATA /9-14/ TAKEN FROM NESTOR-2/8/. 1125 1451 MAINLY BASED ON THE EXPERIMENTAL DATA OF ADAMSKI/6/. 1125 1451 1125 1451 MF=33 COVARIANCES OF NEUTRON CROSS SECTIONS 1125 1451 MT=16 1125 1451 GENERATED USING SPLINE FITTING METHOD /6/ 1125 1451 1125 1451 REFERENCES 1125 1451 6) S. IWASAKI, NUCLEAR DATA FOR SCINECE AND TECHNOLOGY, 1125 1451 GATLINBURG,TENNESSEE, MAY 9 13, 1994. PROCEEDINGS, VOL 2, 614,1125 1451 ANS, INC., (1994). 1125 1451 8) T. NAKAGAWA : THE JAERI NUCLEAR DATA CENTER, UNPUBLISHED. 1125 1451 9) H.O. MENLOVE, ET AL. PHYS. REV. 163,1308 (1967). 1125 1451 10) L. ADAMSKI, ET AL. ANN. NUCL. ENERGY, 7, (7), 397, (1980). 1125 1451 11) Y. IKEDA, ET AL., JAERI-1312 (1988). 1125 1451 12) LU HAN-LIN , ET AL. INDC(CRP)-16 (1989). 1125 1451 13) B.STROHMAIER, ET AL. ANN. NUCL. ENERG., 20,8 , 533-545 (1991).1125 1451 14) M. SAKUMA, S. IWASAKI, H. SHIMADA, N. ODANO, K. SUDA, 1125 1451 J.R. DUMAIS,AND K. SUGIYAMA, JAERI-M 92-027, P. 278 (1992). 1125 1451 ================================================================ 1125 1451 ******* Processing history of JENDL/D99 (n,2n) component ******* 1125 1451 ================================================================ 1125 1451 ***************** Program LINEAR (VERSION 2002-1) ***************1125 1451 For All Data Greater than 1.0000E-10 barns in Absolute Value 1125 1451 Data Linearized to Within an Accuracy of .100000000 per-cent 1125 1451 ***************** Program SIGMA1 (VERSION 2002-1) ***************1125 1451 Data Doppler Broadened to 300.000000 Kelvin 1125 1451 for All Data Greater than 1.0000E-10 barns in Absolute Value 1125 1451 Data Linearized to Within an Accuracy pf .100000000 per-cent 1125 1451 ***************** Program FIXUP (Version 2002-1) ****************1125 1451 Corrected ZA/AWR in All Sections-----------------------------Yes 1125 1451 Corrected Thresholds-----------------------------------------Yes 1125 1451 Extended Cross Sections to 20 MeV----------------------------No 1125 1451 Allow Cross Section Deletion---------------------------------No 1125 1451 Allow Cross Section Reconstruction---------------------------No 1125 1451 Make All Cross Sections Non-Negative-------------------------Yes 1125 1451 Delete Energies Not in Ascending Order-----------------------Yes 1125 1451 Deleted Duplicate Points-------------------------------------Yes 1125 1451 Check for Ascending MAT/MF/MT Order--------------------------Yes 1125 1451 Check for Legal MF/MT Numbers--------------------------------Yes 1125 1451 Allow Creation of Missing Sections---------------------------No 1125 1451 Allow Insertion of Energy Points-----------------------------No 1125 1451 Create Uniform Energy Grid-----------------------------------No 1125 1451 Delete Section if Cross Section =0 at All Energies-----------Yes 1125 1451 ================================================================ 1125 1451 *********** End of JENDL/D-99 (n,2n) Bibliography *********** 1125 1451 ================================================================ 1125 1451 1125 1451 1125 1451 1125 1451 1125 1451 ***************************************************************** 1125 1451 ************************ C O N T E N T S *********************** 1125 1451 ***************** Program GROUPIE (VERSION 2002-1) **************1125 1451 Unshielded Group Averages Using 640 Groups 1125 1451 Weighting Spectrum: Flat (Constant) Spectrum 1125 1451 1 451 223 21125 1451 3 16 27 11125 1451 3 102 217 21125 1451 33 16 32 11125 1451 33 102 5 01125 1451 1125 1 0 1125 0 0 1.10230E+4 2.27923E+1 0 0 0 01125 3 16 -1.24184E+7-1.24184E+7 0 0 1 721125 3 16 72 1 1125 3 16 12900000.0 .000146800 13000000.0 .001190000 13100000.0 .0019700001125 3 16 13200000.0 .002750000 13300000.0 .003530000 13400000.0 .0043100001125 3 16 13500000.0 .005940000 13600000.0 .008420000 13700000.0 .0109000001125 3 16 13800000.0 .013380000 13900000.0 .015860000 14000000.0 .0187000001125 3 16 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1112533102 0.000000+0 0.000000+0 0 102 0 1112533102 0.000000+0 0.000000+0 0 1 12 6112533102 1.000000-5 4.000000-4 5.000000+1 2.500000-3 6.000000+2 1.000000-2112533102 5.000000+5 4.000000-2 5.000000+6 6.250000-2 2.000000+7 0.000000+0112533102 112533 0 1125 0 0 0 0 0 1.20240E+4 2.37790E+1 0 0 34 101225 1451 0.0 0.0 0 0 0 61225 1451 1.00000E+0 2.10000E+7 0 0 10 20021225 1451 3.00000E+2 0.0 1 0 40 31225 1451 12-Mg- 24 IRK-VIENNA EVAL-APR90 1225 1451 DIST-Feb2004 1225 1451 ----IRDF-2002 MATERIAL 1225 1225 1451 -----INCIDENT NEUTRON DATA 1225 1451 ------ENDF-6 FORMAT 1225 1451 *****************************************************************1225 1451 12-MG- 24 IRK-VIENNA EVAL-APR90 1225 1451 DIST-JUN90 1225 1451 IRK-EVAL.NLIB 25 1225 1225 1451 *****************************************************************1225 1451 The Q values and threshold energies were updated prior to pro- 1225 1451 cessing through the codes to comply with the values obtained 1225 1451 using the NNDC calculation program which is based on the 1995 1225 1451 Update to the Atomic mass Evaluation. 1225 1451 *****************************************************************11225 1451 ***************** Program LINEAR (VERSION 2002-1) ***************1225 1451 For All Data Greater than 1.0000E-10 barns in Absolute Value 1225 1451 Data Linearized to Within an Accuracy of .100000000 per-cent 1225 1451 ***************** Program SIGMA1 (VERSION 2002-1) ***************1225 1451 Data Doppler Broadened to 300.000000 Kelvin 1225 1451 for All Data Greater than 1.0000E-10 barns in Absolute Value 1225 1451 Data Linearized to Within an Accuracy pf .100000000 per-cent 1225 1451 ***************** Program FIXUP (Version 2002-1) ****************1225 1451 Corrected ZA/AWR in All Sections-----------------------------Yes 1225 1451 Corrected Thresholds-----------------------------------------Yes 1225 1451 Extended Cross Sections to 20 MeV----------------------------No 1225 1451 Allow Cross Section Deletion---------------------------------No 1225 1451 Allow Cross Section Reconstruction---------------------------No 1225 1451 Make All Cross Sections Non-Negative-------------------------Yes 1225 1451 Delete Energies Not in Ascending Order-----------------------Yes 1225 1451 Deleted Duplicate Points-------------------------------------Yes 1225 1451 Check for Ascending MAT/MF/MT Order--------------------------Yes 1225 1451 Check for Legal MF/MT Numbers--------------------------------Yes 1225 1451 Allow Creation of Missing Sections---------------------------No 1225 1451 Allow Insertion of Energy Points-----------------------------No 1225 1451 Create Uniform Energy Grid-----------------------------------No 1225 1451 Delete Section if Cross Section =0 at All Energies-----------Yes 1225 1451 ***************** Program GROUPIE (VERSION 2002-1) **************1225 1451 Unshielded Group Averages Using 640 Groups 1225 1451 Weighting Spectrum: Flat (Constant) Spectrum 1225 1451 1 451 47 11225 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4.179000-4122533103 7.842000-4 5.345000-4 6.224000-4 3.405000-4 1.346000-4 1.630000-4122533103 2.301000-4 2.022000-4 2.989000-4 2.362000-4 3.532000-4 3.066000-4122533103 0.000000+0 0.000000+0 7.673000-5 0.000000+0 0.000000+0 0.000000+0122533103 0.000000+0 1.457000-5 0.000000+0 0.000000+0 0.000000+0 0.000000+0122533103 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0122533103 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0 3.049000-3122533103 8.556000-4 1.492000-3 1.028000-3 1.239000-3 7.962000-4 5.181000-4122533103 5.795000-4 7.489000-4 6.440000-4 3.669000-4 2.900000-4 4.336000-4122533103 3.764000-4 0.000000+0 0.000000+0 9.419000-5 0.000000+0 0.000000+0122533103 0.000000+0 0.000000+0 1.789000-5 0.000000+0 0.000000+0 0.000000+0122533103 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0122533103 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0122533103 7.661000-4 7.167000-4 4.947000-4 6.669000-4 3.912000-4 2.889000-4122533103 2.926000-4 3.759000-4 3.227000-4 1.624000-4 1.284000-4 1.919000-4122533103 1.666000-4 0.000000+0 0.000000+0 4.169000-5 0.000000+0 0.000000+0122533103 0.000000+0 0.000000+0 7.918000-6 0.000000+0 0.000000+0 0.000000+0122533103 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0122533103 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0122533103 2.140000-3 8.614000-4 1.039000-3 6.689000-4 4.346000-4 4.855000-4122533103 6.265000-4 5.385000-4 2.983000-4 2.358000-4 3.525000-4 3.060000-4122533103 0.000000+0 0.000000+0 7.658000-5 0.000000+0 0.000000+0 0.000000+0122533103 0.000000+0 1.454000-5 0.000000+0 0.000000+0 0.000000+0 0.000000+0122533103 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0122533103 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0 1.044000-3122533103 7.213000-4 4.652000-4 2.957000-4 3.292000-4 4.231000-4 3.633000-4122533103 1.846000-4 1.460000-4 2.182000-4 1.894000-4 0.000000+0 0.000000+0122533103 4.741000-5 0.000000+0 0.000000+0 0.000000+0 0.000000+0 9.004000-6122533103 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0122533103 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0122533103 0.000000+0 0.000000+0 0.000000+0 1.838000-3 5.663000-4 4.225000-4122533103 4.067000-4 5.215000-4 4.475000-4 2.162000-4 1.709000-4 2.555000-4122533103 2.218000-4 0.000000+0 0.000000+0 5.550000-5 0.000000+0 0.000000+0122533103 0.000000+0 0.000000+0 1.054000-5 0.000000+0 0.000000+0 0.000000+0122533103 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0122533103 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0122533103 6.293000-4 2.604000-4 2.874000-4 3.654000-4 3.129000-4 1.215000-4122533103 9.602000-5 1.436000-4 1.246000-4 0.000000+0 0.000000+0 3.119000-5122533103 0.000000+0 0.000000+0 0.000000+0 0.000000+0 5.923000-6 0.000000+0122533103 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0122533103 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0122533103 0.000000+0 0.000000+0 5.057000-4 3.775000-4 3.967000-4 3.691000-4122533103 3.262000-4 2.618000-4 3.538000-4 2.978000-4 1.194000-4 3.910000-5122533103 7.673000-5 5.358000-5 3.501000-5 8.203000-5 0.000000+0 1.514000-5122533103 2.017000-5 0.000000+0 2.580000-6 0.000000+0 0.000000+0 2.982000-6122533103 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0122533103 0.000000+0 0.000000+0 0.000000+0 9.496000-4 4.958000-4 4.760000-4122533103 5.165000-4 4.152000-4 5.553000-4 4.658000-4 2.077000-4 6.798000-5122533103 1.204000-4 9.316000-5 6.087000-5 1.426000-4 0.000000+0 2.386000-5122533103 3.506000-5 0.000000+0 4.486000-6 0.000000+0 0.000000+0 5.185000-6122533103 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0122533103 0.000000+0 0.000000+0 0.000000+0 8.269000-4 4.946000-4 4.285000-4122533103 3.433000-4 4.700000-4 3.973000-4 1.372000-4 4.491000-5 1.019000-4122533103 6.155000-5 4.021000-5 9.422000-5 0.000000+0 2.002000-5 2.316000-5122533103 0.000000+0 2.964000-6 0.000000+0 0.000000+0 3.425000-6 0.000000+0122533103 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0122533103 0.000000+0 0.000000+0 7.143000-4 4.580000-4 3.675000-4 4.981000-4122533103 4.197000-4 1.628000-4 5.331000-5 1.080000-4 7.306000-5 4.773000-5122533103 1.118000-4 0.000000+0 2.130000-5 2.749000-5 0.000000+0 3.518000-6122533103 0.000000+0 0.000000+0 4.066000-6 0.000000+0 0.000000+0 0.000000+0122533103 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0122533103 1.398000-3 7.188000-4 9.864999-4 8.085000-4 3.534000-4 1.197000-4122533103 2.088000-4 1.586000-4 1.078000-4 2.427000-4 0.000000+0 4.136000-5122533103 5.967000-5 0.000000+0 7.962000-6 0.000000+0 0.000000+0 8.824000-6122533103 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0122533103 0.000000+0 0.000000+0 0.000000+0 8.888000-4 7.738000-4 6.493000-4122533103 2.855000-4 9.347000-5 1.678000-4 1.281000-4 8.368000-5 1.961000-4122533103 0.000000+0 3.323000-5 4.820000-5 0.000000+0 6.168000-6 0.000000+0122533103 0.000000+0 7.128000-6 0.000000+0 0.000000+0 0.000000+0 0.000000+0122533103 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0 1.631000-3122533103 8.744000-4 3.691000-4 1.256000-4 2.256000-4 1.656000-4 1.132000-4122533103 2.535000-4 0.000000+0 4.462000-5 6.231000-5 0.000000+0 8.360000-6122533103 0.000000+0 0.000000+0 9.215000-6 0.000000+0 0.000000+0 0.000000+0122533103 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0122533103 1.131000-3 3.060000-4 1.002000-4 1.896000-4 1.373000-4 8.970000-5122533103 2.102000-4 0.000000+0 3.747000-5 5.167000-5 0.000000+0 6.611000-6122533103 0.000000+0 0.000000+0 7.640000-6 0.000000+0 0.000000+0 0.000000+0122533103 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0122533103 7.059000-4 9.051000-5 1.168000-4 1.025000-4 8.654000-5 1.850000-4122533103 4.997000-5 3.263000-5 3.437000-5 4.754000-6 7.243000-6 4.016000-6122533103 6.730000-7 9.572000-6 1.904000-5 3.476000-5 5.464000-5 3.826000-5122533103 3.348000-5 3.995000-5 9.178000-5 3.089000-5 0.000000+0 2.849000-4122533103 1.090000-4 1.013000-4 1.121000-4 1.346000-4 6.140000-5 4.787000-5122533103 4.006000-5 8.346000-6 2.419000-6 1.289000-6 2.179000-7 3.102000-6122533103 6.121000-6 1.116000-5 1.756000-5 1.228000-5 1.075000-5 1.284000-5122533103 2.950000-5 1.000000-5 0.000000+0 3.994000-4 1.073000-4 1.165000-4122533103 1.538000-4 6.293000-5 6.038000-5 4.136000-5 1.129000-5 3.033000-6122533103 4.522000-6 3.690000-6 6.919000-6 8.987000-6 1.587000-5 2.106000-5122533103 1.747000-5 4.102000-5 3.878000-5 1.003000-4 1.366000-4 7.423000-4122533103 3.689000-4 1.499000-4 1.471000-4 1.234000-4 6.027000-5 7.013000-5122533103 7.715000-6 7.101000-6 6.377000-7 6.900000-6 8.037000-6 9.057000-6122533103 7.065000-5 9.992000-5 7.427000-5 6.883000-5 5.504000-5 1.462000-4122533103 2.198000-4 0.000000+0 3.917000-4 1.415000-4 1.163000-4 6.212000-5122533103 6.163000-5 9.348000-6 5.029000-6 1.392000-6 4.268000-6 6.262000-6122533103 6.583000-6 4.990000-5 6.943000-5 5.552000-5 4.795000-5 3.995000-5122533103 1.171000-4 1.481000-4 0.000000+0 6.314000-4 7.655000-5 6.174000-5122533103 5.043000-5 1.095000-5 6.171000-6 2.942000-6 8.958000-7 7.855000-6122533103 2.145000-5 4.271000-5 5.936000-5 4.065000-5 4.325000-5 3.943000-5122533103 5.687000-5 0.000000+0 0.000000+0 4.217000-4 5.631000-5 6.805000-5122533103 9.824000-6 6.191000-6 4.993000-6 1.063000-5 1.087000-5 2.519000-5122533103 8.153000-5 1.136000-4 9.074000-5 7.833000-5 6.488000-5 1.916000-4122533103 2.484000-4 0.000000+0 1.364000-4 2.742000-5 5.675000-6 4.078000-6122533103 2.047000-6 3.265000-6 6.286000-6 7.306000-6 3.756000-5 3.566000-5122533103 3.666000-5 3.988000-5 3.708000-5 6.163000-5 8.552000-5 1.954000-4122533103 2.193000-4 4.042000-5 3.101000-5 3.731000-5 4.210000-5 2.190000-5122533103 6.633000-6 2.502000-5 3.546000-5 2.793000-5 2.402000-5 1.868000-5122533103 5.986000-5 9.835000-5 0.000000+0 2.990000-5 1.234000-5 1.606000-5122533103 1.677000-5 8.438000-6 2.638000-6 2.727000-6 4.223000-6 3.176000-6122533103 5.216000-6 5.098000-6 1.078000-5 1.432000-5 7.229000-5 2.146000-5122533103 1.217000-5 1.304000-5 7.310000-6 1.286000-6 1.058000-5 1.034000-5122533103 1.053000-5 9.525000-6 8.492000-6 1.116000-5 1.378000-5 0.000000+0122533103 2.728000-5 1.661000-5 8.260000-6 1.834000-6 1.729000-6 2.294000-6122533103 1.904000-6 3.962000-6 4.040000-6 9.790000-6 1.247000-5 6.928000-5122533103 3.051000-5 8.964000-6 1.622000-6 5.904000-6 8.294000-6 5.994000-6122533103 8.828000-6 7.746000-6 2.092000-5 3.809000-5 9.510000-5 3.034000-5122533103 1.586000-5 1.302000-5 1.255000-5 1.343000-5 1.401000-5 1.257000-5122533103 2.020000-5 2.765000-5 6.808000-5 1.409000-4 1.474000-5 2.281000-5122533103 1.493000-5 1.676000-5 1.412000-5 2.123000-5 1.232000-5 2.577000-5122533103 4.629000-4 1.240000-4 1.180000-4 1.074000-4 9.052000-5 1.470000-4122533103 1.828000-4 0.000000+0 4.242000-4 1.303000-4 1.194000-4 9.884000-5122533103 2.045000-4 2.627000-4 0.000000+0 4.462000-4 1.101000-4 9.018000-5122533103 1.636000-4 2.041000-4 0.000000+0 3.819000-4 1.030000-4 1.966000-4122533103 2.921000-4 6.338000-4 3.281000-4 1.682000-4 2.465000-4 5.976000-4122533103 1.025000-3 7.385000-4 1.611000-3 2.111000-3 3.391000-3 2.720000-2122533103 122533 0 1225 0 0 0 0 0 1.30270E+4 2.67497E+1 0 0 34 101325 1451 0.0 0.0 0 0 0 61325 1451 1.00000E+0 2.00000E+7 0 0 10 20021325 1451 3.00000E+2 0.0 1 0 325 51325 1451 13-Al- 27 FEI/IRK EVAL-APR90 K.I.Zolotarev, M.Wagner et al. 1325 1451 DIST-Feb2004 1325 1451 ----IRDF-2002 MATERIAL 1325 1325 1451 -----INCIDENT NEUTRON DATA 1325 1451 ------ENDF-6 FORMAT 1325 1451 *****************************************************************1325 1451 13-AL- 27 FEI EVAL-May03 K.I.Zolotarev 1325 1451 DIST-Sep03 1325 1451 ----BROND-3 MATERIAL 1325 1325 1451 -----INCIDENT NEUTRON DATA 1325 1451 ------ENDF-6 FORMAT 1325 1451 ***************************************************************** 1325 1451 ------Russian Reactor Dosimetry File RRDF-2002 1325 1451 ***************************************************************** 1325 1451 Authors of evaluation: K.Zolotarev, A.Pashchenko, J.Csikai 1325 1451 ***************************************************************** 1325 1451 MF=3 1325 1451 MT=103 - (n,p) cross section 1325 1451 *****************************************************************1325 1451 13-AL- 27 IRK-VIENNA EVAL-APR90 1325 1451 DIST-JUN90 1325 1451 IRK-EVAL.NLIB 25 1325 1325 1451 MF=3 1325 1451 MT=107 - (n,a) cross section 1325 1451 ******************************************************************1325 1451 ***************************************************************** 1325 1451 ********** Start of (N,P) bibliographical component ********** 1325 1451 ***************************************************************** 1325 1451 ------Russian Reactor Dosimetry File RRDF-2002 1325 1451 ***************************************************************** 1325 1451 Authors of evaluation: K.Zolotarev, A.Pashchenko, J.Csikai 1325 1451 ***************************************************************** 1325 1451 MF=3 1325 1451 MT=103 - (n,p) cross section 1325 1451 ------------------------------------- 1325 1451 Microscopic experimental data [1-76] were analyzed in the 1325 1451 process of preparation of input data base for the evaluation of 1325 1451 cross sections and their uncertainty for the Al-27(n,p)Mg-27 1325 1451 reaction. During this procedure all experimental data if it was 1325 1451 possible were corrected to the new recommended cross section data 1325 1451 for monitor reactions used in the measurements and to the new re- 1325 1451 commended decay data from ref. [77]. 1325 1451 Excitation function for the Al-27(n,p)Mg-27 reaction in the 1325 1451 energy region from threshold to 23.0 MeV was evaluated by means 1325 1451 of statistical analysis of experimental cross section data [1-58].1325 1451 Special correction was done with experimental data [3], [8], 1325 1451 [12], [13], [15], [17], [27], [28], [47] and [58]. 1325 1451 Experimental data of Hudson and Morgan [3], Gabbard et al.[8],1325 1451 Ferguson and Albergotti [15] , Cuzzocrea et al. [17] , Csikai and 1325 1451 Chimoe et al. [47] were renormalized to the results of precise 1325 1451 absolute measurements of Ikeda et al.[53] in the overlapping ener-1325 1451 gy ranges. Correction factors for the experimental data [3], [8], 1325 1451 [15], [17] were Fc=0.87555, Fc=1.41467, Fc=1.60840, Fc=0.89474, 1325 1451 respectively. Data of independent measurements of Csikai and 1325 1451 Chimoe et al. [47] were multiplied to the coefficients Fc=0.89370 1325 1451 and Fc=0.92320, respectively. 1325 1451 Cross sections for the Al-27(n,p)Mg-27 reaction measured by 1325 1451 Bass et al. [12] in the neutron energy range 6.00 - 9.00 MeV with 1325 1451 25 keV step were recalculated by averaging original experimental 1325 1451 data over 100 keV energy intervals. 1325 1451 Experimental data of Calvi et al. [13] and Shimizu et al.[58] 1325 1451 were corrected to the results of Smith and Meadows measurements 1325 1451 with Li-7(p,n)Be-7 neutron source [27] in the overlapping energy 1325 1451 intervals. Correction factors were Fc=0.96506 and Fc=1.47768, res-1325 1451 pectively. For the experimental data [58] value Fc=1.47768 is a 1325 1451 total correction factor. On the first step of correction data of 1325 1451 Shimizu et al. [58] were renormalized to the new evaluated cross 1325 1451 sections for the monitor reaction In-115(n,n')In-115m [78]. 1325 1451 Data of Smith and Meadows [27] measured with using neutrons 1325 1451 from D(d,n)He3 reaction were renormalized to the results of this 1325 1451 experiment obtained with Li7(p,n)Be7 neutron source in the over- 1325 1451 lapping interval 5.398 - 5.870 MeV. D(d,n)He3 data in the energy 1325 1451 range 5.398 - 9.897 MeV were increased to the factor Fc=1.083 . 1325 1451 Data given in ref. [28] by Mostafa were renormalized to the 1325 1451 absolute cross section value for Al-27(n,p)Mg-27 reaction evalu- 1325 1451 ated at 7.1 MeV with taking into account experimental data [12], 1325 1451 [14] and [27]. 1325 1451 Experimental data [5], [8], [12] and [44] were used partially.1325 1451 Data of Mani et al. [5] were taken into account only for the 9 1325 1451 neutron energies: 11.92, 13.82, 14.75, 16.20, 16.63, 17.83, 18.13,1325 1451 20.15 and 21.72 MeV. Renormalized experimental data of Gabbard et 1325 1451 al. [8] were used only for the 8 neutron energies: 12.90, 13.10,1325 1451 13.70, 14.40, 14.90, 15.45, 16.80 and 17.40 MeV. Information ob- 1325 1451 tained from Bass et al. experiment [12] was taken into account in 1325 1451 the evaluation in the energy region 6.0 - 8.6 MeV. In the energy 1325 1451 interval 8.6 - 9.0 MeV this experiment gives significantly over- 1325 1451 estimated cross sections in a comparison with corrected experimen-1325 1451 tal data [24], [27] and new experimental data [55]. Cross section 1325 1451 measured by Bradley et al. [44] for the neutron energy 11 MeV was 1325 1451 not input in the data base for evaluation due to a very big dis- 1325 1451 crepancy with new experimental data of Csikai et al. [55]. 1325 1451 Experimental cross section data [59-76] were rejected due to 1325 1451 their discrepancy with the main bulk of experimental data [1-58]. 1325 1451 In the rejected experiments [59-70] and [75-76] cross section 1325 1451 values were measured only in a one energy point in the interval 1325 1451 14 - 15 MeV. 1325 1451 Statistical analysis of input cross section data was carried 1325 1451 out by means of PADE-2 code [79]. Rational function was used as 1325 1451 the model function [80]. 1325 1451 Evaluated excitation function for the reaction Al27(n,p)Mg27 1325 1451 was tested with using integral experimental data [81-83] for 1325 1451 U-235 thermal fission neutron spectrum and evaluated integral ex- 1325 1451 perimental data [83] for Cf-252 spontaneous fission neutron spec- 1325 1451 trum. Calculated and measured average cross section values for 1325 1451 U-235 thermal fission neutron spectrum [84] and Cf-252 sponta- 1325 1451 neous fission neutron spectrum [85] are given in the table 1. 1325 1451 Table 1 1325 1451 ÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÂÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÂÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄ 1325 1451 TYPE OF SPECTRUM ³ ,mb (calc.) ³ , mb (measured) 1325 1451 ÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÅÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÅÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄ 1325 1451 U-235 neutron fission ³ 4.0768 ³ 4.133 +- 0.074 [81] 1325 1451 ³ ³ 3.914 +- 0.070 [82] 1325 1451 ³ ³ 3.902 +- 0.069 [83] 1325 1451 ÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÅÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÅÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄ 1325 1451 CF-252 spont. fission ³ 4.9070 ³ 4.880 +- 0.104 [83] 1325 1451 ÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÁÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÁÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄ 1325 1451 1325 1451 MF=33 1325 1451 MT=103 - (n,p) cross section cov. matrix 1325 1451 ---------------------------------------- 1325 1451 Uncertainties in the evaluated excitation function for the 1325 1451 reaction Al-27(n,p)Mg-27 are given in the form of relative covari-1325 1451 ance matrix for the 49-neutron energy groups (LB=5). Covariance 1325 1451 matrix of uncertainties was calculated simultaneously with 1325 1451 recommended cross section data by means of PADE-2 code [79]. 1325 1451 Eigenvalues of the 6-th digits relative covariance matrix 1325 1451 given in the 33-file are the following: 1325 1451 1325 1451 9.62022E-08 9.94230E-08 1.04190E-07 1.09967E-07 1325 1451 1.15699E-07 1.20174E-07 1.24951E-07 1.30570E-07 1325 1451 1.37068E-07 1.43716E-07 1.50173E-07 1.58001E-07 1325 1451 1.68205E-07 1.76372E-07 1.86739E-07 1.94590E-07 1325 1451 2.06816E-07 2.14527E-07 2.29525E-07 2.40453E-07 1325 1451 2.53581E-07 2.82127E-07 2.99422E-07 3.49389E-07 1325 1451 3.68505E-07 4.35939E-07 5.42544E-07 7.20462E-07 1325 1451 9.45090E-07 2.44284E-06 1.77944E-05 8.88268E-05 1325 1451 1.54306E-04 2.21515E-04 2.92421E-04 3.30954E-04 1325 1451 3.45302E-04 3.84750E-04 4.56266E-04 6.07458E-04 1325 1451 1.31442E-03 1.53210E-03 1.76053E-03 2.87076E-03 1325 1451 4.69639E-03 5.64972E-03 1.04479E-02 1.48842E-02 1325 1451 5.19587E-02 1325 1451 1325 1451 References : 1325 1451 1. S.G.Forbes Phys. Rev., v.88, p.1309, December 1952 1325 1451 2. R.L.Henkel EXFOR 11524.002, 1954 1325 1451 3. O.M.Hudson jr, I.L.Morgan Bull. Am. Phys. Soc., v.4, p.97 1325 1451 (G2), March 1959 1325 1451 4. M.J.Depraz et al. Journal de Physique-Colloque, v.21, p.377, 1325 1451 May 1960 1325 1451 5. G.S.Mani et.al. Nucl. Phys., v.19, n.5, p.535, November 1960 1325 1451 6. H.Pollehn, H.Neuert Zeitschrift fuer Naturforschung, Sect. A,1325 1451 v.16, p.227, 1961 1325 1451 7. S.K.Mukherjee et al. Proc. of the Physical Society, v.77, 1325 1451 p.508, February 1961 1325 1451 8. F.Gabbard, B.D.Kern Phys. Rev., v.128, p.1276, 1962 1325 1451 9. J.Csikai et al. Atomki Koezlemenyek v.4, p.137, June 1962 ; 1325 1451 J.Csikai et al. Nucl. Phys., v.46, p.141, July 1963 1325 1451 10. C.G.Bonazzola et al. Nucl. Phys., v.51, p.337, February 1964 1325 1451 11. J.E.Strain, W.J.Ross Report ORNL-3672, January 1965 1325 1451 12. R.Bass et al. Progress Report EANDC(E)-66, p.64, Feb. 1966 1325 1451 13. 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Symposium on Neutron- 1325 1451 induced Reactions, Smolenice,CSSR, 25-29 June 1979, v.6,p.415 1325 1451 37. P.Welch et al. Bull. Amer. Phys. Soc., v.26, p.708, May 1981 1325 1451 38. R.C.Harper, W.L.Alford Jour. of Physics, Part G, v.8, p.153, 1325 1451 January 1982 1325 1451 39. S.M.Qaim Nucl. Phys. A, v.382, p.255, July 1982 1325 1451 40. A.Chiadli et al. Proc. of International Conference Nuclear 1325 1451 Data for Science and Technology, 6 - 10 September 1982, 1325 1451 Antwerp, Holland, D.Reidel Publishing Company, p.404, 1983 1325 1451 41. J.Janczyszyn Proc. of Int. Conference Nuclear Data for 1325 1451 Science and Technology, 6 - 10 September 1982, Antwerp, 1325 1451 Holland, D.Reidel Publishing Company, p.869, 1983 1325 1451 42. H.A.Husain, S.E.Hunt International J. of Applied Radiation 1325 1451 and Isotopes, v.34, no.4, p.731, 1983 1325 1451 43. V.T.Shchebolev et al. Atomnaya Energiya, v.54, no.6, p.417, 1325 1451 June 1983 1325 1451 44. D.A.Bradley et al. Proc. of Int. Symp. on Fast Neutrons in 1325 1451 Science and Technology, Chiang Mai, 4-8 February 1985, p.19 1325 1451 45. W.Enz et al. Annalen der Physik, v.42, no.3, p.283, 1985 1325 1451 46. I.Garlea et al. Rev. Roum. Phys., v.30, no.8, p.673, 1985 1325 1451 47. J.Csikai, T.Chimoye et al. Zeitschrift fuer Physik, Sec.A, 1325 1451 v.325, p.69, September 1986 1325 1451 48. J.W.Meadows et al. Ann. Nucl. Energ., v.14, p.489, Sep. 1987 1325 1451 49. Y.Ikeda,C.Konno,K.Oishi et al. Report JAERI-1312, March 1988 1325 1451 50. K.Kudo et.al. Proc. of International Conference Nuclear Data 1325 1451 for Science and Technology, Mito, Japan, 30 May - 3 June 1988,1325 1451 Saikon Publishing Co., LTD, pp.1021-1024, 1989 1325 1451 51. I.Kimura, K.Kobayashi Nucl. Sci. Eng., v.106, p.332, 1990 ; 1325 1451 K.Kobayashi, I.Kimura Proc. of the Intern. Conf. on Nuclear 1325 1451 Data for Science and Technology, 30 May - 3 June 1988, Mito, 1325 1451 Japan, Saikon Publishing Co., LTD, pp.261-265, 1989 ; 1325 1451 K.Kobayashi, I.Kimura Progress Report NEANDC(J)-116, Sep.1985 1325 1451 52. A.Ercan et al. Proc. of an Intern. Conf. on Nuclear Data for 1325 1451 Science and Technology, 13-17 May 1991, Julich, FRG, Springer-1325 1451 Verlag, 1992 1325 1451 53. Y.Ikeda et al. Journal of Nuclear Science and Technology, 1325 1451 v.30, n.9, pp.870-880, September 1993 1325 1451 54. Zhou Hongyu et al. Proc. of Int. Conf. on Nuclear Data for 1325 1451 Science and Technology, Gatlinburg, Tennessee, USA, May 9-13, 1325 1451 1994, Vol.1, pp. 166-169 1325 1451 55. J.Csikai et al. Private communication, Debrecen, April 1998 1325 1451 56. A.A.Filatenkov et al. Report RI-252, St.Petersburg, May 1999; 1325 1451 A.A.Filatenkov et al. VANT, Ser.:Yadernye Konstanty, v.2, p.8,1325 1451 Moscow, 1996 1325 1451 57. A.Fessler, A.J.M.Plompen et al. Nucl. Sci. Eng., v.134, no.2, 1325 1451 pp.171-200, February 2000 1325 1451 58. T.Shimizu, S.Furuichi, H.Sakane, M.Shibata, K.Kawade 1325 1451 Proc. of The 2000 Symposium on Nuclear Data, November 16-17, 1325 1451 2000 JAERI, Tokai, Japan, pp.194-199 1325 1451 59. E.B.Paul, R.L.Clarke Canadian Journal of Physics, v.31, 1325 1451 p.267, 1953 1325 1451 60. S.Yasumi Journal of the Physical Society of Japan, v.12, 1325 1451 p.443, May 1957 1325 1451 61. G.Brown et al. Philosophical Magazine, v.2, p.785, 1957 1325 1451 62. A.Poularikas, R.W.Fink Phys. Rev., v.115, p.989, 1959 1325 1451 63. C.S.Khurana, H.S.Hans Proc. of 4th Nuclear Physics and Solid 1325 1451 Stata Physics Symp., 24-26 Feb. 1960, Waltair, India, p.297 1325 1451 64. R.S.Storey,W.Jack,A.Ward Proc. Phys. Soc., v.75, p.526, 1960 1325 1451 65. M.Sakisaka et al. Journal of the Physical Society of Japan, 1325 1451 v.16, p.1869, October 1961 1325 1451 66. J.Kantele, D.G.Gardner Nucl. Phys, v.35, p.353, 1962 1325 1451 67. W.Langmann EXFOR 20903.003, September 1962 1325 1451 68. F.L.Hassler, R.A.Peck jr Phys. Rev., v.125, p.1011, 1962 1325 1451 69. B.Mitra, A.M.Ghose Nucl. Phys., v.83, p.157, July 1966 1325 1451 70. R.Prasad, D.C.Sarkar Nuovo Cimento, v.A3, no.3, p.467, 1971 1325 1451 71. J.C.Robertson, K.J.Zieba Annals of Nucl. Energy, v.26, no.1, 1325 1451 p.1, 1972 1325 1451 72. R.A.Jarjis J. of Physics, pt.G, v.4, n.3, p.445, 1978 1325 1451 73. J.Csikai Proc. of International Conference Nuclear Data for 1325 1451 Science and Technology, 6 - 10 September 1982, Antwerp, 1325 1451 Holland, D.Reidel Publishing Company, p.414, 1983 1325 1451 74. Tahir Indian Journal of Pure and Applied Physics, v.23, 1325 1451 p.439, September 1985 1325 1451 75. J.P.Gupta et.al. Indian J. Pramana, v.24, p.637, 1985 1325 1451 76. L.I.Klochkova et al. Vopr. Atomn. Nauki i Tekhn., Serija: 1325 1451 Jadernye Konstanty v.1, p.27, 1992 ; 1325 1451 L.I.Klochkova et al. Proc. of the 1-st Int. Conf. on Neutron 1325 1451 Phys., Kiev, USSR, 14 - 18 Sep.1987, v.3, p.315, Moscow 1988 1325 1451 77. R.B.Firestone Table of Isotopes, Eighth edition, Vol. 1, 1325 1451 John Wiley & Sons, Inc., New York, 1995 1325 1451 78. K.I.Zolotarev, P.K.Zolotarev The In-115(n,n')In-115m reaction 1325 1451 excitation function in the energy range from threshold to 1325 1451 20 MeV, Obninsk, IPPE, eval. March 2003 1325 1451 79. S.A.Badikov et al. Preprint FEI-1686, Obninsk, 1985 1325 1451 80. S.Badikov, N.Rabotnov, K.Zolotarev Proc. of NEANSC Speciali- 1325 1451 st's Meeting on Evaluation and Processing of Covariance Data, 1325 1451 Oak Ridge, USA, 1992, OECD, Paris, 1993, p.105 1325 1451 81. O.Horibe et al. Proc. of Conf.:50 Years with Nuclear Fission, 1325 1451 Washington D.C., 25-28 April 1989, v.2, p.923 1325 1451 82. W.Mannhart Prog. Report INDC(Ger)-045, pp.40-43, June 1999 1325 1451 83. W.Mannhart Validation of Differential Cross Sections with 1325 1451 Integral Data , Report INDC(NDS)-435, pp.59-64, IAEA, Vienna, 1325 1451 September 2002 1325 1451 84. L.W.Weston et al. Evaluated Neutron Data for Uranium-235, 1325 1451 ENDF/B-VI Library, MAT=9228, MF=5, MT=18, eval. April 1989 1325 1451 85. W.Mannhart IAEA-TECDOC-410, p.158, IAEA, Vienna, 1987 1325 1451 ***************************************************************** 1325 1451 ********** End of (N,2N) bibliographical component ********** 1325 1451 ================================================================= 1325 1451 ***************************************************************** 1325 1451 The Q values and threshold energies were updated prior to pro- 1325 1451 cessing through the codes to comply with the values obtained 1325 1451 using the NNDC calculation program which is based on the 1995 1325 1451 Update to the Atomic mass Evaluation.Additional points were 1325 1451 added near threshold for MF/MT=3/107 1325 1451 File 2 added to the pointwise file containing only the effective 1325 1451 scattering radius with no resonance parameters given. 1325 1451 Taken from ENDF/B-VI 1325 1451 ***************************************************************** 1325 1451 ***************** Program LINEAR (VERSION 2002-1) ***************1325 1451 For All Data Greater than 1.0000E-10 barns in Absolute Value 1325 1451 Data Linearized to Within an Accuracy of .100000000 per-cent 1325 1451 ***************** Program SIGMA1 (VERSION 2002-1) ***************1325 1451 Data Doppler Broadened to 300.000000 Kelvin 1325 1451 for All Data Greater than 1.0000E-10 barns in Absolute Value 1325 1451 Data Linearized to Within an Accuracy pf .100000000 per-cent 1325 1451 ***************** Program FIXUP (Version 2002-1) ****************1325 1451 Corrected ZA/AWR in All Sections-----------------------------Yes 1325 1451 Corrected Thresholds-----------------------------------------Yes 1325 1451 Extended Cross Sections to 20 MeV----------------------------No 1325 1451 Allow Cross Section Deletion---------------------------------No 1325 1451 Allow Cross Section Reconstruction---------------------------No 1325 1451 Make All Cross Sections Non-Negative-------------------------Yes 1325 1451 Delete Energies Not in Ascending Order-----------------------Yes 1325 1451 Deleted Duplicate Points-------------------------------------Yes 1325 1451 Check for Ascending MAT/MF/MT Order--------------------------Yes 1325 1451 Check for Legal MF/MT Numbers--------------------------------Yes 1325 1451 Allow Creation of Missing Sections---------------------------No 1325 1451 Allow Insertion of Energy Points-----------------------------No 1325 1451 Create Uniform Energy Grid-----------------------------------No 1325 1451 Delete Section if Cross Section =0 at All Energies-----------Yes 1325 1451 ***************** Program GROUPIE (VERSION 2002-1) **************1325 1451 Unshielded Group Averages Using 640 Groups 1325 1451 Weighting Spectrum: Flat (Constant) Spectrum 1325 1451 1 451 334 11325 1451 3 103 64 11325 1451 3 107 60 11325 1451 33 103 224 11325 1451 33 107 224 11325 1451 1325 1 0 1325 0 0 1.30270E+4 2.67497E+1 0 0 0 01325 3103 -1.82797E+6-1.82797E+6 0 0 1 1831325 3103 183 1 1325 3103 1800000.00 7.5592E-10 1900000.00 5.94923E-7 2000000.00 1.70291E-61325 3103 2100000.00 2.81090E-6 2200000.00 3.91889E-6 2300000.00 5.02688E-61325 3103 2400000.00 1.36533E-5 2500000.00 5.31887E-5 2600000.00 9.60656E-51325 3103 2700000.00 .000197747 2800000.00 .000422101 2900000.00 .0007430851325 3103 3000000.00 .001163323 3100000.00 .001664055 3200000.00 .0021825181325 3103 3300000.00 .002660038 3400000.00 .005813040 3500000.00 .0059058401325 3103 3600000.00 .005923953 3700000.00 .007376778 3800000.00 .0053420301325 3103 3900000.00 .006590330 4000000.00 .006400630 4100000.00 .0058240901325 3103 4200000.00 .010016250 4300000.00 .011382425 4400000.00 .0120621751325 3103 4500000.00 .017318350 4600000.00 .018967950 4700000.00 .0161955251325 3103 4800000.00 .015435425 4900000.00 .023968600 5000000.00 .0268780501325 3103 5100000.00 .028133700 5200000.00 .028140700 5300000.00 .0276018001325 3103 5400000.00 .036424900 5500000.00 .041505550 5600000.00 .0478750101325 3103 5700000.00 .058418765 5800000.00 .042922684 5900000.00 .0452353451325 3103 6000000.00 .048209582 6100000.00 .051044500 6200000.00 .0536079001325 3103 6300000.00 .055899200 6400000.00 .057914600 6500000.00 .0596261001325 3103 6600000.00 .061243488 6700000.00 .062934663 6800000.00 .0646258381325 3103 6900000.00 .066317013 7000000.00 .067955880 7100000.00 .0695424401325 3103 7200000.00 .071129000 7300000.00 .072715560 7400000.00 .0743021201325 3103 7500000.00 .075839390 7600000.00 .077327370 7700000.00 .0788153501325 3103 7800000.00 .080303330 7900000.00 .081791310 8000000.00 .0832272701325 3103 8100000.00 .084611210 8200000.00 .085995150 8300000.00 .0873790901325 3103 8400000.00 .088763030 8500000.00 .090087063 8600000.00 .0913511881325 3103 8700000.00 .092615313 8800000.00 .093879438 8900000.00 .0950713751325 3103 9000000.00 .096191125 9100000.00 .097310875 9200000.00 .0984306251325 3103 9300000.00 .099468917 9400000.00 .100425750 9500000.00 .1013825831325 3103 9600000.00 .102253820 9700000.00 .103039461 9800000.00 .1038251011325 3103 9900000.00 .104511118 10000000.0 .105097471 10100000.0 .1056838241325 3103 10200000.0 .106158000 10300000.0 .106520000 10400000.0 .1068820001325 3103 10500000.0 .107122667 10600000.0 .107242000 10700000.0 .1073613331325 3103 10800000.0 .107354500 10900000.0 .107221500 11000000.0 .1070885001325 3103 11100000.0 .106830333 11200000.0 .106447000 11300000.0 .1060636671325 3103 11400000.0 .105561500 11500000.0 .104940500 11600000.0 .1043195001325 3103 11700000.0 .103591167 11800000.0 .102755500 11900000.0 .1019198331325 3103 12000000.0 .100992733 12100000.0 .099974200 12200000.0 .0989556671325 3103 12300000.0 .097854738 12400000.0 .096671413 12500000.0 .0954880881325 3103 12600000.0 .094304763 12700000.0 .093050575 12800000.0 .0917255251325 3103 12900000.0 .090400475 13000000.0 .089075425 13100000.0 .0877503751325 3103 13200000.0 .086425325 13300000.0 .085082972 13400000.0 .0837233171325 3103 13500000.0 .082363661 13600000.0 .081004006 13700000.0 .0796443501325 3103 13800000.0 .078284694 13900000.0 .076925039 14000000.0 .0755653831325 3103 14100000.0 .074205728 14200000.0 .072891140 14300000.0 .0716216201325 3103 14400000.0 .070352100 14500000.0 .069082580 14600000.0 .0678130601325 3103 14700000.0 .066586183 14800000.0 .065401950 14900000.0 .0642177171325 3103 15000000.0 .063087163 15100000.0 .062010288 15200000.0 .0609334131325 3103 15300000.0 .059856538 15400000.0 .058833300 15500000.0 .0578637001325 3103 15600000.0 .056894100 15700000.0 .055924500 15800000.0 .0550073001325 3103 15900000.0 .054142500 16000000.0 .053277700 16100000.0 .0524129001325 3103 16200000.0 .051592388 16300000.0 .050816163 16400000.0 .0500399381325 3103 16500000.0 .049263713 16600000.0 .048525338 16700000.0 .0478248131325 3103 16800000.0 .047124288 16900000.0 .046423763 17000000.0 .0457559751325 3103 17100000.0 .045120925 17200000.0 .044485875 17300000.0 .0438508251325 3103 17400000.0 .043247580 17500000.0 .042676140 17600000.0 .0421047001325 3103 17700000.0 .041533260 17800000.0 .040961820 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DIST-JAN03 1400 1451 ----IRDF-90/NMF-G MATERIAL 1400 1400 1451 -----INCIDENT NEUTRON DATA 1400 1451 ------ENDF-6 FORMAT 1400 1451 DAMAGE CROSS SECTION OF SILICON (ELECTRONICS DAMAGE)1400 1451 ASTM E722 MT=900 1400 1451 1400 1451 1 451 14 01400 1451 3 900 217 01400 1451 1400 1 0 1400 0 0 1.40000+ 4 0.0 + 0 0 0 0 01400 3900 0.0 + 0 0.0 + 0 0 0 1 6411400 3900 641 1 1400 3900 1.0000E-04 1.5471E+00 1.0500E-04 1.5117E+00 1.1000E-04 1.4764E+001400 3900 1.1500E-04 1.4450E+00 1.2000E-04 1.4081E+00 1.2750E-04 1.3668E+001400 3900 1.3500E-04 1.3294E+00 1.4250E-04 1.2948E+00 1.5000E-04 1.2575E+001400 3900 1.6000E-04 1.2189E+00 1.7000E-04 1.1834E+00 1.8000E-04 1.1511E+001400 3900 1.9000E-04 1.1225E+00 2.0000E-04 1.0945E+00 2.1000E-04 1.0678E+001400 3900 2.2000E-04 1.0447E+00 2.3000E-04 1.0217E+00 2.4000E-04 9.9545E-011400 3900 2.5500E-04 9.6654E-01 2.7000E-04 9.4426E-01 2.8000E-04 9.1947E-011400 3900 3.0000E-04 8.8938E-01 3.2000E-04 8.6182E-01 3.4000E-04 8.3681E-011400 3900 3.6000E-04 8.1396E-01 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1.8300E+07 1.8397E+021400 3900 1.8400E+07 1.8431E+02 1.8500E+07 1.8445E+02 1.8600E+07 1.8437E+021400 3900 1.8700E+07 1.8428E+02 1.8800E+07 1.8420E+02 1.8900E+07 1.8411E+021400 3900 1.9000E+07 1.8400E+02 1.9100E+07 1.8388E+02 1.9200E+07 1.8375E+021400 3900 1.9300E+07 1.8363E+02 1.9400E+07 1.8351E+02 1.9500E+07 1.8338E+021400 3900 1.9600E+07 1.8325E+02 1.9700E+07 1.8312E+02 1.9800E+07 1.8300E+021400 3900 1.9900E+07 1.8287E+02 2.0000E+07 0.0000E+00 1400 3900 1400 3 0 1400 0 0 0 0 0 1.50310E+4 3.07076E+1 0 0 34 101525 1451 0.0 0.0 0 0 0 61525 1451 1.00000E+0 2.05000E+7 0 0 10 20021525 1451 3.00000E+2 0.0 1 0 40 31525 1451 15-P - 31 IRK-VIENNA EVAL-JUN80 1525 1451 DIST-Feb2004 1525 1451 ----IRDF-2002 MATERIAL 1525 1525 1451 -----INCIDENT NEUTRON DATA 1525 1451 ------ENDF-6 FORMAT 1525 1451 *****************************************************************1525 1451 DIST-JUN90 1525 1451 IRK-EVAL.NLIB 25 1525 1525 1451 *****************************************************************1525 1451 The Q values and threshold energies were updated prior to pro- 1525 1451 cessing through the codes to comply with the values obtained 1525 1451 using the NNDC calculation program which is based on the 1995 1525 1451 Update to the Atomic mass Evaluation. 1525 1451 Additional points were added near threshold for mf/mt= 3/103 1525 1451 *****************************************************************1525 1451 ***************** Program LINEAR (VERSION 2002-1) ***************1525 1451 For All Data Greater than 1.0000E-10 barns in Absolute Value 1525 1451 Data Linearized to Within an Accuracy of .100000000 per-cent 1525 1451 ***************** Program SIGMA1 (VERSION 2002-1) ***************1525 1451 Data Doppler Broadened to 300.000000 Kelvin 1525 1451 for All Data Greater than 1.0000E-10 barns in Absolute Value 1525 1451 Data Linearized to Within an Accuracy pf .100000000 per-cent 1525 1451 ***************** Program FIXUP (Version 2002-1) ****************1525 1451 Corrected ZA/AWR in All Sections-----------------------------Yes 1525 1451 Corrected Thresholds-----------------------------------------Yes 1525 1451 Extended Cross Sections to 20 MeV----------------------------No 1525 1451 Allow Cross Section Deletion---------------------------------No 1525 1451 Allow Cross Section Reconstruction---------------------------No 1525 1451 Make All Cross Sections Non-Negative-------------------------Yes 1525 1451 Delete Energies Not in Ascending Order-----------------------Yes 1525 1451 Deleted Duplicate Points-------------------------------------Yes 1525 1451 Check for Ascending MAT/MF/MT Order--------------------------Yes 1525 1451 Check for Legal MF/MT Numbers--------------------------------Yes 1525 1451 Allow Creation of Missing Sections---------------------------No 1525 1451 Allow Insertion of Energy Points-----------------------------No 1525 1451 Create Uniform Energy Grid-----------------------------------No 1525 1451 Delete Section if Cross Section =0 at All Energies-----------Yes 1525 1451 ***************** Program GROUPIE (VERSION 2002-1) **************1525 1451 Unshielded Group Averages Using 640 Groups 1525 1451 Weighting Spectrum: Flat (Constant) Spectrum 1525 1451 1 451 47 11525 1451 3 103 69 11525 1451 33 103 108 11525 1451 1525 1 0 1525 0 0 1.50310E+4 3.07076E+1 0 0 0 01525 3103 -7.09680E+5-7.09680E+5 0 0 1 1981525 3103 198 1 1525 3103 720000.000 4.99267E-7 760000.000 2.54695E-6 800000.000 4.70496E-61525 3103 840000.000 6.86298E-6 880000.000 9.02099E-6 920000.000 1.43222E-51525 3103 960000.000 2.27667E-5 1000000.00 3.75444E-5 1100000.00 8.10667E-51525 3103 1200000.00 .000175412 1300000.00 .000440231 1400000.00 .0007334621525 3103 1500000.00 .001126279 1600000.00 .003710000 1700000.00 .0064587501525 3103 1800000.00 .007420000 1900000.00 .008352500 2000000.00 .0093375001525 3103 2100000.00 .014567500 2200000.00 .024042500 2300000.00 .0318600001525 3103 2400000.00 .038020000 2500000.00 .046500000 2600000.00 .0573000001525 3103 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6.217000-4-1.125000-3-3.301000-3-5.194000-3152533103 -6.656000-3-7.841000-3 1.439000-3 4.791000-4-7.138000-5-9.163000-4152533103 -1.679000-3-2.276000-3-2.766000-3 3.212000-4 5.254000-4 5.616000-4152533103 5.582000-4 5.450000-4 5.245000-4 1.899000-3 3.081000-3 4.045000-3152533103 4.771000-3 5.329000-3 5.515000-3 7.581000-3 9.161000-3 1.039000-2152533103 1.064000-2 1.300000-2 1.484000-2 1.597000-2 1.828000-2 2.098000-2152533103 152533 0 1525 0 0 0 0 0 1.60320E+4 3.16974E+1 0 0 34 101625 1451 0.0 0.0 0 0 0 61625 1451 1.00000E+0 2.00000E+7 0 0 10 20021625 1451 3.00000E+2 0.0 1 0 45 31625 1451 16-S - 32 1625 1451 DIST-Feb2004 1625 1451 ----IRDF-2002 MATERIAL 1625 1625 1451 -----INCIDENT NEUTRON DATA 1625 1451 ------ENDF-6 FORMAT 1625 1451 ******************************************************************1625 1451 US EVALUATION 1989 DIST-FEB91 1625 1451 ENDF/B-VI MATERIAL 1625 1625 1451 *****************EXTRACT FOR SPECIAL PURPOSE FILE*****************1625 1451 DOSIMETRY 1625 1451 ******************************************************************1625 1451 The Q values and threshold energies were updated prior to pro- 1625 1451 cessing through the codes to comply with the values obtained 1625 1451 using the NNDC calculation program which is based on the 1995 1625 1451 Update to the Atomic mass Evaluation. 1625 1451 1625 1451 File 2 added to the pointwise file containing only the effective 1625 1451 scattering radius with no resonance parameters given. 1625 1451 Taken from ENDF/B-VI 1625 1451 ******************************************************************1625 1451 ***************** Program LINEAR (VERSION 2002-1) ***************1625 1451 For All Data Greater than 1.0000E-10 barns in Absolute Value 1625 1451 Data Linearized to Within an Accuracy of .100000000 per-cent 1625 1451 ***************** Program SIGMA1 (VERSION 2002-1) ***************1625 1451 Data Doppler Broadened to 300.000000 Kelvin 1625 1451 for All Data Greater than 1.0000E-10 barns in Absolute Value 1625 1451 Data Linearized to Within an Accuracy pf .100000000 per-cent 1625 1451 ***************** Program FIXUP (Version 2002-1) ****************1625 1451 Corrected ZA/AWR in All Sections-----------------------------Yes 1625 1451 Corrected Thresholds-----------------------------------------Yes 1625 1451 Extended Cross Sections to 20 MeV----------------------------No 1625 1451 Allow Cross Section Deletion---------------------------------No 1625 1451 Allow Cross Section Reconstruction---------------------------No 1625 1451 Make All Cross Sections Non-Negative-------------------------Yes 1625 1451 Delete Energies Not in Ascending Order-----------------------Yes 1625 1451 Deleted Duplicate Points-------------------------------------Yes 1625 1451 Check for Ascending MAT/MF/MT Order--------------------------Yes 1625 1451 Check for Legal MF/MT Numbers--------------------------------Yes 1625 1451 Allow Creation of Missing Sections---------------------------No 1625 1451 Allow Insertion of Energy Points-----------------------------No 1625 1451 Create Uniform Energy Grid-----------------------------------No 1625 1451 Delete Section if Cross Section =0 at All Energies-----------Yes 1625 1451 ***************** Program GROUPIE (VERSION 2002-1) **************1625 1451 Unshielded Group Averages Using 640 Groups 1625 1451 Weighting Spectrum: Flat (Constant) Spectrum 1625 1451 1 451 52 11625 1451 3 103 68 11625 1451 33 103 183 11625 1451 1625 1 0 1625 0 0 1.60320E+4 3.16974E+1 0 0 0 01625 3103 0.0 -9.28310E+5 0 0 1 1931625 3103 193 1 1625 3103 920000.000 2.17952E-8 960000.000 6.76274E-6 1000000.00 1.89000E-51625 3103 1100000.00 3.75000E-5 1200000.00 8.10000E-5 1300000.00 .0001710001625 3103 1400000.00 .000350000 1500000.00 .000688000 1600000.00 .0014590001625 3103 1700000.00 .002604000 1800000.00 .006027000 1900000.00 .0090950001625 3103 2000000.00 .013095000 2100000.00 .031680400 2200000.00 .0543095001625 3103 2300000.00 .068355500 2400000.00 .071605250 2500000.00 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1.2742E-01162533103 6.0000E+06 8.9710E-02 6.5000E+06 5.4451E-02 7.0000E+06 3.8618E-02162533103 8.5000E+06 3.7710E-02 9.0000E+06 3.8965E-02 9.5000E+06 3.9717E-02162533103 1.0000E+07 3.6543E-02 1.0500E+07 3.6762E-02 1.2000E+07 4.0115E-02162533103 1.3500E+07 4.2948E-02 1.4000E+07 1.1829E-01 1.6000E+07 1.3460E-01162533103 1.7000E+07 1.3626E-01 1.8000E+07 1.4464E-01 1.9000E+07 1.5243E-01162533103 2.0000E+07 0.0000E+00 0.0000E+00 0.0000E+00 0.0000E+00 0.0000E+00162533103 0.0000E+00 0.0000E+00 0 1 198 99162533103 1.0000E-05 0.0000E+00 9.5700E+05 1.0324E-01 1.0000E+06 1.0324E-01162533103 1.1000E+06 1.0324E-01 1.2000E+06 1.0324E-01 1.3000E+06 1.0324E-01162533103 1.4000E+06 1.0324E-01 1.5000E+06 1.0324E-01 1.5800E+06 1.4837E-01162533103 1.6300E+06 4.6608E-02 1.6800E+06 7.7112E-02 1.7800E+06 1.7314E-01162533103 1.8300E+06 4.6150E-02 1.8800E+06 7.3861E-02 1.9200E+06 1.1358E-01162533103 1.9600E+06 1.5377E-01 2.0000E+06 2.1910E-01 2.0400E+06 3.2471E-01162533103 2.0800E+06 2.6614E-01 2.1200E+06 2.0222E-01 2.1600E+06 1.9961E-02162533103 2.2000E+06 7.4455E-03 2.2500E+06 1.4202E-02 2.3000E+06 1.1971E-02162533103 2.3500E+06 3.8080E-02 2.4000E+06 9.7937E-03 2.4500E+06 1.4403E-02162533103 2.5000E+06 2.3565E-02 2.5500E+06 9.4860E-03 2.6000E+06 2.3963E-02162533103 2.6500E+06 1.4888E-02 2.7000E+06 6.0406E-02 2.7500E+06 2.3094E-02162533103 2.8000E+06 9.7642E-03 2.8400E+06 1.4338E-02 2.9000E+06 1.1818E-02162533103 2.9500E+06 1.6879E-02 3.0000E+06 1.9804E-02 3.0500E+06 3.0368E-02162533103 3.1000E+06 2.3681E-02 3.1500E+06 3.4770E-02 3.2000E+06 2.0475E-02162533103 3.2400E+06 5.1692E-02 3.2800E+06 2.6278E-02 3.3300E+06 3.0475E-02162533103 3.3700E+06 2.0147E-02 3.4000E+06 1.7299E-02 3.4400E+06 2.3325E-02162533103 3.4700E+06 2.6845E-02 3.5100E+06 2.0098E-02 3.5500E+06 9.6267E-03162533103 3.6000E+06 2.0321E-02 3.6500E+06 1.4531E-02 3.7000E+06 2.7538E-02162533103 3.7500E+06 4.8133E-02 3.8000E+06 3.0535E-02 3.8500E+06 4.1763E-02162533103 3.9000E+06 7.4438E-02 3.9400E+06 1.2848E-01 3.9800E+06 6.2225E-02162533103 4.0400E+06 1.7156E-02 4.1000E+06 1.0695E-01 4.1900E+06 1.3013E-01162533103 4.3000E+06 4.6579E-02 4.4100E+06 4.6579E-02 4.4500E+06 4.6579E-02162533103 4.6000E+06 4.6579E-02 4.7500E+06 4.6579E-02 4.8200E+06 4.6579E-02162533103 4.8900E+06 4.6579E-02 5.0000E+06 4.6579E-02 5.2000E+06 7.7413E-02162533103 5.4000E+06 7.7413E-02 5.6000E+06 1.4347E-02 5.8000E+06 1.4611E-02162533103 6.0000E+06 7.2431E-03 6.5000E+06 2.6684E-03 7.0000E+06 7.6506E-04162533103 7.5000E+06 7.6506E-04 8.0000E+06 7.6506E-04 8.5000E+06 1.2798E-03162533103 9.0000E+06 1.3664E-03 9.5000E+06 1.4197E-03 1.0000E+07 1.2019E-03162533103 1.0500E+07 5.4733E-04 1.1000E+07 5.4733E-04 1.1500E+07 5.4733E-04162533103 1.2000E+07 1.0573E-03 1.2500E+07 1.0573E-03 1.3000E+07 1.0573E-03162533103 1.3500E+07 1.6601E-03 1.4000E+07 1.0578E-02 1.4500E+07 1.0578E-02162533103 1.5000E+07 1.0578E-02 1.6000E+07 1.6305E-02 1.7000E+07 1.6710E-02162533103 1.8000E+07 1.8829E-02 1.9000E+07 2.0912E-02 2.0000E+07 0.0000E+00162533103 0.0000E+00 0.0000E+00 0 8 198 99162533103 1.0000E-05 0.0000E+00 9.5700E+05 4.6987E-13 1.0000E+06 4.0977E-12162533103 1.1000E+06 1.6132E-11 1.2000E+06 7.5263E-11 1.3000E+06 3.3543E-10162533103 1.4000E+06 1.4052E-09 1.5000E+06 4.6255E-09 1.5800E+06 1.8175E-08162533103 1.6300E+06 1.1652E-08 1.6800E+06 4.5325E-08 1.7800E+06 3.0780E-07162533103 1.8300E+06 2.0352E-07 1.8800E+06 5.1869E-07 1.9200E+06 9.9963E-07162533103 1.9600E+06 1.6242E-06 2.0000E+06 2.7096E-06 2.0400E+06 6.4783E-06162533103 2.0800E+06 1.1072E-05 2.1200E+06 1.9868E-05 2.1600E+06 3.3565E-06162533103 2.2000E+06 1.9116E-06 2.2500E+06 5.7852E-06 2.3000E+06 5.7230E-06162533103 2.3500E+06 2.1399E-05 2.4000E+06 5.8037E-06 2.4500E+06 7.8821E-06162533103 2.5000E+06 1.2628E-05 2.5500E+06 5.9731E-06 2.6000E+06 1.4956E-05162533103 2.6500E+06 5.6757E-06 2.7000E+06 3.0742E-05 2.7500E+06 2.1495E-05162533103 2.8000E+06 9.6057E-06 2.8400E+06 1.4095E-05 2.9000E+06 1.4186E-05162533103 2.9500E+06 2.7937E-05 3.0000E+06 4.9213E-05 3.0500E+06 9.8839E-05162533103 3.1000E+06 7.4618E-05 3.1500E+06 9.1742E-05 3.2000E+06 4.4272E-05162533103 3.2400E+06 1.2001E-04 3.2800E+06 9.5391E-05 3.3300E+06 1.4439E-04162533103 3.3700E+06 1.0161E-04 3.4000E+06 8.4846E-05 3.4400E+06 1.1310E-04162533103 3.4700E+06 1.3775E-04 3.5100E+06 1.0996E-04 3.5500E+06 5.8369E-05162533103 3.6000E+06 1.2237E-04 3.6500E+06 7.8251E-05 3.7000E+06 1.1350E-04162533103 3.7500E+06 1.4525E-04 3.8000E+06 8.7453E-05 3.8500E+06 1.7881E-04162533103 3.9000E+06 4.5482E-04 3.9400E+06 9.2900E-04 3.9800E+06 6.0536E-04162533103 4.0400E+06 2.0134E-04 4.1000E+06 1.3697E-03 4.1900E+06 1.7611E-03162533103 4.3000E+06 6.2745E-04 4.4100E+06 5.0611E-04 4.4500E+06 4.0547E-04162533103 4.6000E+06 4.0395E-04 4.7500E+06 3.7820E-04 4.8200E+06 3.4763E-04162533103 4.8900E+06 3.0327E-04 5.0000E+06 2.5545E-04 5.2000E+06 3.9126E-04162533103 5.4000E+06 4.9800E-04 5.6000E+06 1.2711E-04 5.8000E+06 1.4110E-04162533103 6.0000E+06 7.1879E-05 6.5000E+06 2.8119E-05 7.0000E+06 8.8116E-06162533103 7.5000E+06 8.4801E-06 8.0000E+06 8.4522E-06 8.5000E+06 1.5461E-05162533103 9.0000E+06 1.7429E-05 9.5000E+06 2.0092E-05 1.0000E+07 1.8530E-05162533103 1.0500E+07 9.0846E-06 1.1000E+07 9.2760E-06 1.1500E+07 9.1510E-06162533103 1.2000E+07 1.6018E-05 1.2500E+07 1.3364E-05 1.3000E+07 1.1289E-05162533103 1.3500E+07 1.4055E-05 1.4000E+07 6.9831E-05 1.4500E+07 5.7844E-05162533103 1.5000E+07 4.2055E-05 1.6000E+07 3.8698E-05 1.7000E+07 2.3904E-05162533103 1.8000E+07 1.8309E-05 1.9000E+07 1.5467E-05 2.0000E+07 0.0000E+00162533103 0.0000E+00 0.0000E+00 0 1 118 59162533103 1.0000E-05 0.0000E+00 9.5700E+05 1.9119E-01 1.5800E+06 1.6485E-01162533103 1.6300E+06 5.1787E-02 1.6800E+06 8.5681E-02 1.7800E+06 1.9237E-01162533103 1.8300E+06 5.1277E-02 1.8800E+06 8.2068E-02 1.9200E+06 1.2620E-01162533103 1.9600E+06 1.7086E-01 2.0000E+06 2.4344E-01 2.0400E+06 3.6079E-01162533103 2.0800E+06 2.9571E-01 2.1200E+06 2.2469E-01 2.1600E+06 2.2179E-02162533103 2.2000E+06 8.2727E-03 2.2500E+06 1.5780E-02 2.3000E+06 1.3301E-02162533103 2.3500E+06 4.2311E-02 2.4000E+06 1.0882E-02 2.4500E+06 1.6003E-02162533103 2.5000E+06 2.6183E-02 2.5500E+06 1.0540E-02 2.6000E+06 2.6626E-02162533103 2.6500E+06 1.6542E-02 2.7000E+06 6.7118E-02 2.7500E+06 2.5660E-02162533103 2.8000E+06 1.0849E-02 2.8400E+06 1.5931E-02 2.9000E+06 1.3131E-02162533103 2.9500E+06 1.8754E-02 3.0000E+06 2.2004E-02 3.0500E+06 3.3742E-02162533103 3.1000E+06 2.6312E-02 3.1500E+06 3.8633E-02 3.2000E+06 2.2750E-02162533103 3.2400E+06 5.7436E-02 3.2800E+06 2.9198E-02 3.3300E+06 3.3861E-02162533103 3.3700E+06 2.2386E-02 3.4000E+06 1.9221E-02 3.4400E+06 2.5917E-02162533103 3.4700E+06 2.9828E-02 3.5100E+06 2.2332E-02 3.5500E+06 1.0696E-02162533103 3.6000E+06 2.2579E-02 3.6500E+06 1.6146E-02 3.7000E+06 3.0597E-02162533103 3.7500E+06 5.3481E-02 3.8000E+06 3.3928E-02 3.8500E+06 4.6404E-02162533103 3.9000E+06 8.2709E-02 3.9400E+06 1.4276E-01 3.9800E+06 6.9139E-02162533103 4.0400E+06 1.9062E-02 4.1000E+06 1.1884E-01 4.1900E+06 1.4459E-01162533103 4.3000E+06 5.3355E-02 5.2000E+06 0.0 162533103 0.0 0.0 1 5 210 20162533103 1.0000E-05 5.2000E+06 5.6000E+06 5.8000E+06 6.0000E+06 6.5000E+06162533103 7.0000E+06 8.5000E+06 9.0000E+06 9.5000E+06 1.0000E+07 1.0500E+07162533103 1.2000E+07 1.3500E+07 1.4000E+07 1.6000E+07 1.7000E+07 1.8000E+07162533103 1.9000E+07 2.0000E+07 0.0000E+00 0.0000E+00 0.0000E+00 0.0000E+00162533103 0.0000E+00 0.0000E+00 0.0000E+00 0.0000E+00 0.0000E+00 0.0000E+00162533103 0.0000E+00 0.0000E+00 0.0000E+00 0.0000E+00 0.0000E+00 0.0000E+00162533103 0.0000E+00 0.0000E+00 0.0000E+00 2.6879E-01-1.6917E-04 7.6244E-05162533103 7.6222E-04 6.2428E-04 5.0197E-04 5.2466E-04 5.5936E-04 5.5109E-04162533103 5.3023E-04 5.2522E-04 4.1067E-04 3.3731E-04 7.0923E-04 7.5401E-04162533103 1.2360E-03 1.2362E-03 1.2365E-03 1.5941E-02 1.3651E-02 9.8297E-04162533103 6.7245E-04 5.9382E-04 6.0350E-04 6.3004E-04 6.2252E-04 6.0748E-04162533103 5.6306E-04 3.8830E-04 2.8840E-04 7.8421E-04 8.4368E-04 1.5155E-03162533103 1.5156E-03 1.5158E-03 1.6235E-02 9.6110E-04 6.8671E-04 5.8948E-04162533103 6.0468E-04 6.3316E-04 6.2636E-04 6.1117E-04 5.6685E-04 3.9017E-04162533103 2.9064E-04 7.9613E-04 8.5614E-04 1.5382E-03 1.5384E-03 1.5389E-03162533103 8.0479E-03 2.5416E-03 6.7123E-04 7.8051E-04 8.0128E-04 7.5961E-04162533103 7.2126E-04 6.5442E-04 4.3721E-04 3.8278E-04 5.5604E-04 5.8133E-04162533103 8.6206E-04 8.6210E-04 8.6223E-04 2.9649E-03 5.5056E-04 7.2580E-04162533103 7.6567E-04 7.1700E-04 6.7961E-04 6.3216E-04 4.3992E-04 3.8498E-04162533103 4.6967E-04 4.8689E-04 6.5527E-04 6.5528E-04 6.5537E-04 1.4913E-03162533103 6.4250E-04 6.5678E-04 6.2474E-04 5.9695E-04 5.2029E-04 3.8122E-04162533103 3.4843E-04 3.9503E-04 4.0583E-04 5.1030E-04 5.1045E-04 5.1050E-04162533103 1.4220E-03 7.3842E-04 6.8011E-04 7.0157E-04 6.0537E-04 4.1142E-04162533103 3.7075E-04 4.1240E-04 4.2421E-04 5.3492E-04 5.3508E-04 5.3511E-04162533103 1.5183E-03 9.1191E-04 6.9948E-04 6.0628E-04 4.2545E-04 3.8548E-04162533103 4.3714E-04 4.4949E-04 5.7461E-04 5.7469E-04 5.7485E-04 1.5774E-03162533103 7.0082E-04 6.9341E-04 4.1684E-04 3.6597E-04 4.2464E-04 4.3794E-04162533103 5.6898E-04 5.6912E-04 5.6922E-04 1.3354E-03 8.5551E-04 4.5574E-04162533103 3.6290E-04 4.0894E-04 4.2291E-04 5.4441E-04 5.4452E-04 5.4461E-04162533103 1.3514E-03 5.3505E-04 3.6596E-04 3.9729E-04 4.1346E-04 5.1956E-04162533103 5.1969E-04 5.1975E-04 1.6092E-03 3.8766E-04 3.3510E-04 3.4532E-04162533103 3.5854E-04 3.5862E-04 3.5868E-04 1.8445E-03-1.5721E-03-4.2629E-04162533103 2.6490E-04 2.6489E-04 2.6498E-04 1.3992E-02 5.9123E-03 7.7589E-04162533103 7.7619E-04 7.7627E-04 1.8117E-02 8.3496E-04 8.3499E-04 8.3513E-04162533103 1.8567E-02 1.3578E-02 1.3579E-02 2.0922E-02 1.3581E-02 2.3235E-02162533103 0.0000E+00 0.0000E+00 0 1 8 4162533103 1.0000E-05 0.0000E+00 9.2000E+05 1.9119E-01 9.5700E+05 0.0000E+00162533103 2.0000E+07 0.0000E+00 162533103 162533 0 1625 0 0 0 0 0 2.10450E+4 4.45700E+1 2 0 1 12126 1451 0.0 0.0 0 0 0 62126 1451 1.00000E+0 2.00000E+7 0 0 10 20022126 1451 3.00000E+2 0.0 1 0 104 32126 1451 21-Sc- 45 CNDC EVAL-FEB91 Z.X.ZHAO 2126 1451 DIST- 2126 1451 ----- DOSIMETRY FILE 2126 1451 ----- INCIDENT NEUTRON DATA 2126 1451 ----- ENDF-6 FORMAT 2126 1451 THIS EVALUATION IS MADE MAINLY BASED ON THE RESONANCE 2126 1451 PARAMETERS GIVEN BY REF.[1], MEASURED DATA OF REFS.[2-10] 2126 1451 AND THE SYSTEMATICS OF EXCITATION FUNCTION FOR (N,GAMMA) 2126 1451 REACTION OF REF.[11]. A SET OF RESOLVED RESONANCE PARAMETERS 2126 1451 ARE GIVEN FOR EN < 100 KEV. IN THE REGION OF EN > 100 KEV, 2126 1451 THE SMOOTH CROSS SECTIONS ARE GIVEN. 2126 1451 MF=2 MT=251 RESONANCE PARAMETER 2126 1451 POSITIVE ENERGY RESONANCE PARAMETERS LISTED BELOW ARE 2126 1451 BASED ON THOSE OF REF.[1]. TO FIT KENNY'S AVERAGE CROSS 2126 1451 SECTIONS [2] (CORRECTED IN REF.[3]), THE CAPTURE WIDTHS 2126 1451 HAVE BEEN ADJUSTED. THE NEGATIVE ENERGY LEVEL PARAMETERS 2126 1451 ARE TAKEN FROM JENDL-3 EVALUATION [12] TO REPRODUCE 2126 1451 THERMAL CROSS SECTION 27.2 B WHICH IS IN GOOD AGREEMENT 2126 1451 WITH 27.2 +- 0.2 B GIVEN BY REF.[1]. 2126 1451 2126 1451 MF=3 MT=102 SMOOTH CROSS SECTION 2126 1451 THE SMOOTH CROSS SECTION CALCULATED FROM ZHAO'S 2126 1451 SYSTEMATICS [11] ARE CONSISTENT WITH THE MEASURED DATA OF 2126 1451 REFS.[2-9] FROM 1 KEV TO 20 MEV. THEREFORE THE SYSTEMATICS 2126 1451 IS USED IN THIS EVALUATION TO GIVE THE SMOOTH CROSS 2126 1451 SECTIONS FOR EN > 100 KEV. 2126 1451 2126 1451 MF=8 MT=102 RADIOACTIVE DECAY DATA 2126 1451 HALF-TIME OF SC-46 IS TAKEN FROM REF.[13] 2126 1451 2126 1451 MF=32 AND 33 MT=102 COVARIANCE DATA 2126 1451 LONG-RANGE COMPONENT (GIVEN IN FILE 32): 2126 1451 10-5 EV TO 1 KEV----- THE ERROR OF THERMAL CROSS 2126 1451 SECTION. 2126 1451 1 KEV TO 100 KEV----- THE NORMALIZATION ERROR OF 2126 1451 KENNY'S MEASUREMENT. 2126 1451 100 KEV TO 20 MEV---- THE SHAPE ERROR OF THE 2126 1451 SYSTEMATICS. 2126 1451 SHORT-RANGE COMPONENT: 2126 1451 FOR EN < 100 KEV, THIS COMPONENT IS GIVEN IN FILE 32. 2126 1451 IN THIS FILE, ONLY THE ERRORS OF SOME LARGE S-WAVE 2126 1451 RESONANCE PARAMETERS ARE CONSIDERED. THE SHORT-RANGE 2126 1451 COMPONENT FOR EN > 100 KEV ARE GIVEN IN FILE 33 BASED ON 2126 1451 THE ERRORS OF PERKIN'S MEASUREMENT [6]. 2126 1451 2126 1451 REFERENCES 2126 1451 [1] S. F. MUGHABGHAB ET AL., BNL-325, 4TH, VOL.1 (1981) 2126 1451 [2] M. J. KENNY ET AL., AUST. J. PHYS., 30, 605 (1977) 2126 1451 [3] B. J. ALLEN ET AL., NUCL. SCI. AND ENG., 82, 230 (1982) 2126 1451 [4] S. A. ROMANOV ET AL., YF, 1, 229 (1965) 2126 1451 [5] R. BOOTH ET AL., PHYS. REV., 112, 226 (1958) 2126 1451 [6] J. L. PERKIN, J. NUCL. ENERGY, 17, 349 (1963) 2126 1451 [7] J. CSIKAI ET AL., NUCL. PHYS., A95, 229 (1967) 2126 1451 [8] M. WAGNER ET AL., APA, 52, 30 (1980) 2126 1451 [9] M. BUDNAR ET AL., INDC(YUG)-6 (1979) 2126 1451 [10] W. MANNHART, ZP/A, 272, 273 (1975) 2126 1451 [11] Z. X. ZHAO ET AL., CHINESE NUCL. PHYS.,VOL.11, 71 (1989)2126 1451 [12] T. NAKAGAWA ET AL., JAERI-M, 90-099 (1990) 2126 1451 [13] J. TULI, NUCLEAR WALLET CARDS, NNDC, 1990 2126 1451 2126 1451 ================================================================ 2126 1451 The Q values and threshold energies were updated prior to pro- 2126 1451 cessing through the codes to comply with the values obtained 2126 1451 using the NNDC calculation program which is based on the 1995 2126 1451 Update to the Atomic mass Evaluation. 2126 1451 2126 1451 Total and Elastic cross sections were only given upto 100 Kev 2126 1451 in the original data.For these two reactions the energy region 2126 1451 between 100 KeV and 20 Mev was taken the ENDF/B-6 library, 2126 1451 linearised and added to this file. 2126 1451 2126 1451 The original SC45G cross section set had uncertainties for the 2126 1451 resonance parameters in file MF=32, which were converted in the 2126 1451 extended SAND-II group structure and inserted in file MF=33 as 2126 1451 extra "NI type" sub-subsections. 2126 1451 2126 1451 ***************** Program LINEAR (VERSION 2002-1) ***************2126 1451 For All Data Greater than 1.0000E-10 barns in Absolute Value 2126 1451 Data Linearized to Within an Accuracy of .100000000 per-cent 2126 1451 ***************** Program RECENT (VERSION 2002-1) ***************2126 1451 for All Data Greater than 1.0000E-10 barns in Absolute Value 2126 1451 Data Linearized to within an Accuracy of .100000000 per-cent 2126 1451 ***************** Program SIGMA1 (VERSION 2002-1) ***************2126 1451 Data Doppler Broadened to 300.000000 Kelvin 2126 1451 for All Data Greater than 1.0000E-10 barns in Absolute Value 2126 1451 Data Linearized to Within an Accuracy pf .100000000 per-cent 2126 1451 ***************** Program FIXUP (Version 2002-1) ****************2126 1451 Corrected ZA/AWR in All Sections-----------------------------Yes 2126 1451 Corrected Thresholds-----------------------------------------Yes 2126 1451 Extended Cross Sections to 20 MeV----------------------------No 2126 1451 Allow Cross Section Deletion---------------------------------No 2126 1451 Allow Cross Section Reconstruction---------------------------No 2126 1451 Make All Cross Sections Non-Negative-------------------------Yes 2126 1451 Delete Energies Not in Ascending Order-----------------------Yes 2126 1451 Deleted Duplicate Points-------------------------------------Yes 2126 1451 Check for Ascending MAT/MF/MT Order--------------------------Yes 2126 1451 Check for Legal MF/MT Numbers--------------------------------Yes 2126 1451 Allow Creation of Missing Sections---------------------------No 2126 1451 Allow Insertion of Energy Points-----------------------------No 2126 1451 Create Uniform Energy Grid-----------------------------------No 2126 1451 Delete Section if Cross Section =0 at All Energies-----------Yes 2126 1451 ***************** Program GROUPIE (VERSION 2002-1) **************2126 1451 Unshielded Group Averages Using 640 Groups 2126 1451 Weighting Spectrum: Flat (Constant) Spectrum 2126 1451 1 451 111 12126 1451 3 102 217 12126 1451 33 102 150 12126 1451 2126 1 0 2126 0 0 2.10450E+4 4.45700E+1 0 99 0 02126 3102 8.76070E+6 8.76070E+6 0 0 1 6412126 3102 641 1 2126 3102 .000100000 427.732022 .000105000 417.757985 .000110000 408.2825332126 3102 .000115000 399.551148 .000120000 389.344069 .000127500 378.0196042126 3102 .000135000 367.672957 .000142500 358.164586 .000150000 347.8738902126 3102 .000160000 337.153670 .000170000 327.399593 .000180000 318.4460852126 3102 .000190000 310.115605 .000200000 302.509818 .000210000 295.3540522126 3102 .000220000 288.707485 .000230000 282.532179 .000240000 275.2837032126 3102 .000255000 267.318279 .000270000 261.171754 .000280000 254.3295382126 3102 .000300000 245.985271 .000320000 238.429839 .000340000 231.5102902126 3102 .000360000 225.136972 .000380000 219.318956 .000400000 213.2533622126 3102 .000425000 207.046482 .000450000 201.375468 .000475000 196.1761222126 3102 .000500000 191.286471 .000525000 186.817655 .000550000 182.5961802126 3102 .000575000 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3.1075E-07 1.2496E-06 1.0171E-05 2.4915E-07 8.2664E-07212633102 1.5264E-06 5.1139E-06 8.6721E-06 3.1180E-06 9.5912E-07 4.3529E-07212633102 2.4124E-07 1.5010E-07 3.9620E-08 8.1735E-08 2.1458E-07 7.8258E-07212633102 8.3717E-08 3.4105E-08 3.3464E-07 3.7590E-07 4.4039E-08 3.8738E-07212633102 1.6208E-07 4.7458E-08 5.3197E-08 5.9656E-08 6.6998E-08 7.5439E-08212633102 8.5275E-08 9.6920E-08 1.1097E-07 1.3387E-07 1.7052E-07 2.6962E-08212633102 3.5749E-07 1.4617E-06 1.1984E-05 2.8959E-07 9.3604E-07 1.7170E-06212633102 5.7460E-06 9.7464E-06 3.5065E-06 1.0796E-06 4.9026E-07 2.7177E-07212633102 1.6899E-07 4.4517E-08 9.1184E-08 2.3382E-07 8.4762E-07 9.1740E-08212633102 3.7552E-08 3.6276E-07 4.0100E-07 4.7260E-08 4.1299E-07 1.7306E-07212633102 5.9687E-08 6.7002E-08 7.5329E-08 8.4920E-08 9.6116E-08 1.0940E-07212633102 1.2547E-07 1.5173E-07 1.9402E-07 3.0802E-08 4.1184E-07 1.7139E-06212633102 1.4157E-05 3.3724E-07 1.0593E-06 1.9283E-06 6.4440E-06 1.0933E-05212633102 3.9357E-06 1.2128E-06 5.5113E-07 3.0558E-07 1.8992E-07 4.9927E-08212633102 1.0153E-07 2.5399E-07 9.1475E-07 1.0026E-07 4.1256E-08 3.9178E-07212633102 4.2612E-07 5.0545E-08 4.3852E-07 1.8406E-07 7.5294E-08 8.4750E-08212633102 9.5661E-08 1.0842E-07 1.2360E-07 1.4201E-07 1.7221E-07 2.2112E-07212633102 3.5259E-08 4.7573E-07 2.0171E-06 1.6790E-05 3.9400E-07 1.1992E-06212633102 2.1638E-06 7.2182E-06 1.2248E-05 4.4120E-06 1.3608E-06 6.1882E-07212633102 3.4320E-07 2.1320E-07 5.5934E-08 1.1291E-07 2.7521E-07 9.8413E-07212633102 1.0935E-07 4.5257E-08 4.2173E-07 4.5126E-07 5.3907E-08 4.6397E-07212633102 1.9510E-07 9.5514E-08 1.0796E-07 1.2255E-07 1.3994E-07 1.6110E-07212633102 1.9595E-07 2.5274E-07 4.0499E-08 5.5184E-07 2.3869E-06 2.0025E-05212633102 4.6255E-07 1.3596E-06 2.4280E-06 8.0814E-06 1.3713E-05 4.9431E-06212633102 1.5261E-06 6.9450E-07 3.8530E-07 2.3925E-07 6.2642E-08 1.2552E-07212633102 2.9764E-07 1.0560E-06 1.1910E-07 4.9607E-08 4.5274E-07 4.7649E-07212633102 5.7365E-08 4.8940E-07 2.0621E-07 1.2221E-07 1.3895E-07 1.5897E-07212633102 1.8340E-07 2.2383E-07 2.9017E-07 4.6746E-08 6.4391E-07 2.8449E-06212633102 2.4065E-05 5.4663E-07 1.5454E-06 2.7264E-06 9.0492E-06 1.5355E-05212633102 5.5386E-06 1.7117E-06 7.7963E-07 4.3269E-07 2.6858E-07 7.0184E-08212633102 1.3958E-07 3.2147E-07 1.1307E-06 1.2963E-07 5.4373E-08 4.8491E-07212633102 5.0188E-07 6.0943E-08 5.1483E-07 2.1742E-07 1.5829E-07 1.8148E-07212633102 2.0988E-07 2.5709E-07 3.3520E-07 5.4325E-08 7.5728E-07 3.4230E-06212633102 2.9203E-05 6.5165E-07 1.7636E-06 3.0663E-06 1.0141E-05 1.7206E-05212633102 6.2104E-06 1.9216E-06 8.7602E-07 4.8642E-07 3.0184E-07 7.8726E-08212633102 1.5538E-07 3.4692E-07 1.2085E-06 1.4108E-07 5.9642E-08 5.1839E-07212633102 5.2752E-07 6.4673E-08 5.4034E-07 2.2877E-07 2.0856E-07 2.4186E-07212633102 2.9754E-07 3.9044E-07 6.3705E-08 8.9987E-07 4.1684E-06 3.5878E-05212633102 7.8568E-07 2.0238E-06 3.4571E-06 1.1381E-05 1.9305E-05 6.9733E-06212633102 2.1604E-06 9.8596E-07 5.4781E-07 3.3988E-07 8.8484E-08 1.7330E-07212633102 3.7429E-07 1.2899E-06 1.5365E-07 6.5522E-08 5.5334E-07 5.5353E-07212633102 6.8596E-08 5.6597E-07 2.4030E-07 2.8137E-07 3.4784E-07 4.5981E-07212633102 7.5598E-08 1.0838E-06 5.1546E-06 4.4775E-05 9.6111E-07 2.3401E-06212633102 3.9112E-06 1.2800E-05 2.1702E-05 7.8454E-06 2.4343E-06 1.1124E-06212633102 6.1856E-07 3.8377E-07 9.9739E-08 1.9382E-07 4.0395E-07 1.3752E-06212633102 1.6756E-07 7.2161E-08 5.8994E-07 5.8000E-07 7.2761E-08 5.9177E-07212633102 2.5209E-07 4.3328E-07 5.7922E-07 9.6332E-08 1.4114E-06 6.9676E-06212633102 6.1270E-05 1.2794E-06 2.8602E-06 4.6076E-06 1.4919E-05 2.5270E-05212633102 9.1462E-06 2.8447E-06 1.3028E-06 7.2551E-07 4.5029E-07 1.1680E-07212633102 2.2470E-07 4.4550E-07 1.4893E-06 1.8766E-07 8.2002E-08 6.3885E-07212633102 6.1425E-07 7.8482E-08 6.2455E-07 2.6740E-07 7.8714E-07 1.3309E-07212633102 2.0096E-06 1.0413E-05 9.2939E-05 1.8739E-06 3.7044E-06 5.6069E-06212633102 1.7800E-05 3.0087E-05 1.0910E-05 3.4069E-06 1.5661E-06 8.7465E-07212633102 5.4357E-07 1.4078E-07 2.6761E-07 4.9614E-07 1.6146E-06 2.1370E-07212633102 9.5360E-08 6.9271E-07 6.4845E-07 8.5139E-08 6.5563E-07 2.8290E-07212633102 2.2866E-08 3.5517E-07 1.9201E-06 1.7342E-05 3.3943E-07 5.9449E-07212633102 8.3507E-07 2.5821E-06 4.3514E-06 1.5812E-06 4.9629E-07 2.2925E-07212633102 1.2853E-07 8.0069E-08 2.0730E-08 3.9037E-08 6.7924E-08 2.1488E-07212633102 2.9936E-08 1.3649E-08 9.2255E-08 8.4481E-08 1.1418E-08 8.4861E-08212633102 3.6933E-08 5.7820E-06 3.3369E-05 3.0616E-04 5.7305E-06 8.0793E-06212633102 9.4646E-06 2.7073E-05 4.5174E-05 1.6515E-05 5.2597E-06 2.4645E-06212633102 1.3982E-06 8.7856E-07 2.2830E-07 4.2504E-07 6.6183E-07 1.9753E-06212633102 3.0563E-07 1.4544E-07 8.5161E-07 7.5803E-07 1.0883E-07 7.4987E-07212633102 3.3273E-07 2.0963E-04 1.9379E-03 3.4155E-05 3.4355E-05 2.3371E-05212633102 4.3038E-05 6.6431E-05 2.5386E-05 8.9490E-06 4.5915E-06 2.7974E-06212633102 1.8542E-06 4.9895E-07 9.2670E-07 1.0662E-06 2.4945E-06 5.8495E-07212633102 3.1928E-07 1.1255E-06 9.7470E-07 1.7988E-07 8.8768E-07 4.3584E-07212633102 1.8310E-02 3.2576E-04 2.8938E-04 1.3803E-04 1.1212E-04 1.2320E-04212633102 5.7938E-05 2.8618E-05 1.8113E-05 1.2562E-05 9.0537E-06 2.5688E-06212633102 4.8481E-06 3.9722E-06 5.1392E-06 2.7835E-06 1.7502E-06 2.8066E-06212633102 2.5018E-06 7.5487E-07 1.7106E-06 1.1796E-06 6.0844E-06 6.9310E-06212633102 6.2869E-06 1.5456E-05 2.5212E-05 9.3414E-06 3.0736E-06 1.4842E-06212633102 8.6228E-07 5.5061E-07 1.4373E-07 2.5891E-07 2.7967E-07 6.2654E-07212633102 1.5322E-07 8.3156E-08 2.7751E-07 2.2733E-07 4.3656E-08 2.0351E-07212633102 1.0156E-07 2.1420E-05 3.9156E-05 1.3226E-04 2.2500E-04 8.1404E-05212633102 2.5305E-05 1.1547E-05 6.3848E-06 3.9080E-06 9.8334E-07 1.7063E-06212633102 1.7295E-06 3.7268E-06 9.3236E-07 4.9759E-07 1.6066E-06 1.2634E-06212633102 2.4505E-07 1.1182E-06 5.5889E-07 8.2376E-05 2.8835E-04 4.9153E-04212633102 1.7730E-04 5.4890E-05 2.4924E-05 1.3711E-05 8.3456E-06 2.0843E-06212633102 3.5510E-06 3.0519E-06 5.2665E-06 1.7699E-06 9.9046E-07 2.3172E-06212633102 1.8166E-06 4.2118E-07 1.4473E-06 8.0185E-07 1.0186E-03 1.7344E-03212633102 6.2471E-04 1.9346E-04 8.7819E-05 4.8271E-05 2.9334E-05 7.3006E-06212633102 1.2276E-05 8.9112E-06 1.0675E-05 5.6377E-06 3.3209E-06 5.0173E-06212633102 4.1306E-06 1.2302E-06 2.6350E-06 1.8117E-06 2.9922E-03 1.0875E-03212633102 3.3573E-04 1.5214E-04 8.3555E-05 5.0747E-05 1.2622E-05 2.1180E-05212633102 1.4928E-05 1.6361E-05 9.5985E-06 5.7041E-06 7.8220E-06 6.5105E-06212633102 2.0601E-06 3.8693E-06 2.8536E-06 3.9777E-04 1.2262E-04 5.5543E-05212633102 3.0517E-05 1.8559E-05 4.6303E-06 7.8604E-06 6.4671E-06 1.0325E-05212633102 3.8327E-06 2.1689E-06 4.3923E-06 3.1927E-06 8.5061E-07 2.3383E-06212633102 1.4256E-06 3.7889E-05 1.7207E-05 9.4876E-06 5.8048E-06 1.4671E-06212633102 2.6099E-06 3.3811E-06 9.0932E-06 1.6277E-06 7.8228E-07 3.3150E-06212633102 1.9151E-06 3.7767E-07 1.8012E-06 8.8737E-07 7.8447E-06 4.3534E-06212633102 2.6974E-06 7.0139E-07 1.3760E-06 3.0715E-06 1.0712E-05 1.2243E-06212633102 4.6912E-07 3.4699E-06 1.6492E-06 2.7246E-07 1.7667E-06 7.9571E-07212633102 2.4461E-06 1.5543E-06 4.2703E-07 9.8619E-07 3.5720E-06 1.3965E-05212633102 1.2612E-06 3.8683E-07 3.9869E-06 1.5685E-06 2.4285E-07 1.8345E-06212633102 7.9547E-07 1.0380E-06 3.1496E-07 9.1294E-07 4.7837E-06 1.9682E-05212633102 1.5729E-06 3.9925E-07 4.7579E-06 1.5139E-06 2.3795E-07 1.9312E-06212633102 8.2324E-07 1.1255E-07 4.2344E-07 2.8478E-06 1.1982E-05 8.9628E-07212633102 1.9429E-07 2.2608E-06 5.5343E-07 9.5443E-08 7.8940E-07 3.3449E-07212633102 2.0742E-06 1.6466E-05 6.9671E-05 4.9909E-06 9.2554E-07 7.9133E-06212633102 1.3090E-06 2.8949E-07 2.2393E-06 9.5624E-07 1.4433E-04 5.9024E-04212633102 4.0501E-05 6.5832E-06 1.5000E-05 2.4583E-06 7.9553E-07 3.1192E-06212633102 1.5155E-06 2.5686E-03 1.8228E-04 2.8441E-05 3.1009E-05 8.8389E-06212633102 2.6695E-06 5.5532E-06 3.2812E-06 1.3371E-05 2.5234E-06 4.0262E-05212633102 6.1624E-06 5.0010E-07 4.1452E-06 1.7801E-06 1.8179E-06 1.4569E-04212633102 5.2352E-05 5.4077E-08 4.7205E-06 1.9613E-06 1.7427E-02 9.9573E-03212633102 8.4587E-05 7.9004E-06 3.3483E-06 9.0737E-03 8.8884E-05 4.9176E-05212633102 2.2947E-05 4.7205E-06 4.3502E-05 1.7438E-05 4.8803E-04 1.9935E-04212633102 8.4828E-05 212633102 0.0 0.0 1 5 10 4212633102 1.1000E+04 1.1500E+04 1.2000E+04 1.2750E+04 3.0436E-05 1.0200E-04212633102 7.2626E-06 3.8928E-04 2.8587E-05 2.1192E-06 212633102 0.0 0.0 1 5 15 5212633102 2.3000E+04 2.4000E+04 2.5500E+04 2.7000E+04 2.8000E+04 2.0221E-03212633102 1.1416E-04 2.6389E-07 8.7044E-08 1.0575E-03 8.1002E-07 1.7944E-07212633102 6.8298E-04 3.0746E-05 9.1324E-05 212633102 0.0 0.0 0 1 12 6212633102 3.4000E+04 1.5819E-05 3.6000E+04 4.9680E-12 3.8000E+04 6.5496E-05212633102 4.0000E+04 4.8828E-05 4.2500E+04 1.6762E-05 4.5000E+04 0.0000E+00212633102 0.0 0.0 0 1 12 6212633102 6.9000E+04 1.6282E-05 7.2000E+04 2.4062E-09 7.6000E+04 2.6452E-08212633102 8.0000E+04 1.7513E-08 8.4000E+04 2.4018E-04 8.8000E+04 0.0000E+00212633102 212633 0 2126 0 0 0 0 0 2.20460E+4 4.55579E+1 0 0 34 102225 1451 0.0 0.0 0 0 0 62225 1451 1.00000E+0 2.00000E+7 0 0 10 20022225 1451 3.00000E+2 0.0 1 0 233 52225 1451 22-Ti- 46 FEI EVAL-Jan02 K.I.Zolotarev 2225 1451 DIST-Feb2004 2225 1451 ----IRDF-2002 MATERIAL 2225 2225 1451 -----INCIDENT NEUTRON DATA 2225 1451 ------ENDF-6 FORMAT 2225 1451 ***************************************************************** 2225 1451 22-TI- 46 FEI EVAL-Jan02 K.I.Zolotarev 2225 1451 DIST-Feb02 20020203 2225 1451 ----BROND-2 MATERIAL 2225 2225 1451 -----INCIDENT NEUTRON DATA 2225 1451 ------ENDF-6 FORMAT 2225 1451 ------Russian Reactor Dosimetry File RRDF-2002 2225 1451 ***************************************************************** 2225 1451 Author of evaluation: K.I.Zolotarev 2225 1451 ***************************************************************** 2225 1451 MF=3 2225 1451 MT= 16 -(n,2n) cross section 2225 1451 ------------------------------------- 2225 1451 Excitation function for the Ti-46(n,2n)Ti-45 reaction in the 2225 1451 energy region from threshold to 20 MeV was evaluated by means of 2225 1451 statistical analysis of experimental cross section data [1-18]. 2225 1451 All experimental data were renormalized to the new standards 2225 1451 for monitor reactions cross sections and decay data. In the measu-2225 1451 rements [5,14-16] uncertainties were increased until the data had 2225 1451 become mutually consistent. 2225 1451 The final procedure of evaluation the Ti-46(n,2n)Ti-45 reac- 2225 1451 tion excitation function has been carried out within the frame- 2225 1451 work of generalized least squares method. Statistical analysis of 2225 1451 input cross section data was carried out by means of PADE-2 code 2225 1451 [19]. Rational function was used as the model function [20]. 2225 1451 The evaluated Ti-46(n,2n)Ti-45 excitation function averaged 2225 1451 on U-235 thermal neutron fission spectrum [21] and Cf-252 sponta- 2225 1451 neous fission neutron spectrum [22] gives the following values : 2225 1451 ÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÂÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÂÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄ 2225 1451 TYPE OF SPECTRUM ³ ,mb (calc.) ³ , mb (measured) 2225 1451 ÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÅÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÅÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄ 2225 1451 U-235 neutron fission ³ 0.0044686 ³ 2225 1451 ÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÅÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÅÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄ 2225 1451 CF-252 spont. fission ³ 0.011982 ³ 0.093 +- 0.031 [23] 2225 1451 ÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÁÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÁÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄ 2225 1451 2225 1451 MT=103 -(n,p) cross section 2225 1451 ------------------------------------- 2225 1451 Excitation function for the Ti-46(n,p)Sc-46m+g reaction in 2225 1451 the energy region from threshold to 20 MeV was evaluated by means 2225 1451 of statistical analysis of experimental cross section data [6,17, 2225 1451 24-38] and data from GNASH calculation [39]. 2225 1451 Analyzed microscopic experimental data were renormalized to 2225 1451 the new recommended standards for monitor reaction cross sections 2225 1451 and decay data. All experiments performed with Ti-natural samples 2225 1451 were corrected for the contribution of the Ti-47(n,np+pn)Sc-46m+g 2225 1451 and Ti-47(n,d)Sc-46m+g reactions. Total cross section for these 2225 1451 reactions were taken from ref.[40]. Experimental data from ref. 2225 1451 [6,27] were used partially. Data of Bormann et al. [6] were taken 2225 1451 only for 14.1 MeV point. In the work of Lukic and Carroll [27] it 2225 1451 were used data measured relative Fe56(n,p)Mn56 and Al27(n,a)Na24 2225 1451 monitor reactions. These data were renormalized to the cross sec- 2225 1451 sections obtained in the ref.[28]. Data of Smith and Meadows [28] 2225 1451 measured with using D(d,n)He3 neutron source were renormalized 2225 1451 in the energy range 6.996-9.950 MeV to the preliminary evaluated 2225 1451 integral of cross section from new experimental data [34,37,38]. 2225 1451 Experimental cross section data [8] , [41-46] were rejected 2225 1451 due to their discrepancy with the main bulk of experimental data 2225 1451 [6,17,24-38] and data from theoretical model calculation. 2225 1451 Statistical analysis of input cross section data was carried 2225 1451 out by means of PADE-2 code [19]. Rational function was used as 2225 1451 the model function [20]. 2225 1451 U-235 thermal fission [21] and Cf-252 spontaneous fission 2225 1451 neutron spectra [22] averaged cross sections calculated from the 2225 1451 evaluated Ti-46(n,p)Sc-46m+g reaction excitation function are 2225 1451 the following: 2225 1451 ÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÂÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÂÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄ 2225 1451 TYPE OF SPECTRUM ³ ,mb (calc.) ³ , mb (measured) 2225 1451 ÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÅÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÅÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄ 2225 1451 U-235 neutron fission ³ 11.447 ³ 11.57 +- 0.37 [47] 2225 1451 ³ ³ 11.51 +- 0.40 [48] 2225 1451 ³ ³ 11.55 +- 0.20 [49] 2225 1451 ÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÅÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÅÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄ 2225 1451 CF-252 spont. fission ³ 13.818 ³ 14.20 +- 0.24 [50] 2225 1451 ³ ³ 14.07 +- 0.25 [51] 2225 1451 ÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÁÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÁÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄ 2225 1451 2225 1451 MF=33 2225 1451 MT= 16 -(n,2n) cross section cov. matrix 2225 1451 ---------------------------------------- 2225 1451 Uncertainties in the evaluated excitation function for the 2225 1451 reaction Ti-46(n,2n)Ti-45 are given in the form of relative cova- 2225 1451 riance matrix for the 17-neutron energy groups (LB=5). Covariance 2225 1451 matrix of uncertainties was calculated simultaneously with 2225 1451 recommended cross section data by means of PADE-2 code. 2225 1451 Eigenvalues of the 6-th digits relative covariance matrix 2225 1451 given in the 33-file are the following: 2225 1451 2225 1451 2.26418E-07 2.55017E-07 2.97541E-07 3.62667E-07 2225 1451 4.74369E-07 6.90723E-07 1.00928E-06 1.38635E-06 2225 1451 2.05902E-06 3.42154E-06 6.73918E-06 1.77794E-05 2225 1451 1.16721E-04 5.82882E-03 7.49965E-03 2.28734E-02 2225 1451 4.20028E-02 2225 1451 2225 1451 MT=103 -(n,p) cross section cov. matrix 2225 1451 ---------------------------------------- 2225 1451 Uncertainties in the evaluated excitation function for the 2225 1451 reaction Ti-46(n,p)Sc-45m+g are given in the form of relative co- 2225 1451 variance matrix for the 31-neutron energy groups (LB=5). Covari- 2225 1451 ance matrix of uncertainties was calculated simultaneously with 2225 1451 recommended cross section data by means of PADE-2 code. 2225 1451 Eigenvalues of the 6-th digits relative covariance matrix 2225 1451 given in the 33-file are the following: 2225 1451 2225 1451 3.10934E-07 3.16123E-07 3.24666E-07 3.37022E-07 2225 1451 3.51765E-07 3.72727E-07 3.91781E-07 4.23365E-07 2225 1451 4.47112E-07 4.91264E-07 5.29779E-07 6.15458E-07 2225 1451 6.75168E-07 7.93660E-07 9.79888E-07 1.29384E-06 2225 1451 1.98373E-06 3.65835E-06 2.36258E-04 5.76816E-04 2225 1451 7.66259E-04 1.00346E-03 1.10942E-03 1.43365E-03 2225 1451 2.89439E-03 3.32523E-03 4.20534E-03 7.47772E-03 2225 1451 1.06722E-02 1.67016E-02 7.87709E-02 2225 1451 2225 1451 References : 2225 1451 1. A.Poularikas, R.W.Fink Physical Review, v.115, p.989, 1959 2225 1451 2. R.J.Prestwood,B.P.Bayhurst Phys. Rev, v.121, p.1438,Mar.1961 2225 1451 3. L.A.Rayburn Phys. Rev., v.122, p.168, 1961 2225 1451 3. M.Cevolani, S.Petralia Nuovo Cimento, v.26, p.1328, Dec. 1962 2225 1451 5. J.E.Strain, W.J.Ross Report ORNL-3672, January 1965 2225 1451 6. M.Bormann et al. Nuclear Physics, v.63, p.438, March 1965 2225 1451 7. J.Csikai Progress Report EANDC-50S, v.2, p.102, July 1965 2225 1451 8. H.L.Pai Canadian J. of Physics, v.44, p.2337, 1966 2225 1451 9. G.N.Maslov et al. Yadernye Konstanty, No.9, p.50, 1972 2225 1451 10. J.Araminowicz, J.Dresler Report INR-1464, p.14, May 1973 2225 1451 11. A.Paulsen et al. Atomkernenergie , v.26, p.34, August 1975 2225 1451 12. S.M.Qaim, N.I.Molla Proc of 9th Symposium on Fusion Technolo-2225 1451 gy, Garmisch, Germany, 14-18 June 1976, p.589 2225 1451 13. R.A.Sigg J. Diss. Abst., v.B37, p.2237, November 1976 2225 1451 14. J.Csikai Proc. of Int. Conf. on Nuclear Data for Science and 2225 1451 Technology, Antwerp, 6-10 September 1982, p.414, Reidel Publ. 2225 1451 Company, 1983 2225 1451 15. N.T.Molla et al. Progress Report INDC(BAN)-002, p.1, Feb.1983 2225 1451 16. Zhou Muyao et al. Chinese J. of Nucl. Phys., v.9, p.34, 1987 2225 1451 17. Y.Ikeda et al. Report JAERI-1312, March 1988 2225 1451 18. P.M.Dighe e.a. Indian J. of Pure and Applied Physics, v.29, 2225 1451 p.665, October 1991 2225 1451 19. S.A.Badikov et.al. Preprint FEI-1686, Obninsk, 1985 2225 1451 20. S.Badikov, N.Rabotnov, K.Zolotarev Proc. of NEANSC Speciali- 2225 1451 st's Meeting on Evaluation and Processing of Covariance Data, 2225 1451 Oak Ridge, USA, 7-9 September 1992, OECD, Paris, 1993, p.105 2225 1451 21. L.W.Weston et al. Evaluated Neutron Data for U-235, ENDF/B-VI 2225 1451 Library, MAT=9228, MF=5, MT=18, eval. April 1989 2225 1451 22. W.Mannhart IAEA-TECDOC-410, p.158, 1987 2225 1451 23. J.Csikai, Z.Dezso Proc. of 4th All Union Conf. on Neutron 2225 1451 Physics, Kiev, 18-22 April 1977, v.3, p.32 2225 1451 24. H.Liskien, A.Paulsen Nucl. Phys., v.63, p.393, March 1965 2225 1451 25. V.N.Levkovskij et al. Yadernaja Fizika (Sov.), v.10, n.1, 2225 1451 p.44, July 1969 2225 1451 26. S.K.Ghorai et al. J. Nucl. Energ., v.25, p.319, August 1971 2225 1451 27. Y.Lukic, E.E.Carroll Nucl. Sci. Eng., v.43, p.233, 1971 2225 1451 28. D.L.Smith, J.W.Meadows Nucl. Sci. Eng., v.58, p.314, 1975 2225 1451 29. Lu Hamlin et al. Chinese J. of Atomic Energy Sci. Technology,2225 1451 v.9, no.2, p.113, May 1975 2225 1451 30. K.Kayashima et al. Prog. Report NEANDC(J)-61U, p.94, Sep.1979 2225 1451 31. M.Viennot et al. Proc of Int. Conference on Nuclear Data for 2225 1451 Science and Technology, Antwerp, 6-10 September 1982, p.406 2225 1451 32. N.I.Molla et al. Progress Report INDC(BAN)-003, Sept. 1986 2225 1451 33. Lu Hanlin et al. Report INDC(CPR)-16, IAEA, August 1989 2225 1451 34. N.I.Molla, S.M.Qaim, M.Uhl Phys. Rev. C, v.42, n.4, p.1540, 2225 1451 October 1990 2225 1451 35. M.Viennot et al. Nucl. Sci. Eng., v.108, p.289, July 1991 2225 1451 36. Yuan Junqian e.a. High Energy Physics and Nucl.Phys.(China), 2225 1451 v.16, n.1, p.57, January 1992 2225 1451 37. J.W.Meadows, D.L.Smith, L.R.Greenwood, R.C.Haight, Y.Ikeda, 2225 1451 C.Konno Annals of Nuclear Energy, v.23, p.877, July 1996 2225 1451 38. Lu Hanlin et al. Report INDC(CPR)-045, IAEA, October 1998 2225 1451 39. P.G.Young, E.D.Arthur A Preequilibrium Statistical Nuclear 2225 1451 Model Code for Calculation of Cross Section and Emission 2225 1451 Spectra. Report LA-6947, Los Alamos, 1977 ; 2225 1451 E.L.Trykiv, G.Ya.Tertychnyi Private communication, IPPE, 2225 1451 Obninsk, May 1999 2225 1451 40. K.I.Zolotrarev RRDF-2002, MAT=2222, eval. January 2002 2225 1451 41. D.L.Allan Nucl. Phys., v.24, p.274, April 1961 2225 1451 42. W.G.Cross, H.L.Pai Prog. Report EANDC(CAN)-16, p.1, Jan. 1963 2225 1451 43. S.M.Qaim, N.I.Molla Nucl. Phys., v.A283, p.269, June 1977 2225 1451 44. I.Ribansky, S.Gmuca J. Phys. G, v.9, p.1537, December 1983 2225 1451 45. R.Pepelnik et al. Progress Report, NEANDC(E)-262U,(5), p.32, 2225 1451 June 1985 2225 1451 46. I.Kimura, K.Kobayashi Nucl. Sci. Eng., v.106, p.332, 1990 2225 1451 47. O.Horibe et al. Proc. of Conf.:50 Years with Nuclear Fission, 2225 1451 Washington D.C., 25-28 April 1989, v.2, p.923 2225 1451 48. W.Mannhart Proc. 5-th ASTM-EUR. Symp. on Reactor Dosimetry, 2225 1451 Geesthacht, FRG, September 24-28, 1984, Vol.2, p.813, 1985 2225 1451 49. W.Mannhart Progress Report INDC(Ger)-045, pp.40-43, 1999 2225 1451 50. W.Mannhart Handbook on Nuclear Activation Data . IAEA Tech- 2225 1451 nical Report series No.273, p.413, 1987 2225 1451 51. W.Mannhart Validation of Differential Cross Sections with 2225 1451 Integral Data , Report INDC(NDS)-435, pp.59-64, IAEA, Vienna, 2225 1451 September 2002 2225 1451 ***************************************************************** 2225 1451 The Q values and threshold energies were updated prior to pro- 2225 1451 cessing through the codes to comply with the values obtained 2225 1451 using the NNDC calculation program which is based on the 1995 2225 1451 Update to the Atomic mass Evaluation. 2225 1451 2225 1451 File 2 added to the pointwise file containing only the effective 2225 1451 scattering radius with no resonance parameters given. 2225 1451 Taken from ENDF/B-VI 2225 1451 2225 1451 Psuedo Threshold of 0.0 added at 2.1E+6 ev to MF/MT=3/103 2225 1451 in original evaluation 2225 1451 2225 1451 ***************************************************************** 2225 1451 ***************** Program LINEAR (VERSION 2002-1) ***************2225 1451 For All Data Greater than 1.0000E-10 barns in Absolute Value 2225 1451 Data Linearized to Within an Accuracy of .100000000 per-cent 2225 1451 ***************** Program SIGMA1 (VERSION 2002-1) ***************2225 1451 Data Doppler Broadened to 300.000000 Kelvin 2225 1451 for All Data Greater than 1.0000E-10 barns in Absolute Value 2225 1451 Data Linearized to Within an Accuracy pf .100000000 per-cent 2225 1451 ***************** Program FIXUP (Version 2002-1) ****************2225 1451 Corrected ZA/AWR in All Sections-----------------------------Yes 2225 1451 Corrected Thresholds-----------------------------------------Yes 2225 1451 Extended Cross Sections to 20 MeV----------------------------No 2225 1451 Allow Cross Section Deletion---------------------------------No 2225 1451 Allow Cross Section Reconstruction---------------------------No 2225 1451 Make All Cross Sections Non-Negative-------------------------Yes 2225 1451 Delete Energies Not in Ascending Order-----------------------Yes 2225 1451 Deleted Duplicate Points-------------------------------------Yes 2225 1451 Check for Ascending MAT/MF/MT Order--------------------------Yes 2225 1451 Check for Legal MF/MT Numbers--------------------------------Yes 2225 1451 Allow Creation of Missing Sections---------------------------No 2225 1451 Allow Insertion of Energy Points-----------------------------No 2225 1451 Create Uniform Energy Grid-----------------------------------No 2225 1451 Delete Section if Cross Section =0 at All Energies-----------Yes 2225 1451 ***************** Program GROUPIE (VERSION 2002-1) **************2225 1451 Unshielded Group Averages Using 640 Groups 2225 1451 Weighting Spectrum: Flat (Constant) Spectrum 2225 1451 1 451 242 12225 1451 3 16 26 12225 1451 3 103 63 12225 1451 33 16 35 12225 1451 33 103 97 12225 1451 2225 1 0 2225 0 0 2.20460E+4 4.55579E+1 0 0 0 02225 3 16 -1.31898E+7-1.31898E+7 0 0 1 672225 3 16 67 1 2225 3 16 13400000.0 5.44722E-6 13500000.0 .000269054 13600000.0 .0013538942225 3 16 13700000.0 .003167185 13800000.0 .005539635 13900000.0 .0084628702225 3 16 14000000.0 .011885800 14100000.0 .015756100 14200000.0 .0200212002225 3 16 14300000.0 .024629400 14400000.0 .029530600 14500000.0 .0346768502225 3 16 14600000.0 .040022950 14700000.0 .045526750 14800000.0 .0511660252225 3 16 14900000.0 .056872075 15000000.0 .062608313 15100000.0 .0683747382225 3 16 15200000.0 .074141163 15300000.0 .079907588 15400000.0 .0855952332225 3 16 15500000.0 .091204100 15600000.0 .096812967 15700000.0 .1022853332225 3 16 15800000.0 .107621200 15900000.0 .112957067 16000000.0 .1181235002225 3 16 16100000.0 .123120500 16200000.0 .128117500 16300000.0 .1329310002225 3 16 16400000.0 .137561000 16500000.0 .142191000 16600000.0 .1466356672225 3 16 16700000.0 .150895000 16800000.0 .155154333 16900000.0 .1592341672225 3 16 17000000.0 .163134500 17100000.0 .167034833 17200000.0 .1707660002225 3 16 17300000.0 .174328000 17400000.0 .177890000 17500000.0 .1812951672225 3 16 17600000.0 .184543500 17700000.0 .187791833 17800000.0 .1908743752225 3 16 17900000.0 .193791125 18000000.0 .196707875 18100000.0 .1996246252225 3 16 18200000.0 .202373250 18300000.0 .204953750 18400000.0 .2075342502225 3 16 18500000.0 .210114750 18600000.0 .212532000 18700000.0 .2147860002225 3 16 18800000.0 .217040000 18900000.0 .219294000 19000000.0 .2215480002225 3 16 19100000.0 .223648400 19200000.0 .225595200 19300000.0 .2275420002225 3 16 19400000.0 .229488800 19500000.0 .231435600 19600000.0 .2332651252225 3 16 19700000.0 .234977375 19800000.0 .236689625 19900000.0 .2384018752225 3 16 20000000.0 0.0 2225 3 16 2225 3 0 2.20460E+4 4.55579E+1 0 0 0 02225 3103 -1.58430E+6-1.58430E+6 0 0 1 1802225 3103 180 1 2225 3103 2100000.00 3.42206E-6 2200000.00 1.02662E-5 2300000.00 1.71103E-52225 3103 2400000.00 2.39544E-5 2500000.00 3.98904E-5 2600000.00 9.93846E-52225 3103 2700000.00 .000237889 2800000.00 .000478634 2900000.00 .0008521572225 3103 3000000.00 .001398970 3100000.00 .002172540 3200000.00 .0032406352225 3103 3300000.00 .004678875 3400000.00 .006545195 3500000.00 .0088332052225 3103 3600000.00 .011449050 3700000.00 .014282900 3800000.00 .0173307502225 3103 3900000.00 .020716350 4000000.00 .024597800 4100000.00 .0290779002225 3103 4200000.00 .034170000 4300000.00 .039792900 4400000.00 .0457663252225 3103 4500000.00 .051856975 4600000.00 .057775050 4700000.00 .0632341002225 3103 4800000.00 .068060950 4900000.00 .072219050 5000000.00 .0757965502225 3103 5100000.00 .079116450 5200000.00 .082497900 5300000.00 .0861876502225 3103 5400000.00 .090482100 5500000.00 .095555300 5600000.00 .1014841502225 3103 5700000.00 .108255000 5800000.00 .115777000 5900000.00 .1239030002225 3103 6000000.00 .132481167 6100000.00 .141251500 6200000.00 .1500218332225 3103 6300000.00 .158677500 6400000.00 .167038500 6500000.00 .1749920002225 3103 6600000.00 .182459000 6700000.00 .189393500 6800000.00 .1957795002225 3103 6900000.00 .201623000 7000000.00 .206947500 7100000.00 .2116860002225 3103 7200000.00 .216090000 7300000.00 .220127250 7400000.00 .2237977502225 3103 7500000.00 .227143500 7600000.00 .230164500 7700000.00 .2331855002225 3103 7800000.00 .235943125 7900000.00 .238437375 8000000.00 .2409316252225 3103 8100000.00 .243425875 8200000.00 .245762125 8300000.00 .2479403752225 3103 8400000.00 .250118625 8500000.00 .252296875 8600000.00 .2544751252225 3103 8700000.00 .256653375 8800000.00 .258831625 8900000.00 .2610098752225 3103 9000000.00 .263188125 9100000.00 .265366375 9200000.00 .2675446252225 3103 9300000.00 .269722875 9400000.00 .271901125 9500000.00 .2740793752225 3103 9600000.00 .276257625 9700000.00 .278435875 9800000.00 .2804955832225 3103 9900000.00 .282436750 10000000.0 .284377917 10100000.0 .2863190832225 3103 10200000.0 .288260250 10300000.0 .290201417 10400000.0 .2919556252225 3103 10500000.0 .293522875 10600000.0 .295090125 10700000.0 .2966573752225 3103 10800000.0 .297992875 10900000.0 .299096625 11000000.0 .3002003752225 3103 11100000.0 .301304125 11200000.0 .302152667 11300000.0 .3027460002225 3103 11400000.0 .303339333 11500000.0 .303725000 11600000.0 .3039030002225 3103 11700000.0 .303857833 11800000.0 .303589500 11900000.0 .3033211672225 3103 12000000.0 .302775333 12100000.0 .301952000 12200000.0 .3011286672225 3103 12300000.0 .300027833 12400000.0 .298649500 12500000.0 .2972711672225 3103 12600000.0 .295628333 12700000.0 .293721000 12800000.0 .2918136672225 3103 12900000.0 .289664667 13000000.0 .287274000 13100000.0 .2848833332225 3103 13200000.0 .282253250 13300000.0 .279383750 13400000.0 .2765142502225 3103 13500000.0 .273644750 13600000.0 .270544600 13700000.0 .2672138002225 3103 13800000.0 .263883000 13900000.0 .260552200 14000000.0 .2572214002225 3103 14100000.0 .253720375 14200000.0 .250049125 14300000.0 .2463778752225 3103 14400000.0 .242706625 14500000.0 .239035375 14600000.0 .2353641252225 3103 14700000.0 .231692875 14800000.0 .228021625 14900000.0 .2243471882225 3103 15000000.0 .220669563 15100000.0 .216991938 15200000.0 .2133143132225 3103 15300000.0 .209636688 15400000.0 .205959063 15500000.0 .2022814382225 3103 15600000.0 .198603813 15700000.0 .195037600 15800000.0 .1915828002225 3103 15900000.0 .188128000 16000000.0 .184673200 16100000.0 .1812184002225 3103 16200000.0 .177889500 16300000.0 .174686500 16400000.0 .1714835002225 3103 16500000.0 .168280500 16600000.0 .165077500 16700000.0 .1619997502225 3103 16800000.0 .159047250 16900000.0 .156094750 17000000.0 .1531422502225 3103 17100000.0 .150303750 17200000.0 .147579250 17300000.0 .1448547502225 3103 17400000.0 .142130250 17500000.0 .139517250 17600000.0 .1370157502225 3103 17700000.0 .134514250 17800000.0 .132012750 17900000.0 .1296180002225 3103 18000000.0 .127330000 18100000.0 .125042000 18200000.0 .1227540002225 3103 18300000.0 .120566000 18400000.0 .118478000 18500000.0 .1163900002225 3103 18600000.0 .114302000 18700000.0 .112306625 18800000.0 .1104038752225 3103 18900000.0 .108501125 19000000.0 .106598375 19100000.0 .1047807882225 3103 19200000.0 .103048363 19300000.0 .101315938 19400000.0 .0995835132225 3103 19500000.0 .097928538 19600000.0 .096351013 19700000.0 .0947734882225 3103 19800000.0 .093195963 19900000.0 .091663500 20000000.0 0.0 2225 3103 2225 3 0 2225 0 0 2.20460E+4 4.55579E+1 0 0 0 1222533 16 0.000000+0 0.000000+0 0 16 0 1222533 16 0.000000+0 0.000000+0 1 5 190 19222533 16 1.000000-5 1.340000+7 1.400000+7 1.425000+7 1.450000+7 1.475000+7222533 16 1.500000+7 1.525000+7 1.550000+7 1.575000+7 1.600000+7 1.650000+7222533 16 1.700000+7 1.750000+7 1.800000+7 1.850000+7 1.900000+7 1.950000+7222533 16 2.000000+7 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0222533 16 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0222533 16 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0222533 16 0.000000+0 2.781420-2 1.283920-2 7.150180-3 4.197760-3 2.624760-3222533 16 1.798060-3 1.390020-3 1.219080-3 1.180400-3 1.246260-3 1.393920-3222533 16 1.529020-3 1.618830-3 1.660180-3 1.659830-3 1.626720-3 1.569240-3222533 16 8.069150-3 5.546650-3 3.876040-3 2.695150-3 1.848660-3 1.244700-3222533 16 8.222770-4 5.378570-4 3.076620-4 2.302510-4 3.187630-4 5.002420-4222533 16 7.302930-4 9.821299-4 1.239760-3 1.493850-3 4.441590-3 3.461700-3222533 16 2.659900-3 2.009820-3 1.496660-3 1.103050-3 8.113270-4 5.333680-4222533 16 3.719090-4 3.756580-4 4.829110-4 6.520230-4 8.556370-4 1.076080-3222533 16 1.302130-3 2.992840-3 2.525270-3 2.115430-3 1.766930-3 1.480710-3222533 16 1.252600-3 1.007340-3 8.189470-4 7.457220-4 7.499250-4 8.043140-4222533 16 8.900250-4 9.942610-4 1.108480-3 2.356520-3 2.171160-3 1.996940-3222533 16 1.836860-3 1.693860-3 1.511730-3 1.326660-3 1.194050-3 1.101810-3222533 16 1.039690-3 9.996680-4 9.757380-4 9.634650-4 2.189000-3 2.170040-3222533 16 2.129830-3 2.072140-3 1.961190-3 1.796150-3 1.623460-3 1.453990-3222533 16 1.293310-3 1.143860-3 1.006340-3 8.805220-4 2.287180-3 2.348570-3222533 16 2.367250-3 2.325030-3 2.189140-3 1.994320-3 1.770100-3 1.535250-3222533 16 1.301110-3 1.074200-3 8.580140-4 2.496620-3 2.577930-3 2.597300-3222533 16 2.495160-3 2.293670-3 2.036390-3 1.752220-3 1.459490-3 1.169360-3222533 16 8.883550-4 2.712310-3 2.783360-3 2.716800-3 2.521360-3 2.250200-3222533 16 1.939420-3 1.612380-3 1.283690-3 9.621380-4 2.917810-3 2.904460-3222533 16 2.737660-3 2.476860-3 2.163820-3 1.826070-3 1.481200-3 1.140080-3222533 16 2.958400-3 2.849680-3 2.639550-3 2.369860-3 2.069060-3 1.755710-3222533 16 1.441570-3 2.814730-3 2.683000-3 2.492950-3 2.269720-3 2.030300-3222533 16 1.785710-3 2.645740-3 2.555400-3 2.433170-3 2.292510-3 2.142650-3222533 16 2.575470-3 2.565690-3 2.536570-3 2.494660-3 2.673410-3 2.760570-3222533 16 2.831970-3 2.964830-3 3.149770-3 3.446400-3 222533 16 222533 0 2.20460E+4 4.55579E+1 0 0 0 1222533103 0.000000+0 0.000000+0 0 103 0 1222533103 0.000000+0 0.000000+0 1 5 561 33222533103 1.000000-5 2.100000+6 3.500000+6 4.000000+6 4.500000+6 5.000000+6222533103 5.500000+6 6.000000+6 6.500000+6 7.000000+6 7.500000+6 8.000000+6222533103 8.500000+6 9.000000+6 9.500000+6 1.000000+7 1.050000+7 1.100000+7222533103 1.150000+7 1.200000+7 1.250000+7 1.300000+7 1.350000+7 1.400000+7222533103 1.450000+7 1.500000+7 1.550000+7 1.600000+7 1.650000+7 1.700000+7222533103 1.800000+7 1.900000+7 2.000000+7 0.000000+0 0.000000+0 0.000000+0222533103 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0222533103 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0222533103 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0222533103 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0222533103 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0 7.547140-2222533103 7.738710-3 5.300980-3 3.343090-3 3.229410-3 3.191540-3 3.980150-3222533103 -1.086430-3-2.650150-3-9.973090-4 7.536140-4 1.330980-3 8.231030-4222533103 -2.400220-4-1.329500-3-2.073460-3-2.296170-3-2.007770-3-1.361480-3222533103 -5.881670-4 7.774100-5 4.595800-4 4.707900-4 1.159630-4-5.310000-4222533103 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1.404600-5 1.966510-5 2.484750-5 2.761730-5222533103 2.707990-5 2.340320-5 1.752840-5 1.080900-5 4.690890-6 4.965480-7222533103 5.774100-7 1.464440-5 4.788230-5 1.893610-3 1.330940-3 8.157900-4222533103 5.722070-4 3.490700-4 1.849240-4 9.739180-5 6.438320-5 6.008410-5222533103 6.643730-5 7.300830-5 7.479840-5 7.044050-5 6.093780-5 4.864880-5222533103 3.639840-5 2.670740-5 2.127720-5 2.080510-5 2.508940-5 3.331480-5222533103 4.436620-5 5.708370-5 7.042430-5 8.937890-5 1.111760-4 1.263500-4222533103 2.282840-3 1.279210-3 2.857520-4 6.556030-5 1.348380-4 1.968960-4222533103 1.935680-4 1.458640-4 8.401970-5 2.974710-5-5.509160-6-1.880930-5222533103 -1.345430-5 3.472290-6 2.362380-5 3.968860-5 4.694420-5 4.378170-5222533103 3.132350-5 1.249240-5-8.982860-6-2.936480-5-4.539530-5-5.468300-5222533103 -3.563130-5 2.603940-5 2.847560-3 2.411660-3 1.176110-3 3.278340-4222533103 -1.829580-5-6.311860-5 1.867090-5 1.249600-4 2.059570-4 2.434140-4222533103 2.376330-4 1.995920-4 1.452970-4 9.116720-5 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-8.970580-5-1.398910-4-1.786300-4-2.044440-4-1.784490-4-7.872330-5222533103 1.445120-3 1.083580-3 6.775910-4 3.228050-4 6.477870-5-8.405100-5222533103 -1.334110-4-1.076300-4-3.862740-5 4.120460-5 1.054510-4 1.378040-4222533103 1.330070-4 9.502850-5 3.371470-5-3.860100-5-1.097120-4-1.690850-4222533103 -2.150220-4-1.835830-4-2.528830-5 9.449480-4 7.149460-4 4.759940-4222533103 2.760980-4 1.382640-4 6.658500-5 5.147420-5 7.484400-5 1.155630-4222533103 1.546680-4 1.791070-4 1.831920-4 1.679270-4 1.389590-4 1.042630-4222533103 7.221500-5 5.034500-5 5.195700-5 1.295790-4 3.101290-4 6.811540-4222533103 6.084510-4 5.221670-4 4.374450-4 3.632780-4 3.038860-4 2.601880-4222533103 2.310620-4 2.144760-4 2.083910-4 2.113000-4 2.224300-4 2.416450-4222533103 2.692310-4 3.056580-4 3.514030-4 4.379840-4 5.881100-4 7.771760-4222533103 6.991220-4 7.367750-4 7.216420-4 6.605220-4 5.672850-4 4.601860-4222533103 3.585400-4 2.790880-4 2.332240-4 2.259000-4 2.562190-4 3.191390-4222533103 4.074230-4 5.132960-4 6.295330-4 8.071950-4 1.037690-3 1.238670-3222533103 8.790310-4 9.295870-4 8.911490-4 7.816890-4 6.311210-4 4.740970-4222533103 3.424850-4 2.593850-4 2.363030-4 2.736960-4 3.638140-4 4.942810-4222533103 6.512370-4 8.215070-4 1.072590-3 1.374030-3 1.596580-3 1.030230-3222533103 1.017920-3 9.115610-4 7.451500-4 5.608290-4 3.988850-4 2.897880-4222533103 2.500630-4 2.824190-4 3.789560-4 5.254750-4 7.054170-4 9.026670-4222533103 1.194860-3 1.543810-3 1.794240-3 1.031340-3 9.444440-4 7.900760-4222533103 6.093440-4 4.433830-4 3.246260-4 2.720760-4 2.908640-4 3.750690-4222533103 5.118720-4 6.855040-4 8.801580-4 1.175590-3 1.539350-3 1.814440-3222533103 8.888590-4 7.683740-4 6.179780-4 4.727510-4 3.619820-4 3.042870-4222533103 3.066400-4 3.663150-4 4.741640-4 6.179180-4 7.847390-4 1.048310-3222533103 1.391040-3 1.675830-3 6.951350-4 5.922860-4 4.861340-4 3.987540-4222533103 3.458160-4 3.349860-4 3.668070-4 4.367340-4 5.373920-4 6.604250-4222533103 8.666360-4 1.155910-3 1.425790-3 5.437350-4 4.853040-4 4.317030-4222533103 3.935570-4 3.777810-4 3.872170-4 4.214350-4 4.778520-4 5.527710-4222533103 6.898540-4 9.026440-4 1.128670-3 4.746400-4 4.584020-4 4.431900-4222533103 4.334040-4 4.325120-4 4.425760-4 4.644360-4 4.980230-4 5.687730-4222533103 6.952440-4 8.510190-4 4.782020-4 4.903100-4 4.974660-4 5.014420-4222533103 5.045490-4 5.088950-4 5.162030-4 5.356930-4 5.801400-4 6.480210-4222533103 5.318150-4 5.641130-4 5.877260-4 6.024600-4 6.093140-4 6.096570-4222533103 6.009780-4 5.800660-4 5.550780-4 6.279110-4 6.821970-4 7.248890-4222533103 7.532320-4 7.662530-4 7.557490-4 6.950360-4 5.859180-4 7.756650-4222533103 8.574140-4 9.223740-4 9.650720-4 9.782880-4 9.082420-4 7.360870-4222533103 9.863660-4 1.097040-3 1.182050-3 1.240980-3 1.193460-3 9.889720-4222533103 1.260120-3 1.394240-3 1.515950-3 1.521600-3 1.321900-3 1.583350-3222533103 1.778520-3 1.865200-3 1.710910-3 2.097680-3 2.348220-3 2.342350-3222533103 2.885560-3 3.210870-3 4.018350-3 222533103 222533 0 2225 0 0 0 0 0 2.20470E+4 4.65484E+1 0 0 34 102228 1451 0.0 0.0 0 0 0 62228 1451 1.00000E+0 2.00000E+7 0 0 10 20022228 1451 3.00000E+2 0.0 1 0 163 62228 1451 22-Ti- 47 FEI/ANL EVAL-Oct95 K.I.Zolotarev, C.Philis et al. 2228 1451 DIST-Feb2004 2228 1451 ----IRDF-2002 MATERIAL 2228 2228 1451 -----INCIDENT NEUTRON DATA 2228 1451 ------ENDF-6 FORMAT 2228 1451 ***************************************************************** 2228 1451 22-TI-47 FEI EVAL-Oct95 K.I.Zolotarev 2228 1451 DIST-Feb02 Mod1 20020203 2228 1451 ----BROND-2 MATERIAL 2228 REVISION 1 2228 1451 -----INCIDENT NEUTRON DATA 2228 1451 ------ENDF-6 FORMAT 2228 1451 ------Russian Reactor Dosimetry File RRDF-2002 2228 1451 ***************************************************************** 2228 1451 22-TI- 47 ANL EVAL-JAN77 C.PHILIS,O.BERSILLON,D.SMITH,ETC2228 1451 DIST-FEB91 910201 2228 1451 ----IRDF-90 MATERIAL 2228 2228 1451 ENDF/B-VI MATERIAL 2228 2228 1451 ENDF/B-V MATERIAL CONVERTED TO ENDF-6 FORMAT BY NNDC 2228 1451 ***************************************************************** 2228 1451 *****************EXTRACT FOR SPECIAL PURPOSE FILE*****************2228 1451 DOSIMETRY 2228 1451 ***************************************************************** 2228 1451 ******** Start of (N,NP) bibliographical component ******* 2228 1451 ***************************************************************** 2228 1451 Authors of evaluation: K.I.Zolotarev and A.B.Pashchenko 2228 1451 2228 1451 For this special purpose library it was decided the reaction 2228 1451 MF=3 MT=28+104 would use MF=10, MT=5 to present this data. 2228 1451 This was done after processing through the codes. The 2228 1451 co-variance file was converted from MF/MT=33/28 to MF/MT=40/5 2228 1451 ******************************************************************2228 1451 MF=3 2228 1451 MT= 28 -(n,np+pn+d) cross section data 2228 1451 -------------------------------------- 2228 1451 In this section is given the sum of cross section of the reac-2228 1451 tions Ti47(n,np)Sc46m+g , Ti47(n,pn)Sc46m+g and Ti47(n,d)Sc46m+g. 2228 1451 Excitation function for the Ti47(n,x)Sc46m+g reaction in the 2228 1451 energy region from threshold to 20 MeV was evaluated by means of 2228 1451 statistical analysis of experimental cross section data [1-6] and 2228 1451 data from STAPRE [7] calculation. 2228 1451 All experimental data were renormalized to the new standards 2228 1451 for monitor reactions cross sections and decay data. 2228 1451 The final procedure of evaluation Ti47(n,x)Sc46m+g excitation 2228 1451 function from threshold to 20 MeV has been carried out within the 2228 1451 framework of generalized least squares method. Rational function 2228 1451 was used as model function [8]. Calculations was performed by 2228 1451 means of Pade-2 code [9]. 2228 1451 U-235 thermal fission [10] and Cf-252 spontaneous fission 2228 1451 neutron spectra [11] averaged cross-sections calculated from the 2228 1451 evaluated Ti47(n,x)Sc46m+g excitation function are the following: 2228 1451 2228 1451 -------------------------------------------- 2228 1451 TYPE OF SPECTRUM I , mb (calc.) 2228 1451 --------------------------I----------------- 2228 1451 U-235 neutron fission I 8.1158E-3 2228 1451 CF-252 spontan. fission I 1.9201E-2 2228 1451 2228 1451 MF=33 2228 1451 MT= 28 -(n,np+pn+d) cross section cov. matrix 2228 1451 --------------------------------------------- 2228 1451 Uncertainties in the evaluated excitation function for the 2228 1451 reaction Ti-47(n,x)Sc-46m+g are given in the form of relative 2228 1451 covariance matrix for the 17-neutron energy groups (LB=5). Cova- 2228 1451 riance matrix of uncertainties was calculated simultaneously with 2228 1451 recommended cross section data by means of PADE-2 code. 2228 1451 Eigenvalues of the 6-th digits relative covariance matrix 2228 1451 given in the 33-file are the following: 2228 1451 2228 1451 3.91044E-07 4.46920E-07 5.37497E-07 7.02044E-07 2228 1451 9.85160E-07 1.54860E-06 2.86775E-06 6.97782E-06 2228 1451 2.67223E-05 1.90995E-04 9.14779E-04 5.12707E-03 2228 1451 8.27566E-03 1.15029E-02 1.68918E-02 5.65241E-02 2228 1451 8.27240E-01 2228 1451 2228 1451 References : 2228 1451 1. W.G.Cross, H.L.Pai Progress Rep. EANDC(CAN)-16, p.1, Jan.1963 2228 1451 2. H.L.Pai Canadian J. of Physics, v.44, p.2337, 1966 2228 1451 3. I.Ribansky, S.Gmuca J. Phys.G, v.9, p.1537, December 1983 2228 1451 4. N.I.Molla et al. Report INDC(BAN)-003, September 1986 2228 1451 5. Y.Ikeda et al. Report JAERI-1312, March 1988 2228 1451 6. Y.Uno et al. Report JAERI-M-93-046, p.247-256, 1993 2228 1451 7. M.Uhl, B.Strohmaier Computer Code STAPRE for Particle Induced 2228 1451 Activation Cross Section and Related Quantities, Report 2228 1451 IRK 76-01, Vienna, 1976 2228 1451 8. S.Badikov, N.Rabotnov, K.Zolotarev Proc. of NEANSC Speciali- 2228 1451 st's Meeting on Evaluation and Processing of Covariance Data, 2228 1451 Oak Ridge, USA, 7-9 September 1992, OECD, Paris, 1993, p.105 2228 1451 9. S.A.Badikov et al. Preprint FEI-1686, Obninsk, 1985 2228 1451 10. L.W.Weston et al. Evaluated Neutron Data for U-235, ENDF/B-VI 2228 1451 Library, MAT=9228, MF=5, MT=18, eval. April 1989 2228 1451 11. W.Mannhart IAEA-TECDOC-410, p.158, IAEA, Vienna, 1987 2228 1451 ******************************************************************2228 1451 ******** End of (N,NP) bibliographical component ******* 2228 1451 ***************************************************************** 2228 1451 ***************************************************************** 2228 1451 ******** Start of (N,P) bibliographical component ******* 2228 1451 ***************************************************************** 2228 1451 ** 2228 1451 DOCUMENTATION ANL/NDM-27 (N,P) 2228 1451 ** 2228 1451 COMMENTS- (N,P) EVAL IN ENERGY RANGES 1.10E+06 TO 1.0E+07 AND 2228 1451 1.35E+07 TO 2.0E+07 BASED ON EXPERIMENTAL DATA.INTERPOLATION AND2228 1451 EXTRAPOLATION TO OTHER ENERGIES BASED ON THEORETICAL CALCULATION2228 1451 PERFORMED AT BRUYERES LE CHATEL USING CODE DEVELOPED BY JARRE. 2228 1451 PSEUDO-THRESHOLD=0.5 MEV PER P.C. MACFARLANE TO MAGURNO 6/20/83 2228 1451 ** 2228 1451 REFERENCES 2228 1451 AL61 ALLAN D. NP24,274,61 2228 1451 AR78 ARTHUR E. PC 1978 2228 1451 AR75 ARMITAGE P.C. NNDC 2228 1451 CR62 CROSS W.AND PAI H. EANDC(CAN)-16,2,62 2228 1451 FU79 C.Y.FU PRIVATE COMM 1979 2228 1451 GH71 GHORAI S.AND CARROLL E. NSE 43,233,71 2228 1451 GO62 GONZALES ET.AL. PR126,271,62 2228 1451 HI HILLMAN M. NP37,78,62 2228 1451 LE69 LEVKOVSKII V. Y.F.10,44,69 2228 1451 PA66 PAI H.CJP 44 2337,66 2228 1451 PO59 POULARIKIS A. AND FINK R. PR115,989,59 2228 1451 SM75 SMITH D.L.AND MEADOWS J. ANL/NDM10,75 2228 1451 TI68 TIKKU ET.AL. SYMP ON NUCL CROSS SECT AND TECH WASH.1968 2228 1451 ******************************************************************2228 1451 ******** End of (N,P) bibliographical component ******* 2228 1451 ***************************************************************** 2228 1451 The Q values and threshold energies were updated prior to pro- 2228 1451 cessing through the codes to comply with the values obtained 2228 1451 using the NNDC calculation program which is based on the 1995 2228 1451 Update to the Atomic mass Evaluation. 2228 1451 2228 1451 File 2 added to the pointwise file containing only the effective 2228 1451 scattering radius with no resonance parameters given. 2228 1451 Taken from ENDF/B-VI 2228 1451 2228 1451 Threshold of 0.0 added at 1.0E-5 ev to MF/MT=3/103 after 2228 1451 processing thro. the pre-procssing codes. 2228 1451 2228 1451 2228 1451 ***************************************************************** 2228 1451 ***************** Program LINEAR (VERSION 2002-1) ***************2228 1451 For All Data Greater than 1.0000E-10 barns in Absolute Value 2228 1451 Data Linearized to Within an Accuracy of .100000000 per-cent 2228 1451 ***************** Program SIGMA1 (VERSION 2002-1) ***************2228 1451 Data Doppler Broadened to 300.000000 Kelvin 2228 1451 for All Data Greater than 1.0000E-10 barns in Absolute Value 2228 1451 Data Linearized to Within an Accuracy pf .100000000 per-cent 2228 1451 ***************** Program FIXUP (Version 2002-1) ****************2228 1451 Corrected ZA/AWR in All Sections-----------------------------Yes 2228 1451 Corrected Thresholds-----------------------------------------Yes 2228 1451 Extended Cross Sections to 20 MeV----------------------------No 2228 1451 Allow Cross Section Deletion---------------------------------No 2228 1451 Allow Cross Section Reconstruction---------------------------No 2228 1451 Make All Cross Sections Non-Negative-------------------------Yes 2228 1451 Delete Energies Not in Ascending Order-----------------------Yes 2228 1451 Deleted Duplicate Points-------------------------------------Yes 2228 1451 Check for Ascending MAT/MF/MT Order--------------------------Yes 2228 1451 Check for Legal MF/MT Numbers--------------------------------Yes 2228 1451 Allow Creation of Missing Sections---------------------------No 2228 1451 Allow Insertion of Energy Points-----------------------------No 2228 1451 Create Uniform Energy Grid-----------------------------------No 2228 1451 Delete Section if Cross Section =0 at All Energies-----------Yes 2228 1451 ***************************************************************** 2228 1451 ***************** Program GROUPIE (VERSION 2002-1) **************2228 1451 Unshielded Group Averages Using 640 Groups 2228 1451 Weighting Spectrum: Flat (Constant) Spectrum 2228 1451 1 451 173 12228 1451 3 103 70 02228 1451 8 5 2 12228 1451 10 5 42 12228 1451 33 103 115 12228 1451 40 5 36 02228 1451 2228 1 0 2228 0 0 2.20470E+4 4.65484E+1 0 0 0 02228 3103 1.82300E+5 1.82300E+5 0 0 1 1992228 3103 199 1 2228 3103 690000.000 5.16780E-5 720000.000 .000310068 760000.000 .0006201362228 3103 800000.000 .000805043 840000.000 .000864789 880000.000 .0009245352228 3103 920000.000 .000984281 960000.000 .001044027 1000000.00 .0013463902228 3103 1100000.00 .001891370 1200000.00 .002436350 1300000.00 .0029813302228 3103 1400000.00 .003526310 1500000.00 .004697020 1600000.00 .0064934602228 3103 1700000.00 .008289900 1800000.00 .010086340 1900000.00 .0118827802228 3103 2000000.00 .013905700 2100000.00 .016155100 2200000.00 .0184045002228 3103 2300000.00 .020653900 2400000.00 .022903300 2500000.00 .0250525002228 3103 2600000.00 .027101500 2700000.00 .029097500 2800000.00 .0310405002228 3103 2900000.00 .032983500 3000000.00 .034727500 3100000.00 .0362725002228 3103 3200000.00 .037817500 3300000.00 .039362500 3400000.00 .0409075002228 3103 3500000.00 .042653900 3600000.00 .044601700 3700000.00 .0465495002228 3103 3800000.00 .048497300 3900000.00 .050445100 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.129450000 12000000.0 .128879750 12100000.0 .1281192502228 3103 12200000.0 .127358750 12300000.0 .126598250 12400000.0 .1258377502228 3103 12500000.0 .125077250 12600000.0 .124316750 12700000.0 .1235562502228 3103 12800000.0 .122795750 12900000.0 .122035250 13000000.0 .1212747502228 3103 13100000.0 .120514250 13200000.0 .119753750 13300000.0 .1189932502228 3103 13400000.0 .118232750 13500000.0 .117472250 13600000.0 .1167117502228 3103 13700000.0 .115951250 13800000.0 .115190750 13900000.0 .1144302502228 3103 14000000.0 .113575000 14100000.0 .112625000 14200000.0 .1116750002228 3103 14300000.0 .110725000 14400000.0 .109775000 14500000.0 .1087300002228 3103 14600000.0 .107590000 14700000.0 .106450000 14800000.0 .1053100002228 3103 14900000.0 .104170000 15000000.0 .103124400 15100000.0 .1021732002228 3103 15200000.0 .101222000 15300000.0 .100270800 15400000.0 .0993196002228 3103 15500000.0 .098273800 15600000.0 .097133400 15700000.0 .0959930002228 3103 15800000.0 .094852600 15900000.0 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4.0000E+06-1.0000E+00222833103 4.5000E+06-1.0000E+00 5.0000E+06-1.0000E+00 5.5000E+06-1.0000E+00222833103 6.0000E+06-1.0000E+00 6.5000E+06-5.6000E-01 8.0000E+06-1.0000E+00222833103 8.5000E+06-3.5000E-01 1.0000E+07-2.4000E-01 1.2000E+07-2.7000E-01222833103 1.4000E+07-2.6000E-01 1.5500E+07-1.0000E+00 1.6000E+07-2.9000E-01222833103 1.8000E+07-2.8000E-01 2.0000E+07 0.0000E+00 1.0000E-05 0.0000E+00222833103 8.0000E+05 1.5931E-01 8.5000E+05 1.4664E-01 9.0000E+05 9.4772E-02222833103 1.0000E+06 3.1943E-02 1.5000E+06 3.1622E-02 2.0000E+06 4.3982E-02222833103 2.5000E+06 6.7150E-02 2.7000E+06 5.4584E-02 3.0000E+06 5.0435E-02222833103 3.5000E+06 4.9106E-02 4.0000E+06 5.6605E-02 4.5000E+06 7.1451E-02222833103 5.0000E+06 8.2388E-02 5.5000E+06 8.1997E-02 6.0000E+06 9.7032E-02222833103 6.5000E+06 9.2268E-02 8.0000E+06 8.9291E-02 8.5000E+06 1.1267E-01222833103 1.0000E+07 1.1269E-01 1.2000E+07 1.1312E-01 1.4000E+07 1.1326E-01222833103 1.5500E+07 1.1823E-01 1.6000E+07 1.1337E-01 1.8000E+07 1.1309E-01222833103 2.0000E+07 0.0000E+00 0.0000E+00 0.0000E+00 0.0000E+00 0.0000E+00222833103 0.0000E+00 0.0000E+00 0 1 88 44222833103 1.0000E-05 0.0000E+00 8.0000E+05 2.2843E-02 8.5000E+05 1.9354E-02222833103 9.0000E+05 8.0835E-03 1.0000E+06 9.1833E-04 1.5000E+06 8.9998E-04222833103 2.0000E+06 1.7410E-03 2.5000E+06 4.0582E-03 2.7000E+06 2.6815E-03222833103 3.0000E+06 2.2893E-03 3.5000E+06 2.1702E-03 4.0000E+06 2.8837E-03222833103 4.5000E+06 4.5947E-03 5.0000E+06 6.1090E-03 5.5000E+06 6.0512E-03222833103 6.0000E+06 8.4737E-03 6.5000E+06 4.2908E-03 7.0000E+06 4.2908E-03222833103 7.5000E+06 4.2908E-03 8.0000E+06 7.1756E-03 8.5000E+06 3.9985E-03222833103 9.0000E+06 3.9985E-03 9.5000E+06 3.9985E-03 1.0000E+07 2.7432E-03222833103 1.0500E+07 2.7432E-03 1.1000E+07 2.7432E-03 1.1500E+07 2.7432E-03222833103 1.2000E+07 3.1093E-03 1.2500E+07 3.1093E-03 1.3000E+07 3.1093E-03222833103 1.3500E+07 3.1093E-03 1.4000E+07 3.0015E-03 1.4500E+07 3.0015E-03222833103 1.5000E+07 3.0015E-03 1.5500E+07 1.2580E-02 1.6000E+07 3.3545E-03222833103 1.6500E+07 3.3545E-03 1.7000E+07 3.3545E-03 1.7500E+07 3.3545E-03222833103 1.8000E+07 3.2228E-03 1.8500E+07 3.2228E-03 1.9000E+07 3.2228E-03222833103 1.9500E+07 3.2228E-03 2.0000E+07 0.0000E+00 0.0000E+00 0.0000E+00222833103 0.0000E+00 0.0000E+00 0 8 88 44222833103 1.0000E-05 0.0000E+00 8.0000E+05 1.7840E-09 8.5000E+05 1.7944E-09222833103 9.0000E+05 9.4782E-10 1.0000E+06 6.1588E-10 1.5000E+06 6.9313E-09222833103 2.0000E+06 6.6180E-08 2.5000E+06 3.0335E-07 2.7000E+06 2.7625E-07222833103 3.0000E+06 3.4819E-07 3.5000E+06 5.0856E-07 4.0000E+06 9.3419E-07222833103 4.5000E+06 1.8450E-06 5.0000E+06 3.1013E-06 5.5000E+06 3.6005E-06222833103 6.0000E+06 5.7907E-06 6.5000E+06 3.5738E-06 7.0000E+06 4.1828E-06222833103 7.5000E+06 4.9454E-06 8.0000E+06 9.7757E-06 8.5000E+06 6.0984E-06222833103 9.0000E+06 6.5926E-06 9.5000E+06 7.2632E-06 1.0000E+07 5.3544E-06222833103 1.0500E+07 5.4347E-06 1.1000E+07 5.3985E-06 1.1500E+07 5.2463E-06222833103 1.2000E+07 5.6908E-06 1.2500E+07 5.3562E-06 1.3000E+07 5.0318E-06222833103 1.3500E+07 4.7183E-06 1.4000E+07 4.2268E-06 1.4500E+07 3.8407E-06222833103 1.5000E+07 3.4731E-06 1.5500E+07 1.3094E-05 1.6000E+07 3.1245E-06222833103 1.6500E+07 2.8157E-06 1.7000E+07 2.5230E-06 1.7500E+07 2.2418E-06222833103 1.8000E+07 1.9701E-06 1.8500E+07 1.8648E-06 1.9000E+07 1.7624E-06222833103 1.9500E+07 1.6595E-06 2.0000E+07 0.0000E+00 0.0000E+00 0.0000E+00222833103 0.0000E+00 0.0000E+00 0 1 4 2222833103 6.9000E+05 2.2843E-02 8.0000E+05 0.0 222833103 222833 0 2228 0 0 2.20470E+4 4.65484E+1 0 0 0 1222833231 0.0000E+00 0.0000E+00 0 231 0 1222833231 0.0000E+00 0.0000E+00 1 5 190 19222833231 1.000000-5 8.400000+6 1.200000+7 1.250000+7 1.300000+7 1.350000+7222833231 1.400000+7 1.450000+7 1.500000+7 1.550000+7 1.600000+7 1.650000+7222833231 1.700000+7 1.750000+7 1.800000+7 1.850000+7 1.900000+7 1.950000+7222833231 2.000000+7 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0222833231 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0222833231 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0222833231 0.000000+0 8.224320-1 3.047800-2-5.099880-3 2.660400-2 2.567630-2222833231 8.789060-3-2.089150-4 4.143110-4 5.627230-3 1.120560-2 1.502500-2222833231 1.653970-2 1.606880-2 1.428180-2 1.189650-2 9.535380-3 7.674830-3222833231 2.317220-2 1.579900-2 4.911180-3 9.215290-4 7.415560-4 1.566370-3222833231 2.456920-3 3.187640-3 3.730820-3 4.106900-3 4.345820-3 4.476950-3222833231 4.526420-3 4.516220-3 4.464290-3 4.384770-3 2.410370-2 1.179110-2222833231 2.085650-3-4.356060-5 1.197690-3 2.968140-3 4.280220-3 4.991600-3222833231 5.234730-3 5.189020-3 5.005500-3 4.790790-3 4.611590-3 4.504530-3222833231 4.485670-3 1.255610-2 7.443600-3 3.314840-3 1.501210-3 1.219560-3222833231 1.623630-3 2.199950-3 2.702100-3 3.042030-3 3.212750-3 3.242780-3222833231 3.171400-3 3.036720-3 2.870740-3 7.555650-3 4.699770-3 2.094550-3222833231 5.902510-4 3.201110-5 5.900340-5 3.656710-4 7.479170-4 1.087880-3222833231 1.328070-3 1.448730-3 1.451680-3 1.349620-3 3.780760-3 2.278690-3222833231 1.087240-3 3.782550-4 7.219620-5 3.330810-5 1.404120-4 3.044130-4222833231 4.669730-4 5.938050-4 6.675780-4 6.820470-4 2.094230-3 1.740670-3222833231 1.376410-3 1.067600-3 8.374700-4 6.850910-4 5.990970-4 5.649730-4222833231 5.687140-4 5.983170-4 6.442110-4 2.137400-3 2.230550-3 2.111580-3222833231 1.877100-3 1.606180-3 1.352240-3 1.146690-3 1.004800-3 9.311090-4222833231 9.235670-4 2.710740-3 2.837360-3 2.718770-3 2.461500-3 2.152640-3222833231 1.853540-3 1.602850-3 1.421840-3 1.319530-3 3.190670-3 3.236060-3222833231 3.076280-3 2.802720-3 2.488110-3 2.183810-3 1.922800-3 1.723960-3222833231 3.447240-3 3.420070-3 3.236110-3 2.964070-3 2.657910-3 2.356140-3222833231 2.084160-3 3.528090-3 3.454820-3 3.259670-3 2.991250-3 2.688170-3222833231 2.378610-3 3.491600-3 3.386500-3 3.181570-3 2.911070-3 2.602730-3222833231 3.371280-3 3.242790-3 3.031150-3 2.760710-3 3.195830-3 3.061290-3222833231 2.860660-3 3.016960-3 2.911930-3 2.923980-3 222833231 222833 0 2228 0 0 0 0 0 2.20480E+4 4.75360E+1 0 0 34 102231 1451 0.0 0.0 0 0 0 62231 1451 1.00000E+0 2.00000E+7 0 0 10 20022231 1451 3.00000E+2 0.0 1 0 242 62231 1451 22-Ti- 48 FEI EVAL-Feb02 K.I.Zolotarev 2231 1451 DIST-Feb2004 2231 1451 ----IRDF-2002 MATERIAL 2231 2231 1451 -----INCIDENT NEUTRON DATA 2231 1451 ------ENDF-6 FORMAT 2231 1451 ***************************************************************** 2231 1451 ********************** SPECIAL PURPOSE FILE ***************** 2231 1451 DOSIMETRY 2231 1451 2231 1451 For this special purpose library it was decided the reaction 2231 1451 MF=3 MT=28+104 would use MF=10, MT=5 to present this data. 2231 1451 This was done after processing through the codes. The 2231 1451 co-variance file was converted from MF/MT=33/28 to MF/MT=40/5 2231 1451 ***************************************************************** 2231 1451 22-TI-48 FEI EVAL-Feb02 K.I.Zolotarev 2231 1451 DIST-Feb02 20020214 2231 1451 ----BROND-2 MATERIAL 2231 2231 1451 -----INCIDENT NEUTRON DATA 2231 1451 ------ENDF-6 FORMAT 2231 1451 ------Russian Reactor Dosimetry File RRDF-2002 2231 1451 ***************************************************************** 2231 1451 Author of evaluation: K.I.Zolotarev 2231 1451 ***************************************************************** 2231 1451 MF= 3 2231 1451 MT= 28 -(n,np+pn+d) cross section data 2231 1451 -------------------------------------- 2231 1451 In this section is given the sum of cross sections of the re- 2231 1451 actions Ti-48(n,np)Sc-47 , Ti-48(n,pn)Sc-47 and Ti-48(n,d)Sc-47. 2231 1451 Excitation function for the Ti-48(n,x)Sc-47 reaction in the 2231 1451 energy region from threshold to 20 MeV was evaluated by means of 2231 1451 statistical analysis of experimental cross section data [1-12] 2231 1451 and data from STAPRE [13] calculation. 2231 1451 Analised experimental data were corrected to the new stan- 2231 1451 dards for monitor reactions cross sections and decay data. 2231 1451 The final procedure of evaluation Ti-48(n,x)Sc-47 excitation 2231 1451 function from threshold to 20 MeV has been carried out within the 2231 1451 framework of generalized least squares method. Rational function 2231 1451 was used as model function [14]. Calculations was performed by 2231 1451 means of Pade-2 code [15]. 2231 1451 U-235 thermal fission [16] and Cf-252 spontanious fission 2231 1451 neutron spectra [17] averaged cross-sections calculated from the 2231 1451 evaluated Ti-48(n,x)Sc-47 excitation function are the following: 2231 1451 2231 1451 -------------------------------------------- 2231 1451 TYPE OF SPECTRUM I , mb (calc.) 2231 1451 --------------------------I----------------- 2231 1451 U-235 neutron fission I 1.6558E-3 2231 1451 CF-252 spontan. fission I 4.2891E-3 2231 1451 2231 1451 MT=103 -(n,p) cross section 2231 1451 ------------------------------------- 2231 1451 Excitation function for the Ti-48(n,p)Sc-48 reaction in the 2231 1451 energy region from threshold to 20 MeV was evaluated by means of 2231 1451 of statistical analysis of experimental cross section data [1-2], 2231 1451 [4], [6-11], [18-41] and data from STAPRE [13] calculation. 2231 1451 The energy dependence of cross-section from 4.7 MeV to thres- 2231 1451 hold was extrapolated with L=0 penetrability function for the 2231 1451 the outgoing p + Sc48 channal [42]. 2231 1451 Analised microscopic experimental data were renormalized to 2231 1451 the new recommended standards for monitor reaction cross sections 2231 1451 and decay data. All experiments performed with Ti-natural samples 2231 1451 were corrected for the contribution of the Ti-49(n,np+pn)Sc-48 2231 1451 and Ti-49(n,d)Sc-48 reactions. Total cross section for these reac-2231 1451 tions were taken from ref. [43]. Experimental data from ref. [24] 2231 1451 were used partially. It was taken into account only data measured 2231 1451 relative Fe-56(n,p)Mn-56 monitor reaction. Data of Firkin [6] 2231 1451 obtained in the experiment with neutrons from D(d,n)He3 reaction 2231 1451 were renormalized to the results of his measurements with 14.1 2231 1451 MeV neutrons from T(d,n)He4 reaction. Cross section data of Smith 2231 1451 and Meadows [26] measured in the energy range 5.964 - 9.952 MeV 2231 1451 with using D(d,n)He3 neutron source were renormalized to the 2231 1451 value 30.7 mb at 10 MeV [38]. The correction factors for the 2231 1451 experimental data [6] and [26] were Fc=0.91853 and Fc=1.11111 , 2231 1451 respectively 2231 1451 Experimental cross section data [44-51] were rejected due to 2231 1451 their discrepancy with the main bulk of experimental data [1-2], 2231 1451 [4], [6-11], [18-41] and data from theoretical model calculatiod. 2231 1451 Statistical analysis of input cross section data was carried 2231 1451 out by means of PADE-2 code [15]. Rational function was used as 2231 1451 the model function [14]. 2231 1451 U-235 thermal fission [16] and Cf-252 spontanious fission 2231 1451 neutron spectra [17] averaged cross-sections calculated from the 2231 1451 evaluated Ti-48(n,p)Sc-48 excitation function are the following: 2231 1451 2231 1451 ÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÂÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÂÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄ 2231 1451 TYPE OF SPECTRUM ³ ,mb (calc.) ³ , mb (measured) 2231 1451 ÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÅÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÅÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄ 2231 1451 U-235 neutron fission ³ 0.30430 ³ 0.305 +- 0.020 [52] 2231 1451 ³ ³ 0.302 +- 0.010 [53] 2231 1451 ³ ³ 0.3007+- 0.0054 [54] 2231 1451 ÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÅÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÅÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄ 2231 1451 Cf-252 spont. fission ³ 0.42629 ³ 0.4275+- 0.0078 [55] 2231 1451 ³ ³ 0.4247+- 0.0080 [56] 2231 1451 ÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÁÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÁÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄ 2231 1451 2231 1451 MF=33 2231 1451 MT= 28 -(n,np+pn+d) cross section cov. matrix 2231 1451 --------------------------------------------- 2231 1451 Uncertainties in the evaluated excitation function for the 2231 1451 reaction Ti-48(n,x)Sc-47 are given in the form of relative cova- 2231 1451 riance matrix for the 15-neutron energy groups (LB=5). Covariance 2231 1451 matrix of uncertainties was calculated simultaneously with 2231 1451 recommended cross section data by means of PADE-2 code. 2231 1451 Eigenvalues of the 6-th digits relative covariance matrix 2231 1451 given in the 33-file are the following: 2231 1451 2231 1451 3.25031E-07 3.96552E-07 7.16579E-07 1.55391E-06 2231 1451 4.04176E-06 1.39260E-05 8.49003E-05 2.17726E-03 2231 1451 8.18710E-03 1.48647E-02 2.37004E-02 3.37652E-02 2231 1451 4.90298E-02 9.66009E-02 5.46438E-01 2231 1451 2231 1451 MT=103 -(n,p) cross section cov. matrix 2231 1451 --------------------------------------------- 2231 1451 Uncertainties in the evaluated excitation function for the 2231 1451 reaction Ti-48(n,p)Sc-48 are given in the form of relative cova- 2231 1451 riance matrix for the 26-neutron energy groups (LB=5). Covariance 2231 1451 matrix of uncertainties was calculated simultaneously with 2231 1451 recommended cross section data by means of PADE-2 code. 2231 1451 Eigenvalues of the 6-th digits relative covariance matrix 2231 1451 given in the 33-file are the following: 2231 1451 2231 1451 5.39341E-09 5.73858E-09 5.82529E-09 6.42044E-09 2231 1451 6.80480E-09 7.28300E-09 7.51108E-09 8.50118E-09 2231 1451 9.53827E-09 1.09813E-08 1.24867E-08 1.39806E-08 2231 1451 1.87772E-08 2.72244E-08 4.05820E-08 6.49966E-08 2231 1451 1.23779E-07 3.87183E-07 6.43873E-06 1.17950E-03 2231 1451 6.38433E-03 8.03793E-03 1.04937E-02 1.43313E-02 2231 1451 2.41167E-02 3.79706E-02 2231 1451 2231 1451 References : 2231 1451 1. W.G.Cross, H.L.Pai Prog. Report EANDC(CAN)-16, p.1, Jan. 1963 2231 1451 2. H.L.Pai Canadian J. of Physics, v.44, p.2337, 1966 2231 1451 3. S.M.Grimes et al. Nucl. Sci. Eng., v.62, p.187, Feb. 1977 2231 1451 4. S.M.Qaim, N.I.Molla Nucl. Phys., v.A283, p.269, June 1977 2231 1451 5. M.Viennot et al. 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Eng., v.108, p.289, July 1991 2231 1451 38. S.M.Qaim et al. Proc. of Int. Conference on Nuclear Data for 2231 1451 Science and Technology, Julich, FRG, 13-17 May 1991. Springer 2231 1451 Verlag, Berlin - Heidelberg, 1992, p.297-300 2231 1451 39. I.Garlea et al. J. Rev. Roum. Phys., v.37, n.1, p.19, 1992 2231 1451 40. Yuan Junqian et al. High Energy Physics and Nuclear Physics 2231 1451 (China), v.16, n.1, p.57, January 1992 2231 1451 41. A.Grallert et al. Progress Report,INDC(NDS)-286, p.131, 1993 2231 1451 42. S.A.Badikov, A.B.Pashchenko Voprosy Atomnoy Nauki i Tekhniki 2231 1451 Ser. Jadernye Konstanty, 2(53), p.70, 1987 2231 1451 43. K.Zolotarev RRDF-2002, MAT=2243, eval. February 2002 2231 1451 44. E.B.Paul, R.L.Clarke Canadian J. Phys., v.31, p.267, 1953 2231 1451 45. D.L.Allan Nucl. Phys., v.24, p.274, April 1961 2231 1451 46. J.E.Strain, W.J.Ross Report ORNL-3672, January 1965 2231 1451 47. H.K.Vonach et al. Proc. of 2-nd Conference on Nuclear Cross- 2231 1451 Sections and Technology, Washington D.C., 4-7 March 1968, 2231 1451 v.2, p.885 2231 1451 48. D.Crumpton J. Inorg. Nucl. Chem., v.31, p.3727, Dec. 1969 2231 1451 49. V.K.Tikku et al. Proc. of Nucl. Phys. and Solid State Phys. 2231 1451 Symp., Chandigarh, v.2, p.115, December 1972 2231 1451 50. R.Spangler et al. J. Trans. Amer. Nucl. Soc., v.22, p.818, 2231 1451 November 1975 2231 1451 51. K.T.Osman, F.I.Habbani Report, INDC(SUD)-001, October 1996 2231 1451 52. O.Horibe et al. Proc. of Conf.:50 Years with Nuclear Fission, 2231 1451 Washington D.C., 25-28 April 1989, v.2, p.923 2231 1451 53. W.Mannhart Proc. 5-th ASTM-EUR. Symp. on Reactor Dosimetry, 2231 1451 Geesthacht, FRG, September 24-28, 1984, Vol.2, p.813, 1985 2231 1451 54. W.Mannhart Progress Report INDC(Ger)-045, pp.40-43, 1999 2231 1451 55. W.Mannhart Handbook on Nuclear Activation Cross Sections , 2231 1451 IAEA Tech. Report Ser. No.273, p.413, 1987 2231 1451 56. W.Mannhart Validation of Differential Cross Sections with 2231 1451 Integral Data , Report INDC(NDS)-435, pp.59-64, IAEA, Vienna, 2231 1451 September 2002 2231 1451 ***************************************************************** 2231 1451 File 2 added to the pointwise file containing only the effective 2231 1451 scattering radius with no resonance parameters given. 2231 1451 Taken from ENDF/B-VI 2231 1451 ***************************************************************** 2231 1451 2231 1451 ***************** Program LINEAR (VERSION 2002-1) ***************2231 1451 For All Data Greater than 1.0000E-10 barns in Absolute Value 2231 1451 Data Linearized to Within an Accuracy of .100000000 per-cent 2231 1451 ***************** Program SIGMA1 (VERSION 2002-1) ***************2231 1451 Data Doppler Broadened to 300.000000 Kelvin 2231 1451 for All Data Greater than 1.0000E-10 barns in Absolute Value 2231 1451 Data Linearized to Within an Accuracy pf .100000000 per-cent 2231 1451 ***************** Program FIXUP (Version 2002-1) ****************2231 1451 Corrected ZA/AWR in All Sections-----------------------------Yes 2231 1451 Corrected Thresholds-----------------------------------------Yes 2231 1451 Extended Cross Sections to 20 MeV----------------------------No 2231 1451 Allow Cross Section Deletion---------------------------------No 2231 1451 Allow Cross Section Reconstruction---------------------------No 2231 1451 Make All Cross Sections Non-Negative-------------------------Yes 2231 1451 Delete Energies Not in Ascending Order-----------------------Yes 2231 1451 Deleted Duplicate Points-------------------------------------Yes 2231 1451 Check for Ascending MAT/MF/MT Order--------------------------Yes 2231 1451 Check for Legal MF/MT Numbers--------------------------------Yes 2231 1451 Allow Creation of Missing Sections---------------------------No 2231 1451 Allow Insertion of Energy Points-----------------------------No 2231 1451 Create Uniform Energy Grid-----------------------------------No 2231 1451 Delete Section if Cross Section =0 at All Energies-----------Yes 2231 1451 ***************** Program GROUPIE (VERSION 2002-1) **************2231 1451 Unshielded Group Averages Using 640 Groups 2231 1451 Weighting Spectrum: Flat (Constant) Spectrum 2231 1451 1 451 252 12231 1451 3 103 60 12231 1451 8 5 2 12231 1451 10 5 39 12231 1451 33 103 71 12231 1451 40 5 30 12231 1451 2231 1 0 2231 0 0 2.20480E+4 4.75360E+1 0 0 0 02231 3103 -3.21180E+6-3.21180E+6 0 0 1 1692231 3103 169 1 2231 3103 3200000.00 4.44705E-9 3300000.00 1.46148E-7 3400000.00 3.52717E-72231 3103 3500000.00 5.59287E-7 3600000.00 7.65856E-7 3700000.00 9.72426E-72231 3103 3800000.00 1.17900E-6 3900000.00 1.75487E-6 4000000.00 2.60258E-62231 3103 4100000.00 3.40719E-6 4200000.00 4.52734E-6 4300000.00 6.44311E-62231 3103 4400000.00 9.78106E-6 4500000.00 1.53460E-5 4600000.00 2.41598E-52231 3103 4700000.00 3.75024E-5 4800000.00 5.69529E-5 4900000.00 8.44353E-52231 3103 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1.083510-3223133103 8.630930-4 6.555010-4 4.857210-4 3.683870-4 3.084850-4 3.039140-4223133103 3.897200-4 6.216970-4 9.436640-4 1.029260-3 9.394040-4 8.298590-4223133103 7.159590-4 6.098560-4 5.194640-4 4.487370-4 3.839040-4 3.563470-4223133103 3.847270-4 9.747120-4 9.683430-4 9.248700-4 8.519660-4 7.582400-4223133103 6.516400-4 4.833150-4 2.585650-4 5.063660-5 1.056860-3 1.089950-3223133103 1.069780-3 1.003410-3 9.001680-4 6.972650-4 3.769060-4 3.602680-5223133103 1.198100-3 1.245680-3 1.236370-3 1.178080-3 1.018650-3 7.236530-4223133103 3.779790-4 1.370840-3 1.444020-3 1.469220-3 1.429200-3 1.281260-3223133103 1.063020-3 1.619290-3 1.760150-3 1.905790-3 2.014100-3 2.044310-3223133103 2.041250-3 2.425520-3 2.879870-3 3.259940-3 3.234920-3 4.327790-3223133103 5.371820-3 6.404900-3 8.492620-3 1.170350-2 223133103 223133 0 2231 0 0 2.20480E+4 4.75360E+1 0 0 0 1223133231 0.0000E+00 0.0000E+00 0 231 0 1223133231 0.0000E+00 0.0000E+00 1 5 153 17223133231 1.000000-5 9.400000+6 1.300000+7 1.350000+7 1.400000+7 1.450000+7223133231 1.500000+7 1.550000+7 1.600000+7 1.650000+7 1.700000+7 1.750000+7223133231 1.800000+7 1.850000+7 1.900000+7 1.950000+7 2.000000+7 0.000000+0223133231 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0223133231 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0223133231 0.000000+0 0.000000+0 0.000000+0 4.707110-1 1.759780-1 1.697230-2223133231 -8.146730-3 1.038100-2-5.224470-3-2.034860-2-1.446150-2-9.242090-4223133231 6.358550-3 3.918520-3-4.555240-3-1.358030-2-1.920200-2-1.987550-2223133231 1.052700-1 2.855580-2-1.423330-3 4.824040-3 5.676170-3-7.073830-4223133231 -3.600630-3-2.063640-3 5.308030-4 1.841570-3 1.461860-3 1.058170-4223133231 -1.304640-3-2.135350-3 2.285560-2 7.762240-3 1.086850-3 2.996610-3223133231 4.013350-3 2.410030-3 5.183160-4-1.889870-4 3.334230-4 1.449850-3223133231 2.490700-3 3.052080-3 3.021780-3 1.008370-2 4.078560-3-7.285120-4223133231 -1.046370-3 3.213810-4 1.205820-3 1.175750-3 5.878370-4-9.490980-5223133231 -5.712350-4-7.367790-4-6.212280-4 8.396540-3 7.438880-3 3.738740-3223133231 1.061670-3 2.405110-4 6.122910-4 1.371180-3 2.005320-3 2.316480-3223133231 2.311780-3 2.087370-3 1.621120-2 1.515040-2 8.632830-3 3.121490-3223133231 1.138420-3 2.178660-3 4.619640-3 6.965430-3 8.348450-3 8.521370-3223133231 1.895660-2 1.478520-2 8.727250-3 4.770970-3 3.793980-3 4.917860-3223133231 6.814000-3 8.444710-3 9.280860-3 1.554920-2 1.270370-2 8.999380-3223133231 6.138550-3 4.681970-3 4.463460-3 5.026080-3 5.915220-3 1.336970-2223133231 1.145390-2 8.468040-3 5.709430-3 3.904690-3 3.230420-3 3.514540-3223133231 1.161810-2 1.015980-2 8.083180-3 6.198510-3 4.939490-3 4.404970-3223133231 1.095260-2 1.089590-2 1.022320-2 9.212560-3 8.099470-3 1.315310-2223133231 1.425810-2 1.409830-2 1.289180-2 1.700900-2 1.794020-2 1.714270-2223133231 1.989720-2 1.988000-2 2.083550-2 223133231 223133 0 2231 0 0 0 0 0 2.20490E+4 4.85274E+1 0 0 34 202234 1451 0.0 0.0 0 0 0 62234 1451 1.00000E+0 2.00000E+7 0 0 10 20022234 1451 3.00000E+2 0.0 1 0 116 42234 1451 22-Ti- 49 FEI EVAL-Feb02 K.I.Zolotarev 2234 1451 DIST-Feb2004 2234 1451 ----IRDF-2002 MATERIAL 2234 2234 1451 -----INCIDENT NEUTRON DATA 2234 1451 ------ENDF-6 FORMAT 2234 1451 ***************************************************************** 2234 1451 ********************** SPECIAL PURPOSE FILE ***************** 2234 1451 DOSIMETRY 2234 1451 2234 1451 For this special purpose library it was decided the reaction 2234 1451 MF=3 MT=28+104 would use MF=10, MT=5 to present this data. 2234 1451 This was done after processing through the codes. The 2234 1451 co-variance file was converted from MF/MT=33/28 to MF/MT=40/5 2234 1451 ***************************************************************** 2234 1451 22-TI-49 FEI EVAL-Feb02 K.I.Zolotarev 2234 1451 DIST-Feb02 2234 1451 ----BROND-2 MATERIAL 2234 2234 1451 -----INCIDENT NEUTRON DATA 2234 1451 ------ENDF-6 FORMAT 2234 1451 ------Russian Reactor Dosimetry File RRDF-2002 2234 1451 ***************************************************************** 2234 1451 Author of evaluation: K.I.Zolotarev 2234 1451 ***************************************************************** 2234 1451 MF= 3 2234 1451 MT= 28 -(n,np+pn+d) cross section data 2234 1451 -------------------------------------- 2234 1451 In this section is given the sum of cross section of the reac-2234 1451 tions Ti-49(n,np)Sc-48 , Ti-49(n,pn)Sc-48 and Ti-49(n,d)Sc-48. 2234 1451 Excitation function for the Ti-49(n,x)Sc-48 reaction in the 2234 1451 energy region from threshold to 20 MeV was evaluated by means of 2234 1451 statistical analysis of experimental cross section data [1-5] and 2234 1451 data from STAPRE [6] calculation. 2234 1451 Analysed experimental data were renormalized to the new stan- 2234 1451 dards for monitor reactions cross sections and decay data. Data 2234 1451 of Pai [2] measured in the energy region 16-19.5 MeV with using 2234 1451 Van de Graaff accelerator were renormalized to the results of 2234 1451 the theoretical model calculation. 2234 1451 The final procedure of evaluation Ti-49(n,x)Sc-48 excitation 2234 1451 function from threshold to 20 MeV has been carried out within the 2234 1451 framework of generalized least squares method. Rational function 2234 1451 was used as model function [7]. Calculations was performed by 2234 1451 means of Pade-2 code [8]. 2234 1451 U-235 thermal fission [9] and Cf-252 spontaneous fission 2234 1451 neutron spectra [10] averaged cross-sections calculated from the 2234 1451 evaluated Ti-49(n,x)Sc-48 excitation function are the following: 2234 1451 2234 1451 -------------------------------------------- 2234 1451 TYPE OF SPECTRUM I , mb (calc.) 2234 1451 --------------------------I----------------- 2234 1451 U-235 neutron fission I 1.0041E-3 2234 1451 CF-252 spontan. fission I 2.6070E-3 2234 1451 2234 1451 MF=33 2234 1451 MT= 28 -(n,np+pn+d) cross section cov. matrix 2234 1451 --------------------------------------------- 2234 1451 Uncertainties in the evaluated excitation function for the 2234 1451 reaction Ti-49(n,x)Sc-48 are given in the form of relative cova- 2234 1451 riance matrix for the 14-neutron energy groups (LB=5). Covariance 2234 1451 matrix of uncertainties was calculated simultaneously with 2234 1451 recommended cross section data by means of PADE-2 code. 2234 1451 Eigenvalues of the 6-th digits relative covariance matrix 2234 1451 given in the 33-file are the following: 2234 1451 2234 1451 2.34704E-08 3.44912E-08 5.30971E-08 9.47934E-08 2234 1451 1.99158E-07 5.01449E-07 1.49592E-06 5.55617E-06 2234 1451 1.56887E-04 2.23251E-03 8.77261E-03 1.50447E-02 2234 1451 3.12585E-02 8.36148E-02 2234 1451 2234 1451 References : 2234 1451 1. W.G.Cross, H.L.Pai Progress Report, EANDC(CAN)-16, p.1, 2234 1451 January 1963 2234 1451 2. H.L.Pai Canadian J. of Physics, v.44, p.2337, 1966 2234 1451 3. S.M.Qaim Nucl. Phys., v.A382, n.2, p.255, July 1982 2234 1451 4. I.Ribansky, S.Gmuca J. Phys.G, v.9, p.1537, December 1983 2234 1451 5. Y.Ikeda et al. Report JAERI-1312, March 1988 2234 1451 6. M.Uhl, B.Strohmaier Computer Code STAPRE for Particle Induced 2234 1451 Activation Cross Section and Related Quantities, Report 2234 1451 IRK 76-01, Vienna, 1976 2234 1451 7. S.Badikov, N.Rabotnov, K.Zolotarev Proc. of NEANSC Speciali- 2234 1451 st's Meeting on Evaluation and Processing of Covariance Data, 2234 1451 Oak Ridge, USA, 7-9 September 1992, OECD, Paris, 1993, p.105 2234 1451 8. S.A.Badikov et al. Preprint FEI-1686, Obninsk, 1985 2234 1451 9. L.W.Weston et al. Evaluated Neutron Data for U-235, ENDF/B-VI 2234 1451 Library, MAT=9228, MF=5, MT=18, eval. April 1989 2234 1451 10. W.Mannhart IAEA-TECDOC-410, p.158, IAEA, Vienna, 1987 2234 1451 ***************************************************************** 2234 1451 File 2 added to the pointwise file containing only the effective 2234 1451 scattering radius with no resonance parameters given. 2234 1451 Taken from JENDL-3.2 2234 1451 ***************************************************************** 2234 1451 2234 1451 ***************** Program LINEAR (VERSION 2002-1) ***************2234 1451 For All Data Greater than 1.0000E-10 barns in Absolute Value 2234 1451 Data Linearized to Within an Accuracy of .100000000 per-cent 2234 1451 ***************** Program SIGMA1 (VERSION 2002-1) ***************2234 1451 Data Doppler Broadened to 300.000000 Kelvin 2234 1451 for All Data Greater than 1.0000E-10 barns in Absolute Value 2234 1451 Data Linearized to Within an Accuracy pf .100000000 per-cent 2234 1451 ***************** Program FIXUP (Version 2002-1) ****************2234 1451 Corrected ZA/AWR in All Sections-----------------------------Yes 2234 1451 Corrected Thresholds-----------------------------------------Yes 2234 1451 Extended Cross Sections to 20 MeV----------------------------No 2234 1451 Allow Cross Section Deletion---------------------------------No 2234 1451 Allow Cross Section Reconstruction---------------------------No 2234 1451 Make All Cross Sections Non-Negative-------------------------Yes 2234 1451 Delete Energies Not in Ascending Order-----------------------Yes 2234 1451 Deleted Duplicate Points-------------------------------------Yes 2234 1451 Check for Ascending MAT/MF/MT Order--------------------------Yes 2234 1451 Check for Legal MF/MT Numbers--------------------------------Yes 2234 1451 Allow Creation of Missing Sections---------------------------No 2234 1451 Allow Insertion of Energy Points-----------------------------No 2234 1451 Create Uniform Energy Grid-----------------------------------No 2234 1451 Delete Section if Cross Section =0 at All Energies-----------Yes 2234 1451 ***************** Program GROUPIE (VERSION 2002-1) **************2234 1451 Unshielded Group Averages Using 640 Groups 2234 1451 Weighting Spectrum: Flat (Constant) Spectrum 2234 1451 1 451 124 12234 1451 8 5 2 12234 1451 10 5 39 12234 1451 40 5 27 12234 1451 2234 1 0 2234 0 0 2.20490E+4 4.85274E+1 0 0 1 02234 3231 -9.12956E+6-9.12956E+6 21048 0 1 1082234 3231 108 1 2234 3231 9300000.00 1.10601E-9 9400000.00 4.31560E-9 9500000.00 7.57585E-92234 3231 9600000.00 1.08361E-8 9700000.00 1.40964E-8 9800000.00 1.73566E-82234 3231 9900000.00 2.06169E-8 10000000.0 4.97413E-8 10100000.0 1.04730E-72234 3231 10200000.0 1.59719E-7 10300000.0 2.14707E-7 10400000.0 2.69696E-72234 3231 10500000.0 7.67187E-7 10600000.0 1.70718E-6 10700000.0 2.64718E-62234 3231 10800000.0 3.58717E-6 10900000.0 4.52716E-6 11000000.0 6.74413E-62234 3231 11100000.0 1.09550E-5 11200000.0 1.67516E-5 11300000.0 2.44527E-52234 3231 11400000.0 3.44093E-5 11500000.0 4.70069E-5 11600000.0 6.26681E-52234 3231 11700000.0 8.18556E-5 11800000.0 .000105075 11900000.0 .0001328762234 3231 12000000.0 .000165858 12100000.0 .000204674 12200000.0 .0002500282234 3231 12300000.0 .000302684 12400000.0 .000363468 12500000.0 .0004332672234 3231 12600000.0 .000513036 12700000.0 .000603801 12800000.0 .0007066562234 3231 12900000.0 .000822770 13000000.0 .000953387 13100000.0 .0010998252234 3231 13200000.0 .001263485 13300000.0 .001445825 13400000.0 .0016483902234 3231 13500000.0 .001872795 13600000.0 .002120700 13700000.0 .0023938452234 3231 13800000.0 .002694020 13900000.0 .003023020 14000000.0 .0033827102234 3231 14100000.0 .003774980 14200000.0 .004201675 14300000.0 .0046646352234 3231 14400000.0 .005165665 14500000.0 .005706495 14600000.0 .0062887702234 3231 14700000.0 .006914025 14800000.0 .007583640 14900000.0 .0082988252234 3231 15000000.0 .009060585 15100000.0 .009869710 15200000.0 .0107266502234 3231 15300000.0 .011631650 15400000.0 .012584700 15500000.0 .0135853002234 3231 15600000.0 .014632650 15700000.0 .015725600 15800000.0 .0168627502234 3231 15900000.0 .018042250 16000000.0 .019270825 16100000.0 .0205278752234 3231 16200000.0 .021818075 16300000.0 .023141425 16400000.0 .0244962002234 3231 16500000.0 .025882400 16600000.0 .027268600 16700000.0 .0286769782234 3231 16800000.0 .030107533 16900000.0 .031538089 17000000.0 .0329686442234 3231 17100000.0 .034399200 17200000.0 .035829756 17300000.0 .0372603112234 3231 17400000.0 .038690867 17500000.0 .040121422 17600000.0 .0415165752234 3231 17700000.0 .042876325 17800000.0 .044236075 17900000.0 .0455958252234 3231 18000000.0 .046914233 18100000.0 .048191300 18200000.0 .0494683672234 3231 18300000.0 .050704483 18400000.0 .051899650 18500000.0 .0530948172234 3231 18600000.0 .054247567 18700000.0 .055357900 18800000.0 .0564682332234 3231 18900000.0 .057530588 19000000.0 .058544963 19100000.0 .0595593382234 3231 19200000.0 .060573713 19300000.0 .061538125 19400000.0 .0624525752234 3231 19500000.0 .063367025 19600000.0 .064281475 19700000.0 .0651578332234 3231 19800000.0 .065996100 19900000.0 .066834367 20000000.0 0.0 2234 3231 2234 3 0 2234 0 0 2.20490E+4 4.85274E+1 0 0 0 1223433231 0.0000E+00 0.0000E+00 0 231 0 1223433231 0.0000E+00 0.0000E+00 1 5 136 16223433231 1.000000-5 9.300000+6 1.350000+7 1.400000+7 1.450000+7 1.500000+7223433231 1.550000+7 1.600000+7 1.650000+7 1.700000+7 1.750000+7 1.800000+7223433231 1.850000+7 1.900000+7 1.950000+7 2.000000+7 0.000000+0 0.000000+0223433231 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0223433231 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0223433231 0.000000+0 2.340470-2 1.573760-2 1.165090-2 8.156660-3 5.541340-3223433231 3.969430-3 3.457650-3 3.824990-3 4.757050-3 5.897230-3 6.966960-3223433231 7.813930-3 8.420550-3 8.858660-3 1.315090-2 1.087230-2 8.474760-3223433231 6.164580-3 4.207140-3 2.837340-3 2.197190-3 2.273200-3 2.909620-3223433231 3.862550-3 4.881200-3 5.757010-3 6.348110-3 9.682370-3 8.120140-3223433231 6.337310-3 4.558730-3 3.043450-3 2.025410-3 1.624510-3 1.820480-3223433231 2.467110-3 3.358720-3 4.283100-3 5.063930-3 7.358580-3 6.232520-3223433231 4.886070-3 3.524410-3 2.383810-3 1.652690-3 1.421190-3 1.664730-3223433231 2.277620-3 3.115020-3 4.034000-3 5.813390-3 5.091770-3 4.162120-3223433231 3.178860-3 2.326160-3 1.761320-3 1.580970-3 1.811250-3 2.427040-3223433231 3.374180-3 5.065940-3 4.757790-3 4.186930-3 3.465820-3 2.759470-3223433231 2.250120-3 2.087010-3 2.377510-3 3.180650-3 5.131690-3 5.149110-3223433231 4.804660-3 4.208600-3 3.567080-3 3.106570-3 3.039490-3 3.527780-3223433231 5.806850-3 6.020090-3 5.794240-3 5.284890-3 4.721960-3 4.359860-3223433231 4.419540-3 6.857770-3 7.193930-3 7.072420-3 6.651570-3 6.153380-3223433231 5.798070-3 8.165580-3 8.611450-3 8.566600-3 8.159510-3 7.554320-3223433231 9.678720-3 1.017690-2 1.011670-2 9.575680-3 1.129210-2 1.182470-2223433231 1.177030-2 1.316420-2 1.408870-2 1.651520-2 223433231 223433 0 2234 0 0 0 0 0 2.30510E+4 5.05063E+1 0 0 34 102328 1451 0.0 0.0 0 0 0 62328 1451 1.00000E+0 2.00000E+7 0 0 10 20022328 1451 3.00000E+2 0.0 1 0 185 32328 1451 23-V - 51 FEI EVAL-Dec01 K.I.Zolotarev 2328 1451 DIST-Feb2004 2328 1451 ----IRDF-2002 MATERIAL 2328 2328 1451 -----INCIDENT NEUTRON DATA 2328 1451 ------ENDF-6 FORMAT 2328 1451 ***************************************************************** 2328 1451 23-V-51 FEI EVAL-Dec01 K.I.Zolotarev 2328 1451 DIST-Jan02 20020118 2328 1451 ----BROND-2 MATERIAL 2328 2328 1451 -----INCIDENT NEUTRON DATA 2328 1451 ------ENDF-6 FORMAT 2328 1451 ------Russian Reactor Dosimetry File RRDF-2002 2328 1451 ***************************************************************** 2328 1451 Author of evaluation: K.I.Zolotarev 2328 1451 ***************************************************************** 2328 1451 MF=3 2328 1451 MT=107 -(n,a) cross section 2328 1451 ------------------------------------- 2328 1451 Excitation function for the V-51(n,a)Sc-48 reaction in the 2328 1451 energy region from threshold to 20 MeV was evaluated by means of 2328 1451 statistical analysis of experimental cross section data [1-21] 2328 1451 and data from GNASH calculation [45]. 2328 1451 All analysed microscopic experimental data [1-38] and integ- 2328 1451 ral experimental data [39-44] were renormalized to the new recom- 2328 1451 mended standards for monitor reactions cross sections and decay 2328 1451 data. 2328 1451 Experimental cross section data from ref. [22-38] were rejec- 2328 1451 ted due to their big discrepancy with the main bulk of experimen- 2328 1451 al data [1-21], data from theoretical model calculation and data 2328 1451 from (n,a) cross section systematics. 2328 1451 Statistical analysis of input cross section data was carried 2328 1451 out by means of PADE-2 code [46]. Rational function was used as 2328 1451 the model function [47]. 2328 1451 The evaluated V-51(n,a)Sc-48 excitation function averaged 2328 1451 on U-235 neutron fission spectrum [48] and Cf-252 spontaneous 2328 1451 fission neutron spectrum [49] gives the next values : 2328 1451 ÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÂÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÂÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄ 2328 1451 TYPE OF SPECTRUM ³ ,mb (calc.) ³ , mb (measured) 2328 1451 ÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÅÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÅÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄ 2328 1451 U-235 neutron fission ³ 0.024414 ³ 0.0235 +-0.0015 [40] 2328 1451 ³ ³ 0.0215 +-0.0013 [41] 2328 1451 ³ ³ 0.02438+-0.00056 [42] 2328 1451 ÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÅÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÅÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄ 2328 1451 CF-252 spont. fission ³ 0.038514 ³ 0.03904+-0.00086 [43] 2328 1451 ÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÁÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÁÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄ 2328 1451 2328 1451 MF=33 2328 1451 MT=107 -(n,a) cross section cov. matrix 2328 1451 --------------------------------------- 2328 1451 Uncertainties in the evaluated excitation function for the 2328 1451 reaction V-51(n,a)Sc-48 are given in the form of relative covari- 2328 1451 ance matrix for the 26-neutron energy groups (LB=5). Covariance 2328 1451 matrix of uncertainties was calculated simultaneously with recom- 2328 1451 mended cross section data by means of PADE-2 code. 2328 1451 Eigenvalues of the 6-th digits relative covariance matrix 2328 1451 given in the 33-file are the following: 2328 1451 2328 1451 4.46067E-09 4.97872E-09 5.09809E-09 5.50951E-09 2328 1451 6.22582E-09 6.26808E-09 6.80068E-09 7.58613E-09 2328 1451 9.63723E-09 1.29149E-08 1.82432E-08 2.72742E-08 2328 1451 4.31460E-08 7.37989E-08 1.54144E-07 3.75108E-07 2328 1451 2.13791E-06 1.13740E-04 6.78127E-04 9.37900E-04 2328 1451 1.40076E-03 2.54577E-03 4.01102E-03 5.22789E-03 2328 1451 1.00172E-02 1.34609E-02 2328 1451 2328 1451 References : 2328 1451 1. H.Vonach, H.Muenzer Oesterr. Akad. Wiss., Math + Naturw. 2328 1451 Anzeiger, v.95, p.199, November 1958 2328 1451 2. H.K.Vonach et al. Proc. 2nd Conf. on Nuclear Cross-Sections 2328 1451 and Technology, Washington D.C., 4-7 March 1968, v.2, p.885 2328 1451 3. V.N.Levkovskiy et al. Yadernaja Fizika (Sov.), v.10, n.1, 2328 1451 p.44, July 1968 2328 1451 4. J.C.Robertson,B.Audric,P.Kolkowski J. of Nucl. Energy, v.27, 2328 1451 p.531, August 1973 2328 1451 5. A.Paulsen et al. Atomkernenergie, v.22, p.291, Sep. 1974 2328 1451 6. W.Mannhart, H.Vonach Zeitschrift fuer Physik, Section A, 2328 1451 v.272, p.279, March 1975 2328 1451 7. S.M.Qaim, N.I.Molla Proc. of 9th Symposium on Fusion Techno- 2328 1451 logy, Garmisch, 14-18 June 1976, FRG, Pergamon Press, p.589 2328 1451 8. O.Schwerer et al. Oesterr. Akad. Wiss., Math + Naturw. 2328 1451 Anzeiger, v.113, p.153, June 1976 2328 1451 9. E.Zupranska et al. Progress Report INR-1821/I/PL/A, 1980 2328 1451 E.Zupranska et al. Acta. Phys. Pol., v.B11, p.853, Nov. 1980 2328 1451 10. I.Kanno, J.W.Meadows, D.L.Smith Annals of Nuclear Energy, 2328 1451 v.11, p.623, December 1984 2328 1451 11. J.W.Meadows, D.L.Smith et al. Annals of Nuclear Energy, v.14, 2328 1451 p.489, September 1987 2328 1451 12. Y.Ikeda et al. Report JAERI-1312, March 1988 2328 1451 13. K.Kobayashi, I.Kimura Proc. of an Intern. Conf. on Nuclear 2328 1451 Data for Science and Technology, 30 May - 3 June 1988, Mito, 2328 1451 Japan, Saikon Publishing Co., LTD, pp.261-265, 1989 ; 2328 1451 14. Lu Hanlin et al. Report INDC(CPR)-16, August 1989 2328 1451 15. I.Kimura, K.Kobayashi Nucl. Sci. Eng., v.106, p.332, 1990 2328 1451 16. N.I.Molla et al. Proc. of International Conference on Nuclear 2328 1451 Data for Science and Technology, Gatlinburg, Tennessee, USA, 2328 1451 9-13 May 1994, v.2, pp.938-940 2328 1451 17. A.A.Filatenkov et al. VANT, Ser.: Yadernye Konstanty, v.2, 2328 1451 p.8, Moscow, 1996 2328 1451 18. D.Subasi et al. Nucl. Sci. Eng., to be published in 1997 2328 1451 19. A.D.Majdeddin "Measurement and analysis of excitation func- 2328 1451 tions for fast neutron induced reactions from threshold to 2328 1451 20 MeV". PHD dissertation, Inst. of Experimental Physics , 2328 1451 Kossuth University, Hungary, June 1998 2328 1451 20. A.A.Filatenkov et al. Report RI-252, St.Petersburg, May 1999 2328 1451 21. P.Reimer et al. Progress Report On Nuclear Data Research in 2328 1451 the FRG, INDC(Ger)-047, July 2001 2328 1451 22. I.Kumabe J. Phys. Soc. Jap., v.13, p.325, 1958 2328 1451 23. E.B.Paul, R.L.Clarke Canadian J. of Phys., v.31, p.267, 1953 2328 1451 24. A.Poularikas, R.W.Fink Progress Report A-ARK-60, p.3, 1960 2328 1451 25. F.Gabbard et al. Bull. Amer. Phys. Soc., v.5, p.42(C8), 1960 2328 1451 26. M.Bormann et al. Phys. Rev., v.22, p.602, October 1961 2328 1451 27. M.Hillman Nucl. Phys., v.37, p.78, 1962 2328 1451 28. E.T.Bramlitt, R.W.Fink Phys. Rev., v.131, p.2649, Sep. 1963 2328 1451 29. J.E.Strain, W.J.Ross Report ORNL-3672, January 1965 2328 1451 30. D.Crumpton J. Inorg. Nucl. Chem., v.31, p.3727, Dec. 1969 2328 1451 31. R.Spangler et al. Transactions of the American Nuclear 2328 1451 Society, v.22, p.818, November 1975 2328 1451 32. U.Garuska et al. Progress Report INDC(POL)-9, p.16, Sep. 1978 2328 1451 33. W.H.Warren, W.L.Alford Annals of Nucl. Energy, v.9,p.369,1982 2328 1451 34. G.Helfer et al. Czechoslovak Journal of Physics, Part B, 2328 1451 v.34, p.30, 1984 2328 1451 35. R.Pepelnik et al. Int. Conf. on Nuclear Data for Basic and 2328 1451 Applied Science, Santa Fe, N.M., 13-17 May 1985, v.1, p.211 2328 1451 36. N.I.Molla et al. Progress Report INDC(BAN)-003 , Sep. 1986 2328 1451 37. Y.Ikeda et al. Progress Report INDC(JPN)-162/U,p.24,Aug. 1992 2328 1451 38. J.Cezar Suita et al. Nucl. Sci. Eng., v.126, p.101, 1997 2328 1451 39. F.Nasyrov,B.D.Sciborskij Atomnaya Energija (Sov.), v.25, n.5,2328 1451 p.437, November 1968 2328 1451 40. I.Kimura, K.Kobayashi, T.Shibata Nucl. Sci. Technology., v.8,2328 1451 p.59, February 1971 2328 1451 41. K.Kobayashi, I.Kimura et al. Nucl. Sci. Technology., v.13, 2328 1451 p.531, October 1976 2328 1451 42. W.Mannhart Progress Report INDC(Ger)-045, pp.40-43, 1999 2328 1451 43. W.Mannhart Proc. of 5-th ASTM-EURATOM Symposium on Reactor 2328 1451 Dosimetry. Geesthacht, Sep.24-28, 1984, Vol.2, p.813, 1985 2328 1451 44. J.Csikai, Z.Dezso Proc. of 4th All Union Conf. on Neutron 2328 1451 Physics, Kiev, 18-22 April 1977, v.3, p.32 2328 1451 45. P.G.Young, E.D.Arthur A Preequilibrium Statistical Nuclear 2328 1451 Model Code for Calculation of Cross Section and Emission 2328 1451 Spectra. Report LA-6947, Los Alamos, 1977 ; 2328 1451 E.L.Trykiv, G.Ya.Tertychnyi Private communication, IPPE, 2328 1451 Obninsk, May 1999 2328 1451 46. S.A.Badikov et.al. Preprint FEI-1686, Obninsk, 1985 2328 1451 47. S.Badikov, N.Rabotnov, K.Zolotarev Proc. of NEANSC Speciali- 2328 1451 st's Meeting on Evaluation and Processing of Covariance Data, 2328 1451 Oak Ridge, USA, 7-9 September 1992, OECD, Paris, 1993, p.105 2328 1451 48. L.W.Weston et al. Evaluated Neutron Data for U-235, ENDF/B-VI 2328 1451 Library, MAT=9228, MF=5, MT=18, eval. April 1989 2328 1451 49. W.Mannhart IAEA-TECDOC-410, p.158, IAEA, Vienna, 1987 2328 1451 ***************************************************************** 2328 1451 The Q values and threshold energies were updated prior to pro- 2328 1451 cessing through the codes to comply with the values obtained 2328 1451 using the NNDC calculation program which is based on the 1995 2328 1451 Update to the Atomic mass Evaluation. 2328 1451 2328 1451 File 2 added to the pointwise file containing only the effective 2328 1451 scattering radius with no resonance parameters given. 2328 1451 Taken from ENDF/B-VI 2328 1451 ***************************************************************** 2328 1451 2328 1451 ***************** Program LINEAR (VERSION 2002-1) ***************2328 1451 For All Data Greater than 1.0000E-10 barns in Absolute Value 2328 1451 Data Linearized to Within an Accuracy of .100000000 per-cent 2328 1451 ***************** Program SIGMA1 (VERSION 2002-1) ***************2328 1451 Data Doppler Broadened to 300.000000 Kelvin 2328 1451 for All Data Greater than 1.0000E-10 barns in Absolute Value 2328 1451 Data Linearized to Within an Accuracy pf .100000000 per-cent 2328 1451 ***************** Program FIXUP (Version 2002-1) ****************2328 1451 Corrected ZA/AWR in All Sections-----------------------------Yes 2328 1451 Corrected Thresholds-----------------------------------------Yes 2328 1451 Extended Cross Sections to 20 MeV----------------------------No 2328 1451 Allow Cross Section Deletion---------------------------------No 2328 1451 Allow Cross Section Reconstruction---------------------------No 2328 1451 Make All Cross Sections Non-Negative-------------------------Yes 2328 1451 Delete Energies Not in Ascending Order-----------------------Yes 2328 1451 Deleted Duplicate Points-------------------------------------Yes 2328 1451 Check for Ascending MAT/MF/MT Order--------------------------Yes 2328 1451 Check for Legal MF/MT Numbers--------------------------------Yes 2328 1451 Allow Creation of Missing Sections---------------------------No 2328 1451 Allow Insertion of Energy Points-----------------------------No 2328 1451 Create Uniform Energy Grid-----------------------------------No 2328 1451 Delete Section if Cross Section =0 at All Energies-----------Yes 2328 1451 ***************** Program GROUPIE (VERSION 2002-1) **************2328 1451 Unshielded Group Averages Using 640 Groups 2328 1451 Weighting Spectrum: Flat (Constant) Spectrum 2328 1451 1 451 192 12328 1451 3 107 64 12328 1451 33 107 71 12328 1451 2328 1 0 2328 0 0 2.30510E+4 5.05063E+1 0 0 0 02328 3107 -2.05800E+6-2.05800E+6 0 0 1 1812328 3107 181 1 2328 3107 2000000.00 1.8941E-13 2100000.00 1.23745E-9 2200000.00 3.65187E-92328 3107 2300000.00 6.06629E-9 2400000.00 8.48070E-9 2500000.00 1.08951E-82328 3107 2600000.00 1.33095E-8 2700000.00 1.57240E-8 2800000.00 1.81384E-82328 3107 2900000.00 2.05528E-8 3000000.00 2.17904E-8 3100000.00 2.18512E-82328 3107 3200000.00 2.27955E-8 3300000.00 2.90406E-8 3400000.00 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6.691760-4232833107 6.282040-4 5.043390-4 3.275690-4 1.439350-4 5.926440-6-3.705100-5232833107 7.612450-4 7.584940-4 6.595730-4 4.869100-4 2.815200-4 9.398340-5232833107 -2.521280-5 8.108710-4 7.715490-4 6.460520-4 4.571060-4 2.389510-4232833107 3.009620-5 8.139380-4 7.703780-4 6.367250-4 4.204600-4 1.378650-4232833107 8.239680-4 7.757420-4 6.043510-4 3.003020-4 8.270000-4 7.491350-4232833107 5.126110-4 8.137850-4 7.643050-4 1.041490-3 232833107 232833 0 2328 0 0 0 0 0 2.40000+ 4 0.0 + 0 -1 0 34 12400 1451 0.0 + 0 0.0 + 0 0 0 0 62400 1451 1.0 + 0 0.0 + 0 0 0 10 22400 1451 0.0 + 0 0.0 + 0 0 0 15 22400 1451 24-CR- 0 PETTEN EVAL-MAR85 W.L.ZIJP 2400 1451 PRIVATE COMM. DIST-MAR85 850320 2400 1451 ----IRDF-90/NMF-G MATERIAL 2400 2400 1451 -----INCIDENT-NEUTRON DATA 2400 1451 ------ENDF-6 FORMAT 2400 1451 2400 1451 DAMAGE CROSS SECTIONS OF CHROMIUM IN STEEL 2400 1451 ASTM STANDARD 2400 1451 2400 1451 ***************** PROGRAM LINEAR (VERSION 84-2) *****************2400 1451 DATA LINEARIZED TO WITHIN AN ACCURACY OF 0.100 PER-CENT 2400 1451 **************** PROGRAM GROUPIE (VERSION 84-2) *****************2400 1451 UNSHIELDED GROUP AVERAGES USING 640 GROUPS (SAND-II EXTEND) 2400 1451 2400 1451 2400 1451 1 451 21 02400 1451 3 900 116 02400 1451 2400 1 0 2400 0 0 2.40000+ 4 0.0 + 0 0 0 0 02400 3900 0.0 + 0 0.0 + 0 0 0 1 3382400 3900 338 1 2400 3900 5.75000+ 2 8.64800- 1 6.00000+ 2 8.64800- 1 6.30000+ 2 8.64800- 12400 3900 6.60000+ 2 8.64800- 1 6.90000+ 2 8.64800- 1 7.20000+ 2 8.64800- 12400 3900 7.60000+ 2 1.75900+ 0 8.00000+ 2 1.75900+ 0 8.40000+ 2 1.75900+ 02400 3900 8.80000+ 2 1.75900+ 0 9.20000+ 2 1.75900+ 0 9.60000+ 2 2.57100+ 02400 3900 1.00000+ 3 2.57100+ 0 1.05000+ 3 2.57100+ 0 1.10000+ 3 2.57100+ 02400 3900 1.15000+ 3 2.57100+ 0 1.20000+ 3 2.57100+ 0 1.27500+ 3 3.15400+ 02400 3900 1.35000+ 3 3.15400+ 0 1.42500+ 3 3.15400+ 0 1.50000+ 3 3.15400+ 02400 3900 1.60000+ 3 4.06800+ 0 1.70000+ 3 4.06800+ 0 1.80000+ 3 4.06800+ 02400 3900 1.90000+ 3 4.06800+ 0 2.00000+ 3 6.38400+ 0 2.10000+ 3 6.38400+ 02400 3900 2.20000+ 3 6.38400+ 0 2.30000+ 3 6.38400+ 0 2.40000+ 3 6.38400+ 02400 3900 2.55000+ 3 6.38400+ 0 2.70000+ 3 1.28700+ 1 2.80000+ 3 1.28700+ 12400 3900 3.00000+ 3 1.28700+ 1 3.20000+ 3 1.28700+ 1 3.40000+ 3 3.30100+ 12400 3900 3.60000+ 3 3.30100+ 1 3.80000+ 3 3.30100+ 1 4.00000+ 3 3.30100+ 12400 3900 4.25000+ 3 3.30100+ 1 4.50000+ 3 4.25400+ 1 4.75000+ 3 4.25400+ 12400 3900 5.00000+ 3 4.25400+ 1 5.25000+ 3 4.25400+ 1 5.50000+ 3 5.25700+ 12400 3900 5.75000+ 3 5.25700+ 1 6.00000+ 3 5.25700+ 1 6.30000+ 3 5.25700+ 12400 3900 6.60000+ 3 5.25700+ 1 6.90000+ 3 5.25700+ 1 7.20000+ 3 5.51400+ 12400 3900 7.60000+ 3 5.51400+ 1 8.00000+ 3 5.51400+ 1 8.40000+ 3 5.51400+ 12400 3900 8.80000+ 3 5.51400+ 1 9.20000+ 3 2.44900+ 1 9.60000+ 3 2.44900+ 12400 3900 1.00000+ 4 2.44900+ 1 1.05000+ 4 2.44900+ 1 1.10000+ 4 2.44900+ 12400 3900 1.15000+ 4 2.44900+ 1 1.20000+ 4 1.74700+ 1 1.27500+ 4 1.74700+ 12400 3900 1.35000+ 4 1.74700+ 1 1.42500+ 4 1.74700+ 1 1.50000+ 4 1.59100+ 12400 3900 1.60000+ 4 1.59100+ 1 1.70000+ 4 1.59100+ 1 1.80000+ 4 1.59100+ 12400 3900 1.90000+ 4 1.90700+ 1 2.00000+ 4 1.90700+ 1 2.10000+ 4 1.90700+ 12400 3900 2.20000+ 4 1.90700+ 1 2.30000+ 4 1.90700+ 1 2.40000+ 4 1.90700+ 12400 3900 2.55000+ 4 3.34000+ 1 2.70000+ 4 3.34000+ 1 2.80000+ 4 3.34000+ 12400 3900 3.00000+ 4 3.34000+ 1 3.20000+ 4 3.01800+ 1 3.40000+ 4 3.01800+ 12400 3900 3.60000+ 4 3.01800+ 1 3.80000+ 4 3.01800+ 1 4.00000+ 4 1.15100+ 22400 3900 4.25000+ 4 1.15100+ 2 4.50000+ 4 1.15100+ 2 4.75000+ 4 1.15100+ 22400 3900 5.00000+ 4 1.15100+ 2 5.25000+ 4 7.67800+ 1 5.50000+ 4 7.67800+ 12400 3900 5.75000+ 4 7.67800+ 1 6.00000+ 4 7.67800+ 1 6.30000+ 4 7.67800+ 12400 3900 6.60000+ 4 4.44600+ 1 6.90000+ 4 4.44600+ 1 7.20000+ 4 4.44600+ 12400 3900 7.60000+ 4 4.44600+ 1 8.00000+ 4 4.44600+ 1 8.40000+ 4 4.44600+ 12400 3900 8.80000+ 4 3.19900+ 2 9.20000+ 4 3.19900+ 2 9.60000+ 4 3.19900+ 22400 3900 1.00000+ 5 3.19900+ 2 1.05000+ 5 3.19900+ 2 1.10000+ 5 1.42200+ 22400 3900 1.15000+ 5 1.42200+ 2 1.20000+ 5 1.42200+ 2 1.27500+ 5 1.42200+ 22400 3900 1.35000+ 5 3.62100+ 2 1.42500+ 5 3.62100+ 2 1.50000+ 5 3.62100+ 22400 3900 1.60000+ 5 2.43200+ 2 1.70000+ 5 2.43200+ 2 1.80000+ 5 2.43200+ 22400 3900 1.90000+ 5 2.26100+ 2 2.00000+ 5 2.26100+ 2 2.10000+ 5 2.26100+ 22400 3900 2.20000+ 5 2.67500+ 2 2.30000+ 5 2.67500+ 2 2.40000+ 5 2.67500+ 22400 3900 2.55000+ 5 2.04200+ 2 2.70000+ 5 2.04200+ 2 2.80000+ 5 1.84750+ 22400 3900 3.00000+ 5 1.65300+ 2 3.20000+ 5 2.36700+ 2 3.40000+ 5 2.36700+ 22400 3900 3.60000+ 5 2.00900+ 2 3.80000+ 5 2.00900+ 2 4.00000+ 5 4.29200+ 22400 3900 4.25000+ 5 4.29200+ 2 4.50000+ 5 4.05700+ 2 4.75000+ 5 4.05700+ 22400 3900 5.00000+ 5 3.60500+ 2 5.25000+ 5 3.60500+ 2 5.50000+ 5 4.10800+ 22400 3900 5.75000+ 5 4.10800+ 2 6.00000+ 5 3.12500+ 2 6.30000+ 5 3.12500+ 22400 3900 6.60000+ 5 2.87500+ 2 6.90000+ 5 2.87500+ 2 7.20000+ 5 5.22100+ 22400 3900 7.60000+ 5 5.12650+ 2 8.00000+ 5 5.03200+ 2 8.40000+ 5 6.39100+ 22400 3900 8.80000+ 5 6.39100+ 2 9.20000+ 5 5.05300+ 2 9.60000+ 5 5.05300+ 22400 3900 1.00000+ 6 5.60500+ 2 1.10000+ 6 5.60500+ 2 1.20000+ 6 8.21900+ 22400 3900 1.30000+ 6 8.21900+ 2 1.40000+ 6 9.93800+ 2 1.50000+ 6 9.93800+ 22400 3900 1.60000+ 6 9.85500+ 2 1.70000+ 6 9.85500+ 2 1.80000+ 6 9.34200+ 22400 3900 1.90000+ 6 9.34200+ 2 2.00000+ 6 1.17600+ 3 2.10000+ 6 1.17600+ 32400 3900 2.20000+ 6 1.17600+ 3 2.30000+ 6 1.28400+ 3 2.40000+ 6 1.28400+ 32400 3900 2.50000+ 6 1.28400+ 3 2.60000+ 6 1.43100+ 3 2.70000+ 6 1.43100+ 32400 3900 2.80000+ 6 1.43100+ 3 2.90000+ 6 1.41900+ 3 3.00000+ 6 1.41900+ 32400 3900 3.10000+ 6 1.41900+ 3 3.20000+ 6 1.41900+ 3 3.30000+ 6 1.51100+ 32400 3900 3.40000+ 6 1.51100+ 3 3.50000+ 6 1.51100+ 3 3.60000+ 6 1.51100+ 32400 3900 3.70000+ 6 1.61400+ 3 3.80000+ 6 1.61400+ 3 3.90000+ 6 1.61400+ 32400 3900 4.00000+ 6 1.61400+ 3 4.10000+ 6 1.70100+ 3 4.20000+ 6 1.70100+ 32400 3900 4.30000+ 6 1.70100+ 3 4.40000+ 6 1.70100+ 3 4.50000+ 6 1.82700+ 32400 3900 4.60000+ 6 1.82700+ 3 4.70000+ 6 1.82700+ 3 4.80000+ 6 1.82700+ 32400 3900 4.90000+ 6 1.82700+ 3 5.00000+ 6 1.87500+ 3 5.10000+ 6 1.87500+ 32400 3900 5.20000+ 6 1.87500+ 3 5.30000+ 6 1.87500+ 3 5.40000+ 6 1.87500+ 32400 3900 5.50000+ 6 1.92900+ 3 5.60000+ 6 1.92900+ 3 5.70000+ 6 1.92900+ 32400 3900 5.80000+ 6 1.92900+ 3 5.90000+ 6 1.92900+ 3 6.00000+ 6 1.98100+ 32400 3900 6.10000+ 6 1.98100+ 3 6.20000+ 6 1.98100+ 3 6.30000+ 6 1.98100+ 32400 3900 6.40000+ 6 1.98100+ 3 6.50000+ 6 1.98100+ 3 6.60000+ 6 1.98100+ 32400 3900 6.70000+ 6 2.04100+ 3 6.80000+ 6 2.04100+ 3 6.90000+ 6 2.04100+ 32400 3900 7.00000+ 6 2.04100+ 3 7.10000+ 6 2.04100+ 3 7.20000+ 6 2.04100+ 32400 3900 7.30000+ 6 2.04100+ 3 7.40000+ 6 2.11500+ 3 7.50000+ 6 2.11500+ 32400 3900 7.60000+ 6 2.11500+ 3 7.70000+ 6 2.11500+ 3 7.80000+ 6 2.11500+ 32400 3900 7.90000+ 6 2.11500+ 3 8.00000+ 6 2.11500+ 3 8.10000+ 6 2.11500+ 32400 3900 8.20000+ 6 2.17100+ 3 8.30000+ 6 2.17100+ 3 8.40000+ 6 2.17100+ 32400 3900 8.50000+ 6 2.17100+ 3 8.60000+ 6 2.17100+ 3 8.70000+ 6 2.17100+ 32400 3900 8.80000+ 6 2.17100+ 3 8.90000+ 6 2.17100+ 3 9.00000+ 6 2.23600+ 32400 3900 9.10000+ 6 2.23600+ 3 9.20000+ 6 2.23600+ 3 9.30000+ 6 2.23600+ 32400 3900 9.40000+ 6 2.23600+ 3 9.50000+ 6 2.23600+ 3 9.60000+ 6 2.23600+ 32400 3900 9.70000+ 6 2.23600+ 3 9.80000+ 6 2.23600+ 3 9.90000+ 6 2.23600+ 32400 3900 1.00000+ 7 2.34900+ 3 1.01000+ 7 2.34900+ 3 1.02000+ 7 2.34900+ 32400 3900 1.03000+ 7 2.34900+ 3 1.04000+ 7 2.34900+ 3 1.05000+ 7 2.34900+ 32400 3900 1.06000+ 7 2.34900+ 3 1.07000+ 7 2.34900+ 3 1.08000+ 7 2.34900+ 32400 3900 1.09000+ 7 2.34900+ 3 1.10000+ 7 2.47000+ 3 1.11000+ 7 2.47000+ 32400 3900 1.12000+ 7 2.47000+ 3 1.13000+ 7 2.47000+ 3 1.14000+ 7 2.47000+ 32400 3900 1.15000+ 7 2.47000+ 3 1.16000+ 7 2.47000+ 3 1.17000+ 7 2.47000+ 32400 3900 1.18000+ 7 2.47000+ 3 1.19000+ 7 2.47000+ 3 1.20000+ 7 2.64500+ 32400 3900 1.21000+ 7 2.64500+ 3 1.22000+ 7 2.64500+ 3 1.23000+ 7 2.64500+ 32400 3900 1.24000+ 7 2.64500+ 3 1.25000+ 7 2.64500+ 3 1.26000+ 7 2.64500+ 32400 3900 1.27000+ 7 2.64500+ 3 1.28000+ 7 2.64500+ 3 1.29000+ 7 2.64500+ 32400 3900 1.30000+ 7 2.73000+ 3 1.31000+ 7 2.73000+ 3 1.32000+ 7 2.73000+ 32400 3900 1.33000+ 7 2.73000+ 3 1.34000+ 7 2.73000+ 3 1.35000+ 7 2.73000+ 32400 3900 1.36000+ 7 2.73000+ 3 1.37000+ 7 2.73000+ 3 1.38000+ 7 2.73000+ 32400 3900 1.39000+ 7 2.73000+ 3 1.40000+ 7 2.78400+ 3 1.41000+ 7 2.78400+ 32400 3900 1.42000+ 7 2.78400+ 3 1.43000+ 7 2.78400+ 3 1.44000+ 7 2.78400+ 32400 3900 1.45000+ 7 2.78400+ 3 1.46000+ 7 2.78400+ 3 1.47000+ 7 2.78400+ 32400 3900 1.48000+ 7 2.78400+ 3 1.49000+ 7 2.78400+ 3 1.50000+ 7 2.90500+ 32400 3900 1.51000+ 7 2.90500+ 3 1.52000+ 7 2.90500+ 3 1.53000+ 7 2.90500+ 32400 3900 1.54000+ 7 2.90500+ 3 1.55000+ 7 2.90500+ 3 1.56000+ 7 2.90500+ 32400 3900 1.57000+ 7 2.90500+ 3 1.58000+ 7 2.90500+ 3 1.59000+ 7 2.90500+ 32400 3900 1.60000+ 7 2.92900+ 3 1.61000+ 7 2.92900+ 3 1.62000+ 7 2.92900+ 32400 3900 1.63000+ 7 2.92900+ 3 1.64000+ 7 2.92900+ 3 1.65000+ 7 2.92900+ 32400 3900 1.66000+ 7 2.92900+ 3 1.67000+ 7 2.92900+ 3 1.68000+ 7 2.92900+ 32400 3900 1.69000+ 7 2.92900+ 3 1.70000+ 7 2.95400+ 3 1.71000+ 7 2.95400+ 32400 3900 1.72000+ 7 2.95400+ 3 1.73000+ 7 2.95400+ 3 1.74000+ 7 2.95400+ 32400 3900 1.75000+ 7 2.95400+ 3 1.76000+ 7 2.95400+ 3 1.77000+ 7 2.95400+ 32400 3900 1.78000+ 7 2.95400+ 3 1.79000+ 7 2.95400+ 3 1.80000+ 7 2.96500+ 32400 3900 1.81000+ 7 2.96500+ 3 1.82000+ 7 2.96500+ 3 1.83000+ 7 2.96500+ 32400 3900 1.84000+ 7 2.96500+ 3 1.85000+ 7 2.96500+ 3 1.86000+ 7 2.96500+ 32400 3900 1.87000+ 7 2.96500+ 3 1.88000+ 7 2.96500+ 3 1.89000+ 7 2.96500+ 32400 3900 1.90000+ 7 2.98900+ 3 1.91000+ 7 2.98900+ 3 1.92000+ 7 2.98900+ 32400 3900 1.93000+ 7 2.98900+ 3 1.94000+ 7 2.98900+ 3 1.95000+ 7 2.98900+ 32400 3900 1.96000+ 7 2.98900+ 3 1.97000+ 7 2.98900+ 3 1.98000+ 7 2.98900+ 32400 3900 1.99000+ 7 2.98900+ 3 2.00000+ 7 0.0 + 0 2400 3900 2400 3 0 2400 0 0 0 0 0 2.40520E+4 5.14943E+1 0 0 34 102431 1451 0.0 0.0 0 0 0 62431 1451 1.00000E+0 2.00000E+7 0 0 10 20022431 1451 3.00000E+2 0.0 1 0 38 32431 1451 24-Cr- 52 IRK-VIENNA EVAL-APR90 2431 1451 DIST-Feb2004 2431 1451 ----IRDF-2002 MATERIAL 2431 2431 1451 -----INCIDENT NEUTRON DATA 2431 1451 ------ENDF-6 FORMAT 2431 1451 *****************************************************************2431 1451 DIST-JUN90 910201 2431 1451 IRK-EVAL.NLIB 25 2431 2431 1451 *****************************************************************2431 1451 The Q values and threshold energies were updated prior to pro- 2431 1451 cessing through the codes to comply with the values obtained 2431 1451 using the NNDC calculation program which is based on the 1995 2431 1451 Update to the Atomic mass Evaluation. 2431 1451 ***************** Program LINEAR (VERSION 2002-1) ***************2431 1451 For All Data Greater than 1.0000E-10 barns in Absolute Value 2431 1451 Data Linearized to Within an Accuracy of .100000000 per-cent 2431 1451 ***************** Program SIGMA1 (VERSION 2002-1) ***************2431 1451 Data Doppler Broadened to 300.000000 Kelvin 2431 1451 for All Data Greater than 1.0000E-10 barns in Absolute Value 2431 1451 Data Linearized to Within an Accuracy pf .100000000 per-cent 2431 1451 ***************** Program FIXUP (Version 2002-1) ****************2431 1451 Corrected ZA/AWR in All Sections-----------------------------Yes 2431 1451 Corrected Thresholds-----------------------------------------Yes 2431 1451 Extended Cross Sections to 20 MeV----------------------------No 2431 1451 Allow Cross Section Deletion---------------------------------No 2431 1451 Allow Cross Section Reconstruction---------------------------No 2431 1451 Make All Cross Sections Non-Negative-------------------------Yes 2431 1451 Delete Energies Not in Ascending Order-----------------------Yes 2431 1451 Deleted Duplicate Points-------------------------------------Yes 2431 1451 Check for Ascending MAT/MF/MT Order--------------------------Yes 2431 1451 Check for Legal MF/MT Numbers--------------------------------Yes 2431 1451 Allow Creation of Missing Sections---------------------------No 2431 1451 Allow Insertion of Energy Points-----------------------------No 2431 1451 Create Uniform Energy Grid-----------------------------------No 2431 1451 Delete Section if Cross Section =0 at All Energies-----------Yes 2431 1451 ***************** Program GROUPIE (VERSION 2002-1) **************2431 1451 Unshielded Group Averages Using 640 Groups 2431 1451 Weighting Spectrum: Flat (Constant) Spectrum 2431 1451 1 451 45 12431 1451 3 16 30 12431 1451 33 16 26 12431 1451 2431 1 0 2431 0 0 2.40520E+4 5.14943E+1 0 0 0 02431 3 16 -1.20394E+7-1.20394E+7 0 0 1 792431 3 16 79 1 2431 3 16 12200000.0 .000136222 12300000.0 .002913303 12400000.0 .0090350002431 3 16 12500000.0 .018505000 12600000.0 .029945000 12700000.0 .0421942862431 3 16 12800000.0 .054457131 12900000.0 .070114000 13000000.0 .0871593332431 3 16 13100000.0 .104204667 13200000.0 .121073583 13300000.0 .1368840002431 3 16 13400000.0 .152518000 13500000.0 .168152000 13600000.0 .1837860002431 3 16 13700000.0 .200229500 13800000.0 .221530000 13900000.0 .2436400002431 3 16 14000000.0 .265750000 14100000.0 .287860000 14200000.0 .3092252502431 3 16 14300000.0 .326122000 14400000.0 .342274000 14500000.0 .3584260002431 3 16 14600000.0 .374578000 14700000.0 .390333833 14800000.0 .4037126672431 3 16 14900000.0 .416695333 15000000.0 .429678000 15100000.0 .4426606672431 3 16 15200000.0 .455643333 15300000.0 .468626000 15400000.0 .4816086672431 3 16 15500000.0 .491881500 15600000.0 .499444500 15700000.0 .5070075002431 3 16 15800000.0 .514570500 15900000.0 .522133500 16000000.0 .5296965002431 3 16 16100000.0 .537259500 16200000.0 .544822500 16300000.0 .5523855002431 3 16 16400000.0 .559948500 16500000.0 .566084500 16600000.0 .5707935002431 3 16 16700000.0 .575502500 16800000.0 .580211500 16900000.0 .5849205002431 3 16 17000000.0 .589629500 17100000.0 .594338500 17200000.0 .5990475002431 3 16 17300000.0 .603756500 17400000.0 .608465500 17500000.0 .6119980002431 3 16 17600000.0 .614354000 17700000.0 .616710000 17800000.0 .6190660002431 3 16 17900000.0 .621422000 18000000.0 .623778000 18100000.0 .6261340002431 3 16 18200000.0 .628490000 18300000.0 .630846000 18400000.0 .6332020002431 3 16 18500000.0 .635069000 18600000.0 .636447000 18700000.0 .6378250002431 3 16 18800000.0 .639203000 18900000.0 .640581000 19000000.0 .6419590002431 3 16 19100000.0 .643337000 19200000.0 .644715000 19300000.0 .6460930002431 3 16 19400000.0 .647471000 19500000.0 .648849000 19600000.0 .6502270002431 3 16 19700000.0 .651605000 19800000.0 .652983000 19900000.0 .6543610002431 3 16 20000000.0 0.0 2431 3 16 2431 3 0 2431 0 0 2.40520E+4 5.14943E+1 0 0 0 1243133 16 0.000000+0 0.000000+0 0 16 0 1243133 16 0.000000+0 0.000000+0 1 5 136 16243133 16 1.000000-5 1.220000+7 1.245000+7 1.255000+7 1.265000+7 1.275000+7243133 16 1.300000+7 1.350000+7 1.400000+7 1.450000+7 1.500000+7 1.600000+7243133 16 1.700000+7 1.800000+7 1.900000+7 2.000000+7 0.000000+0 0.000000+0243133 16 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0243133 16 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0243133 16 0.000000+0 4.031000-2 3.015000-2 2.011000-2 4.441000-3 1.005000-2243133 16 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0243133 16 0.000000+0 0.000000+0 0.000000+0 2.256000-2 1.504000-2 3.322000-3243133 16 7.519000-3 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0243133 16 0.000000+0 0.000000+0 0.000000+0 0.000000+0 1.003000-2 2.216000-3243133 16 5.015000-3 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0243133 16 0.000000+0 0.000000+0 0.000000+0 0.000000+0 1.088000-3 1.107000-3243133 16 7.104000-4 1.385000-4 3.054000-4 1.269000-4 1.050000-3 6.131000-4243133 16 5.750000-4 5.345000-4 5.072000-4 2.506000-3 0.000000+0 0.000000+0243133 16 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0243133 16 0.000000+0 1.823000-3 2.416000-4 5.329000-4 2.213000-4 1.832000-3243133 16 1.070000-3 1.003000-3 9.325000-4 8.849000-4 4.904000-4 3.163000-4243133 16 3.318000-4 3.571000-4 2.085000-4 1.956000-4 1.818000-4 1.725000-4243133 16 7.342000-4 2.818000-4 7.875000-4 4.599000-4 4.313000-4 4.009000-4243133 16 3.805000-4 4.986000-4 3.270000-4 1.910000-4 1.791000-4 1.665000-4243133 16 1.580000-4 3.980000-3 1.581000-3 1.482000-3 1.378000-3 1.308000-3243133 16 1.358000-3 8.658000-4 8.048000-4 7.637000-4 1.194000-3 7.548000-4243133 16 7.163000-4 1.032000-3 6.658000-4 9.292000-4 243133 16 243133 0 2431 0 0 0 0 0 2.50550E+4 5.44661E+1 0 0 34 102525 1451 0.0 0.0 0 0 0 62525 1451 1.00000E+0 2.00000E+7 0 5 10 20022525 1451 3.00000E+2 0.0 1 0 157 32525 1451 25-Mn- 55 JAERI,ORNL EVAL-MAR88 K.SHIBATA 2525 1451 DIST-Feb2004 2525 1451 ----IRDF-2002 MATERIAL 2525 2525 1451 -----INCIDENT NEUTRON DATA 2525 1451 ------ENDF-6 FORMAT 2525 1451 ******************************************************************2525 1451 25-MN- 55 JAERI,ORNL EVAL-MAR88 K.SHIBATA 2525 1451 DIST-SEP 1 REV2-SEP98 20010926 2525 1451 ----ENDF/B-VI MATERIAL 2525 REVISION 2 2525 1451 *****************EXTRACT FOR SPECIAL PURPOSE FILE*****************2525 1451 DOSIMETRY 2525 1451 ******************************************************************2525 1451 2525 1451 **************************************************************** 2525 1451 2525 1451 ENDF/B-VI MOD 3 Revision, June 2000, S.C. Frankle, R.C. Reedy, 2525 1451 P.G. Young (LANL) 2525 1451 2525 1451 The secondary gamma-ray spectrum for radiative capture (MF 12, 2525 1451 MT 102) has been updated for new experimental data at incident 2525 1451 neutron energies up to 100 keV. 2525 1451 The previous (MOD 2) pure continuum at thermal neutron energy 2525 1451 is replaced by 321 discrete photons. 2525 1451 The Q-value for radiative capture was also updated in File 3. 2525 1451 Details of these changes are described in Frankel et al. [Fr01]. 2525 1451 2525 1451 REFERENCES 2525 1451 2525 1451 [Fr01] S.C. Frankle, R.C. Reedy, and P.G. Young, Los ALamos 2525 1451 National Laboratory Report, LA-13812 (2001). 2525 1451 2525 1451 **************************************************************** 2525 1451 2525 1451 ENDF/B-VI MOD 2 Revision, October 1997, V. McLane (NNDC) 2525 1451 2525 1451 1. Corrected residual nucleus and AWR in File 6, MT=22,107. 2525 1451 2. Updated file 1 comments and and corrected references. 2525 1451 2525 1451 **************************************************************** 2525 1451 2525 1451 ENDF/B-VI MOD 1 Evaluation, March 1988, K. Shibata (ORNL,JAERI) 2525 1451 2525 1451 File 1 General Information ------------------------------------ 2525 1451 MT=451 Descriptive data and dictionary 2525 1451 2525 1451 File 2 Resonance Parameters ----------------------------------- 2525 1451 MT=151 Resolved resonance parameters for MLBW formula. 2525 1451 The parameters of the lowest four resonances were taken 2525 1451 from the work of Macklin [1]. Others were taken from the 2525 1451 compilation of mughabghab et al.[2] except that the 2525 1451 parameters of two negative resonances were adjusted so as 2525 1451 to fit to experimental thermal cross sections. 2525 1451 Resonance region : 1.0E-5 eV to 100 keV. 2525 1451 2525 1451 Scattering radius: 5.15 fm 2525 1451 2525 1451 Calculated 2200-m/s cross sections and res. integrals 2525 1451 2200-M/S RES. INTEG. 2525 1451 Elastic 2.167 b - 2525 1451 Capture 13.413 b 11.73 b 2525 1451 Total 15.579 b - 2525 1451 2525 1451 File 3 Neutron Cross Sections --------------------------------- 2525 1451 MT=1 Total 2525 1451 Below 100 keV : no background 2525 1451 Above 100 keV : based on the experimental data [3,4,5]. 2525 1451 MT=2 Elastic scattering 2525 1451 (total) - (nonelastic cross section). 2525 1451 2525 1451 MT=102 Radiative capture cross section 2525 1451 Below 100 keV: Resonance parameters given (no background). 2525 1451 Above 100 keV: Based on the experimental data [11-15] 2525 1451 (taken from JENDL-2 [20]). 2525 1451 2525 1451 2525 1451 File 33. Covariance data 2525 1451 MT=102 2525 1451 Estimated from experimental data and calculation. 2525 1451 2525 1451 ---------------------------------------------------------------- 2525 1451 REFERENCES 2525 1451 2525 1451 [1] R.L. Macklin, Nucl.Sci.Eng. 89, 362 (1985). 2525 1451 [2] S. F. Mughabghab, M. Divadeenam, and N.E. Holden, Neutron 2525 1451 Cross Sections, Vol. 1, Part A (Academic Press, 1981). 2525 1451 [3] S. Cierjacks, P. Forti, D. Kopsch et al., "High Resolution 2525 1451 Total Cross Sections for Na, Cl, K, V, Mn and Co between 0.5 2525 1451 and 30 MeV", report KFK-1000 (1968). 2525 1451 [4] W.F.E. Pineo, M. Divadeenam, E.G. Bilpuch et al., Ann.Phys. 2525 1451 84, 165 (1974). 2525 1451 [5] J.B. Garg, J. Rainwater, and W.W. Havens Jr., Nucl.Sci.Eng. 2525 1451 65, 76 (1978). 2525 1451 [6] C.Y. Fu, "A Consistent Nuclear Model for Compound and Pre- 2525 1451 compound Reactions with Conservation of Angular Momentum," 2525 1451 Oak Ridge National Laboratory report ORNL/TM-7042 (1980). 2525 1451 [7] C.Y. Fu, private communication (1985). 2525 1451 [8] Zhou Enchen, Huo Junde, Zhou Chunmei et al., Nucl.Data 2525 1451 Sheets 44, 463 (1985). 2525 1451 [9] F.G. Perey, Phys.Rev. 131, 745 (1963). 2525 1451 [10] J.R. Huizenga, and G.J. Igo, Nucl.Phys. 29, 462 (1962). 2525 1451 [11] J.B. Garg, R.L. Macklin, and J. Halperin, Phys.Rev. C18, 2525 1451 2079 (1978). 2525 1451 [12] A.G. Dovbenko, V.E. Kolesov, V.P. Koroleva et al., At.En. 26,2525 1451 67 (1969). 2525 1451 [13] H.O. Menlove, K.L. Coop, H.A. Grench et al. Phys.Rev. 163, 2525 1451 1299 (1967). 2525 1451 [14] O. Schwerer, M. Winkler-Rohatsch, H. Warhanek et al., Nucl. 2525 1451 Phys. A264, 105 (1976). 2525 1451 [15] M. Budnar, F. Cvelbar, E. Hodgson et al., "Prompt Gamma-ray 2525 1451 Spectra and Integarted Cross Sections for the Radiative 2525 1451 Capture of 14 MeV Neutrons for 28 Natural Targets ...," 2525 1451 report INDC(YUG)-6 (1979). 2525 1451 [16] L. Colli, I. Iori, S. Micheletti et al., Nuovo.cim. 21, 966 2525 1451 (1962). 2525 1451 [17] S. Sudar and Csikai, Nucl.Phys. A319, 157 (1979). 2525 1451 [18] M. Diksic, P. Strohal, and I. Slaus, J.Inorg.Nucl.Chem. 36, 2525 1451 477 (1974). 2525 1451 [19] C.H. Wu, R. Woelfle, and S.M. Qaim, Nucl.Phys. A329, 63 2525 1451 (1979). 2525 1451 [20] T. Nakagawa, "Summary of JENDL-2 General Purpose File," 2525 1451 report JAERI-M 84-103 (1984). 2525 1451 2525 1451 ******************************************************************2525 1451 2525 1451 2525 1451 2525 1451 2525 1451 ******************************************************************2525 1451 ************************ C O N T E N T S *********************** 2525 1451 ***************** Program LINEAR (VERSION 2002-1) ***************2525 1451 For All Data Greater than 1.0000E-10 barns in Absolute Value 2525 1451 Data Linearized to Within an Accuracy of .100000000 per-cent 2525 1451 ***************** Program RECENT (VERSION 2002-1) ***************2525 1451 for All Data Greater than 1.0000E-10 barns in Absolute Value 2525 1451 Data Linearized to within an Accuracy of .100000000 per-cent 2525 1451 ***************** Program SIGMA1 (VERSION 2002-1) ***************2525 1451 Data Doppler Broadened to 300.000000 Kelvin 2525 1451 for All Data Greater than 1.0000E-10 barns in Absolute Value 2525 1451 Data Linearized to Within an Accuracy pf .100000000 per-cent 2525 1451 ***************** Program FIXUP (Version 2002-1) ****************2525 1451 Corrected ZA/AWR in All Sections-----------------------------Yes 2525 1451 Corrected Thresholds-----------------------------------------Yes 2525 1451 Extended Cross Sections to 20 MeV----------------------------No 2525 1451 Allow Cross Section Deletion---------------------------------No 2525 1451 Allow Cross Section Reconstruction---------------------------No 2525 1451 Make All Cross Sections Non-Negative-------------------------Yes 2525 1451 Delete Energies Not in Ascending Order-----------------------Yes 2525 1451 Deleted Duplicate Points-------------------------------------Yes 2525 1451 Check for Ascending MAT/MF/MT Order--------------------------Yes 2525 1451 Check for Legal MF/MT Numbers--------------------------------Yes 2525 1451 Allow Creation of Missing Sections---------------------------No 2525 1451 Allow Insertion of Energy Points-----------------------------No 2525 1451 Create Uniform Energy Grid-----------------------------------No 2525 1451 Delete Section if Cross Section =0 at All Energies-----------Yes 2525 1451 ***************** Program GROUPIE (VERSION 2002-1) **************2525 1451 Unshielded Group Averages Using 640 Groups 2525 1451 Weighting Spectrum: Flat (Constant) Spectrum 2525 1451 1 451 164 32525 1451 3 102 217 32525 1451 33 102 17 12525 1451 2525 1 0 2525 0 0 2.50550E+4 5.44661E+1 0 0 0 02525 3102 7.27050E+6 7.27050E+6 0 0 1 6412525 3102 641 1 2525 3102 .000100000 210.836355 .000105000 205.919990 .000110000 201.2493862525 3102 .000115000 196.945550 .000120000 191.914305 .000127500 186.3322702525 3102 .000135000 181.232244 .000142500 176.545422 .000150000 171.4729812525 3102 .000160000 166.188822 .000170000 161.380891 .000180000 156.9675742525 3102 .000190000 152.861356 .000200000 149.112342 .000210000 145.5851502525 3102 .000220000 142.308952 .000230000 139.265033 .000240000 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DIST-JAN82 811211 2600 1451 ----IRDF-90/NMF-G MATERIAL 2600 2600 1451 -----INCIDENT NEUTRON DATA 2600 1451 ------ENDF-6 FORMAT 2600 1451 DAMAGE CROSS SECTION OF IRON IN STEEL 2600 1451 E693 ASTM DPA FOR IRON MT=900 (FROM P.GRIFFIN 2002) 2600 1451 DAMAGE ENERGY TO DISPLACEMENT CONVERSION CONSISTENT 2600 1451 WITH ASTM E521 AND E281 2600 1451 EURATOM STANDARD MT=901 2600 1451 ***************** PROGRAM LINEAR (VERSION 84-2) *****************2600 1451 DATA LINEARIZED TO WITHIN AN ACCURACY OF 0.100 PER-CENT 2600 1451 **************** PROGRAM GROUPIE (VERSION 84-2) *****************2600 1451 UNSHIELDED GROUP AVERAGES USING 640 GROUPS (SAND-II EXTEND) 2600 1451 2600 1451 2600 1451 2600 1451 1 451 24 02600 1451 3 900 217 02600 1451 3 901 217 02600 1451 2600 1 0 2600 0 0 2.60000+ 4 0.0 + 0 0 0 0 02600 3900 0.0 + 0 0.0 + 0 0 0 1 6412600 3900 641 1 2600 3900 1.0000E-04 1.5835E+02 1.0500E-04 1.5462E+02 1.1000E-04 1.5114E+022600 3900 1.1500E-04 1.4789E+02 1.2000E-04 1.4411E+02 1.2750E-04 1.3992E+022600 3900 1.3500E-04 1.3609E+02 1.4250E-04 1.3254E+02 1.5000E-04 1.2875E+022600 3900 1.6000E-04 1.2479E+02 1.7000E-04 1.2117E+02 1.8000E-04 1.1785E+022600 3900 1.9000E-04 1.1481E+02 2.0000E-04 1.1196E+02 2.1000E-04 1.0932E+022600 3900 2.2000E-04 1.0686E+02 2.3000E-04 1.0457E+02 2.4000E-04 1.0189E+022600 3900 2.5500E-04 9.8933E+01 2.7000E-04 9.6660E+01 2.8000E-04 9.4127E+012600 3900 3.0000E-04 9.1052E+01 3.2000E-04 8.8249E+01 3.4000E-04 8.5688E+012600 3900 3.6000E-04 8.3339E+01 3.8000E-04 8.1173E+01 4.0000E-04 7.8925E+012600 3900 4.2500E-04 7.6636E+01 4.5000E-04 7.4537E+01 4.7500E-04 7.2599E+012600 3900 5.0000E-04 7.0818E+01 5.2500E-04 6.9148E+01 5.5000E-04 6.7592E+012600 3900 5.7500E-04 6.6138E+01 6.0000E-04 6.4642E+01 6.3000E-04 6.3120E+012600 3900 6.6000E-04 6.1702E+01 6.9000E-04 6.0373E+01 7.2000E-04 5.8927E+012600 3900 7.6000E-04 5.7397E+01 8.0000E-04 5.5979E+01 8.4000E-04 5.4660E+012600 3900 8.8000E-04 5.3432E+01 9.2000E-04 5.2287E+01 9.6000E-04 5.1215E+012600 3900 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1.98000+ 7 2.14889+ 32600 3901 1.99000+ 7 2.14889+ 3 2.00000+ 7 0.0 + 0 2600 3901 2600 3 0 2600 0 0 0 0 0 2.60540E+4 5.34762E+1 0 0 34 102625 1451 0.0 0.0 0 0 0 62625 1451 1.00000E+0 2.00000E+7 0 0 10 20022625 1451 3.00000E+2 0.0 1 0 527 72625 1451 26-Fe- 54 FEI/LANL EVAL-Jun01 K.I.Zolotarev,M.Chadwick et al. 2625 1451 DIST-Feb2004 2625 1451 ----IRDF-2002 MATERIAL 2625 2625 1451 -----INCIDENT NEUTRON DATA 2625 1451 ------ENDF-6 FORMAT 2625 1451 ***************************************************************** 2625 1451 26-Fe- 54 FEI EVAL-Jun01 K.I.Zolotarev 2625 1451 DIST-May03 Mod1 20030512 2625 1451 ----BROND-2 MATERIAL 2625 REVISION 1 2625 1451 -----INCIDENT NEUTRON DATA 2625 1451 ------ENDF-6 FORMAT 2625 1451 ------Russian Reactor Dosimetry File RRDF-2002 2625 1451 ***************************************************************** 2625 1451 Authors of evaluation: K.I.Zolotarev and A.B.Pashchenko 2625 1451 MF=3 2625 1451 MT= 16 - (n,2n) cross section 2625 1451 MT=107 - (n,a) cross section 2625 1451 ***************************************************************** 2625 1451 *****************EXTRACT FOR SPECIAL PURPOSE FILE*****************2625 1451 DOSIMETRY 2625 1451 ******************************************************************2625 1451 26-Fe- 54 LANL,ORNL EVAL-SEP96 M.B.CHADWICK,P.G.YOUNG,D.HETRICK 2625 1451 Ch99,He91,Fu91 DIST-SEP 1 REV4- 20010926 2625 1451 ----ENDF/B-VI MATERIAL 2625 REVISION 4 2625 1451 MF=3 2625 1451 MT= 103 - (n,p) cross section 2625 1451 ******************************************************************2625 1451 ******** Start of (N,2N), (N,A) bibliographical component ******* 2625 1451 ***************************************************************** 2625 1451 ------Russian Reactor Dosimetry File RRDF-2002 2625 1451 ***************************************************************** 2625 1451 Authors of evaluation: K.I.Zolotarev and A.B.Pashchenko 2625 1451 MF=3 2625 1451 MT= 16 - (n,2n) cross section 2625 1451 ------------------------------------- 2625 1451 Excitation function for the Fe-54(n,2n)Fe-53m+g reaction in 2625 1451 the energy region from threshold to 21 MeV was evaluated by means 2625 1451 of statistical analysis of experimental cross section data [1-16] 2625 1451 and data from GNASH [31-32] calculation. 2625 1451 All experimental data if it was possible were renormalized to 2625 1451 the new standards for monitor reactions cross sections and decay 2625 1451 data. Decay data for the iron-53 were taken from ref. [17]. Recom-2625 1451 mended cross sections for the reaction Al-27(n,p)Mg-27 used as a 2625 1451 monitor in the measurements [7-8], [11-14] and [16] were taken 2625 1451 from new evaluation [18]. 2625 1451 Special correction was done with experimental data [1], [6]. 2625 1451 Data of Terrel and Holm [1] measured at 16.89 and 17.89 MeV 2625 1451 have been corrected introducing the Fc= 0.5 coefficient, because 2625 1451 of authors used double decreased value for positron yield. 2625 1451 Data of Andreev et al.[6] were renormalized using preliminary 2625 1451 evaluated integral of cross section value in the neutron energy 2625 1451 range from 15.0 to 16.5 MeV. Renormalised experimental data of 2625 1451 Andreev et al. are well consistent with Ryves measurements [9-10] 2625 1451 and with new experimental data for Fe-54(n,2n)Fe-53m+g reaction 2625 1451 obtained by Fessler [16]. 2625 1451 Experimental data from ref.[19-27] were rejected due to their 2625 1451 discrepancy with the main bulk of experimental data [1-16] and 2625 1451 data from theoretical model calculation. 2625 1451 The final procedure of evaluation Fe54(n,2n)Fe53m+g excitati- 2625 1451 on function from threshold to 21 MeV has been carried out within 2625 1451 the framework of generalized least squares method. Rational func- 2625 1451 tion was used as model function [28]. 2625 1451 U-235 thermal fission [29] and Cf-252 spontaneous fission 2625 1451 neutron spectra [30] averaged cross sections calculated from the 2625 1451 evaluated Fe54(n,2n)Fe53m+g excitation function are the following:2625 1451 2625 1451 ---------------------------------------------- 2625 1451 TYPE OF SPECTRUM I , mb (calc.) 2625 1451 ---------------------------I------------------ 2625 1451 U-235 neutron fission I 1.2839E-3 2625 1451 CF-252 spontan. fission I 3.6219E-3 2625 1451 ---------------------------------------------- 2625 1451 2625 1451 2625 1451 MT=107 - (n,a) cross section 2625 1451 ------------------------------------- 2625 1451 Excitation function for the Fe-54(n,a)Cr-51 reaction in the 2625 1451 energy region from 3 MeV to 20 MeV was evaluated by means of 2625 1451 statistical analysis of experimental cross section data [3], [4], 2625 1451 [11], [15], [21], [34-49] and data from GNASH [31-32] calculation.2625 1451 All experimental data were renormalized to the new standards 2625 1451 for monitor reactions cross sections and decay data. Experimental 2625 1451 data of S.R.Salisbury and R.A.Chalmers [5] were rejected due to 2625 1451 their inconsistency with theoretical model calculations data in 2625 1451 the energy region 3-5 MeV. Data of Y.M.Gledenov et al. [50] at 2625 1451 7 MeV were rejected due to their very big discrepancy with measu- 2625 1451 rements [39,46,49] and data from theoretical model calculation. 2625 1451 The final procedure of evaluation Fe-54(n,a)Cr-51 excitation 2625 1451 function from 3 MeV to 20 MeV have been carried out within the 2625 1451 framework of generalized least squares method. Rational function 2625 1451 was used as a model function [28]. 2625 1451 Integral experimental data for U-235 neutron fission spectrum 2625 1451 [51-53] was used for testing evaluated Fe-54(n,a)Cr-51 excitation 2625 1451 function. The results of testing are given in Table 1. 2625 1451 Data for U-235 thermal fission neutron spectrum and Cf-252 2625 1451 spontaneous fission neutron spectrum were taken from ref.[29] and 2625 1451 [30], respectively. 2625 1451 Table 1. 2625 1451 --------------------------------------------------------------- 2625 1451 TYPE OF SPECTRUM I ,MB (calc.) I , MB (measured) 2625 1451 ----------------------I-----------------I---------------------- 2625 1451 I I 2625 1451 U-235 neutron fission I 0.8459 I 0.790 +- 0.119 [51] 2625 1451 I I 0.850 +- 0.100 [52] 2625 1451 I I 0.891 +- 0.070 [53] 2625 1451 I I 0.850 +- 0.050 [* ] 2625 1451 I I 2625 1451 ----------------------I-----------------I---------------------- 2625 1451 I I 2625 1451 CF-252 spont. fission I 1.1114 I 2625 1451 I I 2625 1451 --------------------------------------------------------------- 2625 1451 [*] - Averaged cross section value obtained from ref. [51-53] 2625 1451 2625 1451 MF=33 2625 1451 MT= 16 - (n,2n) cross section cov. matrix 2625 1451 ------------------------------------------ 2625 1451 Uncertainties in the evaluated excitation function for the 2625 1451 reaction Fe-54(n,2n)Fe-53m+g are given in the form of relative co-2625 1451 variance matrix for the 12-neutron energy groups (LB=5). Covari- 2625 1451 ance matrix of uncertainties was calculated simultaneously with 2625 1451 recommended cross section data by means of PADE-2 code [33]. 2625 1451 Eigenvalues of the 6-th digits relative covariance matrix 2625 1451 given in the 33-file are the following: 2625 1451 2625 1451 3.65349E-06 4.84130E-06 6.69851E-06 1.07852E-05 2625 1451 2.27150E-05 7.28051E-05 4.72794E-04 5.38548E-03 2625 1451 1.13987E-02 1.32443E-02 3.40784E-02 1.48451E-01 2625 1451 2625 1451 MT=107 - (n,a) cross section cov. matrix 2625 1451 ------------------------------------------ 2625 1451 Uncertainties in the evaluated excitation function for the 2625 1451 reaction Fe-54(n,a)Cr-51 are given in the form of relative covari-2625 1451 ance matrix for the 17-neutron energy groups (LB=5). Covariance 2625 1451 matrix of uncertainties was calculated simultaneously with 2625 1451 recommended cross section data by means of PADE-2 code [33]. 2625 1451 Eigenvalues of the 6-th digits relative covariance matrix 2625 1451 given in the 33-file are the following: 2625 1451 2625 1451 1.85377E-09 2.23510E-09 2.38367E-09 2.64130E-09 2625 1451 3.38901E-09 3.88584E-09 5.57195E-09 9.37333E-09 2625 1451 2.56042E-08 1.44977E-06 8.05497E-04 1.66197E-03 2625 1451 2.03640E-03 2.27545E-03 4.21154E-03 1.42809E-02 2625 1451 2.58358E-02 2625 1451 2625 1451 References : 2625 1451 1. J.Terrell, D.M.Holm Physical Review, v.109, p.2031, 1958 2625 1451 2. M.J.Depraz et al. Journal de Physique, v.21, p.377, May 1960 2625 1451 3. D.M.Chittenden et al. Physical Review, v.122, p.860, 1961 2625 1451 4. W.G.Cross et al. Prog. Report EANDC(CAN)-16, p., January 1963 2625 1451 3. W.G.Cross et al. Prog. Report EANDC(CAN)-16, p.1, Jan. 1963 2625 1451 5. S.R.Salisbury, R.A.Chalmers Phys. Rev., v.B140, p.305, 1965 2625 1451 6. M.F.Andreev, V.I.Serov Jadernaja Fizika (Sov.), v.7, n.4, 2625 1451 p.745, April 1968 2625 1451 7. S.M.Qaim Nuclear Physics, v.A185, p.614, May 1972 2625 1451 8. R.A.Sigg, P.K.Kuroda Inorg. Nucl. Chem., v.37, p.631, 1975 2625 1451 9. T.B.Ryves et al. J. Metrologia, v.14, n.3, p.127, June 1978 2625 1451 10. T.B.Ryves et al. J. of Physics, pt.G, v.4, p.1783, 1978 2625 1451 11. B.M.Bahal, R.Pepelnik Report GKSS-84-E-, 1984 ; 2625 1451 B.M.Bahal, R.Pepelnik Progress Report NEANDC(E)-252/U,(5), 2625 1451 p.28, June 1984 2625 1451 12. L.R.Greenwood, R.K.Smither Proc. of Intern. Conf. on Nuclear 2625 1451 Data for Basic and Applied Science, Santa Fe, New Mexico, USA,2625 1451 13-17 May 1985, v.1, p.163 2625 1451 13. T.Katoh et al. Report JAERI-M-89-083, 1989 2625 1451 14. M.Viennot et al. Nucl. Sci. Eng., v.108, p.289, July 1991 2625 1451 15. A.Ercan et al. Proc. of International Conference on Nuclear 2625 1451 Data for Science and Technology, 13-17 May 1991, Julich, FRG, 2625 1451 Springer-Verlag, 1992, p.376-377 2625 1451 16. A.Fessler Report JUL-3502, FZ Julich GmbH, Germany, 1998 2625 1451 17. E.Browne, R.B.Firestone Table of Radioactive Isotopes, 2625 1451 John Wiley & Sons, New York, 1986 ; 2625 1451 R.B.Firestone Table of Isotopes, Eighth edition, Vol. 1, 2625 1451 John Wiley & Sons, Inc., New York, 1995 2625 1451 18. K.I.Zolotarev Report INDC(CCP)-431, Distr.: J+R/EL, IAEA, 2625 1451 Vienna, August 2002 2625 1451 19. D.L.Allan J. Proc. Phys. Soc., v.70, p.195, 1956 2625 1451 20. L.A.Rayburn Phys. Rev., v.122, p.168, 1961 2625 1451 21. H.Pollehn, H.Neuert Zeitschrift f. Naturforschung, sect.A, 2625 1451 v.16, p.227, 1961 2625 1451 22. C.Carles Comptes Rendus, v.257, p.659, July 1963 2625 1451 23. J.Csikai Report EANDC-50S, v.2, p.102, July 1965 2625 1451 24. J.E.Strain, W.J.Ross Report ORNL-3672, January 1965 2625 1451 25. J.Araminowicz, J.Dresler Rep.INR-1464, Swierk-Warsew,May 1973 2625 1451 26. M.Bormann et al. Zeitsch. f. Physik, sect.A, v.277, p.203, 2625 1451 June 1976 2625 1451 27. Zhou Muyao et al. Chinese J. of Nucl. Phys., v.9, p.34, 1987 2625 1451 28. S.Badikov, N.Rabotnov, K.Zolotarev Proc. of NEANSC Speciali- 2625 1451 st's Meeting on Evaluation and Processing of Covariance Data, 2625 1451 Oak Ridge, USA, 7-9 September 1992, OECD, Paris, 1993, p.105 2625 1451 29. L.W.Weston et al. Evaluated Neutron Data for U-235, ENDF/B-VI 2625 1451 Library, MAT-9228, MF=5, MT=18, eval. April 1989 2625 1451 30. W.Mannhart Report IAEA-TECDOC-410, p.158, 1987 2625 1451 31. P.G.Young, E.D.Arthur A Preequilibrium Statistical Nuclear 2625 1451 Model Code for Calculation of Cross Section and Emission 2625 1451 Spectra. Report LA-6947, Los Alamos, 1977 2625 1451 32. E.L.Trykov, G.Ya.Tertychnyi Private communication, IPPE, 2625 1451 Obninsk, May 1999 2625 1451 33. S.A.Badikov et al. Preprint FEI-1686, Obninsk, 1985 2625 1451 34. P.Venugopala Rao, R.W.Fink Phys. Rev., v.154, p.1023, 1967 2625 1451 35. S.M.Qaim et al. Conf. on Chemical Nuclear Data, Measurements 2625 1451 and Applicat., Univ. of Kent, Canterbury,20-22 September 1971,2625 1451 p.121 2625 1451 36. J.J.Singh Trans. Amer. Nucl. Soc., v.15, p.147, June 1972 2625 1451 37. G.N.Maslov et al. Sov. J. Yadernye Konstanty, v.9, p.50, 1972 2625 1451 38. K.Fukuda et al. Prog. Report NEANDC(J)-56/U, p.44, Sep. 1978 2625 1451 39. A.Paulsen et al. Nucl. Sci. Eng., v.72, v.1, p.113, 1979 2625 1451 40. O.I.Artem'ev et al. Atomnaja Energija (Sov.), v.49, no.3, 2625 1451 p.195, September 1980 2625 1451 41. L.R.Greenwood Progress Report DOE-ER-0046-21, p.15, May 1985 2625 1451 42. J.W.Meadows et al. Annals of Nuclear Energy, v.14, p.489, 2625 1451 September 1987 2625 1451 43. Y.Ikeda, C.Konno, K.Oishi et al. Report JAERI-1312, 1988 2625 1451 44. Lu Hanlin et al. Report INDC(CPR)-16, August 1989 2625 1451 45. S.K.Saraf et al. Nucl. Sci. Eng., v.107, p.365, April 1991 2625 1451 46. J.W.Meadows, W.Mannhart et.al. Proc. of Intern. Conference 2625 1451 on Nuclear Data for Science and Technology, 13-17 May 1991, 2625 1451 Julich, FRG, Springer-Verlag, 1992, p.288-290 2625 1451 47. A.Grallert et al. Report INDC(NDS)-286, p.193, IAEA, 1993 2625 1451 48. J.W.Meadows et al. Annals of Nuclear Energy, v.23, p.877, 2625 1451 July 1996 2625 1451 49. Lu Hanlin et al. Report INDC(CPR)-045, IAEA, Vienna, 1998 2625 1451 50. Y.M.Gledenov et al. Report INDC(CPR)-042, IAEA, Vienna, 1997 2625 1451 51. H.Braun, L.Nagy Radiochimica Acta, v.10, p.15, 1968 2625 1451 52. A.V.Bushuev et al. Atomnaja Energija (Sov.), v.63, no.3, 2625 1451 p.207, September 1987 2625 1451 53. O.Horibe et al. Proc. of Conference: 50 Years with Nuclear 2625 1451 Fission, Washington D.C., 25-28 April 1989, v.2, p.923 2625 1451 ******************************************************************2625 1451 ******** End of (N,2N), (N,A) bibliographical component ******** 2625 1451 ***************************************************************** 2625 1451 ******************************************************************2625 1451 ********* Start of (N,P) bibliographical component ********* 2625 1451 ***************************************************************** 2625 1451 2625 1451 ENDF/B-VI MOD 5 Revision, May 2000, S.C. Frankle, R.C. Reedy, 2625 1451 P.G. Young (LANL) 2625 1451 2625 1451 The secondary gamma-ray spectrum for radiative capture (MF 12, 2625 1451 MT 102) has been updated for new experimental data at incident 2625 1451 neutron energies up to 1 keV. 2625 1451 The MF=12, MT=102 yields above 1 keV were adjusted slightly to 2625 1451 force energy conservation. 2625 1451 The Q-value for radiative capture was also updated in File 3. 2625 1451 Details of these changes are described in Frankel et al. [Fr01]. 2625 1451 2625 1451 ******************************************************************2625 1451 2625 1451 ENDF/B-VI MOD 4 Evaluation, September 1997, M.B. Chadwick, 2625 1451 P.G. Young (LANL) and A.J. Koning (ECN) 2625 1451 2625 1451 Los Alamos LA150 Library, produced with FKK/GNASH/GSCAN code 2625 1451 in cooperation with ECN Petten. 2625 1451 2625 1451 This evaluation provides a complete representation of the 2625 1451 nuclear data needed for transport, damage, heating, 2625 1451 radioactivity, and shielding applications over the incident 2625 1451 neutron energy range from 1.0E-11 to 150 MeV. The discussion 2625 1451 here is divided into the region below and above 20 MeV. 2625 1451 2625 1451 INCIDENT NEUTRON ENERGIES < 20 MeV 2625 1451 2625 1451 Below 20 MeV the evaluation is based completely on the ENDF/B- 2625 1451 VI.1 (Release 1) evaluation by [He91] (see also [Fu86]). The 2625 1451 following modifications were made to the ENDF/B-VI.1 evaluation: 2625 1451 2625 1451 1. The covariance files (MF=33) were removed from the file. 2625 1451 2625 1451 2. The derived MF=3 files for MT=203,205,207 were removed. 2625 1451 2625 1451 2625 1451 INCIDENT NEUTRON ENERGIES > 20 MeV 2625 1451 2625 1451 The evaluation above 20 MeV utilizes MF=6, MT=5 to represent 2625 1451 all reaction data. Production cross sections and emission 2625 1451 spectra are given for neutrons, protons, deuterons, tritons, 2625 1451 alpha particles, gamma rays, and all residual nuclides produced 2625 1451 (A>5) in the reaction chains. To summarize, the ENDF sections 2625 1451 with non-zero data above En = 20 MeV are: 2625 1451 2625 1451 MF=3 MT= 1 Total Cross Section 2625 1451 MT= 2 Elastic Scattering Cross Section 2625 1451 MT= 3 Nonelastic Cross Section 2625 1451 MT= 5 Sum of Binary (n,n') and (n,x) Reactions 2625 1451 2625 1451 MF=4 MT= 2 Elastic Angular Distributions 2625 1451 2625 1451 MF=6 MT= 5 Production Cross Sections and Energy-Angle 2625 1451 Distributions for Emission Neutrons, Protons, 2625 1451 Deuterons, and Alphas; and Angle-Integrated 2625 1451 Spectra for Gamma Rays and Residual Nuclei That 2625 1451 Are Stable Against Particle Emission 2625 1451 2625 1451 The evaluation is based on nuclear model calculations that 2625 1451 have been benchmarked to experimental data, especially for n + 2625 1451 Fe56 and p + Fe56 reactions [Ch96a]. We use the GNASH code 2625 1451 system [Yo92], which utilizes Hauser-Feshbach statistical, 2625 1451 preequilibrium and direct-reaction theories. Spherical optical 2625 1451 model calculations are used to obtain particle transmission 2625 1451 coefficients for the Hauser-Feshbach calculations, as well as 2625 1451 for the elastic neutron angular distributions. 2625 1451 Cross sections and spectra for producing individual residual 2625 1451 nuclei are included for reactions that exceed a cross section of 2625 1451 approximately 1 nb at any energy. The energy-angle-correlations 2625 1451 for all outgoing particles are based on Kalbach systematics 2625 1451 [Ka88]. 2625 1451 A model was developed to calculate the energy distributions of 2625 1451 all recoil nuclei in the GNASH calculations [Ch96b]. The recoil 2625 1451 energy distributions are represented in the laboratory system in 2625 1451 MT=5, MF=6, and are given as isotropic in the lab system. Note 2625 1451 that all other data in MT=5,MF=6 are given in the center-of-mass 2625 1451 system. This method of representation requires a modification of 2625 1451 the original ENDF-6 format. 2625 1451 Preequilibrium corrections were performed in the course of the 2625 1451 GNASH calculations using the exciton model of Kalbach [Ka77, 2625 1451 Ka85], validated by comparison with calculations using Feshbach, 2625 1451 Kerman, Koonin (FKK) theory [Ch93]. Discrete level data from 2625 1451 nuclear data sheets were matched to continuum level densities 2625 1451 using the formulation of Ignatyuk [Ig75] and pairing and shell 2625 1451 parameters from the Cook [Co67] analysis. Neutron and charged- 2625 1451 particle transmission coefficients were obtained from the 2625 1451 optical potentials, as discussed below. Gamma-ray transmission 2625 1451 coefficients were calculated using the Kopecky-Uhl model [Ko90]. 2625 1451 The neutron optical model potential of Arthur et al. [Ar80] 2625 1451 was used to calculate transmission coefficients and cross 2625 1451 sections with the SCAT2 code [Be92] up to a neutron energy of 26 2625 1451 MeV. Between 26 and 52 MeV, the imaginary volume component of 2625 1451 Arthur's potential was modified to better account for nonelastic 2625 1451 cross section measurements, and above 52 MeV the Semmering 2625 1451 potential of Madland [Ma88] was used. For protons, the 2625 1451 Beccetti-Greenlees potential [Be69] was utilized below 28 MeV, 2625 1451 and the Madland potential [Ma88] was used at higher energies. 2625 1451 The global spherical potential of Perey [Pe63] was utilized for 2625 1451 deuterons, and the potential of Beccetti-Greenlees [Be71] 2625 1451 was adopted for tritons. Finally, the alpha potential of Lemos 2625 1451 [Le72], as adapted by Arthur et al. [Ar80], was used for alpha 2625 1451 particles. 2625 1451 Direct reaction cross sections to discrete states were 2625 1451 calculated with the ECIS95 code [Re95] using deformation 2625 1451 parameters compiled in Nuclear Data Sheets. 2625 1451 We used the same values for the total, elastic, and nonelastic 2625 1451 cross sections (MF=3, MT=1,2,3) above 20 MeV as were used in the 2625 1451 n + 56Fe evaluation. 2625 1451 2625 1451 ***************************************************************** 2625 1451 2625 1451 REFERENCES 2625 1451 2625 1451 [Ar80] E.D. Arthur and P.G. Young, 'Evaluation of Neutron 2625 1451 Cross Sections to 40 MeV for 54,56Fe," Proc. Sym. on Neutron 2625 1451 Cross Sections from 10 to 50 MeV, 12-14 May 1980, Brookhaven 2625 1451 National Laboratory [Eds. M. R. Bhat and S. Pearlstein, BNL- 2625 1451 NCS-51245, 1980] p. 731. 2625 1451 [Be69] F.D. Becchetti, Jr., and G.W. Greenlees, Phys.Rev. 182 2625 1451 1190 (1969) 2625 1451 [Be71] F.D. Becchetti, Jr., and G.W. Greenlees, "Polarization 2625 1451 Phenomena in Nuclear Reactions," (Ed: H.H. Barschall and 2625 1451 W. Haeberli, The University of Wisconsin Press, 1971) p.682 2625 1451 [Be92] O. Bersillon, "SCAT2 - A Spherical Optical Model Code," 2625 1451 in Proc. ICTP Workshop on Computation and Analysis of Nuclear 2625 1451 Data Relevant to Nuclear Energy and Safety, 10 February-13 2625 1451 March, 1992, Trieste, Italy, to be published in World Scientific2625 1451 Press, and Progress Report of the Nuclear Physics Division, 2625 1451 Bruyeres-le-Chatel 1977, CEA-N-2037, p.111 (1978). 2625 1451 [Ch93] M.B. Chadwick and P.G. Young, Phys.Rev. C 47, 2255 (1993) 2625 1451 [Ch96a] M.B. Chadwick and P.G. Young, "GNASH Calculations of 2625 1451 n,p + 54,56,57,58Fe and Benchmarking of Results" in APT 2625 1451 progress report: 1 August - 1 September 1996, internal Los 2625 1451 Alamos National Laboratory memo T-2-96/MS-52, 6 Aug. 1996 from 2625 1451 R.E. MacFarlane to L. Waters. 2625 1451 [Ch96b] M.B. Chadwick, P.G. Young, R.E. MacFarlane, and A.J. 2625 1451 Koning, "High-Energy Nuclear Data Libraries for Accelerator- 2625 1451 Driven Technologies: Calculational Method for Heavy Recoils," 2625 1451 Proc. of 2nd Int. Conf. on Accelerator Driven Transmutation 2625 1451 Technology and Applications, Kalmar, Sweden, 3-7 June 1996 2625 1451 [Ch99] M.B. Chadwick, P G. Young, G. M. Hale, et al., Los Alamos 2625 1451 National Laboratory report, LA-UR-99-1222 (1999) 2625 1451 [Co67] J.L. Cook, H. Ferguson, and A.R. deL Musgrove, Aust.J. 2625 1451 Phys. 20, 477 (1967) 2625 1451 [Fr01] S.C. Frankle, R.C. Reedy, and P.G. Young, Los ALamos 2625 1451 National Laboratory Report, LA-13812 (2001). 2625 1451 [Fu86] C.Y. Fu and D.M. Hetrick, report ORNL/TM-9964 [ENDF-341] 2625 1451 (1986) 2625 1451 [He91] D.M. Hetrick, C.Y. Fu, and N.M. Larson, ENDF/B-VI.1 2625 1451 Evaluation of n + 54Fe, personal comm. (1991) 2625 1451 [Ig75] A.V. Ignatyuk, G.N. Smirenkin, and A.S. Tishin, Sov.J. 2625 1451 Nucl.Phys. 21, 255 (1975); translation of Yad.Fiz. 21, 485 2625 1451 (1975) 2625 1451 [Ka77] C. Kalbach, Z.Phys.A 283, 401 (1977) 2625 1451 [Ka85] C. Kalbach, Los Alamos report LA-10248-MS (1985) 2625 1451 [Ka88] C. Kalbach, Phys.Rev. C 37, 2350 (1988); see also 2625 1451 C. Kalbach and F.M. Mann, Phys.Rev. C 23, 112 (1981) 2625 1451 [Ko90] J. Kopecky and M. Uhl, Phys.Rev. C 41, 1941 (1990) 2625 1451 [Le72] O.F. Lemos, "Diffusion Elastique de Particules Alpha 2625 1451 de 21 a 29.6 MeV sur des Noyaux de la Region Ti-Zn," Orsay 2625 1451 report, Series A, No. 136 (1976). 2625 1451 [Ma88] D.G. Madland, Proc. OECD/NEANDC Specialist's Meeting on 2625 1451 Preequilibrium Nuclear Reactions, Semmering, Austria, Feb. 2625 1451 1988, NEANDC-245 (1988) p.103 2625 1451 [Pe63] C.M. Perey and F.G. Perey, Phys.Rev. 132, 755 (1963) 2625 1451 [Re95] J. Raynal, "Notes on ECIS94," CEA informal 2625 1451 report,Saclay (1995). 2625 1451 [Yo92] P.G. Young, E.D. Arthur, and M.B. Chadwick, Los Alamos 2625 1451 National Laboratory report LA-12343-MS (1992) 2625 1451 2625 1451 **************************************************************** 2625 1451 2625 1451 ENDF/B-VI MOD 3 Revision, October 1997, V. McLane (NNDC). 2625 1451 2625 1451 1. Corrected residual nucleus in File 6, MT=16,103. 2625 1451 2. Updated File 1 comments and corrected references. 2625 1451 2625 1451 **************************************************************** 2625 1451 2625 1451 ENDF/B-VI MOD 2 Revision, July 1991, (ORNL) 2625 1451 2625 1451 The elastic transformation matrix was removed. 2625 1451 2625 1451 **************************************************************** 2625 1451 2625 1451 ENDF/B-VI MOD 1 Evaluation, November 1989, D.M. Hetrick, C.Y. Fu,2625 1451 and N.M. Larson (ORNL) 2625 1451 2625 1451 This work employed nuclear model codes including the 2625 1451 Distorted Wave Born Approximation (DWBA) program DWUCK [1] 2625 1451 and the Hauser-Feshbach code TNG [2,3,4]. The TNG code provides 2625 1451 energy and angular distributions of particles emitted in the 2625 1451 compound and pre-compound reactions, ensures consistency among 2625 1451 all reactions, and maintains energy balance. details pertinent 2625 1451 to the contents of this evaluation will be published at a later 2625 1451 date. 2625 1451 Resonance parameters and thermal cross sections are from 2625 1451 Mughabghab [5] and adjusted through SAMMY by N. Larson using J.A.2625 1451 Harvey data [11] between 110 and 700 keV, capture cross section 2625 1451 background in File 3 was based on Allen [6]. Above 700 keV, 2625 1451 54Fe data from J.A. Harvey [11] used. 2625 1451 2625 1451 DESCRIPTION OF FILES 2625 1451 2625 1451 File 1 GENERAL INFORMATION ------------------------------------ 2625 1451 MT=451 General information, references, and definitions. 2625 1451 2625 1451 2625 1451 File 3 NEUTRON CROSS SECTIONS --------------------------------- 2625 1451 MT=103 (n,p) cross sections 2625 1451 Taken from the GLUCS [8] calculation in which this reaction 2625 1451 was studied simultaneously with 12 other dosimetry 2625 1451 reaction cross sections [9]. 2625 1451 2625 1451 File 33 UNCERTAINTY FILES ------------------------------------- 2625 1451 An LB=8 section is included for all non-derived files as 2625 1451 required by ENDF/B-VI. 2625 1451 MT=103 (n,p) covariances were taken from the GLUCS [8] 2625 1451 calculation in which this reaction was studied 2625 1451 simultaneously with 12 other dosimetry reaction cross 2625 1451 sections [9]. 2625 1451 ---------------------------------------------------------------- 2625 1451 REFERENCES: 2625 1451 2625 1451 [1] P.D. Kunz, "Distorted Wave Code DWUCK72," Univ. of Colorado, 2625 1451 unpublished (1972). 2625 1451 [2] C.Y. Fu, "A Consistent Nuclear Model For Compound and Pre- 2625 1451 compound Reactions with Conservation of Angular Momentum," 2625 1451 report ORNL/TM-7042 (1980), and Nucl.Sci.Eng. 100, 61 (1988).2625 1451 [3] C.Y Fu, "Development and Application of Multi-step 2625 1451 Hauser-Feshbach/Pre-equilibrium Model Theory," Symp. 2625 1451 Neutron Cross Sections from 10 to 50 MeV, Upton, N.Y., 2625 1451 May 12-14,1980, report BNL-NCS-51425, p.675. 2625 1451 [4] K. Shibata and C.Y. Fu, "Recent Improvements of the TNG 2625 1451 Statistical Model Code", report ORNL/TM-10093 (1986). 2625 1451 [5] S.F. Mughabghab, Neutron Cross Sections, Vol. 1: Resonance 2625 1451 Parameters and Thermal Cross Sections, Part A (Academic 2625 1451 Press, 1981). 2625 1451 [6] B.J. Allen, Geel Conference, 1977. 2625 1451 [7] P.T. Guenther, D.L. Smith, A.B. Smith et al., Ann.Nucl. En. 2625 1451 13 (11), 601-610 (1986). 2625 1451 [8] D.M. Hetrick and C.Y. Fu, "GLUCS: A Generalized Least Squares2625 1451 Program for Updating Cross Section Evaluations with 2625 1451 Correlated Data Sets," report ORNL/TM-7341 [ENDF-303] (1980).2625 1451 [9] C.Y. Fu and D.M. Hetrick, "Experience in Using the 2625 1451 Covariances of Some ENDF/B-V Dosimetry Cross Sections: 2625 1451 Proposed Improvements and Additions of Cross-reaction 2625 1451 Covariances," Proc. Fourth ASTM-Euratom Symp. on Reactor 2625 1451 Dosimetry, Gaithersburg, Maryland, March 22-26,1982, 2625 1451 (US National Bureau Of Standards, 1982) p.877. 2625 1451 ([0] S.M. Grimes, R.C. Haight, K.R. Alvar et al., Phys.Rev., C19, 2625 1451 2127 (1979). 2625 1451 [11] J.A. Harvey, private communication, 1989. 2625 1451 2625 1451 ******************************************************************2625 1451 ********* End of (N,P) bibliographical component ********* 2625 1451 ***************************************************************** 2625 1451 The Q values and threshold energies were updated prior to pro- 2625 1451 cessing through the codes to comply with the values obtained 2625 1451 using the NNDC calculation program which is based on the 1995 2625 1451 Update to the Atomic mass Evaluation. 2625 1451 2625 1451 File 2 added to the pointwise file containing only the effective 2625 1451 scattering radius with no resonance parameters given. 2625 1451 Taken from ENDF/B-VI 2625 1451 ************************ C O N T E N T S *********************** 2625 1451 2625 1451 ***************** Program LINEAR (VERSION 2002-1) ***************2625 1451 For All Data Greater than 1.0000E-10 barns in Absolute Value 2625 1451 Data Linearized to Within an Accuracy of .100000000 per-cent 2625 1451 ***************** Program SIGMA1 (VERSION 2002-1) ***************2625 1451 Data Doppler Broadened to 300.000000 Kelvin 2625 1451 for All Data Greater than 1.0000E-10 barns in Absolute Value 2625 1451 Data Linearized to Within an Accuracy pf .100000000 per-cent 2625 1451 ***************** Program FIXUP (Version 2002-1) ****************2625 1451 Corrected ZA/AWR in All Sections-----------------------------Yes 2625 1451 Corrected Thresholds-----------------------------------------Yes 2625 1451 Extended Cross Sections to 20 MeV----------------------------No 2625 1451 Allow Cross Section Deletion---------------------------------No 2625 1451 Allow Cross Section Reconstruction---------------------------No 2625 1451 Make All Cross Sections Non-Negative-------------------------Yes 2625 1451 Delete Energies Not in Ascending Order-----------------------Yes 2625 1451 Deleted Duplicate Points-------------------------------------Yes 2625 1451 Check for Ascending MAT/MF/MT Order--------------------------Yes 2625 1451 Check for Legal MF/MT Numbers--------------------------------Yes 2625 1451 Allow Creation of Missing Sections---------------------------No 2625 1451 Allow Insertion of Energy Points-----------------------------No 2625 1451 Create Uniform Energy Grid-----------------------------------No 2625 1451 Delete Section if Cross Section =0 at All Energies-----------Yes 2625 1451 ***************** Program GROUPIE (VERSION 2002-1) **************2625 1451 Unshielded Group Averages Using 640 Groups 2625 1451 Weighting Spectrum: Flat (Constant) Spectrum 2625 1451 1 451 538 12625 1451 3 16 25 12625 1451 3 103 70 12625 1451 3 107 62 12625 1451 33 16 21 12625 1451 33 103 57 12625 1451 33 107 35 12625 1451 2625 1 0 2625 0 0 2.60540E+4 5.34762E+1 0 0 0 02625 3 16 -1.33785E+7-1.33785E+7 0 0 1 652625 3 16 65 1 2625 3 16 13600000.0 .000103649 13700000.0 .000301866 13800000.0 .0005359992625 3 16 13900000.0 .000884865 14000000.0 .001354120 14100000.0 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7.4355E-03 1.0657E-04 2.9882E-04 1.8815E-04 1.8663E-04 1.6565E-04262533103 2.3528E-04 1.7206E-04 6.2483E-04 6.4762E-03 3.5286E-04 2.8481E-04262533103 2.1734E-04 2.1479E-04 1.6072E-04 1.3802E-04 1.0328E-04 1.8866E-04262533103 1.0632E-03 3.3244E-04 2.2580E-04 2.2490E-04 1.8391E-04 6.6682E-05262533103 5.3549E-05 7.1187E-05 7.5034E-04 3.3845E-04 3.3483E-04 2.3650E-04262533103 9.7942E-05 7.7582E-05 5.3421E-05 6.7111E-04 5.1950E-04 2.9475E-04262533103 3.7441E-05 3.7615E-05 9.3780E-05 6.4010E-04 2.9008E-04 4.1529E-05262533103 4.1092E-05 9.1147E-05 6.1427E-04 1.4060E-04 1.1416E-04-1.7842E-05262533103 2.0209E-03 1.0198E-03 1.0894E-03 1.1388E-03 7.7129E-04 9.9624E-03262533103 0.0000E+00 0.0000E+00 12 4 48 24262533103 1.0000E-05 0.0000E+00 6.9000E+05-9.0000E-02 1.0000E+06-1.0000E-01262533103 2.0000E+06-2.7000E-01 4.0000E+06-2.5000E-01 6.0000E+06-1.0000E+00262533103 7.0000E+06-1.0000E+00 8.0000E+06-3.1000E-01 1.2000E+07-2.9000E-01262533103 1.4000E+07-3.3000E-01 1.6000E+07-1.1000E-01 2.0000E+07 0.0000E+00262533103 1.0000E-05 0.0000E+00 5.0000E+05 1.7503E-01 1.0000E+06 8.0475E-02262533103 2.0000E+06 3.2607E-02 4.0000E+06 2.7392E-02 6.0000E+06 2.5906E-02262533103 7.0000E+06 2.5300E-02 8.0000E+06 2.4784E-02 1.2000E+07 4.4955E-02262533103 1.4000E+07 3.3747E-02 1.6000E+07 9.9812E-02 2.0000E+07 0.0000E+00262533103 0.000000+0 0.000000+0 0 1 88 44262533103 1.000000-5 0.000000+0 6.900000+5 2.481500-3 8.000000+5 2.481500-3262533103 1.000000+6 5.828500-4 1.250000+6 5.828500-4 1.500000+6 5.828500-4262533103 1.750000+6 5.828500-4 2.000000+6 2.583600-4 2.250000+6 2.583600-4262533103 2.500000+6 2.583600-4 2.750000+6 2.583600-4 3.000000+6 2.583600-4262533103 3.220000+6 2.583600-4 3.250000+6 2.583600-4 3.350000+6 2.583600-4262533103 3.360000+6 2.583600-4 3.400000+6 2.583600-4 3.450000+6 2.583600-4262533103 3.500000+6 2.583600-4 3.750000+6 2.583600-4 4.000000+6 1.688300-4262533103 4.250000+6 1.688300-4 4.500000+6 1.688300-4 5.000000+6 1.688300-4262533103 5.500000+6 1.688300-4 6.000000+6 6.040000-4 7.000000+6 5.760900-4262533103 8.000000+6 1.713800-4 8.500000+6 1.713800-4 9.000000+6 1.713800-4262533103 9.500000+6 1.713800-4 1.000000+7 1.713800-4 1.050000+7 1.713800-4262533103 1.100000+7 1.713800-4 1.150000+7 1.713800-4 1.200000+7 5.274600-4262533103 1.250000+7 5.274600-4 1.300000+7 5.274600-4 1.350000+7 5.274600-4262533103 1.400000+7 3.382400-4 1.500000+7 3.382400-4 1.600000+7 9.862800-4262533103 1.800000+7 9.862800-4 2.000000+7 0.000000+0 262533103 0.000000+0 0.000000+0 0 8 88 44262533103 1.000000-5 0.000000+0 6.900000+5 2.35040-15 8.000000+5 1.71840-12262533103 1.000000+6 1.95530-11 1.250000+6 2.22450-10 1.500000+6 1.147100-9262533103 1.750000+6 1.375300-8 2.000000+6 2.922700-8 2.250000+6 7.618500-8262533103 2.500000+6 1.917900-7 2.750000+6 5.167200-7 3.000000+6 9.244100-7262533103 3.220000+6 1.009500-6 3.250000+6 1.023200-6 3.350000+6 1.076900-6262533103 3.360000+6 1.086200-6 3.400000+6 1.080800-6 3.450000+6 1.070700-6262533103 3.500000+6 1.331000-6 3.750000+6 2.066100-6 4.000000+6 1.708300-6262533103 4.250000+6 1.882200-6 4.500000+6 2.511700-6 5.000000+6 3.306500-6262533103 5.500000+6 3.915200-6 6.000000+6 1.542800-5 7.000000+6 1.456600-5262533103 8.000000+6 4.226500-6 8.500000+6 4.273700-6 9.000000+6 4.251200-6262533103 9.500000+6 4.188000-6 1.000000+7 4.185100-6 1.050000+7 4.171800-6262533103 1.100000+7 4.161300-6 1.150000+7 3.840400-6 1.200000+7 1.040900-5262533103 1.250000+7 9.461600-6 1.300000+7 8.428100-6 1.350000+7 7.345100-6262533103 1.400000+7 3.401000-6 1.500000+7 2.165000-6 1.600000+7 3.813700-6262533103 1.800000+7 2.035300-6 2.000000+7 0.000000+0 262533103 262533 0 2.60540E+4 5.34762E+1 0 0 0 1262533107 0.000000+0 0.000000+0 0 107 0 1262533107 0.000000+0 0.000000+0 1 5 190 19262533107 1.000000-5 2.500000+6 5.000000+6 6.000000+6 7.000000+6 8.000000+6262533107 9.000000+6 1.000000+7 1.100000+7 1.200000+7 1.300000+7 1.400000+7262533107 1.450000+7 1.500000+7 1.600000+7 1.700000+7 1.800000+7 1.900000+7262533107 2.000000+7 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0262533107 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0262533107 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0262533107 0.000000+0 1.967980-2 9.633730-3 1.642170-3-7.873680-4 1.763730-4262533107 1.190170-3 1.391200-3 9.990920-4 4.107560-4-7.196720-6-7.635850-5262533107 2.317610-5 3.165150-4 6.879890-4 6.296170-4-2.098400-4-1.980280-3262533107 6.920690-3 2.102260-3-7.406730-5 2.004840-4 8.283510-4 1.023790-3262533107 8.128000-4 4.201790-4 9.153950-5-9.428760-6 2.000020-5 1.747130-4262533107 4.071900-4 4.020070-4-9.273340-5-1.202550-3 1.693190-3 1.036540-3262533107 6.190390-4 4.273150-4 3.244890-4 2.414200-4 1.586830-4 8.148000-5262533107 3.617670-5 1.695870-5 9.143770-6 3.129460-5 7.794820-5 1.265680-4262533107 1.582870-4 1.363610-3 1.012880-3 5.698220-4 2.544810-4 7.642630-5262533107 1.509480-6-8.578530-6 3.519450-6 1.342820-5 2.123660-5 1.402950-5262533107 -8.208480-6-2.787740-5-2.663740-5 1.050540-3 8.480750-4 5.713450-4262533107 3.081410-4 1.087930-4 6.133330-6-4.996150-6 1.701860-5 7.081030-5262533107 1.229820-4 8.175890-5-1.015020-4-4.373240-4 9.279640-4 8.104860-4262533107 5.821060-4 3.289850-4 1.294250-4 4.657870-5 3.281980-5 6.404150-5262533107 1.406320-4 1.699830-4 6.664270-5-2.138110-4 8.554680-4 7.401580-4262533107 5.261530-4 2.858140-4 1.331290-4 6.311590-5 2.314560-5 7.289320-5262533107 2.015140-4 3.382760-4 4.238150-4 7.539170-4 6.328030-4 4.126190-4262533107 2.242130-4 1.093590-4-4.162330-6-8.076930-6 1.879820-4 5.566300-4262533107 1.043390-3 6.135990-4 4.640330-4 2.936590-4 1.706290-4 2.345590-5262533107 -2.956160-5 1.618240-4 6.136620-4 1.290690-3 4.150570-4 3.211380-4262533107 2.403130-4 1.283090-4 6.183300-5 1.676890-4 4.809320-4 9.913180-4262533107 3.109670-4 2.900050-4 2.492490-4 2.077720-4 2.190780-4 3.096660-4262533107 4.856670-4 3.172600-4 3.412840-4 3.374690-4 2.897020-4 2.071610-4262533107 1.028700-4 4.633580-4 5.410820-4 4.564190-4 1.841470-4-2.618490-4262533107 7.451530-4 7.686590-4 5.507190-4 8.915960-5 1.151220-3 1.553840-3262533107 1.941310-3 3.226410-3 5.491560-3 1.062090-2 262533107 262533 0 2625 0 0 0 0 0 2.60560E+4 5.54544E+1 0 0 34 102631 1451 0.0 0.0 0 0 0 62631 1451 1.00000E+0 2.00000E+7 0 0 10 20022631 1451 3.00000E+2 0.0 1 0 261 32631 1451 26-Fe- 56 FEI EVAL-Oct02 K.I.Zolotarev 2631 1451 DIST-Feb2004 2631 1451 ----IRDF-2002 MATERIAL 2631 2631 1451 -----INCIDENT NEUTRON DATA 2631 1451 ------ENDF-6 FORMAT 2631 1451 ***************************************************************** 2631 1451 26-FE- 56 FEI EVAL-Oct02 K.I.Zolotarev 2631 1451 DIST-Nov03 20031105 2631 1451 ----BROND-3 MATERIAL 2631 Revision 2 2631 1451 -----INCIDENT NEUTRON DATA 2631 1451 ------ENDF-6 FORMAT 2631 1451 ------Russian Reactor Dosimetry File RRDF-2002 2631 1451 ***************************************************************** 2631 1451 Author of evaluation: K.I.Zolotarev 2631 1451 ***************************************************************** 2631 1451 MF=3 2631 1451 MT=103 -(n,p) cross section 2631 1451 ------------------------------------- 2631 1451 Excitation function for the Fe-56(n,p)Mn-56 reaction in the 2631 1451 energy region from threshold to 20 MeV was evaluated by means of 2631 1451 statistical analysis of experimental cross section data [1-40]. 2631 1451 The energy dependence of cross-section from 4.0 MeV to the 2631 1451 threshold was extrapolated with L=0 penetrability function for 2631 1451 the outgoing p + Mn-56 channel [41]. 2631 1451 Analised microscopic experimental data [1-40], [42-69] were 2631 1451 renormalized if it were possible to the new recommended standards 2631 1451 for monitor reaction cross sections and decay data. 2631 1451 Special correction was done with experimental data [2],[9-10] 2631 1451 and [17]. Experimental data of Terrell and Holm [2] corresponding 2631 1451 to neutron energies 6.54 MeV,7.41 MeV, 8.21 MeV were renormalized 2631 1451 to the preliminary evaluated cross section value at En=8.2 MeV. 2631 1451 Data of Liskien and Paulsen [9] measured in the energy range 2631 1451 12.60 - 19.58 MeV were corrected to the preliminary evaluated 2631 1451 integral of cross section in the energy interval 14-15 MeV. Data 2631 1451 obtained by Liskien and Paulsen in the energy range 6.06-8.20 MeV 2631 1451 [10] were renormalized to the preliminary evaluated cross section 2631 1451 value at 8.0 MeV. The correction factors for the experimental 2631 1451 data [2], [9] and [10] were Fc=1.23487, Fc=1.09674 and Fc=1.05835,2631 1451 respectively. Data of Smith and Meadows [17] obtained in the expe-2631 1451 riment with neutrons from D(d,n)He3 reaction in the energy range 2631 1451 6.486 - 9.945 MeV were renormalized to the preliminary evaluated 2631 1451 cross section value at 9.945 MeV. The correction factor for this 2631 1451 data was Fc=1.14066 . 2631 1451 Experimental information about Fe-56(n,p)Mn-56 reaction exci- 2631 1451 tation function are given in the ref. [13],[21],[23] in the form 2631 1451 of cross section ratios to monitor reactions. The results of pre- 2631 1451 cise relative measurements of Vonach et al. [13] in the energy 2631 1451 range 13.6-14.7 MeV were normalized to the preliminary evaluated 2631 1451 absolute cross section value 107.13 mb for Fe-56(n,p)Mn-56 reac- 2631 1451 tion at 14.7 MeV point. Raics et al. [21] and Antov et al. [23] 2631 1451 measured ratios of Fe-56(n,p)Mn-56 cross section to U-238(n,f) 2631 1451 and Al-27(n,a)Na-24 reactions cross section. Recommended absolute 2631 1451 cross section data for U-238(n,f) and Al-27(n,a)Na-24 reactions 2631 1451 were taken from ref. [70] and [71], respectively. 2631 1451 Experimental data from refs.[5] and [11] were used partially. 2631 1451 It were used only data obtained at 14.5 MeV [5] and in the energy 2631 1451 range 3.95 - 10.0 MeV [11]. Cross et al. data [5] for the neutron 2631 1451 energies 13.78 MeV, 14.07 MeV and 14.73 MeV were rejected due 2631 1451 to their systematically underestimation Fe-56(n,p)Mn-56 cross 2631 1451 section. The result of Grundl measurements at 14.1 MeV [11] was 2631 1451 not taken into account due to a very overestimated cross section 2631 1451 value obtained for this energy point. 2631 1451 Experimental cross section data [42-69] were rejected due to 2631 1451 their discrepancy with the main bulk of experimental data [1-40]. 2631 1451 In the rejected experiments [42-51], [53], [56-58], [60-62], 2631 1451 [64-67] and [69] the cross section values were measured only in a 2631 1451 one energy point in the interval 14 - 15 MeV. 2631 1451 Statistical analysis of input cross section data was carried 2631 1451 out by means of PADE-2 code [72]. Rational function was used as 2631 1451 the model function [73]. 2631 1451 U-235 thermal fission [74] and Cf-252 spontaneous fission 2631 1451 neutron spectra [75] averaged cross-sections calculated from the 2631 1451 evaluated Fe-56(n,p)Mn-56 excitation function are the following: 2631 1451 2631 1451 ÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÂÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÂÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄ 2631 1451 TYPE OF SPECTRUM ³ ,mb (calc.) ³ , mb (measured) 2631 1451 ÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÅÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÅÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄ 2631 1451 U-235 neutron fission ³ 1.0997 ³ 1.09 +- 0.04 [76] 2631 1451 ³ ³ 1.13 +- 0.07 [77] 2631 1451 ³ ³ 1.083 +- 0.017 [78] 2631 1451 ÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÅÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÅÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄ 2631 1451 Cf-252 spont. fission ³ 1.4626 ³ 1.471 +- 0.025 [79] 2631 1451 ³ ³ 1.465 +- 0.026 [80] 2631 1451 ÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÁÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÁÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄ 2631 1451 2631 1451 MT=33 2631 1451 MT=103 -(n,p) cross section cov. matrix 2631 1451 --------------------------------------- 2631 1451 Uncertainties in the evaluated excitation function for the 2631 1451 reaction Fe-56(n,p)Mn-56 are given in the form of relative covari-2631 1451 ance matrix for the 25-neutron energy groups (LB=5). Covariance 2631 1451 matrix of uncertainties was calculated simultaneously with 2631 1451 recommended cross section data by means of PADE-2 code. 2631 1451 Eigenvalues of the 6-th digits relative covariance matrix 2631 1451 given in the 33-file are the following: 2631 1451 2631 1451 4.35195E-09 4.49367E-09 4.76919E-09 5.19445E-09 2631 1451 5.64595E-09 6.15408E-09 7.11594E-09 7.89561E-09 2631 1451 9.40785E-09 1.19394E-08 1.48739E-08 1.86574E-08 2631 1451 2.18905E-08 1.32640E-06 2.77587E-04 3.86559E-04 2631 1451 6.52483E-04 1.02062E-03 1.26078E-03 1.33119E-03 2631 1451 1.81407E-03 2.05264E-03 3.00443E-03 4.57086E-03 2631 1451 8.12954E-03 2631 1451 2631 1451 References : 2631 1451 1. G.Brown Philosophical Magazine, v.2, p.785, 1957 2631 1451 2. J.Terrell, D.M.Holm Phys. Rev., v.109, p.2031, 1958 2631 1451 3. H.Pollehn, H.Neuert Zeitschrift f. Naturforschung, sect. A, 2631 1451 v.16, p.227, 1961 2631 1451 4. F.Gabbard, B.D.Kern Phys. Rev., v.128, p.1276, 1962 2631 1451 5. W.G.Cross et al. 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Soc., v.75, p.526, 1960 2631 1451 51. D.M.Chittenden et al. Phys. Rev., v.122, p.860, May 1961 2631 1451 52. M.Bormann et al. Zeitschrift fuer Physik, v.166, p.477, 1962 2631 1451 53. J.E.Strain, W.J.Ross Report ORNL-3672, January 1965 2631 1451 54. J.D.Hemingway et al. Proc. Royal Society, section A, v.292, 2631 1451 p.180, May 1966 2631 1451 55. P.Cuzzocrea et al. Nuovo Cimento B, v.54, p.53, March 1968 2631 1451 56. S.M.Qaim et al. Proc. of Conference on Chemical Nuclear Data, 2631 1451 Measurements and Applicat., Univ. of Kent, Canterbury, 20-22 2631 1451 September 1971, p.121 2631 1451 57. J.J.Singh Trans. Amer. Nucl. Soc., v.15, p.147, June 1972 2631 1451 58. R.Spangler et al. Ann. Nucl. Sci., v.22, p.818, November 1975 2631 1451 59. A.B.M.G.Mostafa Nuclear Science and Applications, Series B: 2631 1451 Physical sciences, v.9, p.10, October 1976 2631 1451 60. Z.A.Ramendik et al. Atomnaja Energija (Sov.), v.42, no.2, 2631 1451 p.136, February 1977 2631 1451 61. S.Sothras Dissertation Abstracts, v.B38, p.280, July 1978 2631 1451 62. D.Sharma et al. Proc. of the 21-st Nuclear Physics and Solid 2631 1451 State Physics Symposium, Bombay, India, 28-31 December 1978 , 2631 1451 v.2, p.349 2631 1451 63. P.N.Ngoc et al. Progress Report INDC(HUN)-20,p.3, March 1983; 2631 1451 P.N.Ngoc et al. Nucleonika (Pol), v.29, p.87, 1984 2631 1451 64. P.N.Ngoc et al. Report INDC(VN)-2, November 1983 2631 1451 65. I.Garlea et al. Rev. Roum. Phys., v.29, p.421, 1984 2631 1451 66. J.P.Gupta et al. Pramana, v.24, p.637, 1985 2631 1451 67. M.Viennot et al. Nucl. Sci. Eng., v.108, p.289, July 1991 2631 1451 68. A.Ercan et al. Proc. of an Int. Conference on Nuclear Data 2631 1451 for Science and Technology, Juelich, FRG, 13-17 May 1991, 2631 1451 Springer-Verlag, 1992 2631 1451 69. L.I.Klochkova et al. Voprosy Atomnoy Nauki i Techniki, Ser.: 2631 1451 Yadernye Konstanty, v.1, p.27, 1992 2631 1451 70. ENDF/B-VI library,MAT=9237, eval. L.W.Weston et al. Nov.1989; 2631 1451 H.Conde editor Nuclear Data Standards for Nuclear Measu- 2631 1451 rements, Report NEANDC-311 U, pp.70-74, OECD, Paris, 1992. 2631 1451 71. M.Wagner et al. Evaluation of cross sections for 14 important 2631 1451 neutron-dosimetry reactions. Physics Data No.13-5, 1990 2631 1451 72. S.A.Badikov et al. Preprint FEI-1686, Obninsk, 1985 2631 1451 73. S.Badikov, N.Rabotnov, K.Zolotarev Proc. of NEANSC Speciali- 2631 1451 st's Meeting on Evaluation and Processing of Covariance Data, 2631 1451 Oak Ridge, USA, 7-9 September 1992, OECD, Paris, 1993, p.105 2631 1451 74. L.W.Weston et al. Evaluated Neutron Data for Uranium-235, 2631 1451 ENDF/B-VI Library, MAT=9228, MF=5, MT=18, eval. April 1989 2631 1451 75. W.Mannhart IAEA-TECDOC-410, p.158, IAEA, Vienna, 1987 2631 1451 76. W.Mannhart Proc. 5-th ASTM-EUR. Symp. on Reactor Dosimetry, 2631 1451 Geesthacht, FRG, September 24-28, 1984, Vol.2, p.813, 1985 2631 1451 77. O.Horibe et al. Proc. of Conf.:50 Years with Nuclear Fission, 2631 1451 Washington D.C., 25-28 April 1989, v.2, p.923 2631 1451 78. W.Mannhart Progress Report INDC(Ger)-045, pp.40-43, 1999 2631 1451 79. W.Mannhart Handbook on Nuclear Activation Cross Sections , 2631 1451 IAEA Tech. Report Ser. No.273, p.413, 1987 2631 1451 80. W.Mannhart Validation of Differential Cross Sections with 2631 1451 Integral Data , Report INDC(NDS)-435, pp.59-64, IAEA, Vienna, 2631 1451 September 2002 2631 1451 ***************************************************************** 2631 1451 The Q values and threshold energies were updated prior to pro- 2631 1451 cessing through the codes to comply with the values obtained 2631 1451 using the NNDC calculation program which is based on the 1995 2631 1451 Update to the Atomic mass Evaluation. 2631 1451 2631 1451 File 2 added to the pointwise file containing only the effective 2631 1451 scattering radius with no resonance parameters given. 2631 1451 Taken from ENDF/B-VI 2631 1451 ***************************************************************** 2631 1451 ***************** Program LINEAR (VERSION 2002-1) ***************2631 1451 For All Data Greater than 1.0000E-10 barns in Absolute Value 2631 1451 Data Linearized to Within an Accuracy of .100000000 per-cent 2631 1451 ***************** Program SIGMA1 (VERSION 2002-1) ***************2631 1451 Data Doppler Broadened to 300.000000 Kelvin 2631 1451 for All Data Greater than 1.0000E-10 barns in Absolute Value 2631 1451 Data Linearized to Within an Accuracy pf .100000000 per-cent 2631 1451 ***************** Program FIXUP (Version 2002-1) ****************2631 1451 Corrected ZA/AWR in All Sections-----------------------------Yes 2631 1451 Corrected Thresholds-----------------------------------------Yes 2631 1451 Extended Cross Sections to 20 MeV----------------------------No 2631 1451 Allow Cross Section Deletion---------------------------------No 2631 1451 Allow Cross Section Reconstruction---------------------------No 2631 1451 Make All Cross Sections Non-Negative-------------------------Yes 2631 1451 Delete Energies Not in Ascending Order-----------------------Yes 2631 1451 Deleted Duplicate Points-------------------------------------Yes 2631 1451 Check for Ascending MAT/MF/MT Order--------------------------Yes 2631 1451 Check for Legal MF/MT Numbers--------------------------------Yes 2631 1451 Allow Creation of Missing Sections---------------------------No 2631 1451 Allow Insertion of Energy Points-----------------------------No 2631 1451 Create Uniform Energy Grid-----------------------------------No 2631 1451 Delete Section if Cross Section =0 at All Energies-----------Yes 2631 1451 ***************** Program GROUPIE (VERSION 2002-1) **************2631 1451 Unshielded Group Averages Using 640 Groups 2631 1451 Weighting Spectrum: Flat (Constant) Spectrum 2631 1451 1 451 268 12631 1451 3 103 61 12631 1451 33 103 66 12631 1451 2631 1 0 2631 0 0 2.60560E+4 5.54544E+1 0 99 0 02631 3103 -2.91310E+6-2.91310E+6 0 0 1 1722631 3103 172 1 2631 3103 2900000.00 5.9516E-12 3000000.00 8.5021E-11 3100000.00 1.8579E-102631 3103 3200000.00 2.8657E-10 3300000.00 3.8734E-10 3400000.00 4.8811E-102631 3103 3500000.00 3.19099E-9 3600000.00 2.37217E-8 3700000.00 2.30222E-72631 3103 3800000.00 8.58242E-7 3900000.00 3.26945E-6 4000000.00 9.50803E-62631 3103 4100000.00 2.13588E-5 4200000.00 4.12059E-5 4300000.00 7.22242E-52631 3103 4400000.00 .000118637 4500000.00 .000186046 4600000.00 .0002818582631 3103 4700000.00 .000415817 4800000.00 .000600658 4900000.00 .0008528442631 3103 5000000.00 .001193315 5100000.00 .001648055 5200000.00 .0022481552631 3103 5300000.00 .003028750 5400000.00 .004026080 5500000.00 .0052718752631 3103 5600000.00 .006784800 5700000.00 .008560220 5800000.00 .0105616952631 3103 5900000.00 .012519200 6000000.00 .014238450 6100000.00 .0161200002631 3103 6200000.00 .018181150 6300000.00 .020072250 6400000.00 .0217811502631 3103 6500000.00 .023826650 6600000.00 .025746950 6700000.00 .0270865002631 3103 6800000.00 .028396700 6900000.00 .029706900 7000000.00 .0310583252631 3103 7100000.00 .032450975 7200000.00 .033907625 7300000.00 .0354282752631 3103 7400000.00 .037005975 7500000.00 .038640725 7600000.00 .0403027402631 3103 7700000.00 .041992020 7800000.00 .043681300 7900000.00 .0453705802631 3103 8000000.00 .047059860 8100000.00 .048700150 8200000.00 .0502914502631 3103 8300000.00 .051829775 8400000.00 .053315125 8500000.00 .0547361002631 3103 8600000.00 .056092700 8700000.00 .057449300 8800000.00 .0587423902631 3103 8900000.00 .059971970 9000000.00 .061201550 9100000.00 .0624311302631 3103 9200000.00 .063660710 9300000.00 .064888420 9400000.00 .0661142602631 3103 9500000.00 .067340100 9600000.00 .068565940 9700000.00 .0697917802631 3103 9800000.00 .071069213 9900000.00 .072398238 10000000.0 .0737272632631 3103 10100000.0 .075056288 10200000.0 .076448125 10300000.0 .0779027752631 3103 10400000.0 .079357425 10500000.0 .080812075 10600000.0 .0823283602631 3103 10700000.0 .083906280 10800000.0 .085484200 10900000.0 .0870621202631 3103 11000000.0 .088640040 11100000.0 .090238714 11200000.0 .0918581432631 3103 11300000.0 .093477571 11400000.0 .095097000 11500000.0 .0967164292631 3103 11600000.0 .098335857 11700000.0 .099955286 11800000.0 .1014997502631 3103 11900000.0 .102969250 12000000.0 .104438750 12100000.0 .1059082502631 3103 12200000.0 .107264833 12300000.0 .108508500 12400000.0 .1097521672631 3103 12500000.0 .110859000 12600000.0 .111829000 12700000.0 .1127990002631 3103 12800000.0 .113632500 12900000.0 .114329500 13000000.0 .1149057502631 3103 13100000.0 .115361250 13200000.0 .115689750 13300000.0 .1158912502631 3103 13400000.0 .115996000 13500000.0 .115908000 13600000.0 .1157240002631 3103 13700000.0 .115415750 13800000.0 .114983250 13900000.0 .1144076672631 3103 14000000.0 .113689000 14100000.0 .112970333 14200000.0 .1121030002631 3103 14300000.0 .111087000 14400000.0 .110071000 14500000.0 .1089375002631 3103 14600000.0 .107686500 14700000.0 .106435500 14800000.0 .1050908752631 3103 14900000.0 .103652625 15000000.0 .102214375 15100000.0 .1007761252631 3103 15200000.0 .099278768 15300000.0 .097722305 15400000.0 .0961658412631 3103 15500000.0 .094609377 15600000.0 .093052914 15700000.0 .0914964502631 3103 15800000.0 .089939986 15900000.0 .088383523 16000000.0 .0868270592631 3103 16100000.0 .085270595 16200000.0 .083714132 16300000.0 .0822083502631 3103 16400000.0 .080753250 16500000.0 .079298150 16600000.0 .0778430502631 3103 16700000.0 .076387950 16800000.0 .074994575 16900000.0 .0736629252631 3103 17000000.0 .072331275 17100000.0 .070999625 17200000.0 .0697291002631 3103 17300000.0 .068519700 17400000.0 .067310300 17500000.0 .0661009002631 3103 17600000.0 .064953475 17700000.0 .063868025 17800000.0 .0627825752631 3103 17900000.0 .061697125 18000000.0 .060671725 18100000.0 .0597063752631 3103 18200000.0 .058741025 18300000.0 .057775675 18400000.0 .0568670002631 3103 18500000.0 .056015000 18600000.0 .055163000 18700000.0 .0543110002631 3103 18800000.0 .053511638 18900000.0 .052764913 19000000.0 .0520181882631 3103 19100000.0 .051271463 19200000.0 .050573338 19300000.0 .0499238132631 3103 19400000.0 .049274288 19500000.0 .048624763 19600000.0 .0480200382631 3103 19700000.0 .047460113 19800000.0 .046900188 19900000.0 .0463402632631 3103 20000000.0 0.0 2631 3103 2631 3 0 2631 0 0 2.60560E+4 5.54544E+1 0 0 0 1263133103 0.000000+0 0.000000+0 0 103 0 1263133103 0.000000+0 0.000000+0 1 5 378 27263133103 1.000000-5 2.900000+6 5.000000+6 6.000000+6 7.000000+6 8.000000+6263133103 9.000000+6 1.000000+7 1.100000+7 1.150000+7 1.200000+7 1.250000+7263133103 1.300000+7 1.350000+7 1.400000+7 1.450000+7 1.500000+7 1.550000+7263133103 1.600000+7 1.650000+7 1.700000+7 1.750000+7 1.800000+7 1.850000+7263133103 1.900000+7 1.950000+7 2.000000+7 0.000000+0 0.000000+0 0.000000+0263133103 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0263133103 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0263133103 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0263133103 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0 6.980350-3263133103 2.338270-3 1.702360-4 4.353220-4-6.002120-5 3.473780-4 1.342800-4263133103 -1.013590-4-1.453170-4-1.032110-4-1.910370-5 5.580590-5 8.566180-5263133103 6.229070-5 2.961970-6-6.242280-5-1.056550-4-1.094340-4-7.037860-5263133103 1.788090-6 8.621690-5 1.524240-4 1.605400-4 6.014590-5-2.124840-4263133103 2.467980-3 6.043630-4 3.578130-4 2.055220-4 1.680110-4 2.128390-4263133103 1.769540-4 1.366700-4 9.916600-5 7.004770-5 5.015720-5 3.742600-5263133103 2.898990-5 2.276790-5 1.801980-5 1.502850-5 1.442320-5 1.658450-5263133103 2.127870-5 2.748400-5 3.331180-5 3.591800-5 3.136190-5 1.435050-5263133103 2.089910-3 5.811850-4 5.051160-4 4.463010-4 2.985330-4 2.415270-4263133103 2.109770-4 1.764960-4 1.378030-4 9.933400-5 6.703660-5 4.523120-5263133103 3.491560-5 3.399020-5 3.868840-5 4.511620-5 5.024920-5 5.232580-5263133103 5.086090-5 4.651010-5 4.095130-5 3.684360-5 3.791020-5 2.324770-3263133103 8.909680-4 2.243430-4 3.463970-4 3.586080-4 2.979140-4 2.137770-4263133103 1.316880-4 7.020240-5 3.717340-5 2.982110-5 3.851070-5 5.200710-5263133103 6.154930-5 6.257380-5 5.457370-5 4.014340-5 2.400670-5 1.237870-5263133103 1.271590-5 3.381980-5 8.624440-5 1.319090-3 6.075680-4 2.355910-4263133103 1.831400-4 1.797820-4 1.702720-4 1.477000-4 1.151050-4 8.057420-5263133103 5.231750-5 3.522030-5 2.987090-5 3.362030-5 4.237480-5 5.207540-5263133103 5.950970-5 6.257730-5 6.025220-5 5.245210-5 3.991740-5 2.416230-5263133103 9.567740-4 6.313680-4 3.548430-4 2.308720-4 1.569670-4 1.197830-4263133103 1.026490-4 9.147350-5 7.836800-5 6.207430-5 4.573040-5 3.376500-5263133103 2.952730-5 3.417020-5 4.646570-5 6.302500-5 7.854620-5 8.588480-5263133103 7.587850-5 3.687780-5 8.246530-4 7.196220-4 5.696170-4 4.048130-4263133103 2.567500-4 1.473680-4 8.506790-5 6.460670-5 7.167200-5 8.965720-5263133103 1.051550-4 1.104520-4 1.034970-4 8.671160-5 6.568700-5 4.830190-5263133103 4.434860-5 6.563950-5 1.265420-4 7.702400-4 6.761850-4 5.244410-4263133103 3.568230-4 2.113590-4 1.126140-4 6.666460-5 6.340340-5 8.424730-5263133103 1.104340-4 1.283210-4 1.311310-4 1.184390-4 9.489290-5 6.907830-5263133103 5.291210-5 6.162700-5 1.143590-4 6.324280-4 5.237220-4 3.833900-4263133103 2.465290-4 1.404880-4 7.815570-5 5.731600-5 6.563580-5 8.760000-5263133103 1.098400-4 1.236410-4 1.252570-4 1.152190-4 9.746320-5 7.872510-5263133103 6.832810-5 7.841040-5 4.660800-4 3.703690-4 2.615780-4 1.637400-4263133103 9.329270-5 5.585710-5 4.758000-5 5.925580-5 8.050700-5 1.024550-4263133103 1.187560-4 1.256000-4 1.212720-4 1.056640-4 7.989500-5 4.610700-5263133103 3.221980-4 2.512470-4 1.743220-4 1.074110-4 6.116180-5 3.911000-5263133103 3.878810-5 5.417430-5 7.796430-5 1.029630-4 1.226310-4 1.310680-4263133103 1.227080-4 9.188330-5 3.231030-5 2.173730-4 1.684250-4 1.165000-4263133103 7.298230-5 4.527820-5 3.576720-5 4.260230-5 6.120560-5 8.556410-5263133103 1.089960-4 1.244090-4 1.241890-4 9.983000-5 4.136060-5 1.477430-4263133103 1.185060-4 8.879480-5 6.555160-5 5.284920-5 5.160670-5 6.020940-5263133103 7.533050-5 9.255920-5 1.067160-4 1.118820-4 1.012050-4 6.649160-5263133103 1.138400-4 1.048040-4 9.446080-5 8.548980-5 7.961900-5 7.752210-5263133103 7.899010-5 8.316230-5 8.870200-5 9.388370-5 9.659180-5 9.423730-5263133103 1.173920-4 1.242110-4 1.246130-4 1.193480-4 1.101060-4 9.908370-5263133103 8.873970-5 8.171760-5 8.089820-5 8.954360-5 1.115220-4 1.477160-4263133103 1.605150-4 1.612980-4 1.513370-4 1.337330-4 1.127760-4 9.360740-5263133103 8.216990-5 8.538600-5 1.115530-4 1.858920-4 1.970250-4 1.936170-4263133103 1.779120-4 1.539130-4 1.268590-4 1.030300-4 8.983880-5 9.619200-5263133103 2.208930-4 2.302800-4 2.251120-4 2.070470-4 1.789350-4 1.444760-4263133103 1.081310-4 7.522490-5 2.563740-4 2.685600-4 2.647750-4 2.436270-4263133103 2.040130-4 1.447640-4 6.437880-5 3.017630-4 3.184520-4 3.123860-4263133103 2.767690-4 2.037260-4 8.360900-5 3.586650-4 3.751200-4 3.563580-4263133103 2.888990-4 1.563190-4 4.204370-4 4.352960-4 4.045150-4 3.095700-4263133103 5.050460-4 5.556140-4 5.751470-4 7.485460-4 9.914130-4 1.606050-3263133103 263133 0 2631 0 0 0 0 0 2.60580E+4 5.74356E+1 0 0 34 102637 1451 0.0 0.0 0 0 0 62637 1451 1.00000E+0 2.00000E+7 0 0 10 20022637 1451 3.00000E+2 0.0 1 0 162 32637 1451 26-Fe- 58 JAERI EVAL-OCT95 T.NAKAGAWA 2637 1451 DIST-Feb2004 2637 1451 ----IRDF-2002 MATERIAL 2637 2637 1451 -----INCIDENT NEUTRON DATA 2637 1451 ------ENDF-6 FORMAT 2637 1451 ******************************************************************2637 1451 26-FE- 58 JAERI EVAL-OCT95 T.NAKAGAWA 2637 1451 DIST-JUL98 2637 1451 ----JENDL/D-99 MATERIAL 2637 2637 1451 *****************EXTRACT FOR SPECIAL PURPOSE FILE*****************2637 1451 DOSIMETRY 2637 1451 ******************************************************************2637 1451 HISTORY 2637 1451 93-10 JENDL-3.2. 2637 1451 RE-EVALUATION WAS MADE BY 2637 1451 T.NAKAGAWA (NDC/JAERI): RESONANCE PARAMS, CROSS SECTIONS 2637 1451 S.IGARASI (NEDAC): GAMMA-RAY PRODUCTION DATA BELOW 10 KEV2637 1451 COMPILED BY T.NAKAGAWA 2637 1451 95-10 COVARIANCES WERE ESTIMATED BY T.NAKAGAWA. 2637 1451 98-04 COMPILED TO JENDL DOSIMETRY FILE 99. 2637 1451 02-10 COVARIANCES WERE MODIFIED BY K.SHIBATA. 2637 1451 2637 1451 ===== ORIGINAL DATA FILE ===== 2637 1451 2637 1451 2637 1451 FE-58 CAPTURE FE-59 (HALF-LIFE = 44.503 H) 2637 1451 2637 1451 MF=1 GENERAL INFORMATION 2637 1451 MT=451 DESCRIPTIVE DATA AND DICTIONARY 2637 1451 2637 1451 2637 1451 MF=2 RESONANCE PARAMETERS 2637 1451 MT = 151 RESOLVED RESONANCES 2637 1451 TAKEM FROM JENDL-3.2/1/. 2637 1451 RESONANCE REGION = 1.0E-5 EV TO 350.0 KEV 2637 1451 THE MULTILEVEL BREIT-WIGNER FORMULA WAS USED. PARAMETERS 2637 1451 WERE DETERMINED ON THE BASIS OF DATA BY GARG ET AL./2/, 2637 1451 KAEPPELER ET AL./3/, ALLEN AND MACKLIN/4/. 2637 1451 2637 1451 CALCULATED 2200-M/S CROSS SECTIONS AND RES. INTEGRALS. 2637 1451 2200-M/S RES. INTEG. 2637 1451 ELASTIC 6.470 B - 2637 1451 CAPTURE 1.300 B 1.36 B 2637 1451 TOTAL 7.770 B - 2637 1451 2637 1451 MF = 3 CROSS SECTIONS 2637 1451 MT = 1 TOTAL CROSS SECTION 2637 1451 TAKEN FROM JENDL-3.2/1/. 2637 1451 2637 1451 MT=2 ELASTIC 2637 1451 TOTAL CROSS SECTION - SUM OF PARTIAL CROSS SECTIONS 2637 1451 TAKEN FROM JENDL-3.2/1/. 2637 1451 2637 1451 MT = 102 CAPTURE CROSS SECTION 2637 1451 THE CAPTURE CROSS SECTION WAS ADOPTED FROM JENDL-3.2/1/. 2637 1451 2637 1451 IN THE RESONANCE REGION BELOW 350 KEV, THE CROSS SECTION WAS 2637 1451 CALCULATED FROM THE MULTILEVEL BREIT-WIGNER FORMULA WITH 2637 1451 RESONANCE PARAMETERS DETERMINED ON THE BASIS OF DATA BY GARG ET 2637 1451 AL./2/, KAEPPELER ET AL./3/, AND ALLEN AND MACKLIN/4/. ABOVE 2637 1451 350 KEV, THE CROSS SECTION WAS CALCULATED WITH THE OPTICAL AND 2637 1451 STATISTICAL MODEL CODE CASTHY/5/ BY NORMALIZING TO 3 MB AT 500 2637 1451 KEV/6/. CONTRIBUTIONS TO THE INELASTIC SCATTERING PROCESSES 2637 1451 WERE CONSIDERED BY ADOPTING THE EXCITED LEVELS IN THE FOLLOWING 2637 1451 TABLE. AND THE CONTRIBUTIONS TO THE (N,2N), (N,N'A), (N,N'P), 2637 1451 (N,P) AND (N,A) REACTION CROSS SECTIONS WERE ALSO CONSIDERED IN 2637 1451 THE CASTHY CALCULATION. THESE THRESHOLD REACTION CROSS SECTIONS 2637 1451 WERE CALCULATED WITH GNASH/7/. THE DIRECT CAPTURE CROSS SECTION 2637 1451 WAS CALCULATED WITH A SIMPLE FORMULA DERIVED BY BENZI AND 2637 1451 REFFO/8/ AND ADDED TO THE RESULT OF CASTHY CALCULATION. 2637 1451 2637 1451 OPTICAL POTENTIAL PARAMETERS 2637 1451 ----------------------------- 2637 1451 V = 46.0-0.25*EN (MEV), 2637 1451 WS = 14.0-0.2*EN (MEV), (IN THE GAUSSIAN FORM) 2637 1451 WI = 0.125*E-0.0004*E**2 (MEV), 2637 1451 VSO= 6.0 (MEV), 2637 1451 R = 1.286 (FM), A0 = 0.62 (FM) 2637 1451 RS = 1.387 (FM), AS = 0.7 (FM) 2637 1451 RSO= 1.07 (FM), ASO= 0.62 (FM) 2637 1451 2637 1451 LEVEL SCHEME 2637 1451 ------------ 2637 1451 NO. ENERGY(MEV) SPIN-PARITY 2637 1451 G.S. 0.0 0 + 2637 1451 1. 0.8108 2 + 2637 1451 2. 1.6747 2 + 2637 1451 3. 2.0765 4 + 2637 1451 4. 2.1339 3 + 2637 1451 5. 2.2581 0 + 2637 1451 6. 2.6004 4 + 2637 1451 7. 2.7819 1 + 2637 1451 8. 2.8764 2 + 2637 1451 9. 3.0840 2 + 2637 1451 10. 3.1330 4 + 2637 1451 11. 3.2330 2 + 2637 1451 12. 3.2440 0 + 2637 1451 LEVELS ABOVE 3.389 MEV WERE ASSUMED TO BE OVERLAPPING. 2637 1451 2637 1451 MF = 33 COVARIANCES OF NEUTRON CROSS SECTIONS 2637 1451 MT=102 CAPTURE CROSS SECTION 2637 1451 2637 1451 IN THE RESONANCE REGION, THE VARIANCE OF CROSS SECTION WAS 2637 1451 ESTIMATED FROM STANDARD DEVIATIONS OF THE RESONANCE PARAMETERS. 2637 1451 THE STANDARD DEVIATIONS WERE TAKEN FROM THE EXPERIMENTAL DATA 2637 1451 WHICH WERE CONSIDERED IN THE PARAMETER EVALUATION. FOR THE 2637 1451 RESONANCE PARAMETERS DETERMINED FROM THE CAPTURE AREA, THEIR 2637 1451 ERRORS WERE ESTIMATED FROM THE ERROR OF CAPTURE AREA. AN ERROR 2637 1451 OF 10 % WAS ASSUMED FOR OTHER CASES. 2637 1451 2637 1451 ABOVE 350 KEV, COVARIANCE MATRIX WAS OBTAINED BY USING 2637 1451 KALMAN/10/. ERRORS WERE CONSIDERED TO THE OPTICAL POTENTIAL AND 2637 1451 LEVEL DENSITY PARAMETERS, AND THE NORMALIZATION CROSS SECTION. 2637 1451 2637 1451 2637 1451 REFERENCES 2637 1451 1) NAKAGAWA T. ET AL.: J. NUCL. SCI. TECHNOL., 32, 1259 (1995). 2637 1451 2) GARG J.B. ET AL.: PHYS. REV., C18, 1141 (1978). 2637 1451 3) KAEPPELER F. ET AL.: NUCL. SCI. ENG., 84, 234 (1983). 2637 1451 4) ALLEN B.J. AND MACKLIN R.L.: J. PHYS. G., 6, 381 (1980). 2637 1451 5) IGARASI S. AND FUKAHORI T.: JAERI 1321 (1991). 2637 1451 6) TROFIMOV JU.N.: ATOMNAJA ENERGIJA, 58, 278 (1985). 2637 1451 7) YOUNG P.G. AND ARTHUR E.D.: LA-6974 (1977). 2637 1451 8) BENZI V. AND REFFO G.: CCDN/NW/10 (1969). 2637 1451 9) YAMAKOSHI H.: JAERI 1261, P.30 (1979). 2637 1451 10) KAWANO T. AND SHIBATA K.: JAERI-DATA/CODE 97-037 (1997) 2637 1451 [IN JAPANESE] 2637 1451 2637 1451 ******************************************************************2637 1451 The Q values and threshold energies were updated prior to pro- 2637 1451 cessing through the codes to comply with the values obtained 2637 1451 using the NNDC calculation program which is based on the 1995 2637 1451 Update to the Atomic mass Evaluation. 2637 1451 ******************************************************************2637 1451 ***************** Program LINEAR (VERSION 2002-1) ***************2637 1451 For All Data Greater than 1.0000E-10 barns in Absolute Value 2637 1451 Data Linearized to Within an Accuracy of .100000000 per-cent 2637 1451 ***************** Program RECENT (VERSION 2002-1) ***************2637 1451 for All Data Greater than 1.0000E-10 barns in Absolute Value 2637 1451 Data Linearized to within an Accuracy of .100000000 per-cent 2637 1451 ***************** Program SIGMA1 (VERSION 2002-1) ***************2637 1451 Data Doppler Broadened to 300.000000 Kelvin 2637 1451 for All Data Greater than 1.0000E-10 barns in Absolute Value 2637 1451 Data Linearized to Within an Accuracy pf .100000000 per-cent 2637 1451 ***************** Program FIXUP (Version 2002-1) ****************2637 1451 Corrected ZA/AWR in All Sections-----------------------------Yes 2637 1451 Corrected Thresholds-----------------------------------------Yes 2637 1451 Extended Cross Sections to 20 MeV----------------------------No 2637 1451 Allow Cross Section Deletion---------------------------------No 2637 1451 Allow Cross Section Reconstruction---------------------------No 2637 1451 Make All Cross Sections Non-Negative-------------------------Yes 2637 1451 Delete Energies Not in Ascending Order-----------------------Yes 2637 1451 Deleted Duplicate Points-------------------------------------Yes 2637 1451 Check for Ascending MAT/MF/MT Order--------------------------Yes 2637 1451 Check for Legal MF/MT Numbers--------------------------------Yes 2637 1451 Allow Creation of Missing Sections---------------------------No 2637 1451 Allow Insertion of Energy Points-----------------------------No 2637 1451 Create Uniform Energy Grid-----------------------------------No 2637 1451 Delete Section if Cross Section =0 at All Energies-----------Yes 2637 1451 ***************** Program GROUPIE (VERSION 2002-1) **************2637 1451 Unshielded Group Averages Using 640 Groups 2637 1451 Weighting Spectrum: Flat (Constant) Spectrum 2637 1451 1 451 169 12637 1451 3 102 217 12637 1451 33 102 200 12637 1451 2637 1 0 2637 0 0 2.60580E+4 5.74356E+1 0 0 0 02637 3102 6.58043E+6 6.58043E+6 0 0 1 6412637 3102 641 1 2637 3102 .000100000 20.4392055 .000105000 19.9625958 .000110000 19.50981102637 3102 .000115000 19.0925823 .000120000 18.6048355 .000127500 18.06369282637 3102 .000135000 17.5692785 .000142500 17.1149218 .000150000 16.62318192637 3102 .000160000 16.1109172 .000170000 15.6448196 .000180000 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8.324920-4 5.325110-4263733102 1.729720-4-2.437260-4-1.493790-4 1.608010-3 1.626470-3 1.640970-3263733102 1.452110-3 1.526050-3 1.497130-3 1.653870-3 1.591680-3 1.626360-3263733102 1.745200-3 1.865800-3 1.942930-3 1.967390-3 1.888840-3 1.747240-3263733102 1.496650-3 1.273130-3 1.079570-3 1.216020-3 2.115460-3 1.755610-3263733102 1.771520-3 1.369290-3 1.601530-3 1.556760-3 1.699100-3 1.631830-3263733102 1.746660-3 1.934560-3 2.127160-3 2.235890-3 2.278390-3 2.201960-3263733102 1.900010-3 1.288550-3 7.795860-4 2.641850-4 0.000000+0 9.788250-4263733102 1.962490-3 1.117600-3 1.693930-3 1.608710-3 1.609390-3 1.572910-3263733102 1.792730-3 2.158490-3 2.513250-3 2.720840-3 2.784210-3 2.693820-3263733102 2.221200-3 1.039010-3 9.561710-5-8.607330-4-1.822950-3-9.269570-4263733102 7.089630-3 1.465430-3 1.637290-3 2.893500-3 2.716890-3 2.003500-3263733102 1.195020-3 5.785340-4 3.649000-4 4.572150-4 7.079900-4 1.750060-3263733102 4.057770-3 5.866950-3 7.838270-3 1.067520-2 1.038370-2 1.780840-3263733102 1.792260-3 2.161630-3 2.263660-3 2.346080-3 2.495750-3 2.601530-3263733102 2.643620-3 2.603900-3 2.520760-3 2.290590-3 1.852580-3 1.554630-3263733102 1.273120-3 9.360720-4 6.909550-4 1.962490-3 2.565350-3 2.718850-3263733102 2.738130-3 2.737270-3 2.706870-3 2.617010-3 2.543910-3 2.452930-3263733102 2.311910-3 2.188750-3 2.172640-3 2.191620-3 2.157160-3 1.411970-3263733102 4.664890-3 4.921800-3 4.558810-3 3.928810-3 3.214320-3 2.671990-3263733102 2.315360-3 2.166760-3 2.369560-3 3.283990-3 4.245070-3 5.342850-3263733102 6.652700-3 4.731700-3 5.685160-3 5.591200-3 5.274010-3 4.850020-3263733102 4.581470-3 4.495990-3 4.685950-3 5.286950-3 4.999010-3 5.436770-3263733102 6.162770-3 6.981340-3 4.208060-3 7.242010-3 8.144960-3 8.988140-3263733102 9.918580-3 1.092380-2 1.199030-2 1.333340-2 8.692290-3 6.805680-3263733102 5.439890-3 4.040780-3 8.571640-4 1.102500-2 1.306150-2 1.509890-2263733102 1.724460-2 1.918590-2 2.120880-2 1.205000-2 7.614630-3 3.937350-3263733102 1.748330-4-3.032670-3 1.716100-2 2.007790-2 2.296380-2 2.600090-2263733102 2.868950-2 1.526500-2 8.626110-3 2.794700-3-2.965560-3-6.111480-3263733102 2.496400-2 2.837000-2 3.205500-2 3.559640-2 1.889820-2 1.035410-2263733102 2.874230-3-4.179100-3-6.893270-3 3.422500-2 3.780380-2 4.258630-2263733102 2.280370-2 1.270380-2 4.104470-3-3.570120-3-5.771380-3 4.452100-2263733102 4.894050-2 2.666180-2 1.559110-2 6.239220-3-1.641400-3-3.268030-3263733102 5.904900-2 3.828800-2 2.968610-2 2.309590-2 2.096800-2 2.670880-2263733102 6.150400-2 7.269370-2 8.635430-2 1.136110-1 1.503490-1 1.004890-1263733102 1.229590-1 1.670490-1 2.212880-1 1.648360-1 2.211800-1 2.929660-1263733102 3.226240-1 4.279660-1 6.352090-1 263733102 263733 0 2637 0 0 0 0 0 2.70590E+4 5.84269E+1 0 0 34 102725 1451 0.0 0.0 0 0 0 62725 1451 1.00000E+0 2.00000E+7 0 0 10 20022725 1451 3.00000E+2 0.0 1 0 233 72725 1451 27-Co- 59 IRK/FEI EVAL-APR90 VONACH ET AL., K.I.Zolotarev 2725 1451 DIST-Feb2004 2725 1451 ----IRDF-2002 MATERIAL 2725 2725 1451 -----INCIDENT NEUTRON DATA 2725 1451 ------ENDF-6 FORMAT 2725 1451 ******************************************************************2725 1451 27-CO- 59 ANL,ORNL EVAL-JUL89 A.SMITH+,G.DESAUSSURE+ 2725 1451 ANL/NDM-107 DIST-JUN93 REV1-JUN92 19930930 2725 1451 ----ENDF/B-VI MATERIAL 2725 REVISION 1 2725 1451 ******************************************************************2725 1451 *****************EXTRACT FOR SPECIAL PURPOSE FILE*****************2725 1451 DOSIMETRY 2725 1451 ******************************************************************2725 1451 ***************************************************************** 2725 1451 ********** Start of (N,G) bibliographical component *********** 2725 1451 MF=2,MT=151, MF=3,MT=1,2,102 taken from ENDF/B-VI Release 2 2725 1451 MF=33,MT=102 taken from IRDF-90 2725 1451 ***************************************************************** 2725 1451 REVISION 1 -- NEW RESONANCE REGION EVALUATION BY G.DESAUSSURE, 2725 1451 N.M.LARSON, J.A.HARVEY, N.W.HILL, ANN. NUCL. ENERGY 19, 393 2725 1451 (1992) 2725 1451 ORIGINAL EVALUATION 2725 1451 A.SMITH,D.SMITH,P.GUENTHER,J.MEADOWS, R.LAWSON(ANL), 2725 1451 R.HOWERTON(LLNL),M.SUGIMOTO(JAERI) 2725 1451 ******************************************************************2725 1451 ENDF/B-V MATERIAL CONVERTED TO ENDF-6 FORMAT BY NNDC 2725 1451 * * * * * * *2725 1451 R.LAWSON(ANL),R.HOWERTON(LLNL),M.SUGIMOTO(JAERI) 2725 1451 ***************************************************************** 2725 1451 ********** End of (N,G) bibliographical component *********** 2725 1451 ***************************************************************** 2725 1451 ***************************************************************** 2725 1451 ********** Start of (N,2N) bibliographical component ********** 2725 1451 ***************************************************************** 2725 1451 ++++++++++++++++++++++++++++++++++++++++++++++++++++++++++++++++++2725 1451 MF/MT 2151, MF/MT 3 16 AND MF/MT 33 16 FROM FOLLOWING 2725 1451 EVALUATION 2725 1451 27-CO- 59 IRK-VIENNA EVAL-APR90 2725 1451 DIST-JUN90 2725 1451 IRK-EVAL.NLIB 25 2725 2725 1451 ++++++++++++++++++++++++++++++++++++++++++++++++++++++++++++++++++2725 1451 ***************************************************************** 2725 1451 ********** End of (N,2N) bibliographical component *********** 2725 1451 ***************************************************************** 2725 1451 ***************************************************************** 2725 1451 ********** Start of (N,A) bibliographical component ********** 2725 1451 ***************************************************************** 2725 1451 ***************************************************************** 2725 1451 ------Russian Reactor Dosimetry File RRDF-2002 2725 1451 ***************************************************************** 2725 1451 Author of evaluation: K.I.Zolotarev 2725 1451 ***************************************************************** 2725 1451 MF=3 2725 1451 MT=107 -(n,a) cross section 2725 1451 ------------------------------------- 2725 1451 Excitation function for the Co-59(n,a)Mn-56 reaction in the 2725 1451 energy region from 2.5 MeV to 21 MeV was evaluated by means of 2725 1451 statistical analysis of experimental cross section data [1-28], 2725 1451 and data from GNASH [29] calculation. 2725 1451 All experimental data were renormalized to the new standards 2725 1451 for monitor reactions cross sections and decay data. Uncertainty 2725 1451 in the monitor reaction cross section was added to the total 2725 1451 uncertainty for Santry and Butler data [5]. Cross section data 2725 1451 measured by Huang Jianzhou et al. [18] were renormalized to the 2725 1451 preliminary evaluated integral of excitation function in the 2725 1451 energy range 13.0 - 15.0 MeV. Correction factor for this data 2725 1451 was equal Fc= 1.0507 . 2725 1451 Experimental data from ref.[30-41] were rejected due to their 2725 1451 discrepancy with the main bulk of experimental data [1-28] and 2725 1451 data from theoretical model calculation. 2725 1451 The final procedure of evaluation Co-59(n,a)Mn-56 excitation 2725 1451 function from threshold to 21 MeV has been carried out within the 2725 1451 framework of generalized least squares method. Rational function 2725 1451 was used as model function [42]. Calculations was performed by 2725 1451 means of Pade-2 code [43]. 2725 1451 The evaluated Co-59(n,a)Mn-56 excitation function averaged 2725 1451 on U-235 neutron fission spectrum [44] and Cf-252 spontaneous 2725 1451 fission neutron spectrum [45] gives the next values : 2725 1451 ÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÂÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÂÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄ 2725 1451 TYPE OF SPECTRUM ³ ,mb (calc.) ³ , mb (measured) 2725 1451 ÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÅÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÅÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄ 2725 1451 ³ ³ 0.1500 +- 0.0080 [46] 2725 1451 U-235 neutron fission ³ 0.15823 ³ 0.1700 +- 0.0120 [47] 2725 1451 ³ ³ 0.1568 +- 0.0035 [48] 2725 1451 ÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÅÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÅÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄ 2725 1451 ³ ³ 0.2208 +- 0.0014 [49] 2725 1451 Cf-252 spont. fission ³ 0.22095 ³ 0.2221 +- 0.0023 [50] 2725 1451 ³ ³ 0.2218 +- 0.0041 [51] 2725 1451 ÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÁÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÁÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄ 2725 1451 2725 1451 MT=33 2725 1451 MT=107 -(n,a) cross section cov. matrix 2725 1451 ------------------------------------------- 2725 1451 Uncertainties in the evaluated excitation function for the 2725 1451 reaction Co-59(n,a)Mn-56 are given in the form of relative cova- 2725 1451 riance matrix for the 32-neutron energy groups (LB=5). Covariance 2725 1451 matrix of uncertainties was calculated simultaneously with recom- 2725 1451 mended cross section data by means of PADE-2 code. 2725 1451 Eigenvalues of the 6-th digits relative covariance matrix 2725 1451 given in the 33-file are the following: 2725 1451 2725 1451 1.11275E-08 1.13832E-08 1.19424E-08 1.27401E-08 2725 1451 1.37409E-08 1.52188E-08 1.67720E-08 1.86812E-08 2725 1451 2.09005E-08 2.38172E-08 2.69210E-08 3.07595E-08 2725 1451 3.46853E-08 4.07806E-08 4.51537E-08 5.55085E-08 2725 1451 6.19014E-08 7.83507E-08 9.39864E-08 1.01329E-07 2725 1451 2.19793E-07 8.71735E-07 1.71048E-05 1.93185E-04 2725 1451 1.05065E-03 1.19481E-03 2.01786E-03 2.53733E-03 2725 1451 3.28491E-03 1.06037E-02 2.32281E-02 9.30559E-01 2725 1451 2725 1451 References : 2725 1451 1. H.G.Blosser et.al. Phys. Rev., v.110, p.531, 1958 2725 1451 2. I.L.Preiss,R.W.Fink Nucl. Phys., v.15, p.326, February 1960 2725 1451 3. C.S.Khurana, H.S.Hans Proc.of 4th Nuclear Physics and Solid 2725 1451 State Physics Symp., 24-26 Feb. 1960, Waltair, India, p.297 2725 1451 4. E.Weigold Australian J. Phys., v.13, p.186, 1960 2725 1451 5. D.C.Santry, J.P.Butler Can. J. Phys., v.42, p.1030, 1964 2725 1451 6. H.Liskien, A.Paulsen J. Nucl. of Energy, v.19, p.73, 1965 2725 1451 7. K.S.Khurana,I.M.Govil Nucl. Phys., v.69, p.153, July 1965 2725 1451 8. H.Liskien, A.Paulsen Nucleonics, v.8, p.315, June 1966 2725 1451 9. V.N.Levkovskiy et.al. Yadernaja Fizika (Sov.), v.8, n.1, p.7, 2725 1451 July 1968 2725 1451 10. J.C.Robertson, B.Audric, P.Kolkowski J. Nucl. Energ., v.27, 2725 1451 p.531, August 1973 2725 1451 11. S.K.Ghorai, J.E.Gaiser, W.L.Alford Annals of Nucl. Energy, 2725 1451 v.7, p.41, 1980 2725 1451 12. E.Zupranska et al. Progress Report INR-1821/I/PL/A, 1980 2725 1451 E.Zupranska et al. Acta. Phys. Pol., v.B11, p.853, Nov. 1980 2725 1451 13. B.M.Bahal, R.Pepelnik Report GKSS-84-E-, 1984 2725 1451 14. J.W.Meadows, D.L.Smith et al. Annals of Nucl. Energy, v.14, 2725 1451 p.489, 1987 2725 1451 15. J.W.Meadows, D.L.Smith, R.D.Lawson Annals of Nucl. Energy, 2725 1451 v.14, p.603, 1987 2725 1451 16. Y.Ikeda, C.Konno, K.Oishi et al. Report JAERI-1312, 1988 2725 1451 17. T.B.Ryves, P.Kolkowski, S.M.Judge Annals of Nuclear Energy, 2725 1451 v.15, p.561, Dec.1988 2725 1451 18. Huang Jianzhou et al. Report INDC(CPR)-16, Vienna,August 1989 2725 1451 19. I.Kimura, K.Kobayashi Nucl. Sci. Eng., v.106, p.332, 1990 2725 1451 20. Li Tingyan et al. High Energy Physics and Nuclear Physics, 2725 1451 v.14(6), p.542, June 1990 2725 1451 21. A.Ercan et al. Proc. of an Intern. Conf. on Nuclear Data for 2725 1451 Science and Technology, 13-17 May 1991, Julich, FRG, Springer-2725 1451 Verlag, 1992, p.376-377 2725 1451 22. Y.Ikeda et al. Progress Report INDC(JPN)-162/U,p.24, Aug.1992 2725 1451 23. A.Grallert et al. Report INDC(NDS)-286, p.193, IAEA, 1993 2725 1451 24. W.Mannhart et al. Proc.of International Conference on Nuclear 2725 1451 Data for Science and Technology, Gatlinburg, Tennessee, May 2725 1451 9-13, 1994, v.1, pp.285-287 2725 1451 25. A.A.Filatenkov et al. VANT, Ser.:Yadernye Konstanty, v.2, p.8,2725 1451 Moscow, 1996 2725 1451 26. K.T.Osman, F.I.Habbani Report INDC(SUD)-001, Distr.L, IAEA, 2725 1451 Vienna, October 1996 2725 1451 27. A.D.Majdeddin Measurement and analysis of excitation func- 2725 1451 tions for fast neutron induced reactions from threshold to 2725 1451 20 MeV. PHD dissertation, Inst. of Experimental Physics , 2725 1451 Kossuth University, Hungary, June 1998 2725 1451 28. A.A.Filatenkov et al. Report RI-252, St. Petersburg, May 1999 2725 1451 29. P.G.Young, E.D.Arthur A Preequilibrium Statistical Nuclear 2725 1451 Model Code for Calculation of Cross Section and Emission 2725 1451 Spectra. Report LA-6947, Los Alamos, 1977 ; 2725 1451 E.L.Trykiv, G.Ya.Tertychnyi Private communication, IPPE, 2725 1451 Obninsk, May 1999 2725 1451 30. E.B.Paul,R.L.Clarke Canad. J. of Phys., v.31, p.267, 1953 2725 1451 31. M.Bormann et al. Journal de Physique, v.22, p.602, Oct. 1961 2725 1451 32. F.Gabbard, B.D.Kern Phys. Rev., v.128, n.3, p.1276, 1962 2725 1451 33. J.M.F.Jeronymo et al. Nucl. Phys., v.47(1), p.157, July 1963 2725 1451 34. J.E.Strain, W.J.Ross Report ORNL-3672, January 1965 2725 1451 35. M.Bormann et al. Nucl. Phys., v.63, p.438, March 1965 2725 1451 36. U.Garuska, J.Dresler, H.Malecki Prog. Report INR-1773/I/PL/A,2725 1451 p.16, September 1978 2725 1451 37. H.M.Agrawal, K.R.Zasadny, G.F.Knoll Trans. Amer. Nucl. Soc., 2725 1451 v.47, p.431, November 1984 save 2725 1451 38. M.Berrada Progress Report to the IAEA NDS on Research 2725 1451 Contract -3311.R1/RB, 15 April 1984 2725 1451 39. I.Garlea et al. Revue Roumaine de Physique, v.30, p.673, 1985 2725 1451 40. N.I.Molla et al. Proc.of International Conference on Nuclear 2725 1451 Data for Science and Technology, Gatlinburg, Tennessee, May 2725 1451 9-13, 1994, v.2, pp.938-940 2725 1451 41. R.Doczi et al. Nucl. Sci. Eng., v.129, p.164, June 1998 2725 1451 42. S.Badikov, N.Rabotnov, K.Zolotarev Proc. of NEANSC Speciali- 2725 1451 st's Meeting on Evaluation and Processing of Covariance Data, 2725 1451 Oak Ridge, USA, 7-9 September 1992, OECD, Paris, 1993, p.105 2725 1451 43. S.A.Badikov et.al. Preprint FEI-1686, Obninsk, 1985 2725 1451 44. L.W.Weston et al. Evaluated Neutron Data for U-235, ENDF/B-VI 2725 1451 Library, MAT=9228, MF=5, MT=18, eval. April 1989 2725 1451 45. W.Mannhart IAEA-TECDOC-410, p.158, IAEA, Vienna, 1987 2725 1451 46. K.Kobayashi, I.Kimura Report NEANDC(J)-61U, p.81, Sep. 1979 2725 1451 47. O.Horibe et al. Proc. of Conf.:50 Years with Nuclear Fission, 2725 1451 Washington D.C., 25-28 April 1989, v.2, p.923 2725 1451 48. W.Mannhart Progress Report INDC(Ger)-045, pp.40-43, 1999 2725 1451 49. K.Kobayashi,I.Kimura,W.Mannhart Nucl. Sci. Technology.,v.19, 2725 1451 n.5, p.341, May 1982 2725 1451 50. W.Mannhart Handbook on Nuclear Activation Data , IAEA 2725 1451 Technical Report series No.273, p.413, 1987 2725 1451 51. W.Mannhart Validation of Differential Cross Sections with 2725 1451 Integral Data , Report INDC(NDS)-435, pp.59-64, IAEA, Vienna, 2725 1451 September 2002 2725 1451 ***************************************************************** 2725 1451 ********** End of (N,A) bibliographical component ********** 2725 1451 ***************************************************************** 2725 1451 The Q values and threshold energies were updated prior to pro- 2725 1451 cessing through the codes to comply with the values obtained 2725 1451 using the NNDC calculation program which is based on the 1995 2725 1451 Update to the Atomic mass Evaluation. 2725 1451 ***************************************************************** 2725 1451 ***************** Program LINEAR (VERSION 2002-1) ***************2725 1451 For All Data Greater than 1.0000E-10 barns in Absolute Value 2725 1451 Data Linearized to Within an Accuracy of .100000000 per-cent 2725 1451 ***************** Program RECENT (VERSION 2002-1) ***************2725 1451 for All Data Greater than 1.0000E-10 barns in Absolute Value 2725 1451 Data Linearized to within an Accuracy of .100000000 per-cent 2725 1451 ***************** Program SIGMA1 (VERSION 2002-1) ***************2725 1451 Data Doppler Broadened to 300.000000 Kelvin 2725 1451 for All Data Greater than 1.0000E-10 barns in Absolute Value 2725 1451 Data Linearized to Within an Accuracy pf .100000000 per-cent 2725 1451 ***************** Program FIXUP (Version 2002-1) ****************2725 1451 Corrected ZA/AWR in All Sections-----------------------------Yes 2725 1451 Corrected Thresholds-----------------------------------------Yes 2725 1451 Extended Cross Sections to 20 MeV----------------------------No 2725 1451 Allow Cross Section Deletion---------------------------------No 2725 1451 Allow Cross Section Reconstruction---------------------------No 2725 1451 Make All Cross Sections Non-Negative-------------------------Yes 2725 1451 Delete Energies Not in Ascending Order-----------------------Yes 2725 1451 Deleted Duplicate Points-------------------------------------Yes 2725 1451 Check for Ascending MAT/MF/MT Order--------------------------Yes 2725 1451 Check for Legal MF/MT Numbers--------------------------------Yes 2725 1451 Allow Creation of Missing Sections---------------------------No 2725 1451 Allow Insertion of Energy Points-----------------------------No 2725 1451 Create Uniform Energy Grid-----------------------------------No 2725 1451 Delete Section if Cross Section =0 at All Energies-----------Yes 2725 1451 ***************** Program GROUPIE (VERSION 2002-1) **************2725 1451 Unshielded Group Averages Using 640 Groups 2725 1451 Weighting Spectrum: Flat (Constant) Spectrum 2725 1451 1 451 244 12725 1451 3 16 35 12725 1451 3 102 217 12725 1451 3 107 75 12725 1451 33 16 42 12725 1451 33 102 11 12725 1451 33 107 103 12725 1451 2725 1 0 2725 0 0 2.70590E+4 5.84269E+1 0 0 0 02725 3 16 -1.04530E+7-1.04530E+7 0 0 1 952725 3 16 95 1 2725 3 16 10600000.0 .001438929 10700000.0 .007329788 10800000.0 .0135365962725 3 16 10900000.0 .026794286 11000000.0 .047102857 11100000.0 .0674114292725 3 16 11200000.0 .088045429 11300000.0 .110632000 11400000.0 .1335440002725 3 16 11500000.0 .156456000 11600000.0 .179368000 11700000.0 .2030660002725 3 16 11800000.0 .231480000 11900000.0 .260680000 12000000.0 .2898800002725 3 16 12100000.0 .319080000 12200000.0 .348198750 12300000.0 .3768300002725 3 16 12400000.0 .405380000 12500000.0 .433930000 12600000.0 .4624800002725 3 16 12700000.0 .489566125 12800000.0 .507869000 12900000.0 .5247080002725 3 16 13000000.0 .541547000 13100000.0 .558386000 13200000.0 .5752250002725 3 16 13300000.0 .592064000 13400000.0 .608903000 13500000.0 .6257420002725 3 16 13600000.0 .642581000 13700000.0 .658802125 13800000.0 .6713160002725 3 16 13900000.0 .683212000 14000000.0 .695108000 14100000.0 .7087826252725 3 16 14200000.0 .726805000 14300000.0 .744663875 14400000.0 .7545780002725 3 16 14500000.0 .761242000 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DIST-MAR85 850320 2800 1451 ----IRDF-90/NMF-G MATERIAL 2800 2800 1451 -----INCIDENT NEUTRON DATA 2800 1451 ------ENDF-6 FORMAT 2800 1451 2800 1451 DAMAGE CROSS SECTIONS OF NICKEL IN STEEL 2800 1451 ASTM STANDARD 2800 1451 2800 1451 ***************** PROGRAM LINEAR (VERSION 84-2) *****************2800 1451 DATA LINEARIZED TO WITHIN AN ACCURACY OF 0.100 PER-CENT 2800 1451 **************** PROGRAM GROUPIE (VERSION 84-2) *****************2800 1451 UNSHIELDED GROUP AVERAGES USING 640 GROUPS (SAND-II EXTEND) 2800 1451 2800 1451 ##################################################################2800 1451 MODIFICATIONS RELATED TO THE VERSION IRDF-90/NMF-G ARE DESCRIBED 2800 1451 IN THE REPORT INDC(HUN)-31/G+R BY E. J. 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3900 1.72000+ 7 3.24600+ 3 1.73000+ 7 3.24600+ 3 1.74000+ 7 3.24600+ 32800 3900 1.75000+ 7 3.24600+ 3 1.76000+ 7 3.24600+ 3 1.77000+ 7 3.24600+ 32800 3900 1.78000+ 7 3.24600+ 3 1.79000+ 7 3.24600+ 3 1.80000+ 7 3.33500+ 32800 3900 1.81000+ 7 3.33500+ 3 1.82000+ 7 3.33500+ 3 1.83000+ 7 3.33500+ 32800 3900 1.84000+ 7 3.33500+ 3 1.85000+ 7 3.33500+ 3 1.86000+ 7 3.33500+ 32800 3900 1.87000+ 7 3.33500+ 3 1.88000+ 7 3.33500+ 3 1.89000+ 7 3.33500+ 32800 3900 1.90000+ 7 3.37500+ 3 1.91000+ 7 3.37500+ 3 1.92000+ 7 3.37500+ 32800 3900 1.93000+ 7 3.37500+ 3 1.94000+ 7 3.37500+ 3 1.95000+ 7 3.37500+ 32800 3900 1.96000+ 7 3.37500+ 3 1.97000+ 7 3.37500+ 3 1.98000+ 7 3.37500+ 32800 3900 1.99000+ 7 3.37500+ 3 2.00000+ 7 0.0 + 0 2800 3900 2800 3 0 2800 0 0 0 0 0 2.80580E+4 5.74376E+1 0 0 34 102825 1451 0.0 0.0 0 0 0 62825 1451 1.00000E+0 2.00000E+7 0 0 10 20022825 1451 3.00000E+2 0.0 1 0 378 52825 1451 28-Ni- 58 IRK-IJS/FEIEVAL-AUG99 EUR.JOINT COLLAB. K.I.Zolotarev 2825 1451 DIST-Feb2004 2825 1451 ----IRDF-2002 MATERIAL 2825 2825 1451 -----INCIDENT NEUTRON DATA 2825 1451 ------ENDF-6 FORMAT 2825 1451 2825 1451 ***************************** JEFF-3.0 ***********************2825 1451 *****************EXTRACT FOR SPECIAL PURPOSE FILE*****************2825 1451 DOSIMETRY 2825 1451 ******************************************************************2825 1451 28-NI- 58 IRK-IJS EVAL-AUG99 EUROPEAN JOINT COLLABORATION 2825 1451 2825 1451 DATA TAKEN FROM :- EFF-3.1 (DIST-AUG99 REV1-SEP00) 2825 1451 MF=3 2825 1451 MT= 16 - (n,2n) 2825 1451 2825 1451 ******************************************************************2825 1451 28-Ni- 58 FEI EVAL-Dec02 K.I.Zolotarev 2825 1451 DIST-Jan03 2825 1451 ----BROND-3 MATERIAL 2825 2825 1451 -----INCIDENT NEUTRON DATA 2825 1451 ------ENDF-6 FORMAT 2825 1451 ***************************************************************** 2825 1451 ------Russian Reactor Dosimetry File RRDF-2002 2825 1451 ***************************************************************** 2825 1451 Author of evaluation: K.I.Zolotarev 2825 1451 ***************************************************************** 2825 1451 MF=3 2825 1451 MT=103 - Ni58(n,p)Co58m+g reaction 2825 1451 2825 1451 ***************************************************************** 2825 1451 ******* Start of JEFF-3.0 (N,2N) bibliographical component ****** 2825 1451 ***************************************************************** 2825 1451 ***************************** JEFF-3.0 ***********************2825 1451 28-NI- 58 IRK-IJS EVAL-AUG99 EUROPEAN JOINT COLLABORATION 2825 1451 2825 1451 DATA TAKEN FROM :- EFF-3.1 (DIST-AUG99 REV1-SEP00) 2825 1451 2825 1451 ******************************************************************2825 1451 Authors and Responsibilities: 2825 1451 S.Tagesen, H.Vonach and A.Wallner, I.R.K.: 2825 1451 - Complete evaluation of the cross sections including covariance 2825 1451 matrices by generalized least squares cross section update 2825 1451 code GLUCS (9, 10). The total cross section is evaluated in 2825 1451 broad energy bins. 2825 1451 A.Trkov, I.J.S.: 2825 1451 - Final assembly of the file. 2825 1451 - Consistency corrections (interp.law in MF6 starter file). 2825 1451 - Implementation of the resonance fluctuations on the smooth 2825 1451 newly evaluated cross sections. 2825 1451 - Preliminary data verification and benchmarking. 2825 1451 2825 1451 Evaluation Details: 2825 1451 The Oak Ridge Ni-58 ENDF/B-VI Revision 1 evaluation (MAT 2825) 2825 1451 by Larson et.al. was the chosen starter file. All neutron 2825 1451 cross sections above 810 keV were re-evaluated and include 2825 1451 covariance data. The following data sets were selected as 2825 1451 "priors": 2825 1451 MT16 IRDF-90 (evaluation Pavlik)(11) 2825 1451 2825 1451 Resonance Fluctuations in the Cross Sections: 2825 1451 The cross sections below 810 keV are not affected by the re- 2825 1451 evaluation process. Above this energy, broad bin average total 2825 1451 cross section from the starter file was calculated. The bins 2825 1451 correspond to those used in the re-evaluation process. A smooth 2825 1451 cross section curve was generated, conserving the bin average 2825 1451 values. Fluctuations modulating function was defined as the 2825 1451 ratio of the original and the smoothed cross section. This 2825 1451 modulating function is applied on the re-evaluated smooth total 2825 1451 cross section and the inelastic cross sections. 2825 1451 2825 1451 The remaining comments are taken over from the starter file, 2825 1451 except for the sections referring to the data, which have been 2825 1451 superseeded. Modified sections are identified by the use of 2825 1451 lower case characters. 2825 1451 2825 1451 ******************************************************************2825 1451 CAPTURE WIDTHS CORRECTED FOR 58.7 AND 439.52 KEV RESONANCES. 2825 1451 THE ELASTIC TRANSFORMATION MATRIX WAS REMOVED. 2825 1451 FIXED TYPO IN MINIMUM ENERGY IN MF=6, MT=51 2825 1451 ******************************************************************2825 1451 2825 1451 THIS WORK EMPLOYED NUCLEAR MODEL CODES INCLUDING THE 2825 1451 DISTORTED WAVE BORN APPROXIMATION (DWBA) PROGRAM DWUCK (1) 2825 1451 AND THE HAUSER-FESHBACH CODE TNG (2,3,4). THE TNG CODE PROVIDES 2825 1451 ENERGY AND ANGULAR DISTRIBUTIONS OF PARTICLES EMITTED IN THE 2825 1451 COMPOUND AND PRE-COMPOUND REACTIONS, ENSURES CONSISTENCY AMONG ALL2825 1451 REACTIONS, AND MAINTAINS ENERGY BALANCE. DETAILS PERTINENT TO THE 2825 1451 CONTENTS OF THIS EVALUATION AND EXTENSIVE COMPARISONS OF 2825 1451 CALCULATIONS WITH EXPERIMENTAL DATA CAN BE FOUND IN REFERENCE (5).2825 1451 2825 1451 ----- DESCRIPTION OF FILES 2825 1451 (MF-MT) 2825 1451 1-451 GENERAL INFORMATION, REFERENCES, AND DEFINITIONS. 2825 1451 3-16 (n,2n) cross sections were re-evaluated. 2825 1451 -------------------------------------------------------------- 2825 1451 UNCERTAINTY FILES 2825 1451 ALL NON-DERIVED FILES CONTAIN AN LB=8 COMPONENT, AS 2825 1451 REQUIRED BY ENDF/B-VI FORMATS 2825 1451 For all evaluated reactions full covariance matrices are 2825 1451 given as calculated by the bayesian evaluation update code2825 1451 GLUCS. This includes full inter-reaction covariances. 2825 1451 2825 1451 33-16 (N,2N) covariances from GLUCS. 2825 1451 2825 1451 REFERENCES: 2825 1451 (11) M. Wagner et al., Physics Data 13-5, Fachinformationszentrum 2825 1451 Karlsruhe, 1990 2825 1451 ***************************************************************** 2825 1451 ******* End of JEFF-3.0 (N,2N) bibliographical component ******* 2825 1451 ***************************************************************** 2825 1451 ***************************************************************** 2825 1451 ******* Start of RRDF (N,P) bibliographical component ******* 2825 1451 ***************************************************************** 2825 1451 28-Ni- 58 FEI EVAL-Dec02 K.I.Zolotarev 2825 1451 DIST-Jan03 2825 1451 ----BROND-3 MATERIAL 2825 2825 1451 -----INCIDENT NEUTRON DATA 2825 1451 ------ENDF-6 FORMAT 2825 1451 ***************************************************************** 2825 1451 ------Russian Reactor Dosimetry File RRDF-2002 2825 1451 ***************************************************************** 2825 1451 Author of evaluation: K.I.Zolotarev 2825 1451 ***************************************************************** 2825 1451 MF=3 2825 1451 MT=103 - Ni58(n,p)Co58m+g reaction 2825 1451 ------------------------------------- 2825 1451 Microscopic experimental data [1-64] were analyzed in the 2825 1451 process of preparation of input data base for the evaluation of 2825 1451 cross sections and their uncertainty for the Ni-58(n,p)Co-58m+g 2825 1451 reaction. During this procedure all experimental data if it was 2825 1451 possible were corrected to the new recommended cross section data 2825 1451 for monitor reactions used in the measurements and to the new re- 2825 1451 commended decay data. 2825 1451 Excitation function for the Ni-58(n,p)Co-58m+g reaction in 2825 1451 the energy region from threshold to 20 MeV was evaluated by means 2825 1451 of statistical analysis of experimental cross section data [1-41].2825 1451 Special correction was done with experimental data [9], [12], 2825 1451 [13], [15], [16], [20], [23], [26], [34], [35]. Experimental data 2825 1451 of Decowski et al. [9] corresponding to the neutron energy region 2825 1451 12.42 - 17.08 MeV were renormalized to the preliminary evaluated 2825 1451 cross section value at En=14.77 MeV. Original cross sections from 2825 1451 ref. [9] were multiplied to the correction factor Fc=0.53807 . 2825 1451 Data of Paulsen and Widera [12] measured in the energy ranges 2825 1451 3.35 - 6.27 MeV and 12.8 - 16.4 MeV were corrected to the preli- 2825 1451 minary evaluated cross section value at 6.3 MeV and integral of 2825 1451 cross section in the energy interval 14-15 MeV. The correction 2825 1451 factors for the experimental data [12] were Fc=1.059 and Fc=1.058,2825 1451 respectively. 2825 1451 Data of Smith and Meadows [13] measured with using neutrons 2825 1451 from D(d,n)He3 reaction were renormalized to the results of this 2825 1451 experiment obtained with Li7(p,n)Be7 neutron source in the over- 2825 1451 lapping interval 5.398 - 5.870 MeV. D(d,n)He3 data in the energy 2825 1451 range 5.398 - 9.897 MeV were increased to the factor Fc=1.083 . 2825 1451 Data of Wu and Chou [15] and Husain and Hunt [20] were also 2825 1451 renormalized to the Smith and Meadows measurements carried out 2825 1451 with Li7(p,n)Be7 neutron source. Correction factors for experi- 2825 1451 mental data were Fc=1.04592 [15] and Fc=1.03634 [20]. 2825 1451 Cross sections measured by Hudson et al. [16] in the energy 2825 1451 range 13.3 - 17.10 MeV were corrected to Filatenkov et al. experi-2825 1451 mental data at 14.1 MeV [40], Fc=0.95367. 2825 1451 Experimental data of Rondio et al. [23,26] were corrected to 2825 1451 the preliminary evaluated Ni-58(n,p)Co-58m+g reaction cross sec- 2825 1451 tion integral in the energy interval 17.3 - 18.5 MeV (Fc=0.1889). 2825 1451 Data of Yuan Junqian et al. [34] and Li Tingyan et al. [35] 2825 1451 were multiplied to the factors Fc=0.9000 and Fc=0.9259 , respec- 2825 1451 tively, using preliminary evaluated cross section integral in the 2825 1451 energy interval 13.6 - 14.6 MeV. 2825 1451 Experimental cross section data [42-64] were rejected due to 2825 1451 their discrepancy with the main bulk of experimental data [1-41]. 2825 1451 Data of Decowski et al. [9] obtained in the neutron energy range 2825 1451 from 1.95 to 4.79 MeV were rejected for the same reason. In the 2825 1451 rejected experiments [42-43], [51-55], [57-60], [62-64] the cross 2825 1451 section values were measured only in a one energy point in the 2825 1451 interval 14 - 15 MeV. 2825 1451 Statistical analysis of input cross section data was carried 2825 1451 out by means of PADE-2 code [65]. Rational function was used as 2825 1451 the model function [66]. 2825 1451 U-235 thermal fission [67] and Cf-252 spontaneous fission 2825 1451 neutron spectra [68] averaged cross sections calculated from the 2825 1451 the evaluated Ni-58(n,p)Co-58m+g excitation function are the 2825 1451 following: 2825 1451 2825 1451 ÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÂÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÂÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄ 2825 1451 TYPE OF SPECTRUM ³ ,mb (calc.) ³ , mb (measured) 2825 1451 ÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÅÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÅÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄ 2825 1451 U-235 neutron fission ³ 107.44 ³ 108.9 +- 5.2 [69] 2825 1451 ³ ³ 106.0 +- 7.0 [70] 2825 1451 ³ ³ 108.5 +- 1.4 [71] 2825 1451 ÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÅÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÅÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄ 2825 1451 Cf-252 spont. fission ³ 117.36 ³ 117.6 +- 1.5 [72] 2825 1451 ³ ³ 117.5 +- 1.53 [73] 2825 1451 ÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÁÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÁÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄ 2825 1451 2825 1451 MT=33 2825 1451 MT=103 -(n,p) cross section cov. matrix 2825 1451 --------------------------------------- 2825 1451 Uncertainties in the evaluated excitation function for the 2825 1451 reaction Ni-58(n,p)Co-58m+g are given in the form of relative co- 2825 1451 variance matrix for the 37-neutron energy groups (LB=5). Covari- 2825 1451 ance matrix of uncertainties was calculated simultaneously with 2825 1451 recommended cross section data by means of PADE-2 code. 2825 1451 Eigenvalues of the 6-th digits relative covariance matrix 2825 1451 given in the 33-file are the following: 2825 1451 2825 1451 3.02525E-08 3.04415E-08 3.07179E-08 3.11702E-08 2825 1451 3.17809E-08 3.24784E-08 3.36976E-08 3.53449E-08 2825 1451 3.74027E-08 4.12908E-08 4.46423E-08 5.24310E-08 2825 1451 5.88770E-08 7.22029E-08 8.98365E-08 1.06273E-07 2825 1451 1.47057E-07 1.97171E-07 2.32636E-07 3.12653E-07 2825 1451 4.31346E-07 5.83111E-07 7.69089E-07 9.84635E-07 2825 1451 1.17764E-06 1.35691E-06 6.11687E-05 6.46368E-04 2825 1451 7.13990E-04 1.20451E-03 1.41032E-03 1.54898E-03 2825 1451 2.03298E-03 3.77118E-03 4.31020E-03 5.28282E-03 2825 1451 3.31282E-02 2825 1451 2825 1451 References: 2825 1451 1. I.Kumabe, R.W.Fink Nucl. Phys., v.15, p.316, February 1960 2825 1451 2. R.N.Glover, E.Weigold Nucl. Phys., v.29, p.309, 1962 2825 1451 3. J.F.Barry J. Nucl. Energy AB, v.16, p.467, 1962 2825 1451 4. W.G.Cross et al. Progress Report EANDC(CAN)-16, p.1, 2825 1451 January 1963 2825 1451 5. J.W.Meadows, J.F.Whalen Phys. Rev., v.130, p.2022, 1963 2825 1451 6. K.Debertin, E.Rossle Nucl. Phys., v.70, p.89, August 1965 2825 1451 7. M.Bormann et al. Zeitschrift f. Naturforschung, section A, 2825 1451 v.21, p.988, 1966 ; 2825 1451 8. A.Paulsen, H.Liskien Proc. of the 1-st IAEA Conference on 2825 1451 Nuclear Data for Reactors, Paris, 17-21 October 1966, p.217 2825 1451 9. P.Decowski et al. Nucl. Phys. A, v.112, p.513, May 1968 2825 1451 10. R.C.Barrall et al. Report AFWL-TR-68-134, Albuquerque, NM, 2825 1451 March 1969 2825 1451 11. R.W.Fink, W.-D.Lu Bull. Amer. Phys. Soc., v.15, p.1372(EH6), 2825 1451 July 1970 2825 1451 12. A.Paulsen, R.Widera Proc. of Conference on Chemical Nuclear 2825 1451 Data, Measurements and Applicat., Univ. of Kent, Canterbury, 2825 1451 20-22 September 1971, p.129 2825 1451 13. D.L.Smith,J.W.Meadows Nucl. Sci. Eng., v.58, p.314, Nov. 1975 2825 1451 14. R.A.Sigg Dissertation Abstracts section B, v.37, p.2237, 2825 1451 November 1976 2825 1451 15. M.W.Wu, J.C.Chou Nucl. Sci. Eng., v.63, p.268, 1977 2825 1451 16. C.G.Hudson et al. Annals of Nuclear Energy, v.5, p.589, 1978 2825 1451 17. M.T.Swinhoe, C.A.Uttley Report AERE-R-9929, September 1980 2825 1451 18. P.N.Ngoc et al. Nukleonika, v.29, p.87, 1984 2825 1451 19. P.Raics et al. Atomki Koezlemenyek, v.23, p.45, June 1981 2825 1451 20. H.A.Husain, S.E.Hunt Internat. J. of Applied Radiation and 2825 1451 Isotopes, v.34, n.4, p.731, 1983 2825 1451 21. A.Pavlik et al. Nucl. Sci. Eng., v.90, p.186, June 1985 2825 1451 22. Lu Hanlin et al. Chinese J. of Nuclear Physics, v.7, n.3, 2825 1451 p.242, August 1985; 2825 1451 Lu Hanlin et al. Report INDC(CPR)-16, September 1989 2825 1451 23. J.Rondio et al. Journ. Phys. G: Nucl. Part. Phys., v.11, 2825 1451 pp.549-564, March 1985 2825 1451 24. N.V.Kornilov et al. Atomnaya Energiya, v.58, n.2, p.128, 1985 2825 1451 25. L.R.Greenwood Progress Report ASTM-STP-956, p.743, 1987 2825 1451 26. J.Rondio et al. Acta Physica Polonica Section B, v.20, 2825 1451 No.7, p.627, 1987 ; 2825 1451 J.Rondio et al. Acta Physica Polonica Section B, v.18, 2825 1451 p.1065, November 1987 2825 1451 27. Y.Ikeda et al. Report JAERI-1312, March 1988 2825 1451 28. H.Vonach, M.Wagner, R.C.Haight Report NEANDC-259, p.165, 2825 1451 September 1989 2825 1451 29. S.Cabral, G.Boerker, H.Klein, W.Mannhart 2825 1451 Nucl. Sci. Eng., v.106, p.308, 1990 2825 1451 30. Y.Ikeda et al. Prog. Report NEANDC(J)-155, p.11, August, 1990 2825 1451 31. D.L.Smith, J.W.Meadows, H.Vonach, M.Wagner, R.C.Haight, 2825 1451 W.Mannhart Proc. of Int. Conf. on Nuclear Data for Science 2825 1451 and Technology, Juelich, Germany, 13 - 17 May 1991, p.282 2825 1451 32. M.Viennot et al. Nucl. Sci. Eng., v.108, p.289, July 1991 2825 1451 33. P.M.Dighe et al. Journ. Phys. G: Nucl. Part. Phys., v.17, 2825 1451 p.169-172, 1991 2825 1451 34. Yuan Junqian et al. High Energy Physics and Nuclear Physics, 2825 1451 v.16(1), p.57, January 1992 2825 1451 35. Li Tingyan et.al. Chinese J. High Energy Physics and Nuclear 2825 1451 Physics, v.16, n.2, p.151, Feb.1992 2825 1451 36. Cs.M.Buczko, J.Csikai et al. Physical Review C, v.52, no.4, 2825 1451 p.1940, 1995 2825 1451 37. K.T.Osman, F.I.Habbani Report INDC(SUD)-001,Distr.L, IAEA, 2825 1451 Vienna, October 1996 2825 1451 38. J.Cezar Suita et al. Nucl. Sci. Eng., v.126, p.101, 1997 2825 1451 39. Lu Hanlin et al. Report INDC(CPR)-045, IAEA, October 1998 2825 1451 40. A.A.Filatenkov et al. Report RI-252, May 1999 2825 1451 41. T.Senga et al. Report JAERI-Conf 2000-005, p.208, March 2000 2825 1451 42. D.L.Allan Proc. Phys. Soc., sec.A, v.70, p.195, March 1957 2825 1451 43. K.H.Purser, F.W.Titterton Australian J. Physic, v.12, 2825 1451 p.103, 1959 2825 1451 44. L.Gonzalez et al. Phys. Rev., v.120, p.1319, 1960 2825 1451 45. K.Nakai et al. J. Phys. Soc. Japan, v.17, p.1215, July 1962 2825 1451 46. J.M.F.Jeronimo et al. Nucl. Phys., v.47, p.157, July 1963 2825 1451 47. J.Konijn, A.Lauber Nucl. Phys., v.48, p.191, November 1963 2825 1451 48. S.Okumura Nucl. Phys., v.A93, p.74, March 1967 2825 1451 49. M.Bormann et al. Proc. of 1-st IAEA Conf. on Nuclear Data for 2825 1451 Reactors, Paris, 17-21 October 1966, v.1, p.225, April 1967 2825 1451 50. J.K.Temperley Nucl. Sci. Eng., v.32, p.195, 1968 2825 1451 51. V.N.Levkovskij et al. Yadernaja Fizika (Sov.), v.10, no.1, 2825 1451 p.44, July 1969 2825 1451 52. G.N.Maslov et al. Yadernye Konstanty, v.9, p.50, 1972 2825 1451 53. J.D.Hemingway J. Nucl. Energy, v.27, p.241, April 1973 2825 1451 54. J.Dresler et al. Progress Report, INR-1464, p.12, May 1973 2825 1451 55. K.Fukuda et al. Progress Report NEANDC(J)-56/U, p.44, 2825 1451 September 1978 2825 1451 56. Ju.A.Nemilov, Ju.N.Trofimov Progress Report YFI-26,p.25, 2825 1451 November 1978 2825 1451 57. I.Garlea et al. Rev. Roum. Phys., v.30, no.8, p.673, 1985 2825 1451 58. I.Ribansky et al. Report INDC(CSR)-7, June 1985 2825 1451 59. N.I.Molla et al. Report INDC(BAN)-003, September 1986 ; 2825 1451 M.Rahman et al. Nuclear Science and Applications, v.1, no.1, 2825 1451 p.32, January 1989 2825 1451 60. K.Kobayashi, I.Kimura Proc. of an Intern. Conf. on Nuclear 2825 1451 Data for Science and Technology, 30 May - 3 June 1988, Mito, 2825 1451 Japan, Saikon Publishing Co., LTD, pp.261-265, 1989 2825 1451 61. N.I.Molla et al. Proc. of Int. Conf. on Nuclear Data for 2825 1451 Science and Technology, Juelich, FRG, 13-17 May 1991, p. 355 2825 1451 62. A.Ercan et.al. Proc. of an Intern. Conf. on Nuclear Data for 2825 1451 Science and Technology, 13-17 May 1991, Julich, FRG, Springer-2825 1451 Verlag, 1992 2825 1451 63. I.Garlea et al. Rev. Roum. Phys., v.37, no.1, pp.19-25, 1992 2825 1451 64. L.I.Klochkova et al. Vop. At. Nauki i Tekhn., Ser. Yadernye 2825 1451 Konstanty, v.1, p.27, 1992 2825 1451 65. S.A.Badikov et al. Preprint FEI-1686, Obninsk, 1985 2825 1451 66. S.Badikov, N.Rabotnov, K.Zolotarev Proc. of NEANSC Speciali- 2825 1451 st's Meeting on Evaluation and Processing of Covariance Data, 2825 1451 Oak Ridge, USA, 1992, OECD, Paris, 1993, p.105 2825 1451 67. L.W.Weston et al. Evaluated Neutron Data for Uranium-235, 2825 1451 ENDF/B-VI Library, MAT=9228, MF=5, MT=18, eval. April 1989 2825 1451 68. W.Mannhart IAEA-TECDOC-410, p.158, IAEA, Vienna, 1987 2825 1451 69. W.Mannhart Proc. 5-th ASTM-EUR. Symp. on Reactor Dosimetry, 2825 1451 Geesthacht, FRG, September 24-28, 1984, Vol.2, p.813, 1985 2825 1451 70. O.Horibe et al. Proc. of Conf.:50 Years with Nuclear Fission, 2825 1451 Washington D.C., 25-28 April 1989, v.2, p.923 2825 1451 71. W.Mannhart Prog. Report INDC(Ger)-045, pp.40-43, June 1999 2825 1451 72. W.Mannhart Handbook on Nuclear Activation Cross Sections , 2825 1451 IAEA Technical Report Ser. No.273, p.413, 1987 2825 1451 73. W.Mannhart Validation of Differential Cross Sections with 2825 1451 Integral Data , Report INDC(NDS)-435, pp.59-64, IAEA, Vienna, 2825 1451 September 2002 2825 1451 ***************************************************************** 2825 1451 ******* End of RRDF (N,P) bibliographical component ******* 2825 1451 ***************************************************************** 2825 1451 The Q values and threshold energies were updated prior to pro- 2825 1451 cessing through the codes to comply with the values obtained 2825 1451 using the NNDC calculation program which is based on the 1995 2825 1451 Update to the Atomic mass Evaluation. 2825 1451 2825 1451 Psuedo Threshold of 1.0E+5 changed to 4.0E+5 ev in MF/MT=3/103 2825 1451 in original input file before processing through codes. 2825 1451 2825 1451 File 2 added to the pointwise file containing only the effective 2825 1451 scattering radius with no resonance parameters given. 2825 1451 Taken from ENDF/B-VI 2825 1451 ***************************************************************** 2825 1451 ***************** Program LINEAR (VERSION 2002-1) ***************2825 1451 For All Data Greater than 1.0000E-10 barns in Absolute Value 2825 1451 Data Linearized to Within an Accuracy of .100000000 per-cent 2825 1451 ***************** Program RECENT (VERSION 2002-1) ***************2825 1451 for All Data Greater than 1.0000E-10 barns in Absolute Value 2825 1451 Data Linearized to within an Accuracy of .100000000 per-cent 2825 1451 ***************** Program SIGMA1 (VERSION 2002-1) ***************2825 1451 Data Doppler Broadened to 300.000000 Kelvin 2825 1451 for All Data Greater than 1.0000E-10 barns in Absolute Value 2825 1451 Data Linearized to Within an Accuracy pf .100000000 per-cent 2825 1451 ***************** Program FIXUP (Version 2002-1) ****************2825 1451 Corrected ZA/AWR in All Sections-----------------------------Yes 2825 1451 Corrected Thresholds-----------------------------------------Yes 2825 1451 Extended Cross Sections to 20 MeV----------------------------No 2825 1451 Allow Cross Section Deletion---------------------------------No 2825 1451 Allow Cross Section Reconstruction---------------------------No 2825 1451 Make All Cross Sections Non-Negative-------------------------Yes 2825 1451 Delete Energies Not in Ascending Order-----------------------Yes 2825 1451 Deleted Duplicate Points-------------------------------------Yes 2825 1451 Check for Ascending MAT/MF/MT Order--------------------------Yes 2825 1451 Check for Legal MF/MT Numbers--------------------------------Yes 2825 1451 Allow Creation of Missing Sections---------------------------No 2825 1451 Allow Insertion of Energy Points-----------------------------No 2825 1451 Create Uniform Energy Grid-----------------------------------No 2825 1451 Delete Section if Cross Section =0 at All Energies-----------Yes 2825 1451 ***************** Program GROUPIE (VERSION 2002-1) **************2825 1451 Unshielded Group Averages Using 640 Groups 2825 1451 Weighting Spectrum: Flat (Constant) Spectrum 2825 1451 1 451 387 02825 1451 3 16 29 02825 1451 3 103 73 02825 1451 33 16 58 02825 1451 33 103 133 02825 1451 2825 1 0 2825 0 0 2.80580E+4 5.74376E+1 0 0 0 02825 3 16 -1.22190E+7-1.22190E+7 0 0 1 772825 3 16 77 1 2825 3 16 12400000.0 .000186234 12500000.0 .000945261 12600000.0 .0017445372825 3 16 12700000.0 .002543812 12800000.0 .003343087 12900000.0 .0041423622825 3 16 13000000.0 .005363417 13100000.0 .007006250 13200000.0 .0086490832825 3 16 13300000.0 .010291917 13400000.0 .011934750 13500000.0 .0135775832825 3 16 13600000.0 .015483500 13700000.0 .017652500 13800000.0 .0198330002825 3 16 13900000.0 .022025000 14000000.0 .024031250 14100000.0 .0258517502825 3 16 14200000.0 .028017000 14300000.0 .030527000 14400000.0 .0327212502825 3 16 14500000.0 .034599750 14600000.0 .036406500 14700000.0 .0381415002825 3 16 14800000.0 .039815500 14900000.0 .041428500 15000000.0 .0430853002825 3 16 15100000.0 .044785900 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1.863350-4 1.799310-4 1.755750-4 1.744380-4 1.761340-4 1.782190-4282533103 1.762210-4 1.647870-4 1.402290-4 1.035290-4 6.169210-5 2.584340-5282533103 6.998090-6 1.227500-5 4.351170-5 9.813280-5 1.709930-4 2.561150-4282533103 3.478690-4 4.415720-4 5.336810-4 6.217340-4 7.406080-4 4.698480-4282533103 4.261510-4 3.105910-4 2.084990-4 1.530770-4 1.414180-4 1.567140-4282533103 1.808120-4 2.001940-4 2.080530-4 2.037870-4 1.910920-4 1.756680-4282533103 1.631630-4 1.575100-4 1.596250-4 1.665770-4 1.717430-4 1.667020-4282533103 1.451310-4 1.072230-4 6.151530-5 2.203050-5 2.289250-6 1.025870-5282533103 4.684130-5 1.074370-4 1.846640-4 2.707070-4 3.587670-4 4.436710-4282533103 5.219210-4 5.914870-4 6.755440-4 5.170950-4 4.878640-4 3.941380-4282533103 2.906260-4 2.081330-4 1.574460-4 1.367100-4 1.374010-4 1.489810-4282533103 1.621920-4 1.708060-4 1.720450-4 1.661260-4 1.553290-4 1.428380-4282533103 1.314040-4 1.220480-4 1.132700-4 1.015920-4 8.378250-5 5.972660-5282533103 3.385720-5 1.387210-5 7.716430-6 2.090880-5 5.544840-5 1.101620-4282533103 1.817660-4 2.659880-4 3.584410-4 4.551630-4 5.528810-4 6.935630-4282533103 5.768110-4 5.531400-4 4.556030-4 3.323940-4 2.207550-4 1.417480-4282533103 1.011220-4 9.323110-5 1.062830-4 1.272180-4 1.451270-4 1.530470-4282533103 1.484980-4 1.331520-4 1.117920-4 9.048050-5 7.406550-5 6.387610-5282533103 5.713620-5 4.909770-5 3.676990-5 2.149920-5 8.674810-6 5.318290-6282533103 1.751250-5 4.888850-5 1.003540-4 1.706000-4 2.568760-4 3.557620-4282533103 4.637120-4 6.333340-4 6.069070-4 5.605150-4 4.534650-4 3.278320-4282533103 2.158140-4 1.354730-4 9.141260-5 7.842730-5 8.615350-5 1.031270-4282533103 1.195210-4 1.286180-4 1.273560-4 1.162180-4 9.845830-5 7.863670-5282533103 6.065940-5 4.622520-5 3.474380-5 2.489180-5 1.654910-5 1.161150-5282533103 1.335700-5 2.517250-5 4.948540-5 8.730490-5 1.382790-4 2.010270-4282533103 2.735760-4 3.536950-4 4.817800-4 5.812910-4 5.312480-4 4.377140-4282533103 3.288570-4 2.274940-4 1.481040-4 9.672000-5 7.264260-5 7.081360-5282533103 8.390600-5 1.038060-4 1.225950-4 1.333900-4 1.313770-4 1.150770-4282533103 8.728570-5 5.472140-5 2.582750-5 7.650840-6 3.665640-6 1.362650-5282533103 3.491240-5 6.412560-5 9.816150-5 1.346460-4 1.719590-4 2.090840-4282533103 2.454320-4 2.806830-4 3.303550-4 5.509690-4 5.154950-4 4.391610-4282533103 3.415510-4 2.425000-4 1.581420-4 9.888940-5 6.904830-5 6.730480-5282533103 8.747470-5 1.193440-4 1.499790-4 1.661890-4 1.585030-4 1.256480-4282533103 7.668980-5 2.811230-5-3.501070-6-8.510460-6 1.331180-5 5.508020-5282533103 1.069860-4 1.597770-4 2.065590-4 2.431800-4 2.678620-4 2.805190-4282533103 2.821410-4 2.667540-4 5.417870-4 5.136820-4 4.421840-4 3.463000-4282533103 2.471180-4 1.628990-4 1.061840-4 8.253620-5 9.022130-5 1.204470-4282533103 1.584500-4 1.863210-4 1.883670-4 1.580330-4 1.027010-4 4.182840-5282533103 -1.978330-6-1.327090-5 1.116700-5 6.402690-5 1.326760-4 2.041440-4282533103 2.680810-4 3.177620-4 3.498570-4 3.636240-4 3.600870-4 3.260650-4282533103 5.356030-4 5.047340-4 4.339530-4 3.427590-4 2.515430-4 1.772730-4282533103 1.309670-4 1.163850-4 1.294420-4 1.584730-4 1.862670-4 1.949470-4282533103 1.733930-4 1.238480-4 6.254270-5 1.232750-5-8.083160-6 8.890820-6282533103 5.948470-5 1.328500-4 2.160810-4 2.977160-4 3.693570-4 4.259120-4282533103 4.650730-4 4.866190-4 4.870360-4 5.190510-4 4.882560-4 4.249420-4282533103 3.458300-4 2.676380-4 2.040560-4 1.637790-4 1.491530-4 1.553460-4282533103 1.707500-4 1.798490-4 1.691530-4 1.342560-4 8.344380-5 3.432480-5282533103 5.388940-6 8.354180-6 4.544650-5 1.112280-4 1.962950-4 2.905760-4282533103 3.853630-4 4.741820-4 5.528270-4 6.190620-4 6.913430-4 5.024060-4282533103 4.788270-4 4.267050-4 3.585150-4 2.874830-4 2.254110-4 1.808090-4282533103 1.570670-4 1.510140-4 1.530260-4 1.501600-4 1.322530-4 9.798390-5282533103 5.640470-5 2.258470-5 1.070060-5 2.911650-5 7.923250-5 1.570580-4282533103 2.556770-4 3.673960-4 4.851380-4 6.031250-4 7.170790-4 8.712440-4282533103 4.970580-4 4.795620-4 4.324380-4 3.657970-4 2.920140-4 2.237100-4282533103 1.713170-4 1.402170-4 1.283050-4 1.259800-4 1.202720-4 1.019710-4282533103 7.128330-5 3.787990-5 1.585420-5 1.766380-5 5.050340-5 1.157090-4282533103 2.100860-4 3.278250-4 4.621730-4 6.066060-4 7.554590-4 9.737830-4282533103 4.965710-4 4.768580-4 4.259810-4 3.542670-4 2.752360-4 2.031230-4282533103 1.494550-4 1.191020-4 1.078110-4 1.038920-4 9.468250-5 7.415140-5282533103 4.629310-5 2.249660-5 1.582900-5 3.625710-5 8.855090-5 1.725350-4282533103 2.844970-4 4.187960-4 5.692020-4 7.297510-4 9.749350-4 4.857170-4282533103 4.593680-4 4.041610-4 3.309560-4 2.534110-4 1.850300-4 1.352320-4282533103 1.058290-4 9.053390-5 7.906960-5 6.388380-5 4.475090-5 2.842960-5282533103 2.474100-5 4.238370-5 8.649170-5 1.581960-4 2.554100-4 3.740190-4282533103 5.090310-4 6.553620-4 8.830400-4 4.622530-4 4.355670-4 3.846720-4282533103 3.182150-4 2.468290-4 1.807690-4 1.271380-4 8.806000-5 6.133630-5282533103 4.348530-5 3.294650-5 3.123000-5 4.181120-5 6.821470-5 1.125520-4282533103 1.750150-4 2.540930-4 3.471210-4 4.509020-4 5.621480-4 7.341490-4282533103 4.443900-4 4.285000-4 3.883860-4 3.274380-4 2.526150-4 1.740560-4282533103 1.032440-4 5.000760-5 1.996620-5 1.392060-5 2.917560-5 6.150500-5282533103 1.066340-4 1.608980-4 2.213640-4 2.857300-4 3.521810-4 4.192940-4282533103 4.859520-4 5.816950-4 4.533230-4 4.499330-4 4.120140-4 3.396280-4282533103 2.425640-4 1.399160-4 5.435630-5 3.576710-6-5.659410-6 2.242670-5282533103 7.696180-5 1.454850-4 2.173370-4 2.849970-4 3.440510-4 3.925320-4282533103 4.301430-4 4.575930-4 4.813510-4 4.857840-4 4.796520-4 4.229260-4282533103 3.212120-4 1.967440-4 8.186660-5 5.950200-6-1.591330-5 1.429150-5282533103 8.259810-5 1.707280-4 2.620060-4 3.441990-4 4.099680-4 4.560900-4282533103 4.822470-4 4.899470-4 4.712900-4 5.075800-4 4.787570-4 3.916490-4282533103 2.653020-4 1.347580-4 3.667320-5-5.566650-6 1.196800-5 7.748840-5282533103 1.717390-4 2.753360-4 3.729060-4 4.543540-4 5.144220-4 5.515150-4282533103 5.665130-4 5.513000-4 4.853190-4 4.324480-4 3.299840-4 2.066050-4282533103 9.851580-5 3.426120-5 2.566870-5 6.833550-5 1.479640-4 2.473420-4282533103 3.510700-4 4.476680-4 5.298930-4 5.940640-4 6.391730-4 6.707940-4282533103 4.268960-4 3.717220-4 2.815670-4 1.833800-4 1.050470-4 6.517590-5282533103 6.955450-5 1.135560-4 1.869020-4 2.778930-4 3.760410-4 4.732300-4282533103 5.638770-4 6.446410-4 7.413690-4 3.756610-4 3.387300-4 2.724920-4282533103 1.974450-4 1.345130-4 9.832300-5 9.510360-5 1.240620-4 1.800840-4282533103 2.562310-4 3.454750-4 4.416530-4 5.398550-4 6.807700-4 3.609990-4282533103 3.428820-4 2.928950-4 2.278710-4 1.658530-4 1.206730-4 1.000320-4282533103 1.060900-4 1.371230-4 1.892260-4 2.576610-4 3.377130-4 4.689640-4282533103 3.773120-4 3.706070-4 3.302750-4 2.703650-4 2.058530-4 1.486660-4282533103 1.063320-4 8.229690-5 7.695740-5 8.879390-5 1.152710-4 1.761240-4282533103 4.157570-4 4.230810-4 3.969360-4 3.463240-4 2.815830-4 2.118020-4282533103 1.437810-4 8.194610-5 2.873500-5-1.490280-5-6.107470-5 4.907200-4282533103 5.238500-4 5.211570-4 4.859640-4 4.245560-4 3.440740-4 2.512660-4282533103 1.518550-4 5.033710-5-9.636020-5 6.307180-4 7.025830-4 7.333670-4282533103 7.226360-4 6.745020-4 5.953670-4 4.923350-4 3.722470-4 1.753660-4282533103 8.655530-4 9.914960-4 1.072120-3 1.105210-3 1.093810-3 1.043780-3282533103 9.621740-4 7.967120-4 1.233200-3 1.438380-3 1.597930-3 1.708340-3282533103 1.771340-3 1.791690-3 1.753320-3 1.793410-3 2.115770-3 2.395310-3282533103 2.626540-3 2.809320-3 2.992600-3 2.629110-3 3.114210-3 3.559090-3282533103 3.955790-3 4.443820-3 3.832780-3 4.524150-3 5.174450-3 6.033490-3282533103 5.486660-3 6.417000-3 7.695010-3 7.646890-3 9.373270-3 1.179240-2282533103 282533 0 2825 0 0 0 0 0 2.80600E+4 5.94159E+1 0 0 34 102831 1451 0.0 0.0 0 0 0 62831 1451 1.00000E+0 1.50000E+8 0 0 10 20022831 1451 3.00000E+2 0.0 1 0 395 32831 1451 28-Ni- 60 LANL,ORNL EVAL-SEP97 S.CHIBA,M.B.CHADWICK,LARSON 2831 1451 DIST-Feb2004 2831 1451 ----IRDF-2002 MATERIAL 2831 2831 1451 -----INCIDENT NEUTRON DATA 2831 1451 ------ENDF-6 FORMAT 2831 1451 2831 1451 ******************************************************************2831 1451 ----ENDF/B-VI MATERIAL 2831 REVISION 3 2831 1451 Ch97,Ch99 DIST-SEP 1 REV3- 20010926 2831 1451 2831 1451 ENDF/B-VI MOD 4 Revision, June 2000, S.C. Frankle, R.C. Reedy, 2831 1451 P.G. Young (LANL) 2831 1451 2831 1451 The secondary gamma-ray spectrum for radiative capture (MF 12, 2831 1451 MT 102) has been updated for new experimental data at incident 2831 1451 neutron energies up to 1 keV. 2831 1451 The Q-value for radiative capture was also updated in File 3. 2831 1451 Details of these changes are described in Frankel et al. [Fr01]. 2831 1451 2831 1451 ******************************************************************2831 1451 *****************EXTRACT FOR SPECIAL PURPOSE FILE*****************2831 1451 DOSIMETRY 2831 1451 ******************************************************************2831 1451 2831 1451 ENDF/B-VI MOD 3 Evaluation, September 1997, S. Chiba, 2831 1451 M.B. Chadwick, P.G. Young (LANL), and 2831 1451 A.J. Koning (ECN) 2831 1451 2831 1451 Los Alamos LA150 Library, produced with FKK/GNASH/GSCAN code 2831 1451 in cooperation with ECN Petten. 2831 1451 2831 1451 This evaluation provides a complete representation of the 2831 1451 nuclear data needed for transport, damage, heating, 2831 1451 radioactivity, and shielding applications over the incident 2831 1451 neutron energy range from 1.0E-11 to 150 MeV. The discussion 2831 1451 here is divided into the region below and above 20 MeV. 2831 1451 2831 1451 INCIDENT NEUTRON ENERGIES < 20 MeV 2831 1451 2831 1451 Below 20 MeV the evaluation is based completely on the ENDF/B- 2831 1451 VI.5(Rev1,Release 2) evaluation by Larson, C. Perey, Hetrich, and 2831 1451 Fu. 2831 1451 2831 1451 INCIDENT NEUTRON ENERGIES > 20 MeV 2831 1451 2831 1451 The ENDF/B-VI Release 2 evaluation extends to 20 MeV and 2831 1451 includes cross sections and energy-angle data for all 2831 1451 significant reactions. The present evaluation utilizes a more 2831 1451 compact composite reaction spectrum representation above 20 MeV 2831 1451 in order to reduce the length of the file. No essential data for 2831 1451 applications is lost with this representation. 2831 1451 The evaluation above 20 MeV utilizes MF=6, MT=5 to represent 2831 1451 all reaction data. Production cross sections and emission 2831 1451 spectra are given for neutrons, protons, deuterons, tritons, 2831 1451 alpha particles, gamma rays, and all residual nuclides produced 2831 1451 (A>5) in the reaction chains. To summarize, the ENDF sections 2831 1451 with non-zero data above En = 20 MeV are: 2831 1451 2831 1451 MF=3 MT= 1 Total Cross Section 2831 1451 MT= 2 Elastic Scattering Cross Section 2831 1451 MT= 3 Nonelastic Cross Section 2831 1451 MT= 5 Sum of Binary (n,n') and (n,x) Reactions 2831 1451 2831 1451 MF=4 MT= 2 Elastic Angular Distributions 2831 1451 2831 1451 MF=6 MT= 5 Production Cross Sections and Energy-Angle 2831 1451 Distributions for Emission Neutrons, Protons, 2831 1451 Deuterons, Tritons, and Alphas; and Angle- 2831 1451 Integrated Spectra for Gamma Rays and Residual 2831 1451 Nuclei That Are Stable Against Particle Emission 2831 1451 2831 1451 The evaluation is based on nuclear model calculations that 2831 1451 have been benchmarked to experimental data, especially for n + 2831 1451 Ni58 and p + Ni58 reactions [Ch97]. We use the GNASH code system 2831 1451 [Yo92], which utilizes Hauser-Feshbach statistical, 2831 1451 preequilibrium and direct-reaction theories. Spherical optical 2831 1451 model calculations are used to obtain particle transmission 2831 1451 coefficients for the Hauser-Feshbach calculations, as well as 2831 1451 for the elastic neutron angular distributions. 2831 1451 Cross sections and spectra for producing individual residual 2831 1451 nuclei are included for reactions. The energy-angle-correlations 2831 1451 for all outgoing particles are based on Kalbach systematics 2831 1451 [Ka88]. 2831 1451 A model was developed to calculate the energy distributions of 2831 1451 all recoil nuclei in the GNASH calculations [Ch96]. The recoil 2831 1451 energy distributions are represented in the laboratory system in 2831 1451 MT=5, MF=6, and are given as isotropic in the lab system. All 2831 1451 other data in MT=5,MF=6 are given in the center-of-mass system. 2831 1451 This method of representation utilizes the LCT=3 option approved 2831 1451 at the November, 1996, CSEWG meeting. 2831 1451 Preequilibrium corrections were performed in the course of the 2831 1451 GNASH calculations using the exciton model of Kalbach [Ka77, 2831 1451 Ka85], validated by comparison with calculations using Feshbach, 2831 1451 Kerman, Koonin (FKK) theory [Ch93]. Discrete level data from 2831 1451 nuclear data sheets were matched to continuum level densities 2831 1451 using the formulation of Gilbert and Cameron [Gi65] and pairing 2831 1451 and shell parameters from the Cook [Co67] analysis. Neutron and 2831 1451 charged- particle transmission coefficients were obtained from 2831 1451 the optical potentials, as discussed below. Gamma-ray 2831 1451 transmission coefficients were calculated using the Kopecky-Uhl 2831 1451 model [Ko90]. 2831 1451 2831 1451 SOME Ni-SPECIFIC INFORMATION CONCERNING THE EVAL. 2831 1451 2831 1451 The neutron total cross section was evaluated based on the 2831 1451 least-squares method taking account of the experimental 2831 1451 data [Du67, Bo71, St71, Sm79, Fe80, Pe82, Ha82a, Di97]. The data 2831 1451 for natural Ni [Di97] was also used because there was not enough 2831 1451 data for Ni-60 above 20 MeV. The data for natural Ni were 2831 1451 transformed to the Ni-60 cross section according to A*(2/3) law. 2831 1451 Result of this estimation was used as the evaluated total cross 2831 1451 section data above 20 MeV. 2831 1451 2831 1451 The evaluated total cross section data (1 to 250 MeV), s-wave 2831 1451 strength function [Mu81] and elastic scattering angular 2831 1451 distribution data [Bo71, Gu85, Tu73, Ya79] were used to obtain the2831 1451 neutron optical potential parameters. The parameter estimation was2831 1451 carried out based on Marquart-Bayesian approach [Sm91], where 2831 1451 ECIS95 code [Ra96] was used for the optical model calculation. We2831 1451 have employed the energy dependence of the optical potential 2831 1451 similar to Delaroche's work [De89]. The initial potential 2831 1451 parameters were adopted from Koning and Delaroche [Ko97]. A total2831 1451 of 7 parameters concerning the central potential depth were 2831 1451 estimated with associated covariance matrix, while the geometrical2831 1451 parameters were fixed to the result of a similar search for n + 2831 1451 Ni-58. Presently obtained potential was used for the calculation 2831 1451 of neutron transmission coefficients and DWBA cross sections in 2831 1451 the energy region above 20 MeV. Below 20 MeV, the Harper neutron 2831 1451 potential [Ha82b] was used for the calculation of transmission 2831 1451 coefficients. 2831 1451 2831 1451 The proton optical potential was also searched to obtain a good 2831 1451 description of proton-total reaction cross section as predicted by2831 1451 Wellisch-Axen systematic [We96] above 50 MeV. The parameter 2831 1451 estimation was carried out by the Marquart-Bayesian approach 2831 1451 similar to the neutron OMP, but trying to seek the best parameter 2831 1451 to reproduce the reaction cross sections compiled by Carlson 2831 1451 [Ca96] and Wellisch values. The experimental data in Carlson 2831 1451 [Ca96] was scaled for Ni-60 according to A**(2/3) law. In this 2831 1451 search, the geometrical parameters were fixed to be same as the 2831 1451 neutron potential. The present potential gives a good description2831 1451 of the proton total reaction cross section from 10 MeV to 250 MeV.2831 1451 However, after some trial and error to reproduce both the elastic 2831 1451 scattering and reaction cross section data for Ni-58, we have 2831 1451 employed the following combination of proton potentials: 2831 1451 2831 1451 0 to 5 MeV : Harper potential [Ha82b] 2831 1451 6 to 47 MeV : Koning and Delaroche [Ko97] 2831 1451 48 to 260 MeV : Present OMP 2831 1451 2831 1451 For deuterons, the Lohr-Haeberli [Lo74] global potential was used;2831 1451 for alpha particles the McFadden-Satchler [Mc66] potential was 2831 1451 used; and for tritons the Becchetti-Greenlees [Be71] potential was2831 1451 used. The He-3 channel was ignored. 2831 1451 2831 1451 The direct collective inelastic scattering to the following level 2831 1451 in Ni-60 was considered by the DWBA-mode calculation of ECIS95 2831 1451 [Ra96]: 2831 1451 2831 1451 Jpi Ex(MeV) Deformation length 2831 1451 2+ 1.331 0.8535 2831 1451 2831 1451 The deformation length was determined to much the ENDF/B-VI value 2831 1451 at 20 MeV. 2831 1451 2831 1451 Certain nuclear level densities were modified from the default 2831 1451 values in accordance with experimental level-density information 2831 1451 as follows. The level density of 57Fe was matched to the observed 2831 1451 D-0 value by modifying the pairing energy to 0.15 MeV; the level 2831 1451 density of the main competition channel, neutron emission to 60Ni,2831 1451 was adjusted to match Fischer et al.'s [Fi86] value of 3.85e3+-25%2831 1451 at 11 MeV excitation energy, but then increased slightly (a total 2831 1451 density increase of 11%) by using a pairing energy of 1.1 MeV, 2831 1451 since Haight's alpha production at 14 MeV is approx. 11% smaller 2831 1451 than Fischer et al.'s measurements. Production of residual nuclei 2831 1451 59Ni and 56Fe, through (n,2n) and (n,na) reactions, become 2831 1451 particularly important above 14 MeV. Pairing energies were 2831 1451 adjusted to match the level density for 59Ni at the neutron 2831 1451 binding energy (D0=12.9 keV), and Fischer et al.'s result for the 2831 1451 56Fe level density at 11 MeV. 2831 1451 2831 1451 The new Haight et al. [Ha97] LANL/WNR 60Ni(n,x alpha) data up to 2831 1451 50 MeV was used to benchmark our model calculations [Ch97] - 2831 1451 agreement with experiment was good. 2831 1451 2831 1451 **************************************************************** 2831 1451 2831 1451 REFERENCES 2831 1451 2831 1451 [Be71] F.D. Becchetti, Jr., and G.W. Greenlees in 2831 1451 "Polarization Phenomena in Nuclear Reactions," (Ed: H.H. 2831 1451 Barschall and W. Haeberli, The University of Wisconsin Press, 2831 1451 1971) p.682 2831 1451 [Bo71] P. Boschung et al, Nucl.Phys. A161, 593 (1971) 2831 1451 [Ca96] R.F. Carlson, Atomic Data and Nuclear Data Tables, 63, 2831 1451 93 (1996) 2831 1451 [Ch93] M.B. Chadwick and P.G. Young, Phys.Rev. C 47, 2255 (1993) 2831 1451 [Ch96] M.B. Chadwick, P.G. Young, R.E. MacFarlane, and A.J. 2831 1451 Koning, "High-Energy Nuclear Data Libraries for Accelerator- 2831 1451 Driven Technologies: Calculational Method for Heavy Recoils," 2831 1451 Proc. of 2nd Int. Conf. on Accelerator Driven Transmutation 2831 1451 Technology and Applications, Kalmar, Sweden, 3-7 June 1996 2831 1451 [Ch97] M.B. Chadwick and P.G. Young, "Model Calculations of 2831 1451 n,p + 58,60,61,62,64Ni" in APT PROGRESS REPORT: 1 August - 1 2831 1451 September 1997, internal Los Alamos National Laboratory memo T-2831 1451 2-97/MS-51, 8 September 1997 from R.E. MacFarlane to L. Waters.2831 1451 [Ch99] M.B. Chadwick, P.G. Young, G.M. Hale, et al., Los Alamos 2831 1451 National Laboratory report, LA-UR-99-1222 (1999) 2831 1451 [Ci68] S. Cierjack et al, report KFK-1000 (1968) 2831 1451 [Co67] J.L. Cook, H. Ferguson, and A.R. DeL Musgrove, Aust.J. 2831 1451 Phys. 20, 477 (1967) 2831 1451 [De89] J.P. Delaroche, Y. Wang and J. Rapaport, Phys.Rev.C 39, 2831 1451 391 (1989) 2831 1451 [Di97] F. Dietrich et al., private communication (1997). 2831 1451 [Du67] Yu.V. Dukarevich et al., Nucl.Phys. A92, 433 (1967) 2831 1451 [Fe80] M.B. Fedorov et al., 80Kiev, 1, 309(1980) 2831 1451 [Fi86] R. Fischer, G. Traxler, M. Uhl at al., Phys Rev C34, 460 2831 1451 (1986) 2831 1451 [Fr01] S.C. Frankle, R.C. Reedy, and P.G. Young, Los ALamos 2831 1451 National Laboratory Report, LA-13812 (2001). 2831 1451 [Fu96] C.Y. Fu, "Effects of shape differences in the level 2831 1451 densities of three formalisms on calculated cross sections", 2831 1451 personal communication, 1996 (to be published by the IAEA). 2831 1451 [Gi65] A. Gilbert and A.G.W. Cameron, Can.J.Phys. 43, 1446 (1965)2831 1451 [Gu85] P.P. Guss et al, Nucl.Phys. A438, 187 (1985) 2831 1451 [Ha82a] J.A. Harvey et al, 82Anterp (1982) p. 856 2831 1451 [Ha82b] R.C. Harper and W.L. Alford, J.Phys.G. 8, 153 (1982) 2831 1451 [Ha97] R.C. Haight, F.B. Bateman, S.M. Sterbenz et al., Nuclear 2831 1451 Data for Science and Technol., Proc. Conf. Trieste, Italy, May 2831 1451 1997, Vol. I (Edit. Compositori, Bologna, Italy, 1997) p. 603 2831 1451 [Ka77] C. Kalbach, Z.Phys.A 283, 401 (1977) 2831 1451 [Ka85] C. Kalbach, Los Alamos National Laboratory report 2831 1451 LA-10248-MS (1985) 2831 1451 [Ka88] C. Kalbach, Phys.Rev.C 37, 2350 (1988); see also 2831 1451 C. Kalbach and F. M. Mann, Phys.Rev.C 23, 112 (1981) 2831 1451 [Ko90] J. Kopecky and M. Uhl, Phys.Rev.C 41, 1941 (1990) 2831 1451 [Ko97] A. Koning and J.P. Delaroche, private communication. 2831 1451 [La83] D.C. Larson et al, report ORNL-TM-8203 (1983) 2831 1451 [Lo74] J.M.Lohr and W.Haeberli, Nucl.Phys. A232, 381 (1974) 2831 1451 [Mc66] L. McFadden and G.R. Satchler, Nucl.Phys. 84, 177 (1966) 2831 1451 [Mu81] S.F. Mughabghab, M. Divadeenam and N.E. Holden, "Neutron 2831 1451 Cross Sections", Vol. 1, Part A (Academic Press, 1981) 2831 1451 [Pe73] F.G. Perey, private communication (1973) [EXFOR 10342] 2831 1451 [Pe82] C.M. Perey et al, report ORNL-5893 (1982) 2831 1451 [Pe88] Pedroni et al, Phys.Rev.C 38, 2052 (1988) 2831 1451 [Po81] W. Poenitz, Proc. Conf. on Nuclear Data Evaluation Methods2831 1451 and Procedures, Brookhaven National Laboratory Report BNL-NCS- 2831 1451 51363, p.249(1981). 2831 1451 [Ra96] J. Raynal, "Notes on ECIS94", Service de Physique 2831 1451 Theorique, Saclay, France (personal communication through A. J.2831 1451 Koning, 1996). 2831 1451 [Sc73] W. Schimmerling et al., Phys.Rev.C 7, 248 (1973) 2831 1451 [Sm79] A.B. Smith et al., Nucl.Sci.Eng., 72, 293 (1979) 2831 1451 [Sm91] D.L. Smith, "Probability, Statistics, and Data Uncertainty2831 1451 in Nuclear Science and Technology" (American Nuclear Society, 2831 1451 1991) 2831 1451 [Sm92] A.B. Smith et al., J.Phys.G, 18, 629 (1992) 2831 1451 [St71] P. Stoler et al,, 71Knox, 1, 311 (1971) 2831 1451 [Tu73] A.I. Tutubalin et al., Neutron Physics, Proc. Conf., Kiev,2831 1451 Ukraine, 1973, Vol. 3 [EXFOR 40417] p. 62 2831 1451 [Ul83] J.J. Ullmann, F.P. Brady, C.M. Castaneda, et al., Nucl. 2831 1451 Phys. A427, 493 (1984) 2831 1451 [We96] H.P. Wellisch and D. Axen, Phys.Rev.C 54, 1329 (1996) 2831 1451 [Ya79] Y. Yamanouti, J. Rapaport, S.M. Grimes et al, Nuclear 2831 1451 Cross Sections for Technology, Proc. Conf., Knoxville, TN, 2831 1451 1979, NBS Special Publication 594 (1980) [EXFOR 10953] 2831 1451 [Yo92] P.G. Young, E.D. Arthur, and M.B. Chadwick, Los Alamos 2831 1451 report LA-12343-MS (1992) 2831 1451 2831 1451 **************************************************************** 2831 1451 2831 1451 ENDF/B-VI MOD 2 Revision, July 1991, D.C. Larson, C.M. Perey, 2831 1451 D.M. Hetrick, and C.Y. Fu (ORNL) 2831 1451 2831 1451 REVISION 1 CHANGES 2831 1451 The resonance region from 1.E-5 eV to 450 keV has been changed,2831 1451 based on a SAMMY analysis by F. G. Perey of new 60Ni ORELA data 2831 1451 up to 100 keV of Harvey, Hill and Perey. This new work reproduces2831 1451 the thermal cross sections and includes the small resonances 2831 1451 inadvertently left out in the original file. No background 2831 1451 cross sections are required in file 3 in the resonance region. 2831 1451 Removed elastic transformation matrix. 2831 1451 2831 1451 **************************************************************** 2831 1451 2831 1451 ENDF/B-VI MOD 1 Evaluation, October 1989, D.C. Larson, 2831 1451 C.M. Perey, D.M. Hetrick, and C.Y. Fu (ORNL) 2831 1451 2831 1451 This work employed nuclear model codes including the 2831 1451 Distorted Wave Born Approximation (DWBA) program DWUCK (1) 2831 1451 and the Hauser-Feshbach code TNG (2,3,4). The TNG code provides 2831 1451 energy and angular distributions of particles emitted in the 2831 1451 compound and pre-compound reactions, ensures consistency among 2831 1451 all reactions, and maintains energy balance. Details pertinent 2831 1451 to the contents of this evaluation and extensive comparisons of 2831 1451 calculations with experimental data can be found in Hetrick (5). 2831 1451 2831 1451 ----- DESCRIPTION OF FILES 2831 1451 (MF-MT) 2831 1451 1-451 GENERAL INFORMATION, REFERENCES, AND DEFINITIONS. 2831 1451 2-151 RESONANCE PARAMETERS -- TAKEN FROM REF (6). THE NEGATIVE 2831 1451 ENERGY RESONANCES WERE ADJUSTED BY THE AUTHOR, C. M. PEREY2831 1451 TO GIVE THE PROPER THERMAL SCATTERING AND CAPTURE. USING 2831 1451 THESE MODIFIED RESULTS, NO BACKGROUND IS REQUIRED IN 3/1 2831 1451 IE, THE TOTAL, SCATTERING AND CAPTURE IS GIVEN COMPLETELY 2831 1451 BY THE RESONANCE PARAMETERS FROM 1.E-5 EV TO 450 KEV. 2831 1451 RESONANCE PARAMETERS ARE FROM A REICH-MOORE ANALYSIS WITH 2831 1451 THE CODE SAMMY (REF 7). 2831 1451 THUS, THE THERMAL CROSS SECTIONS ARE GIVEN BY THE 2831 1451 RESONANCE PARAMETERS AND HAVE VALUES: TOTAL 3.9 B, 2831 1451 ELASTIC SCATTERING 0.98 B, AND CAPTURE 2.92 B. 2831 1451 NOTE THAT THE FLAG HAS BEEN SET TO ALLOW USER CALCULATION 2831 1451 OF THE ANGULAR DISTRIBUTIONS FROM THE R-M RESONANCE 2831 1451 PARAMETERS, IF THE USER WANTS ANGULAR DISTRIBUTIONS ON 2831 1451 A FINER ENERGY GRID THAN GIVEN IN 4/2. 2831 1451 3-103 (N,P) CROSS SECTIONS FOR EN LESS THAN OR EQUAL TO 12 MEV 2831 1451 WERE TAKEN FROM A CURVE DRAWN THROUGH THE DATA OF VONACH 2831 1451 ET AL. (10). THIS CURVE WAS CONNECTED TO THE ANL DATA OF 2831 1451 GREENWOOD (11) FROM 14.5 TO 15.0 MEV. FROM 15.0 TO 20.0 2831 1451 MEV THE DATA OF PAULSEN AND LISKIEN (12) WERE USED 2831 1451 AS A GUIDE. THE NEW DATA BY VONACH ET AL. (10) 2831 1451 RESULT IN AN INTEGRAL VALUE CLOSE TO THE 2.39 VALUE OF 2831 1451 MANNHART (13). 2831 1451 -------------------------------------------------------------- 2831 1451 UNCERTAINTY FILES 2831 1451 AN LB=8 SECTION IS INCLUDED FOR ALL NON-DERIVED FILES AS 2831 1451 REQUIRED BY ENDF/B-VI. 2831 1451 2831 1451 33-103 COVARIANCES - DATA OF VONACH ET AL. (10) AND CSISRS DATA 2831 1451 USED AS A GUIDE. 2831 1451 ---------------------------------------------------------------- 2831 1451 REFERENCES: 2831 1451 (1) P.D. Kunz, "DISTORTED WAVE CODE DWUCK72," UNIV. OF 2831 1451 COLORADO, UNPUBLISHED (1972). 2831 1451 (2) C.Y. Fu, Oak Ridge report ORNL/TM-7042 (1980) 2831 1451 (3) C.Y Fu, Neutron Cross Sections from 10 to 50 MeV, Proc. Symp.2831 1451 Upton, NY, May 1980, Brookhaven report BNL-NCS-51425 (1980) 2831 1451 p. 675 2831 1451 (4) K. Shibata and C.Y. Fu, Oak Ridge National Laboratory report 2831 1451 ORNL/TM-10093 (1986) 2831 1451 (5) D.M. Hetrick, C.Y. Fu, and D.C. Larson, Oak Ridge National 2831 1451 Laboratory report ORNL/TM-10219 [ENDF-344] (1987) 2831 1451 (6) C.M. Perey, J.A. Harvey, R.L. Macklin, et al., Phys.Rev.C, 2831 1451 27, 2556 (1983) 2831 1451 (7) N.M. Larson and F.G. Perey, Oak Ridge National Laboratory 2831 1451 reports ORNL/TM-7485 (1980); ORNL/TM-9179 (1984); 2831 1451 ORNL/TM-9179/R1 (1985); ORNL/TM-9179/R2 (1988) 2831 1451 (8) J.A. Harvey, ORELA, 80-M NE110 TA TARGET DATA, SENT TO NNDC 2831 1451 (9) J.A. Harvey, ORELA, 80-M NE110 BE BLOCK DATA, SENT TO NNDC 2831 1451 (10) H. Vonach, M. Wagner, and R.C. Haight, "Neutron cross 2831 1451 sections of 58Ni and 60Ni for 8-12 MeV neutrons," private 2831 1451 communication, 1989 2831 1451 (11) L.R. Greenwood, report DOE-ER-0046-21 (1985) p. 15 2831 1451 (12) A. Paulsen, Nukleonik 10, 91 (1967) 2831 1451 (13) W. Mannhart, Proc. Fourth ASTM-Euratom Symp. on Reactor 2831 1451 Dosimetry, Vol. II, March 22-26, 1982 (1982) p. 637 2831 1451 (14) M. Divadeenam, Brookhaven National Laboratory report 2831 1451 BNL-NCS-51346 [ENDF-294] (1979) 2831 1451 (15) S.M. Grimes, R.C. Haight, K.R. Alvar, et al., Phys.Rev.C 2831 1451 19, 2127 (1979) 2831 1451 ****************************************************************2831 1451 Psuedo Threshold of 0.0 added at 2.7E+6 ev to MF/MT=3/103 after 2831 1451 processing thro. codes. 2831 1451 2831 1451 2831 1451 2831 1451 ************************ C O N T E N T S *********************** 2831 1451 ***************** Program LINEAR (VERSION 2002-1) ***************2831 1451 For All Data Greater than 1.0000E-10 barns in Absolute Value 2831 1451 Data Linearized to Within an Accuracy of .100000000 per-cent 2831 1451 ***************** Program RECENT (VERSION 2002-1) ***************2831 1451 for All Data Greater than 1.0000E-10 barns in Absolute Value 2831 1451 Data Linearized to within an Accuracy of .100000000 per-cent 2831 1451 ***************** Program SIGMA1 (VERSION 2002-1) ***************2831 1451 Data Doppler Broadened to 300.000000 Kelvin 2831 1451 for All Data Greater than 1.0000E-10 barns in Absolute Value 2831 1451 Data Linearized to Within an Accuracy pf .100000000 per-cent 2831 1451 ***************** Program FIXUP (Version 2002-1) ****************2831 1451 Corrected ZA/AWR in All Sections-----------------------------Yes 2831 1451 Corrected Thresholds-----------------------------------------Yes 2831 1451 Extended Cross Sections to 20 MeV----------------------------No 2831 1451 Allow Cross Section Deletion---------------------------------No 2831 1451 Allow Cross Section Reconstruction---------------------------No 2831 1451 Make All Cross Sections Non-Negative-------------------------Yes 2831 1451 Delete Energies Not in Ascending Order-----------------------Yes 2831 1451 Deleted Duplicate Points-------------------------------------Yes 2831 1451 Check for Ascending MAT/MF/MT Order--------------------------Yes 2831 1451 Check for Legal MF/MT Numbers--------------------------------Yes 2831 1451 Allow Creation of Missing Sections---------------------------No 2831 1451 Allow Insertion of Energy Points-----------------------------No 2831 1451 Create Uniform Energy Grid-----------------------------------No 2831 1451 Delete Section if Cross Section =0 at All Energies-----------Yes 2831 1451 ***************** Program GROUPIE (VERSION 2002-1) **************2831 1451 Unshielded Group Averages Using 640 Groups 2831 1451 Weighting Spectrum: Flat (Constant) Spectrum 2831 1451 1 451 402 42831 1451 3 103 61 12831 1451 33 103 14 12831 1451 2831 1 0 2831 0 0 2.80600E+4 5.94159E+1 0 0 0 02831 3103 -2.04150E+6-2.04150E+6 0 0 1 1742831 3103 174 1 2831 3103 2700000.00 1.27290E-8 2800000.00 3.81870E-8 2900000.00 6.36450E-82831 3103 3000000.00 2.39704E-6 3100000.00 7.03836E-6 3200000.00 1.16797E-52831 3103 3300000.00 1.63210E-5 3400000.00 2.09623E-5 3500000.00 5.81607E-52831 3103 3600000.00 .000127916 3700000.00 .000197672 3800000.00 .0002674272831 3103 3900000.00 .000337182 4000000.00 .000553104 4100000.00 .0009151922831 3103 4200000.00 .001277280 4300000.00 .001639368 4400000.00 .0020014562831 3103 4500000.00 .002751840 4600000.00 .003890520 4700000.00 .0050292002831 3103 4800000.00 .006167880 4900000.00 .007306560 5000000.00 .0085883102831 3103 5100000.00 .010013130 5200000.00 .011437950 5300000.00 .0128627702831 3103 5400000.00 .014287590 5500000.00 .015850000 5600000.00 .0175500002831 3103 5700000.00 .019250000 5800000.00 .020950000 5900000.00 .0226500002831 3103 6000000.00 .024550000 6100000.00 .026650000 6200000.00 .0287500002831 3103 6300000.00 .030850000 6400000.00 .032950000 6500000.00 .0350000002831 3103 6600000.00 .037000000 6700000.00 .039000000 6800000.00 .0410000002831 3103 6900000.00 .043000000 7000000.00 .045200000 7100000.00 .0476000002831 3103 7200000.00 .050000000 7300000.00 .052400000 7400000.00 .0548000002831 3103 7500000.00 .056900000 7600000.00 .058700000 7700000.00 .0605000002831 3103 7800000.00 .062300000 7900000.00 .064100000 8000000.00 .0660000002831 3103 8100000.00 .068000000 8200000.00 .070000000 8300000.00 .0720000002831 3103 8400000.00 .074000000 8500000.00 .075800000 8600000.00 .0774000002831 3103 8700000.00 .079000000 8800000.00 .080600000 8900000.00 .0822000002831 3103 9000000.00 .083950000 9100000.00 .085850000 9200000.00 .0877500002831 3103 9300000.00 .089650000 9400000.00 .091550000 9500000.00 .0933500002831 3103 9600000.00 .095050000 9700000.00 .096750000 9800000.00 .0984500002831 3103 9900000.00 .100150000 10000000.0 .101850000 10100000.0 .1035500002831 3103 10200000.0 .105250000 10300000.0 .106950000 10400000.0 .1086500002831 3103 10500000.0 .110350000 10600000.0 .112050000 10700000.0 .1137500002831 3103 10800000.0 .115450000 10900000.0 .117150000 11000000.0 .1188500002831 3103 11100000.0 .120550000 11200000.0 .122250000 11300000.0 .1239500002831 3103 11400000.0 .125650000 11500000.0 .127350000 11600000.0 .1290500002831 3103 11700000.0 .130750000 11800000.0 .132450000 11900000.0 .1341500002831 3103 12000000.0 .135550000 12100000.0 .136650000 12200000.0 .1377500002831 3103 12300000.0 .138850000 12400000.0 .139950000 12500000.0 .1410500002831 3103 12600000.0 .142150000 12700000.0 .143250000 12800000.0 .1443500002831 3103 12900000.0 .145450000 13000000.0 .146200000 13100000.0 .1466000002831 3103 13200000.0 .147000000 13300000.0 .147400000 13400000.0 .1478000002831 3103 13500000.0 .147850000 13600000.0 .147550000 13700000.0 .1472500002831 3103 13800000.0 .146950000 13900000.0 .146650000 14000000.0 .1460520002831 3103 14100000.0 .145156000 14200000.0 .144260000 14300000.0 .1433640002831 3103 14400000.0 .142468000 14500000.0 .140559000 14600000.0 .1376370002831 3103 14700000.0 .134715000 14800000.0 .131793000 14900000.0 .1288710002831 3103 15000000.0 .125949000 15100000.0 .123027000 15200000.0 .1201050002831 3103 15300000.0 .117183000 15400000.0 .114261000 15500000.0 .1113390002831 3103 15600000.0 .108417000 15700000.0 .105495000 15800000.0 .1025730002831 3103 15900000.0 .099651000 16000000.0 .097483667 16100000.0 .0960710002831 3103 16200000.0 .094658333 16300000.0 .093245667 16400000.0 .0918330002831 3103 16500000.0 .090420333 16600000.0 .089007667 16700000.0 .0875950002831 3103 16800000.0 .086182333 16900000.0 .084769667 17000000.0 .0833570002831 3103 17100000.0 .081944333 17200000.0 .080531667 17300000.0 .0791190002831 3103 17400000.0 .077706333 17500000.0 .076620000 17600000.0 .0758600002831 3103 17700000.0 .075100000 17800000.0 .074340000 17900000.0 .0735800002831 3103 18000000.0 .072820000 18100000.0 .072060000 18200000.0 .0713000002831 3103 18300000.0 .070540000 18400000.0 .069780000 18500000.0 .0690200002831 3103 18600000.0 .068260000 18700000.0 .067500000 18800000.0 .0667400002831 3103 18900000.0 .065980000 19000000.0 .065220000 19100000.0 .0644600002831 3103 19200000.0 .063700000 19300000.0 .062940000 19400000.0 .0621800002831 3103 19500000.0 .061420000 19600000.0 .060660000 19700000.0 .0599000002831 3103 19800000.0 .059140000 19900000.0 .058380000 20000000.0 0.0 2831 3103 2831 3 0 2831 0 0 2.80600E+4 5.94160E+1 0 0 0 1283133103 0.000000+0 0.000000+0 0 103 0 4283133103 0.000000+0 0.000000+0 0 0 6 3283133103 1.000000-5 0.000000+0 2.700000+6 1.00000-11 2.000000+7 0.000000+0283133103 0.000000+0 0.000000+0 0 1 6 3283133103 1.000000-5 0.000000+0 2.700000+6 5.000000-3 2.000000+7 0.000000+0283133103 0.000000+0 0.000000+0 0 1 16 8283133103 1.000000-5 0.000000+0 2.700000+6 7.200000-2 5.000000+6 1.800000-2283133103 7.000000+6 1.012500-2 1.000000+7 1.012500-2 1.200000+7 4.500000-3283133103 1.600000+7 1.800000-2 2.000000+7 0.000000+0 283133103 0.000000+0 0.000000+0 0 8 16 8283133103 1.000000-5 0.000000+0 2.700000+6 4.66640-17 5.000000+6 1.240600-7283133103 7.000000+6 2.178000-6 1.000000+7 1.147600-5 1.200000+7 9.112500-6283133103 1.600000+7 1.928300-5 2.000000+7 0.000000+0 283133103 283133 0 2831 0 0 0 0 0 2.90630E+4 6.23890E+1 0 0 34 102925 1451 0.0 0.0 0 0 0 62925 1451 1.00000E+0 2.00000E+7 0 0 10 20022925 1451 3.00000E+2 0.0 1 0 550 72925 1451 29-Cu- 63 LANL,ORNL,FEI EVAL-FEB98 A.KONING ET AL,K.Zolotarev 2925 1451 DIST-Feb2004 2925 1451 ----IRDF-2002 MATERIAL 2925 2925 1451 -----INCIDENT NEUTRON DATA 2925 1451 ------ENDF-6 FORMAT 2925 1451 *****************EXTRACT FOR SPECIAL PURPOSE FILE*****************2925 1451 DOSIMETRY 2925 1451 ******************************************************************2925 1451 29-Cu- 63 LANL,ORNL EVAL-FEB98 A.KONING,M.CHADWICK,HETRICK 2925 1451 Ch98,Ch99 DIST-SEP 1 REV4- 20010926 2925 1451 ----ENDF/B-VI MATERIAL 2925 REVISION 3 2925 1451 **************************************************************** 2925 1451 ***************************************************************** 2925 1451 29-Cu- 63 FEI EVAL-Nov01 K.I.Zolotarev 2925 1451 DIST-Jan02 20020110 2925 1451 ----BROND-2 MATERIAL 2925 2925 1451 -----INCIDENT NEUTRON DATA 2925 1451 ------ENDF-6 FORMAT 2925 1451 ------Russian Reactor Dosimetry File RRDF-2002 2925 1451 ***************************************************************** 2925 1451 Author of evaluation: K.I.Zolotarev 2925 1451 ----- MF=3 MT=107 ----- 2925 1451 ***************************************************************** 2925 1451 ******** Start of (N,2N), (N.G) bibliographical component ******* 2925 1451 ***************************************************************** 2925 1451 2925 1451 ENDF/B-VI MOD 5 Revision, May 2000, S.C. Frankle, R.C. Reedy, 2925 1451 P.G. Young (LANL) 2925 1451 2925 1451 The secondary gamma-ray spectrum for radiative capture (MF 12, 2925 1451 MT 102) has been updated for new experimental data at incident 2925 1451 neutron energies up to 1 keV. The Q-value for radiative capture 2925 1451 was also updated in File 3. 2925 1451 Details of these changes are described in Frankel et al. [Fr01]. 2925 1451 2925 1451 **************************************************************** 2925 1451 2925 1451 ENDF/B-VI MOD 4 Evaluation, February 1998, A.J. Koning (ECN), 2925 1451 M.B. Chadwick, P.G. Young (LANL) 2925 1451 2925 1451 Los Alamos LA150 Library, produced with FKK/GNASH/GSCAN code 2925 1451 in cooperation with ECN Petten. 2925 1451 2925 1451 This evaluation provides a complete representation of the 2925 1451 nuclear data needed for transport, damage, heating, 2925 1451 radioactivity, and shielding applications over the incident 2925 1451 neutron energy range from 1.0E-11 to 150 MeV. The discussion 2925 1451 here is divided into the region below and above 20 MeV. 2925 1451 2925 1451 INCIDENT NEUTRON ENERGIES < 20 MeV 2925 1451 2925 1451 Below 20 MeV the evaluation is based completely on the ENDF/B- 2925 1451 VI (Mod 3) evaluation by D. Hetrick, C.Y. Fu, and D. Larson. 2925 1451 2925 1451 INCIDENT NEUTRON ENERGIES > 20 MeV 2925 1451 2925 1451 The ENDF/B-VI Release 2 evaluation extends to 20 MeV and 2925 1451 includes cross sections and energy-angle data for all 2925 1451 significant reactions. The present evaluation utilizes a more 2925 1451 compact composite reaction spectrum representation above 20 MeV 2925 1451 in order to reduce the length of the file. No essential data for 2925 1451 applications is lost with this representation. 2925 1451 The evaluation above 20 MeV utilizes MF=6, MT=5 to represent 2925 1451 all reaction data. Production cross sections and emission 2925 1451 spectra are given for neutrons, protons, deuterons, tritons, 2925 1451 alpha particles, gamma rays, and all residual nuclides produced 2925 1451 (A>5) in the reaction chains. To summarize, the ENDF sections 2925 1451 with non-zero data above En = 20 MeV are: 2925 1451 2925 1451 MF=3 MT= 1 Total Cross Section 2925 1451 MT= 2 Elastic Scattering Cross Section 2925 1451 MT= 3 Nonelastic Cross Section 2925 1451 MT= 5 Sum of Binary (n,n') and (n,x) Reactions 2925 1451 2925 1451 MF=4 MT= 2 Elastic Angular Distributions 2925 1451 2925 1451 MF=6 MT= 5 Production Cross Sections and Energy-Angle 2925 1451 Distributions for Emission Neutrons, Protons, 2925 1451 Deuterons, Tritons, and Alphas; and Angle- 2925 1451 Integrated Spectra for Gamma Rays and Residual 2925 1451 Nuclei That Are Stable Against Particle Emission 2925 1451 2925 1451 The evaluation is based on nuclear model calculations that 2925 1451 have been benchmarked to experimental data, especially for n + 2925 1451 Cu65 and p + Cu65 reactions [Ch98]. We use the GNASH code system 2925 1451 [Yo92], which utilizes Hauser-Feshbach statistical, preequilib- 2925 1451 rium and direct-reaction theories. Spherical optical model 2925 1451 calculations are used to obtain particle transmission 2925 1451 coefficients for the Hauser-Feshbach calculations, as well as 2925 1451 for the elastic neutron angular distributions. 2925 1451 Cross sections and spectra for producing individual residual 2925 1451 nuclei are included for reactions. The energy-angle-correlations 2925 1451 for all outgoing particles are based on Kalbach systematics 2925 1451 [Ka88]. 2925 1451 A model was developed to calculate the energy distributions of 2925 1451 all recoil nuclei in the GNASH calculations [Ch96a]. The recoil 2925 1451 energy distributions are represented in the laboratory system in 2925 1451 MT=5, MF=6, and are given as isotropic in the lab system. All 2925 1451 other data in MT=5,MF=6 are given in the center-of-mass system. 2925 1451 This method of representation utilizes the LCT=3 option approved 2925 1451 at the November, 1996, CSEWG meeting. 2925 1451 Preequilibrium corrections were performed in the course of the 2925 1451 GNASH calculations using the exciton model of Kalbach [Ka77, 2925 1451 Ka85], validated by comparison with calculations using Feshbach, 2925 1451 Kerman, Koonin (FKK) theory [Ch93]. Discrete level data from 2925 1451 nuclear data sheets were matched to continuum level densities 2925 1451 using the formulation of Ignatyuk et al. [Ig75] and pairing and 2925 1451 shell parameters from the Cook [Co67] analysis. Neutron and 2925 1451 charged- particle transmission coefficients were obtained from 2925 1451 the optical potentials, as discussed below. Gamma-ray 2925 1451 transmission coefficients were calculated using the Kopecky-Uhl 2925 1451 model [Ko90]. 2925 1451 2925 1451 SPECIFIC INFORMATION CONCERNING THE 63Cu EVALUATION 2925 1451 2925 1451 This evaluation is documented in some detail in Ref. [Ko98b]. 2925 1451 2925 1451 The neutron total cross section above 20 MeV was obtained by 2925 1451 evaluating experimental data, with a particular emphasis on the 2925 1451 Finlay [Fi93] elemental data. This resulted in an evaluated 2925 1451 elemental Cu total cross section; to obtain an isotopic 63Cu total2925 1451 cross section, it was assumed that 63Cu and 65Cu have total cross 2925 1451 sections in an A**2/3 ratio to one another. The total neutron 2925 1451 nonelastic cross section was obtained directly from an optical 2925 1451 model calculation (see below), after verifying that it was in good2925 1451 agreement with the experimental data [Ko98b]. 2925 1451 2925 1451 To obtain the neutron optical potential we used total cross 2925 1451 section data from 1.2 to 4.5 MeV [Gu86] and from 5.3 to 600 MeV 2925 1451 [Fi93], and elastic scattering angular distribution data from 1.6 2925 1451 to 96 MeV [Br50, Sa60, Ki74, El82, Gu86]. The optical potential 2925 1451 parameters were obtained using a combination of a grid search code2925 1451 and the interactive optical model viewer ECISVIEW [Ko97], both 2925 1451 built around the coupled channels code ECIS96 [Ra94]. The energy 2925 1451 dependence of the optical model parameters is as described in 2925 1451 [Ko98]. This optical potential was used for the calculation, with 2925 1451 ECIS96, of neutron transmission coefficients and DWBA cross 2925 1451 sections for the entire energy region above 20 MeV. 2925 1451 2925 1451 Due to the lack of proton elastic scattering data in numerical 2925 1451 form, we used a combination of global optical models for the 2925 1451 proton channel. The Becchetti-Greenlees potential [Be69]was 2925 1451 adopted below 47 MeV, and the non-relativistic version of the 2925 1451 Madland potential [Ma88] above 47 MeV. At this particular energy 2925 1451 point the two potentials join smoothly. 2925 1451 2925 1451 For deuterons, the Lohr-Haeberli global potential [Lo74] was used;2925 1451 for alpha particles the Moyen potential (MacFadden-Satchler 2925 1451 [Ma66]) was used; and for tritons the Becchetti-Greenlees 2925 1451 potential [Be71] was used. The He-3 channel was ignored, due to 2925 1451 its small importance. 2925 1451 2925 1451 Following Delaroche et al. [De82], we adopted the weak-coupling 2925 1451 model for direct collective inelastic scattering for Cu-63, using 2925 1451 Ni-64 as a basis. For the calculation of the cross sections, 2925 1451 ECIS96 was used in DWBA mode. We used the following direct 2925 1451 transitions for Cu-63 (ground state 3/2- ) : 2925 1451 2925 1451 Jpi Ex(MeV) Deformation lengths 2925 1451 0.5- 0.669 Delta(2)=0.319 2925 1451 2.5- 0.962 Delta(2)=0.552 2925 1451 3.5- 1.327 Delta(2)=0.638 2925 1451 1.5- 1.547 Delta(2)=0.451 2925 1451 1.5- 3.382 Delta(3)=0.313 2925 1451 2.5- 3.632 Delta(3)=0.384 2925 1451 3.5- 3.882 Delta(3)=0.444 2925 1451 4.5- 4.132 Delta(3)=0.496 2925 1451 2925 1451 No measurements exist for neutron-induced emission spectra above 2925 1451 20 MeV for 63Cu. However, for Cu-65 there exists 25.7 MeV (n,xn) 2925 1451 data by Marcinkowski et al [Ma83]. This has been used to benchmark2925 1451 the Cu-65 data. Without adjusting any of the level density or pre-2925 1451 equilibrium parameters the GNASH calculation was in good agreement2925 1451 with these data. Hence we also adopted these parameters for the 2925 1451 whole energy region for Cu-63. 2925 1451 2925 1451 **************************************************************** 2925 1451 2925 1451 REFERENCES 2925 1451 2925 1451 [Ab93] W. Abfalterer, R.W. Finlay, S.M. Grimes, and V. Mishra, 2925 1451 Phys.Rev. C47, 1033 (1993) 2925 1451 [Al83] R. Alarcon and J. Rapaport, Nucl.Phys. A458, 502 (1986) 2925 1451 [Ar80] E.D. Arthur and P.G. Young, 'Evaluation of Neutron Cross 2925 1451 Sections to 40 MeV for 54,56Fe," Proc. Sym. on Neutron Cross 2925 1451 Sections from 10 to 50 MeV, 12-14 May 1980, Brookhaven National2925 1451 Laboratory [Eds. M. R. Bhat and S. Pearlstein, BNL-NCS- 51245, 2925 1451 1980] p. 731. 2925 1451 [Be69] F.D. Becchetti, Jr., and G.W. Greenlees, Phys.Rev. 182, 2925 1451 1190 (1969) 2925 1451 [Be71] F.D. Becchetti, Jr., and G.W. Greenlees in "Polarization 2925 1451 Phenomena in Nuclear Reactions," (Ed: H.H. Barschall and W. 2925 1451 Haeberli, The University of Wisconsin Press, 1971) p.682. 2925 1451 [Be92] O. Bersillon, "SCAT2 - A Spherical Optical Model Code," 2925 1451 in Proc. ICTP Workshop on Computation and Analysis of Nuclear 2925 1451 data Relevant to Nuclear Energy and Safety, February-March, 2925 1451 1999 Trieste, Italy, to be published in World Scientific Press,2925 1451 and Progress Report of the Nuclear Physics Division, Bruyeres- 2925 1451 le-Chatel 1977, CEA-N-2037 (1978) p.111 2925 1451 [Br50] S. Bratenahl, S. Fernbach, R.H. Hildebrand et al., 2925 1451 Phys.Rev. 77, 597 (1950) 2925 1451 [Ch93] M.B. Chadwick and P.G. Young, Phys.Rev. C 47, 2255 (1993) 2925 1451 [Ch96] M.B. Chadwick, P.G. Young, R.E. MacFarlane, and A.J. 2925 1451 Koning, "High-Energy Nuclear Data Libraries for Accelerator- 2925 1451 Driven Technologies: Calculational Method for Heavy Recoils," 2925 1451 Proc. of 2nd Int. Conf. on Accelerator Driven Transmutation 2925 1451 Technology and Applications, Kalmar, Sweden, 3-7 June 1996 2925 1451 [Ch98] M. B. Chadwick and P. G. Young, "GNASH Calculations of 2925 1451 n,p + Cu isotopes and Benchmarking of Results" in APT PROGRESS 2925 1451 REPORT: 1 February - 1 March 1998, internal Los Alamos National2925 1451 Laboratory memo, 6 Mar.1998 from R.E. MacFarlane to L. Waters. 2925 1451 [Ch99] M.B. Chadwick, P G. Young, G. M. Hale, et al., Los Alamos 2925 1451 National Laboratory report, LA-UR-99-1222 (1999) 2925 1451 [Co67] J.L. Cook, H. Ferguson, and A.R. DeL Musgrove, Aust.J. 2925 1451 Phys. 20, 477 (1967) 2925 1451 [De82] J.P. Delaroche, S.M. El-Kadi, P.P. Guss, C.E. Floyd and 2925 1451 R.L. Walter, Nucl. Phys. A390, 541 (1982). 2925 1451 [El82] S.M. El-Kadi, C.E. Nelson, F.O. Purser et al., Nucl.Phys. 2925 1451 A390, 509 (1982) 2925 1451 [Fi93] R. W. Finlay, W. P. Abfalterer, G. Fink et al., Phys.Rev. 2925 1451 C 47, 237 (1993) 2925 1451 [Fr01] S.C. Frankle, R.C. Reedy, and P.G. Young, Los ALamos 2925 1451 National Laboratory Report, LA-13812 (2001). 2925 1451 [Gu86] P. Guenther, D.L. Smith, A.B. Smith, J.F. Whalen, Nucl. 2925 1451 Phys. A448, 280 (1986) 2925 1451 [Ig75] A.V. Ignatyuk, G.N. Smirenkin, and A.S. Tishin, Sov.J. 2925 1451 Nucl.Phys. 21, 255 (1975); translation of Yad.Fiz. 21, 485 2925 1451 (1975) 2925 1451 [Ka77] C. Kalbach, Z.Phys.A 283, 401 (1977) 2925 1451 [Ka85] C. Kalbach, Los Alamos National Laboratory report 2925 1451 LA-10248-MS (1985) 2925 1451 [Ka88] C. Kalbach, Phys.Rev.C 37, 2350 (1988); see also 2925 1451 C. Kalbach and F. M. Mann, Phys.Rev.C 23, 112 (1981) 2925 1451 [Ki74] W.E. Kinney, F.G. Perey, Oak Ridge report ORNL-4908 (1974)2925 1451 [Ko90] J. Kopecky and M. Uhl, Phys.Rev.C 41, 1941 (1990) 2925 1451 [Ko97] A.J. Koning, J.J. van Wijk and J.-P. Delaroche, "ECISVIEW:2925 1451 A Graphical Interface for ECIS95", Proceedings of the NEA 2925 1451 Specialists' Meeting on the Nucleon Nucleus Optical Model up to2925 1451 200 MeV, Bruyeres-le-Chatel, November 13-15 1996. Available at 2925 1451 http://db.nea.fr/html/science/om200/. 2925 1451 [Ko98] A.J. Koning, J.-P. Delaroche and O. Bersillon, "Nuclear 2925 1451 Data for Accelerator-Driven Systems: Nuclear models, Experiment2925 1451 and Data Libraries", to appear in Mucl. Instr. Meth. A (1998). 2925 1451 [Ko98b] A.J. Koning, M.B. Chadwick, and P.G. Young, "ENDF/B-VI 2925 1451 neutron and proton datafiles up to 150 MeV for 63Cu and 65Cu", 2925 1451 Los Alamos National Laboratory report LAUR- (1998); ECN lab and2925 1451 JEFF report (1998). 2925 1451 [Lo74] J.M. Lohr and W. Haeberli, Nucl.Phys. A232, 381 (1974) 2925 1451 [Ma66] Macfadden and Satchler, Nuc.Phys. 84, 177 (1966) 2925 1451 [Ma83] A. Marcinkowski, R.W. Finlay, G. Randers-Pehrson et al., 2925 1451 Nucl.Phys. A402, 220 (1983) 2925 1451 [Ma88] D.G. Madland, "Recent Results in the Development of a 2925 1451 Global Medium-Energy Nucleon-Nucleus Optical-Model Potential, 2925 1451 "Proc. OECD/NEANDC Specialist's Mtg. on Preequilibrium Nuclear 2925 1451 Reactions, Semmering, Austria, 10-12 Feb. 1988, NEANDC-245 'U' 2925 1451 (1988). 2925 1451 [Pe63] C.M. Perey and F.G. Perey, Phys.Rev. 132, 755 (1963) 2925 1451 [Ra94] J. Raynal, Notes on ECIS94, CEA Saclay Report CEA-N-2772 2925 1451 (1994) 2925 1451 [Sa60] G.L. Salmon, Nucl.Phys. 21, 15 (1960) 2925 1451 [Yo92] P.G. Young, E.D. Arthur, and M.B. Chadwick, report 2925 1451 LA-12343-MS (1992) 2925 1451 2925 1451 **************************************************************** 2925 1451 2925 1451 ENDF/B-VI MOD 3 Revision, July 1991 (ORNL) 2925 1451 2925 1451 MOD 3 changes 2925 1451 1) Corrections to MF=6, MT=65 at 17.0 MeV to prevent negative 2925 1451 values in the angular distribution. 2925 1451 2) Corrections to MF=33, MT=102 2925 1451 2925 1451 **************************************************************** 2925 1451 2925 1451 * Note there was no MOD 2 released. 2925 1451 2925 1451 **************************************************************** 2925 1451 2925 1451 ENDF/B-VI MOD 1 Evaluation, October 1989, D. Hetrick, F.Y. Fu, 2925 1451 D. Larson (ORNL) 2925 1451 2925 1451 This work employed several nuclear model codes including the 2925 1451 optical-model code GENOA [1], the Distorted Wave Born 2925 1451 Approximation (DWBA) program DWUCK [2], and the Hauser-Feshbach 2925 1451 code TNG [3,4]. The TNG code provides energy and angular 2925 1451 distributions of particles emitted in the compound and pre- 2925 1451 compound reactions, ensures consistency among all reactions, and 2925 1451 maintains energy balance. Details pertinent to the contents of 2925 1451 this evaluation and extensive comparisons of calculations with 2925 1451 experimental data can be found in reference [5]. 2925 1451 2925 1451 ----- DESCRIPTION OF FILES 2925 1451 (MF-MT) 2925 1451 1-451 GENERAL INFORMATION, REFERENCES, AND DEFINITIONS. 2925 1451 2-151 RESONANCE PARAMETERS WERE TAKEN FROM MUGHABGHAB[6]. POINT2925 1451 WISE RECONSTRUCTION COMPARED WITH DATA [7] SHOWED POORER 2925 1451 FIT ABOVE 100 KEV, SO THE RESONANCE REGION WAS CUT OFF AT 2925 1451 99.5 KEV. REICH-MOORE PARAMETERS ARE GIVEN. AGREEMENT 2925 1451 WITH DATA COULD BE IMPROVED WITH ADDITION OF A BACKGROUND 2925 1451 FILE IN 3/1, BUT THIS IN GENERAL GIVES TOO LARGE AN 2925 1451 AVERAGE CROSS SECTION, WHEN BINNED IN 10 KEV BINS AND 2925 1451 COMPARED WITH THE BINNED DATA. THIS IS PROBABLY DUE TO 2925 1451 TOO LARGE AN ESTIMATE OF NEUTRON WIDTHS FOR RESONANCES 2925 1451 SEEN ONLY IN CAPTURE AND NOT IN TRANSMISSION. 2925 1451 NOTE THAT THE FLAG HAS BEEN SET TO ALLOW USER CALCULATION 2925 1451 OF THE ANGULAR DISTRIBUTIONS FROM THE R-M RESONANCE 2925 1451 PARAMETERS, IF THE USER WANTS ANGULAR DISTRIBUTIONS ON 2925 1451 A FINER ENERGY GRID THAN GIVEN IN 4/2. 2925 1451 3-1 THE TOTAL CROSS SECTION IS GIVEN BY RESONANCE PARAMETERS 2925 1451 FROM 1.E-5 EV TO 99.5 KEV. FROM 1.E-5 TO 1 EV, -0.9B IS 2925 1451 GIVEN TO REDUCE THE ELASTIC AND GIVE THE CORRECT TOTAL 2925 1451 THERMAL CROSS SECTION (9.6 B). FROM 1 EV TO 170 EV THIS 2925 1451 GOES LINEARLY TO ZERO. A SMALL CONTRIBUTION FROM 3/102 IS2925 1451 REQUIRED FROM 60 TO 99.5 KEV, WHICH WHEN ADDED TO 2/151 2925 1451 REPRODUCES THE CAPTURE DATA. FROM 99.5 KEV TO 1.12 MEV 2925 1451 CU63 DATA FROM [7] IS USED, AFTER APPROPRIATE AVERAGING. 2925 1451 ABOVE THIS, NO ISOTOPIC DATA IS AVAILABLE. FROM 1.12 TO 2925 1451 4.0 MEV,NAT CU DATA OF PEREY [8] USED IN V5 IS RETAINED. 2925 1451 FROM 4.0 TO 20 MEV, NAT CU DATA OF LARSON ET.AL [9] IS 2925 1451 AVERAGED AND USED. COMPARISONS FROM 1.2 TO 4.5 2925 1451 MEV WITH AVERAGED ARGONNE DATA FOR NAT CU [10] SHOW 2925 1451 1% AGREEMENT. 2925 1451 3-2 ELASTIC SCATTERING CROSS SECTIONS WERE OBTAINED BY 2925 1451 SUBTRACTING THE NONELASTIC (3-3) FROM THE TOTAL. THE 2925 1451 THERMAL VALUE OF 5.1 B IS REPRODUCED. 2925 1451 3-102 (N,G) DATA TAKEN FROM RESONANCE PARAMETERS FROM 1.E-5 EVTO2925 1451 99.5 KEV. A SMALL BACKGROUND IS GIVEN HERE WHICH WHEN 2925 1451 ADDED TO THE RESONACE CONTRIBUTION REPRODUCES EXPERIMENTAL2925 1451 DATA,INCLUDING THE THERMAL VALUE OF 4.50B. V5 DATA FROM 2925 1451 99.5 KEV TO 20.0 MEV WERE REPLACED BY POINTS ON A CURVE 2925 1451 DRAWN THROUGH DATA FROM THE CSISRS LIBRARY [5,13]; RESULTS2925 1451 COMPARABLE TO EYE GUIDE IN REF [14]. 2925 1451 ---------------------------------------------------------------- 2925 1451 UNCERTAINTY FILES 2925 1451 ALL NON-DERIVED FILES CONTAIN AN LB=8 COMPONENT, AS 2925 1451 REQUIRED BY ENDF/B-VI FORMATS 2925 1451 2925 1451 33-1 TOTAL UNCERTAINTIES GIVEN AS DERIVED FROM 1E-5 TO 200 EV 2925 1451 EXPLICIT FROM 200 EV TO 20 MEV, USING LB=0,1 AND 8. 2925 1451 33-2 EXPLICIT FROM 1E-5 T0 200 EV, DERIVED FROM 200EV TO 20 MEV2925 1451 33-102 CAPTURE UNCERTAINTIES ESTIMATED FROM THERMAL VALUE AT LOW 2925 1451 ENERGIES, BINNED DATA IN THE RESONANCE REGION, AND CSISRS 2925 1451 DATA [5,13,14] FROM 99.5 KEV TO 20 MEV. 2925 1451 **************************************************************** 2925 1451 2925 1451 REFERENCES: 2925 1451 2925 1451 [1] F.G. Perey, computer code GENOA, ORNL, unpublished (1967) 2925 1451 [2] P.D. Kunz, "Distorted Wave Code DWUCK72," Univ. of 2925 1451 Colorado, unpublished (1972) 2925 1451 [3] C.Y. Fu, report ORNL/TM-7042 (1980); also, C.Y Fu, 2925 1451 Symp. on Neutron Cross Sections from 10 to 50 MeV, Upton, NY,2925 1451 May 1980, Brookhaven National Lab. report BNL-NCS-51245 2925 1451 (1980) p.675 2925 1451 [4] K. Shibata and C.Y. Fu, report ORNL/TM-10093 (1986) 2925 1451 [5] D.M. Hetrick, C.Y. Fu, and D.C. Larson, Oak Ridge report 2925 1451 ORNL/TM-9083 [ENDF-337] (1984) 2925 1451 [6] S.F. Mughabghab, M. Divadeenam, and N.E. Holden, "Neutron 2925 1451 Cross Sections, Vol. 1, Neutron Resonance Parameters and 2925 1451 Thermal Cross Sections, Part A, Z=1-60," (Academic Press, 2925 1451 1981) 2925 1451 [7] M.S. Pandey, J.B. Garg and J.A. Harvey, Phys.Rev. C 15, 600 2925 1451 (1977), and private communication. 2925 1451 [8] F.G. Perey, private communication (1977) 2925 1451 [9] D.C. Larson, Symp. on Neutron Cross Sections from 10 to 50 2925 1451 MeV, Upton, NY, May 1980, Brookhaven National Lab. report 2925 1451 BNL-NCS-51245 (1980) p.277 2925 1451 [10] P. Guenther, D.L. Smith, A.B. Smith and J.F. Whalen, Nucl. 2925 1451 Phys. A, 448, 280 (1986) [CSISRS data set 12869/002], and 2925 1451 W.P. Poenitz and J.F. Whalen, Argonne report ANL/NDM-80 2925 1451 (1983) [CSISRS data set 12853] 2925 1451 [11] D.M. Hetrick and C.Y. Fu, Oak Ridge report ORNL/TM-7341 2925 1451 [ENDF-303] (1980) 2925 1451 [12] C.Y. Fu and D.M. Hetrick, Proc. Fourth ASTM-Euratom Symp. 2925 1451 on Reactor Dosimetry, Gaithersburg, Maryland, March 22-26, 2925 1451 1982 (U.S. National Bureau of Standards) p.877 2925 1451 [13] CSISRS Library, National Nuclear Data Center, Brookhaven 2925 1451 National Laboratory, Upton, N.Y. 11973. 2925 1451 [14] V. McLane, C.L. Dunford and P.F. Rose, "Neutron Cross 2925 1451 Sections, Vol. 2, Neutron Cross Section Curves" (Academic 2925 1451 Press, 1988) 2925 1451 [15] S.M. Qaim, Radiochimica Acta, 25, 13 (1978) 2925 1451 [16] M.G. Delfini, J. Kopecky, R.E. Chrien et al., Nucl.Phys. 2925 1451 A404, 250 (1983) 2925 1451 2925 1451 ***************************************************************** 2925 1451 ******** End of (N,2N), (N.G) bibliographical component ******* 2925 1451 ***************************************************************** 2925 1451 ***************************************************************** 2925 1451 ******** Start of (N,A) bibliographical component ******* 2925 1451 ***************************************************************** 2925 1451 ------Russian Reactor Dosimetry File RRDF-2002 2925 1451 ***************************************************************** 2925 1451 Author of evaluation: K.I.Zolotarev 2925 1451 ***************************************************************** 2925 1451 2925 1451 ----- MF=3 MT=107 ----- 2925 1451 Evaluation of Cu-63(n,a)Co-60m+g- excitation function was car-2925 1451 ried out by means of statistical analysis of cross sections from 2925 1451 data base prepared in the energy range 2 - 20 Mev. In the energy 2925 1451 range 3.56 - 19.55 MeV input data base was formed with using of 2925 1451 experimental data from ref. [1-19]. Cross section data in the 2925 1451 interval 2.0 - 3.5 MeV were taken from theoretical model calcula- 2925 1451 tion. Experimental data included in the input data base were 2925 1451 renormalized using new cross sections standards for monitor reac- 2925 1451 tions and new standards for decay data. 2925 1451 The special correction was applied to the experimental data 2925 1451 [2,3,12,13,15,18]. Cross section data of A.Paulsen and H.Liskien 2925 1451 [2-3] measured in the energy region 12.09 - 19.55 MeV with using 2925 1451 T(d,n)He4 neutron source were multiplied to the factor 1.20805 . 2925 1451 Experimental data of Lu Hanlin et al. [12,18] , Wang Yongchag et 2925 1451 al. [13] and Konno et al. [15] were multiplied to the factors 2925 1451 0.88110, 0.86000, 0.84377 and 1.060, respectively. The correction 2925 1451 factors were derived from preliminary evaluated cross sections 2925 1451 integrals in the energy intervals 8.4 - 11.4 and 13 - 15 MeV . 2925 1451 Data of A.Paulsen and H.Liskien [2] in the energy range 2925 1451 5.76 - 11.48 MeV measured with using D(d,n)He3 , Be9(a,n)C12 , 2925 1451 C14(d,n)N15 , N15(d,n)O16 neutron sources were rejected due to 2925 1451 their inconsistency with precision measurements of G.Winkler et 2925 1451 al. [8] and integral experimental data for U-235 fission neutron 2925 1451 spectrum [20-23] and Cf-252 spontaneous fission neutron spectrum 2925 1451 [25-26]. Cross sections measured by J.Kantele and D.Gardner [27], 2925 1451 M.Bormann et al. [28], G.Maslov et al. [29] and K.Kayashima et al.2925 1451 [30] were also rejected due to the big discrepancy with the main 2925 1451 bulk experimental data. 2925 1451 The final procedure evaluation of (n,a) excitation function 2925 1451 was carried out by means of Pade-2 code [31]. 2925 1451 Evaluated excitation function for the reaction Cu63(n,a)Co60 2925 1451 was tested with using integral experimental data [20-24] for 2925 1451 U-235 thermal fission neutron spectrum and evaluated integral ex- 2925 1451 perimental data [32-33] for Cf-252 spontaneous fission neutron 2925 1451 spectrum. Calculated and measured average cross section values 2925 1451 for U-235 thermal fission neutron spectrum [34] and Cf-252 sponta-2925 1451 neous fission neutron spectrum [35] are given in the table 1. 2925 1451 Table 1 2925 1451 ÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÂÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÂÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄ 2925 1451 TYPE OF SPECTRUM ³ ,mb (calc.) ³ , mb (measured) 2925 1451 ÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÅÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÅÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄ 2925 1451 U-235 neutron fission ³ 0.53294 ³ 0.5269 +- 0.0316 [20] 2925 1451 ³ ³ 0.520 +- 0.040 [21] 2925 1451 ³ ³ 0.519 +- 0.035 [22] 2925 1451 ³ ³ 0.5273 +- 0.0156 [23] 2925 1451 ³ ³ 0.538 +- 0.015 [23] 2925 1451 ³ ³ 0.4935 +- 0.0242 [24] 2925 1451 ÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÅÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÅÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄ 2925 1451 CF-252 spont. fission ³ 0.69248 ³ 0.709 +- 0.017 [25] 2925 1451 ³ ³ 0.675 +- 0.018 [26] 2925 1451 ³ ³ 0.6897 +- 0.0130 [32] 2925 1451 ³ ³ 0.6887 +- 0.0135 [33] 2925 1451 ÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÁÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÁÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄ 2925 1451 2925 1451 ----- MF=33 MT=107 ----- 2925 1451 Uncertainties in the evaluated excitation function for the 2925 1451 reaction Cu-63(n,a)Co-60m+g are given in the form of relative co- 2925 1451 variance matrix for the 18-neutron energy groups (LB=5). Covari- 2925 1451 ance matrix of uncertainties was calculated simultaneously with 2925 1451 recommended cross section data by means of PADE-2 code. 2925 1451 Eigenvalues of the 6-th digits relative covariance matrix 2925 1451 given in the 33-file are the following: 2925 1451 2925 1451 9.39333E-10 1.02943E-09 1.18681E-09 1.49652E-09 2925 1451 2.03282E-09 2.85265E-09 3.69906E-09 6.46604E-09 2925 1451 7.96491E-09 7.69999E-08 4.12497E-06 6.99817E-04 2925 1451 1.07280E-03 1.33316E-03 1.89667E-03 4.15078E-03 2925 1451 6.87556E-03 8.81528E-03 2925 1451 2925 1451 References : 2925 1451 1. B.Czapp, H.Vonach Oesterr. Akad. Wiss, Math, Naturw. 2925 1451 Anzeiger, v.97, p.13, January 1960 2925 1451 2. A.Paulsen, H.Liskien Nukleonik, v.10, p.91, July 1967 2925 1451 3. A.Paulsen Zeitschrift f. Phys., v.205, p.226, August 1967 2925 1451 4. R.C.Barrall et al. Report AFWL-TR-68-134, March 1969 2925 1451 6. G.Winkler Nucl. Sci. Eng., v.67, n.2, p.260, August 1978 2925 1451 8. G.Winkler, D.L.Smith, J.W.Meadows Nucl. Sci. Eng., v.76, 2925 1451 p.30, October 1980 2925 1451 9. U.Garuska et al. Prog. Report INDC(POL)-11, p.15, July 1980 2925 1451 10. O.I.Artem'ev et al. Atomnaja Energija (Sov.), v.49, n.3, 2925 1451 p.195, September 1980 2925 1451 11. L.R.Greenwood Progress Report ASTM-STP-956, p.743, 1987 2925 1451 12. Lu Hanlin et al. China Journal of Nuclear Phys., v.12. n.4, 2925 1451 p. 373, 1990 2925 1451 13. Wang Yongchang et al. Chinese J. High Energy Phys. and Nucl. 2925 1451 Phys., v.14, p.919, October 1990 2925 1451 14. J.Csikai, C.M.Buczko, R.Pepelnik, H.M.Agrawal Annals of Nucl. 2925 1451 Energy, v.18, n.1, p.1, 1991 2925 1451 15. C.Konno et al. Report JAERI-1329, October 1993 2925 1451 16. J.W.Meadows et al. Annals of Nucl. Energy, v.23, p.877, 1996 2925 1451 17. A.A.Filatenkov et al. VANT, Ser.: Yadernye Konstanty, v.2, 2925 1451 p.8, Moscow, 1996 2925 1451 18. Lu Hanlin et al. Report INDC(CPR)-045, IAEA, October 1998 2925 1451 19. A.A.Filatenkov et al. Report RI-252, St.Petersburg, May 1999 2925 1451 20. R.Lloret Progress Report EANDC(E)-57, p.172, February 1965 2925 1451 21. A.Fabry, J.P.Deworm Progress Report EANDC(E)-66, p.125, 2925 1451 February 1966 2925 1451 22. K.Kobayashi et al. Nucl. Sci. Techn., v.13, p.531, Oct. 1976 2925 1451 23. L.P.Geraldo et al. Radiochimica Acta, v.57, pp.63-67, 1992 2925 1451 24. W.Mannhart Progress Report INDC(Ger)-045, pp.40-43, 1999 2925 1451 25. G.Winkler et al. Nucl. Sci. Eng., v.78, p.415, August 1981 2925 1451 26. W.Mannhart Proc. of Int. Conf. Nuclear Data for Science and 2925 1451 Technology, 6-10 September 1982, Antwerp, Holland, D.Reidel 2925 1451 Publishing Company, p.429 2925 1451 27. J.Kantele, D.Gardner Nucl. Phys., v.35, p.353, 1962 2925 1451 28. M.Bormann et al. Nucl. Phys., v.A186, p.65, May 1972 2925 1451 29. G.N.Maslov, F.Nasyrov, N.F.Pashkin Yadernye Konstanty, v.9, 2925 1451 p.50, Obninsk, 1972 2925 1451 30. K.Kayashima et al. Prog. Rep. NEANDC(J)-61U, p.94, Sep. 1979 2925 1451 31. S.A.Badikov et al. Preprint FEI-1686, Obninsk, 1985 2925 1451 32. W.Mannhart Handbook on Nuclear Activation Cross Sections, 2925 1451 IAEA Technical Report series No.273, p.413, 1987 2925 1451 33. W.Mannhart Validation of Differential Cross Sections with 2925 1451 Integral Data , Report INDC(NDS)-435, pp.59-64, IAEA, Vienna, 2925 1451 September 2002 2925 1451 34. L.W.Weston et al. Evaluated Neutron Data for U-235, ENDF/B-VI 2925 1451 Library, MAT=9228, MF=5, MT=18, eval. April 1989 2925 1451 35. W.Mannhart IAEA-TECDOC-410, p.158, IAEA, Vienna, 1987 2925 1451 ***************************************************************** 2925 1451 ******** End of (N,A)bibliographical component ******* 2925 1451 ***************************************************************** 2925 1451 2925 1451 2925 1451 2925 1451 ************************ C O N T E N T S *********************** 2925 1451 ***************** Program LINEAR (VERSION 2002-1) ***************2925 1451 For All Data Greater than 1.0000E-10 barns in Absolute Value 2925 1451 Data Linearized to Within an Accuracy of .100000000 per-cent 2925 1451 ***************** Program RECENT (VERSION 2002-1) ***************2925 1451 for All Data Greater than 1.0000E-10 barns in Absolute Value 2925 1451 Data Linearized to within an Accuracy of .100000000 per-cent 2925 1451 ***************** Program SIGMA1 (VERSION 2002-1) ***************2925 1451 Data Doppler Broadened to 300.000000 Kelvin 2925 1451 for All Data Greater than 1.0000E-10 barns in Absolute Value 2925 1451 Data Linearized to Within an Accuracy pf .100000000 per-cent 2925 1451 ***************** Program FIXUP (Version 2002-1) ****************2925 1451 Corrected ZA/AWR in All Sections-----------------------------Yes 2925 1451 Corrected Thresholds-----------------------------------------Yes 2925 1451 Extended Cross Sections to 20 MeV----------------------------No 2925 1451 Allow Cross Section Deletion---------------------------------No 2925 1451 Allow Cross Section Reconstruction---------------------------No 2925 1451 Make All Cross Sections Non-Negative-------------------------Yes 2925 1451 Delete Energies Not in Ascending Order-----------------------Yes 2925 1451 Deleted Duplicate Points-------------------------------------Yes 2925 1451 Check for Ascending MAT/MF/MT Order--------------------------Yes 2925 1451 Check for Legal MF/MT Numbers--------------------------------Yes 2925 1451 Allow Creation of Missing Sections---------------------------No 2925 1451 Allow Insertion of Energy Points-----------------------------No 2925 1451 Create Uniform Energy Grid-----------------------------------No 2925 1451 Delete Section if Cross Section =0 at All Energies-----------Yes 2925 1451 ***************** Program GROUPIE (VERSION 2002-1) **************2925 1451 Unshielded Group Averages Using 640 Groups 2925 1451 Weighting Spectrum: Flat (Constant) Spectrum 2925 1451 1 451 561 52925 1451 3 16 34 12925 1451 3 102 217 52925 1451 3 107 63 12925 1451 33 16 58 12925 1451 33 102 23 32925 1451 33 107 38 12925 1451 2925 1 0 2925 0 0 2.90630E+4 6.23890E+1 0 0 0 02925 3 16 -1.08540E+7-1.08540E+7 0 0 1 912925 3 16 91 1 2925 3 16 11000000.0 .000429103 11100000.0 .002018629 11200000.0 .0036728762925 3 16 11300000.0 .006750000 11400000.0 .011250000 11500000.0 .0182500002925 3 16 11600000.0 .027750000 11700000.0 .039500000 11800000.0 .0535000002925 3 16 11900000.0 .067948571 12000000.0 .082845714 12100000.0 .0977428572925 3 16 12200000.0 .113333107 12300000.0 .133082000 12400000.0 .1535240002925 3 16 12500000.0 .173966000 12600000.0 .194408000 12700000.0 .2150142502925 3 16 12800000.0 .236606000 12900000.0 .258362000 13000000.0 .2801180002925 3 16 13100000.0 .301874000 13200000.0 .322831375 13300000.0 .3389970002925 3 16 13400000.0 .354364000 13500000.0 .369731000 13600000.0 .3850980002925 3 16 13700000.0 .400465000 13800000.0 .415832000 13900000.0 .4311990002925 3 16 14000000.0 .446566000 14100000.0 .461933000 14200000.0 .4772576252925 3 16 14300000.0 .492328000 14400000.0 .507356000 14500000.0 .5223840002925 3 16 14600000.0 .537412000 14700000.0 .551971000 14800000.0 .5637160002925 3 16 14900000.0 .574992000 15000000.0 .586268000 15100000.0 .5975440002925 3 16 15200000.0 .608696250 15300000.0 .619106000 15400000.0 .6293920002925 3 16 15500000.0 .639678000 15600000.0 .649964000 15700000.0 .6598785002925 3 16 15800000.0 .667564000 15900000.0 .674878000 16000000.0 .6821920002925 3 16 16100000.0 .689506000 16200000.0 .696376250 16300000.0 .7005840002925 3 16 16400000.0 .704348000 16500000.0 .708112000 16600000.0 .7118760002925 3 16 16700000.0 .715640000 16800000.0 .719404000 16900000.0 .7231680002925 3 16 17000000.0 .726932000 17100000.0 .730696000 17200000.0 .7344600002925 3 16 17300000.0 .738224000 17400000.0 .741988000 17500000.0 .7456035002925 3 16 17600000.0 .749070500 17700000.0 .752537500 17800000.0 .7560045002925 3 16 17900000.0 .759471500 18000000.0 .762938500 18100000.0 .7664055002925 3 16 18200000.0 .769872500 18300000.0 .773339500 18400000.0 .7768065002925 3 16 18500000.0 .780356000 18600000.0 .783988000 18700000.0 .7876200002925 3 16 18800000.0 .791252000 18900000.0 .794884000 19000000.0 .7985160002925 3 16 19100000.0 .802148000 19200000.0 .805780000 19300000.0 .8094120002925 3 16 19400000.0 .813044000 19500000.0 .814860000 19600000.0 .8148600002925 3 16 19700000.0 .814860000 19800000.0 .814860000 19900000.0 .8148600002925 3 16 20000000.0 0.0 2925 3 16 2925 3 0 2.90630E+4 6.23890E+1 0 0 0 02925 3102 7.91602E+6 7.91602E+6 0 0 1 6412925 3102 641 1 2925 3102 .000100000 70.2622035 .000105000 68.6237994 .000110000 67.06729662925 3102 .000115000 65.6330228 .000120000 63.9563318 .000127500 62.09608432925 3102 .000135000 60.3964753 .000142500 58.8345693 .000150000 57.14415432925 3102 .000160000 55.3831827 .000170000 53.7809161 .000180000 52.31015672925 3102 .000190000 50.9417396 .000200000 49.6923602 .000210000 48.51690432925 3102 .000220000 47.4250938 .000230000 46.4106884 .000240000 45.21999192925 3102 .000255000 43.9115387 .000270000 42.9018717 .000280000 41.77792572925 3102 .000300000 40.4072412 .000320000 39.1661345 .000340000 38.02948142925 3102 .000360000 36.9825557 .000380000 36.0268482 .000400000 35.03047112925 3102 .000425000 34.0108822 .000450000 33.0793104 .000475000 32.22522862925 3102 .000500000 31.4220240 .000525000 30.6879457 .000550000 29.99449532925 3102 .000575000 29.3495653 .000600000 28.6885484 .000630000 28.00852932925 3102 .000660000 27.3863025 .000690000 26.7894143 .000720000 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.045000002 3.30811484 .047499999 3.221662622925 3102 .050000001 3.14208839 .052499998 3.06842366 .055000000 2.998664342925 3102 .057500001 2.93480171 .059999999 2.86765333 .063000001 2.800501852925 3102 .066000000 2.73722878 .068999998 2.67830068 .071999997 2.614436882925 3102 .075999998 2.54606707 .079999998 2.48366401 .083999999 2.425014432925 3102 .088000000 2.37046825 .092000000 2.31988364 .096000001 2.271530822925 3102 .100000001 2.22155040 .104999997 2.16862115 .109999999 2.120249322925 3102 .115000002 2.07404477 .119999997 2.02112934 .127499998 1.962654412925 3102 .135000005 1.90844599 .142499998 1.85872498 .150000006 1.805830022925 3102 .159999996 1.75027369 .170000002 1.69942005 .180000007 1.652808602925 3102 .189999998 1.60995002 .200000003 1.56967971 .209999993 1.532852322925 3102 .219999999 1.49810051 .230000004 1.46573550 .239999995 1.428372152925 3102 .254999995 1.38659312 .270000011 1.35462976 .280000001 1.319114872925 3102 .300000012 1.27564619 .319999993 1.23640287 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8.134670-5 1.278940-4 1.424820-4 1.145190-4292533107 1.153480-3 1.034720-3 7.810150-4 4.620260-4 1.832590-4 5.101270-5292533107 1.413080-5 2.017000-5 8.668080-5 1.691200-4 2.283950-4 2.472330-4292533107 1.019520-3 8.586540-4 5.832950-4 2.813990-4 1.006260-4 2.759680-5292533107 -6.493400-6 4.651990-5 1.727150-4 3.297290-4 4.943410-4 8.217660-4292533107 6.464730-4 3.759330-4 1.718880-4 6.780650-5-1.364010-5 9.720900-7292533107 1.407710-4 3.698800-4 6.635980-4 5.910990-4 4.124830-4 2.402880-4292533107 1.360380-4 3.274150-5 2.902680-6 1.040010-4 3.151620-4 6.171420-4292533107 3.612330-4 2.787220-4 2.169140-4 1.413740-4 9.211770-5 1.194900-4292533107 2.202080-4 3.862950-4 2.811990-4 2.707110-4 2.463720-4 2.133570-4292533107 1.939780-4 1.950960-4 2.180510-4 2.960220-4 3.144670-4 3.107020-4292533107 2.817370-4 2.377330-4 1.849270-4 3.947940-4 4.559440-4 4.588000-4292533107 4.126890-4 3.275050-4 6.282220-4 7.551020-4 8.366480-4 8.809310-4292533107 1.092850-3 1.456080-3 1.843700-3 2.235280-3 3.158140-3 4.788580-3292533107 292533 0 2925 0 0 0 0 0 2.90650E+4 6.43700E+1 0 0 34 102931 1451 0.0 0.0 0 0 0 62931 1451 1.00000E+0 1.50000E+8 0 0 10 20022931 1451 3.00000E+2 0.0 1 0 384 32931 1451 29-Cu- 65 LANL,ORNL EVAL-FEB98 A.KONING,M.CHADWICK,HETRICK 2931 1451 DIST-Feb2002 2931 1451 ----IRDF-2002 MATERIAL 2931 2931 1451 -----INCIDENT NEUTRON DATA 2931 1451 ------ENDF-6 FORMAT 2931 1451 ******************************************************************2931 1451 29-Cu- 65 LANL,ORNL EVAL-FEB98 A.KONING,M.CHADWICK,HETRICK 2931 1451 Ch98,Ch99 DIST-SEP 1 REV4- 20010926 2931 1451 ----ENDF/B-VI MATERIAL 2931 REVISION 4 2931 1451 *****************EXTRACT FOR SPECIAL PURPOSE FILE*****************2931 1451 DOSIMETRY 2931 1451 ******************************************************************2931 1451 2931 1451 **************************************************************** 2931 1451 2931 1451 ENDF/B-VI MOD 5 Revision, May 2000, S.C. Frankle, R.C. Reedy, 2931 1451 P.G. Young (LANL) 2931 1451 2931 1451 The secondary gamma-ray spectrum for radiative capture (MF 12, 2931 1451 MT 102) has been updated for new experimental data at incident 2931 1451 neutron energies up to 1 keV. 2931 1451 The MF=12, MT=102 yields above 1 keV were adjusted slightly to 2931 1451 force energy conservation. 2931 1451 The Q-value for radiative capture was also updated in File 3. 2931 1451 Details of these changes are described in Frankel et al. [Fr01]. 2931 1451 2931 1451 **************************************************************** 2931 1451 2931 1451 ENDF/B-VI MOD 4 Evaluation, February 1998, A.J. Koning (ECN), 2931 1451 M.B. Chadwick, P.G. Young (LANL) 2931 1451 2931 1451 Los Alamos LA150 Library, produced with FKK/GNASH/GSCAN code 2931 1451 in cooperation with ECN Petten. 2931 1451 2931 1451 This evaluation provides a complete representation of the 2931 1451 nuclear data needed for transport, damage, heating, 2931 1451 radioactivity, and shielding applications over the incident 2931 1451 neutron energy range from 1.0E-11 to 150 MeV. The discussion 2931 1451 here is divided into the region below and above 20 MeV. 2931 1451 2931 1451 INCIDENT NEUTRON ENERGIES < 20 MeV 2931 1451 2931 1451 Below 20 MeV the evaluation is based completely on the ENDF/B- 2931 1451 VI (Release 2) evaluation by D. Hetrick, C.Y. Fu, and D. Larson. 2931 1451 2931 1451 INCIDENT NEUTRON ENERGIES > 20 MeV 2931 1451 2931 1451 The ENDF/B-VI Release 2 evaluation extends to 20 MeV and 2931 1451 includes cross sections and energy-angle data for all 2931 1451 significant reactions. The present evaluation utilizes a more 2931 1451 compact composite reaction spectrum representation above 20 MeV 2931 1451 in order to reduce the length of the file. No essential data for 2931 1451 applications is lost with this representation. 2931 1451 The evaluation above 20 MeV utilizes MF=6, MT=5 to represent 2931 1451 all reaction data. Production cross sections and emission 2931 1451 spectra are given for neutrons, protons, deuterons, tritons, 2931 1451 alpha particles, gamma rays, and all residual nuclides produced 2931 1451 (A>5) in the reaction chains. To summarize, the ENDF sections 2931 1451 with non-zero data above En = 20 MeV are: 2931 1451 2931 1451 MF=3 MT= 1 Total Cross Section 2931 1451 MT= 2 Elastic Scattering Cross Section 2931 1451 MT= 3 Nonelastic Cross Section 2931 1451 MT= 5 Sum of Binary (n,n') and (n,x) Reactions 2931 1451 2931 1451 MF=4 MT= 2 Elastic Angular Distributions 2931 1451 2931 1451 MF=6 MT= 5 Production Cross Sections and Energy-Angle 2931 1451 Distributions for Emission Neutrons, Protons, 2931 1451 Deuterons, Tritons, and Alphas; and Angle- 2931 1451 Integrated Spectra for Gamma Rays and Residual 2931 1451 2931 1451 Nuclei That Are Stable Against Particle Emission 2931 1451 2931 1451 The evaluation is based on nuclear model calculations that 2931 1451 have been benchmarked to experimental data, especially for n + 2931 1451 Cu65 and p + Cu65 reactions (Ch98). We use the GNASH code system 2931 1451 (Yo92), which utilizes Hauser-Feshbach statistical, preequilib- 2931 1451 rium and direct-reaction theories. Spherical optical model 2931 1451 calculations are used to obtain particle transmission 2931 1451 coefficients for the Hauser-Feshbach calculations, as well as 2931 1451 for the elastic neutron angular distributions. 2931 1451 Cross sections and spectra for producing individual residual 2931 1451 nuclei are included for reactions. The energy-angle-correlations 2931 1451 for all outgoing particles are based on Kalbach systematics 2931 1451 (Ka88). 2931 1451 A model was developed to calculate the energy distributions of 2931 1451 all recoil nuclei in the GNASH calculations (Ch96a). The recoil 2931 1451 energy distributions are represented in the laboratory system in 2931 1451 MT=5, MF=6, and are given as isotropic in the lab system. All 2931 1451 other data in MT=5,MF=6 are given in the center-of-mass system. 2931 1451 This method of representation utilizes the LCT=3 option approved 2931 1451 at the November, 1996, CSEWG meeting. 2931 1451 Preequilibrium corrections were performed in the course of the 2931 1451 GNASH calculations using the exciton model of Kalbach (Ka77, 2931 1451 Ka85), validated by comparison with calculations using Feshbach, 2931 1451 Kerman, Koonin (FKK) theory [Ch93]. Discrete level data from 2931 1451 nuclear data sheets were matched to continuum level densities 2931 1451 using the formulation of Ignatyuk et al. (Ig75) and pairing and 2931 1451 shell parameters from the Cook (Co67) analysis. Neutron and 2931 1451 charged- particle transmission coefficients were obtained from 2931 1451 the optical potentials, as discussed below. Gamma-ray 2931 1451 transmission coefficients were calculated using the Kopecky-Uhl 2931 1451 model (Ko90). 2931 1451 2931 1451 SPECIFIC INFORMATION CONCERNING THE 65Cu EVALUATION 2931 1451 2931 1451 This evaluation is documented in some detail in Ref. (Ko98b). 2931 1451 2931 1451 The neutron total cross section above 20 MeV was obtained by 2931 1451 evaluating experimental data, with a particular emphasis on the 2931 1451 Finlay (Fi93) elemental data. This resulted in an evaluated 2931 1451 elemental Cu total cross section; to obtain an isotopic 65Cu total2931 1451 cross section, it was assumed that 63Cu and 65Cu have total cross 2931 1451 sections in an A**2/3 ratio to one another. The total neutron 2931 1451 nonelastic cross section was obtained directly from an optical 2931 1451 model calculation (see below), after verifying that it was in good2931 1451 agreement with the experimental data (Ko98b). 2931 1451 2931 1451 To obtain the neutron optical potential we used total cross 2931 1451 section data from 1.2 to 4.5 MeV (Gu86) and from 5.3 to 600 MeV 2931 1451 (Fi93), and elastic scattering angular distribution data from 1.6 2931 1451 to 96 MeV (Br50, Sa60, Ki74, El82, Gu86). The optical potential 2931 1451 parameters were obtained using a combination of a grid search code2931 1451 and the interactive optical model viewer ECISVIEW [Ko97], both 2931 1451 built around the coupled channels code ECIS96 [Ra94]. The energy 2931 1451 dependence of the optical model parameters is as described in 2931 1451 [Ko98]. This optical potential was used for the calculation, with 2931 1451 ECIS96, of neutron transmission coefficients and DWBA cross 2931 1451 sections for the entire energy region above 20 MeV. 2931 1451 2931 1451 Due to the lack of proton elastic scattering data in numerical 2931 1451 form, we used a combination of global optical models for the 2931 1451 proton channel. The Becchetti-Greenlees potential [Be69]was 2931 1451 adopted below 47 MeV, and the non-relativistic version of the 2931 1451 Madland potential [Ma88] above 47 MeV. At this particular energy 2931 1451 point the two potentials join smoothly. 2931 1451 2931 1451 For deuterons, the Lohr-Haeberli global potential [Lo74] was used;2931 1451 for alpha particles the Moyen potential (MacFadden-Satchler 2931 1451 [Ma66]) was used; and for tritons the Becchetti-Greenlees 2931 1451 potential [Be71] was used. The He-3 channel was ignored, due to 2931 1451 its small importance. 2931 1451 2931 1451 Following Delaroche et al. [De82], we adopted the weak-coupling 2931 1451 model for direct collective inelastic scattering for Cu-65, using 2931 1451 Ni-64 as a basis. For the calculation of the cross sections, 2931 1451 ECIS96 was used in DWBA mode. We used the following direct 2931 1451 transitions for Cu-65 (ground state 3/2- ) : 2931 1451 2931 1451 Jpi Ex(MeV) Deformation length (Delta) or parameter (Beta) 2931 1451 0.5- 0.771 Beta(2)=0.0566 2931 1451 2.5- 1.116 Beta(2)=0.0980 2931 1451 3.5- 1.481 Beta(2)=0.1132 2931 1451 1.5- 1.743 Beta(2)=0.0800 2931 1451 1.5- 3.185 Delta(3)=0.3167 2931 1451 2.5- 3.435 Delta(3)=0.3879 2931 1451 3.5- 3.685 Delta(3)=0.4479 2931 1451 4.5- 3.935 Delta(3)=0.5008 2931 1451 2931 1451 Only one measurement exists for neutron-induced emission spectra 2931 1451 above 20 MeV for 65Cu: the 25.7 MeV (n,xn) data by Marcinkowski et2931 1451 al (Ma83). Without adjusting any of the level density or pre- 2931 1451 equilibrium parameters the GNASH calculation was in good agreement2931 1451 with these data (Ko98b). Hence we adopted these parameters for the2931 1451 whole energy region. 2931 1451 2931 1451 **************************************************************** 2931 1451 2931 1451 REFERENCES 2931 1451 2931 1451 [Ab93] W. Abfalterer, R.W. Finlay, S.M. Grimes, and V. Mishra, 2931 1451 Phys.Rev. C47, 1033 (1993) 2931 1451 [Al83] R. Alarcon and J. Rapaport, Nucl.Phys. A458, 502 (1986) 2931 1451 [Ar80] E.D. Arthur and P.G. Young, 'Evaluation of Neutron Cross 2931 1451 Sections to 40 MeV for 54,56Fe," Proc. Sym. on Neutron Cross 2931 1451 Sections from 10 to 50 MeV, 12-14 May 1980, Brookhaven National2931 1451 Laboratory [Eds. M. R. Bhat and S. Pearlstein, BNL-NCS- 51245, 2931 1451 1980] p. 731. 2931 1451 [Be69] F.D. Becchetti, Jr., and G.W. Greenlees, Phys.Rev. 182, 2931 1451 1190 (1969) 2931 1451 [Be71] F.D. Becchetti, Jr., and G.W. Greenlees in "Polarization 2931 1451 Phenomena in Nuclear Reactions," (Ed: H.H. Barschall and W. 2931 1451 Haeberli, The University of Wisconsin Press, 1971) p.682. 2931 1451 [Be92] O. Bersillon, "SCAT2 - A Spherical Optical Model Code," 2931 1451 in Proc. ICTP Workshop on Computation and Analysis of Nuclear 2931 1451 data Relevant to Nuclear Energy and Safety, February-March, 2931 1451 1999 Trieste, Italy, to be published in World Scientific Press,2931 1451 and Progress Report of the Nuclear Physics Division, Bruyeres- 2931 1451 le-Chatel 1977, CEA-N-2037 (1978) p.111 2931 1451 [Br50] S. Bratenahl, S. Fernbach, R.H. Hildebrand et al., 2931 1451 Phys.Rev. 77, 597 (1950) 2931 1451 [Ch93] M.B. Chadwick and P.G. Young, Phys.Rev. C 47, 2255 (1993) 2931 1451 [Ch96] M.B. Chadwick, P.G. Young, R.E. MacFarlane, and A.J. 2931 1451 Koning, "High-Energy Nuclear Data Libraries for Accelerator- 2931 1451 Driven Technologies: Calculational Method for Heavy Recoils," 2931 1451 Proc. of 2nd Int. Conf. on Accelerator Driven Transmutation 2931 1451 Technology and Applications, Kalmar, Sweden, 3-7 June 1996 2931 1451 [Ch98] M.B. Chadwick and P.G. Young, "GNASH Calculations of 2931 1451 n,p + Cu isotopes and Benchmarking of Results" in APT PROGRESS 2931 1451 REPORT: 1 February - 1 March 1998, internal Los Alamos National2931 1451 Laboratory memo, 6 Mar.1998 from R.E. MacFarlane to L. Waters. 2931 1451 [Ch99] M.B. Chadwick, P G. Young, G. M. Hale, et al., Los Alamos 2931 1451 National Laboratory report, LA-UR-99-1222 (1999) 2931 1451 [Co67] J.L. Cook, H. Ferguson, and A.R. DeL Musgrove, Aust.J. 2931 1451 Phys. 20, 477 (1967) 2931 1451 [De82] J.P. Delaroche, S.M. El-Kadi, P.P. Guss, C.E. Floyd and 2931 1451 R.L. Walter, Nucl. Phys. A390, 541 (1982). 2931 1451 [El82] S.M. El-Kadi, C.E. Nelson, F.O. Purser et al., Nucl.Phys. 2931 1451 A390, 509 (1982) 2931 1451 [Fi93] R. W. Finlay, W. P. Abfalterer, G. Fink et al., Phys. Rev 2931 1451 C 47, 237 (1993) 2931 1451 [Fr01] S.C. Frankle, R.C. Reedy, and P.G. Young, Los ALamos 2931 1451 National Laboratory Report, LA-13812 (2001). 2931 1451 [Gu86] P. Guenther, D.L. Smith, A.B. Smith, J.F. Whalen, Nucl. 2931 1451 Phys. A448, 280 (1986) 2931 1451 [Ig75] A.V. Ignatyuk, G.N. Smirenkin, and A.S. Tishin, Sov.J. 2931 1451 Nucl.Phys. 21, 255 (1975); translation of Yad.Fiz. 21, 485 2931 1451 (1975) 2931 1451 [Ka77] C. Kalbach, Z.Phys.A 283, 401 (1977) 2931 1451 [Ka85] C. Kalbach, Los Alamos National Laboratory report 2931 1451 LA-10248-MS (1985) 2931 1451 [Ka88] C. Kalbach, Phys.Rev.C 37, 2350 (1988); see also 2931 1451 C. Kalbach and F. M. Mann, Phys.Rev.C 23, 112 (1981) 2931 1451 [Ki74] W.E. Kinney, F.G. Perey, report ORNL-4908 (1974) 2931 1451 [Ko90] J. Kopecky and M. Uhl, Phys.Rev.C 41, 1941 (1990) 2931 1451 [Ko97] A.J. Koning, J.J. van Wijk and J.-P. Delaroche, "ECISVIEW:2931 1451 A Graphical Interface for ECIS95", Proceedings of the NEA 2931 1451 Specialists' Meeting on the Nucleon Nucleus Optical Model up to2931 1451 200 MeV, Bruyeres-le-Chatel, November 13-15 1996. Available at 2931 1451 http://db.nea.fr/html/science/om200/. 2931 1451 [Ko98] A.J. Koning, J.-P. Delaroche and O. Bersillon, "Nuclear 2931 1451 Data for Accelerator-Driven Systems: Nuclear models, Experiments2931 1451 and Data Libraries", to appear in Mucl. Instr. Meth. A (1998). 2931 1451 [Ko98b] A.J. Koning, M.B. Chadwick, and P.G. Young, "ENDF/B-VI 2931 1451 neutron and proton datafiles up to 150 MeV for 63Cu and 65Cu", 2931 1451 Los Alamos National Laboratory report LAUR- (1998); ECN lab and2931 1451 JEFF report (1998). 2931 1451 [Lo74] J.M. Lohr and W. Haeberli, Nucl.Phys. A232, 381 (1974) 2931 1451 [Ma66] Macfadden and Satchler, Nuc.Phys. 84, 177 (1966) 2931 1451 [Ma83] A. Marcinkowski, R.W. Finlay, G. Randers-Pehrson et al., 2931 1451 Nucl.Phys. A402, 220 (1983) 2931 1451 [Ma88] D.G. Madland, "Recent Results in the Development of a 2931 1451 Global Medium-Energy Nucleon-Nucleus Optical-Model Potential, 2931 1451 "Proc. OECD/NEANDC Specialist's Mtg. on Preequilibrium Nuclear 2931 1451 Reactions, Semmering, Austria, 10-12 Feb. 1988, NEANDC-245 'U' 2931 1451 (1988). 2931 1451 [Pe63] C.M. Perey and F.G. Perey, Phys.Rev. 132, 755 (1963) 2931 1451 [Ra94] J. Raynal, Notes on ECIS94, CEA Saclay Report CEA-N-2772 2931 1451 (1994) 2931 1451 [Sa60] G.L. Salmon, Nucl.Phys. 21, 15 (1960) 2931 1451 [Yo92] P.G. Young, E.D. Arthur, and M.B. Chadwick, report 2931 1451 LA-12343-MS (1992) 2931 1451 2931 1451 **************************************************************** 2931 1451 2931 1451 ENDF/B-VI MOD 3 Revision, March 1991, ORNL 2931 1451 2931 1451 MOD 3 changes 2931 1451 1) Corrections to MF=6, MT=63 at 17.5 MeV to prevent negative 2931 1451 values in the angular distribution. 2931 1451 2931 1451 **************************************************************** 2931 1451 2931 1451 * Note there was no MOD 2 released. 2931 1451 2931 1451 **************************************************************** 2931 1451 2931 1451 ENDF/B-VI MOD 1 Evaluation, November 1989, D. Hetrick, F.Y. Fu, 2931 1451 D. Larson (ORNL) 2931 1451 2931 1451 This work employed several nuclear model codes including the 2931 1451 optical-model code GENOA [1], the Distorted Wave Born 2931 1451 Approximation (DWBA) program DWUCK [2], and the Hauser-Feshbach 2931 1451 code TNG [3,4]. The TNG code provides energy and angular 2931 1451 distributions of particles emitted in the compound and pre- 2931 1451 compound reactions, ensures consistency among all reactions, and 2931 1451 maintains energy balance. Details pertinent to the contents of 2931 1451 this evaluation and extensive comparisons of calculations with 2931 1451 experimental data can be found in reference [5]. 2931 1451 2931 1451 ----- DESCRIPTION OF FILES 2931 1451 (MF-MT) 2931 1451 1-451 GENERAL INFORMATION, REFERENCES, AND DEFINITIONS. 2931 1451 2-151 RESONANCE PARAMETERS WERE TAKEN FROM MUGHABGHAB [6]. POINT2931 1451 WISE RECONSTRUCTION COMPARED WITH DATA [7] SHOWED POORER 2931 1451 FIT ABOVE 100 KEV, SO THE RESONANCE REGION WAS CUT OFF AT 2931 1451 99.5 KEV. REICH-MOORE PARAMETERS ARE GIVEN. AGREEMENT 2931 1451 WITH DATA COULD BE IMPROVED WITH ADDITION OF A BACKGROUND 2931 1451 FILE IN 3/1, BUT THIS IN GENERAL GIVES TOO LARGE AN 2931 1451 AVERAGE CROSS SECTION, WHEN BINNED IN 10 KEV BINS AND 2931 1451 COMPARED WITH THE BINNED DATA. THIS IS PROBABLY DUE TO 2931 1451 TOO LARGE AN ESTIMATE OF NEUTRON WIDTHS FOR RESONANCES 2931 1451 SEEN ONLY IN CAPTURE AND NOT IN TRANSMISSION. 2931 1451 NOTE THAT THE FLAG HAS BEEN SET TO ALLOW USER CALCULATION 2931 1451 OF THE ANGULAR DISTRIBUTIONS FROM THE R-M RESONANCE 2931 1451 PARAMETERS, IF THE USER WANTS ANGULAR DISTRIBUTIONS ON 2931 1451 A FINER ENERGY GRID THAN GIVEN IN 4/2. 2931 1451 3-16 (N,2N) CROSS SECTIONS WERE TAKEN FROM THE GLUCS [12] 2931 1451 CALCULATION IN WHICH THIS REACTION WAS STUDIED SIMUL- 2931 1451 TANEOUSLY WITH 12 OTHER DOSIMETRY REACTION CROSS 2931 1451 SECTIONS [13]. 2931 1451 33-16 (N,2N) COVARIANCES WERE TAKEN FROM THE GLUCS [12] 2931 1451 CALCULATION IN WHICH THIS REACTION WAS STUDIED SIMUL- 2931 1451 TANEOUSLY WITH 12 OTHER DOSIMETRY REACTION CROSS 2931 1451 SECTIONS [13]. 2931 1451 **************************************************************** 2931 1451 2931 1451 REFERENCES: 2931 1451 2931 1451 [1] F.G. Perey, computer code GENOA, ORNL, unpublished (1967) 2931 1451 [2] P.D. Kunz, "Distorted Wave Code DWUCK72," Univ. of 2931 1451 Colorado, unpublished (1972) 2931 1451 [3] C.Y. Fu, report ORNL/TM-7042 (1980); also, C.Y Fu, 2931 1451 Symp. on Neutron Cross Sections from 10 to 50 MeV, Upton, NY,2931 1451 May 1980, Brookhaven National Lab. report BNL-NCS-51245 2931 1451 (1980) p.675 2931 1451 [4] K. Shibata and C.Y. Fu, report ORNL/TM-10093 (1986) 2931 1451 [5] D.M. Hetrick, C.Y. Fu, and D.C. Larson, Oak Ridge report 2931 1451 ORNL/TM-9083 [ENDF-337] (1984) 2931 1451 [6] S.F. Mughabghab, M. Divadeenam, and N.E. Holden, "Neutron 2931 1451 Cross Sections, Vol. 1, Neutron Resonance Parameters and 2931 1451 Thermal Cross Sections, Part A, Z=1-60," (Academic Press, 2931 1451 1981) 2931 1451 [7] M.S. Pandey, J.B. Garg and J.A. Harvey, Phys.Rev. C 15, 600 2931 1451 (1977), and private communication. 2931 1451 [8] F.G. Perey, private communication (1977) 2931 1451 [9] D.C. Larson, Symp. on Neutron Cross Sections from 10 to 50 2931 1451 MeV, Upton, NY, May 1980, Brookhaven National Lab. report 2931 1451 BNL-NCS-51245 (1980) p.277 2931 1451 [10] P. Guenther, D.L. Smith, A.B. Smith and J.F. Whalen, Nucl. 2931 1451 Phys. A, 448, 280 (1986) [CSISRS data set 12869/002], and 2931 1451 W.P. Poenitz and J.F. Whalen, Argonne report ANL/NDM-80 2931 1451 (1983) [CSISRS data set 12853] 2931 1451 [11] A.I. Dyumin, D.M. Kaminker, G.N. Popova, and V.A. Smolin, 2931 1451 Izv.Akad.Nauk SSSR, Ser.Fiz. 36, 852 (1972) 2931 1451 [12] D.M. Hetrick and C.Y. Fu, Oak Ridge report ORNL/TM-7341 2931 1451 [ENDF-303] (1980) 2931 1451 [13] C.Y. Fu and D.M. Hetrick, Proc. Fourth ASTM-Euratom Symp. 2931 1451 on Reactor Dosimetry, Gaithersburg, Maryland, March 22-26, 2931 1451 1982 (U.S. National Bureau of Standards) p.877 2931 1451 [14] CSISRS Library, National Nuclear Data Center, Brookhaven 2931 1451 National Laboratory, Upton, N.Y. 11973. 2931 1451 [15] V. McLane, C.L. Dunford and P.F. Rose, "Neutron Cross 2931 1451 Sections, Vol. 2, Neutron Cross Section Curves" (Academic 2931 1451 Press, 1988) 2931 1451 [16] S.M. Qaim and G. Stoecklin, Nucl.Phys. A257, 233 (1976) 2931 1451 [17] S.M. Qaim, Radiochimica Acta, 25, 13 (1978) 2931 1451 [18] M.G. Delfini, J. Kopecky, R.E. Chrien et al., Nucl.Phys. 2931 1451 A404, 250 (1983) 2931 1451 **************************************************************** 2931 1451 The Q values and threshold energies were updated prior to pro- 2931 1451 cessing through the codes to comply with the values obtained 2931 1451 using the NNDC calculation program which is based on the 1995 2931 1451 Update to the Atomic mass Evaluation. 2931 1451 ************************ C O N T E N T S *********************** 2931 1451 ***************** Program LINEAR (VERSION 2002-1) ***************2931 1451 For All Data Greater than 1.0000E-10 barns in Absolute Value 2931 1451 Data Linearized to Within an Accuracy of .100000000 per-cent 2931 1451 ***************** Program RECENT (VERSION 2002-1) ***************2931 1451 for All Data Greater than 1.0000E-10 barns in Absolute Value 2931 1451 Data Linearized to within an Accuracy of .100000000 per-cent 2931 1451 ***************** Program SIGMA1 (VERSION 2002-1) ***************2931 1451 Data Doppler Broadened to 300.000000 Kelvin 2931 1451 for All Data Greater than 1.0000E-10 barns in Absolute Value 2931 1451 Data Linearized to Within an Accuracy pf .100000000 per-cent 2931 1451 ***************** Program FIXUP (Version 2002-1) ****************2931 1451 Corrected ZA/AWR in All Sections-----------------------------Yes 2931 1451 Corrected Thresholds-----------------------------------------Yes 2931 1451 Extended Cross Sections to 20 MeV----------------------------No 2931 1451 Allow Cross Section Deletion---------------------------------No 2931 1451 Allow Cross Section Reconstruction---------------------------No 2931 1451 Make All Cross Sections Non-Negative-------------------------Yes 2931 1451 Delete Energies Not in Ascending Order-----------------------Yes 2931 1451 Deleted Duplicate Points-------------------------------------Yes 2931 1451 Check for Ascending MAT/MF/MT Order--------------------------Yes 2931 1451 Check for Legal MF/MT Numbers--------------------------------Yes 2931 1451 Allow Creation of Missing Sections---------------------------No 2931 1451 Allow Insertion of Energy Points-----------------------------No 2931 1451 Create Uniform Energy Grid-----------------------------------No 2931 1451 Delete Section if Cross Section =0 at All Energies-----------Yes 2931 1451 ***************** Program GROUPIE (VERSION 2002-1) **************2931 1451 Unshielded Group Averages Using 640 Groups 2931 1451 Weighting Spectrum: Flat (Constant) Spectrum 2931 1451 1 451 391 52931 1451 3 16 37 12931 1451 33 16 62 12931 1451 2931 1 0 2931 0 0 2.90650E+4 6.43700E+1 0 0 0 02931 3 16 -9.91020E+6-9.91020E+6 0 0 1 1012931 3 16 101 1 2931 3 16 10000000.0 .000366266 10100000.0 .004894626 10200000.0 .0105964472931 3 16 10300000.0 .016298268 10400000.0 .022000089 10500000.0 .0367329002931 3 16 10600000.0 .060496700 10700000.0 .084260500 10800000.0 .1080243002931 3 16 10900000.0 .131788100 11000000.0 .159085000 11100000.0 .1899150002931 3 16 11200000.0 .220745000 11300000.0 .251575000 11400000.0 .2824050002931 3 16 11500000.0 .314504000 11600000.0 .347872000 11700000.0 .3812400002931 3 16 11800000.0 .414608000 11900000.0 .447976000 12000000.0 .4801010002931 3 16 12100000.0 .510983000 12200000.0 .541865000 12300000.0 .5727470002931 3 16 12400000.0 .603629000 12500000.0 .630458000 12600000.0 .6532340002931 3 16 12700000.0 .676010000 12800000.0 .698786000 12900000.0 .7215620002931 3 16 13000000.0 .741987000 13100000.0 .760061000 13200000.0 .7781350002931 3 16 13300000.0 .796209000 13400000.0 .814283000 13500000.0 .8305190002931 3 16 13600000.0 .844917000 13700000.0 .859315000 13800000.0 .8737130002931 3 16 13900000.0 .888111000 14000000.0 .900663000 14100000.0 .9113690002931 3 16 14200000.0 .922075000 14300000.0 .932781000 14400000.0 .9434870002931 3 16 14500000.0 .952211000 14600000.0 .958953000 14700000.0 .9656950002931 3 16 14800000.0 .972437000 14900000.0 .979179000 15000000.0 .9856050002931 3 16 15100000.0 .991715000 15200000.0 .997825000 15300000.0 1.003935002931 3 16 15400000.0 1.01004500 15500000.0 1.01484000 15600000.0 1.018320002931 3 16 15700000.0 1.02180000 15800000.0 1.02528000 15900000.0 1.028760002931 3 16 16000000.0 1.03373000 16100000.0 1.04019000 16200000.0 1.046650002931 3 16 16300000.0 1.05311000 16400000.0 1.05957000 16500000.0 1.063630002931 3 16 16600000.0 1.06529000 16700000.0 1.06695000 16800000.0 1.068610002931 3 16 16900000.0 1.07027000 17000000.0 1.07157000 17100000.0 1.072510002931 3 16 17200000.0 1.07345000 17300000.0 1.07439000 17400000.0 1.075330002931 3 16 17500000.0 1.07734000 17600000.0 1.08042000 17700000.0 1.083500002931 3 16 17800000.0 1.08658000 17900000.0 1.08966000 18000000.0 1.089750002931 3 16 18100000.0 1.08685000 18200000.0 1.08395000 18300000.0 1.081050002931 3 16 18400000.0 1.07815000 18500000.0 1.07277167 18600000.0 1.064915002931 3 16 18700000.0 1.05705833 18800000.0 1.04920167 18900000.0 1.041345002931 3 16 19000000.0 1.03348833 19100000.0 1.02563167 19200000.0 1.017775002931 3 16 19300000.0 1.00991833 19400000.0 1.00206167 19500000.0 .9942050002931 3 16 19600000.0 .986348333 19700000.0 .978491667 19800000.0 .9706350002931 3 16 19900000.0 .962778333 20000000.0 0.0 2931 3 16 2931 3 0 2931 0 0 2.90650E+4 6.43700E+1 0 0 0 1293133 16 0.0000E+00 0.0000E+00 0 16 0 2293133 16 0.0000E+00 0.0000E+00 1 5 300 24293133 16 1.0000E-05 1.0000E+07 1.0500E+07 1.1000E+07 1.1500E+07 1.2000E+07293133 16 1.2500E+07 1.3000E+07 1.3500E+07 1.4000E+07 1.4500E+07 1.5000E+07293133 16 1.5500E+07 1.6000E+07 1.6500E+07 1.7000E+07 1.7500E+07 1.8000E+07293133 16 1.8500E+07 1.9000E+07 1.9200E+07 1.9400E+07 1.9600E+07 2.0000E+07293133 16 0.0000E+00 0.0000E+00 0.0000E+00 0.0000E+00 0.0000E+00 0.0000E+00293133 16 0.0000E+00 0.0000E+00 0.0000E+00 0.0000E+00 0.0000E+00 0.0000E+00293133 16 0.0000E+00 0.0000E+00 0.0000E+00 0.0000E+00 0.0000E+00 0.0000E+00293133 16 0.0000E+00 0.0000E+00 0.0000E+00 0.0000E+00 0.0000E+00 9.2849E-02293133 16 -3.9100E-03 9.8124E-04 3.2315E-04 2.5973E-04 1.6943E-04 1.0837E-04293133 16 5.2219E-05 4.5076E-05 3.6396E-05 1.4418E-05 2.1146E-05 2.1805E-05293133 16 2.1931E-05 2.1318E-05 2.1522E-05 2.2348E-05 2.9580E-05 2.2389E-05293133 16 2.5489E-05 2.5389E-05 2.8747E-05 1.1582E-03 9.4407E-05 2.2976E-04293133 16 1.6573E-04 1.1914E-04 8.3873E-05 4.3373E-05 3.6631E-05 3.1684E-05293133 16 1.1438E-05 1.8007E-05 1.8412E-05 1.8521E-05 1.7912E-05 1.8306E-05293133 16 1.8801E-05 2.5446E-05 1.8989E-05 2.1853E-05 2.1701E-05 2.4893E-05293133 16 1.0426E-03 4.1291E-04 3.5993E-04 2.4206E-04 1.6476E-04 8.2857E-05293133 16 7.0722E-05 6.0039E-05 2.2260E-05 3.4439E-05 3.5246E-05 3.5284E-05293133 16 3.4112E-05 3.4653E-05 3.5725E-05 4.7927E-05 3.4885E-05 4.0981E-05293133 16 4.0699E-05 4.6995E-05 1.6804E-03 5.1367E-04 3.7077E-04 2.3022E-04293133 16 1.0876E-04 9.5181E-05 7.5902E-05 3.0612E-05 4.4708E-05 4.6020E-05293133 16 4.5981E-05 4.4753E-05 4.5221E-05 4.6593E-05 6.1095E-05 4.5008E-05293133 16 5.2412E-05 5.2045E-05 5.9698E-05 1.5662E-03 4.6015E-04 3.0034E-04293133 16 1.3181E-04 1.1846E-04 8.7882E-05 3.8955E-05 5.3423E-05 5.5298E-05293133 16 5.5174E-05 5.4093E-05 5.4391E-05 5.6078E-05 7.1784E-05 5.4017E-05293133 16 6.1977E-05 6.1527E-05 6.9508E-05 1.6388E-03 3.8739E-04 1.8183E-04293133 16 1.6833E-04 1.0963E-04 5.4807E-05 6.8444E-05 7.1184E-05 7.0989E-05293133 16 6.9696E-05 7.0329E-05 7.2399E-05 9.1321E-05 7.1362E-05 7.9338E-05293133 16 7.8939E-05 8.7688E-05 1.6381E-03 2.7820E-04 2.9569E-04 1.5183E-04293133 16 8.4051E-05 9.2687E-05 9.9014E-05 9.7400E-05 9.5524E-05 9.6112E-05293133 16 9.8898E-05 1.2500E-04 9.5565E-05 1.0835E-04 1.0739E-04 1.2076E-04293133 16 7.2642E-04 1.3861E-04 2.1090E-04 1.3154E-04 1.2590E-04 1.3899E-04293133 16 1.3473E-04 1.3248E-04 1.3243E-04 1.3654E-04 1.7246E-04 1.2855E-04293133 16 1.4860E-04 1.4753E-04 1.6696E-04 7.4020E-04 2.2844E-04 1.3272E-04293133 16 1.3480E-04 1.4740E-04 1.4311E-04 1.4019E-04 1.4023E-04 1.4475E-04293133 16 1.8498E-04 1.3536E-04 1.5870E-04 1.5766E-04 1.7994E-04 2.5205E-04293133 16 6.7831E-05 1.5774E-04 1.5403E-04 1.5480E-04 1.4692E-04 1.4845E-04293133 16 1.5385E-04 2.1458E-04 1.4114E-04 1.7986E-04 1.7866E-04 2.1467E-04293133 16 4.9848E-04 7.7213E-05 1.8584E-04 1.5113E-04 1.7339E-04 1.6481E-04293133 16 1.6703E-04 1.1799E-04 1.6769E-04 1.2330E-04 1.2252E-04 8.4028E-05293133 16 3.4942E-04 1.7432E-04 1.9975E-04 2.0482E-04 1.8271E-04 1.7572E-04293133 16 1.6743E-04 1.7022E-04 1.5696E-04 1.5582E-04 1.4329E-04 5.2876E-04293133 16 1.9487E-04 2.1525E-04 1.9401E-04 1.9702E-04 1.6642E-04 1.6184E-04293133 16 1.5555E-04 1.5458E-04 1.4089E-04 2.9887E-04 1.8750E-04 2.0981E-04293133 16 1.8656E-04 1.7332E-04 1.7942E-04 1.6365E-04 1.6255E-04 1.4448E-04293133 16 3.3467E-04 2.2613E-04 2.0243E-04 1.7545E-04 1.8875E-04 1.6806E-04293133 16 1.6691E-04 1.4225E-04 1.1229E-03-7.4862E-05 3.1309E-04 2.2013E-04293133 16 2.7176E-04 2.6985E-04 1.8472E-04 4.9986E-04 1.6697E-04 2.0915E-04293133 16 1.7615E-04 1.7495E-04 1.4909E-04 1.3873E-03 1.1714E-04 5.8852E-04293133 16 5.8437E-04 3.4575E-04 4.7121E-04 2.5339E-04 2.5176E-04 1.7365E-04293133 16 3.0426E-03 1.1867E-03 9.8857E-04 3.0015E-03 9.8197E-04 2.7731E-03293133 16 0.0000E+00 0.0000E+00 0 8 48 24293133 16 1.0000E-05 0.0000E+00 1.0000E+07 1.8167E-06 1.0500E+07 9.2309E-07293133 16 1.1000E+07 5.7456E-06 1.1500E+07 2.8213E-05 1.2000E+07 5.2413E-05293133 16 1.2500E+07 8.4423E-05 1.3000E+07 1.1260E-04 1.3500E+07 6.0382E-05293133 16 1.4000E+07 7.0182E-05 1.4500E+07 2.6190E-05 1.5000E+07 5.5253E-05293133 16 1.5500E+07 4.0563E-05 1.6000E+07 6.4379E-05 1.6500E+07 3.7846E-05293133 16 1.7000E+07 4.2917E-05 1.7500E+07 1.4672E-04 1.8000E+07 6.5269E-05293133 16 1.8500E+07 1.7207E-04 1.9000E+07 5.8043E-05 1.9200E+07 3.7985E-04293133 16 1.9400E+07 3.6370E-04 1.9600E+07 3.0778E-04 2.0000E+07 2.6557E-04293133 16 293133 0 2931 0 0 0 0 0 3.00640E+4 6.33800E+1 0 0 34 103025 1451 0.0 0.0 0 0 0 63025 1451 1.00000E+0 2.00000E+7 0 0 10 20023025 1451 3.00000E+2 0.0 1 0 42 33025 1451 30-Zn- 64 IRK-VIENNA EVAL-APR90 3025 1451 DIST-Feb2004 3025 1451 ----IRDF-2002 MATERIAL 3025 3025 1451 -----INCIDENT NEUTRON DATA 3025 1451 ------ENDF-6 FORMAT 3025 1451 *****************************************************************3025 1451 30-ZN- 64 IRK-VIENNA EVAL-APR90 3025 1451 DIST-JUN90 3025 1451 IRK-EVAL.NLIB 25 3025 3025 1451 *****************************************************************3025 1451 The Q value was updated prior to processing through the pre- 3025 1451 processing codes to comply with the values obtained using the 3025 1451 NNDC calculation program which is based on the 1995 Update to 3025 1451 the Atomic mass Evaluation. 3025 1451 Threshold value of 0.0 inserted at 1.0e-5 ev for MF/MT=3/103 3025 1451 after processing using the pre-processing codes. 3025 1451 *****************************************************************3025 1451 ***************** Program LINEAR (VERSION 2002-1) ***************3025 1451 For All Data Greater than 1.0000E-10 barns in Absolute Value 3025 1451 Data Linearized to Within an Accuracy of .100000000 per-cent 3025 1451 ***************** Program SIGMA1 (VERSION 2002-1) ***************3025 1451 Data Doppler Broadened to 300.000000 Kelvin 3025 1451 for All Data Greater than 1.0000E-10 barns in Absolute Value 3025 1451 Data Linearized to Within an Accuracy pf .100000000 per-cent 3025 1451 ***************** Program FIXUP (Version 2002-1) ****************3025 1451 Corrected ZA/AWR in All Sections-----------------------------Yes 3025 1451 Corrected Thresholds-----------------------------------------Yes 3025 1451 Extended Cross Sections to 20 MeV----------------------------No 3025 1451 Allow Cross Section Deletion---------------------------------No 3025 1451 Allow Cross Section Reconstruction---------------------------No 3025 1451 Make All Cross Sections Non-Negative-------------------------Yes 3025 1451 Delete Energies Not in Ascending Order-----------------------Yes 3025 1451 Deleted Duplicate Points-------------------------------------Yes 3025 1451 Check for Ascending MAT/MF/MT Order--------------------------Yes 3025 1451 Check for Legal MF/MT Numbers--------------------------------Yes 3025 1451 Allow Creation of Missing Sections---------------------------No 3025 1451 Allow Insertion of Energy Points-----------------------------No 3025 1451 Create Uniform Energy Grid-----------------------------------No 3025 1451 Delete Section if Cross Section =0 at All Energies-----------Yes 3025 1451 ***************** Program GROUPIE (VERSION 2002-1) **************3025 1451 Unshielded Group Averages Using 640 Groups 3025 1451 Weighting Spectrum: Flat (Constant) Spectrum 3025 1451 1 451 49 13025 1451 3 103 72 13025 1451 33 103 103 13025 1451 3025 1 0 3025 0 0 3.00640E+4 6.33800E+1 0 0 0 03025 3103 2.03600E+5 2.03600E+5 0 0 1 2063025 3103 206 1 3025 3103 500000.000 2.82258E-6 525000.000 8.46774E-6 550000.000 1.41129E-53025 3103 575000.000 1.97581E-5 600000.000 2.59677E-5 630000.000 3.27419E-53025 3103 660000.000 3.95161E-5 690000.000 4.62903E-5 720000.000 5.41935E-53025 3103 760000.000 6.32258E-5 800000.000 7.22581E-5 840000.000 8.12903E-53025 3103 880000.000 9.03226E-5 920000.000 9.93548E-5 960000.000 .0001083873025 3103 1000000.00 .000124194 1100000.00 .000182215 1200000.00 .0003133333025 3103 1300000.00 .000456625 1400000.00 .000819000 1500000.00 .0012710003025 3103 1600000.00 .001886125 1700000.00 .002900000 1800000.00 .0039791253025 3103 1900000.00 .006095000 2000000.00 .008635000 2100000.00 .0116475003025 3103 2200000.00 .015815000 2300000.00 .020140000 2400000.00 .0267750003025 3103 2500000.00 .034355000 2600000.00 .043101625 2700000.00 .0547000003025 3103 2800000.00 .066454875 2900000.00 .078801000 3000000.00 .0913890003025 3103 3100000.00 .101665125 3200000.00 .106290000 3300000.00 .1107457503025 3103 3400000.00 .117132000 3500000.00 .124308000 3600000.00 .1297177503025 3103 3700000.00 .130810000 3800000.00 .131846375 3900000.00 .1359710003025 3103 4000000.00 .141359000 4100000.00 .146190125 4200000.00 .1496600003025 3103 4300000.00 .153109375 4400000.00 .157469000 4500000.00 .1622010003025 3103 4600000.00 .166793500 4700000.00 .171045000 4800000.00 .1752502503025 3103 4900000.00 .178779000 5000000.00 .182031000 5100000.00 .1846305003025 3103 5200000.00 .185635000 5300000.00 .186509125 5400000.00 .1861100003025 3103 5500000.00 .185190000 5600000.00 .184270000 5700000.00 .1837981673025 3103 5800000.00 .186015333 5900000.00 .188680667 6000000.00 .1913460003025 3103 6100000.00 .194011333 6200000.00 .196676667 6300000.00 .1993420003025 3103 6400000.00 .202007333 6500000.00 .203560000 6600000.00 .2040000003025 3103 6700000.00 .204440000 6800000.00 .204880000 6900000.00 .2053200003025 3103 7000000.00 .205760000 7100000.00 .206200000 7200000.00 .2066400003025 3103 7300000.00 .207080000 7400000.00 .207520000 7500000.00 .2085860003025 3103 7600000.00 .210278000 7700000.00 .211970000 7800000.00 .2136620003025 3103 7900000.00 .215354000 8000000.00 .217046000 8100000.00 .2187380003025 3103 8200000.00 .220430000 8300000.00 .222122000 8400000.00 .2238140003025 3103 8500000.00 .225047333 8600000.00 .225822000 8700000.00 .2265966673025 3103 8800000.00 .227371333 8900000.00 .228146000 9000000.00 .2289206673025 3103 9100000.00 .229695333 9200000.00 .230470000 9300000.00 .2312446673025 3103 9400000.00 .232019333 9500000.00 .232794000 9600000.00 .2335686673025 3103 9700000.00 .234343333 9800000.00 .235118000 9900000.00 .2358926673025 3103 10000000.0 .237326750 10100000.0 .239420250 10200000.0 .2415137503025 3103 10300000.0 .243607250 10400000.0 .245700750 10500000.0 .2477942503025 3103 10600000.0 .249887750 10700000.0 .251981250 10800000.0 .2540747503025 3103 10900000.0 .256168250 11000000.0 .258261750 11100000.0 .2603552503025 3103 11200000.0 .262448750 11300000.0 .264542250 11400000.0 .2666357503025 3103 11500000.0 .268729250 11600000.0 .270822750 11700000.0 .2729162503025 3103 11800000.0 .275009750 11900000.0 .277103250 12000000.0 .2756823333025 3103 12100000.0 .270747000 12200000.0 .265811667 12300000.0 .2608763333025 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1.649000-3302533103 7.343000-4 2.925000-4 4.268000-4 1.552000-3 2.783000-3 3.330000-3302533103 4.958000-3 1.162000-3 5.634000-4 2.219000-4 4.848000-4 3.209000-4302533103 6.358000-4 7.888000-4 1.274000-3 1.183000-3 2.002000-4 5.519000-4302533103 5.540000-4 6.012000-4 1.011000-3 2.382000-3 5.401000-4 1.775000-4302533103 7.415000-5 1.623000-4 2.084000-4 3.578000-4 1.548000-3 6.073000-4302533103 4.197000-4 7.127000-4 1.670000-3 2.862000-3 1.544000-3 2.636000-3302533103 6.208000-3 3.103000-3 3.019000-3 6.528000-3 8.573000-3 1.380000-2302533103 5.830000-2 302533103 302533 0 3025 0 0 0 0 0 3.10000+ 4 0.0000 + 0 -1 0 34 13100 1451 0.0 + 0 0.0 + 0 0 0 0 63100 1451 1.0 + 0 0.0 + 0 0 0 10 23100 1451 0.0 + 0 0.0 + 0 0 0 8 23100 1451 31-GA- 0 ASTM EVAL- 91 P.J.GRIFFIN 3100 1451 PRIVATE COM. DIST-JAN03 811211 3100 1451 ----IRDF-02/NMF-G MATERIAL 3100 3100 1451 -----INCIDENT NEUTRON DATA 3100 1451 ------ENDF-6 FORMAT 3100 1451 DAMAGE CROSS SECTION OF GaAs (ELECTRONICS DAMAGE) 3100 1451 ASTM E722 MT=900 3100 1451 3100 1451 1 451 14 03100 1451 3 900 217 03100 1451 3100 1 0 3100 0 0 3.10000+ 4 0.0 + 0 0 0 0 03100 3900 0.0 + 0 0.0 + 0 0 0 1 6413100 3900 641 1 3100 3900 1.0000E-04 1.3716E+02 1.0500E-04 1.3727E+02 1.1000E-04 1.3738E+023100 3900 1.1500E-04 1.3749E+02 1.2000E-04 1.3760E+02 1.2750E-04 1.3770E+023100 3900 1.3500E-04 1.3780E+02 1.4250E-04 1.3789E+02 1.5000E-04 1.3799E+023100 3900 1.6000E-04 1.3808E+02 1.7000E-04 1.3818E+02 1.8000E-04 1.3827E+023100 3900 1.9000E-04 1.3836E+02 2.0000E-04 1.3845E+02 2.1000E-04 1.3854E+023100 3900 2.2000E-04 1.3864E+02 2.3000E-04 1.3878E+02 2.4000E-04 1.3891E+023100 3900 2.5500E-04 1.3905E+02 2.7000E-04 1.3919E+02 2.8000E-04 1.3901E+023100 3900 3.0000E-04 1.3853E+02 3.2000E-04 1.3804E+02 3.4000E-04 1.3756E+023100 3900 3.6000E-04 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1.6000E-01 1.1163E+02 1.7000E-01 1.1093E+02 1.8000E-01 1.1023E+023100 3900 1.9000E-01 1.0952E+02 2.0000E-01 1.0878E+02 2.1000E-01 1.0803E+023100 3900 2.2000E-01 1.0205E+02 2.3000E-01 9.9233E+01 2.4000E-01 9.8960E+013100 3900 2.5500E-01 9.8666E+01 2.7000E-01 9.8355E+01 2.8000E-01 9.8024E+013100 3900 3.0000E-01 9.7657E+01 3.2000E-01 9.7257E+01 3.4000E-01 9.6832E+013100 3900 3.6000E-01 9.6384E+01 3.8000E-01 9.5300E+01 4.0000E-01 9.3602E+013100 3900 4.2500E-01 9.1901E+01 4.5000E-01 9.0191E+01 4.7500E-01 8.8463E+013100 3900 5.0000E-01 8.7288E+01 5.2500E-01 8.6660E+01 5.5000E-01 8.6026E+013100 3900 5.7500E-01 8.5389E+01 6.0000E-01 8.4753E+01 6.3000E-01 8.1218E+013100 3900 6.6000E-01 7.8036E+01 6.9000E-01 7.7941E+01 7.2000E-01 7.7786E+013100 3900 7.6000E-01 7.7619E+01 8.0000E-01 7.7453E+01 8.4000E-01 7.7178E+013100 3900 8.8000E-01 7.7126E+01 9.2000E-01 7.8234E+01 9.6000E-01 7.9780E+013100 3900 1.0000E+00 7.9775E+01 1.0500E+00 7.6397E+01 1.1000E+00 7.3491E+013100 3900 1.1500E+00 7.2477E+01 1.2000E+00 7.3017E+01 1.2750E+00 7.1021E+013100 3900 1.3500E+00 6.9159E+01 1.4250E+00 6.9890E+01 1.5000E+00 7.0163E+013100 3900 1.6000E+00 6.9972E+01 1.7000E+00 6.8695E+01 1.8000E+00 6.6386E+013100 3900 1.9000E+00 6.4332E+01 2.0000E+00 6.3911E+01 2.1000E+00 6.3727E+013100 3900 2.2000E+00 6.3432E+01 2.3000E+00 6.2665E+01 2.4000E+00 6.1244E+013100 3900 2.5500E+00 6.0412E+01 2.7000E+00 5.9687E+01 2.8000E+00 5.8895E+013100 3900 3.0000E+00 5.8237E+01 3.2000E+00 5.7714E+01 3.4000E+00 5.7130E+013100 3900 3.6000E+00 5.6487E+01 3.8000E+00 5.5666E+01 4.0000E+00 5.4672E+013100 3900 4.2500E+00 5.3638E+01 4.5000E+00 5.2566E+01 4.7500E+00 5.1540E+013100 3900 5.0000E+00 5.0558E+01 5.2500E+00 4.9550E+01 5.5000E+00 4.8337E+013100 3900 5.7500E+00 4.7067E+01 6.0000E+00 4.5631E+01 6.3000E+00 4.4439E+013100 3900 6.6000E+00 4.3436E+01 6.9000E+00 4.2163E+01 7.2000E+00 4.0910E+013100 3900 7.6000E+00 3.9890E+01 8.0000E+00 3.8870E+01 8.4000E+00 3.7851E+013100 3900 8.8000E+00 3.6896E+01 9.2000E+00 3.6007E+01 9.6000E+00 3.5099E+013100 3900 1.0000E+01 3.4059E+01 1.0500E+01 3.2995E+01 1.1000E+01 3.1874E+013100 3900 1.1500E+01 3.0694E+01 1.2000E+01 2.9514E+01 1.2750E+01 2.8316E+013100 3900 1.3500E+01 2.7264E+01 1.4250E+01 2.6413E+01 1.5000E+01 2.5561E+013100 3900 1.6000E+01 2.4701E+01 1.7000E+01 2.3855E+01 1.8000E+01 2.3024E+013100 3900 1.9000E+01 2.2192E+01 2.0000E+01 2.1352E+01 2.1000E+01 2.0509E+013100 3900 2.2000E+01 1.9659E+01 2.3000E+01 1.8800E+01 2.4000E+01 1.8044E+013100 3900 2.5500E+01 1.7385E+01 2.7000E+01 1.6723E+01 2.8000E+01 1.6061E+013100 3900 3.0000E+01 1.5445E+01 3.2000E+01 1.4875E+01 3.4000E+01 1.4305E+013100 3900 3.6000E+01 1.3734E+01 3.8000E+01 1.3165E+01 4.0000E+01 1.2595E+013100 3900 4.2500E+01 1.2024E+01 4.5000E+01 1.1449E+01 4.7500E+01 1.0988E+013100 3900 5.0000E+01 1.0638E+01 5.2500E+01 1.0287E+01 5.5000E+01 9.9305E+003100 3900 5.7500E+01 9.5438E+00 6.0000E+01 9.1394E+00 6.3000E+01 8.8213E+003100 3900 6.6000E+01 8.5463E+00 6.9000E+01 8.2122E+00 7.2000E+01 7.9198E+003100 3900 7.6000E+01 7.6836E+00 8.0000E+01 7.4465E+00 8.4000E+01 7.1973E+003100 3900 8.8000E+01 6.9399E+00 9.2000E+01 6.6811E+00 9.6000E+01 6.4211E+003100 3900 1.0000E+02 6.1490E+00 1.0500E+02 5.8714E+00 1.1000E+02 5.6203E+003100 3900 1.1500E+02 5.3994E+00 1.2000E+02 5.1737E+00 1.2750E+02 4.9434E+003100 3900 1.3500E+02 4.7500E+00 1.4250E+02 4.5899E+00 1.5000E+02 4.4266E+003100 3900 1.6000E+02 4.2632E+00 1.7000E+02 4.3032E+00 1.8000E+02 4.4669E+003100 3900 1.9000E+02 4.7471E+00 2.0000E+02 3.8537E+00 2.1000E+02 4.3474E+003100 3900 2.2000E+02 3.6819E+00 2.3000E+02 3.5706E+00 2.4000E+02 4.0995E+003100 3900 2.5500E+02 3.0083E+00 2.7000E+02 5.0913E+00 2.8000E+02 5.6135E+003100 3900 3.0000E+02 4.5205E+00 3.2000E+02 6.3687E+00 3.4000E+02 2.6198E+003100 3900 3.6000E+02 3.4985E+00 3.8000E+02 3.6612E+00 4.0000E+02 3.1876E+003100 3900 4.2500E+02 2.1236E+00 4.5000E+02 2.1440E+00 4.7500E+02 4.5796E+003100 3900 5.0000E+02 2.9168E+00 5.2500E+02 4.3518E+00 5.5000E+02 1.4714E+003100 3900 5.7500E+02 2.0069E+00 6.0000E+02 2.2452E+00 6.3000E+02 4.2128E+003100 3900 6.6000E+02 3.1071E+00 6.9000E+02 1.5217E+00 7.2000E+02 1.2373E+003100 3900 7.6000E+02 1.1523E+00 8.0000E+02 1.1249E+00 8.4000E+02 1.4630E+003100 3900 8.8000E+02 1.9094E+00 9.2000E+02 3.0359E+00 9.6000E+02 1.3138E+003100 3900 1.0000E+03 3.6528E+00 1.0500E+03 9.7157E-01 1.1000E+03 2.9769E+003100 3900 1.1500E+03 1.0585E+00 1.2000E+03 4.0391E+00 1.2750E+03 6.7298E-013100 3900 1.3500E+03 6.7764E-01 1.4250E+03 1.3796E+00 1.5000E+03 6.7876E-013100 3900 1.6000E+03 6.7518E-01 1.7000E+03 7.0528E-01 1.8000E+03 4.1174E+003100 3900 1.9000E+03 7.0857E-01 2.0000E+03 7.4331E-01 2.1000E+03 1.9085E+003100 3900 2.2000E+03 9.7233E-01 2.3000E+03 1.0859E+01 2.4000E+03 8.7289E-013100 3900 2.5500E+03 3.9196E+00 2.7000E+03 7.9842E-01 2.8000E+03 7.5788E-013100 3900 3.0000E+03 7.9912E-01 3.2000E+03 1.0212E+00 3.4000E+03 1.4056E+013100 3900 3.6000E+03 1.0500E+00 3.8000E+03 9.8169E-01 4.0000E+03 1.8146E+003100 3900 4.2500E+03 7.5901E-01 4.5000E+03 7.1954E-01 4.7500E+03 7.7158E-013100 3900 5.0000E+03 8.2704E-01 5.2500E+03 9.7526E-01 5.5000E+03 1.0957E+013100 3900 5.7500E+03 1.1932E+01 6.0000E+03 8.0398E-01 6.3000E+03 6.6710E-013100 3900 6.6000E+03 7.2019E-01 6.9000E+03 5.1709E+00 7.2000E+03 5.2950E-013100 3900 7.6000E+03 4.3053E-01 8.0000E+03 4.0029E-01 8.4000E+03 3.8638E-013100 3900 8.8000E+03 3.9023E-01 9.2000E+03 4.0210E-01 9.6000E+03 4.3962E-013100 3900 1.0000E+04 5.2598E-01 1.0500E+04 6.7015E-01 1.1000E+04 8.2229E-013100 3900 1.1500E+04 9.4465E-01 1.2000E+04 1.1042E+00 1.2750E+04 1.3147E+003100 3900 1.3500E+04 1.4648E+00 1.4250E+04 1.5236E+00 1.5000E+04 1.6630E+003100 3900 1.6000E+04 1.8367E+00 1.7000E+04 1.9586E+00 1.8000E+04 1.4132E+013100 3900 1.9000E+04 3.5726E+00 2.0000E+04 1.0644E+00 2.1000E+04 7.5453E-013100 3900 2.2000E+04 5.9425E-01 2.3000E+04 4.9881E-01 2.4000E+04 4.4759E-013100 3900 2.5500E+04 4.2479E-01 2.7000E+04 4.1753E-01 2.8000E+04 4.2775E-013100 3900 3.0000E+04 4.6204E-01 3.2000E+04 5.5056E-01 3.4000E+04 8.5951E-013100 3900 3.6000E+04 1.5997E+00 3.8000E+04 1.3238E+01 4.0000E+04 1.0191E+023100 3900 4.2500E+04 2.1149E+00 4.5000E+04 8.6689E-01 4.7500E+04 3.9098E-013100 3900 5.0000E+04 3.9491E-01 5.2500E+04 3.9884E-01 5.5000E+04 4.0281E-013100 3900 5.7500E+04 4.0683E-01 6.0000E+04 4.1084E-01 6.3000E+04 4.1389E-013100 3900 6.6000E+04 4.1647E-01 6.9000E+04 4.1956E-01 7.2000E+04 4.2216E-013100 3900 7.6000E+04 4.2429E-01 8.0000E+04 4.2642E-01 8.4000E+04 4.2855E-013100 3900 8.8000E+04 4.3085E-01 9.2000E+04 4.3332E-01 9.6000E+04 4.3585E-013100 3900 1.0000E+05 4.3844E-01 1.0500E+05 4.4112E-01 1.1000E+05 4.4351E-013100 3900 1.1500E+05 4.4564E-01 1.2000E+05 4.4783E-01 1.2750E+05 4.5009E-013100 3900 1.3500E+05 4.5204E-01 1.4250E+05 4.5363E-01 1.5000E+05 4.5532E-013100 3900 1.6000E+05 4.5702E-01 1.7000E+05 4.5940E-01 1.8000E+05 4.6246E-013100 3900 1.9000E+05 4.6553E-01 2.0000E+05 4.6868E-01 2.1000E+05 4.7191E-013100 3900 2.2000E+05 4.7515E-01 2.3000E+05 4.7853E-01 2.4000E+05 4.8152E-013100 3900 2.5500E+05 4.8412E-01 2.7000E+05 4.8686E-01 2.8000E+05 4.8962E-013100 3900 3.0000E+05 4.9219E-01 3.2000E+05 4.9467E-01 3.4000E+05 4.9715E-013100 3900 3.6000E+05 4.9970E-01 3.8000E+05 5.0412E-01 4.0000E+05 5.1023E-013100 3900 4.2500E+05 5.1650E-01 4.5000E+05 5.2290E-01 4.7500E+05 5.2871E-013100 3900 5.0000E+05 5.3410E-01 5.2500E+05 5.3952E-01 5.5000E+05 5.4507E-013100 3900 5.7500E+05 5.5083E-01 6.0000E+05 5.5678E-01 6.3000E+05 5.6135E-013100 3900 6.6000E+05 5.6532E-01 6.9000E+05 5.7023E-01 7.2000E+05 5.7450E-013100 3900 7.6000E+05 5.7802E-01 8.0000E+05 5.8169E-01 8.4000E+05 5.8544E-013100 3900 8.8000E+05 5.8936E-01 9.2000E+05 5.9345E-01 9.6000E+05 5.9769E-013100 3900 1.0000E+06 6.0216E-01 1.1000E+06 6.0685E-01 1.2000E+06 6.1120E-013100 3900 1.3000E+06 6.1510E-01 1.4000E+06 6.1924E-01 1.5000E+06 6.2363E-013100 3900 1.6000E+06 6.2743E-01 1.7000E+06 6.3069E-01 1.8000E+06 6.3409E-013100 3900 1.9000E+06 6.3765E-01 2.0000E+06 6.4213E-01 2.1000E+06 6.4731E-013100 3900 2.2000E+06 6.5271E-01 2.3000E+06 6.5828E-01 2.4000E+06 6.6416E-013100 3900 2.5000E+06 6.7016E-01 2.6000E+06 6.7640E-01 2.7000E+06 6.8214E-013100 3900 2.8000E+06 6.8713E-01 2.9000E+06 6.9241E-01 3.0000E+06 6.9792E-013100 3900 3.1000E+06 7.0329E-01 3.2000E+06 7.0832E-01 3.3000E+06 7.1344E-013100 3900 3.4000E+06 7.1892E-01 3.5000E+06 7.2678E-01 3.6000E+06 7.3700E-013100 3900 3.7000E+06 7.4758E-01 3.8000E+06 7.5868E-01 3.9000E+06 7.6888E-013100 3900 4.0000E+06 7.7855E-01 4.1000E+06 7.8855E-01 4.2000E+06 7.9881E-013100 3900 4.3000E+06 8.0987E-01 4.4000E+06 8.2143E-01 4.5000E+06 8.3051E-013100 3900 4.6000E+06 8.3849E-01 4.7000E+06 8.4870E-01 4.8000E+06 8.5739E-013100 3900 4.9000E+06 8.6500E-01 5.0000E+06 8.7283E-01 5.1000E+06 8.8088E-013100 3900 5.2000E+06 8.8981E-01 5.3000E+06 8.9897E-01 5.4000E+06 9.0855E-013100 3900 5.5000E+06 9.1915E-01 5.6000E+06 9.3000E-01 5.7000E+06 9.4042E-013100 3900 5.8000E+06 9.5009E-01 5.9000E+06 9.6033E-01 6.0000E+06 9.7162E-013100 3900 6.1000E+06 9.8114E-01 6.2000E+06 9.8952E-01 6.3000E+06 9.9837E-013100 3900 6.4000E+06 1.0078E+00 6.5000E+06 1.0189E+00 6.6000E+06 1.0315E+003100 3900 6.7000E+06 1.0447E+00 6.8000E+06 1.0583E+00 6.9000E+06 1.0726E+003100 3900 7.0000E+06 1.0877E+00 7.1000E+06 1.1034E+00 7.2000E+06 1.1178E+003100 3900 7.3000E+06 1.1311E+00 7.4000E+06 1.1448E+00 7.5000E+06 1.1593E+003100 3900 7.6000E+06 1.1731E+00 7.7000E+06 1.1864E+00 7.8000E+06 1.2006E+003100 3900 7.9000E+06 1.2158E+00 8.0000E+06 1.2312E+00 8.1000E+06 1.2482E+003100 3900 8.2000E+06 1.2663E+00 8.3000E+06 1.2848E+00 8.4000E+06 1.3038E+003100 3900 8.5000E+06 1.3213E+00 8.6000E+06 1.3402E+00 8.7000E+06 1.3604E+003100 3900 8.8000E+06 1.3826E+00 8.9000E+06 1.4067E+00 9.0000E+06 1.4263E+003100 3900 9.1000E+06 1.4438E+00 9.2000E+06 1.4683E+00 9.3000E+06 1.4925E+003100 3900 9.4000E+06 1.5131E+00 9.5000E+06 1.5349E+00 9.6000E+06 1.5580E+003100 3900 9.7000E+06 1.5822E+00 9.8000E+06 1.6083E+00 9.9000E+06 1.6361E+003100 3900 1.0000E+07 1.6661E+00 1.0100E+07 1.6983E+00 1.0200E+07 1.7287E+003100 3900 1.0300E+07 1.7570E+00 1.0400E+07 1.7871E+00 1.0500E+07 1.8197E+003100 3900 1.0600E+07 1.8492E+00 1.0700E+07 1.8738E+00 1.0800E+07 1.9004E+003100 3900 1.0900E+07 1.9292E+00 1.1000E+07 1.9574E+00 1.1100E+07 1.9858E+003100 3900 1.1200E+07 2.0158E+00 1.1300E+07 2.0480E+00 1.1400E+07 2.0812E+003100 3900 1.1500E+07 2.1171E+00 1.1600E+07 2.1558E+00 1.1700E+07 2.1921E+003100 3900 1.1800E+07 2.2245E+00 1.1900E+07 2.2593E+00 1.2000E+07 2.2965E+003100 3900 1.2100E+07 2.3330E+00 1.2200E+07 2.3672E+00 1.2300E+07 2.4048E+003100 3900 1.2400E+07 2.4446E+00 1.2500E+07 2.4905E+00 1.2600E+07 2.5429E+003100 3900 1.2700E+07 2.5983E+00 1.2800E+07 2.6581E+00 1.2900E+07 2.7158E+003100 3900 1.3000E+07 2.7717E+00 1.3100E+07 2.8309E+00 1.3200E+07 2.8945E+003100 3900 1.3300E+07 2.9637E+00 1.3400E+07 3.0395E+00 1.3500E+07 3.1006E+003100 3900 1.3600E+07 3.1571E+00 1.3700E+07 3.2270E+00 1.3800E+07 3.2906E+003100 3900 1.3900E+07 3.3451E+00 1.4000E+07 3.4043E+00 1.4100E+07 3.4657E+003100 3900 1.4200E+07 3.5320E+00 1.4300E+07 3.6035E+00 1.4400E+07 3.6814E+003100 3900 1.4500E+07 3.7638E+00 1.4600E+07 3.8551E+00 1.4700E+07 3.9420E+003100 3900 1.4800E+07 4.0228E+00 1.4900E+07 4.1110E+00 1.5000E+07 4.2070E+003100 3900 1.5100E+07 4.2920E+00 1.5200E+07 4.3682E+00 1.5300E+07 4.4448E+003100 3900 1.5400E+07 4.5282E+00 1.5500E+07 4.6114E+00 1.5600E+07 4.6932E+003100 3900 1.5700E+07 4.7819E+00 1.5800E+07 4.8711E+00 1.5900E+07 4.9679E+003100 3900 1.6000E+07 5.0727E+00 1.6100E+07 5.1861E+00 1.6200E+07 5.2913E+003100 3900 1.6300E+07 5.3884E+00 1.6400E+07 5.4949E+00 1.6500E+07 5.6036E+003100 3900 1.6600E+07 5.7124E+00 1.6700E+07 5.8213E+00 1.6800E+07 5.9305E+003100 3900 1.6900E+07 6.0479E+00 1.7000E+07 6.1830E+00 1.7100E+07 6.3300E+003100 3900 1.7200E+07 6.4924E+00 1.7300E+07 6.6592E+00 1.7400E+07 6.8311E+003100 3900 1.7500E+07 6.9891E+00 1.7600E+07 7.1631E+00 1.7700E+07 7.3530E+003100 3900 1.7800E+07 7.5532E+00 1.7900E+07 7.7820E+00 1.8000E+07 7.9627E+003100 3900 1.8100E+07 8.1275E+00 1.8200E+07 8.3410E+00 1.8300E+07 8.5423E+003100 3900 1.8400E+07 8.7103E+00 1.8500E+07 8.8786E+00 1.8600E+07 9.0687E+003100 3900 1.8700E+07 9.2755E+00 1.8800E+07 9.4867E+00 1.8900E+07 9.7326E+003100 3900 1.9000E+07 9.9862E+00 1.9100E+07 1.0262E+01 1.9200E+07 1.0532E+013100 3900 1.9300E+07 1.0782E+01 1.9400E+07 1.1053E+01 1.9500E+07 1.1350E+013100 3900 1.9600E+07 1.1615E+01 1.9700E+07 1.1845E+01 1.9800E+07 1.2091E+013100 3900 1.9900E+07 1.2355E+01 2.0000E+07 0.0000E+00 3100 3900 3100 3 0 3100 0 0 0 0 0 3.30750E+4 7.42780E+1 0 0 34 103325 1451 0.0 0.0 0 0 0 63325 1451 1.00000E+0 2.00000E+7 0 0 10 20023325 1451 3.00000E+2 0.0 1 0 128 33325 1451 33-As- 75 FEI EVAL-Sep94 K.I.Zolotarev 3325 1451 DIST-Feb2004 3325 1451 ----IRDF-2002 MATERIAL 3325 3325 1451 -----INCIDENT NEUTRON DATA 3325 1451 ------ENDF-6 FORMAT 3325 1451 ***************************************************************** 3325 1451 33-AS-75 FEI EVAL-Sep94 K.I.Zolotarev 3325 1451 DIST-May03 20030504 3325 1451 ----BROND-2 MATERIAL 3325 3325 1451 ------Russian Reactor Dosimetry File RRDF-2002 3325 1451 ***************************************************************** 3325 1451 Authors of evaluation: K.Zolotarev, V.Manokhin, A.Pashchenko 3325 1451 ***************************************************************** 3325 1451 MF=3 3325 1451 MT= 16 - As-75(n,2n)As-74 reaction 3325 1451 ------------------------------------- 3325 1451 Evaluation was made on the base of experimental data [1-18]. 3325 1451 Systematics of (n,2n) excitation functions [19] was taken into 3325 1451 account in the energy region 10.5 - 13.0 MeV and 19.0 - 20.0 MeV. 3325 1451 Cross section data of E.Paul and R.Clarke [20] were rejected 3325 1451 due to inconsistency with main bulk of experimental data and 3325 1451 systematics of (n,2n) excitation functions. 3325 1451 Statistical analysis of input cross section data was carried 3325 1451 out by means of PADE-2 code [21]. Rational function was used as 3325 1451 the model function [22]. 3325 1451 Integral experimental data for U-235 thermal fission neutron 3325 1451 spectrum [23-24] was used for testing evaluated As-75(n,2n)As-74 3325 1451 excitation function. The results of testing are given in Table 1. 3325 1451 Data for U-235 thermal fission neutron spectrum and Cf-252 3325 1451 spontaneous fission neutron spectrum were taken from ref.[25] and 3325 1451 [26], respectively. 3325 1451 Table 1. 3325 1451 ÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÂÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÂÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄ 3325 1451 TYPE OF SPECTRUM ³ ,mb (calc.) ³ , mb (measured) 3325 1451 ÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÅÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÅÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄ 3325 1451 U-235 neutron fission ³ 0.30920 ³ 0.304 +- 0.036 [22] 3325 1451 ³ ³ 0.311 +- 0.023 [23] 3325 1451 ÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÅÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÅÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄ 3325 1451 Cf-252 spont. fission ³ 0.61804 ³ 3325 1451 ÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÁÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÁÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄ 3325 1451 3325 1451 MT=33 3325 1451 MT= 16 -(n,2n) cross section cov. matrix 3325 1451 ---------------------------------------- 3325 1451 Uncertainties in the evaluated excitation function for the 3325 1451 reaction As-75(n,2n)As-74 are given in the form of relative cova- 3325 1451 riance matrix for the 19-neutron energy groups (LB=5). Covariance 3325 1451 matrix of uncertainties was calculated simultaneously with 3325 1451 recommended cross section data by means of PADE-2 code. 3325 1451 Eigenvalues of the 6-th digits relative covariance matrix 3325 1451 given in the 33-file are the following: 3325 1451 3325 1451 2.79921E-09 3.33401E-09 3.76284E-09 3.88107E-09 3325 1451 4.19122E-09 4.53852E-09 4.91657E-09 5.52980E-09 3325 1451 7.06444E-09 1.01577E-08 1.66580E-08 2.33017E-08 3325 1451 1.49576E-07 8.86325E-04 3.07310E-03 3.70656E-03 3325 1451 9.37209E-03 1.38518E-02 8.84386E-02 3325 1451 3325 1451 References : 3325 1451 1. R.J.Prestwood, B.P.Bayhurst Phys. Rev., v.121, p.1438, 1961 3325 1451 2. B.Granger, M.Longueve Report EANDC(E)-49L, p.83, Oct. 1963 3325 1451 3. P.Hille, H.Muenzer Acta Phys. Austriaca, v.23, p.44, May 1966 3325 1451 4. J.Csikai, G.Peto Acta Physica Hung., v.23, p.87, May 1967 3325 1451 5. M.Bormann et al. Nucl. Phys., v.A115, p.309, July 1968 3325 1451 M.Bormann et al. Progress Report EANDC(E)-76, January 1967 3325 1451 6. P.R.Prasad et al. Nucl. Phys., v.A138, p.85, November 1969 3325 1451 7. P.Venugopala Rao et al. Phys. Rev., v.C3, p.629, Feb. 1971 3325 1451 8. M.Wagner, M.Uhl Oesterr. Akad. Wiss., Math.+Naturw. Anzeiger,3325 1451 v.108, p.185, November 1971 3325 1451 9. S.M.Qaim Nucl. Phys., Sec.A, v.185, p.614, May 1972 3325 1451 10. D.V.Viktorov, V.L.Sjablin Jadernaja Fizika, v.15,p.1099, 1972 3325 1451 11. C.T.Simpson et al. Inorg. Nucl. Chem., v.35, p.2085,June 1973 3325 1451 12. J.Araminowicz,J.Dresler Peport INR-1464, Warsaw,p.14,May 1973 3325 1451 13. W.Grochulski et al. Acta Physica Polonica, v.B6,p.139, 1975 3325 1451 14. R.A.Sigg Dissertation Abstract, sec.B, v.37,p.2237, Nov.1976 3325 1451 15. J.L.Casanova,M.L.Sanchez Anales de Fisica y Quimica (Spain), 3325 1451 v.72, n.3, p.186, July 1976 3325 1451 16. S.Sothras Dissertation Abstract, sec.B,v.38,p.280, July 1978 3325 1451 17. C.Konno, Y.Ikeda, K.Oishi e.a. Report JAERI-1329, Oct. 1993 3325 1451 18. I.Birn, S.M.Qaim Nucl. Sci. Eng., v.116, p.125-137, 1994 3325 1451 19. V.N.Manokhin VANT, Ser.:Yadernye Konstanty, 1994, No. 1, 3325 1451 pp.18-22 3325 1451 20. E.B.Paul, R.L.Clarke Can. J. Phys., v.31, p.267, 1953 3325 1451 21. S.A.Badikov et al. Preprint FEI-1686, Obninsk, 1985 3325 1451 22. S.Badikov, N.Rabotnov, K.Zolotarev Proc. of NEANSC Speciali- 3325 1451 st's Meeting on Evaluation and Processing of Covariance Data, 3325 1451 Oak Ridge, USA, 7-9 September 1992, OECD, Paris, 1993, p.105 3325 1451 23. F.Nasyrov, B.D.Sciborskij Atomnaja Energija, v.25, no.5, 3325 1451 p.437, November 1968 3325 1451 24. E.Steinnes Radiochimica Acta, v.13, p.169, June 1970 3325 1451 25. L.W.Weston et al. Evaluated Neutron Data for U-235, ENDF/B-VI 3325 1451 Library, MAT=9228, MF=5, MT=18, eval. April 1989 3325 1451 26. W.Mannhart IAEA-TECDOC-410, p.158, IAEA, Vienna, 1987 3325 1451 ***************************************************************** 3325 1451 The Q values and threshold energies were updated prior to pro- 3325 1451 cessing through the codes to comply with the values obtained 3325 1451 using the NNDC calculation program which is based on the 1995 3325 1451 Update to the Atomic mass Evaluation. 3325 1451 3325 1451 File 2 added to the pointwise file containing only the effective 3325 1451 scattering radius with no resonance parameters given. 3325 1451 Taken from ENDF/B-VI 3325 1451 ***************************************************************** 3325 1451 ***************** Program LINEAR (VERSION 2002-1) ***************3325 1451 For All Data Greater than 1.0000E-10 barns in Absolute Value 3325 1451 Data Linearized to Within an Accuracy of .100000000 per-cent 3325 1451 ***************** Program SIGMA1 (VERSION 2002-1) ***************3325 1451 Data Doppler Broadened to 300.000000 Kelvin 3325 1451 for All Data Greater than 1.0000E-10 barns in Absolute Value 3325 1451 Data Linearized to Within an Accuracy pf .100000000 per-cent 3325 1451 ***************** Program FIXUP (Version 2002-1) ****************3325 1451 Corrected ZA/AWR in All Sections-----------------------------Yes 3325 1451 Corrected Thresholds-----------------------------------------Yes 3325 1451 Extended Cross Sections to 20 MeV----------------------------No 3325 1451 Allow Cross Section Deletion---------------------------------No 3325 1451 Allow Cross Section Reconstruction---------------------------No 3325 1451 Make All Cross Sections Non-Negative-------------------------Yes 3325 1451 Delete Energies Not in Ascending Order-----------------------Yes 3325 1451 Deleted Duplicate Points-------------------------------------Yes 3325 1451 Check for Ascending MAT/MF/MT Order--------------------------Yes 3325 1451 Check for Legal MF/MT Numbers--------------------------------Yes 3325 1451 Allow Creation of Missing Sections---------------------------No 3325 1451 Allow Insertion of Energy Points-----------------------------No 3325 1451 Create Uniform Energy Grid-----------------------------------No 3325 1451 Delete Section if Cross Section =0 at All Energies-----------Yes 3325 1451 ***************** Program GROUPIE (VERSION 2002-1) **************3325 1451 Unshielded Group Averages Using 640 Groups 3325 1451 Weighting Spectrum: Flat (Constant) Spectrum 3325 1451 1 451 135 13325 1451 3 16 36 13325 1451 33 16 42 13325 1451 3325 1 0 3325 0 0 3.30750E+4 7.42780E+1 0 0 0 03325 3 16 -1.02442E+7-1.02442E+7 0 0 1 983325 3 16 98 1 3325 3 16 10300000.0 6.04805E-6 10400000.0 .000608735 10500000.0 .0052550503325 3 16 10600000.0 .014392210 10700000.0 .028074850 10800000.0 .0460172503325 3 16 10900000.0 .067864400 11000000.0 .093207450 11100000.0 .1216010003325 3 16 11200000.0 .152580000 11300000.0 .185675000 11400000.0 .2204260003325 3 16 11500000.0 .256394000 11600000.0 .293214625 11700000.0 .3303858753325 3 16 11800000.0 .367557125 11900000.0 .404728375 12000000.0 .4413397503325 3 16 12100000.0 .477391250 12200000.0 .512559250 12300000.0 .5468437503325 3 16 12400000.0 .580032500 12500000.0 .612125500 12600000.0 .6430125003325 3 16 12700000.0 .672693500 12800000.0 .701134000 12900000.0 .7283340003325 3 16 13000000.0 .754310750 13100000.0 .779064250 13200000.0 .8026452503325 3 16 13300000.0 .825053750 13400000.0 .846360250 13500000.0 .8665647503325 3 16 13600000.0 .885747750 13700000.0 .903909250 13800000.0 .9209150003325 3 16 13900000.0 .936765000 14000000.0 .952615000 14100000.0 .9672400003325 3 16 14200000.0 .980640000 14300000.0 .994040000 14400000.0 1.006381673325 3 16 14500000.0 1.01766500 14600000.0 1.02894833 14700000.0 1.039320003325 3 16 14800000.0 1.04878000 14900000.0 1.05824000 15000000.0 1.066797503325 3 16 15100000.0 1.07445250 15200000.0 1.08210750 15300000.0 1.089762503325 3 16 15400000.0 1.09655125 15500000.0 1.10247375 15600000.0 1.108396253325 3 16 15700000.0 1.11431875 15800000.0 1.11943900 15900000.0 1.123757003325 3 16 16000000.0 1.12807500 16100000.0 1.13239300 16200000.0 1.136711003325 3 16 16300000.0 1.14029100 16400000.0 1.14313300 16500000.0 1.145975003325 3 16 16600000.0 1.14881700 16700000.0 1.15165900 16800000.0 1.153825003325 3 16 16900000.0 1.15531500 17000000.0 1.15680500 17100000.0 1.158295003325 3 16 17200000.0 1.15978500 17300000.0 1.16127500 17400000.0 1.162227503325 3 16 17500000.0 1.16264250 17600000.0 1.16305750 17700000.0 1.163472503325 3 16 17800000.0 1.16335214 17900000.0 1.16269643 18000000.0 1.162040713325 3 16 18100000.0 1.16138500 18200000.0 1.16072929 18300000.0 1.160073573325 3 16 18400000.0 1.15941786 18500000.0 1.15810500 18600000.0 1.156135003325 3 16 18700000.0 1.15416500 18800000.0 1.15219500 18900000.0 1.150225003325 3 16 19000000.0 1.14825500 19100000.0 1.14628500 19200000.0 1.143680003325 3 16 19300000.0 1.14044000 19400000.0 1.13720000 19500000.0 1.133960003325 3 16 19600000.0 1.13072000 19700000.0 1.12748000 19800000.0 1.123815003325 3 16 19900000.0 1.11972500 20000000.0 0.0 3325 3 16 3325 3 0 3325 0 0 3.30750E+4 7.42780E+1 0 0 0 1332533 16 0.000000+0 0.000000+0 0 16 0 1332533 16 0.000000+0 0.000000+0 1 5 231 21332533 16 1.000000-5 1.030000+7 1.100000+7 1.150000+7 1.200000+7 1.250000+7332533 16 1.300000+7 1.350000+7 1.400000+7 1.450000+7 1.500000+7 1.550000+7332533 16 1.600000+7 1.650000+7 1.700000+7 1.750000+7 1.800000+7 1.850000+7332533 16 1.900000+7 1.950000+7 2.000000+7 0.000000+0 0.000000+0 0.000000+0332533 16 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0332533 16 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0332533 16 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0 4.254080-2332533 16 3.081380-2 1.931400-2 1.096210-2 5.721690-3 2.776080-3 1.317510-3332533 16 7.548290-4 7.048850-4 9.349630-4 1.310960-3 1.762040-3 2.258760-3332533 16 2.801620-3 3.416040-3 4.153410-3 5.097380-3 6.377610-3 8.194880-3332533 16 2.503660-2 1.710970-2 1.053310-2 5.933080-3 3.031140-3 1.367320-3332533 16 5.445340-4 2.757120-4 3.679670-4 6.958200-4 1.178950-3 1.766310-3332533 16 2.425790-3 3.137300-3 3.888340-3 4.670750-3 5.478140-3 6.303300-3332533 16 1.310600-2 9.036090-3 5.666770-3 3.182410-3 1.508880-3 4.972090-4332533 16 1.198130-6-9.989140-5 9.961100-5 5.258840-4 1.118440-3 1.823060-3332533 16 2.584600-3 3.338870-3 4.001790-3 4.453280-3 4.511120-3 6.875300-3332533 16 4.696000-3 2.853540-3 1.466050-3 5.292000-4-8.635390-6-2.104130-4332533 16 -1.367730-4 1.569210-4 6.190120-4 1.197530-3 1.834630-3 2.458670-3332533 16 2.972620-3 3.235810-3 3.033720-3 3.453490-3 2.266810-3 1.294430-3332533 16 5.856670-4 1.362000-4-8.005410-5-9.600310-5 5.413750-5 3.357960-4332533 16 7.122850-4 1.141420-3 1.569950-3 1.925150-3 2.101170-3 1.936260-3332533 16 1.633900-3 1.073760-3 6.359020-4 3.318380-4 1.553350-4 9.295540-5332533 16 1.286100-4 2.450480-4 4.234780-4 6.421240-4 8.735030-4 1.080130-3332533 16 1.207510-3 1.172240-3 8.558900-4 6.667390-4 5.164420-4 4.069720-4332533 16 3.365790-4 3.019830-4 2.993640-4 3.246310-4 3.733290-4 4.402050-4332533 16 5.184400-4 5.983440-4 6.651090-4 6.747560-4 6.610070-4 6.283810-4332533 16 5.803990-4 5.209130-4 4.541900-4 3.852600-4 3.205060-4 2.685700-4332533 16 2.417760-4 2.584150-4 3.465920-4 7.554750-4 7.992770-4 7.962510-4332533 16 7.522190-4 6.741240-4 5.700760-4 4.499610-4 3.267030-4 2.183230-4332533 16 1.513670-4 1.667090-4 9.133160-4 9.708800-4 9.754440-4 9.318050-4332533 16 8.455750-4 7.232610-4 5.729600-4 4.056160-4 2.372370-4 9.272540-5332533 16 1.098580-3 1.177030-3 1.204940-3 1.180740-3 1.101790-3 9.636420-4332533 16 7.593550-4 4.784870-4 1.055990-4 1.346410-3 1.473240-3 1.545330-3332533 16 1.546490-3 1.454020-3 1.234990-3 8.406180-4 1.969000-4 1.715590-3332533 16 1.907030-3 2.015320-3 1.996100-3 1.785820-3 1.290460-3 3.668920-4332533 16 2.227810-3 2.458970-3 2.534180-3 2.358840-3 1.793140-3 6.238740-4332533 16 2.814850-3 2.998200-3 2.888540-3 2.307930-3 9.846970-4 3.293890-3332533 16 3.287670-3 2.782370-3 1.476680-3 3.433180-3 3.143590-3 2.141300-3332533 16 3.284480-3 3.040640-3 4.268410-3 332533 16 332533 0 3325 0 0 0 0 0 3.90890E+4 8.81421E+1 0 0 34 103925 1451 0.0 0.0 0 0 0 63925 1451 1.00000E+0 2.00000E+7 0 0 10 20023925 1451 3.00000E+2 0.0 1 0 91 33925 1451 39-Y - 89 SRI EVAL-JAN96 N.ODANO (SHIP RES. INST.) 3925 1451 DIST-Feb2004 3925 1451 ----IRDF-2002 MATERIAL 3925 3925 1451 -----INCIDENT NEUTRON DATA 3925 1451 ------ENDF-6 FORMAT 3925 1451 *************************************************************** 3925 1451 39- Y- 89 SRI EVAL-JAN96 N.ODANO (SHIP RES. INST.) 3925 1451 DIST-JUL98 3925 1451 ----JENDL/D-99 MATERIAL 3925 3925 1451 *************************************************************** 3925 1451 HISTORY 3925 1451 96-01 EVALUATION FOR JENDL DOSIMETRY FILE VERSION 2 WAS MADE BY 3925 1451 N.ODANO (SHIP RES. INST.). 3925 1451 97-09 COMPILED TO JENDL DOSIMETRY FILE 99. 3925 1451 3925 1451 ==== POINT-WISE DATA FILE ==== 3925 1451 3925 1451 3925 1451 Y-89 (N,2N) Y-88 (HALF-LIFE = 106.65D) 3925 1451 3925 1451 MF=1 GENERAL INFORMATION 3925 1451 MT=451 DESCRIPTIVE DATA AND DICTIONARY 3925 1451 3925 1451 MF=2 RESONANCE PARAMETERS 3925 1451 MT=151 PARAMETERS 3925 1451 ONLY SPIN AND SCATTERING RADIUS ARE GIVEN. 3925 1451 3925 1451 MF=3 NEUTRON CROSS SECTIONS 3925 1451 MT=16 (N,2N) CROSS SECTION 3925 1451 EXPERIMENTAL DATA/1-17/ IN NESTOR-2/18/ WERE TAKEN FOR 3925 1451 THE EVALUATION USING GMA CODE/19/. 3925 1451 3925 1451 MF=33 COVARIANCES OF NEUTRON CROSS SECTIONS 3925 1451 MT=16 GENERATED USING THE GMA CODE. 3925 1451 3925 1451 REFERENCES 3925 1451 1) O.M.HUDSON, JR.+ : BULLETIN OF THE AMERICAN PHYSICAL SOCIETY, 3925 1451 6, 506(E9) (1961). 3925 1451 2) R.RIEDER+ : EANDC(E)-49L, 83 (1965). 3925 1451 3) D.G.VALLIS : AWRE-O-76/66 (1966). 3925 1451 4) J.CSIKAI+ : ACTA PHYSICA HANGARIA, 23, 87 (1967). 3925 1451 5) D.S.MATHER+ : AWRE-O-47/69 (1969). 3925 1451 6) D.R.NETHAWAY : NUCL. PHYS. A, 190, 635 (1972). 3925 1451 7) S.M.QAIM+ : EUR-5182E, 939 (1974). 3925 1451 8) B.P.BAYHURST+ : PHYS. REV. C, 12, 451 (1975). 3925 1451 9) S.K.GHORAI+ : NUCL. PHYS. A, 266, 53 (1976). 3925 1451 10) M.BORMANN+ : Z. PHYS. A, 277, 203 (1976). 3925 1451 11) L.R.VEESER+ : PHYS. REV. C, 16, 1792 (1977). 3925 1451 12) J.FREHAUT+ : PROC. SYMP. ON NEUTRON CROSS SECTIONS FROM 10-50 3925 1451 MEV, UPTON, USA, 12-14, 1980, P.399 (1980). 3925 1451 13) J.LAUREC+ : CEA-R-5109 (1981). 3925 1451 14) P.RAICS+ : ATOMKI KOZLEMENYEK, 23, 45 (1981). 3925 1451 15) L.R.GREENWOOD : ASTM-STP-956, 743 (1987). 3925 1451 16) HUANG JIAN-ZHOU+ : INDC(CPR)-16 (189). 3925 1451 17) C.KONNO+ : JAERI 1329 (1993). 3925 1451 18) T.NAKAGAWA : THE JAERI NUCLEAR DATA CENTER, UNPUBLISHED. 3925 1451 19) W.P.POENITZ : PROC. CONF. NUCLEAR DATA EVALUATION METHODS 3925 1451 AND PROCEDURES, BROOKHAVEN NATIONAL LAB. 1980, BNL-NCS- 3925 1451 51363, P.249 (1981). 3925 1451 *************************************************************** 3925 1451 The Q values and threshold energies were updated prior to pro- 3925 1451 cessing through the codes to comply with the values obtained 3925 1451 using the NNDC calculation program which is based on the 1995 3925 1451 Update to the Atomic mass Evaluation. 3925 1451 *************************************************************** 3925 1451 3925 1451 ***************** Program LINEAR (VERSION 2002-1) ***************3925 1451 For All Data Greater than 1.0000E-10 barns in Absolute Value 3925 1451 Data Linearized to Within an Accuracy of .100000000 per-cent 3925 1451 ***************** Program SIGMA1 (VERSION 2002-1) ***************3925 1451 Data Doppler Broadened to 300.000000 Kelvin 3925 1451 for All Data Greater than 1.0000E-10 barns in Absolute Value 3925 1451 Data Linearized to Within an Accuracy pf .100000000 per-cent 3925 1451 ***************** Program FIXUP (Version 2002-1) ****************3925 1451 Corrected ZA/AWR in All Sections-----------------------------Yes 3925 1451 Corrected Thresholds-----------------------------------------Yes 3925 1451 Extended Cross Sections to 20 MeV----------------------------No 3925 1451 Allow Cross Section Deletion---------------------------------No 3925 1451 Allow Cross Section Reconstruction---------------------------No 3925 1451 Make All Cross Sections Non-Negative-------------------------Yes 3925 1451 Delete Energies Not in Ascending Order-----------------------Yes 3925 1451 Deleted Duplicate Points-------------------------------------Yes 3925 1451 Check for Ascending MAT/MF/MT Order--------------------------Yes 3925 1451 Check for Legal MF/MT Numbers--------------------------------Yes 3925 1451 Allow Creation of Missing Sections---------------------------No 3925 1451 Allow Insertion of Energy Points-----------------------------No 3925 1451 Create Uniform Energy Grid-----------------------------------No 3925 1451 Delete Section if Cross Section =0 at All Energies-----------Yes 3925 1451 ***************** Program GROUPIE (VERSION 2002-1) **************3925 1451 Unshielded Group Averages Using 640 Groups 3925 1451 Weighting Spectrum: Flat (Constant) Spectrum 3925 1451 1 451 98 03925 1451 3 16 32 03925 1451 33 16 23 03925 1451 3925 1 0 3925 0 0 3.90890E+4 8.81421E+1 0 0 0 03925 3 16 -1.14760E+7-1.14760E+7 0 0 1 853925 3 16 85 1 3925 3 16 11600000.0 .008985210 11700000.0 .030140246 11800000.0 .0559560003925 3 16 11900000.0 .082542000 12000000.0 .109128000 12100000.0 .1357140003925 3 16 12200000.0 .163801750 12300000.0 .200900000 12400000.0 .2395000003925 3 16 12500000.0 .278100000 12600000.0 .316700000 12700000.0 .3553000003925 3 16 12800000.0 .393900000 12900000.0 .432500000 13000000.0 .4743200003925 3 16 13100000.0 .519360000 13200000.0 .564400000 13300000.0 .6094400003925 3 16 13400000.0 .654480000 13500000.0 .692660000 13600000.0 .7239800003925 3 16 13700000.0 .755300000 13800000.0 .786620000 13900000.0 .8179400003925 3 16 14000000.0 .847270000 14100000.0 .874610000 14200000.0 .9019500003925 3 16 14300000.0 .929290000 14400000.0 .956630000 14500000.0 .9775700003925 3 16 14600000.0 .992110000 14700000.0 1.00665000 14800000.0 1.021190003925 3 16 14900000.0 1.03573000 15000000.0 1.04830000 15100000.0 1.058900003925 3 16 15200000.0 1.06950000 15300000.0 1.08010000 15400000.0 1.090700003925 3 16 15500000.0 1.10190000 15600000.0 1.11370000 15700000.0 1.125500003925 3 16 15800000.0 1.13730000 15900000.0 1.14910000 16000000.0 1.158350003925 3 16 16100000.0 1.16505000 16200000.0 1.17175000 16300000.0 1.178450003925 3 16 16400000.0 1.18515000 16500000.0 1.19185000 16600000.0 1.198550003925 3 16 16700000.0 1.20525000 16800000.0 1.21195000 16900000.0 1.218650003925 3 16 17000000.0 1.22130000 17100000.0 1.21990000 17200000.0 1.218500003925 3 16 17300000.0 1.21710000 17400000.0 1.21570000 17500000.0 1.214300003925 3 16 17600000.0 1.21290000 17700000.0 1.21150000 17800000.0 1.210100003925 3 16 17900000.0 1.20870000 18000000.0 1.20790000 18100000.0 1.207700003925 3 16 18200000.0 1.20750000 18300000.0 1.20730000 18400000.0 1.207100003925 3 16 18500000.0 1.20690000 18600000.0 1.20670000 18700000.0 1.206500003925 3 16 18800000.0 1.20630000 18900000.0 1.20610000 19000000.0 1.205900003925 3 16 19100000.0 1.20570000 19200000.0 1.20550000 19300000.0 1.205300003925 3 16 19400000.0 1.20510000 19500000.0 1.20490000 19600000.0 1.204700003925 3 16 19700000.0 1.20450000 19800000.0 1.20430000 19900000.0 1.204100003925 3 16 20000000.0 0.0 3925 3 16 3925 3 0 3925 0 0 3.90890E+4 8.81421E+1 0 0 0 1392533 16 0.000000+0 0.000000+0 0 16 0 1392533 16 0.000000+0 0.000000+0 1 5 120 15392533 16 1.000000-5 1.160000+7 1.167980+7 1.200000+7 1.262500+7 1.325000+7392533 16 1.375000+7 1.425000+7 1.475000+7 1.525000+7 1.575000+7 1.650000+7392533 16 1.750000+7 1.900000+7 2.000000+7 0.000000+0 0.000000+0 0.000000+0392533 16 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0392533 16 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0 2.614690-2392533 16 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0392533 16 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0392533 16 2.614690-2 2.351680-4 2.278630-4 9.042900-5 9.259900-5 8.525630-5392533 16 8.299090-5 7.707100-5 4.918100-5 5.777220-5 5.017710-5 4.334530-5392533 16 2.515020-3 1.482340-4 7.056710-5 6.952990-5 6.360270-5 6.802450-5392533 16 5.867100-5 4.991930-5 5.263290-5 5.027740-5 3.826150-5 1.181300-3392533 16 1.041660-4 9.116020-5 1.023620-4 8.883020-5 1.176510-4 7.032450-5392533 16 8.058560-5 7.055540-5 5.527930-5 3.254420-4 1.149980-4 1.205660-4392533 16 1.177200-4 1.281940-4 7.931010-5 9.144310-5 8.052500-5 6.472540-5392533 16 2.271050-4 1.022200-4 1.102170-4 9.664150-5 6.583630-5 8.648390-5392533 16 7.769970-5 7.830730-5 2.030630-4 1.065710-4 1.315170-4 8.313670-5392533 16 9.450610-5 8.571720-5 6.963890-5 3.359890-4 1.032510-4 6.690090-5392533 16 7.735940-5 6.913090-5 5.707270-5 1.877490-3 1.006380-4 1.132040-4392533 16 9.722290-5 7.147720-5 7.645230-4 9.817050-5 1.108810-4 1.009150-4392533 16 4.986290-4 1.343200-4 1.551700-4 5.697770-4 1.678400-4 4.251840-4392533 16 392533 0 3925 0 0 0 0 0 4.00900E+4 8.91324E+1 0 0 34 104025 1451 0.0 0.0 0 0 0 64025 1451 1.00000E+0 2.00000E+7 0 0 10 20024025 1451 3.00000E+2 0.0 1 0 40 34025 1451 40-Zr- 90 IRK-VIENNA EVAL-APR90 4025 1451 DIST-Feb2004 4025 1451 ----IRDF-2002 MATERIAL 4025 4025 1451 -----INCIDENT NEUTRON DATA 4025 1451 ------ENDF-6 FORMAT 4025 1451 *****************************************************************4025 1451 40-ZR- 90 IRK-VIENNA EVAL-APR90 4025 1451 DIST-JUN90 4025 1451 IRK-EVAL.NLIB 25 4025 4025 1451 *****************************************************************4025 1451 The Q values and threshold energies were updated prior to pro- 4025 1451 cessing through the codes to comply with the values obtained 4025 1451 using the NNDC calculation program which is based on the 1995 4025 1451 Update to the Atomic mass Evaluation. 4025 1451 *****************************************************************4025 1451 ***************** Program LINEAR (VERSION 2002-1) ***************4025 1451 For All Data Greater than 1.0000E-10 barns in Absolute Value 4025 1451 Data Linearized to Within an Accuracy of .100000000 per-cent 4025 1451 ***************** Program SIGMA1 (VERSION 2002-1) ***************4025 1451 Data Doppler Broadened to 300.000000 Kelvin 4025 1451 for All Data Greater than 1.0000E-10 barns in Absolute Value 4025 1451 Data Linearized to Within an Accuracy pf .100000000 per-cent 4025 1451 ***************** Program FIXUP (Version 2002-1) ****************4025 1451 Corrected ZA/AWR in All Sections-----------------------------Yes 4025 1451 Corrected Thresholds-----------------------------------------Yes 4025 1451 Extended Cross Sections to 20 MeV----------------------------No 4025 1451 Allow Cross Section Deletion---------------------------------No 4025 1451 Allow Cross Section Reconstruction---------------------------No 4025 1451 Make All Cross Sections Non-Negative-------------------------Yes 4025 1451 Delete Energies Not in Ascending Order-----------------------Yes 4025 1451 Deleted Duplicate Points-------------------------------------Yes 4025 1451 Check for Ascending MAT/MF/MT Order--------------------------Yes 4025 1451 Check for Legal MF/MT Numbers--------------------------------Yes 4025 1451 Allow Creation of Missing Sections---------------------------No 4025 1451 Allow Insertion of Energy Points-----------------------------No 4025 1451 Create Uniform Energy Grid-----------------------------------No 4025 1451 Delete Section if Cross Section =0 at All Energies-----------Yes 4025 1451 ***************** Program GROUPIE (VERSION 2002-1) **************4025 1451 Unshielded Group Averages Using 640 Groups 4025 1451 Weighting Spectrum: Flat (Constant) Spectrum 4025 1451 1 451 47 14025 1451 3 16 30 14025 1451 33 16 46 14025 1451 4025 1 0 4025 0 0 4.00900E+4 8.91324E+1 0 0 0 04025 3 16 -1.19700E+7-1.19700E+7 0 0 1 804025 3 16 80 1 4025 3 16 12100000.0 .005470824 12200000.0 .017333776 12300000.0 .0342925004025 3 16 12400000.0 .056337500 12500000.0 .081985000 12600000.0 .1112350004025 3 16 12700000.0 .144872500 12800000.0 .182897500 12900000.0 .2225200004025 3 16 13000000.0 .263740000 13100000.0 .303092500 13200000.0 .3405775004025 3 16 13300000.0 .376913750 13400000.0 .412101250 13500000.0 .4472887504025 3 16 13600000.0 .482476250 13700000.0 .516977500 13800000.0 .5507925004025 3 16 13900000.0 .582510000 14000000.0 .612130000 14100000.0 .6424475004025 3 16 14200000.0 .673462500 14300000.0 .703267500 14400000.0 .7318625004025 3 16 14500000.0 .757600000 14600000.0 .780480000 14700000.0 .8009275004025 3 16 14800000.0 .818942500 14900000.0 .837701429 15000000.0 .8572042864025 3 16 15100000.0 .876707143 15200000.0 .895960893 15300000.0 .9137200004025 3 16 15400000.0 .931230000 15500000.0 .948740000 15600000.0 .9662500004025 3 16 15700000.0 .983260583 15800000.0 .997274667 15900000.0 1.010789334025 3 16 16000000.0 1.02430400 16100000.0 1.03781867 16200000.0 1.051333334025 3 16 16300000.0 1.06484800 16400000.0 1.07836267 16500000.0 1.088679004025 3 16 16600000.0 1.09579700 16700000.0 1.10291500 16800000.0 1.110033004025 3 16 16900000.0 1.11715100 17000000.0 1.12426900 17100000.0 1.131387004025 3 16 17200000.0 1.13850500 17300000.0 1.14562300 17400000.0 1.152741004025 3 16 17500000.0 1.15859150 17600000.0 1.16317450 17700000.0 1.167757504025 3 16 17800000.0 1.17234050 17900000.0 1.17692350 18000000.0 1.181506504025 3 16 18100000.0 1.18608950 18200000.0 1.19067250 18300000.0 1.195255504025 3 16 18400000.0 1.19983850 18500000.0 1.20193100 18600000.0 1.201533004025 3 16 18700000.0 1.20113500 18800000.0 1.20073700 18900000.0 1.200339004025 3 16 19000000.0 1.19994100 19100000.0 1.19954300 19200000.0 1.199145004025 3 16 19300000.0 1.19874700 19400000.0 1.19834900 19500000.0 1.197951004025 3 16 19600000.0 1.19755300 19700000.0 1.19715500 19800000.0 1.196757004025 3 16 19900000.0 1.19635900 20000000.0 0.0 4025 3 16 4025 3 0 4025 0 0 4.00900E+4 8.91324E+1 0 0 0 1402533 16 0.000000+0 0.000000+0 0 16 0 1402533 16 0.000000+0 0.000000+0 1 5 253 22402533 16 1.000000-5 1.210000+7 1.240000+7 1.260000+7 1.280000+7 1.300000+7402533 16 1.320000+7 1.340000+7 1.360000+7 1.380000+7 1.400000+7 1.420000+7402533 16 1.440000+7 1.460000+7 1.480000+7 1.500000+7 1.550000+7 1.600000+7402533 16 1.700000+7 1.800000+7 1.900000+7 2.000000+7 0.000000+0 0.000000+0402533 16 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0402533 16 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0402533 16 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0402533 16 0.000000+0 1.159000-2 4.915000-3 2.591000-3 2.281000-3 3.480000-3402533 16 6.788000-4 1.381000-4 1.958000-4 5.803000-5 5.767000-5 2.137000-4402533 16 6.431000-5 5.829000-5 5.247000-5 1.721000-3 1.796000-3 1.230000-3402533 16 1.225000-3 1.489000-3 1.297000-3 4.873000-3 2.573000-3 2.266000-3402533 16 1.893000-3 6.741000-4 1.371000-4 1.066000-4 5.763000-5 5.727000-5402533 16 1.392000-4 6.387000-5 5.789000-5 5.211000-5 1.130000-3 1.177000-3402533 16 7.987000-4 7.954000-4 9.667000-4 7.058000-4 8.925000-3 4.716000-3402533 16 0.000000+0 1.403000-3 2.854000-4 0.000000+0 1.199000-4 1.192000-4402533 16 1.055000-4 1.329000-4 1.205000-4 1.085000-4 8.939000-4 9.207000-4402533 16 5.961000-4 5.941000-4 7.214000-4 0.000000+0 6.922000-3 0.000000+0402533 16 1.236000-3 2.513000-4 0.000000+0 1.056000-4 1.050000-4 9.288000-5402533 16 1.171000-4 1.061000-4 9.551000-5 7.873000-4 8.108000-4 5.250000-4402533 16 5.232000-4 6.353000-4 0.000000+0 2.527000-3 0.000000+0 0.000000+0402533 16 1.010000-4 0.000000+0 0.000000+0 8.392000-5 0.000000+0 0.000000+0402533 16 0.000000+0 6.644000-4 6.964000-4 4.856000-4 4.834000-4 5.878000-4402533 16 6.691000-4 1.279000-3 1.354000-4 3.023000-5 5.849000-5 3.123000-5402533 16 5.003000-5 5.378000-5 5.446000-5 5.173000-5 2.342000-4 2.412000-4402533 16 1.562000-4 1.557000-4 1.890000-4 0.000000+0 3.077000-4 8.119000-5402533 16 7.922000-5 7.440000-5 7.199000-5 5.936000-5 7.322000-5 7.150000-5402533 16 5.648000-5 5.854000-5 3.739000-5 3.784000-5 4.510000-5 0.000000+0402533 16 1.171000-4 4.644000-5 4.378000-5 4.618000-5 3.271000-5 4.260000-5402533 16 4.095000-5 3.739000-5 3.920000-5 2.733000-5 2.721000-5 3.308000-5402533 16 3.766000-5 9.605000-5 4.489000-5 4.244000-5 3.238000-5 4.414000-5402533 16 3.926000-5 2.002000-5 2.062000-5 1.335000-5 1.331000-5 1.616000-5402533 16 0.000000+0 1.114000-4 4.042000-5 3.070000-5 4.347000-5 3.718000-5402533 16 1.990000-5 2.050000-5 2.254000-5 2.203000-5 2.717000-5 1.259000-5402533 16 8.344000-5 2.932000-5 3.968000-5 3.557000-5 5.268000-5 5.499000-5402533 16 3.700000-5 3.711000-5 4.471000-5 3.129000-5 6.920000-5 3.915000-5402533 16 3.852000-5 2.559000-5 2.649000-5 1.696000-5 1.712000-5 2.046000-5402533 16 0.000000+0 1.115000-4 4.736000-5 2.011000-5 2.072000-5 1.341000-5402533 16 1.337000-5 1.623000-5 0.000000+0 1.118000-4 2.243000-5 2.328000-5402533 16 1.482000-5 1.505000-5 1.786000-5 0.000000+0 6.550000-4 4.446000-4402533 16 2.989000-4 2.997000-4 3.613000-4 2.477000-4 6.887000-4 3.120000-4402533 16 3.128000-4 3.770000-4 2.596000-4 4.461000-4 3.007000-4 2.902000-4402533 16 3.123000-4 4.453000-4 2.888000-4 3.121000-4 5.414000-4 2.681000-4402533 16 6.335000-4 402533 16 402533 0 4025 0 0 0 0 0 4.10930E+4 9.21083E+1 0 0 34 104125 1451 0.0 0.0 0 0 0 64125 1451 1.00000E+0 2.00000E+7 0 0 10 20024125 1451 3.00000E+2 0.0 1 0 408 94125 1451 41-Nb- 93 FEI/LANL EVAL-Aug96 Zolotarev et al.,Chadwick et al. 4125 1451 DIST-Feb2004 4125 1451 ----IRDF-2002 MATERIAL 4125 4125 1451 -----INCIDENT NEUTRON DATA 4125 1451 ------ENDF-6 FORMAT 4125 1451 ***************************************************************** 4125 1451 -----RUSSIAN DOSIMETRY FILE RRDF-98 Rev.2 4125 1451 ----- MF=3 MT=16 ----- 4125 1451 ----- MF=3 MT=51 ----- 4125 1451 For the IRDF-2002 file these two reactions were converted at 4125 1451 IAEA/NDS. 4125 1451 The reaction MF/MT=3/16 was converted to MF/MT=10/16 4125 1451 The reaction MF/MT=3/51 was converted to MF/MT=10/ 4 4125 1451 The corresponding co-variance files were also converted 4125 1451 The reaction MF/MT=33/16 was converted to MF/MT=40/16 4125 1451 The reaction MF/MT=33/51 was converted to MF/MT=40/ 4 4125 1451 ***************************************************************** 4125 1451 41-NB-93 FEI EVAL-Aug96 K.Zolotarev and G.Badikov 4125 1451 DIST-Aug96 4125 1451 ---BROND-3 MATERIAL 4112 4125 1451 ----INCIDENT NEUTRON DATA 4125 1451 -----ENDF-6 FORMAT 4125 1451 AUTHORS OF EVALUATION: K.Zolotarev and S.Badikov 4125 1451 ***************************************************************** 4125 1451 41-NB- 93 LANL,ANL EVAL-DEC97 M.CHADWICK,P.YOUNG,D.L.SMITH 4125 1451 Ch97,Ch99 DIST-AUG99 REV2- 19990803 4125 1451 ----ENDF/B-VI MATERIAL 4125 REVISION 2 4125 1451 *****************EXTRACT FOR SPECIAL PURPOSE FILE*****************4125 1451 DOSIMETRY 4125 1451 ******************************************************************4125 1451 **************************************************************** 4125 1451 ******** Start of (N,2N),(N,N') bibliographical component ******* 4125 1451 ***************************************************************** 4125 1451 4125 1451 ----- MF=3 MT=16 ----- 4125 1451 4125 1451 The evaluation Nb93(n,2n)Nb92m excitation function from 4125 1451 9.20 MeV to 20 MeV has been carried out within the framework of 4125 1451 generalized least squares method , rational function was used as 4125 1451 model function [1].The experimental data base covers the results 4125 1451 of measurements [2-37]. 4125 1451 Below 9.2 MeV the evaluation has been carried out by 4125 1451 means of linear interpolation. Experimental data [38-43] were 4125 1451 rejectected due to large discrepancy with the main bulk of the 4125 1451 experimental data. 4125 1451 The evaluated Nb93(n,2n)Nb92m excitation function averaged 4125 1451 on U-235 neutron fission spectrum [44] and Cf-252 spontaneous 4125 1451 fission neutron spectrum [45] gives the next values : 4125 1451 4125 1451 --------------------------------------------------------------- 4125 1451 TYPE OF SPECTRUM I ,MB (calc.) I , MB (measured) 4125 1451 ----------------------I-----------------I---------------------- 4125 1451 I I 4125 1451 U-235 neutron fission I 0.4416 I 0.4796+-0.0293 [46] 4125 1451 I I 4125 1451 ----------------------I-----------------I---------------------- 4125 1451 I I 4125 1451 CF-252 spont. fission I 0.7701 I 0.749 +- 0.038 [65] 4125 1451 I I 4125 1451 4125 1451 4125 1451 ----- MF=3 MT=51 ----- 4125 1451 4125 1451 Evaluation Nb93(n,n')Nb93m excitation function from 30 keV 4125 1451 to 20 MeV was performed by the same method. Evaluation is 4125 1451 based on : 4125 1451 from 170 keV to 680 keV on theoretical model calculations [47], 4125 1451 from 680 keV to 7.9 MeV on the experimental data [48-51] , 4125 1451 from 7.9 MeV to 20.0 MeV on theoretical model calculations [47] 4125 1451 and experimental data [52,53]. Below 170 keV the evaluation has 4125 1451 been carried out by means of linear interpolation. Experimental 4125 1451 data [48] and [52] were renormalized , respectively , to the new 4125 1451 monitor cross-sections data U235(n,f) [54] and Nb93(n,2n)Nb92m 4125 1451 [55]. 4125 1451 Re-evaluated experimental data of M.Wagner e.a. [49-50] were 4125 1451 taken from ref. [56]. 4125 1451 The evaluated Nb93(n,n')Nb93m excitation function averaged 4125 1451 on U-235 neutron fission spectrum [44] and Cf-252 spontaneous 4125 1451 fission neutron spectrum [45] gives the next values : 4125 1451 4125 1451 --------------------------------------------------------------- 4125 1451 TYPE OF SPECTRUM I ,MB (calc.) I , MB (measured) 4125 1451 ----------------------I-----------------I---------------------- 4125 1451 I I 150.0 +- 13.0 [57] 4125 1451 U-235 neutron fission I 143.47 I 147.0 +- 11.0 [58] 4125 1451 I I 146.2 +- 12.5 [59] 4125 1451 ----------------------I-----------------I---------------------- 4125 1451 I I 144.9 +- 4.8 [60] 4125 1451 CF-252 spont. fission I 146.04 I 144.0 +- 5.0 [61] 4125 1451 I I 150.6 +- 3.6 [62] 4125 1451 4125 1451 The integal cross-section data [57 - 59], measured in U-235 4125 1451 neutron fission spectrum, were renormalized to the Nb93m half- 4125 1451 life (16.1 +- 0.2) Year [63] , new values of K X-rays branching 4125 1451 and new monitor cross-section data. 4125 1451 The full description of the cross sections data evaluation 4125 1451 for the both reactions is given in [64]. 4125 1451 4125 1451 References : 4125 1451 1. S.Badikov, K.Zolotarev Proc. of NEANSC Specialist's. Meeting 4125 1451 on Evaluation and Processing of Covariance Data, Oar Ridge , 4125 1451 USA, 1992, OECD, Paris, 1993, p.105 4125 1451 2. H.Vonach,H.Muenzer J,OSA, v.96, p.120, Mar.1959 4125 1451 3. H.A.Tewes e.a. Report UCRL-6028-T, June 1960 4125 1451 4. R.J.Prestwood,B.P.Bayhurst Phys.Rev., v.121, p.1438,Mar.1961 4125 1451 5. V.L.Glagolev,P.A.Jampol'skij J,JET, v.13, p.520, 1961 4125 1451 6. D.G.Vallis Report AWRE-O-76/66, Dec.1966 4125 1451 7. L.Husain e.a. Phys.Rev./C, v.1, p.1233, Apr.1970 4125 1451 12. A.Paulsen, R.Widera Z.Physik, v.238, p.23, Sep.1970 4125 1451 13. M.Bormann e.a. Progress Rep. EANDC(E)-127, p.38, Mar.1970 4125 1451 14. S.M.Qaim e.a. Proc. of Conf.on Chemical Nuclear Data, Measur.4125 1451 and Applicat., Univ.of Kent, Canterbury, 20-22 Sep.1971,p.121 4125 1451 15. D.R.Nethaway Nucl.Phys./A, v.190, p.635, 1972 4125 1451 16. E.Holub,N.Cindro J.of Phys./G, v.2, p.405, June 1976 4125 1451 17. D.R.Nethaway J. of Inorg. and Nucl.Chem., v.40, p.1285, 1978 4125 1451 18. C.G.Hudson e.a. J.Annals of Nucl.Energy, v.5, p.589, 1978 4125 1451 19. J.Laurec e.a. Report CEA-R-5109, June 1981 4125 1451 20. T.B.Ryves,P.Kolkowski J.Phys.G, v.7, n.4, p.529, 1981 4125 1451 21. R.C.Harper,W.L.Alford J,JP/G, v.8, p.153, Jan.1982 4125 1451 22. J.Csikai Proc. of Int.Conf. Nuclear Data for Science and 4125 1451 Technology, Antverp., 6-10 Sept.1982, p.414 4125 1451 23. A.Chiadli,G.Paic Progress Report MOH-5, p.13, 1982 4125 1451 24. S.Daroczy e.a. Conf.,83KIEV, v.3, p.191, Nov.1983 4125 1451 25. Y.Ikeda e.a EXFOR 21945, 1984 4125 1451 26. Lu Han-Lin e.a. Chinese J.Nucl.Phys., v.6,n.1,p.76, Feb.1984 4125 1451 27. I.Garlea e.a. J,RRP, v.29, p.421, 1984 4125 1451 28. I.Garlea e.a. Proc.of the 14-th Intern. Symp.on Nucl.Phys., 4125 1451 Gaussig,GDR,19-23 Nov.1984, ZFK-562, p.126, 1985 4125 1451 29. Y.S.Kim e.a. J.of the Korean Nucl.Soc., v.18, n.2,p.92,1986 4125 1451 30. Y.Ikeda e.a. Report JAERI-1312, 1988 4125 1451 31. K.Kobayashi,I.Kimura Proc.of Int. Conf.on Nuclear Data for 4125 1451 Science and Technology, Mito, Japan 30 May - 3 June 1988 4125 1451 32. R.Woelfle e.a. Appl.Radiat.and Isotopes, v.39, p.407, 1988 4125 1451 33. Lu Hanlin e.a. Report INDC(CPR)-16, Aug.1989 4125 1451 34. Wang Xiuyuan e.a. EXFOR 30935, 1989 4125 1451 35. D.C.Santry,R.D.Werner Can.J.Phys., v.68, p.582, 1990 4125 1451 36. Y.Ikeda e.a. Progress Report INDC(JPN)-162/U, p.16, Aug.1992 4125 1451 37 D.L.Smith e.a. Proc.Int.Conf. Nuclear Data for Science and 4125 1451 Technology, Julich, FRG, 13-17 May 1991. Springer Verlag, 4125 1451 Berlin - Heidelberg, 1992, p.282-284 4125 1451 38. D.R.Koehler,W.L.Alford Report NP-11667, 1962 4125 1451 39. F.Strohal e.a. Nucl.Phys., v.30, p.49, Feb.1962 4125 1451 40. G.T.Western e.a. Report AFWL-TR-65-216, p.2, June 1966 4125 1451 41. W.D.Lu e.a. Phys.Rev.C, v.1, p.350, Jan.1970 4125 1451 42. A.Mannan,S.M.Qaim Phys.Rev./C, v.38, n.2, p.630, Aug.1988 4125 1451 43. S.I.Bhuiyan EXFOR 30936, 1989 4125 1451 44. L.W.Weston e.a. Evaluated Neutron Data for U-235, ENDF/B-VI 4125 1451 Library, MAT=9228, MF=5, MT=18, eval.apr. 1989 4125 1451 45. W.Mannhart IAEA-TECDOC-410, p.158, 1987 4125 1451 46. K.Kobayashi, T.Kobayashi Progress Report NEANDC(J)-155U, 4125 1451 P.52, 1990 4125 1451 47. B.Strohmaier Ann. nucl. Energy, v.16, n.9, p.461, 1989 4125 1451 48. D.B.Gayther e.a. Rep. AERE-R-12612, MAY 1987 4125 1451 49. M.Wagner e.a. J.Annals of Nucl.Energy, v.15, n.7, p.363, 1988 4125 1451 50. M.Wagner e.a. Proc.Int.Conf. Nucl.Data for Sci.Tec.,Mito, 4125 1451 Japan,1988,p.1049 4125 1451 51. M.Wagner e.a. Measurement of the activation cross section for 4125 1451 the reaction 93Nb(n,n')93mNb in the neutron energy range 4125 1451 6 - 9 MeV. Progress Report 1991 4125 1451 52. T.B.Ryves,P.Kolkowski J. of Physics G, v.7, n.4, p.529, 1981 4125 1451 53. Y.Ikeda e.a. Progress Report INDC(JPN)-162/U, p.16, Aug.1992 4125 1451 54. L.W.Weston e.a. Evaluated Neutron Data for U-235, ENDF/B-VI 4125 1451 Library, MAT=9228, MF=5, MT=18, eval.apr. 1989 4125 1451 55. M.Wagner e.a. Physics Data Nr.13-5, p.175, Karlsruhe 1990 4125 1451 56. M.Wagner e.a. Physics Data Nr.13-5, p.135, Karlsruhe 1990 4125 1451 57. K.Sakurai, I.Kondo Nucl.Instr.and Meth., v.187, p.649, 1981 4125 1451 58. K.Kobayashi, I.Kimura Rep. NEANDC(J)-61U, p.78, sep.1979 4125 1451 59. F.Hegedues Proc. of the 1 st ASTM-EURATOM Symposium on Reac- 4125 1451 tor Dosimetry, Petten 1975. EUR 5667e, Part 1, p.757, 1976 4125 1451 60. W.G.Alberts e.a. Proc. of the 6 th ASTM-EURATOM Symposium on 4125 1451 Reactor Dosimetry, ASTM-STP-1OO1, p.223, 1989 4125 1451 61. T.G.Williamson e.a. Proc. of the 6 th ASTM-EURATOM Symposium 4125 1451 on Reactor Dosimetry, ASTM-STP-1OO1, p.229, 1989 4125 1451 62. J.G.Williams e.a. Proc.of the 6 th ASTM-EURATOM Symposium on 4125 1451 Reactor Dosimetry, ASTM-STP-1OO1, p.235, 1989 4125 1451 63. U.Schotzig,H.Schrader R.PTB-RA-16/2, 2 ed.,Braunschweig,1986 4125 1451 64. S.Badikov, K.Zolotarev Proc.of the 8-th ASTM-EURATOM Sympos. 4125 1451 on Reactor Dosimetry, Vail, Colorado, USA,29 Aug.-3 Sep. 1993 4125 1451 65. W.Mannhart Private communication, 1990 4125 1451 ******************************************************************4125 1451 ******** End of (N,2N),(N,N') bibliographical component ******** 4125 1451 ***************************************************************** 4125 1451 ******************************************************************4125 1451 ******** Start of (N,G) bibliographical component ******** 4125 1451 ***************************************************************** 4125 1451 4125 1451 ENDF/B-VI MOD 3 Evaluation, December 1997, M.B. Chadwick and 4125 1451 P.G. Young (LANL) 4125 1451 4125 1451 Los Alamos LA150 Library, produced with FKK/GNASH/GSCAN code 4125 1451 in cooperation with ECN Petten. 4125 1451 4125 1451 This evaluation provides a complete representation of the 4125 1451 nuclear data needed for transport, damage, heating, radioactivity,4125 1451 and shielding applications over the incident proton energy range 4125 1451 from 1 to 150 MeV. The evaluation utilizes MF=6, MT=5 to 4125 1451 represent all reaction data. Production cross sections and 4125 1451 emission spectra are given for neutrons, protons, deuterons, 4125 1451 tritons, alpha particles, gamma rays, and all residual nuclides 4125 1451 produced (A>5) in the reaction chains. To summarize, the ENDF 4125 1451 sections with non-zero data above are: 4125 1451 4125 1451 MF=3 MT= 2 Integral of nuclear plus interference components 4125 1451 of the elastic scattering cross section 4125 1451 4125 1451 MT= 5 Sum of binary (p,n') and (p,x) reactions 4125 1451 4125 1451 MF=6 MT= 2 Elastic (p,p) angular distributions given as 4125 1451 ratios of the differential nuclear-plus- 4125 1451 interference to the integrated value. 4125 1451 4125 1451 MT= 5 Production cross sections and energy-angle 4125 1451 distributions for emission neutrons, protons, 4125 1451 deuterons, and alphas; and angle-integrated 4125 1451 spectra for gamma rays and residual nuclei that 4125 1451 are stable against particle emission 4125 1451 4125 1451 The evaluation is based on nuclear model calculations that have 4125 1451 been benchmarked to experimental data, especially for n +93Nb and 4125 1451 n +93Nb reactions [Ch98]. We use the GNASH code system [Yo92], 4125 1451 which utilizes Hauser-Feshbach statistical, preequilibrium and 4125 1451 direct-reaction theories. Spherical optical model calculations are4125 1451 used to obtain particle transmission coefficients for the Hauser- 4125 1451 Feshbach calculations, as well as for the elastic proton angular 4125 1451 distributions. 4125 1451 Cross sections and spectra for producing individual residual 4125 1451 nuclei are included for reactions. The energy-angle-correlations 4125 1451 for all outgoing particles are based on Kalbach systematics 4125 1451 [Ka88]. 4125 1451 A model was developed to calculate the energy distributions of 4125 1451 all recoil nuclei in the GNASH calculations [Ch96a]. The recoil 4125 1451 energy distributions are represented in the laboratory system in 4125 1451 MT=5, MF=6, and are given as isotropic in the lab system. All 4125 1451 other data in MT=5,MF=6 are given in the center-of-mass system. 4125 1451 This method of representation utilizes the LCT=3 option approved 4125 1451 at the November, 1996, CSEWG meeting. 4125 1451 Preequilibrium corrections were performed in the course of the 4125 1451 GNASH calculations using the exciton model of Kalbach [Ka77, 4125 1451 Ka85], validated by comparison with calculations using Feshbach, 4125 1451 Kerman, Koonin (FKK) theory [Ch93]. Discrete level data from 4125 1451 nuclear data sheets were matched to continuum level densities 4125 1451 using the formulation of Ignatyuk et al. [Ig75] and pairing and 4125 1451 shell parameters from the Cook [Co67] analysis. Neutron and 4125 1451 charged- particle transmission coefficients were obtained from the4125 1451 optical potentials, as discussed below. Gamma-ray transmission 4125 1451 coefficients were calculated using the Kopecky-Uhl model [Ko90]. 4125 1451 4125 1451 SPECIFIC INFORMATION CONCERNING THE 93Nb EVALUATION 4125 1451 4125 1451 The total neutron cross section was obtained from the Finlay[Fi93]4125 1451 measurements. 4125 1451 4125 1451 The following optical potentials were used in the GNASH 4125 1451 calculations. For incident neutrons, the Wilmore-Hodgson 4125 1451 potential was used below 15 MeV, and the Madland potential [Ma88] 4125 1451 was used at higher energies. For incident protons, the 4125 1451 Becchetti-Greenlees [Be69] potential was used up to 50 MeV, above 4125 1451 which the Madland potential [Ma88] was used. In both cases, the 4125 1451 matching energy between the potentials was chosen to result in 4125 1451 continuity of the reaction cross section. For protons at 50 MeV 4125 1451 the reaction cross section (and transmission coefficients) was 4125 1451 renormalized slightly to smoothen the transition between the 4125 1451 potentials. The Perey [Pe63] potential was used for indident 4125 1451 deuterons. For tritons, the Becchetti-Greenlees [Be71] was used 4125 1451 up to 80 MeV, above which the Watanabe potential was used. The 4125 1451 Moyen (McFadden Satchler) [Mc66] potential was used for alpha 4125 1451 particles over the whole energy range. 4125 1451 4125 1451 Direct inelastic scattering to low-lying states in Nb93 was 4125 1451 determined as follows. Coherent excitation of 2+ and 3- 4125 1451 vibrations were assumed to be fragmented over Nb93 states, after 4125 1451 coupling these excitations with the 4.5+ core. The magnitudes of 4125 1451 the deformation lengths of 2+ and 3- excitations was obtained by 4125 1451 fitting values of 34 and 46 mb respectively at 14 MeV, obtained 4125 1451 in ref. [Ch93] and accounting for measurements well. This 4125 1451 strength was then fragmented over Nb states. For the 3- 4125 1451 excitation, the 7 states are in the "continuum" region of the 4125 1451 GNASH calculation at approximately 2.5 MeV, with spins 1.5-,2.5-, 4125 1451 ..,7.5-. For the 2+, the 5 states (2.5+,3.5+,...6.5+) near 1 MeV 4125 1451 were assumed to be those whose inelastic cross section in the 4125 1451 existing ENDF <20 MeV file are significant (note that the ENDF 4125 1451 file below 20 MeV appears to incorporate inelastic information 4125 1451 only up to 5 MeV for many states, after which a value of zero at 4125 1451 20 MeV was inserted). 4125 1451 4125 1451 Experimental data is used to benchmark the calculations. For 4125 1451 incident neutrons, experimental neutron emission spectra data 4125 1451 exist at 20 and 26 MeV by Marcinkowski [Ma83]. For incident 4125 1451 protons, spectra data exist at 14 and 26 MeV by Watanabe et 4125 1451 al. [Wa97], and at 65 MeV by Sakai et al [Sa80]. Our evaluation 4125 1451 agrees reasonably well with these measurements. 4125 1451 4125 1451 **************************************************************** 4125 1451 4125 1451 REFERENCES 4125 1451 4125 1451 [Be69] F.D. Becchetti, Jr., and G.W. Greenlees, Phys.Rev. 182, 4125 1451 1190 (1969) 4125 1451 [Be71] F.D. Becchetti, Jr., and G.W. Greenlees in Polarization 4125 1451 Phenomena in Nuclear Reactions (Ed: H.H. Barschall and W. 4125 1451 Haeberli, The University of Wisconsin Press, 1971) p.682 4125 1451 [Ch93] M.B. Chadwick and P.G. Young, Phys.Rev. C 47, 2255 (1993) 4125 1451 [Ch96a] M.B. Chadwick, P.G. Young, R.E. MacFarlane, and A.J. 4125 1451 Koning, "High-Energy Nuclear Data Libraries for Accelerator- 4125 1451 Driven Technologies: Calculational Method for Heavy Recoils," 4125 1451 Proc. of 2nd Int. Conf. on Accelerator Driven Transmutation 4125 1451 Technology and Applications, Kalmar, Sweden, 3-7 June 1996. 4125 1451 [Ch98] M.B. Chadwick and P.G. Young, "Model Calculations of 4125 1451 n,p + 93Nb" in APT PROGRESS REPORT: 1 November 1997 - 1 January 4125 1451 1998, internal Los Alamos National Laboratory memo 4125 1451 January 1998 from R.E. MacFarlane to L. Waters. 4125 1451 [Ch99] M.B. Chadwick, P.G. Young, G.M. Hale et al., Los Alamos 4125 1451 National Laboratory report, LA-UR-99-1222 (1999) 4125 1451 [Co67] J.L. Cook, H. Ferguson, and A.R. DeL Musgrove, Aust.J. 4125 1451 Phys. 20, 477 (1967) 4125 1451 [Fi93] R. W. Finlay, W. P. Abfalterer, G. Fink et al., Phys. Rev 4125 1451 C 47, 237 (1993) 4125 1451 [Ig75] A.V. Ignatyuk, G.N. Smirenkin, and A.S. Tishin, Sov.J. 4125 1451 Nucl.Phys. 21, 255 (1975); translation of Yad.Fiz. 21, 485 4125 1451 (1975) 4125 1451 [Ka77] C. Kalbach, Z.Phys.A 283, 401 (1977) 4125 1451 [Ka85] C. Kalbach, Los Alamos National Laboratory report 4125 1451 LA-10248-MS (1985) 4125 1451 [Ka88] C. Kalbach, Phys.Rev.C 37, 2350 (1988); see also 4125 1451 C. Kalbach and F. M. Mann, Phys.Rev.C 23, 112 (1981) 4125 1451 [Ko90] J. Kopecky and M. Uhl, Phys.Rev.C 41, 1941 (1990) 4125 1451 [Lo74] J.M. Lohr and W. Haeberli, Nucl.Phys. A232, 381 (1974) 4125 1451 [Ma88] D.G. Madland, "Recent Results in the Development of a 4125 1451 Global Medium-Energy Nucleon-Nucleus Optical-Model Potential," 4125 1451 Proc. OECD/NEANDC Specialist's Mtg. on Preequilibrium Nuclear 4125 1451 Reactions, Semmering, Austria, 10-12 Feb. 1988, NEANDC-245 'U' 4125 1451 (1988). 4125 1451 [Ma83] A. Marcinkowski, R.W. Finlay, G. Randers-Pehrson et al., 4125 1451 Nucl.Phys. A402, 220 (1983) 4125 1451 [Mc66] L. McFadden and G. R. Satchler, Nucl. Phys. 84, 177 4125 1451 (1966). 4125 1451 [Pe63] C.M. Perey and F.G. Perey, Phys.Rev. 132, 755 (1963) 4125 1451 [Sa80] H. Sakai, K. Hosono, N. Matsuoka et al., Nucl.Phys. A344, 4125 1451 41 (1980) 4125 1451 [We96] H.P. Wellisch and D. Axen, Phys.Rev. C 54, 1329(1996) 4125 1451 [Wa97] Y. Watanabe, S. Yoshioka, M. Harada et al, Nuclear Data 4125 1451 for Science and Technology, Proc. Conf. Trieste, May, 1997, 4125 1451 G. Reffo, Ed. (Editrice Compositori, 1997) p.580 4125 1451 [Wi64] D. Wilmore and P.E. Hodgson, Nucl.Phys. 55, 673 (1964) 4125 1451 [Yo92] P.G. Young, E.D. Arthur, and M.B. Chadwick, report 4125 1451 LA-12343-MS (1992) 4125 1451 4125 1451 ***************************************************************** 4125 1451 4125 1451 ENDF/B-VI MOD 2 Revision, August 1991, NNDC 4125 1451 4125 1451 Only the section MOD numbers have been corrected in the 4125 1451 directory. 4125 1451 4125 1451 ***************************************************************** 4125 1451 4125 1451 ENDF/B-VI MOD 1 Evaluation, March 1990, A.B. Smith, D.L. Smith 4125 1451 L.P. Geraldo (ANL), and R. Howerton (LLNL) 4125 1451 4125 1451 Original evaluation fully documented in Smith et al. [1] 4125 1451 4125 1451 ---------------------------------------------------------------- 4125 1451 REFERENCES 4125 1451 4125 1451 [1.] A.B. Smith, D.L. Smith and R.J. Howerton, Argonne report 4125 1451 ANL/NDM-88 (1985) 4125 1451 [2.] D.L. Smith and L.P. Geraldo, Argonne report ANL/NDM-117 4125 1451 (1990) 4125 1451 4125 1451 ******************************************************************4125 1451 ******** End of (N,G) bibliographical component ******** 4125 1451 ***************************************************************** 4125 1451 The Q values and threshold energies were updated prior to pro- 4125 1451 cessing through the codes to comply with the values obtained 4125 1451 using the NNDC calculation program which is based on the 1995 4125 1451 Update to the Atomic mass Evaluation. 4125 1451 4125 1451 File 2 taken from ENDF/B-VI 4125 1451 ************************ C O N T E N T S *********************** 4125 1451 4125 1451 ***************** Program LINEAR (VERSION 2002-1) ***************4125 1451 For All Data Greater than 1.0000E-10 barns in Absolute Value 4125 1451 Data Linearized to Within an Accuracy of .100000000 per-cent 4125 1451 ***************** Program RECENT (VERSION 2002-1) ***************4125 1451 for All Data Greater than 1.0000E-10 barns in Absolute Value 4125 1451 Data Linearized to within an Accuracy of .100000000 per-cent 4125 1451 ***************** Program SIGMA1 (VERSION 2002-1) ***************4125 1451 Data Doppler Broadened to 300.000000 Kelvin 4125 1451 for All Data Greater than 1.0000E-10 barns in Absolute Value 4125 1451 Data Linearized to Within an Accuracy pf .100000000 per-cent 4125 1451 ***************** Program FIXUP (Version 2002-1) ****************4125 1451 Corrected ZA/AWR in All Sections-----------------------------Yes 4125 1451 Corrected Thresholds-----------------------------------------Yes 4125 1451 Extended Cross Sections to 20 MeV----------------------------No 4125 1451 Allow Cross Section Deletion---------------------------------No 4125 1451 Allow Cross Section Reconstruction---------------------------No 4125 1451 Make All Cross Sections Non-Negative-------------------------Yes 4125 1451 Delete Energies Not in Ascending Order-----------------------Yes 4125 1451 Deleted Duplicate Points-------------------------------------Yes 4125 1451 Check for Ascending MAT/MF/MT Order--------------------------Yes 4125 1451 Check for Legal MF/MT Numbers--------------------------------Yes 4125 1451 Allow Creation of Missing Sections---------------------------No 4125 1451 Allow Insertion of Energy Points-----------------------------No 4125 1451 Create Uniform Energy Grid-----------------------------------No 4125 1451 Delete Section if Cross Section =0 at All Energies-----------Yes 4125 1451 ***************** Program GROUPIE (VERSION 2002-1) **************4125 1451 Unshielded Group Averages Using 640 Groups 4125 1451 Weighting Spectrum: Flat (Constant) Spectrum 4125 1451 1 451 421 14125 1451 3 102 217 14125 1451 8 4 2 14125 1451 8 16 2 14125 1451 10 4 90 14125 1451 10 16 41 14125 1451 33 102 9 14125 1451 40 4 104 14125 1451 40 16 39 14125 1451 4125 1 0 4125 0 0 4.10930E+4 9.21083E+1 0 0 0 04125 3102 7.22747E+6 7.22747E+6 0 0 1 6414125 3102 641 1 4125 3102 .000100000 18.0213007 .000105000 17.6011498 .000110000 17.20200384125 3102 .000115000 16.8342057 .000120000 16.4042406 .000127500 15.92720694125 3102 .000135000 15.4913843 .000142500 15.0908610 .000150000 14.65737964125 3102 .000160000 14.2058190 .000170000 13.7949650 .000180000 13.41782814125 3102 .000190000 13.0669303 .000200000 12.7465657 .000210000 12.44515734125 3102 .000220000 12.1651963 .000230000 11.9050847 .000240000 11.59977434125 3102 .000255000 11.2642942 .000270000 11.0054074 .000280000 10.71720284125 3102 .000300000 10.3657565 .000320000 10.0475333 .000340000 9.756094094125 3102 .000360000 9.48766816 .000380000 9.24263675 .000400000 8.987176444125 3102 .000425000 8.72576587 .000450000 8.48693672 .000475000 8.267962614125 3102 .000500000 8.06202911 .000525000 7.87383007 .000550000 7.696048704125 3102 .000575000 7.53070746 .000600000 7.36124432 .000630000 7.186911624125 3102 .000660000 7.02739729 .000690000 6.87437954 .000720000 6.711564314125 3102 .000760000 6.53699045 .000800000 6.37490446 .000840000 6.226363634125 3102 .000880000 6.08518481 .000920000 5.95576328 .000960000 5.831819854125 3102 .001000000 5.70335266 .001050000 5.56924720 .001100000 5.443531344125 3102 .001150000 5.32768145 .001200000 5.19101925 .001275000 5.040835304125 3102 .001350000 4.90352207 .001425000 4.77537192 .001500000 4.639231264125 3102 .001600000 4.49688317 .001700000 4.36693655 .001800000 4.246754304125 3102 .001900000 4.13696432 .002000000 4.03517330 .002100000 3.939658444125 3102 .002200000 3.85218040 .002300000 3.76853368 .002400000 3.673114994125 3102 .002550000 3.56708007 .002700000 3.48413693 .002800000 3.394028254125 3102 .003000000 3.28311047 .003200000 3.18218089 .003400000 3.089889654125 3102 .003600000 3.00571857 .003800000 2.92800045 .004000000 2.846995734125 3102 .004250000 2.76475415 .004500000 2.68968605 .004750000 2.619526324125 3102 .005000000 2.55514441 .005250000 2.49546942 .005500000 2.439062994125 3102 .005750000 2.38736888 .006000000 2.33308297 .006300000 2.278836274125 3102 .006600000 2.22751075 .006900000 2.17992067 .007200000 2.128184914125 3102 .007600000 2.07287015 .008000000 2.02229362 .008400000 1.974836684125 3102 .008800000 1.93068724 .009200000 1.88965925 .009600000 1.850618324125 3102 .010000000 1.81018797 .010500000 1.76743310 .011000000 1.728284414125 3102 .011500000 1.69092729 .012000000 1.64810317 .012750000 1.600693434125 3102 .013500000 1.55678814 .014250000 1.51653646 .015000000 1.473662134125 3102 .016000001 1.42869772 .017000001 1.38751152 .017999999 1.349987664125 3102 .018999999 1.31541292 .020000000 1.28292398 .021000000 1.253192614125 3102 .022000000 1.22518032 .023000000 1.19908065 .024000000 1.168416244125 3102 .025500000 1.13534352 .027000001 1.10913958 .028000001 1.080676714125 3102 .029999999 1.04563566 .032000002 1.01370832 .034000002 .9845640514125 3102 .035999998 .957974892 .037999999 .933469317 .039999999 .9079185714125 3102 .042500000 .882001933 .045000002 .858314863 .047499999 .8362348154125 3102 .050000001 .815903693 .052499998 .797076973 .055000000 .7792451944125 3102 .057500001 .762920918 .059999999 .745757234 .063000001 .7285945024125 3102 .066000000 .712422526 .068999998 .697360082 .071999997 .6810306544125 3102 .075999998 .663545193 .079999998 .647586836 .083999999 .6326033124125 3102 .088000000 .618689959 .092000000 .605799476 .096000001 .5934841284125 3102 .100000001 .580740692 .104999997 .567238550 .109999999 .5548929024125 3102 .115000002 .543099907 .119999997 .529595519 .127499998 .5146728974125 3102 .135000005 .500837179 .142499998 .488141351 .150000006 .4746142334125 3102 .159999996 .460419968 .170000002 .447512045 .180000007 .4358671434125 3102 .189999998 .424682446 .200000003 .414537148 .209999993 .4051062834125 3102 .219999999 .396226892 .230000004 .388071824 .239999995 .3784338824125 3102 .254999995 .367847380 .270000011 .359748630 .280000001 .3506513334125 3102 .300000012 .339586358 .319999993 .329572955 .340000004 .3204548804125 3102 .360000014 .312085870 .379999995 .304512822 .400000006 .2963389924125 3102 .425000012 .288129965 .449999988 .280647711 .474999994 .2736699674125 3102 .500000000 .267249196 .524999976 .261324591 .550000012 .2557032224125 3102 .574999988 .250569355 .600000024 .245164713 .629999995 .2397563244125 3102 .660000026 .234629738 .689999998 .229956081 .720000029 .2249389214125 3102 .759999990 .219377641 .800000012 .214412077 .839999974 .2095743734125 3102 .879999995 .205287449 .920000017 .201117331 .959999979 .1972194084125 3102 1.00000000 .193144171 1.04999995 .188952517 1.10000002 .1849860024125 3102 1.14999998 .181411664 1.20000005 .177069478 1.27499998 .1722936074125 3102 1.35000002 .168067132 1.42499995 .163992345 1.50000000 .1598027644125 3102 1.60000002 .155279789 1.70000005 .151227126 1.79999995 .1473994314125 3102 1.89999998 .143982571 2.00000000 .140708632 2.09999990 .1377903334125 3102 2.20000005 .134930277 2.29999995 .132383342 2.40000010 .1293041244125 3102 2.54999995 .125954380 2.70000005 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***************************************************************** 4525 1451 45-Rh-103 FEI EVAL-Mar03 K.I.Zolotarev 4525 1451 DIST-Apr03 4525 1451 ----BROND-2 MATERIAL 4525 4525 1451 -----INCIDENT NEUTRON DATA 4525 1451 ------ENDF-6 FORMAT 4525 1451 ------Russian Reactor Dosimetry File RRDF-2002 4525 1451 ***************************************************************** 4525 1451 ----- MF=3 MT=51 ----- 4525 1451 For the IRDF-2002 file this reaction was converted at IAEA/NDS 4525 1451 The reaction MF/MT=3/51 was converted to MF/MT=10/ 4 4525 1451 The corresponding co-variance files were also converted 4525 1451 The reaction MF/MT=33/51 was converted to MF/MT=40/ 4 4525 1451 ***************************************************************** 4525 1451 Authors of evaluation: K.I.Zolotarev and P.K.Zolotarev 4525 1451 ***************************************************************** 4525 1451 MF=3 4525 1451 MT= 51 - Rh103(n,n')Rh103m reaction 4525 1451 ------------------------------------- 4525 1451 In the section MT= 51 are given evaluated cross section data 4525 1451 of neutron inelastic excitation of the 56.12-min isomeric state 4525 1451 in the rhodium-103 in the energy range from threshold to 20 MeV. 4525 1451 The isomeric level has energy 39.756 keV with spin and parity 4525 1451 7/2+ [1]. 4525 1451 Microscopic experimental data [4-13] were analyzed in the 4525 1451 process of preparation of input data base for the evaluation of 4525 1451 cross sections and their uncertainty for the Rh-103(n,n')Rh-103m 4525 1451 reaction. During this procedure all experimental data if it was 4525 1451 possible were corrected to the new recommended cross section data 4525 1451 for monitor reactions used in the measurements and to the new re- 4525 1451 commended decay data. Decay data for the rhodium-103m were taken 4525 1451 from ref. [2]. Recommended cross sections for the reaction 4525 1451 In-115(n,n')In-115m used as a monitor in the measurements [4], 4525 1451 [5] and [7] were taken from ref. [3]. 4525 1451 Special correction was done with experimental data [4], [6]. 4525 1451 Data of Cross [4] for the neutron energies 2.20 - 2.86 MeV 4525 1451 were renormalized to the measured by Paulsen et al. cross section 4525 1451 value (994.0+-45.7) mb at 2.60 MeV [8]. 4525 1451 Data of Santry and Butler [6] obtained in the measurements 4525 1451 with D(d,n)He3 source in the energy range 5.00-13.58 MeV were re- 4525 1451 normalized to the integral of Miach et al. experimental data [10] 4525 1451 in the energy interval 6.0-12.0 MeV, Fc= 0.90190 . Cross sections 4525 1451 measured by Santry and Butler with using Li7(p,n)Be7, T(p,n)He3 4525 1451 and T(d,n)He4 neutron sources were not corrected. 4525 1451 Excitation function for the Rh-103(n,n')Rh-103m reaction in 4525 1451 the energy region from threshold to 20 MeV was evaluated by means 4525 1451 of statistical analysis of experimental cross section data [4-10].4525 1451 Experimental cross section data [11-13] were rejected due to 4525 1451 their discrepancy with the main bulk of experimental data [4-10]. 4525 1451 In the rejected experiment [11] the cross section value was deter-4525 1451 mined only in a one energy point 14.20 MeV. 4525 1451 Statistical analysis of input cross section data was carried 4525 1451 out by means of PADE-2 code [14]. Rational function was used as 4525 1451 the model function [15]. 4525 1451 U-235 thermal fission [16] and Cf-252 spontaneous fission 4525 1451 neutron spectra [17] averaged cross sections calculated from the 4525 1451 the evaluated Rh-103(n,n')Rh-103m excitation function are the 4525 1451 following: 4525 1451 ÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÂÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÂÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄ 4525 1451 TYPE OF SPECTRUM ³ ,mb (calc.) ³ , mb (measured) 4525 1451 ÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÅÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÅÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄ 4525 1451 U-235 neutron fission ³ 715.85 ³ 702.16+-28.08 [18] 4525 1451 ³ ³ 721.16+-38.65 [19] 4525 1451 ³ ³ 670.73+-52.05 [20] 4525 1451 ÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÅÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÅÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄ 4525 1451 Cf-252 spont. fission ³ 724.83 ³ 620.82+-67.17 [21] 4525 1451 ³ ³ 813.29+-24.24 [22] 4525 1451 ³ ³ 813.22+-24.15 [22] 4525 1451 ÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÁÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÁÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄ 4525 1451 Integral experimental data [18-22] were corrected to the new 4525 1451 recommended cross sections for monitor reactions from ref. [23] 4525 1451 and [24]. 4525 1451 4525 1451 MT=33 4525 1451 MT= 51 -(n,n') cross section cov. matrix 4525 1451 ---------------------------------------- 4525 1451 Uncertainties in the evaluated excitation function for the 4525 1451 reaction Rh-103(n,n')Rh-103m are given in the form of relative 4525 1451 covariance matrix for the 35-neutron energy groups (LB=5). Cova- 4525 1451 riance matrix of uncertainties was calculated simultaneously with 4525 1451 recommended cross section data by means of PADE-2 code. 4525 1451 Eigenvalues of the 6-th digits relative covariance matrix 4525 1451 given in the 33-file are the following: 4525 1451 4525 1451 4.63337E-06 4.70345E-06 4.81607E-06 4.97060E-06 4525 1451 5.13303E-06 5.31401E-06 5.53768E-06 5.79915E-06 4525 1451 6.08110E-06 6.37403E-06 6.83267E-06 7.59079E-06 4525 1451 8.42883E-06 9.54383E-06 1.23066E-05 1.44930E-05 4525 1451 2.21073E-05 4.79767E-05 9.57950E-05 1.42110E-04 4525 1451 1.77925E-04 2.06906E-04 5.62715E-04 1.57661E-03 4525 1451 1.64672E-03 1.73445E-03 2.10901E-03 2.41391E-03 4525 1451 2.81682E-03 8.17990E-03 1.45397E-02 1.82324E-02 4525 1451 5.39402E-02 1.07397E-01 1.22586E-01 4525 1451 4525 1451 References: 4525 1451 1. R.B.Firestone Table of Isotopes, Eighth edition, Vol. 1, 4525 1451 John Wiley & Sons, Inc., New York, 1995 4525 1451 2. H.Vonach et al. Proc. Inter. Conf. on Nuclear Data for 4525 1451 Science and Technology Gatlinburg, Tennessee, USA May 9-13, 4525 1451 1994, v.1, pp.278-280 ; 4525 1451 U.Schotzig PTB, Braunschweig, FRG, 1994 4525 1451 3. K.I.Zolotarev, F.K.Zolotarev RRDF-2002, MAT=4921, IPPE, 4525 1451 Obninsk, evaluated March 2003 4525 1451 4. W.G.Cross Physics in Canada, v.19, p.39, June 1963 4525 1451 5. A.Pazsit, J.Csikai Sov. Journal of Nuclear Physics, v.15, 4525 1451 p.232, 1972. 4525 1451 6. D.C.Santry, J.P.Butler Canadian Journal of Physics, v.52, 4525 1451 p.1421, August 1974 4525 1451 7. A.Pazsit, G.Peto, J.Csikai, I.Jozsa, J.Bacso 4525 1451 J. of Applied Radiation and Isotopes, v.26, p.621, Oct. 1975 4525 1451 8. A.Paulsen et al. Nucl. Sci. Eng., v.76, p.331, 1980 4525 1451 9. Li Jianwei et al. Proc. of an Int. Conf. on Nuclear Data for 4525 1451 Science and Technology, 30 May - 3 June 1988, Mito, Japan, 4525 1451 Saikon Publishing Co., LTD, p.315, 1988 4525 1451 10. M.M.H.Miah, B.Strohmaier, H.Vonach, W.Mannhart, D.Schmidt 4525 1451 Phys. Rev., pt. C, v.54, No.1, p.222, July 1996 4525 1451 11. W.Nagel J. Nucl. Energ., v.20, p.475, June 1966 4525 1451 12. I.Kimura et al. Journal of Nucl. Science and Technology, 4525 1451 v.6, p.485, September 1969 4525 1451 13. E.Barnard, D.Reitmann Nucl. Phys., v.A303, p.27, June 1978 4525 1451 14. S.A.Badikov et al. Preprint FEI-1686, Obninsk, 1985 4525 1451 15. S.Badikov, N.Rabotnov, K.Zolotarev Proc. of NEANSC Speciali- 4525 1451 st's Meeting on Evaluation and Processing of Covariance Data, 4525 1451 Oak Ridge, USA, 7-9 September 1992, OECD, Paris, 1993, p.105 4525 1451 16. L.W.Weston et al. Evaluated Neutron Data for Uranium-235, 4525 1451 ENDF/B-VI Library, MAT=9228, MF=5, MT=18, eval. April 1989 4525 1451 17. W.Mannhart IAEA-TECDOC-410, p.158, IAEA, Vienna, 1987 4525 1451 18. E.I.Grigor'ev et al. 6th All Union Conference on Neutron 4525 1451 Physics, Kiev, 2-6 October 1983, v.3, p.187, Moscow 1984 4525 1451 19. O.Horibe et al. Proc. of Conf.: 50 Years with Nuclear Fission,4525 1451 Washington D.C., 25-28 April 1989, v.2, p.923 ; 4525 1451 O.Horibe, H.Chatani Proc. of Int. Conf. Nuclear Data for Sci.4525 1451 and Technology, Julich, FRG, 13-17 May 1991, Springer-Verlag, 4525 1451 Berlin Heidelberg, 1992, p.68 4525 1451 20. K.Kobayashi, T.Kobayashi Progress Report NEANDC(J)-155/U, 4525 1451 p.52, August 1990 4525 1451 21. G.J.Kirouac et al. Report - 4005, Knolls Atomic Power Lab., 4525 1451 January 1974 4525 1451 22. G.P.Lamaze et al. Nucl. Sci. Eng., v.100, p.43, Sept. 1988 4525 1451 23. W.Mannhart Prog. Report INDC(Ger)-045, pp.40-43, June 1999 4525 1451 24. W.Mannhart Validation of Differential Cross Sections with 4525 1451 Integral Data , Report INDC(NDS)-435, pp.59-64, IAEA, Vienna, 4525 1451 September 2002 4525 1451 ***************************************************************** 4525 1451 File 2 added to the pointwise file containing only the effective 4525 1451 scattering radius with no resonance parameters given. 4525 1451 Taken from ENDF/B-VI 4525 1451 4525 1451 ***************************************************************** 4525 1451 ***************** Program LINEAR (VERSION 2002-1) ***************4525 1451 For All Data Greater than 1.0000E-10 barns in Absolute Value 4525 1451 Data Linearized to Within an Accuracy of .100000000 per-cent 4525 1451 ***************** Program SIGMA1 (VERSION 2002-1) ***************4525 1451 Data Doppler Broadened to 300.000000 Kelvin 4525 1451 for All Data Greater than 1.0000E-10 barns in Absolute Value 4525 1451 Data Linearized to Within an Accuracy pf .100000000 per-cent 4525 1451 ***************** Program FIXUP (Version 2002-1) ****************4525 1451 Corrected ZA/AWR in All Sections-----------------------------Yes 4525 1451 Corrected Thresholds-----------------------------------------Yes 4525 1451 Extended Cross Sections to 20 MeV----------------------------No 4525 1451 Allow Cross Section Deletion---------------------------------No 4525 1451 Allow Cross Section Reconstruction---------------------------No 4525 1451 Make All Cross Sections Non-Negative-------------------------Yes 4525 1451 Delete Energies Not in Ascending Order-----------------------Yes 4525 1451 Deleted Duplicate Points-------------------------------------Yes 4525 1451 Check for Ascending MAT/MF/MT Order--------------------------Yes 4525 1451 Check for Legal MF/MT Numbers--------------------------------Yes 4525 1451 Allow Creation of Missing Sections---------------------------No 4525 1451 Allow Insertion of Energy Points-----------------------------No 4525 1451 Create Uniform Energy Grid-----------------------------------No 4525 1451 Delete Section if Cross Section =0 at All Energies-----------Yes 4525 1451 ***************** Program GROUPIE (VERSION 2002-1) **************4525 1451 Unshielded Group Averages Using 640 Groups 4525 1451 Weighting Spectrum: Flat (Constant) Spectrum 4525 1451 1 451 187 14525 1451 8 4 2 14525 1451 10 4 88 14525 1451 40 4 122 14525 1451 4525 1 0 4525 0 0 4.51030E+4 1.02021E+2 0 0 1 04525 3291 0.0 -3.97560E+4 45103 0 1 2554525 3291 255 1 4525 3291 40000.0000 .000131876 42500.0000 .000427871 45000.0000 .0007243454525 3291 47500.0000 .001020819 50000.0000 .001317294 52500.0000 .0016137684525 3291 55000.0000 .001910242 57500.0000 .002206716 60000.0000 .0025328374525 3291 63000.0000 .002888606 66000.0000 .003244375 69000.0000 .0036001444525 3291 72000.0000 .004015207 76000.0000 .004489566 80000.0000 .0049639244525 3291 84000.0000 .005438283 88000.0000 .005912641 92000.0000 .0063870004525 3291 96000.0000 .006861358 100000.000 .007395012 105000.000 .0079879604525 3291 110000.000 .008580908 115000.000 .009173856 120000.000 .0113795894525 3291 127500.000 .015198106 135000.000 .019016624 142500.000 .0228351414525 3291 150000.000 .027155800 160000.000 .031978600 170000.000 .0368014004525 3291 180000.000 .041624200 190000.000 .046447000 200000.000 .0510906904525 3291 210000.000 .055555270 220000.000 .060019850 230000.000 .0644844304525 3291 240000.000 .070033483 255000.000 .076286950 270000.000 .0813926004525 3291 280000.000 .087519380 300000.000 .095305320 320000.000 .1027081604525 3291 340000.000 .110028295 360000.000 .116852200 380000.000 .1235934004525 3291 400000.000 .131003000 425000.000 .139081000 450000.000 .1476172504525 3291 475000.000 .156611750 500000.000 .167446250 525000.000 .1801207504525 3291 550000.000 .196934000 575000.000 .217886000 600000.000 .2482181004525 3291 630000.000 .288511267 660000.000 .336357000 690000.000 .3858751334525 3291 720000.000 .440636700 760000.000 .491837800 800000.000 .5305162004525 3291 840000.000 .558433425 880000.000 .578712300 920000.000 .5945503004525 3291 960000.000 .607908400 1000000.00 .628120500 1100000.00 .6549508334525 3291 1200000.00 .680806500 1300000.00 .706662167 1400000.00 .7317365004525 3291 1500000.00 .756029500 1600000.00 .779401000 1700000.00 .8018510004525 3291 1800000.00 .823285500 1900000.00 .843704500 2000000.00 .8631160004525 3291 2100000.00 .881520000 2200000.00 .898753833 2300000.00 .9148175004525 3291 2400000.00 .930881167 2500000.00 .945705167 2600000.00 .9592895004525 3291 2700000.00 .972873833 2800000.00 .985413333 2900000.00 .9969080004525 3291 3000000.00 1.00840267 3100000.00 1.01891625 3200000.00 1.028448754525 3291 3300000.00 1.03798125 3400000.00 1.04751375 3500000.00 1.056186254525 3291 3600000.00 1.06399875 3700000.00 1.07181125 3800000.00 1.079623754525 3291 3900000.00 1.08670417 4000000.00 1.09305250 4100000.00 1.099400834525 3291 4200000.00 1.10574917 4300000.00 1.11209750 4400000.00 1.118445834525 3291 4500000.00 1.12419500 4600000.00 1.12934500 4700000.00 1.134495004525 3291 4800000.00 1.13964500 4900000.00 1.14479500 5000000.00 1.149945004525 3291 5100000.00 1.15509500 5200000.00 1.16024500 5300000.00 1.165395004525 3291 5400000.00 1.17054500 5500000.00 1.17544306 5600000.00 1.180089174525 3291 5700000.00 1.18473528 5800000.00 1.18938139 5900000.00 1.194027504525 3291 6000000.00 1.19867361 6100000.00 1.20331972 6200000.00 1.207965834525 3291 6300000.00 1.21261194 6400000.00 1.21725806 6500000.00 1.221904174525 3291 6600000.00 1.22655028 6700000.00 1.23119639 6800000.00 1.235842504525 3291 6900000.00 1.24048861 7000000.00 1.24513472 7100000.00 1.249780834525 3291 7200000.00 1.25442694 7300000.00 1.25850604 7400000.00 1.262018114525 3291 7500000.00 1.26553018 7600000.00 1.26904225 7700000.00 1.272554324525 3291 7800000.00 1.27521141 7900000.00 1.27701386 8000000.00 1.278816324525 3291 8100000.00 1.28061877 8200000.00 1.28169500 8300000.00 1.281263334525 3291 8400000.00 1.28005000 8500000.00 1.27883667 8600000.00 1.276161674525 3291 8700000.00 1.27202500 8800000.00 1.26788833 8900000.00 1.262252504525 3291 9000000.00 1.25511750 9100000.00 1.24661250 9200000.00 1.236737504525 3291 9300000.00 1.22537250 9400000.00 1.21251750 9500000.00 1.198110004525 3291 9600000.00 1.18215000 9700000.00 1.16464500 9800000.00 1.145595004525 3291 9900000.00 1.12508000 10000000.0 1.10310000 10100000.0 1.079822504525 3291 10200000.0 1.05524750 10300000.0 1.02941133 10400000.0 1.002314004525 3291 10500000.0 .975216667 10600000.0 .947189214 10700000.0 .9182316434525 3291 10800000.0 .889274071 10900000.0 .860316500 11000000.0 .8313589294525 3291 11100000.0 .802401357 11200000.0 .773443786 11300000.0 .7452375004525 3291 11400000.0 .717782500 11500000.0 .690327500 11600000.0 .6638342504525 3291 11700000.0 .638302750 11800000.0 .613694250 11900000.0 .5900087504525 3291 12000000.0 .567314500 12100000.0 .545611500 12200000.0 .5249195004525 3291 12300000.0 .505238500 12400000.0 .486307000 12500000.0 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16500000.0 .213601000 16600000.0 .212043000 16700000.0 .2104850004525 3291 16800000.0 .208927000 16900000.0 .207520000 17000000.0 .2062640004525 3291 17100000.0 .205008000 17200000.0 .203752000 17300000.0 .2024960004525 3291 17400000.0 .201240000 17500000.0 .200100429 17600000.0 .1990772864525 3291 17700000.0 .198054143 17800000.0 .197031000 17900000.0 .1960078574525 3291 18000000.0 .194984714 18100000.0 .193961571 18200000.0 .1930241674525 3291 18300000.0 .192172500 18400000.0 .191320833 18500000.0 .1904691674525 3291 18600000.0 .189617500 18700000.0 .188765833 18800000.0 .1879141674525 3291 18900000.0 .187062500 19000000.0 .186210833 19100000.0 .1854136674525 3291 19200000.0 .184671000 19300000.0 .183928333 19400000.0 .1831856674525 3291 19500000.0 .182443000 19600000.0 .181700333 19700000.0 .1809576674525 3291 19800000.0 .180215000 19900000.0 .179472333 20000000.0 0.0 4525 3291 4525 3 0 4525 0 0 4.51030E+4 1.02021E+2 0 0 0 1452533291 0.0000E+00 0.0000E+00 0 291 0 1452533291 0.0000E+00 0.0000E+00 1 5 703 37452533291 1.000000-5 4.000000+4 2.000000+5 4.000000+5 6.000000+5 8.000000+5452533291 1.000000+6 1.500000+6 2.000000+6 2.500000+6 3.000000+6 3.500000+6452533291 4.000000+6 4.500000+6 5.000000+6 5.500000+6 6.000000+6 6.500000+6452533291 7.000000+6 7.500000+6 8.000000+6 8.500000+6 9.000000+6 9.500000+6452533291 1.000000+7 1.050000+7 1.100000+7 1.150000+7 1.200000+7 1.300000+7452533291 1.400000+7 1.500000+7 1.600000+7 1.700000+7 1.800000+7 1.900000+7452533291 2.000000+7 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0452533291 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0452533291 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0452533291 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0452533291 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0452533291 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0452533291 0.000000+0 1.201660-1 3.626120-3 5.830250-3 2.406510-3 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1.034210-4-4.532930-5 2.185350-3 1.745000-3452533291 1.484750-3 1.385400-3 1.380060-3 1.419750-3 1.470620-3 1.509200-3452533291 1.519370-3 1.490590-3 1.417430-3 1.300150-3 1.146670-3 9.756850-4452533291 8.197340-4 7.246800-4 7.391190-4 8.883700-4 1.141090-3 1.399790-3452533291 1.541910-3 1.489490-3 1.091420-3 4.000170-4-6.305970-6-6.918640-5452533291 1.719700-5 8.992070-5 7.228520-5-5.799610-5 1.705580-3 1.595370-3452533291 1.530820-3 1.492930-3 1.468820-3 1.447610-3 1.420890-3 1.382350-3452533291 1.327560-3 1.254000-3 1.161690-3 1.054450-3 9.418340-4 8.412030-4452533291 7.780520-4 7.807980-4 8.663420-4 1.019870-3 1.186120-3 1.289640-3452533291 1.274940-3 1.030670-3 5.329380-4 1.499900-4-1.318120-5-3.406270-5452533291 -5.586790-6 1.728610-5 1.067630-5 1.634360-3 1.625160-3 1.607390-3452533291 1.574910-3 1.528640-3 1.468740-3 1.395220-3 1.308330-3 1.209170-3452533291 1.100440-3 9.876960-4 8.808630-4 7.955860-4 7.526840-4 7.727340-4452533291 8.635330-4 1.004710-3 1.144140-3 1.218310-3 1.185790-3 9.517920-4452533291 5.073870-4 1.702880-4 1.966640-5-9.743300-6 7.648470-6 2.860580-5452533291 3.445390-5 1.686590-3 1.698690-3 1.689110-3 1.652020-3 1.589090-3452533291 1.501910-3 1.392580-3 1.264400-3 1.123050-3 9.780430-4 8.445670-4452533291 7.445890-4 7.049900-4 7.491080-4 8.799410-4 1.062630-3 1.225610-3452533291 1.292150-3 1.222690-3 9.173280-4 4.205910-4 1.117470-4 2.198870-5452533291 3.810570-5 6.754520-5 6.709010-5 2.377900-5 1.767300-3 1.787600-3452533291 1.778600-3 1.732110-3 1.648380-3 1.529190-3 1.378700-3 1.204750-3452533291 1.020930-3 8.488400-4 7.195540-4 6.705770-4 7.337420-4 9.118860-4452533291 1.155540-3 1.366660-3 1.444020-3 1.340990-3 9.399110-4 3.401500-4452533291 3.115010-5 3.701460-6 7.870690-5 1.295250-4 1.052400-4-3.667290-6452533291 1.861870-3 1.882680-3 1.864080-3 1.796830-3 1.681110-3 1.520660-3452533291 1.324290-3 1.108380-3 8.998800-4 7.381710-4 6.716550-4 7.426720-4452533291 9.579980-4 1.258590-3 1.523090-3 1.624410-3 1.503210-3 1.018640-3452533291 3.048970-4-3.417320-5-2.491860-5 9.680240-5 1.687500-4 1.281170-4452533291 -3.466650-5 1.953700-3 1.962510-3 1.921520-3 1.822400-3 1.667480-3452533291 1.465140-3 1.232610-3 9.995790-4 8.107960-4 7.228870-4 7.879580-4452533291 1.019900-3 1.357890-3 1.666730-3 1.798970-3 1.680060-3 1.145900-3452533291 3.346320-4-6.084060-5-5.413060-5 8.742170-5 1.740050-4 1.296000-4452533291 -5.921130-5 2.018400-3 2.004860-3 1.934740-3 1.802360-3 1.614240-3452533291 1.386200-3 1.147330-3 9.433540-4 8.345250-4 8.798450-4 1.102280-3452533291 1.448040-3 1.781360-3 1.944550-3 1.848350-3 1.309570-3 4.359200-4452533291 -3.516650-5-7.449720-5 5.289760-5 1.437700-4 1.099710-4-6.921970-5452533291 2.040740-3 2.005170-3 1.912560-3 1.760710-3 1.561740-3 1.340470-3452533291 1.138540-3 1.013920-3 1.027960-3 1.213110-3 1.531140-3 1.860530-3452533291 2.046820-3 1.989450-3 1.493570-3 6.047000-4 4.891560-5-7.673680-5452533291 1.371240-6 8.420380-5 7.466270-5-5.785160-5 2.030060-3 1.988070-3452533291 1.894760-3 1.751080-3 1.574090-3 1.396040-3 1.266060-3 1.241910-3452533291 1.364030-3 1.616560-3 1.907010-3 2.099450-3 2.088640-3 1.678040-3452533291 8.271910-4 1.904990-4-5.195030-5-5.417730-5 7.952040-6 3.346140-5452533291 -1.934510-5 2.026430-3 2.007020-3 1.944500-3 1.840030-3 1.711650-3452533291 1.592980-3 1.531150-3 1.570710-3 1.721980-3 1.934050-3 2.105380-3452533291 2.136050-3 1.839400-3 1.079100-3 3.806240-4 7.422110-6-9.667630-5452533291 -6.646090-5-7.827400-8 5.032560-5 2.093200-3 2.129750-3 2.125450-3452533291 2.075490-3 1.991990-3 1.903760-3 1.852790-3 1.874580-3 1.967840-3452533291 2.079970-3 2.129570-3 1.951850-3 1.324730-3 6.001410-4 1.053960-4452533291 -1.067670-4-1.161000-4-1.004330-5 1.517680-4 2.296170-3 2.410830-3452533291 2.468920-3 2.452110-3 2.363350-3 2.232360-3 2.112150-3 2.051350-3452533291 2.056640-3 2.080990-3 1.991540-3 1.516680-3 8.163850-4 2.383610-4452533291 -6.646880-5-1.167200-4 1.891140-5 2.788530-4 2.672770-3 2.860640-3452533291 2.948500-3 2.902960-3 2.729070-3 2.481360-3 2.246960-3 2.093980-3452533291 2.025970-3 1.946320-3 1.600080-3 9.808020-4 3.887410-4 3.325860-5452533291 -4.995920-5 9.539570-5 4.136770-4 3.207960-3 3.435520-3 3.495000-3452533291 3.349050-3 3.028830-3 2.633510-3 2.280710-3 2.036640-3 1.834010-3452533291 1.526700-3 1.031610-3 5.177390-4 1.802330-4 8.386130-5 2.111010-4452533291 5.209780-4 3.846890-3 4.078820-3 4.065070-3 3.778340-3 3.287360-3452533291 2.728170-3 2.229490-3 1.728800-3 1.286970-3 9.106850-4 5.629970-4452533291 3.263650-4 2.481130-4 3.284120-4 5.472190-4 4.558340-3 4.791720-3452533291 4.690670-3 4.237160-3 3.533920-3 2.750150-3 1.782710-3 9.594800-4452533291 6.041830-4 4.544830-4 3.836850-4 3.591780-4 3.767060-4 4.355290-4452533291 5.379550-3 5.621120-3 5.396270-3 4.706480-3 3.711170-3 2.207740-3452533291 7.515130-4 2.027880-4 1.638560-4 2.563050-4 3.091530-4 2.787680-4452533291 1.688620-4 6.326250-3 6.518650-3 6.081040-3 5.085380-3 3.181080-3452533291 9.695790-4-6.047840-5-2.210550-4-7.551840-5 4.549650-5 2.181970-5452533291 -1.665760-4 7.262800-3 7.312700-3 6.627790-3 4.695680-3 1.873840-3452533291 1.345200-4-4.562920-4-4.594530-4-3.217850-4-2.598870-4-3.491170-4452533291 8.021830-3 7.929730-3 6.477870-3 3.489070-3 1.036030-3-2.102110-4452533291 -5.656340-4-4.887160-4-2.780690-4-8.214510-5 8.624160-3 8.079560-3452533291 5.538600-3 2.666940-3 7.744250-4-2.478930-5-7.501780-5 2.929000-4452533291 8.666250-4 9.222250-3 8.299370-3 5.838530-3 3.614610-3 2.386360-3452533291 2.132820-3 2.604510-3 3.570530-3 1.032610-2 9.947210-3 8.723800-3452533291 7.726450-3 7.373060-3 7.680650-3 8.542170-3 1.246280-2 1.317360-2452533291 1.318840-2 1.302470-2 1.298030-2 1.315830-2 1.597610-2 1.695770-2452533291 1.713040-2 1.683090-2 1.633520-2 1.899860-2 1.933620-2 1.899850-2452533291 1.818580-2 2.027880-2 1.992020-2 1.919880-2 2.027840-2 1.978060-2452533291 2.040580-2 452533291 452533 0 4525 0 0 0 0 0 4.71090E+4 1.07969E+2 0 0 34 104731 1451 0.0 0.0 0 0 0 64731 1451 1.00000E+0 2.00000E+7 0 0 10 20024731 1451 3.00000E+2 0.0 1 0 138 44731 1451 47-Ag-109 CNDC EVAL-JUN91 Z.X.ZHAO 4731 1451 DIST-FEB2004 4731 1451 ----IRDF-2002 MATERIAL 4731 4731 1451 -----INCIDENT NEUTRON DATA 4731 1451 ------ENDF-6 FORMAT 4731 1451 ================================================================ 4731 1451 47-AG-109 CNDC EVAL-JUN91 Z.X.ZHAO 4731 1451 DIST- 4731 1451 ----ENDF/A-6 MATERIAL 4731 4731 1451 -----INCIDENT NEUTRON DATA 4731 1451 ------ENDF-6 FORMAT 4731 1451 * * * * * * *4731 1451 THIS EVALUATION WAS PERFORMED BASED ON THE MEASURED DATA 4731 1451 AVAILABLE AND THE CALCULATIONS OF THE SYSTEMATICS AND THEORY. 4731 1451 FILE 2 RESONANCE PARAMETER 4731 1451 ONLY POTENTIAL SCATTERING RADIUS R' OF REF.[1] IS GIVEN. 4731 1451 FILE 10 PRODUCTION CROSS SECTION OF RADIOACTIVE NUCLIDE 4731 1451 MT=102 109-AG(N,GAMMA)110M-AG REACTION 4731 1451 THE PRODUCTION CROSS SECTION OF ISOMER STATE OF 110-AG 4731 1451 ARE OBTAINED BY 4731 1451 C.S.(M) = C.S.(M+G)*R 4731 1451 WHERE C.S.(M+G) REPRESENTS THE TOTAL CAPTURE CROSS SECTION 4731 1451 AND R IS THE RATIO OF THE CAPTURE CROSS SECTION TO ISOMER 4731 1451 STATE AND THE TOTAL CAPTURE CROSS SECTION. THE VALUES OF 4731 1451 R WERE CALCULATED BY USING UNIFY2[2] AND NORMALIZED TO 4731 1451 THE LIMITED DATA FOR C.S.(M) MEASURED AT THERMAL CROSS 4731 1451 SECTION [3-5] AND AROUND 25 KEV [6-7]. 4731 1451 C.S.(M+G) WAS RE-EVALUATED IN THE FOLLOWING WAY: 4731 1451 BELOW 2.5 KEV, C.S.(M+G) WAS CALCULATED FROM A SET 4731 1451 OF RESONANCE PARAMETERS OF JENDL-3 AND AVERAGED IN 4731 1451 SMALL ENERGY INTERVALS. 4731 1451 FROM 2.5 KEV TO 20 MEV, THE SYSTEMATICS OF ZHAO [8] 4731 1451 WAS USED TO FIT THE MEASURED DATA OF REFS.[9-10] IN THE 4731 1451 ENERGY RANGE FROM 2.5 KEV TO 2 MEV. 4731 1451 FILE 8 RADIOACTIVE DECAY DATA 4731 1451 THE RADIOACTIVE DECAY DATA FOR 110M-AG ARE TAKEN FROM 4731 1451 REFS.[11] AND [12]. 4731 1451 FILE 40 COVARIANCE DATA FOR FILE 10 4731 1451 COVARIANCE DATA OF FILE 10 CONSISTES OF TWO COMPONENTS. 4731 1451 ONE OF THEM COME FROM THE UNCERTAINTY OF TOTAL CAPTURE 4731 1451 CROSS SECTION. ANOTHER COME FROM THE NORMALIZATION ERROR 4731 1451 OF R=C.S.(M)/C.S.(M+G). 4731 1451 ================================================================ 4731 1451 =========== Processing done at IAEA/NDS for IRDF-2002 ========== 4731 1451 ================================================================ 4731 1451 The original file was calculated from a set of resonance 4731 1451 parameters taken from JENDL-3. 4731 1451 The JENDL-3.2 file for Ag109 was processed as follows to produce 4731 1451 a pointwise file. 4731 1451 ***************** Program LINEAR (VERSION 2002-1) ***************4731 1451 For All Data Greater than 1.0000E-10 barns in Absolute Value 4731 1451 Data Linearized to Within an Accuracy of .100000000 per-cent 4731 1451 ***************** Program RECENT (VERSION 2002-1) ***************4731 1451 for All Data Greater than 1.0000E-10 barns in Absolute Value 4731 1451 Data Linearized to within an Accuracy of .100000000 per-cent 4731 1451 ***************** Program SIGMA1 (VERSION 2002-1) ***************4731 1451 Data Doppler Broadened to 300.000000 Kelvin 4731 1451 for All Data Greater than 1.0000E-10 barns in Absolute Value 4731 1451 Data Linearized to Within an Accuracy pf 1.00000000 per-cent 4731 1451 ***************** Program FIXUP (Version 2002-1) ****************4731 1451 Corrected ZA/AWR in All Sections-----------------------------Yes 4731 1451 Corrected Thresholds-----------------------------------------No 4731 1451 Extended Cross Sections to 20 MeV----------------------------No 4731 1451 Allow Cross Section Deletion---------------------------------No 4731 1451 Allow Cross Section Reconstruction---------------------------No 4731 1451 Make All Cross Sections Non-Negative-------------------------Yes 4731 1451 Delete Energies Not in Ascending Order-----------------------Yes 4731 1451 Deleted Duplicate Points-------------------------------------Yes 4731 1451 Check for Ascending MAT/MF/MT Order--------------------------Yes 4731 1451 Check for Legal MF/MT Numbers--------------------------------Yes 4731 1451 Allow Creation of Missing Sections---------------------------No 4731 1451 Allow Insertion of Energy Points-----------------------------No 4731 1451 Create Uniform Energy Grid-----------------------------------No 4731 1451 Delete Section if Cross Section =0 at All Energies-----------Yes 4731 1451 4731 1451 The branching ratio of 0.0465 was applied to the Ag109 total 4731 1451 capture cross section data of the JENDL-3.2 file in the energy 4731 1451 range 1.0E-5 to 7.1E+3. The scaled data with the above energy 4731 1451 range was then edited into the original data file as used for 4731 1451 IRDF-90 4731 1451 ================================================================ 4731 1451 4731 1451 REFERENCES 4731 1451 [1] S. F. Mughabghab et al., BNL-325, 4th ed., Vol.(1981) 4731 1451 4731 1451 [2] Zhang Jinshang, " Semi-classical Theory Code UNIFY of 4731 1451 Multi-step Nuclear Reaction", to be published 4731 1451 4731 1451 [3] T. B. Ryves et al., J. Nucl. Energy, 25(1971)129 4731 1451 4731 1451 [4] J. K. Aaldijk et al., RCN-176,1972 4731 1451 4731 1451 [5] A. Simonits et al., JRN, 81(1984)369 4731 1451 4731 1451 [6] M. Sriramachandra et al., JP/A,5(1972)877 4731 1451 4731 1451 [7] W. Poenitz, EANDC(E0-66,6(1966) 4731 1451 4731 1451 [8] Zhao Zhixiang et al., Chinese Nucl. Phys., 11(1989)71 4731 1451 4731 1451 [9] R. L. Macklin, Nucl. Sci. and Eng., 82(1982)400 4731 1451 4731 1451 [10] M. Mizumoto et al., "Neutron Radiative Capture and Transm4731 1451 Measurements of 107-Ag and 109-Ag", Proc. Int. Conf. Nucl4731 1451 for Sci. and Tech., Antwerp, 1982, p.226(1983) 4731 1451 4731 1451 [11] J. K. Tuli, Nuclear Wallet Cards ,1990 4731 1451 4731 1451 [12] C. M. Lederer and V. S. shirley, Table of Isotopes, 4731 1451 Seventh Edition(1978) 4731 1451 4731 1451 *****************************************************************4731 1451 4731 1451 4731 1451 4731 1451 4731 1451 ***************** PROGRAM LINEAR (VERSION 96-1) ************* 4731 1451 FOR ALL DATA GREATER THAN 1.00000-10 BARNS IN ABSOLUTE VALUE 4731 1451 DATA LINEARIZED TO WITHIN AN ACCURACY OF 0.10000000 PER-CENT 4731 1451 ***************** Program FIXUP (Version 2002-1) ****************4731 1451 Corrected ZA/AWR in All Sections-----------------------------Yes 4731 1451 Corrected Thresholds-----------------------------------------Yes 4731 1451 Extended Cross Sections to 20 MeV----------------------------No 4731 1451 Allow Cross Section Deletion---------------------------------No 4731 1451 Allow Cross Section Reconstruction---------------------------No 4731 1451 Make All Cross Sections Non-Negative-------------------------Yes 4731 1451 Delete Energies Not in Ascending Order-----------------------Yes 4731 1451 Deleted Duplicate Points-------------------------------------Yes 4731 1451 Check for Ascending MAT/MF/MT Order--------------------------Yes 4731 1451 Check for Legal MF/MT Numbers--------------------------------Yes 4731 1451 Allow Creation of Missing Sections---------------------------No 4731 1451 Allow Insertion of Energy Points-----------------------------No 4731 1451 Create Uniform Energy Grid-----------------------------------No 4731 1451 Delete Section if Cross Section =0 at All Energies-----------Yes 4731 1451 *****************************************************************4731 1451 ***************** Program GROUPIE (VERSION 2002-1) **************4731 1451 Unshielded Group Averages Using 640 Groups 4731 1451 Weighting Spectrum: Flat (Constant) Spectrum 4731 1451 1 451 146 14731 1451 8 102 2 14731 1451 10 102 217 14731 1451 40 102 10 14731 1451 4731 1 0 4731 0 0 4.71090E+4 1.07969E+2 0 0 1 04731 3293 6.80570E+6 6.68810E+6 47110 2 1 6414731 3293 641 1 4731 3293 .000100000 65.8829137 .000105000 64.3718842 .000110000 62.86085474731 3293 .000115000 61.3498263 .000120000 59.7673282 .000127500 58.20321084731 3293 .000135000 56.6406830 .000142500 55.0781536 .000150000 53.37191724731 3293 .000160000 51.8716541 .000170000 50.4199466 .000180000 48.96823914731 3293 .000190000 47.5717224 .000200000 46.5123774 .000210000 45.51024454731 3293 .000220000 44.5081115 .000230000 43.5059794 .000240000 42.26130974731 3293 .000255000 41.0758325 .000270000 40.2204986 .000280000 39.19409774731 3293 .000300000 37.8255648 .000320000 36.5868623 .000340000 35.63353004731 3293 .000360000 34.6903705 .000380000 33.7472116 .000400000 32.71870274731 3293 .000425000 31.8382357 .000450000 31.0164125 .000475000 30.19458894731 3293 .000500000 29.3727662 .000525000 28.6779698 .000550000 28.11072844731 3293 .000575000 27.5434871 .000600000 26.9195219 .000630000 26.23883294731 3293 .000660000 25.5726206 .000690000 25.0645516 .000720000 24.52244004731 3293 .000760000 23.9028841 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19800000.0 .0001042054731 3293 19900000.0 .000103115 20000000.0 0.0 4731 3293 4731 3 0 4731 0 0 4.71090E+4 1.07969E+2 0 0 0 1473133293 0.0000E+00 0.0000E+00 0 293 0 2473133293 0.0000E+00 0.0000E+00 0 1 22 11473133293 1.000000-5 1.000000-4 1.000000+0 2.500000-3 8.000000+1 1.000000-2473133293 1.500000+2 2.250000-2 3.000000+2 2.250000-2 4.000000+2 2.250000-2473133293 5.000000+2 9.000000-2 6.500000+2 1.000000-2 2.500000+4 2.500000-3473133293 2.000000+6 9.000000-2 2.000000+7 0.000000+0 473133293 0.000000+0 0.000000+0 0 1 4 2473133293 1.000000-5 2.500000-3 2.000000+7 0.000000+0 473133293 473133 0 4731 0 0 0 0 0 4.80000E+4 1.11445E+2 0 0 34 104800 1451 0.0 0.0 0 0 0 64800 1451 1.00000E+0 2.00000E+7 0 0 10 20024800 1451 3.00000E+2 0.0 1 0 77 24800 1451 48-Cd- 0 BNL EVAL-MAY74 S.PEARLSTEIN(TRANS FROM U.K.) 4800 1451 DIST-Feb2004 4800 1451 ----IRDF-2002 MATERIAL 4800 4800 1451 -----INCIDENT NEUTRON DATA 4800 1451 ------ENDF-6 FORMAT 4800 1451 *****************************************************************4800 1451 48-CD- 0 BNL EVAL-MAY74 S.PEARLSTEIN(TRANS FROM U.K.) 4800 1451 DIST-JAN90 19900130 4800 1451 ----ENDF/B-VI MATERIAL 4800 4800 1451 -----INCIDENT NEUTRON DATA 4800 1451 ------ENDF-6 FORMAT 4800 1451 ENDF/B-V MATERIAL CONVERTED TO ENDF-6 FORMAT BY NNDC 4800 1451 * * * * * * *4800 1451 UKNDL TO ENDF/B TRANSLATION FROM 4800 1451 UKNDL DATA FILE 70 RECEIVED JULY 1973 FROM CCDN 4800 1451 4800 1451 TRANSLATION OF UKNDL DFN-70 INTO ENDF FORMAT 4800 1451 4800 1451 RESONANCE REGION EVALUATED BY MF JAMES IN 1969 4800 1451 FICTITIOUS CD-111 RESONANCE AT -0.5 EV TO ADJUST CROSS SECTION 4800 1451 IN EV RANGE AND ACHIEVE CONSISTENCY WITH 20 BARN CD-111 THERMAL 4800 1451 CAPTURE CROSS SECTION (BALDOCK ETAL 1966 ORNL-3994) 4800 1451 ABOVE THE RESOLVED RESONANCE RANGE MAINLY DRAKE GA-6997 WAS USED4800 1451 4800 1451 PLEASE REFER COMMENTS OR QUESTIONS ABOUT THE TRANSLATION 4800 1451 CODE (UKE*** ORNL-TM-2880 OR ENDF-134) TO 4800 1451 R. Q. WRIGHT, OAK RIDGE NATIONAL LABORATORY, 4800 1451 P. O. BOX X, OAK RIDGE, TENNESSEE 37830. 4800 1451 4800 1451 RANGE EXTENDED FROM 1.0E-5 EV TO 20 MEV AND ELASTIC SCATTERING 4800 1451 TRANSFER AND TRANSFORMATION DATA ADDED BY NNCSC AT BNL 4800 1451 0 0 0 4800 1451 ***************** PROGRAM LINEAR (VERSION 89-1) *****************4800 1451 DATA LINEARIZED TO WITHIN AN ACCURACY OF 0.300 PER-CENT 4800 1451 ***************** PROGRAM FIXUP (VERSION 89-2) ******************4800 1451 *RECONSTRUCTED MT NUMBERS 4800 1451 4 =+( 50, 91) 4800 1451 103 =+(600,649) 4800 1451 104 =+(650,699) 4800 1451 105 =+(700,749) 4800 1451 106 =+(750,799) 4800 1451 107 =+(800,849) 4800 1451 101 =+( 18, 18)+(102,116) 4800 1451 27 =+(101,101) 4800 1451 3 =+( 4, 5)+( 16, 17)+( 22, 37)+( 41, 42) 4800 1451 19 =+( 18, 18)-( 20, 21)-( 38, 38) 4800 1451 1 =+( 2, 3) 4800 1451 *****************************************************************4800 1451 4800 1451 4800 1451 4800 1451 4800 1451 ***************** Program LINEAR (VERSION 2002-1) ***************4800 1451 For All Data Greater than 1.0000E-10 barns in Absolute Value 4800 1451 Data Linearized to Within an Accuracy of .100000000 per-cent 4800 1451 ***************** Program SIGMA1 (VERSION 2002-1) ***************4800 1451 Data Doppler Broadened to 300.000000 Kelvin 4800 1451 for All Data Greater than 1.0000E-10 barns in Absolute Value 4800 1451 Data Linearized to Within an Accuracy pf .100000000 per-cent 4800 1451 ***************** Program FIXUP (Version 2002-1) ****************4800 1451 Corrected ZA/AWR in All Sections-----------------------------Yes 4800 1451 Corrected Thresholds-----------------------------------------Yes 4800 1451 Extended Cross Sections to 20 MeV----------------------------No 4800 1451 Allow Cross Section Deletion---------------------------------No 4800 1451 Allow Cross Section Reconstruction---------------------------No 4800 1451 Make All Cross Sections Non-Negative-------------------------Yes 4800 1451 Delete Energies Not in Ascending Order-----------------------Yes 4800 1451 Deleted Duplicate Points-------------------------------------Yes 4800 1451 Check for Ascending MAT/MF/MT Order--------------------------Yes 4800 1451 Check for Legal MF/MT Numbers--------------------------------Yes 4800 1451 Allow Creation of Missing Sections---------------------------No 4800 1451 Allow Insertion of Energy Points-----------------------------No 4800 1451 Create Uniform Energy Grid-----------------------------------No 4800 1451 Delete Section if Cross Section =0 at All Energies-----------Yes 4800 1451 ***************** Program GROUPIE (VERSION 2002-1) **************4800 1451 Unshielded Group Averages Using 640 Groups 4800 1451 Weighting Spectrum: Flat (Constant) Spectrum 4800 1451 1 451 83 04800 1451 3 1 217 04800 1451 4800 1 0 4800 0 0 4.80000E+4 1.11445E+2 0 0 0 04800 3 1 0.0 0.0 0 0 1 6414800 3 1 641 1 4800 3 1 .000100000 29407.0831 .000105000 28722.2715 .000110000 28071.69934800 3 1 .000115000 27472.1873 .000120000 26771.4577 .000127500 25994.10844800 3 1 .000135000 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DATA 4931 1451 ------ENDF-6 FORMAT 4931 1451 ***************************************************************** 4931 1451 49-In-115 FEI EVAL-Mar03 K.I.Zolotarev 4931 1451 DIST-Apr03 4931 1451 ----BROND-2 MATERIAL 4931 4931 1451 -----INCIDENT NEUTRON DATA 4931 1451 ------ENDF-6 FORMAT 4931 1451 ------Russian Reactor Dosimetry File RRDF-2002 4931 1451 ***************************************************************** 4931 1451 Authors of evaluation: K.I.Zolotarev and P.K.Zolotarev 4931 1451 ***************************************************************** 4931 1451 MF=3 4931 1451 MT= 51 - In115(n,n')In115m reaction 4931 1451 ------------------------------------- 4931 1451 *****************************************************************4931 1451 CHINESE EVALUATION OF (N,2N) REACTION ADDED AT NDS 4931 1451 * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * 4931 1451 4931 1451 THE FOLLOWING FILE SECTIONS ARE INCLUDED - 4931 1451 MF= 1 GENERAL INFORMATION (MT=451) 4931 1451 MF= 8 RADIOACTIVITY DATA (MT=51) 4931 1451 MF=10 CROSS SECTION FOR RADIOACTIVE NUCLIDE 4931 1451 PRODUCTION (MT=51) 4931 1451 MF=40 DATA COVARIANCES FOR RADIOACTIVE NUCLIDE 4931 1451 PRODUCTION (MT=51) 4931 1451 49-IN-115 CNDC/CIAE EVAL-JUN91 CAI DUNJIU AND WANG ZISHENG 4931 1451 ---ENDF-6 FORMAT 4931 REVISION 0 4931 1451 4931 1451 * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * 4931 1451 *****************************************************************4931 1451 49-IN-115 HEDL,ANL EVAL-MAR90 F.SCHMITTROTH,D.L.SMITH,S.CHIBA 4931 1451 DIST-JUN90 19900329 4931 1451 ----ENDF/B-VI MATERIAL 4931 4931 1451 MF=2 MT=151 EVALUATION OF RESOLVED RESONANCE PARAMETERS 4931 1451 BASED ON NEW BNL-325, REF(2). 4931 1451 MF=3 MT=102 VERSION-V UNRESOLVED REGION CONTAINS ADJUSTED DATA 4931 1451 *****************EXTRACT FOR SPECIAL PURPOSE FILE*****************4931 1451 DOSIMETRY 4931 1451 ******************************************************************4931 1451 ***************************************************************** 4931 1451 ********** Start of (N,2N) bibliographical component ********** 4931 1451 ***************************************************************** 4931 1451 4931 1451 Evaluation of the 115-In(n,2n) 114m-In Reaction 4931 1451 Cross-section 4931 1451 Cai Dunjiu Yu Baosheng Wang Zisheng 4931 1451 Lu Hanlin Zhao Wenrong 4931 1451 Institute of Atomic Energy, China 4931 1451 The excitation function of 115-In(n, 2n) 114m-In reaction 4931 1451 covers neutron energy range from threshold of 9.228 MeV 4931 1451 ( including the energy of isomeric state which has 4931 1451 0.190 MeV ) to 20 MeV. Experimental knowledge of the 4931 1451 115-In(n,2n) reaction cross sections is entirely based 4931 1451 on the results of activation measurements. There are 4931 1451 still no experimental results obtained by 4931 1451 direct-neutron-detection methods (such as scintillation 4931 1451 tank). For both the indium isotopes (115-In,0.957, 4931 1451 113-In,0.043), the primary activity 4931 1451 resulting from the (n,2n) process is due to the decay of 4931 1451 metastable state. The interaction with the primary isotope, 4931 1451 115-In, results in a metastable state in 114-In which 4931 1451 decays with 49.5d half-life by means of an E4 transition 4931 1451 to the ground state (1+) with a half-life of 71.9s. 4931 1451 The (n,2n) process also directly populates the ground state. 4931 1451 Seventeen sets of data available (1-18) were collected. 4931 1451 For the sake of comparison, all the experimental results were 4931 1451 adjusted by the unified nuclear parameters (including 4931 1451 intensity of r-line, half-life, branching ratio), 4931 1451 the reference cross section and the dependence of cross 4931 1451 section on energy which were taken from (Ref.1). In several 4931 1451 cases, adjustments were made only for the dependence of cross 4931 1451 section on energy since some necessary parameters used in data 4931 1451 analysis were not given by the authors. 4931 1451 The evaluation was made as the following ways: 4931 1451 (1).At 14.7MeV neutron energy . 4931 1451 Based on the results measured by Ge(Li) detector,the 4931 1451 recommended cross section at 14.7 MeV is 1290 mb. 4931 1451 (2). For the neutron energy range from threshold to 20 MeV. 4931 1451 The recommended excitation function was obtained 4931 1451 by fitting the measured data with least-square method in the 4931 1451 neutron energy range of 10.4-20MeV and by extrapolating from 4931 1451 10.4MeV to 9.228MeV based on the physical trend and systematics 4931 1451 calculation. 4931 1451 Covariance matrix is given for the evaluated data considering all 4931 1451 adjusted factors. 4931 1451 References 4931 1451 1.Zhao Wenrong et al., CNDC-89014;INDC(CPR)-16, 1989 4931 1451 2.Ke Wei et al., Chin.J.Nucl.Phys.11, 3, 11(1989) 4931 1451 Lu Hanlin et al., INDC(CPR)-011/GI, P67(1988) 4931 1451 3.J.Csikai et al., INDC(NDS)-232/L, P45(1989) 4931 1451 4.Li Jiangwei et al., Chin.J.Nucl.Phys., 10, 1, 52(1988) 4931 1451 Wu Zhihua et al., 88MITO, P315(1988) 4931 1451 5.T.B.Ryves et al., J.Phys., G9, 1549(1983) 4931 1451 6.T.B.Ryves et al., J.Phys., G7, 115(1981) 4931 1451 7.A.Reggoug et al., 82Antwerp, P873(1982);NIMA, 227, 249(1984) 4931 1451 8.K.Kayashima et al., NEANDC(J)-61U, P94(1979) 4931 1451 9.D.C.Santry et al., Can.J.Phys., 54, 757(1976) 4931 1451 10.Lu Hanlin et al., At.Energy Sci.Technol., 2, 113(1975) 4931 1451 11.A.Paulsen et al., Atomkernenergie, 26, 34(1975) 4931 1451 12.J.K.Temperley et al., BRL-1491(1970) 4931 1451 13.R.C.Barrall et al., Nucl.Phys., A138, 387(1969) 4931 1451 14.R.C.Barrall et al., AFWL-TR-68-134(1969) 4931 1451 15.H.Rotzer, Nucl.Phys., A109, 694(1968);DWA176, 289(1968) 4931 1451 16.B.Minetti et al., Z.Physik., A217, 83(1968) 4931 1451 17.H.O.Menlove et al., Phys.Rev., 163, 1308(1967) 4931 1451 18.R.J.Prestwood et al., Phys.Rev., 121, 1438(1961) 4931 1451 19.T.Kozlowski et al., Acta.Phys.Pol., 33, 409(1968) 4931 1451 20.R.Prasad et al.,Nucl.Phys., A94, 476(1967) 4931 1451 and 88,349(1966). 4931 1451 21.W.Nagel, JNE, 20, 475(1966) 4931 1451 22.K.C.Garg et al.,IPA17, 525(1979) 4931 1451 23.G.N.Salaita et al., ANS, 16, 59(1973) 4931 1451 24.J.Janczyszyn et al., J.Radiational Chem., 14, 201(1973) 4931 1451 25.M.Bormann et al., EANDC(E)76u, 51(1967) 4931 1451 26.W.Grochulski et al., INR-1172(1970);INR-1197, 10(1970) 4931 1451 Acta.Phys.Pol., BI, 271(1970) 4931 1451 27.H.K.Vonach et al., 68WASH, 2, P885(1968) 4931 1451 28.Wang Zisheng et al., Research on the Covariances Matrix 4931 1451 of the Evaluation Data(to be published) 4931 1451 4931 1451 ***************************************************************** 4931 1451 ********** End of (N,2N) bibliographical component ********** 4931 1451 ================================================================= 4931 1451 4931 1451 ***************************************************************** 4931 1451 ********** Start of (N,G) bibliographical component ********** 4931 1451 ***************************************************************** 4931 1451 * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * 4931 1451 * * * * * * *4931 1451 CAPTURE TO 54 MIN. ISOMERIC STATE DESCRIPTION MODIFIED BY 4931 1451 R. SCHENTER ON 2/20/84. THIS WAS DONE BY ADDING FILE9 WITH 4931 1451 Y=.79 AND PUTTING THE TOTAL CAPTURE WIDTH IN FILE2 AND THE TOTAL 4931 1451 CAPTURE CROSS SECTION IN FILE3. 4931 1451 P.C.TO NORMALIZATION+STANDARDS SUBCOMITTEE (N,G)F.SCHMITTROTH 4931 1451 MF=2 MT=151 4931 1451 MF=2 MT=102 4931 1451 * * * * * 4931 1451 4931 1451 FILE INFORMATION 4931 1451 4931 1451 MF=1 MT=451 ATOMIC MASS FROM REF(1). 4931 1451 4931 1451 MF=2 MT=151 EVALUATION OF RESOLVED RESONANCE PARAMETERS 4931 1451 BASED ON NEW BNL-325, REF(2). 4931 1451 MF=3 MT=102 VERSION-V UNRESOLVED REGION CONTAINS ADJUSTED DATA 4931 1451 SEE DOCUMENTATION 4931 1451 RADIATIVE NEUTRON CAPTURE TO THE IN 116M(54 MIN.) 4931 1451 STATE. 4931 1451 E GREATER THAN EH, 4931 1451 EVALUATION BASED ON EXPERIMENTAL DATA, REF.(3-7) 4931 1451 AND THEORETICAL CALCULATIONS, REF.(8,9). 4931 1451 E LESS THAT EH, 4931 1451 A 1/V COMPONENT WAS ADDED TO GIVE THE CORRECT 4931 1451 2200M/S CROSS SECTION TO THE 54 MIN. STATE(THE 4931 1451 2.2 SEC. STATE CROSS SECTION WAS INCLUDED). 4931 1451 RADIATIVE CAPTURE TO THE 2.2 SEC. STATE OF IN 116 WAS 4931 1451 INCLUDED AS PART OF THE CAPTURE TO THE 54 MIN. STATE 4931 1451 FOR BOTH THERMAL AND FAST ENERGIES. 4931 1451 RESULTS WERE DIVIDED BY .79 TO GIVE THE TOTAL CAP- 4931 1451 TURE CROSS SECTION IN FILE3. FILE9 COMBINED WITH 4931 1451 FILE3 IS REQUIRED TO GIVE CAPTURE TO 54 MIN. 4931 1451 ISOMERIC STATE. 4931 1451 4931 1451 2200M/S CAPTURE CROSS SECTION,BARNS TO THE 54 MIN. STATE 4931 1451 (FROM RESONANCE PARAMETERS) = 166.413 4931 1451 4931 1451 COMPUTED RESONANCE INTEGRAL =2587.345 4931 1451 4931 1451 REFERENCES. 4931 1451 1. A.H.WAPSTRA AND N.B.GOVE,NUCL. DATA TABLES,VOL.9,PART 1(1971).4931 1451 2. S.F.MUGHABGHAB AND D.I.GARBER,BNL-325,3RD ED.,VOL.1(1973) 4931 1451 3. H.A.GRENCH AND H.O.MENLOVE,PHYS.REV.,VOL.165(1968)1298. 4931 1451 4. H.O.MENLOVE,ET AL.,PHYS.REV.,VOL.163(1967)1299. 4931 1451 5. S.A.COX, PHYS.REV.,VOL.133(1964)B378. 4931 1451 6. A.E.JOHNSRUD,ET AL.,PHYS.REV.,VOL.116(1959)927. 4931 1451 7. G.PETO ET AL., J.NUCL.EN.,VOL.21(1967)797. 4931 1451 8. F.SCHMITTROTH, HEDL-TME 71-106(AUGUST 1971). 4931 1451 9. F.SCHMITTROTH, HEDL-TME 73-79,ENDF-195(NOVEMBER 1973). 4931 1451 ***************************************************************** 4931 1451 ********** (N,G) processing details ********** 4931 1451 ***************************************************************** 4931 1451 Data Linearized to Within an Accuracy of .100000000 per-cent 4931 1451 ***************** Program RECENT (VERSION 2002-1) ***************4931 1451 for All Data Greater than 1.0000E-10 barns in Absolute Value 4931 1451 Data Linearized to within an Accuracy of .100000000 per-cent 4931 1451 ***************** Program SIGMA1 (VERSION 2002-1) ***************4931 1451 Data Doppler Broadened to 300.000000 Kelvin 4931 1451 for All Data Greater than 1.0000E-10 barns in Absolute Value 4931 1451 Data Linearized to Within an Accuracy pf .100000000 per-cent 4931 1451 ***************** Program FIXUP (Version 2002-1) ****************4931 1451 Corrected ZA/AWR in All Sections-----------------------------Yes 4931 1451 Corrected Thresholds-----------------------------------------Yes 4931 1451 Extended Cross Sections to 20 MeV----------------------------No 4931 1451 Allow Cross Section Deletion---------------------------------No 4931 1451 Allow Cross Section Reconstruction---------------------------No 4931 1451 Make All Cross Sections Non-Negative-------------------------Yes 4931 1451 Delete Energies Not in Ascending Order-----------------------Yes 4931 1451 Deleted Duplicate Points-------------------------------------Yes 4931 1451 Check for Ascending MAT/MF/MT Order--------------------------Yes 4931 1451 Check for Legal MF/MT Numbers--------------------------------Yes 4931 1451 Allow Creation of Missing Sections---------------------------No 4931 1451 Allow Insertion of Energy Points-----------------------------No 4931 1451 Create Uniform Energy Grid-----------------------------------No 4931 1451 Delete Section if Cross Section =0 at All Energies-----------Yes 4931 1451 *************** PROGRAM ACTIVATE (VERSION 2002-1) ***************4931 1451 MF=10 Activation Cross Sections Defined by Combining MF=3 4931 1451 Cross Sections and MF=9 Multipliers. MF=9 Deleted. 4931 1451 ***************************************************************** 4931 1451 The metastable state component (MF/MT=10/102) was extracted from 4931 1451 the ACTIVATE output and merged with the other reactions in this 4931 1451 In-115 evaluation. MF/MT=33/102 was adapted as MF/MT=40/102. 4931 1451 ***************************************************************** 4931 1451 ********** End of (N,G) bibliographical component ********** 4931 1451 ================================================================= 4931 1451 ********** Start of (N,N') bibliographical component ********** 4931 1451 ***************************************************************** 4931 1451 ------Russian Reactor Dosimetry File RRDF-2002 4931 1451 ***************************************************************** 4931 1451 Authors of evaluation: K.I.Zolotarev and P.K.Zolotarev 4931 1451 ***************************************************************** 4931 1451 MF=3 4931 1451 MT= 51 - In115(n,n')In115m reaction 4931 1451 ------------------------------------- 4931 1451 In the section MT= 51 are given evaluated cross section data 4931 1451 of neutron inelastic excitation of the 4.486-hour isomeric state 4931 1451 in the indium-115 in the energy range from threshold to 20 MeV. 4931 1451 The isomeric level has energy 336.24 keV with spin and parity 4931 1451 1/2- [1]. 4931 1451 Microscopic experimental data [2-40] were analyzed in the 4931 1451 process of preparation of input data base for the evaluation of 4931 1451 cross sections and their uncertainty for the In-115(n,n')In-115m 4931 1451 reaction. During this procedure all experimental data if it was 4931 1451 possible were corrected to the new recommended cross section data 4931 1451 for monitor reactions used in the measurements and to the new re- 4931 1451 commended decay data. 4931 1451 Special correction was done with experimental data [5], [6], 4931 1451 [12], [24], [25] and [27]. 4931 1451 Grench and Menlove data [5] and Kimura et al. data [6] bet- 4931 1451 ween 0.5 - 0.9 MeV obtained in the measurements with T(p,n)He3 4931 1451 neutron source were renormalized to the integral of Liskien et al.4931 1451 experimental data [16] in the overlapping energy ranges. Original 4931 1451 cross section data from ref. [5] and [6] were multiplied to the 4931 1451 correction factors Fc=1.15268 and Fc=0.92486 , respectively. 4931 1451 Data of Santry and Butler [12] obtained in the absolute measu-4931 1451 rements with T(p,n)He3 source in the energy range 0.348-5.130 MeV 4931 1451 were renormalized to the integral of Liskien et al. experimental 4931 1451 data [16] in the energy interval 1.85-4.07 MeV, Fc= 1.0516 4931 1451 Cross section data measured by Zhao Wenrong et al. [24] were 4931 1451 renormalized also to the experimental data of Liskien et al. [16] 4931 1451 in the overlapping energy ranges. All original experimental data 4931 1451 from ref. [24] except cross section value at the energy point 4931 1451 En=8.5 MeV were multiplied to the factor Fc=1.065 . 4931 1451 Experimental data of Kimura et al. [6] in the energy interval 4931 1451 1.8-4.6 MeV and experimental data of Santry and Butler [12] obta- 4931 1451 ined relative to the S-32(n,p)P-32 standard were not corrected. 4931 1451 Data of Zhao Wenrong et al. [24] for the energy point 8.5 MeV 4931 1451 was corrected to the recommended cross section value for monitor 4931 1451 tor reaction Al-27(n,a)Na-24 from ref. [41]. 4931 1451 Cross section data for the In-115(n,n')In-115m reaction measu-4931 1451 red by Lu Hanlin et al. [25] and by Csikai et al. [27] were renor-4931 1451 malized to the integral of Konno et al. [30] and Filatenkov et al.4931 1451 [31] experimental data in the overlapping energy interval 13.54 - 4931 1451 14.78 MeV . Original data from ref. [25] and [27] were multiplied 4931 1451 to the correction factors Fc=1.13622 and Fc=1.08624, respectively.4931 1451 Excitation function for the In-115(n,n')In-115m reaction in 4931 1451 the energy region from threshold to 20 MeV was evaluated by means 4931 1451 of statistical analysis of experimental cross section data [2-31].4931 1451 Experimental cross section data [32-40] were rejected due to 4931 1451 their discrepancy with the main bulk of experimental data [2-31]. 4931 1451 Data of Martin et al. [2] in the energy interval 4.55 - 5.26 MeV 4931 1451 were not taken into account in the evaluation due to their syste- 4931 1451 matic underestimation In-115(n,n')In-115m reaction cross section. 4931 1451 In the rejected experiments [33-38] the cross section values were 4931 1451 measured only in a one energy point in the interval 14 - 15 MeV. 4931 1451 Statistical analysis of input cross section data was carried 4931 1451 out by means of PADE-2 code [42]. Rational function was used as 4931 1451 the model function [43]. 4931 1451 U-235 thermal fission [44] and Cf-252 spontaneous fission 4931 1451 neutron spectra [45] averaged cross sections calculated from the 4931 1451 the evaluated In-115(n,n')In-115m excitation function are the 4931 1451 following: 4931 1451 4931 1451 ÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÂÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÂÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄ 4931 1451 TYPE OF SPECTRUM ³ ,mb (calc.) ³ , mb (measured) 4931 1451 ÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÅÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÅÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄ 4931 1451 U-235 neutron fission ³ 188.40 ³ 188.2 +- 2.3 [46] 4931 1451 ÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÅÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÅÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄ 4931 1451 Cf-252 spont. fission ³ 191.66 ³ 197.4 +- 2.704 [47] 4931 1451 ÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÁÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÁÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄ 4931 1451 4931 1451 MT=33 4931 1451 MT= 51 -(n,n') cross section cov. matrix 4931 1451 ---------------------------------------- 4931 1451 Uncertainties in the evaluated excitation function for the 4931 1451 reaction In-115(n,n')In-115m are given in the form of relative 4931 1451 covariance matrix for the 39-neutron energy groups (LB=5). Cova- 4931 1451 riance matrix of uncertainties was calculated simultaneously with 4931 1451 recommended cross section data by means of PADE-2 code. 4931 1451 Eigenvalues of the 6-th digits relative covariance matrix 4931 1451 given in the 33-file are the following: 4931 1451 4931 1451 7.55059E-08 8.01529E-08 8.14328E-08 8.29955E-08 4931 1451 8.51539E-08 8.79248E-08 8.88394E-08 9.29091E-08 4931 1451 9.80564E-08 1.06991E-07 1.19646E-07 1.34061E-07 4931 1451 1.46033E-07 1.71437E-07 2.22942E-07 2.66525E-07 4931 1451 4.04740E-07 7.05147E-07 1.52953E-06 2.70059E-06 4931 1451 2.95312E-06 3.08776E-06 6.35038E-06 4.26321E-04 4931 1451 6.81158E-04 9.21024E-04 9.54240E-04 1.07694E-03 4931 1451 1.11466E-03 1.28566E-03 1.45684E-03 1.62008E-03 4931 1451 1.97248E-03 4.10930E-03 5.75040E-03 7.29359E-03 4931 1451 1.00269E-02 1.64825E-02 2.00540E-02 4931 1451 4931 1451 References: 4931 1451 1. R.B.Firestone Table of Isotopes, Eighth edition, Vol. 1, 4931 1451 John Wiley & Sons, Inc., New York, 1995 4931 1451 2. H.C.Martin et al. Phys. Rev., v.93, p.199, January 1954 4931 1451 3. H.O.Menlove, K.L.Coop, H.A.Grench, R.Sher Phys. Rev., v.163, 4931 1451 p.1299, 1967 4931 1451 4. P.Bornemisza-Pauspertl et al. Atomki Koezlemenyek, v.10, No.2,4931 1451 p.112, July 1968 4931 1451 5. H.A.Grench, H.O.Menlove Physical Review, v.165, p.1298, 1968 4931 1451 6. I.Kimura et al. Journal of Nucl. Science and Technology, v.6,4931 1451 p.485, September 1969 4931 1451 7. R.C.Barrall et al. Report AFWL-TR-68-134, Albuquerque, NM, 4931 1451 March 1969 ; 4931 1451 R.C.Barrall et al. Nucl. Phys. A, v.138, p.387, December 1969 4931 1451 8. J.K.Temperley, D.E.Barnes Report BRL-R-1491, August 1970 4931 1451 9. L.Lakosi, A.Veres 2nd Nat'l Soviet Conf. on Neutron Physics, 4931 1451 Kiev, v.4, p.312, May 1973 4931 1451 10. K.Kobayashi et al. J. Nucl. Energ., v.27, p.741, October 1973 4931 1451 11. D.L.Smith, J.W.Meadows Report ANL-NDM-14, Argonne, July 1975; 4931 1451 D.L.Smith, J.W.Meadows Nucl. Sci. Eng., v.60, p.319,July 1976 4931 1451 12. D.C.Santry, J.P.Butler Can. J. Phys., v.54, p.757, 1976 4931 1451 13. C.G.Hudson, W.L.Alford Bull. Amer. Phys. Soc., v.21, 4931 1451 p.188(DB5), February 1976 4931 1451 14. G.Magnusson, I.Bergqvist Nuclear Technology, v.34, p.114, 4931 1451 June 1977 4931 1451 15. S.Yamamoto et al. Report NEANDC(J)-56, p.25, September 1978 4931 1451 16. H.Liskien et al. Nucl. Sci. Eng., v.67, p.334, September 1978 4931 1451 17. P.Andersson et al. Report LUNF-D6-3021, University of Lund, 4931 1451 November 1978 4931 1451 18. C.F.Ai, J.C.Chou Nuclear Science, Taiwan, v.16, no.3, p.157, 4931 1451 September 1979 4931 1451 19. L.Adamski et al. Annals of Nucl. Energy, v.7, p.397, 1980 4931 1451 20. Fan Pei-Guo et al. Chinese J. of Nuclear Physics, v.2, No.4, 4931 1451 p.337, November 1980 4931 1451 21. V.L.Demekhin et al. 6th All Union Conference on Neutron 4931 1451 Physics, Kiev, 2 - 6 October, v.3, p.195, 1983 4931 1451 22. T.B.Ryves et al. J. of Physics, pt.G, v.9, p.1549, Dec. 1983 4931 1451 23. K.Kudo et al. Report NEANDC(J)-106/U, p.1, September 1984 4931 1451 24. Zhao Wenrong, Lu Hanlin et al. Report INDC(CPR)-16, Sep. 1989 4931 1451 25. Lu Hanlin et al. Chinese J. of Nuclear Physics, v.11, n.2, 4931 1451 p.53, 1989 4931 1451 26. I.Kimura, K.Kobayashi Nucl. Sci. Eng., v.106, p.332, 1990 4931 1451 27. J.Csikai et al. Zeitschrift fuer Physik, v.A337, p.39, 1990 4931 1451 28. N.N.Moiseev et al. Atomnaja Energija, v.68, February 1990 4931 1451 29. Y.Ikeda et al. Proc. of an Int. Conf. Nuclear Data for Sci- 4931 1451 ence and Technology, Julich, FRG, 13-17 May 1991, Springer- 4931 1451 Verlag, 1992, pp.294-296 4931 1451 30. C.Konno et al. Report JAERI-1329, October 1993 4931 1451 31. A.A.Filatenkov et al. Report RI-252, St. Petesburg, May 1999; 4931 1451 A.A.Filatenkov et al. VANT, Ser.:Yadernye Konstanty, v.2, p.8,4931 1451 Moscow, 1996 4931 1451 32. S.G.Cohen Nature, v.161, p.475, March 1948 4931 1451 33. W.Nagel EXFOR 20198.016 , December 1966 4931 1451 34. H.Roetzer Nucl. Phys., v.A109, p.694, March 1968 4931 1451 35. B.Minetti, A.Pasquarelli Zeitschrift f. Physik, v.217, p.83, 4931 1451 October 1968 4931 1451 36. I.Garlea et al. Rev. Roum. Phys., v.29, p.421, 1984 4931 1451 37. R.Pepelnik et al. Report NEANDC(E)-262U, No.5,p.32, June 1985 4931 1451 38. I.Garlea et al. Rev. Roum. Phys., v.37, no.1, pp.19-25, 1992 4931 1451 39. J.Csikai, P.Raics EXFOR 30983.002 July 1992 4931 1451 40. N.I.Molla et al. EXFOR 30984.002 4931 1451 41. H.Vonach in "Nuclear Data Standards for Nuclear Measurements",4931 1451 Report NEANDC-311 U, pp.75-77, OECD, Paris, 1992. ; 4931 1451 M.Wagner et al. Evaluation of cross sections for 14 important 4931 1451 neutron-dosimetry reactions. Physics Data No.13-5, 1990 4931 1451 42. S.A.Badikov et al. Preprint FEI-1686, Obninsk, 1985 4931 1451 43. S.Badikov, N.Rabotnov, K.Zolotarev Proc. of NEANSC Speciali- 4931 1451 st's Meeting on Evaluation and Processing of Covariance Data, 4931 1451 Oak Ridge, USA, 1992, OECD, Paris, 1993, p.105 4931 1451 44. L.W.Weston et al. Evaluated Neutron Data for Uranium-235, 4931 1451 ENDF/B-VI Library, MAT=9228, MF=5, MT=18, eval. April 1989 4931 1451 45. W.Mannhart IAEA-TECDOC-410, p.158, IAEA, Vienna, 1987 4931 1451 46. W.Mannhart Prog. Report INDC(Ger)-045, pp.40-43, June 1999 4931 1451 47. W.Mannhart Validation of Differential Cross Sections with 4931 1451 Integral Data , Report INDC(NDS)-435, pp.59-64, IAEA, Vienna, 4931 1451 September 2002 4931 1451 ***************************************************************** 4931 1451 ********** End of (N,N') bibliographical component ********** 4931 1451 ================================================================= 4931 1451 4931 1451 4931 1451 4931 1451 4931 1451 ***************************************************************** 4931 1451 ***************** Program GROUPIE (VERSION 2002-1) **************4931 1451 Unshielded Group Averages Using 640 Groups 4931 1451 Weighting Spectrum: Flat (Constant) Spectrum 4931 1451 1 451 417 14931 1451 8 4 2 14931 1451 8 16 2 14931 1451 8 102 2 14931 1451 10 4 75 14931 1451 10 16 40 14931 1451 10 102 217 14931 1451 40 4 148 14931 1451 40 16 24 14931 1451 40 102 7 04931 1451 4931 1 0 4931 0 0 4.91150E+4 1.13917E+2 0 0 1 04931 3291 0.0 -3.36240E+5 49115 0 1 2144931 3291 214 1 4931 3291 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1.308320-6 1.906180-4 3.619560-4 3.374430-4 1.279230-4-2.160140-4493133291 2.044200-3 2.034690-3 1.822330-3 1.400220-3 8.333560-4 2.891650-4493133291 -4.530750-5 3.594830-5 9.252940-6 2.722010-5 1.214320-4 1.685150-4493133291 1.254710-4-2.967280-6 2.278240-3 2.346400-3 2.144150-3 1.617530-3493133291 8.591030-4 2.100590-5-3.725150-5 1.991280-5-1.808000-4-2.312720-4493133291 -1.228560-4 5.691170-5 2.535410-4 2.803450-3 3.009470-3 2.750750-3493133291 1.945570-3 5.171010-4-6.650430-5 2.771530-5-3.310770-4-5.477510-4493133291 -4.324660-4-7.477900-5 4.251560-4 3.769220-3 4.055680-3 3.524510-3493133291 1.623270-3 2.112930-5 2.413140-5-2.769420-4-5.905430-4-5.680570-4493133291 -2.218830-4 3.548060-4 5.097970-3 5.247340-3 3.288620-3 2.938230-4493133291 3.478540-6 1.013570-4-1.266150-4-3.055100-4-2.789910-4-4.752080-5493133291 6.316630-3 4.933610-3 7.344400-4-2.188660-5 7.292020-4 8.080040-4493133291 4.010620-4-1.342080-4-6.368090-4 4.945260-3 1.213690-3 3.634430-5493133291 1.177520-3 1.652270-3 1.220020-3 2.789710-4-8.913250-4 8.698960-4493133291 3.475700-4 3.534930-4 4.186150-4 3.550390-4 1.773520-4-7.331360-5493133291 5.153350-4 3.975890-4 2.332660-4 1.091980-4 2.823100-5-2.163150-5493133291 1.382620-3 1.844060-3 1.642420-3 1.011630-3 1.406180-4 2.928410-3493133291 3.028540-3 2.435580-3 1.422320-3 3.697970-3 3.780810-3 3.486050-3493133291 4.941200-3 5.924580-3 8.514690-3 493133291 493133 0 4.91150E+4 1.13917E+2 0 0 0 1493133292 0.0000E+00 0.0000E+00 0 292 0 1493133292 0.0000E+00 0.0000E+00 1 5 120 15493133292 1.000000-5 9.100000+6 1.050000+7 1.150000+7 1.220000+7 1.270000+7493133292 1.320000+7 1.380000+7 1.420000+7 1.470000+7 1.520000+7 1.600000+7493133292 1.720000+7 1.800000+7 2.000000+7 4.000000-1 1.154500-2 2.886700-3493133292 5.207500-4 6.268700-4 2.634000-4 7.576400-5 0.000000+0 0.000000+0493133292 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0 1.360500-3493133292 2.329300-4 4.379300-5 3.313900-5 2.375700-5 6.429400-6 0.000000+0493133292 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0493133292 4.273200-4 5.149000-5 5.671700-5 3.970200-5 1.611800-5 0.000000+0493133292 5.134000-6 5.330100-6 0.000000+0 3.850900-6 2.755800-7 0.000000+0493133292 1.326100-4 7.026400-5 5.057000-5 1.814000-5 1.025800-5 1.091900-5493133292 1.151900-5 6.054300-6 4.032200-6 3.183000-6 2.577000-7 1.061600-4493133292 7.193900-5 1.827700-5 7.806100-6 1.063900-5 1.195100-5 9.623500-6493133292 6.619700-6 5.063800-6 1.300400-6 9.380800-5 2.002300-5 8.788300-6493133292 1.212000-5 1.349000-5 1.062800-5 7.704500-6 6.688400-6 1.473800-6493133292 5.145600-5 2.270400-5 1.436100-5 1.686000-5 4.715200-6 4.452700-6493133292 4.233200-6 1.200700-6 4.804900-5 1.232200-5 1.347800-5 3.765900-6493133292 2.757000-6 3.388400-6 4.405400-7 4.707100-5 2.134000-5 8.565900-6493133292 1.578500-5 9.719500-6 9.119700-6 4.580100-5 1.385300-5 1.154500-5493133292 1.754500-5 3.156100-6 4.554700-5 1.336600-5 2.469400-5 2.724500-6493133292 4.622500-5 2.217600-5 1.374800-5 7.898800-5 5.926800-6 8.012400-5493133292 493133 0 4.91150E+4 1.13917E+2 0 0 0 1493133293 0.0000E+00 0.0000E+00 0 293 0 1493133293 0.0000E+00 0.0000E+00 0 1 14 7493133293 1.000000-5 3.600000-3 1.000000+3 2.250000-2 1.000000+4 8.100000-3493133293 1.000000+5 2.500000-3 1.000000+6 4.900000-3 4.000000+6 1.690000-2493133293 2.000000+7 0.000000+0 493133293 493133 0 4931 0 0 0 0 0 5.31270E+4 1.25814E+2 0 0 34 105325 1451 0.0 0.0 0 0 0 65325 1451 1.00000E+0 2.00000E+7 0 0 10 20025325 1451 3.00000E+2 0.0 1 0 60 35325 1451 53-I -127 CNDC/CIAE EVAL-JUN91 Zhao Wenrong and Lu Hanlin et.al.5325 1451 DIST-Feb2004 5325 1451 ----IRDF-2002 MATERIAL 5325 5325 1451 -----INCIDENT NEUTRON DATA 5325 1451 ------ENDF-6 FORMAT 5325 1451 * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * 5325 1451 53-I -127 CNDC/CIAE EVAL-JUN91 Zhao Wenrong and Lu Hanlin et.al.5325 1451 ---ENDF-6 FORMAT 5325 REVISION 0 5325 1451 * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * 5325 1451 5325 1451 Results of the evaluation in ENDF/B Format 5325 1451 53-I-127 5325 1451 I-127(n,2n)I-126 Dosimetry reaction 5325 1451 Evaluated by Zhao Wenrong Lu Hanlin Yu Weixiang Cai Dunjiu 5325 1451 Wang Zisheng Huang Xiaolong 5325 1451 5325 1451 The reaction of I-127(n,2n)I-126 is used as a threshold detector 5325 1451 for neutron dosimetry applications. Its threshold is 9.137 5325 1451 MeV. The half-life of I-126 is 13.023d. Iodine is a monoisotopic 5325 1451 element. The main gamma ray energies of I-126 are 388.6keV 5325 1451 and 666.4keV with branch ratios of 34.1 percent and 33.1 percent 5325 1451 respectively. In order to get reasonable results, the necessary 5325 1451 adjustment was made first for the collected data. Then the 5325 1451 evaluation was performed by spline fitting procedure. The 5325 1451 uncertainties were derived from experimental errors and 5325 1451 the consideration of systematics. 5325 1451 5325 1451 Evaluation documentation: Zhao Wenrong. 5325 1451 Report CIAE 1991 5325 1451 * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * 5325 1451 The Q values and threshold energies were updated prior to pro- 5325 1451 cessing through the codes to comply with the values obtained 5325 1451 using the NNDC calculation program which is based on the 1995 5325 1451 Update to the Atomic mass Evaluation. 5325 1451 * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * 5325 1451 ***************** Program LINEAR (VERSION 2002-1) ***************5325 1451 For All Data Greater than 1.0000E-10 barns in Absolute Value 5325 1451 Data Linearized to Within an Accuracy of .100000000 per-cent 5325 1451 ***************** Program SIGMA1 (VERSION 2002-1) ***************5325 1451 Data Doppler Broadened to 300.000000 Kelvin 5325 1451 for All Data Greater than 1.0000E-10 barns in Absolute Value 5325 1451 Data Linearized to Within an Accuracy pf .100000000 per-cent 5325 1451 ***************** Program FIXUP (Version 2002-1) ****************5325 1451 Corrected ZA/AWR in All Sections-----------------------------Yes 5325 1451 Corrected Thresholds-----------------------------------------Yes 5325 1451 Extended Cross Sections to 20 MeV----------------------------No 5325 1451 Allow Cross Section Deletion---------------------------------No 5325 1451 Allow Cross Section Reconstruction---------------------------No 5325 1451 Make All Cross Sections Non-Negative-------------------------Yes 5325 1451 Delete Energies Not in Ascending Order-----------------------Yes 5325 1451 Deleted Duplicate Points-------------------------------------Yes 5325 1451 Check for Ascending MAT/MF/MT Order--------------------------Yes 5325 1451 Check for Legal MF/MT Numbers--------------------------------Yes 5325 1451 Allow Creation of Missing Sections---------------------------No 5325 1451 Allow Insertion of Energy Points-----------------------------No 5325 1451 Create Uniform Energy Grid-----------------------------------No 5325 1451 Delete Section if Cross Section =0 at All Energies-----------Yes 5325 1451 ***************** Program GROUPIE (VERSION 2002-1) **************5325 1451 Unshielded Group Averages Using 640 Groups 5325 1451 Weighting Spectrum: Flat (Constant) Spectrum 5325 1451 1 451 67 15325 1451 3 16 40 15325 1451 33 16 19 15325 1451 5325 1 0 5325 0 0 5.31270E+4 1.25814E+2 0 0 1 05325 3 16 -9.14300E+6-9.14300E+6 0 0 1 1095325 3 16 109 1 5325 3 16 9200000.00 .002125030 9300000.00 .003744262 9400000.00 .0058736845325 3 16 9500000.00 .008981316 9600000.00 .031000000 9700000.00 .0660600005325 3 16 9800000.00 .108200000 9900000.00 .155240000 10000000.0 .2070000005325 3 16 10100000.0 .262032500 10200000.0 .320900000 10300000.0 .3824372375325 3 16 10400000.0 .448368421 10500000.0 .517276842 10600000.0 .5849000005325 3 16 10700000.0 .651900000 10800000.0 .718900000 10900000.0 .7858500005325 3 16 11000000.0 .851150000 11100000.0 .915570000 11200000.0 .9773500005325 3 16 11300000.0 1.03776632 11400000.0 1.09542105 11500000.0 1.151272635325 3 16 11600000.0 1.19980000 11700000.0 1.24544000 11800000.0 1.285400005325 3 16 11900000.0 1.32310750 12000000.0 1.35620000 12100000.0 1.387452505325 3 16 12200000.0 1.41480000 12300000.0 1.44060105 12400000.0 1.463263165325 3 16 12500000.0 1.48443579 12600000.0 1.50185000 12700000.0 1.518110005325 3 16 12800000.0 1.53215000 12900000.0 1.54537000 13000000.0 1.556000005325 3 16 13100000.0 1.56600000 13200000.0 1.57600000 13300000.0 1.585794875325 3 16 13400000.0 1.59369231 13500000.0 1.60112821 13600000.0 1.608564105325 3 16 13700000.0 1.61572676 13800000.0 1.62125000 13900000.0 1.626500005325 3 16 14000000.0 1.63175000 14100000.0 1.63679830 14200000.0 1.640636365325 3 16 14300000.0 1.64427273 14400000.0 1.64790909 14500000.0 1.651545455325 3 16 14600000.0 1.65518182 14700000.0 1.65826506 14800000.0 1.660795185325 3 16 14900000.0 1.66332530 15000000.0 1.66585542 15100000.0 1.668385545325 3 16 15200000.0 1.67091566 15300000.0 1.67344578 15400000.0 1.675975905325 3 16 15500000.0 1.67817198 15600000.0 1.67940000 15700000.0 1.680566675325 3 16 15800000.0 1.68173333 15900000.0 1.68290000 16000000.0 1.684066675325 3 16 16100000.0 1.68494750 16200000.0 1.68500000 16300000.0 1.684748725325 3 16 16400000.0 1.68376923 16500000.0 1.68274359 16600000.0 1.681717955325 3 16 16700000.0 1.68006051 16800000.0 1.67710000 16900000.0 1.674100005325 3 16 17000000.0 1.67110000 17100000.0 1.66762000 17200000.0 1.663150005325 3 16 17300000.0 1.65795667 17400000.0 1.65133333 17500000.0 1.644666675325 3 16 17600000.0 1.63800000 17700000.0 1.63028708 17800000.0 1.621050005325 3 16 17900000.0 1.61180000 18000000.0 1.60255000 18100000.0 1.592388755325 3 16 18200000.0 1.58090000 18300000.0 1.56824212 18400000.0 1.553897445325 3 16 18500000.0 1.53953846 18600000.0 1.52517949 18700000.0 1.509750005325 3 16 18800000.0 1.49325000 18900000.0 1.47550000 19000000.0 1.456500005325 3 16 19100000.0 1.43675000 19200000.0 1.41625000 19300000.0 1.393692315325 3 16 19400000.0 1.36907692 19500000.0 1.34446154 19600000.0 1.319880525325 3 16 19700000.0 1.29935484 19800000.0 1.28161290 19900000.0 1.263870975325 3 16 20000000.0 0.0 5325 3 16 5325 3 0 5325 0 0 5.31270E+4 1.25814E+2 0 0 1 1532533 16 0.000000+0 0.000000+0 0 16 0 1532533 16 0.000000+0 0.000000+0 1 5 91 13532533 16 1.000000-5 9.200000+6 1.058000+7 1.118000+7 1.157000+7 1.237000+7532533 16 1.355000+7 1.415000+7 1.454000+7 1.514000+7 1.712000+7 1.811000+7532533 16 1.989000+7 1.000000-1 2.189400-3 1.465900-3 8.864000-4 6.104300-4532533 16 5.784600-4 5.628800-4 4.006100-4 5.706500-4 2.236400-4 3.653400-4532533 16 3.326600-4 2.851600-4 6.491500-5 3.944500-5 2.716400-5 2.613800-5532533 16 2.527600-5 1.805000-5 2.583600-5 9.952100-6 1.661900-5 1.522700-5532533 16 1.278300-4 2.641000-5 1.818700-5 1.750000-5 1.692300-5 1.208500-5532533 16 1.729800-5 6.663300-6 1.112700-5 1.019500-5 6.062500-5 1.953900-5532533 16 2.045100-5 8.819600-6 9.864000-6 7.941700-6 3.976900-6 1.965700-5532533 16 1.911200-5 4.140300-5 1.991900-5 1.119300-5 9.425200-6 1.161100-5532533 16 1.011200-5 1.294000-5 1.031500-5 2.200000-5 9.677000-6 7.798900-6532533 16 1.012000-5 7.187100-6 7.827200-6 6.343900-6 1.636500-5 1.249200-5532533 16 1.206100-5 9.854400-6 9.520400-6 8.511300-6 1.567300-5 1.273500-5532533 16 7.466100-6 8.300400-6 6.147800-6 1.539000-5 1.140600-5 1.040700-5532533 16 8.057200-6 1.543700-5 1.378500-5 8.856800-6 1.830900-5 1.393200-5532533 16 2.509400-5 532533 16 532533 0 5325 0 0 0 0 0 5.71390E+4 1.37713E+2 0 0 34 105728 1451 0.0 0.0 0 0 0 65728 1451 1.00000E+0 2.00000E+7 0 0 10 20025728 1451 3.00000E+2 0.0 1 0 200 35728 1451 57-La-139 FEI EVAL-SEP01 K.I.Zolotarev 5728 1451 DIST-Feb2004 5728 1451 ----IRDF-2002 MATERIAL 5728 5728 1451 -----INCIDENT NEUTRON DATA 5728 1451 ------ENDF-6 FORMAT 5728 1451 ***************************************************************** 5728 1451 57-La-139 FEI EVAL-Sep01 K.I.Zolotarev 5728 1451 DIST-Jan04 5728 1451 ----BROND-3 MATERIAL 5728 Revision 1, Jan. 2004 5728 1451 -----INCIDENT NEUTRON DATA 5728 1451 ------ENDF-6 FORMAT 5728 1451 ***************************************************************** 5728 1451 ------Russian Reactor Dosimetry File RRDF-2002 5728 1451 ***************************************************************** 5728 1451 Authors of evaluation: K.I.Zolotarev and V.G.Pronyaev 5728 1451 ***************************************************************** 5728 1451 5728 1451 ----- MF=2 MT=151 ----- 5728 1451 5728 1451 RESONANCE PARAMETERS 5728 1451 1. Resolved Resonance Region (RRR) 5728 1451 Resolved MLBW resonance parameters up to 20 keV were based 5728 1451 on data from ref.[1-2]. Spins of resonances, which were not iden- 5728 1451 tified, have been assigned by taking into account 2*J+1 law and 5728 1451 approximate linear growth of the number of levels with given spin 5728 1451 and parity from energy. Average gamma-widths 0.055 eV (L=0) and 5728 1451 0.04 eV (L=1) were assigned to those resonances , widths for 5728 1451 which were absent. Fictitious P-resonances were added in the 5728 1451 energy region (3 - 20) keV to obtain the average capture cross 5728 1451 section values observed in the experiments [3],[4],[5]. 5728 1451 5728 1451 2. Unresolved Resonance Region (URR) 5728 1451 Unresolved resonance parameters in the region 20 keV - 160 keV 5728 1451 were evaluated by using 'EVPAR' code (Hauser-Feshbach-Moldauer 5728 1451 statistical model) [6]. URR parameters were calculated from opti- 5728 1451 mized fit to the measured capture [4,5] and total cross section 5728 1451 [7,8,9] data. Average resolved resonance parameters have been 5728 1451 used as zero approximation. Contribution of F-wave in the energy 5728 1451 interval 60 - 160 keV given as smoothed background cross section 5728 1451 in File 3. 5728 1451 5728 1451 ----- MF=3 MT=102 ----- 5728 1451 5728 1451 Capture cross sections from 1.0E-5 eV to 160 keV are reconst- 5728 1451 ructed from evaluated MLBW resolved and unresolved resonance pa- 5728 1451 rameters. Small background was added in RRR for taking into 5728 1451 account non-statistical reaction mechanism contribution. Contri- 5728 1451 bution of F-wave in the energy interval (60 - 160) keV was added 5728 1451 as described above. 5728 1451 Data base for the evaluation La-139(n,g)La-140 excitation func-5728 1451 tion in the neutron energy range 160 keV - 20 MeV was formed from 5728 1451 microscopic experimental data [4-5], [10-28] and data received 5728 1451 from theoretical model calculations. Experimental data [4], [10], 5728 1451 [12-18], [20-22], [27-28] were corrected to the new standards.5728 1451 Uncertainties for cross section data measured by Johnsrud et al.5728 1451 [13] were evaluated between +-(16-19)%. Theoretical model calcula-5728 1451 tions was done by means of GNASH code [29]. New modified version 5728 1451 of GNASH was utilized for calculations [30]. Data from theoreti-5728 1451 cal model calculation were dominant in the neutron energy range 5728 1451 3.5 - 20 MeV. 5728 1451 The evaluation La-139(n,g)La-140 excitation function in the 5728 1451 energy range above 160 keV has been carried out within the frame- 5728 1451 work of generalized least squares method , rational function was 5728 1451 used as model function [31]. Procedure of calculation recommended 5728 1451 cross section data was performed by means of PADE-2 code [32]. 5728 1451 Capture cross section at En=0.0253 eV and Resonance Integral 5728 1451 (0.5 to 2.0E+7 eV) from present evaluation are given below in the 5728 1451 Table 1. in a comparison with data from compilations [1], [33-34].5728 1451 Table 1. 5728 1451 ------------------------------------------------- 5728 1451 Parameters Values, barn References 5728 1451 ------------------------------------------------- 5728 1451 Capture cross section 9.04 +- 0.23 This eval. 5728 1451 at En=0.0253 eV 8.93 +- 0.04 [ 1] 5728 1451 9.04 +- 0.04 [33] 5728 1451 9.2 +- 0.2 [34] 5728 1451 ------------------------------------------------- 5728 1451 Resonance Integral 12.1 This eval. 5728 1451 11.8 +- 0.8 [ 1] 5728 1451 12.1 +- 0.6 [33] 5728 1451 11.8 +- 0.8 [34] 5728 1451 ------------------------------------------------- 5728 1451 5728 1451 ----- MF=33 MT=102----- 5728 1451 Uncertainties in the evaluated excitation function for the 5728 1451 reaction La-139(n,g)La-140 are given in the three independent 5728 1451 matrixes. 5728 1451 In the energy range 1.000E-05 - 200 eV uncertainties are pre- 5728 1451 sented in the form of relative covariance matrix for the 16-neut- 5728 1451 ron energy groups (LB=5). Eigenvalues of this 6-th digits relat- 5728 1451 ive covariance matrix are the following: 5728 1451 5728 1451 1.29984E-04 1.80010E-04 2.41627E-04 2.69103E-04 5728 1451 3.03943E-04 3.53897E-04 4.83948E-04 5.72815E-04 5728 1451 6.17194E-04 6.90136E-04 9.46937E-04 1.01148E-03 5728 1451 1.61882E-03 1.76694E-03 1.98803E-03 2.89526E-02 5728 1451 5728 1451 In the energy range 0.2 - 20 keV uncertainties are given in 5728 1451 the form of diagonal matrix of uncertainties for 7-th neutron 5728 1451 energy intervals (LB=1) 5728 1451 In the energy range 20 keV - 20 MeV uncertainties are presen- 5728 1451 ted in the form of relative covariance matrix for the 23-neutron 5728 1451 energy groups (LB=5). Eigenvalues of this 6-th digits relative 5728 1451 covariance matrix are the following: 5728 1451 5728 1451 1.96223E-10 5.86297E-08 3.57093E-07 1.00052E-05 5728 1451 3.95061E-05 1.50141E-04 2.46548E-04 3.34259E-04 5728 1451 4.59370E-04 6.23219E-04 7.85798E-04 8.75929E-04 5728 1451 1.31621E-03 1.62913E-03 1.81439E-03 1.87643E-03 5728 1451 2.55332E-03 2.70062E-03 3.40954E-03 4.32177E-03 5728 1451 1.13231E-02 3.04915E-02 1.18873E-01 5728 1451 5728 1451 References : 5728 1451 1. S.F.Mughabghab et al. Neutron Cross Sections ,Vol.1,Part A, 5728 1451 New York, Academic Press,1981 5728 1451 2. S.I.Sukhoruchkin et al. Landolt Bornstein New Series, v.I/16B,5728 1451 ed. H.Schopper, Springer, 1998, pp.w19-w23 5728 1451 3. V.A.Konks, Ju.P.Popov, F.L.Shapiro Zurnal Eksperim. i Teor. 5728 1451 Fiziki, (USSR), v.46,(1), p.80, 1963 5728 1451 4. D.C.Stupegia et al. J. Nucl. Energ., v.22, p.267, May 1968 5728 1451 5. B.J.Allen, J.W.Boldeman, R.L.Macklin Nucl. Sci. Eng., v.82, 5728 1451 n.2, p.230, October 1982 5728 1451 6. G.N.Manturov et al. Voprosy Atomnoy Nauki i Tekhniki, Ser.: 5728 1451 Jadernye Konstanty, v.1, p.50, 1983 5728 1451 7. M.Divadeenam et al. Diss. Abstr. B , 1968, v.28, p.3834, 1968 5728 1451 8. E.Islam et al. Nucl. Phys. A, v.209, p.189, 1973 5728 1451 9. K.Nishimura et al. Report JAERI-M-6883, January 1977 5728 1451 10. R.L.Macklin et al. Phys. Rev., v.107, p.504, 1957 5728 1451 11. J.L.Perkin et al. Proc. Phys. Soc., v.72, p.505, 1958 5728 1451 12. W.S.Lyon, R.L.Macklin Phys. Rev., v.114, p.1619, 1959 5728 1451 13. A.E.Johnsrud et al. Phys. Rev., v.116, p.927, 1959 5728 1451 14. R.G.Wille, R.W.Fink Phys. Rev., v.118, p.242, 1960 5728 1451 15. J.H.Gibbons et al. Phys. Rev., v.122, p.182, 1961 5728 1451 16. J.Csikai et al. J. Atomki Kozlemenyek, v.8, p.79, June 1966 5728 1451 J.Csikai et al. Nucl. Phys. A, v.95, p.229, March 1967 5728 1451 17. A.K.Chaubey, M.L.Sehgal Phys. Rev., v.152, p.1055, Dec. 1966 5728 1451 18. G.Peto et al. J. Nucl. Energ., v.21, p.797, October 1967 5728 1451 19. F.Rigaud et al. Nucl. Phys., v.A176, p.545, December 1971 5728 1451 20. G.G.Zaikin et al. Ukrainskij Fizichnij Zhurnal, v.16(7), 5728 1451 p.1205, July 1971 5728 1451 21. F.Rigaud et al. Nucl. Sci. Eng., v.55, p.17, September 1974 5728 1451 22. O.Schwerer et al. Nucl. Phys. A, v.264, p.105, June 1976 5728 1451 23. A.R.Del.Musgrove, B.J.Allen, R.L.Macklin Proc. of Int. Conf. 5728 1451 on Neutron Physics and Nuclear Data for Reactors and Other 5728 1451 Applied Purposes, AERE Harwell, 25-29 September 1978, p.449 5728 1451 24. W.P.Poenitz Progress Report ANL-83-4, p.239, April 1982 5728 1451 25. J.Voignier et al. Nucl. Sci. Eng., v.93, p.43, 1986 5728 1451 26. H.Beer Progress Report NEANDC(E)-272U,(5), p.8, June 1986 5728 1451 27. Y.N.Trofimov Voprosy Atomnoy Nauki i Tekhniki, Serija: 5728 1451 Jadernye Konstanty, v.4, p.10, 1987 5728 1451 28. Y.N.Trofimov Proc. of the 1-st International Conf. on Neutron 5728 1451 Physics, Kiev, 14-18 September 1987, v.3, p.331, Moscow 1988 5728 1451 29. P.G.Young, E.D.Arthur A Preequilibrium Statistical Nuclear 5728 1451 Model Code for Calculation of Cross Section and Emission 5728 1451 Spectra. Report LA-6947, Los Alamos, 1977 5728 1451 30. E.L.Trykov, G.Ya.Tertychnyi Private communication, IPPE, 5728 1451 Obninsk, May 1999 5728 1451 31. S.Badikov, N.Rabotnov, K.Zolotarev Proc. of NEANSC Speciali- 5728 1451 st's Meeting on Evaluation and Processing of Covariance Data, 5728 1451 Oak Ridge, USA, 7-9 September 1992, OECD, Paris, 1993, p.105 5728 1451 32. S.A.Badikov et al. Preprint FEI-1686, Obninsk, 1985 5728 1451 33. S.F.Mughabghab et al. Termal Neutron Capture Cross Sections 5728 1451 Resonance Integrals and G-factors, Report INDC(NDS)-440, 5728 1451 IAEA, Vienna, February 2003 5728 1451 34. A.V.Ignatyuk et al. Landolt Bornstein New Series, v.I/16A, 5728 1451 Part 1, ed. H.Schopper, Springer, 1998, p.8-18 5728 1451 ***************************************************************** 5728 1451 The total and elastic cross-section files gave double counting 5728 1451 of the cross-sections in the resonance region, these two files 5728 1451 were modified for this library. The total and elastic cross- 5728 1451 section files were taken from ENDF/B-VI Release 1. 5728 1451 ***************************************************************** 5728 1451 ***************** Program LINEAR (VERSION 2002-1) ***************5728 1451 For All Data Greater than 1.0000E-10 barns in Absolute Value 5728 1451 Data Linearized to Within an Accuracy of .100000000 per-cent 5728 1451 ***************** Program RECENT (VERSION 2002-1) ***************5728 1451 for All Data Greater than 1.0000E-10 barns in Absolute Value 5728 1451 Data Linearized to within an Accuracy of .100000000 per-cent 5728 1451 ***************** Program SIGMA1 (VERSION 2002-1) ***************5728 1451 Data Doppler Broadened to 300.000000 Kelvin 5728 1451 for All Data Greater than 1.0000E-10 barns in Absolute Value 5728 1451 Data Linearized to Within an Accuracy pf .100000000 per-cent 5728 1451 ***************** Program FIXUP (Version 2002-1) ****************5728 1451 Corrected ZA/AWR in All Sections-----------------------------Yes 5728 1451 Corrected Thresholds-----------------------------------------Yes 5728 1451 Extended Cross Sections to 20 MeV----------------------------No 5728 1451 Allow Cross Section Deletion---------------------------------No 5728 1451 Allow Cross Section Reconstruction---------------------------No 5728 1451 Make All Cross Sections Non-Negative-------------------------Yes 5728 1451 Delete Energies Not in Ascending Order-----------------------Yes 5728 1451 Deleted Duplicate Points-------------------------------------Yes 5728 1451 Check for Ascending MAT/MF/MT Order--------------------------Yes 5728 1451 Check for Legal MF/MT Numbers--------------------------------Yes 5728 1451 Allow Creation of Missing Sections---------------------------No 5728 1451 Allow Insertion of Energy Points-----------------------------No 5728 1451 Create Uniform Energy Grid-----------------------------------No 5728 1451 Delete Section if Cross Section =0 at All Energies-----------Yes 5728 1451 ***************** Program GROUPIE (VERSION 2002-1) **************5728 1451 Unshielded Group Averages Using 640 Groups 5728 1451 Weighting Spectrum: Flat (Constant) Spectrum 5728 1451 1 451 207 15728 1451 3 102 217 15728 1451 33 102 93 15728 1451 5728 1 0 5728 0 0 5.71390E+4 1.37713E+2 0 0 0 05728 3102 5.16110E+6 5.16110E+6 0 0 1 6415728 3102 641 1 5728 3102 .000100000 140.916130 .000105000 137.631355 .000110000 134.5107985728 3102 .000115000 131.635331 .000120000 128.273838 .000127500 124.5443475728 3102 .000135000 121.137008 .000142500 118.005771 .000150000 114.6169365728 3102 .000160000 111.086689 .000170000 107.874636 .000180000 104.9262495728 3102 .000190000 102.183048 .000200000 99.6784848 .000210000 97.32212995728 3102 .000220000 95.1334787 .000230000 93.0999779 .000240000 90.71308485728 3102 .000255000 88.0902320 .000270000 86.0663388 .000280000 83.81339225728 3102 .000300000 81.0658839 .000320000 78.5781377 .000340000 76.29978725728 3102 .000360000 74.2013117 .000380000 72.2857027 .000400000 70.28859095728 3102 .000425000 68.2449527 .000450000 66.3776940 .000475000 64.66584655728 3102 .000500000 63.0560178 .000525000 61.5847291 .000550000 60.19486285728 3102 .000575000 58.9022361 .000600000 57.5773841 .000630000 56.21446155728 3102 .000660000 54.9673858 .000690000 53.7710963 .000720000 52.49819395728 3102 .000760000 51.1333567 .000800000 49.8661477 .000840000 48.70483885728 3102 .000880000 47.6010855 .000920000 46.5892455 .000960000 45.62022485728 3102 .001000000 44.6158143 .001050000 43.5673071 .001100000 42.58438595728 3102 .001150000 41.6787045 .001200000 40.6103452 .001275000 39.43618005728 3102 .001350000 38.3627031 .001425000 37.3608905 .001500000 36.29659065728 3102 .001600000 35.1837725 .001700000 34.1678926 .001800000 33.22832805728 3102 .001900000 32.3699820 .002000000 31.5741079 .002100000 30.82725945728 3102 .002200000 30.1432970 .002300000 29.4892640 .002400000 28.74300675728 3102 .002550000 27.9137684 .002700000 27.2651423 .002800000 26.56036255728 3102 .003000000 25.6927903 .003200000 24.9033544 .003400000 24.18153895728 3102 .003600000 23.5233241 .003800000 22.9156450 .004000000 22.28233845728 3102 .004250000 21.6394275 .004500000 21.0525874 .004750000 20.50405645728 3102 .005000000 20.0005519 .005250000 19.5336671 .005500000 19.09221685728 3102 .005750000 18.6875024 .006000000 18.2624172 .006300000 17.83749075728 3102 .006600000 17.4354503 .006900000 17.0627040 .007200000 16.65762145728 3102 .007600000 16.2246636 .008000000 15.8289941 .008400000 15.45792825728 3102 .008800000 15.1129611 .009200000 14.7923717 .009600000 14.48729375728 3102 .010000000 14.1711393 .010500000 13.8366769 .011000000 13.53026115728 3102 .011500000 13.2378410 .012000000 12.9026382 .012750000 12.53202545728 3102 .013500000 12.1890031 .014250000 11.8743787 .015000000 11.53881425728 3102 .016000001 11.1865003 .017000001 10.8633443 .017999999 10.56820955728 3102 .018999999 10.2967876 .020000000 10.0420336 .021000000 9.808730435728 3102 .022000000 9.58877507 .023000000 9.38366445 .024000000 9.142476865728 3102 .025500000 8.88195969 .027000001 8.67540086 .028000001 8.450706555728 3102 .029999999 8.17372671 .032000002 7.92099010 .034000002 7.689985845728 3102 .035999998 7.47901504 .037999999 7.28440094 .039999999 7.081330265728 3102 .042500000 6.87522787 .045000002 6.68700537 .047499999 6.511801765728 3102 .050000001 6.35088582 .052499998 6.20203607 .055000000 6.061114085728 3102 .057500001 5.93184964 .059999999 5.79580688 .063000001 5.659533175728 3102 .066000000 5.53106507 .068999998 5.41136298 .071999997 5.281554185728 3102 .075999998 5.14250084 .079999998 5.01545054 .083999999 4.896032725728 3102 .088000000 4.78499161 .092000000 4.68226134 .096000001 4.584216205728 3102 .100000001 4.48324276 .104999997 4.37641057 .109999999 4.278466245728 3102 .115000002 4.18477460 .119999997 4.07719395 .127499998 3.958265675728 3102 .135000005 3.84797071 .142499998 3.74673540 .150000006 3.638811005728 3102 .159999996 3.52534063 .170000002 3.42151836 .180000007 3.326642955728 3102 .189999998 3.23955461 .200000003 3.15757224 .209999993 3.082848935728 3102 .219999999 3.01500584 .230000004 2.94815628 .239999995 2.871692795728 3102 .254999995 2.78666076 .270000011 2.72163373 .280000001 2.649409865728 3102 .300000012 2.56098805 .319999993 2.48088377 .340000004 2.408141305728 3102 .360000014 2.34167369 .379999995 2.27990170 .400000006 2.216861425728 3102 .425000012 2.15320686 .449999988 2.09521825 .474999994 2.043085085728 3102 .500000000 1.99748613 .524999976 1.95603723 .550000012 1.922719395728 3102 .574999988 1.89911055 .600000024 1.89200884 .629999995 1.936167425728 3102 .660000026 2.16687669 .689999998 2.97714428 .720000029 3.585621135728 3102 .759999990 2.36696288 .800000012 1.77176305 .839999974 1.606425365728 3102 .879999995 1.52575725 .920000017 1.47129338 .959999979 1.427912665728 3102 1.00000000 1.38690059 1.04999995 1.34698658 1.10000002 1.311277615728 3102 1.14999998 1.27807038 1.20000005 1.24089857 1.27499998 1.199740075728 3102 1.35000002 1.16237991 1.42499995 1.12842344 1.50000000 1.091833595728 3102 1.60000002 1.05355434 1.70000005 1.01871964 1.79999995 .9868971405728 3102 1.89999998 .957195969 2.00000000 .929820007 2.09999990 .9045599205728 3102 2.20000005 .880519178 2.29999995 .858515706 2.40000010 .8324045995728 3102 2.54999995 .803586799 2.70000005 .781502901 2.79999995 .7566564095728 3102 3.00000000 .726321306 3.20000005 .698717025 3.40000010 .6735123945728 3102 3.59999990 .650113662 3.79999995 .628541218 4.00000000 .6062990895728 3102 4.25000000 .583326845 4.50000000 .562202761 4.75000000 .5427964895728 3102 5.00000000 .524548155 5.25000000 .507547920 5.50000000 .4919011885728 3102 5.75000000 .476953842 6.00000000 .461844036 6.30000019 .4460978025728 3102 6.59999990 .431653795 6.90000010 .417874545 7.19999981 .4031649715728 3102 7.59999990 .387421500 8.00000000 .372904078 8.39999962 .3591470755728 3102 8.80000019 .346393405 9.19999981 .334728830 9.60000038 .3234366485728 3102 10.0000000 .311655490 10.5000000 .299489725 11.0000000 .2881600605728 3102 11.5000000 .277598486 12.0000000 .265340267 12.7500000 .2518482985728 3102 13.5000000 .239496295 14.2500000 .228159008 15.0000000 .2160590055728 3102 16.0000000 .203554474 17.0000000 .192244019 18.0000000 .1819503825728 3102 19.0000000 .172599573 20.0000000 .164032558 21.0000000 .1561493165728 3102 22.0000000 .148866331 23.0000000 .142145255 24.0000000 .1345352775728 3102 25.5000000 .126274331 27.0000000 .120040369 28.0000000 .1132982805728 3102 30.0000000 .105381402 32.0000000 .098493094 34.0000000 .0925493295728 3102 36.0000000 .087415552 38.0000000 .083031106 40.0000000 .0789771265728 3102 42.5000000 .075518280 45.0000000 .073173495 47.5000000 .0721248275728 3102 50.0000000 .072673156 52.5000000 .075468582 55.0000000 .0816682855728 3102 57.5000000 .093733114 60.0000000 .121278817 63.0000000 .1964426155728 3102 66.0000000 .490808755 69.0000000 38.8167148 72.0000000 111.1919955728 3102 76.0000000 .304919017 80.0000000 .111430450 84.0000000 .0641543615728 3102 88.0000000 .044870529 92.0000000 .034798476 96.0000000 .0286624405728 3102 100.000000 .024124733 105.000000 .020618111 110.000000 .0180647125728 3102 115.000000 .016096787 120.000000 .014192109 127.500000 .0124010315728 3102 135.000000 .010992289 142.500000 .009849928 150.000000 .0087673255728 3102 160.000000 .007756086 170.000000 .006928042 180.000000 .0062457545728 3102 190.000000 .005669140 200.000000 .005180843 210.000000 .0047636395728 3102 220.000000 .004408720 230.000000 .004138393 240.000000 .1080888905728 3102 255.000000 .003454009 270.000000 .003153382 280.000000 .0028934845728 3102 300.000000 .002615505 320.000000 .065215765 340.000000 .0061120735728 3102 360.000000 .002028551 380.000000 .001885322 400.000000 .0017555255728 3102 425.000000 .001636953 450.000000 .001541719 475.000000 .0014665725728 3102 500.000000 .001412820 525.000000 .001392752 550.000000 .0014553985728 3102 575.000000 .002027849 600.000000 2.07704896 630.000000 .0022773975728 3102 660.000000 .001569452 690.000000 .640595066 720.000000 .0012771835728 3102 760.000000 .001170377 800.000000 .001296779 840.000000 .9758432225728 3102 880.000000 .307883942 920.000000 .004246978 960.000000 .7127309865728 3102 1000.00000 .002136035 1050.00000 .003638613 1100.00000 .0134021445728 3102 1150.00000 1.60569954 1200.00000 .516360996 1275.00000 .0024041835728 3102 1350.00000 .003761734 1425.00000 .136001840 1500.00000 .0009020785728 3102 1600.00000 .374778034 1700.00000 .000940109 1800.00000 .0632213855728 3102 1900.00000 .161078382 2000.00000 .017721336 2100.00000 1.140632905728 3102 2200.00000 .003805450 2300.00000 .250131212 2400.00000 .3266767325728 3102 2550.00000 .105724204 2700.00000 .001786788 2800.00000 .2871656745728 3102 3000.00000 .078072363 3200.00000 .211893608 3400.00000 .2297869445728 3102 3600.00000 .362415424 3800.00000 .091087950 4000.00000 .0710547765728 3102 4250.00000 .185614814 4500.00000 .274804507 4750.00000 .1063477185728 3102 5000.00000 .090699012 5250.00000 .166010721 5500.00000 .0889003525728 3102 5750.00000 .263216372 6000.00000 .062719930 6300.00000 .1870963565728 3102 6600.00000 .094966934 6900.00000 .231505410 7200.00000 .1428706705728 3102 7600.00000 .083189902 8000.00000 .135855675 8400.00000 .1308038955728 3102 8800.00000 .079543336 9200.00000 .057390732 9600.00000 .1698050635728 3102 10000.0000 .095655986 10500.0000 .076190286 11000.0000 .0605745495728 3102 11500.0000 .099569645 12000.0000 .091689585 12750.0000 .0743117505728 3102 13500.0000 .059266584 14250.0000 .075003631 15000.0000 .0496516195728 3102 16000.0000 .065352560 17000.0000 .060008600 18000.0000 .0492534465728 3102 19000.0000 .046632142 20000.0000 .053979643 21000.0000 .0523238465728 3102 22000.0000 .050783408 23000.0000 .049344663 24000.0000 .0476748405728 3102 25500.0000 .045824588 27000.0000 .044399469 28000.0000 .0428424855728 3102 30000.0000 .040906568 32000.0000 .039149431 34000.0000 .0375525075728 3102 36000.0000 .036092256 38000.0000 .034751876 40000.0000 .0333649775728 3102 42500.0000 .031957755 45000.0000 .030675222 47500.0000 .0295016015728 3102 50000.0000 .028434482 52500.0000 .027462854 55000.0000 .0265665975728 3102 57500.0000 .025737611 60000.0000 .024878936 63000.0000 .0240020325728 3102 66000.0000 .023205451 69000.0000 .022446899 72000.0000 .0216415645728 3102 76000.0000 .020797765 80000.0000 .020041074 84000.0000 .0193688115728 3102 88000.0000 .018758350 92000.0000 .018163677 96000.0000 .0176087655728 3102 100000.000 .017053673 105000.000 .016509385 110000.000 .0160154425728 3102 115000.000 .015566041 120000.000 .015066526 127500.000 .0145490835728 3102 135000.000 .014099105 142500.000 .013666508 150000.000 .0132054995728 3102 160000.000 .012550055 170000.000 .011703463 180000.000 .0109078935728 3102 190000.000 .010272784 200000.000 .009794410 210000.000 .0094180525728 3102 220000.000 .009107929 230000.000 .008844160 240000.000 .0085619245728 3102 255000.000 .008270502 270000.000 .008061967 280000.000 .0078451465728 3102 300000.000 .007596099 320000.000 .007385542 340000.000 .0072139015728 3102 360000.000 .007070845 380000.000 .006951513 400000.000 .0068385595728 3102 425000.000 .006742198 450000.000 .006672964 475000.000 .0066163595728 3102 500000.000 .006579159 525000.000 .006561226 550000.000 .0065507515728 3102 575000.000 .006556875 600000.000 .006572463 630000.000 .0065982085728 3102 660000.000 .006638844 690000.000 .006681843 720000.000 .0067492445728 3102 760000.000 .006830836 800000.000 .006928983 840000.000 .0070320165728 3102 880000.000 .007149792 920000.000 .007279764 960000.000 .0074257055728 3102 1000000.00 .007753469 1100000.00 .008363678 1200000.00 .0090764805728 3102 1300000.00 .009508900 1400000.00 .009378977 1500000.00 .0088884055728 3102 1600000.00 .008338516 1700000.00 .007841272 1800000.00 .0074041655728 3102 1900000.00 .007013814 2000000.00 .006658618 2100000.00 .0063311835728 3102 2200000.00 .006027052 2300000.00 .005743452 2400000.00 .0054785075728 3102 2500000.00 .005230797 2600000.00 .005001541 2700000.00 .0047847695728 3102 2800000.00 .004581968 2900000.00 .004391955 3000000.00 .0042142745728 3102 3100000.00 .004047834 3200000.00 .003891622 3300000.00 .0037452535728 3102 3400000.00 .003607809 3500000.00 .003478471 3600000.00 .0033563605728 3102 3700000.00 .003242869 3800000.00 .003134944 3900000.00 .0030314815728 3102 4000000.00 .002936409 4100000.00 .002845264 4200000.00 .0027587345728 3102 4300000.00 .002676819 4400000.00 .002598874 4500000.00 .0025248995728 3102 4600000.00 .002454351 4700000.00 .002387231 4800000.00 .0023230815728 3102 4900000.00 .002261901 5000000.00 .002203299 5100000.00 .0021472745728 3102 5200000.00 .002093494 5300000.00 .002041959 5400000.00 .0019923895728 3102 5500000.00 .001944784 5600000.00 .001898896 5700000.00 .0018547265728 3102 5800000.00 .001812066 5900000.00 .001770916 6000000.00 .0017314885728 3102 6100000.00 .001693783 6200000.00 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18200000.0 .000375874 18300000.0 .0003649435728 3102 18400000.0 .000354170 18500000.0 .000343988 18600000.0 .0003343975728 3102 18700000.0 .000324944 18800000.0 .000316315 18900000.0 .0003078235728 3102 19000000.0 .000299834 19100000.0 .000292347 19200000.0 .0002849755728 3102 19300000.0 .000278285 19400000.0 .000271708 19500000.0 .0002655395728 3102 19600000.0 .000259778 19700000.0 .000254108 19800000.0 .0002489825728 3102 19900000.0 .000243947 20000000.0 0.0 5728 3102 5728 3 0 5728 0 0 5.71390E+4 1.37713E+2 0 0 0 1572833102 0.000000+0 0.000000+0 0 102 0 3572833102 0.000000+0 0.000000+0 1 5 171 18572833102 1.000000-5 1.000000-4 1.000000-3 1.000000-2 2.000000-2 3.000000-2572833102 1.000000-1 3.000000-1 5.000000-1 1.000000+0 3.000000+0 1.000000+1572833102 3.000000+1 6.500000+1 8.000000+1 1.275000+2 2.000000+2 2.000000+7572833102 3.767710-3 2.067300-3 2.139120-3 1.031770-3 6.641270-4 1.917110-3572833102 1.953300-3 1.918890-3 2.400260-3 2.000230-3 1.928820-3 1.972660-3572833102 1.644780-3 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1.601770-3 2.429640-3 1.576590-3572833102 1.668390-3 0.000000+0 2.175610-3 2.372830-3 1.886870-3 1.806130-3572833102 1.868630-3 1.545720-3 2.359900-3 1.537840-3 1.596370-3 0.000000+0572833102 3.401930-3 2.444360-3 2.279290-3 2.370740-3 1.958590-3 2.989980-3572833102 1.947030-3 2.026610-3 0.000000+0 2.267330-3 1.989240-3 1.949760-3572833102 1.647100-3 2.494680-3 1.617180-3 1.718630-3 0.000000+0 2.152470-3572833102 1.957590-3 1.558050-3 2.394310-3 1.564730-3 1.603350-3 0.000000+0572833102 2.316060-3 1.674220-3 2.457340-3 1.587250-3 1.713230-3 0.000000+0572833102 1.593640-3 2.044400-3 1.329110-3 1.370750-3 0.000000+0 4.144240-3572833102 2.022380-3 2.119610-3 0.000000+0 1.549840-3 1.413310-3 0.000000+0572833102 2.109240-3 0.000000+0 0.000000+0 572833102 0.000000+0 0.000000+0 0 1 20 10572833102 1.000000-5 0.000000+0 2.000000+2 2.345650-1 3.000000+2 3.374160-1572833102 5.000000+2 2.215830-1 1.000000+3 1.670580-1 2.200000+3 9.646030-2572833102 4.750000+3 1.530000-1 1.000000+4 2.445570-1 2.000000+4 0.000000+0572833102 2.000000+7 0.000000+0 572833102 0.000000+0 0.000000+0 1 5 325 25572833102 1.000000-5 2.000000+4 4.000000+4 6.000000+4 8.000000+4 1.000000+5572833102 2.000000+5 4.000000+5 6.000000+5 8.000000+5 1.000000+6 2.000000+6572833102 3.000000+6 4.000000+6 5.000000+6 6.000000+6 7.000000+6 8.000000+6572833102 9.000000+6 1.000000+7 1.200000+7 1.400000+7 1.600000+7 1.800000+7572833102 2.000000+7 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0572833102 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0572833102 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0572833102 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0572833102 0.000000+0 4.373210-3 4.257650-3 4.190390-3 4.171110-3 4.148810-3572833102 4.161440-3 3.619020-3 2.457010-3 1.445370-3 1.531360-3 8.031520-4572833102 1.043540-3 1.200720-3 1.215670-3 1.172440-3 1.144140-3 1.175140-3572833102 1.194820-3 1.182890-3 1.164400-3 1.307070-3 1.656590-3 1.955650-3572833102 4.480110-3 4.497000-3 4.428800-3 4.177620-3 4.221330-3 3.656530-3572833102 2.681970-3 1.606600-3 1.534450-3 8.608040-4 1.089460-3 1.307050-3572833102 1.346700-3 1.277400-3 1.206560-3 1.240720-3 1.298720-3 1.259650-3572833102 1.263580-3 1.392600-3 1.800410-3 2.077550-3 4.599630-3 4.605670-3572833102 4.281150-3 4.258790-3 3.657810-3 2.547890-3 1.545390-3 1.541850-3572833102 8.321630-4 1.060020-3 1.239290-3 1.268460-3 1.219390-3 1.174940-3572833102 1.203660-3 1.237920-3 1.216770-3 1.206490-3 1.344290-3 1.716880-3572833102 2.010090-3 4.742920-3 4.459620-3 4.291190-3 3.680760-3 2.327280-3572833102 1.405120-3 1.562540-3 7.833310-4 1.022270-3 1.137270-3 1.142630-3572833102 1.123670-3 1.124920-3 1.148600-3 1.141660-3 1.150330-3 1.115430-3572833102 1.269810-3 1.584080-3 1.905780-3 4.605220-3 4.273810-3 3.718160-3572833102 2.442330-3 1.420210-3 1.576280-3 8.058040-4 1.059450-3 1.197410-3572833102 1.202570-3 1.168140-3 1.156000-3 1.186970-3 1.191920-3 1.190400-3572833102 1.162340-3 1.315120-3 1.652630-3 1.970050-3 4.650030-3 3.785000-3572833102 2.297160-3 1.362930-3 1.607150-3 8.019660-4 1.059440-3 1.177090-3572833102 1.176780-3 1.153250-3 1.155170-3 1.183480-3 1.175920-3 1.183830-3572833102 1.148000-3 1.307290-3 1.631080-3 1.960810-3 4.191990-3 3.716890-3572833102 2.220880-3 1.586800-3 9.805650-4 1.200080-3 1.483940-3 1.552120-3572833102 1.462450-3 1.354480-3 1.388990-3 1.478980-3 1.419310-3 1.438570-3572833102 1.569140-3 2.050230-3 2.338900-3 6.013730-3 5.613730-3 1.648130-3572833102 1.136680-3 1.104900-3 1.402660-3 1.605800-3 1.597240-3 1.462240-3572833102 1.416670-3 1.530810-3 1.479160-3 1.507310-3 1.624740-3 2.135960-3572833102 2.441640-3 9.596400-3 2.815990-3 9.693480-4 1.103800-3 1.023690-3572833102 1.074000-3 1.221770-3 1.341960-3 1.284610-3 1.173230-3 1.279090-3572833102 1.172100-3 1.398410-3 1.645660-3 2.135440-3 5.787750-3 1.657310-3572833102 2.205260-3 2.617220-3 2.669340-3 2.556840-3 2.464170-3 2.529390-3572833102 2.598890-3 2.555700-3 2.531740-3 2.824340-3 3.602280-3 4.223850-3572833102 3.633110-3 4.267780-3 4.641970-3 4.800960-3 4.840430-3 4.849650-3572833102 4.869290-3 4.846410-3 4.898520-3 4.754970-3 5.395380-3 6.744660-3572833102 8.112480-3 5.966690-3 6.739140-3 6.708790-3 6.302550-3 5.970150-3572833102 6.109290-3 6.347980-3 6.211780-3 6.184360-3 6.864580-3 8.790700-3572833102 1.027570-2 8.063440-3 8.221160-3 7.668950-3 7.008490-3 6.952500-3572833102 7.356010-3 7.186020-3 7.197940-3 7.924890-3 1.020480-2 1.189920-2572833102 8.717380-3 8.454870-3 7.820360-3 7.406320-3 7.592980-3 7.632630-3572833102 7.502130-3 8.380310-3 1.061400-2 1.263480-2 8.701350-3 8.438740-3572833102 7.794530-3 7.432500-3 7.718140-3 7.386790-3 8.453560-3 1.047490-2572833102 1.273760-2 8.742840-3 8.359810-3 7.410060-3 7.578710-3 7.242570-3572833102 8.353630-3 1.032520-2 1.250930-2 8.829350-3 8.074660-3 7.385510-3572833102 7.413820-3 8.264620-3 1.053090-2 1.236150-2 8.774640-3 7.717270-3572833102 7.483530-3 8.327420-3 1.057710-2 1.255790-2 8.949390-3 7.595650-3572833102 8.383470-3 1.059340-2 1.256530-2 8.513730-3 8.420100-3 1.022820-2572833102 1.243340-2 1.077640-2 1.217860-2 1.364410-2 1.682810-2 1.804110-2572833102 2.429660-2 572833102 572833 0 5728 0 0 0 0 0 5.91410E+4 1.39697E+2 0 0 34 105925 1451 0.0 0.0 0 0 0 65925 1451 1.00000E+0 2.00000E+7 0 0 10 20025925 1451 3.00000E+2 0.0 1 0 122 35925 1451 59-Pr-141 FEI EVAL-Jan97 K.I.Zolotarev et al. 5925 1451 DIST-Feb2004 5925 1451 ----IRDF-2002 MATERIAL 5925 5925 1451 -----INCIDENT NEUTRON DATA 5925 1451 ------ENDF-6 FORMAT 5925 1451 ***************************************************************** 5925 1451 59-PR-141 FEI EVAL-Jan97 K.I.Zolotarev et al. 5925 1451 DIST-Jan02 20020123 5925 1451 ----BROND-2 MATERIAL 5925 5925 1451 -----INCIDENT NEUTRON DATA 5925 1451 ------ENDF-6 FORMAT 5925 1451 ------Russian Reactor Dosimetry File RRDF-2002 5925 1451 ***************************************************************** 5925 1451 Authors of evaluation: K.Zolotarev, V.Manokhin, A.Pashchenko 5925 1451 ***************************************************************** 5925 1451 MF= 3 5925 1451 MT= 16 - Pr-141(n,2n)Pr-140 reaction 5925 1451 -------------------------------------- 5925 1451 Excitation function for the Pr-141(n,2n)Pr-140 reaction in 5925 1451 the energy region from threshold to 20 MeV was evaluated by means 5925 1451 of statistical analysis of experimental cross section data [1-19] 5925 1451 and data from systematics of (n,2n) excitation function [20] in 5925 1451 the energy regions 9.5 - 12.0 MeV and 18.0 - 20.0 MeV. 5925 1451 All experimental data were renormalized to the new standards 5925 1451 for monitor reactions cross sections and decay data. 5925 1451 The final procedure of evaluation has been carried out within 5925 1451 the framework of generalized least squares method. Rational func- 5925 1451 tion was used as model function [21]. Calculations was performed 5925 1451 by means of Pade-2 code [22]. 5925 1451 U-235 thermal fission [23] and Cf-252 spontaneous fission 5925 1451 neutron spectra [24] averaged cross-sections calculated from the 5925 1451 evaluated Pr141(n,2n)Pr140 excitation function are the following: 5925 1451 5925 1451 -------------------------------------------- 5925 1451 TYPE OF SPECTRUM I CS>, mb (calc.) 5925 1451 --------------------------I----------------- 5925 1451 U-235 neutron fission I 1.0922 5925 1451 CF-252 spontan. fission I 1.9843 5925 1451 5925 1451 MF=33 5925 1451 MT= 16 -(n,2n) cross section cov. matrix 5925 1451 ---------------------------------------- 5925 1451 Uncertainties in the evaluated excitation function for the re-5925 1451 action Pr-141(n,2n)Pr-140 are given in the form of relative cova- 5925 1451 riance matrix for the 18-neutron energy groups (LB=5). Covariance 5925 1451 matrix of uncertainties was calculated simultaneously with 5925 1451 recommended cross section data by means of PADE-2 code. 5925 1451 Eigenvalues of the 6-th digits relative covariance matrix 5925 1451 given in the 33-file are the following: 5925 1451 5925 1451 6.47369E-06 6.50089E-06 6.56924E-06 6.70314E-06 5925 1451 6.90792E-06 7.20594E-06 7.64163E-06 8.30022E-06 5925 1451 9.35563E-06 1.12079E-05 1.49213E-05 2.36154E-05 5925 1451 4.48092E-05 3.28699E-04 6.81059E-03 1.70941E-02 5925 1451 2.55798E-02 1.53270E-01 5925 1451 5925 1451 References : 5925 1451 1. E.B.Paul, R.L.Clarke Canadian J. Phys., v.31,p.267, 1953 5925 1451 2. J.M.Ferguson, W.E.Thompson Phys. Rev., v.118, p.228, 1960 5925 1451 3. R.G.Wille, R.W.Fink Phys. Rev., v.118, p.242, 1960 5925 1451 4. L.A.Rayburn Phys. Rev., v.122, p.168, 1961 5925 1451 5. C.S.Khurana, H.S.Hans Nucl. Phys., v.28, p.560, 1961 5925 1451 6. D.R.Koehler, W.L.Alford Report NP-11667, 1962 5925 1451 7. M.Cevolani, S.Petralia Nucl. Sci. Eng., v.26, p.1328, 1962 5925 1451 8. L.A.Rayburn Bull. American Phys. Soc., v.8, p.60, Jan. 1963 5925 1451 9. M.P.Menon, M.Y.Cuypers Phys. Rev., v.156, p.1340, 1967 5925 1451 10. P.Cuzzocrea e.a. Nuovo Cimento, Sec.B, v.52, n.2, p.476, 1967 5925 1451 11. M.Bormann e.a. Nucl. Phys., Sec.A, v.115, p.309, July 1968 5925 1451 12. G.Peto e.a. Acta Physica Hung., v.24, p.93, April 1968 5925 1451 13. A.Chatterjee e.a. Proc. of 12th Nucl. Phys. and Solid State 5925 1451 Phys. Sympos., Roorkee, India, 28-31 December 1969, v.2,p.117 5925 1451 14. A.Bari Dissertation Abstracts, sec.B, v.32, p.5091, 1972 5925 1451 15. J.Araminowicz, J.Dresler Prog. Rep. INR-1464, p.14, May 1973 5925 1451 16. R.A.Sigg, P.K.Kuroda Inorg. Nucl. Chem., v.37, p.631, 1975 5925 1451 17. M.Valkonen Report JU-RR-1/1976, Jyvaeskylae University, 1976 5925 1451 18. S.Murahira et al. Report INDC(JPN)-175/U, p.171, Nov. 1995 5925 1451 19. Y.Kasugai, Y.Ikeda, Y.Uno Proc. of Int. Conf. on Nuclear Data 5925 1451 for Science and Technology, Trieste, Italy, 19-24 May 1997, 5925 1451 v.1, p.635 5925 1451 20. V.N.Manokhin Report INDC(CCP)-397, IAEA, Vienna, 1997 5925 1451 21. S.Badikov, N.Rabotnov, K.Zolotarev Proc. of NEANSC Speciali- 5925 1451 st's Meeting on Evaluation and Processing of Covariance Data, 5925 1451 Oak Ridge, USA, 7-9 September 1992, OECD, Paris, 1993, p.105 5925 1451 22. S.A.Badikov et al. Preprint FEI-1686, Obninsk, 1985 5925 1451 23. L.W.Weston et al. Evaluated Neutron Data for U-235, ENDF/B-VI 5925 1451 Library, MAT=9228, MF=5, MT=18, eval. April 1989 5925 1451 24. W.Mannhart IAEA-TECDOC-410, p.158, Vienna, 1987 5925 1451 ******************************************************************5925 1451 The Q values and threshold energies were updated prior to pro- 5925 1451 cessing through the codes to comply with the values obtained 5925 1451 using the NNDC calculation program which is based on the 1995 5925 1451 Update to the Atomic mass Evaluation. 5925 1451 5925 1451 File 2 added to the pointwise file containing only the effective 5925 1451 scattering radius with no resonance parameters given. 5925 1451 Taken from ENDF/B-VI 5925 1451 ******************************************************************5925 1451 ***************** Program LINEAR (VERSION 2002-1) ***************5925 1451 For All Data Greater than 1.0000E-10 barns in Absolute Value 5925 1451 Data Linearized to Within an Accuracy of .100000000 per-cent 5925 1451 ***************** Program SIGMA1 (VERSION 2002-1) ***************5925 1451 Data Doppler Broadened to 300.000000 Kelvin 5925 1451 for All Data Greater than 1.0000E-10 barns in Absolute Value 5925 1451 Data Linearized to Within an Accuracy pf .100000000 per-cent 5925 1451 ***************** Program FIXUP (Version 2002-1) ****************5925 1451 Corrected ZA/AWR in All Sections-----------------------------Yes 5925 1451 Corrected Thresholds-----------------------------------------Yes 5925 1451 Extended Cross Sections to 20 MeV----------------------------No 5925 1451 Allow Cross Section Deletion---------------------------------No 5925 1451 Allow Cross Section Reconstruction---------------------------No 5925 1451 Make All Cross Sections Non-Negative-------------------------Yes 5925 1451 Delete Energies Not in Ascending Order-----------------------Yes 5925 1451 Deleted Duplicate Points-------------------------------------Yes 5925 1451 Check for Ascending MAT/MF/MT Order--------------------------Yes 5925 1451 Check for Legal MF/MT Numbers--------------------------------Yes 5925 1451 Allow Creation of Missing Sections---------------------------No 5925 1451 Allow Insertion of Energy Points-----------------------------No 5925 1451 Create Uniform Energy Grid-----------------------------------No 5925 1451 Delete Section if Cross Section =0 at All Energies-----------Yes 5925 1451 ***************** Program GROUPIE (VERSION 2002-1) **************5925 1451 Unshielded Group Averages Using 640 Groups 5925 1451 Weighting Spectrum: Flat (Constant) Spectrum 5925 1451 1 451 129 15925 1451 3 16 39 15925 1451 33 16 38 15925 1451 5925 1 0 5925 0 0 5.91410E+4 1.39697E+2 0 0 0 05925 3 16 -9.39700E+6-9.39700E+6 0 0 1 1075925 3 16 107 1 5925 3 16 9400000.00 .000102039 9500000.00 .004115304 9600000.00 .0164385955925 3 16 9700000.00 .038856950 9800000.00 .070499450 9900000.00 .1102883505925 3 16 10000000.0 .157013000 10100000.0 .209402000 10200000.0 .2661885005925 3 16 10300000.0 .326164000 10400000.0 .388315750 10500000.0 .4513412505925 3 16 10600000.0 .514366750 10700000.0 .577392250 10800000.0 .6395045005925 3 16 10900000.0 .699850500 11000000.0 .758343500 11100000.0 .8147340005925 3 16 11200000.0 .868854500 11300000.0 .920604500 11400000.0 .9699370005925 3 16 11500000.0 1.01684800 11600000.0 1.06136500 11700000.0 1.102987505925 3 16 11800000.0 1.14290250 11900000.0 1.18068500 12000000.0 1.216335005925 3 16 12100000.0 1.25004500 12200000.0 1.28181500 12300000.0 1.311847505925 3 16 12400000.0 1.34014250 12500000.0 1.36689000 12600000.0 1.392090005925 3 16 12700000.0 1.41561500 12800000.0 1.43746500 12900000.0 1.459315005925 3 16 13000000.0 1.47946167 13100000.0 1.49790500 13200000.0 1.516348335925 3 16 13300000.0 1.53338000 13400000.0 1.54900000 13500000.0 1.564620005925 3 16 13600000.0 1.57889750 13700000.0 1.59183250 13800000.0 1.604767505925 3 16 13900000.0 1.61770250 14000000.0 1.62927600 14100000.0 1.639488005925 3 16 14200000.0 1.64970000 14300000.0 1.65991200 14400000.0 1.670124005925 3 16 14500000.0 1.67916600 14600000.0 1.68703800 14700000.0 1.694910005925 3 16 14800000.0 1.70278200 14900000.0 1.71065400 15000000.0 1.717540835925 3 16 15100000.0 1.72344250 15200000.0 1.72934417 15300000.0 1.735245835925 3 16 15400000.0 1.74114750 15500000.0 1.74704917 15600000.0 1.752039295925 3 16 15700000.0 1.75611786 15800000.0 1.76019643 15900000.0 1.764275005925 3 16 16000000.0 1.76835357 16100000.0 1.77243214 16200000.0 1.776510715925 3 16 16300000.0 1.77971063 16400000.0 1.78203188 16500000.0 1.784353135925 3 16 16600000.0 1.78667438 16700000.0 1.78899563 16800000.0 1.791316885925 3 16 16900000.0 1.79363813 17000000.0 1.79595938 17100000.0 1.797471675925 3 16 17200000.0 1.79817500 17300000.0 1.79887833 17400000.0 1.799581675925 3 16 17500000.0 1.80028500 17600000.0 1.80098833 17700000.0 1.800834175925 3 16 17800000.0 1.79982250 17900000.0 1.79881083 18000000.0 1.797799175925 3 16 18100000.0 1.79678750 18200000.0 1.79577583 18300000.0 1.793587005925 3 16 18400000.0 1.79022100 18500000.0 1.78685500 18600000.0 1.783489005925 3 16 18700000.0 1.78012300 18800000.0 1.77536333 18900000.0 1.769210005925 3 16 19000000.0 1.76305667 19100000.0 1.75515667 19200000.0 1.745510005925 3 16 19300000.0 1.73586333 19400000.0 1.72389250 19500000.0 1.709597505925 3 16 19600000.0 1.69321500 19700000.0 1.67284000 19800000.0 1.648080005925 3 16 19900000.0 1.61751500 20000000.0 0.0 5925 3 16 5925 3 0 5925 0 0 5.91410E+4 1.39697E+2 0 0 0 1592533 16 0.000000+0 0.000000+0 0 16 0 1592533 16 0.000000+0 0.000000+0 1 5 210 20592533 16 1.000000-5 9.400000+6 1.050000+7 1.100000+7 1.150000+7 1.200000+7592533 16 1.250000+7 1.300000+7 1.350000+7 1.400000+7 1.450000+7 1.500000+7592533 16 1.550000+7 1.600000+7 1.650000+7 1.700000+7 1.750000+7 1.800000+7592533 16 1.850000+7 1.900000+7 0.000000+0 0.000000+0 0.000000+0 0.000000+0592533 16 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0592533 16 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0592533 16 0.000000+0 0.000000+0 0.000000+0 7.629710-2 5.270640-2 3.376320-2592533 16 2.074520-2 1.252350-2 7.563810-3 4.706180-3 3.188710-3 2.533760-3592533 16 2.446920-3 2.749080-3 3.334790-3 4.148410-3 5.171510-3 6.419360-3592533 16 7.945970-3 9.861990-3 1.237550-2 4.016490-2 2.809550-2 1.900490-2592533 16 1.265780-2 8.366950-3 5.519320-3 3.674800-3 2.539500-3 1.923500-3592533 16 1.707930-3 1.823370-3 2.237300-3 2.948350-3 3.986210-3 5.417190-3592533 16 7.354640-3 9.962330-3 2.145550-2 1.571000-2 1.127530-2 7.970050-3592533 16 5.552600-3 3.813640-3 2.597590-3 1.796650-3 1.340690-3 1.189230-3592533 16 1.326580-3 1.760550-3 2.524460-3 3.682770-3 5.338130-3 7.621620-3592533 16 1.236030-2 9.413210-3 7.029290-3 5.152190-3 3.707030-3 2.623120-3592533 16 1.845220-3 1.335170-3 1.071090-3 1.046710-3 1.272260-3 1.777090-3592533 16 2.614640-3 3.867430-3 5.635080-3 7.582390-3 5.943050-3 4.583860-3592533 16 3.484360-3 2.620050-3 1.966010-3 1.503080-3 1.220170-3 1.115420-3592533 16 1.197700-3 1.489140-3 2.029030-3 2.877480-3 4.106360-3 4.916750-3592533 16 3.993780-3 3.224900-3 2.597250-3 2.101980-3 1.730500-3 1.477540-3592533 16 1.342690-3 1.331920-3 1.459470-3 1.750620-3 2.244570-3 2.989940-3592533 16 3.456170-3 2.967610-3 2.561130-3 2.226210-3 1.959650-3 1.758980-3592533 16 1.623750-3 1.556590-3 1.564460-3 1.660280-3 1.865240-3 2.208980-3592533 16 2.737410-3 2.515350-3 2.327170-3 2.163680-3 2.023420-3 1.905210-3592533 16 1.808580-3 1.734270-3 1.685040-3 1.667360-3 1.695630-3 2.469300-3592533 16 2.400280-3 2.330770-3 2.251710-3 2.160910-3 2.055290-3 1.930710-3592533 16 1.782030-3 1.604310-3 1.399900-3 2.450900-3 2.455830-3 2.435060-3592533 16 2.378170-3 2.280110-3 2.133270-3 1.926860-3 1.647670-3 1.288780-3592533 16 2.543100-3 2.568940-3 2.550460-3 2.474410-3 2.331110-3 2.106380-3592533 16 1.781790-3 1.343610-3 2.656650-3 2.673470-3 2.633140-3 2.518610-3592533 16 2.314590-3 2.000500-3 1.558140-3 2.749450-3 2.752370-3 2.692880-3592533 16 2.550410-3 2.305410-3 1.937640-3 2.834100-3 2.852250-3 2.816520-3592533 16 2.705480-3 2.499580-3 3.001600-3 3.118840-3 3.217400-3 3.275390-3592533 16 3.472370-3 3.866030-3 4.312850-3 4.687830-3 5.673930-3 7.407790-3592533 16 592533 0 5925 0 0 0 0 0 6.40000E+4 1.55901E+2 -1 0 34 106400 1451 0.0 0.0 0 0 0 66400 1451 1.00000E+0 2.00000E+7 0 0 10 20026400 1451 3.00000E+2 0.0 1 0 59 26400 1451 64-Gd- 0 NDS IAEA-JAN04 6400 1451 DIST-Feb2004 6400 1451 ----IRDF-2002 6400 1451 -----INCIDENT NEUTRON DATA 6400 1451 ------ENDF-6 FORMAT 6400 1451 *****************************************************************6400 1451 *****************EXTRACT FOR SPECIAL PURPOSE FILE*****************6400 1451 DOSIMETRY Assembled at NDS from the ENDF/B-VI evaluations for 6400 1451 64-Gd-152,154,155,156,157,158 and 160. The files 6400 1451 were processed through the 2002 pre-processing codes 6400 1451 LINEAR,RECENT,SIGMA1 and FIXUP to produce pointwise 6400 1451 files prior to input to MIXER. 6400 1451 6400 1451 MIXER INPUT FILE 6400 1451 ***************** Program MIXER (VERSION 2002-1) ****************6400 1451 GADOLINIUM,MET. DENSITY 7.901 G/CC. CONSTITUENTS: 160-21.86 6400 1451 158-24.84, 152-0.2, 154-2.18, 155-14.8,156-20.47, 157-15.65 6400 1451 ---------------------------------------- 6400 1451 Composition 6400 1451 ---------------------------------------- 6400 1451 Isotope MF MT Atom-Fract Grams/cc 6400 1451 ---------------------------------------- 6400 1451 64152 3 1 .001999808 .015264700 6400 1451 64154 3 1 .021799790 .168591500 6400 1451 64155 3 1 .147999470 1.15202110 6400 1451 64156 3 1 .204700250 1.60365810 6400 1451 64157 3 1 .156499785 1.23392290 6400 1451 64158 3 1 .248400703 1.97099140 6400 1451 64160 3 1 .218600194 1.75653450 6400 1451 ---------------------------------------- 6400 1451 64000 3 1 1.00000000 7.90098420 6400 1451 ---------------------------------------- 6400 1451 ***************** Program MIXER (VERSION 2002-1) ****************6400 1451 GADOLINIUM,MET. DENSITY 7.901 G/CC. CONSTITUENTS: 160-21.86 6400 1451 158-24.84, 152-0.2, 154-2.18, 155-14.8,156-20.47, 157-15.65 6400 1451 ---------------------------------------- 6400 1451 Composition 6400 1451 ---------------------------------------- 6400 1451 Isotope MF MT Atom-Fract Grams/cc 6400 1451 ---------------------------------------- 6400 1451 64152 3 102 .001999808 .015264700 6400 1451 64154 3 102 .021799790 .168591500 6400 1451 64155 3 102 .147999470 1.15202110 6400 1451 64156 3 102 .204700250 1.60365810 6400 1451 64157 3 102 .156499785 1.23392290 6400 1451 64158 3 102 .248400703 1.97099140 6400 1451 64160 3 102 .218600194 1.75653450 6400 1451 ---------------------------------------- 6400 1451 64000 3 102 1.00000000 7.90098420 6400 1451 ---------------------------------------- 6400 1451 *****************************************************************6400 1451 6400 1451 6400 1451 6400 1451 6400 1451 *****************************************************************6400 1451 ***************** Program GROUPIE (VERSION 2002-1) **************6400 1451 Unshielded Group Averages Using 640 Groups 6400 1451 Weighting Spectrum: Flat (Constant) Spectrum 6400 1451 1 451 65 06400 1451 3 1 217 06400 1451 6400 1 0 6400 0 0 6.40000E+4 1.55901E+2 0 0 0 06400 3 1 0.0 0.0 0 0 1 6416400 3 1 641 1 6400 3 1 .000100000 589975.698 .000105000 576263.743 .000110000 563237.9996400 3 1 .000115000 551236.393 .000120000 537207.265 .000127500 521643.7756400 3 1 .000135000 507426.347 .000142500 494362.490 .000150000 480225.4426400 3 1 .000160000 465500.570 .000170000 452104.974 .000180000 439810.8086400 3 1 .000190000 428373.756 .000200000 417933.482 .000210000 408112.2456400 3 1 .000220000 398991.461 .000230000 390518.607 .000240000 380574.7966400 3 1 .000255000 369650.434 .000270000 361222.513 .000280000 351842.7476400 3 1 .000300000 340406.989 .000320000 330055.677 .000340000 320578.4706400 3 1 .000360000 311852.081 .000380000 303888.913 .000400000 295589.4916400 3 1 .000425000 287099.710 .000450000 279345.803 .000475000 272240.2776400 3 1 .000500000 265560.337 .000525000 259458.199 .000550000 253695.5916400 3 1 .000575000 248338.457 .000600000 242850.160 .000630000 237206.3286400 3 1 .000660000 232045.336 .000690000 227095.998 .000720000 221833.3046400 3 1 .000760000 216193.334 .000800000 210960.030 .000840000 206167.7206400 3 1 .000880000 201614.870 .000920000 197444.084 .000960000 193451.9046400 3 1 .001000000 189318.522 .001050000 185006.271 .001100000 180966.4126400 3 1 .001150000 177247.659 .001200000 172864.165 .001275000 168052.1596400 3 1 .001350000 163658.313 .001425000 159561.986 .001500000 155216.5116400 3 1 .001600000 150679.577 .001700000 146543.973 .001800000 142724.4306400 3 1 .001900000 139242.265 .002000000 136018.231 .002100000 132996.6396400 3 1 .002200000 130235.002 .002300000 127597.042 .002400000 124594.0286400 3 1 .002550000 121263.966 .002700000 118663.917 .002800000 115847.2726400 3 1 .003000000 112388.884 .003200000 109270.806 .003400000 106419.7946400 3 1 .003600000 103799.615 .003800000 101394.497 .004000000 98926.89826400 3 1 .004250000 96446.8108 .004500000 94116.3537 .004750000 92031.69016400 3 1 .005000000 90052.7594 .005250000 88259.9430 .005500000 86584.14836400 3 1 .005750000 84985.7159 .006000000 83423.3912 .006300000 81754.47316400 3 1 .006600000 80285.8467 .006900000 78856.9970 .007200000 77345.67306400 3 1 .007600000 75741.3503 .008000000 74259.2536 .008400000 72905.22576400 3 1 .008800000 71606.4403 .009200000 70452.3569 .009600000 69310.09086400 3 1 .010000000 68168.2878 .010500000 66961.4528 .011000000 65835.29176400 3 1 .011500000 64812.6881 .012000000 63580.7132 .012750000 62262.54436400 3 1 .013500000 61017.9283 .014250000 59899.8789 .015000000 58665.31746400 3 1 .016000001 57381.5012 .017000001 56196.4155 .017999999 55083.30886400 3 1 .018999999 54053.0189 .020000000 53067.6345 .021000000 52144.11396400 3 1 .022000000 51273.0324 .023000000 50401.9583 .024000000 49329.60956400 3 1 .025500000 48142.7567 .027000001 47179.1975 .028000001 46024.12316400 3 1 .029999999 44500.3126 .032000002 43015.3586 .034000002 41552.58366400 3 1 .035999998 40100.7539 .037999999 38649.2567 .039999999 37022.65226400 3 1 .042500000 35228.1518 .045000002 33470.3801 .047499999 31736.19736400 3 1 .050000001 30059.6715 .052499998 28409.0926 .055000000 26820.79836400 3 1 .057500001 25286.7754 .059999999 23676.4725 .063000001 22016.52586400 3 1 .066000000 20454.6852 .068999998 18989.8397 .071999997 17413.42796400 3 1 .075999998 15771.9502 .079999998 14291.7495 .083999999 12962.05766400 3 1 .088000000 11768.9216 .092000000 10697.5306 .096000001 9743.521686400 3 1 .100000001 8791.29030 .104999997 7861.73990 .109999999 7051.680946400 3 1 .115000002 6344.97492 .119999997 5586.76237 .127499998 4821.438736400 3 1 .135000005 4187.76665 .142499998 3659.95592 .150000006 3152.561666400 3 1 .159999996 2680.10654 .170000002 2298.61194 .180000007 1987.812616400 3 1 .189999998 1731.80601 .200000003 1518.87476 .209999993 1340.324736400 3 1 .219999999 1189.70658 .230000004 1061.56016 .239999995 927.6548976400 3 1 .254999995 795.424418 .270000011 704.201722 .280000001 614.3485966400 3 1 .300000012 516.932409 .319999993 440.064304 .340000004 378.5042506400 3 1 .360000014 328.611225 .379999995 287.669565 .400000006 249.9390966400 3 1 .425000012 215.792288 .449999988 188.059115 .474999994 165.2783926400 3 1 .500000000 146.367689 .524999976 130.518725 .550000012 117.1220506400 3 1 .574999988 105.735374 .600000024 95.0807037 .629999995 85.24093946400 3 1 .660000026 76.9287302 .689999998 69.8538784 .720000029 62.88688686400 3 1 .759999990 56.2288703 .800000012 50.7023383 .839999974 46.08568046400 3 1 .879999995 42.2131854 .920000017 38.9203473 .959999979 36.12422896400 3 1 1.00000000 33.4721011 1.04999995 31.0180402 1.10000002 29.00273926400 3 1 1.14999998 27.3535870 1.20000005 25.7500083 1.27499998 24.36919116400 3 1 1.35000002 23.5459552 1.42499995 23.2845714 1.50000000 23.84482546400 3 1 1.60000002 26.1504101 1.70000005 32.4331292 1.79999995 53.78486416400 3 1 1.89999998 155.165344 2.00000000 180.836068 2.09999990 77.07810556400 3 1 2.20000005 72.7161857 2.29999995 122.584743 2.40000010 562.9680816400 3 1 2.54999995 807.932885 2.70000005 278.975585 2.79999995 184.9995716400 3 1 3.00000000 35.5484623 3.20000005 19.5117011 3.40000010 18.28310716400 3 1 3.59999990 17.2413804 3.79999995 10.3079106 4.00000000 8.794554646400 3 1 4.25000000 8.05001316 4.50000000 7.70017817 4.75000000 7.633698076400 3 1 5.00000000 7.89102455 5.25000000 8.71626738 5.50000000 11.00782826400 3 1 5.75000000 19.6004705 6.00000000 183.378816 6.30000019 183.7132516400 3 1 6.59999990 19.3480461 6.90000010 11.8316475 7.19999981 21.19399026400 3 1 7.59999990 114.245385 8.00000000 13.0299188 8.39999962 7.291497776400 3 1 8.80000019 6.25730918 9.19999981 6.06853531 9.60000038 12.22528286400 3 1 10.0000000 11.9591698 10.5000000 6.13540176 11.0000000 15.84106066400 3 1 11.5000000 50.3332202 12.0000000 26.2716604 12.7500000 6.025856556400 3 1 13.5000000 8.60534329 14.2500000 66.2241929 15.0000000 9.265762166400 3 1 16.0000000 290.499829 17.0000000 32.9229550 18.0000000 9.258952006400 3 1 19.0000000 72.7887372 20.0000000 346.790026 21.0000000 197.8306356400 3 1 22.0000000 309.535101 23.0000000 65.9976647 24.0000000 24.39204786400 3 1 25.5000000 11.1178381 27.0000000 14.6794657 28.0000000 47.19567426400 3 1 30.0000000 63.4094372 32.0000000 197.565900 34.0000000 38.79411056400 3 1 36.0000000 40.9938021 38.0000000 15.7428086 40.0000000 8.534046916400 3 1 42.5000000 71.8714886 45.0000000 33.1826967 47.5000000 85.92709666400 3 1 50.0000000 69.9378819 52.5000000 34.2277399 55.0000000 12.52665456400 3 1 57.5000000 97.0438375 60.0000000 19.1908190 63.0000000 22.11990306400 3 1 66.0000000 23.3550317 69.0000000 16.1765505 72.0000000 6.512260046400 3 1 76.0000000 35.3924840 80.0000000 140.727247 84.0000000 24.83407616400 3 1 88.0000000 7.37024937 92.0000000 17.4139551 96.0000000 36.78957746400 3 1 100.000000 52.2961348 105.000000 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7.31586191 475000.000 7.324898266400 3 1 500000.000 7.33111202 525000.000 7.31897430 550000.000 7.303412236400 3 1 575000.000 7.27907017 600000.000 7.24842864 630000.000 7.216354506400 3 1 660000.000 7.20222258 690000.000 7.24480512 720000.000 7.298339186400 3 1 760000.000 7.33223285 800000.000 7.33781676 840000.000 7.324470486400 3 1 880000.000 7.31093482 920000.000 7.30108941 960000.000 7.323477616400 3 1 1000000.00 7.32052865 1100000.00 7.28452665 1200000.00 7.246504336400 3 1 1300000.00 7.18163966 1400000.00 7.08717995 1500000.00 7.005851626400 3 1 1600000.00 6.91033168 1700000.00 6.84917941 1800000.00 6.773429026400 3 1 1900000.00 6.69723840 2000000.00 6.62467069 2100000.00 6.566497116400 3 1 2200000.00 6.50997524 2300000.00 6.45879556 2400000.00 6.414734676400 3 1 2500000.00 6.37078706 2600000.00 6.32704875 2700000.00 6.283322386400 3 1 2800000.00 6.23959601 2900000.00 6.19314033 3000000.00 6.142457286400 3 1 3100000.00 6.09066487 3200000.00 6.03878275 3300000.00 5.986103296400 3 1 3400000.00 5.93170828 3500000.00 5.87555926 3600000.00 5.819069596400 3 1 3700000.00 5.76268152 3800000.00 5.70629345 3900000.00 5.649905386400 3 1 4000000.00 5.59371016 4100000.00 5.53770780 4200000.00 5.481792206400 3 1 4300000.00 5.43159956 4400000.00 5.38486083 4500000.00 5.341090516400 3 1 4600000.00 5.30023290 4700000.00 5.25937631 4800000.00 5.218520406400 3 1 4900000.00 5.18113675 5000000.00 5.14730080 5100000.00 5.113730106400 3 1 5200000.00 5.08444163 5300000.00 5.06068140 5400000.00 5.036943056400 3 1 5500000.00 5.01320469 5600000.00 4.99003581 5700000.00 4.971389656400 3 1 5800000.00 4.95371758 5900000.00 4.93604551 6000000.00 4.918532966400 3 1 6100000.00 4.90117992 6200000.00 4.88925335 6300000.00 4.878335386400 3 1 6400000.00 4.86670072 6500000.00 4.85506606 6600000.00 4.843433476400 3 1 6700000.00 4.83429814 6800000.00 4.82738682 6900000.00 4.820475506400 3 1 7000000.00 4.81365421 7100000.00 4.80698616 7200000.00 4.803311816400 3 1 7300000.00 4.80127939 7400000.00 4.79924696 7500000.00 4.800718896400 3 1 7600000.00 4.80616541 7700000.00 4.81161581 7800000.00 4.817066216400 3 1 7900000.00 4.82251661 8000000.00 4.83000689 8100000.00 4.840432106400 3 1 8200000.00 4.85095573 8300000.00 4.86147935 8400000.00 4.872002986400 3 1 8500000.00 4.88346054 8600000.00 4.89585205 8700000.00 4.908243556400 3 1 8800000.00 4.92063505 8900000.00 4.93302656 9000000.00 4.945436686400 3 1 9100000.00 4.95800229 9200000.00 4.97062120 9300000.00 4.983244306400 3 1 9400000.00 4.99586785 9500000.00 5.00849140 9600000.00 5.021114956400 3 1 9700000.00 5.03373850 9800000.00 5.04636205 9900000.00 5.058985606400 3 1 10000000.0 5.07160915 10100000.0 5.08423271 10200000.0 5.096717116400 3 1 10300000.0 5.10846077 10400000.0 5.12009543 10500000.0 5.131730106400 3 1 10600000.0 5.14336476 10700000.0 5.15499943 10800000.0 5.166634096400 3 1 10900000.0 5.17825492 11000000.0 5.18969603 11100000.0 5.201081876400 3 1 11200000.0 5.21246771 11300000.0 5.22385355 11400000.0 5.231251816400 3 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5.15530023 15600000.0 5.148720686400 3 1 15700000.0 5.14212068 15800000.0 5.13551972 15900000.0 5.128918776400 3 1 16000000.0 5.12306018 16100000.0 5.11794396 16200000.0 5.112827746400 3 1 16300000.0 5.10771151 16400000.0 5.10259529 16500000.0 5.097479076400 3 1 16600000.0 5.09236285 16700000.0 5.08724663 16800000.0 5.082130406400 3 1 16900000.0 5.07701418 17000000.0 5.07189796 17100000.0 5.066781746400 3 1 17200000.0 5.06166552 17300000.0 5.05654929 17400000.0 5.051433076400 3 1 17500000.0 5.04631685 17600000.0 5.04120063 17700000.0 5.036084416400 3 1 17800000.0 5.03096818 17900000.0 5.02585196 18000000.0 5.020675146400 3 1 18100000.0 5.01543771 18200000.0 5.01020029 18300000.0 5.004962876400 3 1 18400000.0 4.99972544 18500000.0 4.99481980 18600000.0 4.990245936400 3 1 18700000.0 4.98580967 18800000.0 4.98405926 18900000.0 4.983345976400 3 1 19000000.0 4.98287553 19100000.0 4.98264794 19200000.0 4.982420356400 3 1 19300000.0 4.98219276 19400000.0 4.98196517 19500000.0 4.981737576400 3 1 19600000.0 4.98150998 19700000.0 4.98128239 19800000.0 4.981054806400 3 1 19900000.0 4.98082721 20000000.0 0.0 6400 3 1 6400 3 0 6400 0 0 0 0 0 6.91690E+4 1.67483E+2 0 0 34 106925 1451 0.0 0.0 0 0 0 66925 1451 1.00000E+0 2.00000E+7 0 0 10 20026925 1451 3.00000E+2 0.0 1 0 88 36925 1451 69-Tm-169 SRI EVAL-JAN96 N.ODANO (SHIP RES. INST.) 6925 1451 DIST-Feb2004 6925 1451 ----IRDF-2002 MATERIAL 6925 6925 1451 -----INCIDENT NEUTRON DATA 6925 1451 ------ENDF-6 FORMAT 6925 1451 **************************************************************** 6925 1451 69-TM-169 SRI EVAL-JAN96 N.ODANO (SHIP RES. INST.) 6925 1451 DIST-JUL98 6925 1451 ----JENDL/D-99 MATERIAL 6925 6925 1451 **************************************************************** 6925 1451 HISTORY 6925 1451 96-01 EVALUATION FOR JENDL DOSIMETRY FILE VERSION 2 WAS MADE BY 6925 1451 N.ODANO (SHIP RES. INST.). 6925 1451 97-09 COMPILED TO JENDL DOSIMETRY FILE 99. 6925 1451 6925 1451 ==== POINT-WISE DATA FILE ==== 6925 1451 6925 1451 6925 1451 TM-169 (N,2N) TM-168 (HALF-LIFE = 93.1 D) 6925 1451 6925 1451 MF=1 GENERAL INFORMATION 6925 1451 MT=451 DESCRIPTIVE DATA AND DICTIONARY 6925 1451 6925 1451 MF=2 RESONANCE PARAMETERS 6925 1451 MT=151 PARAMETERS 6925 1451 ONLY SPIN AND SCATTERING RADIUS ARE GIVEN. 6925 1451 6925 1451 MF=3 NEUTRON CROSS SECTIONS 6925 1451 MT=16 (N,2N) CROSS SECTION 6925 1451 EXPERIMENTAL DATA/1-13/ IN NESTOR-2/14/ WERE TAKEN FOR 6925 1451 THE EVALUATION USING GMA CODE/15/. 6925 1451 6925 1451 MF=33 COVARIANCES OF NEUTRON CROSS SECTIONS 6925 1451 MT=16 GENERATED USING THE GMA CODE. 6925 1451 6925 1451 REFERENCES 6925 1451 1) W.DILG+ : NUCL. PHYS. A, 118, 9 (1968). 6925 1451 2) D.S.MATHER+ : AWRE-O-72/,72 (1972). 6925 1451 3) D.R.NETHERWAY : NUCL. PHYS. A, 190, 635 (1972). 6925 1451 4) R.VOS+ : BULLETIN OF THE AMERICAN PHYSICAL SOCIETY, 18, 6925 1451 775(EB6) (1973). 6931 1451 6925 1451 5) S.M.QAIM : NUCL. PHYS. A, 224, 319 (1974). 6925 1451 6) B.P.BAYHURST+ : PHYS. REV. C, 12, 451 (1975). 6925 1451 7) L.R.VEESER+ : PHYS. REV. C, 16, 1792 (1977). 6925 1451 8) J.FREHAUT+ : PROC. SYMP. ON NEUTRON CROSS SECTIONS FROM 10-50 6925 1451 MEV, UPTON, USA, 12-14 MAY, 1980, P.399 (1980). 6925 1451 9) J.LAUREC+ : CEA-R-5109 (1981). 6925 1451 10) L.R.GREENWOOD : ASTM-STP-956, 743 (1987). 6925 1451 11) LU HANLIN+ : INDC(CRP)-16 (1989). 6925 1451 12) WANG XIUYUAN+ : A COLLECTION OF ACADEMIC EXCHANGE REPORT IN 6925 1451 THE FIELDS OD RADIOCHEMISTRY, CHEMICAL INDUSTRY AND ISOTOPE, 6925 1451 NUCLEAR INSTITUTE OF SICHUAN (1989). 6925 1451 13) C.KONNO+ : JAERI 1329 (1993). 6925 1451 14) T.NAKAGAWA : THE JAERI NUCLEAR DATA CENTER, UNPUBLISHED. 6925 1451 15) W.P.POENITZ : PROC. CONF. NUCLEAR DATA EVALUATION METHODS 6925 1451 AND PROCEDURES, BROOKHAVEN NATIONAL LAB. 1980, BNL-NCS- 6925 1451 51363, P.249 (1981). 6925 1451 **************************************************************** 6925 1451 The Q values and threshold energies were updated prior to pro- 6925 1451 cessing through the codes to comply with the values obtained 6925 1451 using the NNDC calculation program which is based on the 1995 6925 1451 Update to the Atomic mass Evaluation. 6925 1451 **************************************************************** 6925 1451 ***************** Program LINEAR (VERSION 2002-1) ***************6925 1451 For All Data Greater than 1.0000E-10 barns in Absolute Value 6925 1451 Data Linearized to Within an Accuracy of .100000000 per-cent 6925 1451 ***************** Program SIGMA1 (VERSION 2002-1) ***************6925 1451 Data Doppler Broadened to 300.000000 Kelvin 6925 1451 for All Data Greater than 1.0000E-10 barns in Absolute Value 6925 1451 Data Linearized to Within an Accuracy pf .100000000 per-cent 6925 1451 ***************** Program FIXUP (Version 2002-1) ****************6925 1451 Corrected ZA/AWR in All Sections-----------------------------Yes 6925 1451 Corrected Thresholds-----------------------------------------Yes 6925 1451 Extended Cross Sections to 20 MeV----------------------------No 6925 1451 Allow Cross Section Deletion---------------------------------No 6925 1451 Allow Cross Section Reconstruction---------------------------No 6925 1451 Make All Cross Sections Non-Negative-------------------------Yes 6925 1451 Delete Energies Not in Ascending Order-----------------------Yes 6925 1451 Deleted Duplicate Points-------------------------------------Yes 6925 1451 Check for Ascending MAT/MF/MT Order--------------------------Yes 6925 1451 Check for Legal MF/MT Numbers--------------------------------Yes 6925 1451 Allow Creation of Missing Sections---------------------------No 6925 1451 Allow Insertion of Energy Points-----------------------------No 6925 1451 Create Uniform Energy Grid-----------------------------------No 6925 1451 Delete Section if Cross Section =0 at All Energies-----------Yes 6925 1451 ***************** Program GROUPIE (VERSION 2002-1) **************6925 1451 Unshielded Group Averages Using 640 Groups 6925 1451 Weighting Spectrum: Flat (Constant) Spectrum 6925 1451 1 451 95 16925 1451 3 16 44 16925 1451 33 16 23 16925 1451 6925 1 0 6925 0 0 6.91690E+4 1.67483E+2 0 0 0 06925 3 16 -8.03340E+6-8.03340E+6 0 0 1 1216925 3 16 121 1 6925 3 16 8000000.00 .000162775 8100000.00 .006767183 8200000.00 .0167043106925 3 16 8300000.00 .026641437 8400000.00 .070615833 8500000.00 .1486275006925 3 16 8600000.00 .226639167 8700000.00 .304650833 8800000.00 .3826625006925 3 16 8900000.00 .460674167 9000000.00 .535816000 9100000.00 .6080880006925 3 16 9200000.00 .680360000 9300000.00 .752632000 9400000.00 .8249040006925 3 16 9500000.00 .897176000 9600000.00 .969448000 9700000.00 1.041720006925 3 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13800000.0 2.00127333 13900000.0 2.000246676925 3 16 14000000.0 1.99922000 14100000.0 1.99819333 14200000.0 1.997166676925 3 16 14300000.0 1.99614000 14400000.0 1.99511333 14500000.0 1.994686676925 3 16 14600000.0 1.99486000 14700000.0 1.99503333 14800000.0 1.995206676925 3 16 14900000.0 1.99538000 15000000.0 1.99555333 15100000.0 1.995726676925 3 16 15200000.0 1.99590000 15300000.0 1.99607333 15400000.0 1.996246676925 3 16 15500000.0 1.99642000 15600000.0 1.99659333 15700000.0 1.996766676925 3 16 15800000.0 1.99694000 15900000.0 1.99711333 16000000.0 1.994380006925 3 16 16100000.0 1.98874000 16200000.0 1.98310000 16300000.0 1.977460006925 3 16 16400000.0 1.97182000 16500000.0 1.96618000 16600000.0 1.960540006925 3 16 16700000.0 1.95490000 16800000.0 1.94926000 16900000.0 1.943620006925 3 16 17000000.0 1.91730500 17100000.0 1.87031500 17200000.0 1.823325006925 3 16 17300000.0 1.77633500 17400000.0 1.72934500 17500000.0 1.682355006925 3 16 17600000.0 1.63536500 17700000.0 1.58837500 17800000.0 1.541385006925 3 16 17900000.0 1.49439500 18000000.0 1.45198500 18100000.0 1.414155006925 3 16 18200000.0 1.37632500 18300000.0 1.33849500 18400000.0 1.300665006925 3 16 18500000.0 1.26283500 18600000.0 1.22500500 18700000.0 1.187175006925 3 16 18800000.0 1.14934500 18900000.0 1.11151500 19000000.0 1.081435006925 3 16 19100000.0 1.05910500 19200000.0 1.03677500 19300000.0 1.014445006925 3 16 19400000.0 .992115000 19500000.0 .969785000 19600000.0 .9474550006925 3 16 19700000.0 .925125000 19800000.0 .902795000 19900000.0 .8804650006925 3 16 20000000.0 0.0 6925 3 16 6925 3 0 6925 0 0 6.91690E+4 1.67483E+2 0 0 0 1692533 16 0.000000+0 0.000000+0 0 16 0 1692533 16 0.000000+0 0.000000+0 1 5 120 15692533 16 1.000000-5 8.000000+6 8.240180+6 8.700000+6 9.500000+6 1.050000+7692533 16 1.150000+7 1.250000+7 1.375000+7 1.525000+7 1.650000+7 1.750000+7692533 16 1.850000+7 1.950000+7 2.000000+7 0.000000+0 0.000000+0 0.000000+0692533 16 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0692533 16 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0 1.544490-3692533 16 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0692533 16 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0692533 16 1.544490-3 2.161260-4 8.021960-5 8.018380-5 8.965270-5 1.308120-4692533 16 8.611730-5 7.978640-5 1.636990-4 1.667380-4 9.654600-5 2.752760-4692533 16 1.762320-3 1.773580-4 1.789170-4 1.261310-4 1.126210-4 7.632130-5692533 16 6.414960-5 1.179690-4 1.202010-4 7.090170-5 1.789260-4 2.253400-3692533 16 3.400620-4 1.941300-4 1.014070-4 7.258540-5 4.974090-5 6.811760-5692533 16 7.143290-5 4.130170-5 6.744160-5 2.055720-3 1.967720-4 1.028640-4692533 16 7.314980-5 5.047840-5 6.867560-5 7.201810-5 4.176900-5 6.741150-5692533 16 7.739520-4 1.059610-4 7.502940-5 7.712840-5 9.702920-5 1.017520-4692533 16 5.552930-5 7.629130-5 2.742330-4 8.094930-5 8.206160-5 1.132040-4692533 16 1.180210-4 6.526070-5 1.110700-4 1.449620-4 5.756050-5 7.774620-5692533 16 8.052350-5 5.484240-5 7.438890-5 4.765490-4 1.022420-4 1.031120-4692533 16 1.117260-4 7.429010-5 3.972050-4 1.424560-4 1.295420-4 1.435500-4692533 16 4.368100-4 1.091050-4 1.436310-4 2.619390-3 1.132960-4 1.091640-3692533 16 692533 0 6925 0 0 0 0 0 7.31810E+4 1.79393E+2 0 0 34 107328 1451 0.0 0.0 0 0 0 67328 1451 1.00000E+0 2.00000E+7 0 0 10 20027328 1451 3.00000E+2 0.0 1 0 117 37328 1451 73-Ta-181 NAIG EVAL-MAR87 N.YAMAMURO 7328 1451 DIST-Feb2004 7328 1451 ----IRDF-2002 MATERIAL 7328 REVISION 2 7328 1451 -----INCIDENT NEUTRON DATA 7328 1451 ------ENDF-6 FORMAT 7328 1451 **************************************************************** 7328 1451 73-TA-181 NAIG EVAL-MAR87 N.YAMAMURO 7328 1451 DIST-SEP89 REV2-FEB94 7328 1451 ----JENDL-3.2 MATERIAL 7328 REVISION 2 7328 1451 7328 1451 TA-181 CAPTURE TA-181(HALF-LIFE = 114.43D) 7328 1451 **************************************************************** 7328 1451 HISTORY 7328 1451 76-03 THE EVALUATION FOR JENDL-1 /1/ WAS MADE BY H.YAMAKOSHI 7328 1451 (SHIP RESEARCH INSTITUTE) AND JENDL-1 COMPILATION GROUP. 7328 1451 83-03 JENDL-1 DATA WERE ADOPTED FOR JENDL-2 AND EXTENDED TO 20 7328 1451 MEV. MF=5 WAS REVISED, AND UNRESOLVED RESONANCE PARAMETERS 7328 1451 WERE ADDED BY Y.KIKUCHI (JAERI) /2/. 7328 1451 83-11 COMMENT DATA WERE ADDED. 7328 1451 87-03 THE EVALUATION FOR JENDL-3 WAS MADE BY N.YAMAMURO (NAIG). 7328 1451 RESONANCE PARAMETERS WERE ADDED BY NEW EXPERIMENTAL DATA. 7328 1451 NEUTRON CROSS SECTIONS, EXCEPT TOTAL AND ELASTIC SCATTERING 7328 1451 CROSS SECTIONS, AND ENERGY DISTRIBUTIONS OF SECONDARY 7328 1451 NEUTRONS AND PHOTONS WERE CALCULATED WITH GNASH /3/ AND 7328 1451 CASTHY /4/ CODES. 7328 1451 94-02 JENDL-3.2. 7328 1451 COMPILED BY T.NAKAGAWA (NDC/JAERI) 7328 1451 7328 1451 ***** MODIFIED PARTS FOR JENDL-3.2 ******************** 7328 1451 (12,102) 7328 1451 DATA WERE DETERMINED FROM ENERGY BALANCE. 7328 1451 *********************************************************** 7328 1451 7328 1451 7328 1451 MF=1 GENERAL INFORMATION 7328 1451 MT=451 DESCRIPTIVE DATA AND DICTIONARY 7328 1451 7328 1451 MF=2 RESONANCE PARAMETERS 7328 1451 MT=151 RESOLVED AND UNRESOLVED RESONANCE PARAMETERS 7328 1451 RESOLVED PARAMETERS FOR MLBW FORMULA 7328 1451 7328 1451 THE ENERGY REGION FROM 1.0E-5EV TO 1.0 KEV. PARAMETERS WERE 7328 1451 TAKEN FROM REFS./5,6,7/ FOR POSITIVE RESONANCES, AND FROM 7328 1451 ENDF/B-IV FOR A NEGATIVE RESONANCE. THE RADIATIVE WIDTH OF 7328 1451 0.059EV WAS ASSUMED FOR THE RESONANCE WHOSE RADIATIVE WIDTH 7328 1451 WAS UNKNOWN. 7328 1451 7328 1451 UNRESOLVED PARAMETERS 7328 1451 7328 1451 IN THE ENERGY RANGE FROM 1 TO 100KEV, PARAMETERS WERE 7328 1451 DETERMINED TO REPRODUCE THE MEASURED CAPTURE CROSS SECTIONS 7328 1451 /6,8/. THE PARAMETERS ARE AS FOLLOWS, 7328 1451 7328 1451 R= 7.8 FM , DOBS= 4.2 EV , RADIATIVE WIDTH= 0.065 EV, 7328 1451 SO= 1.7E-04 S1= 2.0E-05 S2= 2.3E-04 NL= 3 7328 1451 7328 1451 CALCULATED 2200-M/SEC CROSS SECTIONS AND RESONANCE INTEGRALS7328 1451 2200-M/SEC RES. INTEG 7328 1451 ELASTIC 5.65 B - 7328 1451 CAPTURE 20.67 B 660.43 B 7328 1451 TOTAL 26.32 B - 7328 1451 7328 1451 MF=3 NEUTRON CROSS SECTIONS 7328 1451 MT=1 TOTAL 7328 1451 EVALUATED FROM EXPERIMENTAL DATA. 7328 1451 MT=2 ELASTIC SCATTERING 7328 1451 (TOTAL CROSS SECTION) - (REACTION CROSS SECTION) 7328 1451 MT=102 RADIATIVE CAPTURE CROSS SECTION 7328 1451 CALCULATED WITH CASTHY/4/. 7328 1451 7328 1451 REFERENCES 7328 1451 1) IGARASI S. ET AL.: JAERI 1261 (1979) 7328 1451 2) (ED.) NAKAGAWA T.: JAERI-M 84-103 (1984) 7328 1451 3) YOUNG,P.G. AND ARTHUR,E.D.: "GNASH, A PREEQUILIBRIUM 7328 1451 STATISTICAL NUCLEAR-MODEL CODE FOR CALCULATION OF CROSS 7328 1451 SECTIONS AND EMISSION SPECTRA", LA-6974 (1977). 7328 1451 4) IGARASHI,S. AND FUKAHORI,T.: JAERI 1321 (1991). 7328 1451 5) MUGHABGHAB,S.F, AND GARDER,D.I.: BNL325, 3RD ED. (1973). 7328 1451 6) MACKLIN,R,L,: NUCL,SCI,ENG., 86,362 (1984). 7328 1451 7) TSUBONE,I., NAKAJIMA,Y. AND KANDA,Y.: PRIVATE COMMUNICATION 7328 1451 8) YAMAMURO,N., SAITO,K., EMOTO,T., WADA,T., FUJITA,Y. AND 7328 1451 KOBAYASHI,K.: J.NUCL.SCI.TECHNOL., 17, 582 (1980). 7328 1451 9) WILMORE,D. AND HODGSON,P.E.: NUCL, PHYS., 55, 673 (1964). 7328 1451 10) FIRESTONE,R.B.* NUCL, DATA SHEETS 43, 289 (1984). 7328 1451 *************************************************************** 7328 1451 The Q values and threshold energies were updated prior to pro- 7328 1451 cessing through the codes to comply with the values obtained 7328 1451 using the NNDC calculation program which is based on the 1995 7328 1451 Update to the Atomic mass Evaluation. 7328 1451 ***************** Program LINEAR (VERSION 2002-1) ***************7328 1451 For All Data Greater than 1.0000E-10 barns in Absolute Value 7328 1451 Data Linearized to Within an Accuracy of .100000000 per-cent 7328 1451 ***************** Program RECENT (VERSION 2002-1) ***************7328 1451 for All Data Greater than 1.0000E-10 barns in Absolute Value 7328 1451 Data Linearized to within an Accuracy of .100000000 per-cent 7328 1451 ***************** Program SIGMA1 (VERSION 2002-1) ***************7328 1451 Data Doppler Broadened to 300.000000 Kelvin 7328 1451 for All Data Greater than 1.0000E-10 barns in Absolute Value 7328 1451 Data Linearized to Within an Accuracy pf .100000000 per-cent 7328 1451 ***************** Program FIXUP (Version 2002-1) ****************7328 1451 Corrected ZA/AWR in All Sections-----------------------------Yes 7328 1451 Corrected Thresholds-----------------------------------------Yes 7328 1451 Extended Cross Sections to 20 MeV----------------------------No 7328 1451 Allow Cross Section Deletion---------------------------------No 7328 1451 Allow Cross Section Reconstruction---------------------------No 7328 1451 Make All Cross Sections Non-Negative-------------------------Yes 7328 1451 Delete Energies Not in Ascending Order-----------------------Yes 7328 1451 Deleted Duplicate Points-------------------------------------Yes 7328 1451 Check for Ascending MAT/MF/MT Order--------------------------Yes 7328 1451 Check for Legal MF/MT Numbers--------------------------------Yes 7328 1451 Allow Creation of Missing Sections---------------------------No 7328 1451 Allow Insertion of Energy Points-----------------------------No 7328 1451 Create Uniform Energy Grid-----------------------------------No 7328 1451 Delete Section if Cross Section =0 at All Energies-----------Yes 7328 1451 ***************** Program GROUPIE (VERSION 2002-1) **************7328 1451 Unshielded Group Averages Using 640 Groups 7328 1451 Weighting Spectrum: Flat (Constant) Spectrum 7328 1451 1 451 124 17328 1451 3 102 217 17328 1451 33 102 16 17328 1451 7328 1 0 7328 0 0 7.31810E+4 1.79393E+2 0 0 0 07328 3102 6.06296E+6 6.06296E+6 0 0 1 6417328 3102 641 1 7328 3102 .000100000 323.217903 .000105000 315.681488 .000110000 308.5218207328 3102 .000115000 301.924447 .000120000 294.211880 .000127500 285.6549077328 3102 .000135000 277.837071 .000142500 270.652682 .000150000 262.8772027328 3102 .000160000 254.777193 .000170000 247.407203 .000180000 240.6421387328 3102 .000190000 234.347828 .000200000 228.601006 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.0970308707328 3102 1300000.00 .093152354 1400000.00 .089688166 1500000.00 .0843624307328 3102 1600000.00 .077620419 1700000.00 .071773871 1800000.00 .0666320467328 3102 1900000.00 .062144851 2000000.00 .057693351 2100000.00 .0533494997328 3102 2200000.00 .049514519 2300000.00 .046096349 2400000.00 .0430441537328 3102 2500000.00 .039965962 2600000.00 .036888310 2700000.00 .0341533867328 3102 2800000.00 .031702769 2900000.00 .029504405 3000000.00 .0270695227328 3102 3100000.00 .024470771 3200000.00 .022186918 3300000.00 .0201851367328 3102 3400000.00 .018408410 3500000.00 .016835710 3600000.00 .0154340117328 3102 3700000.00 .014182623 3800000.00 .013064813 3900000.00 .0120582257328 3102 4000000.00 .010776892 4100000.00 .009322457 4200000.00 .0080923817328 3102 4300000.00 .007050960 4400000.00 .006159965 4500000.00 .0053985397328 3102 4600000.00 .004745187 4700000.00 .004181665 4800000.00 .0036955027328 3102 4900000.00 .003273923 5000000.00 .002816271 5100000.00 .0023504037328 3102 5200000.00 .001969123 5300000.00 .001654538 5400000.00 .0013953337328 3102 5500000.00 .001179810 5600000.00 .001000877 5700000.00 .0008514337328 3102 5800000.00 .000726381 5900000.00 .000621363 6000000.00 .0005319557328 3102 6100000.00 .000455780 6200000.00 .000391423 6300000.00 .0003370467328 3102 6400000.00 .000290866 6500000.00 .000251612 6600000.00 .0002180957328 3102 6700000.00 .000189461 6800000.00 .000164938 6900000.00 .0001438777328 3102 7000000.00 .000126426 7100000.00 .000111945 7200000.00 9.93160E-57328 3102 7300000.00 8.82150E-5 7400000.00 7.85053E-5 7500000.00 6.99745E-57328 3102 7600000.00 6.24562E-5 7700000.00 5.58332E-5 7800000.00 4.99792E-57328 3102 7900000.00 4.48090E-5 8000000.00 3.97481E-5 8100000.00 3.48806E-57328 3102 8200000.00 3.06678E-5 8300000.00 2.70014E-5 8400000.00 2.38058E-57328 3102 8500000.00 2.10222E-5 8600000.00 1.85956E-5 8700000.00 1.64654E-57328 3102 8800000.00 1.46027E-5 8900000.00 1.29644E-5 9000000.00 1.14976E-57328 3102 9100000.00 1.01769E-5 9200000.00 9.02338E-6 9300000.00 8.00673E-67328 3102 9400000.00 7.11624E-6 9500000.00 6.33236E-6 9600000.00 5.64148E-67328 3102 9700000.00 5.03184E-6 9800000.00 4.49326E-6 9900000.00 4.01735E-67328 3102 10000000.0 3.61056E-6 10100000.0 3.26201E-6 10200000.0 2.94988E-67328 3102 10300000.0 2.67066E-6 10400000.0 2.41987E-6 10500000.0 2.19506E-67328 3102 10600000.0 1.99242E-6 10700000.0 1.81081E-6 10800000.0 1.64692E-67328 3102 10900000.0 1.49895E-6 11000000.0 1.36998E-6 11100000.0 1.25708E-67328 3102 11200000.0 1.15454E-6 11300000.0 1.06090E-6 11400000.0 9.75771E-77328 3102 11500000.0 8.97808E-7 11600000.0 8.27136E-7 11700000.0 7.62379E-77328 3102 11800000.0 7.02910E-7 11900000.0 6.48963E-7 12000000.0 6.04076E-77328 3102 12100000.0 5.67200E-7 12200000.0 5.33022E-7 12300000.0 5.01074E-77328 3102 12400000.0 4.71252E-7 12500000.0 4.43550E-7 12600000.0 4.17575E-77328 3102 12700000.0 3.93213E-7 12800000.0 3.70602E-7 12900000.0 3.49426E-77328 3102 13000000.0 3.30587E-7 13100000.0 3.13996E-7 13200000.0 2.98327E-77328 3102 13300000.0 2.83547E-7 13400000.0 2.69619E-7 13500000.0 2.56409E-77328 3102 13600000.0 2.43978E-7 13700000.0 2.32360E-7 13800000.0 2.21177E-77328 3102 13900000.0 2.10728E-7 14000000.0 2.01228E-7 14100000.0 1.92650E-77328 3102 14200000.0 1.84535E-7 14300000.0 1.76786E-7 14400000.0 1.69441E-77328 3102 14500000.0 1.62470E-7 14600000.0 1.55761E-7 14700000.0 1.49437E-77328 3102 14800000.0 1.43348E-7 14900000.0 1.37600E-7 15000000.0 1.31418E-77328 3102 15100000.0 1.24932E-7 15200000.0 1.18796E-7 15300000.0 1.12997E-77328 3102 15400000.0 1.07521E-7 15500000.0 1.02329E-7 15600000.0 9.74570E-87328 3102 15700000.0 9.27964E-8 15800000.0 8.84000E-8 15900000.0 8.42572E-87328 3102 16000000.0 8.06496E-8 16100000.0 7.75337E-8 16200000.0 7.45694E-87328 3102 16300000.0 7.17256E-8 16400000.0 6.90170E-8 16500000.0 6.64320E-87328 3102 16600000.0 6.39375E-8 16700000.0 6.15916E-8 16800000.0 5.92985E-87328 3102 16900000.0 5.71001E-8 17000000.0 5.51630E-8 17100000.0 5.34235E-87328 3102 17200000.0 5.17080E-8 17300000.0 5.01101E-8 17400000.0 4.85278E-87328 3102 17500000.0 4.70063E-8 17600000.0 4.55502E-8 17700000.0 4.41732E-87328 3102 17800000.0 4.28313E-8 17900000.0 4.14985E-8 18000000.0 4.03688E-87328 3102 18100000.0 3.94019E-8 18200000.0 3.84351E-8 18300000.0 3.74837E-87328 3102 18400000.0 3.65908E-8 18500000.0 3.57061E-8 18600000.0 3.48673E-87328 3102 18700000.0 3.40452E-8 18800000.0 3.32291E-8 18900000.0 3.24595E-87328 3102 19000000.0 3.16995E-8 19100000.0 3.09575E-8 19200000.0 3.02550E-87328 3102 19300000.0 2.95540E-8 19400000.0 2.88663E-8 19500000.0 2.82189E-87328 3102 19600000.0 2.75742E-8 19700000.0 2.69479E-8 19800000.0 2.63519E-87328 3102 19900000.0 2.57564E-8 20000000.0 0.0 7328 3102 7328 3 0 7328 0 0 7.31810E+4 1.79393E+2 0 0 0 1732833102 0.000000+0 0.000000+0 0 102 0 1732833102 0.000000+0 0.000000+0 0 1 74 37732833102 1.000000-5 9.000000-4 1.000000+0 2.500000-3 5.000000+0 2.500000-3732833102 1.000000+1 1.000000-2 2.000000+1 1.000000-2 3.000000+1 2.250000-2732833102 5.000000+1 2.250000-2 7.000000+1 2.250000-2 1.000000+2 4.000000-2732833102 2.000000+2 4.000000-2 3.000000+2 4.000000-2 5.000000+2 4.000000-2732833102 7.000000+2 4.000000-2 1.000000+3 2.250000-2 2.000000+3 2.250000-2732833102 3.000000+3 1.000000-2 5.000000+3 1.000000-2 7.000000+3 1.000000-2732833102 1.000000+4 1.000000-2 2.000000+4 1.000000-2 3.000000+4 1.000000-2732833102 5.000000+4 1.000000-2 7.000000+4 1.000000-2 1.000000+5 4.000000-2732833102 2.000000+5 4.000000-2 3.000000+5 4.000000-2 5.000000+5 4.000000-2732833102 7.000000+5 4.000000-2 1.000000+6 1.000000-2 2.000000+6 1.000000-2732833102 3.000000+6 1.000000-2 4.000000+6 4.000000-2 6.000000+6 4.000000-2732833102 7.000000+6 4.000000-2 1.000000+7 9.000000-2 1.500000+7 9.000000-2732833102 2.000000+7 0.000000+0 732833102 732833 0 7328 0 0 0 0 0 7.41860E+4 1.84357E+2 0 0 34 107443 1451 0.0 0.0 0 0 0 67443 1451 1.00000E+0 2.00000E+7 0 0 10 20027443 1451 3.00000E+2 0.0 1 0 199 37443 1451 74-W -186 FEI,LANL+ EVAL-SEP01 K.I.Zolotarev,M.B.CHADWICK etal. 7443 1451 DIST-Feb2004 7443 1451 ----IRDF-2002 MATERIAL 7443 7443 1451 -----INCIDENT NEUTRON DATA 7443 1451 ------ENDF-6 FORMAT 7443 1451 ***************************************************************** 7443 1451 74-W -186 FEI EVAL-Sep01 K.I.Zolotarev 7443 1451 DIST-Jan04 7443 1451 ----BROND-3 MATERIAL 7443 Revision 1, Jan. 2004 7443 1451 -----INCIDENT NEUTRON DATA 7443 1451 ------ENDF-6 FORMAT 7443 1451 ------Russian Reactor Dosimetry File RRDF-2002 7443 1451 ***************************************************************** 7443 1451 ********** Start of (N,G) bibliographical component ********** 7443 1451 ***************************************************************** 7443 1451 ***************************************************************** 7443 1451 Authors of evaluation: K.I.Zolotarev and V.G.Pronyaev 7443 1451 ***************************************************************** 7443 1451 7443 1451 ----- MF=2 MT=151 ----- 7443 1451 7443 1451 RESOLVED RESONANCE PARAMETERS 7443 1451 Resolved MLBW resonance parameters up to 8.5 keV were based 7443 1451 on data from ref.[1-2]. Spins of resonances, which were not iden- 7443 1451 tified, have been assigned by taking into account 2*J+1 law and 7443 1451 approximate linear growth of the number of levels with given spin 7443 1451 and parity from energy. Average gamma-widths 0.05 eV (L=0 and L=1)7443 1451 were assigned to those resonances, widths for which were absent. 7443 1451 Fictitious resonances were added in the energy region 4 - 8.5 keV 7443 1451 to obtain the average capture cross section values observed in 7443 1451 the experiments [3],[4],[5],[6]. 7443 1451 7443 1451 UNRESOLVED RESONANCE PARAMETERS 7443 1451 Unresolved resonance parameters in the region 8.5 - 100 keV 7443 1451 were evaluated with using 'EVPAR' code (Hauser-Feshbach-Moldauer 7443 1451 statistical model)[7]. They were calculated from optimized fit to 7443 1451 the measured capture [4,5,6,8,9] and total [10,11] cross sections 7443 1451 data. Average resolved resonance parameters have been used as 7443 1451 zero approximation. Contribution of F-wave in the energy region 7443 1451 30 - 100 keV given as smoothed background cross section in File 3.7443 1451 7443 1451 ----- MF=3 MT=102 ----- 7443 1451 Capture cross sections from 1.000E-5 eV to 100 keV are recon- 7443 1451 structed from evaluated MLBW resolved and unresolved resonance 7443 1451 parameters. Contribution of F-wave in the neutron energy interval 7443 1451 30 - 100 keV is added as described above. 7443 1451 Data base for the evaluation W-186(n,g)W-187 excitation func- 7443 1451 tion in the neutron energy range 100 keV - 20 MeV was formed from 7443 1451 microscopic experimental data [4-6],[8-9],[12-25] and data recei- 7443 1451 ved from theoretical model calculations. Experimental data [8-9], 7443 1451 [16-20], [22], [24] were corrected to the new standards. Uncer- 7443 1451 tainty for cross section values measured by Bartolome et al. [4] 7443 1451 was evaluated as +-16%. Theoretical model calculations was done 7443 1451 by means of GNASH code [26]. New modified version of GNASH was 7443 1451 utilized for calculations [27]. Data from GNASH calculation were 7443 1451 dominant in the neutron energy range 4 - 20 MeV. 7443 1451 The evaluation W-186(n,g)W-187 excitation function in the 7443 1451 energy range above 100 keV has been carried out within the frame- 7443 1451 work of generalized least squares method , rational function was 7443 1451 used as model function [28]. Procedure of calculation recommended 7443 1451 cross section data was performed by means of PADE-2 code [29]. 7443 1451 Capture cross section at En=0.0253 eV and Resonance Integral 7443 1451 (0.5 to 2.0E+7 eV) from present evaluation are given below in the 7443 1451 Table 1. in a comparison with data from compilations [1] and [30].7443 1451 Table 1. 7443 1451 -------------------------------------------------- 7443 1451 Parameters Values, barn References 7443 1451 -------------------------------------------------- 7443 1451 Capture cross section 38.47 +- 0.85 This eval. 7443 1451 at En=0.0253 eV 37.9 +- 0.6 [ 1] 7443 1451 37 +- 2 [30] 7443 1451 38.5 +- 0.5 [31] 7443 1451 -------------------------------------------------- 7443 1451 Resonance Integral 482.2 This eval. 7443 1451 485 +- 15 [ 1] 7443 1451 485 +- 15 [30] 7443 1451 485 +- 15 [31] 7443 1451 -------------------------------------------------- 7443 1451 7443 1451 ----- MF=33 MT=102----- 7443 1451 Uncertainties in the evaluated excitation function for the 7443 1451 reaction W-186(n,g)W-187 are given in the three independent 7443 1451 matrixes. 7443 1451 In the energy range 1.000E-05 - 100 eV uncertainties are pre- 7443 1451 sented in the form of relative covariance matrix for the 14-neut- 7443 1451 ron energy groups (LB=5). Eigenvalues of this 6-th digits rela- 7443 1451 tive covariance matrix are the following: 7443 1451 7443 1451 1.14625E-10 1.09510E-07 7.28111E-07 8.61003E-06 7443 1451 4.80743E-05 5.34800E-05 5.91642E-05 6.60535E-05 7443 1451 8.44288E-05 1.48138E-04 1.95936E-04 2.39607E-04 7443 1451 8.15874E-04 8.38515E-03 7443 1451 7443 1451 In the energy range 0.1 - 8.5 keV uncertainties are given in 7443 1451 the form of diagonal matrix of uncertainties for 19-th neutron 7443 1451 energy intervals (LB=1) 7443 1451 In the energy range 8.5 keV - 20 MeV uncertainties are pre- 7443 1451 sented in the form of relative covariance matrix for the 24-neut- 7443 1451 ron energy groups (LB=5). Eigenvalues of this 6-th digits rela- 7443 1451 tive covariance matrix are the following: 7443 1451 7443 1451 6.81049E-10 1.45633E-05 5.83401E-05 1.88615E-04 7443 1451 6.00477E-04 7.08017E-04 9.54276E-04 1.08032E-03 7443 1451 1.26144E-03 1.35976E-03 1.50022E-03 1.66258E-03 7443 1451 1.66943E-03 2.15756E-03 2.87259E-03 3.83276E-03 7443 1451 6.81420E-03 9.70769E-03 9.99331E-03 1.17664E-02 7443 1451 1.48009E-02 1.93593E-02 2.04003E-02 2.26206E-01 7443 1451 7443 1451 References : 7443 1451 1. S.F.Mughabghab et al. Neutron Cross Sections, vol.1, part B, 7443 1451 New York, Academic Press, 1984 7443 1451 2. S.I.Sukhoruchkin et al. Landolt Bornstein New Series, v.I/16B,7443 1451 ed. H.Schopper, Springer, 1998, pp.w19-w23 7443 1451 3. S.V.Kapchigashev, Ju.P.Popov, F.L.Shapiro EXFOR 40755.005 7443 1451 4. Z.M.Bartolome et al. Nucl. Sci. Eng., v.37, pp.137-156, 1969 7443 1451 5. R.L.Macklin et al. Nucl. Sci. Eng., v.84, p.98, June 1983 7443 1451 6. M.V.Bokhovko et al. Voprosy Atomnoj Nauki i Tekhniki, Ser.: 7443 1451 Yadernye Konstanty, v.1, p.39, February 1986 7443 1451 7. G.N.Manturov et al. Voprosy Atomnoy Nauki i Tekhniki, Ser.: 7443 1451 Jadernye Konstanty, v.1, p.50, 1983 7443 1451 8. Yu.Ya.Stavisskii, V.A.Tolstikov Atomnaja Energija, v.9, 7443 1451 p.401, 1960 7443 1451 9. V.N.Kononov et al. Yadernaja Fizika, v.4, no.2, p.282, 1966 7443 1451 10. P.T.Guenther. et al. Phys. Rev. C., v.26, p.2433, 1982 7443 1451 11. V.N.Kononov et al. Sov J. of Nuclear Physics, v.46, part 1, 7443 1451 p.51, July 1987 7443 1451 12. R.Allen et al. Nature, v.161, p.727, May 1948 7443 1451 13. L.E.Beghian, H.H.Halban Nature, v.163, p.366, March 1949 7443 1451 14. J.L.Perkin et al. Proc. Phys. Soc., v.72, p.505, 1958 7443 1451 15. A.I.Leipunskij et al. Second UN Conf. on the Peaceful Uses of 7443 1451 Atomic Energy, Geneva, 1-13 September 1958, v.15, p.50(2219) 7443 1451 16. M.V.Pasechnik et al. Second UN Conf. on the Peaceful Uses of 7443 1451 Atomic Energy, Geneva, 1-13 September 1958, v.15, p.18(2030) 7443 1451 17. A.E.Johnsrud et al. Phys. Rev., v.116, p.927, 1959 7443 1451 18. G.G.Zaikin et al. Jaderno-Fizicheskie Issledovanija (USSR 7443 1451 progress report), v.6, p.103, 1968 7443 1451 19. M.Lindner et al. Nucl. Sci. Eng., v.59, p.381, April 1976 7443 1451 20. O.Schwerer et al. Nucl. Phys., v.A264, p.105, June 1976 7443 1451 21. M.Valkonen Report JU-RR-1/1976, Jyvaeskylae University, 1976 7443 1451 M.Valkonen et al. J. Inorg. Nucl. Chem., v.36, p.715, 1974 7443 1451 22. G.Magnusson et al. Physica Scripta, v.21, p.21, January 1980 7443 1451 23. J.Voignier et al. Report CEA-R-5089, August 1981 ; 7443 1451 J.Voignier et al. Nucl. Sci. Eng., v.93, p.43, 1986 7443 1451 24. Yu.N.Trofimov 1st International Conf. on Neutron Physics, 7443 1451 Kiev, 14-16 September 1987, v.3, p.331 7443 1451 25. Zhang Guohui, Lu Hanlin et al. Communication of Nuclear Data 7443 1451 Progress, No. 23, Atomic Energy Press, Beijing, June, 2000 7443 1451 26. P.G.Young, E.D.Arthur A Preequilibrium Statistical Nuclear 7443 1451 Model Code for Calculation of Cross Section and Emission 7443 1451 Spectra. Report LA-6947, Los Alamos, 1977 7443 1451 27. E.L.Trykov, G.Ya.Tertychnyi Private communication, IPPE, 7443 1451 Obninsk, May 1999 7443 1451 28. S.Badikov, N.Rabotnov, K.Zolotarev Proc. of NEANSC Speciali- 7443 1451 st's Meeting on Evaluation and Processing of Covariance Data, 7443 1451 Oak Ridge, USA, 7-9 September 1992, OECD, Paris, 1993, p.105 7443 1451 29. S.A.Badikov et al. Preprint FEI-1686, Obninsk, 1985 7443 1451 30. A.V.Ignatyuk et al. Landolt Bornstein New Series, v.I/16A, 7443 1451 Part 1, ed. H.Schopper, Springer, 1998, p.8-24 7443 1451 31. S.F.Mughabghab et al. Termal Neutron Capture Cross Sections 7443 1451 Resonance Integrals and G-factors, Report INDC(NDS)-440, 7443 1451 IAEA, Vienna, February 2003 7443 1451 ***************************************************************** 7443 1451 ********** End of (N,G) bibliographical component ********** 7443 1451 ***************************************************************** 7443 1451 74-W -186 LANL,ANL EVAL-OCT96 M.B.CHADWICK,P.G.YOUNG,ARTHUR 7443 1451 Ch96a,Ch96b,Ar80,Ch99DIST-SEP 1 REV1- 20010926 7443 1451 ----ENDF/B-VI MATERIAL 7443 7443 1451 MF=3 MT= 1 Total Cross Section 7443 1451 MT= 2 Elastic Scattering Cross Section 7443 1451 ***************************************************************** 7443 1451 7443 1451 ***************** Program LINEAR (VERSION 2002-1) ***************7443 1451 For All Data Greater than 1.0000E-10 barns in Absolute Value 7443 1451 Data Linearized to Within an Accuracy of .100000000 per-cent 7443 1451 ***************** Program RECENT (VERSION 2002-1) ***************7443 1451 for All Data Greater than 1.0000E-10 barns in Absolute Value 7443 1451 Data Linearized to within an Accuracy of .100000000 per-cent 7443 1451 ***************** Program SIGMA1 (VERSION 2002-1) ***************7443 1451 Data Doppler Broadened to 300.000000 Kelvin 7443 1451 for All Data Greater than 1.0000E-10 barns in Absolute Value 7443 1451 Data Linearized to Within an Accuracy pf .100000000 per-cent 7443 1451 ***************** Program FIXUP (Version 2002-1) ****************7443 1451 Corrected ZA/AWR in All Sections-----------------------------Yes 7443 1451 Corrected Thresholds-----------------------------------------Yes 7443 1451 Extended Cross Sections to 20 MeV----------------------------No 7443 1451 Allow Cross Section Deletion---------------------------------No 7443 1451 Allow Cross Section Reconstruction---------------------------No 7443 1451 Make All Cross Sections Non-Negative-------------------------Yes 7443 1451 Delete Energies Not in Ascending Order-----------------------Yes 7443 1451 Deleted Duplicate Points-------------------------------------Yes 7443 1451 Check for Ascending MAT/MF/MT Order--------------------------Yes 7443 1451 Check for Legal MF/MT Numbers--------------------------------Yes 7443 1451 Allow Creation of Missing Sections---------------------------No 7443 1451 Allow Insertion of Energy Points-----------------------------No 7443 1451 Create Uniform Energy Grid-----------------------------------No 7443 1451 Delete Section if Cross Section =0 at All Energies-----------Yes 7443 1451 ***************** Program GROUPIE (VERSION 2002-1) **************7443 1451 Unshielded Group Averages Using 640 Groups 7443 1451 Weighting Spectrum: Flat (Constant) Spectrum 7443 1451 1 451 206 17443 1451 3 102 217 17443 1451 33 102 94 17443 1451 7443 1 0 7443 0 0 7.41860E+4 1.84357E+2 0 0 0 07443 3102 5.46700E+6 5.46700E+6 0 0 1 6417443 3102 641 1 7443 3102 .000100000 603.335846 .000105000 589.267631 .000110000 575.9026837443 3102 .000115000 563.587364 .000120000 549.190298 .000127500 533.2169627443 3102 .000135000 518.623372 .000142500 505.212229 .000150000 490.6976817443 3102 .000160000 475.577316 .000170000 461.819668 .000180000 449.1912287443 3102 .000190000 437.441542 .000200000 426.713841 .000210000 416.6208847443 3102 .000220000 407.246173 .000230000 398.535864 .000240000 388.3117177443 3102 .000255000 377.076879 .000270000 368.407585 .000280000 358.7570467443 3102 .000300000 346.987888 .000320000 336.331282 .000340000 326.5715237443 3102 .000360000 317.582189 .000380000 309.376115 .000400000 300.8208007443 3102 .000425000 292.066023 .000450000 284.066608 .000475000 276.7330497443 3102 .000500000 269.836551 .000525000 263.533426 .000550000 257.5790457443 3102 .000575000 252.041159 .000600000 246.365181 .000630000 240.5260707443 3102 .000660000 235.183230 .000690000 230.057942 .000720000 224.6043127443 3102 .000760000 218.756748 .000800000 213.327421 .000840000 208.3517907443 3102 .000880000 203.622726 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2.33953- 3 1.94182- 3744333102 2.11518- 3 2.03561- 3 2.09740- 3 1.95226- 3 2.53405- 3 3.31215- 3744333102 3.60739- 3 3.96238- 3 4.21580- 3 4.16395- 3 4.06369- 3 4.07706- 3744333102 4.15960- 3 4.58220- 3 6.01328- 3 3.40942- 3 2.54538- 3 2.21842- 3744333102 2.21640- 3 1.99538- 3 2.04750- 3 2.01617- 3 2.05438- 3 1.93119- 3744333102 2.51020- 3 3.18624- 3 3.62121- 3 3.94521- 3 4.06882- 3 4.06998- 3744333102 4.06225- 3 4.02439- 3 4.03336- 3 4.55297- 3 5.90581- 3 3.67830- 3744333102 2.41506- 3 2.22968- 3 2.01926- 3 2.06278- 3 2.05497- 3 2.07286- 3744333102 1.94814- 3 2.54615- 3 3.17719- 3 3.67252- 3 4.02448- 3 4.07891- 3744333102 4.08203- 3 4.12597- 3 4.07952- 3 4.04296- 3 4.61531- 3 5.96228- 3744333102 4.07182- 3 3.04717- 3 2.72009- 3 2.85865- 3 2.82259- 3 2.84474- 3744333102 2.72097- 3 3.56800- 3 4.52475- 3 5.12145- 3 5.62007- 3 5.79392- 3744333102 5.76950- 3 5.76004- 3 5.72564- 3 5.73317- 3 6.46176- 3 8.39467- 3744333102 5.20771- 3 3.71857- 3 3.46976- 3 3.46460- 3 3.50192- 3 3.41207- 3744333102 4.46116- 3 5.64115- 3 6.44819- 3 7.01240- 3 7.20858- 3 7.22485- 3744333102 7.22639- 3 7.14667- 3 7.15233- 3 8.09342- 3 1.04846- 2 4.89023- 3744333102 3.75588- 3 3.59533- 3 3.56163- 3 3.47878- 3 4.58232- 3 5.68368- 3744333102 6.55293- 3 7.26525- 3 7.33644- 3 7.29426- 3 7.39179- 3 7.34179- 3744333102 7.25344- 3 8.27546- 3 1.07087- 2 4.83225- 3 3.76382- 3 3.75857- 3744333102 3.69770- 3 4.82828- 3 6.14842- 3 6.95657- 3 7.58153- 3 7.84944- 3744333102 7.84228- 3 7.80493- 3 7.74221- 3 7.77826- 3 8.75400- 3 1.13701- 2744333102 5.08337- 3 3.78225- 3 3.58894- 3 4.74728- 3 6.10842- 3 6.75552- 3744333102 7.47777- 3 7.81263- 3 7.71529- 3 7.63195- 3 7.64217- 3 7.70332- 3744333102 8.58921- 3 1.12211- 2 4.38207- 3 3.75869- 3 4.78257- 3 6.12157- 3744333102 6.88179- 3 7.50752- 3 7.79883- 3 7.77937- 3 7.72847- 3 7.67689- 3744333102 7.72032- 3 8.67445- 3 1.12767- 2 5.79606- 3 4.87618- 3 5.89683- 3744333102 6.84062- 3 7.53188- 3 7.58778- 3 7.57968- 3 7.69012- 3 7.61139- 3744333102 7.51960- 3 8.59844- 3 1.11077- 2 1.05503- 2 9.17240- 3 8.44451- 3744333102 9.82314- 3 1.03972- 2 1.00600- 2 9.92327- 3 1.00586- 2 1.01197- 2744333102 1.12332- 2 1.47331- 2 1.82857- 2 1.56549- 2 1.18555- 2 1.18792- 2744333102 1.32773- 2 1.33542- 2 1.24270- 2 1.25225- 2 1.45984- 2 1.84671- 2744333102 2.38468- 2 2.09590- 2 1.38269- 2 1.27699- 2 1.50633- 2 1.52120- 2744333102 1.36771- 2 1.66054- 2 2.12396- 2 2.73440- 2 2.31073- 2 1.57553- 2744333102 1.37127- 2 1.58871- 2 1.66958- 2 1.70147- 2 2.33616- 2 2.84616- 2744333102 2.33479- 2 1.68693- 2 1.48460- 2 1.68731- 2 1.83454- 2 2.35919- 2744333102 2.69987- 2 2.38452- 2 1.67894- 2 1.47043- 2 1.94187- 2 2.35998- 2744333102 2.63049- 2 2.09755- 2 1.50714- 2 1.83217- 2 2.41037- 2 2.35921- 2744333102 1.97452- 2 1.67774- 2 2.36775- 2 2.31409- 2 2.15559- 2 2.28705- 2744333102 3.01605- 2 2.94807- 2 4.61913- 2 744333102 744333 0 7443 0 0 0 0 0 7.91970E+4 1.95274E+2 0 0 34 107925 1451 0.0 0.0 0 0 0 67925 1451 1.00000E+0 3.00000E+7 0 0 10 20027925 1451 3.00000E+2 0.0 1 0 202 57925 1451 79-Au-197 LANL/IRK EVAL-JAN84 P.G.YOUNG,VONACH ET AL 7925 1451 DIST-Feb2004 7925 1451 ----IRDF-2002 MATERIAL 7925 7925 1451 -----INCIDENT NEUTRON DATA 7925 1451 ------ENDF-6 FORMAT 7925 1451 *****************************************************************7925 1451 79-AU-197 LANL EVAL-JAN84 P.G.YOUNG 7925 1451 LA-10069-PR DIST-SEP91 REV1-JUL91 19930129 7925 1451 ----ENDF/B-VI MATERIAL 7925 REVISION 1 7925 1451 *****************************************************************7925 1451 *****************EXTRACT FOR SPECIAL PURPOSE FILE*****************7925 1451 DOSIMETRY 7925 1451 ******************************************************************7925 1451 7925 1451 MOD1 OF ENDF/B-VI 7925 1451 7925 1451 The following revisions were made for MOD1 of ENDF/B-VI: 7925 1451 7925 1451 1. MF=1,MT=451 - Comments were added regarding estimated 7925 1451 (expanded) covariance for the Standards Cross Sections. 7925 1451 2. MF=3,MT=102 - Q-value corrected. 7925 1451 7925 1451 *****************************************************************7925 1451 ++++++++++++++++++++++++++++++++++++++++++++++++++++++++++++++++++7925 1451 MF/MT 2151, MF/MT 3 16 AND MF/MT 33 16 FROM FOLLOWING 7925 1451 EVALUATION 7925 1451 79-AU-197 IRK-VIENNA EVAL-APR90 7925 1451 DIST-JUN90 7925 1451 IRK-EVAL.NLIB 25 7925 7925 1451 ++++++++++++++++++++++++++++++++++++++++++++++++++++++++++++++++++7925 1451 *************** SUMMARY OF ENDF/B-VI EVALUATION ******************7925 1451 7925 1451 A new evaluation of all neutron and gamma-ray data above the 7925 1451 resonance region is joined with the endf/b-v resolved resonance 7925 1451 region evaluation and with the Version VI standard cross section 7925 1451 for the (n,gamma) reaction below a neutron energy of 2.5 MeV. 7925 1451 7925 1451 *************** GENERAL DESCRIPTION ******************************7925 1451 7925 1451 P.G.Young and E.D.Arthur 7925 1451 7925 1451 the new evaluation for files 3,4,5,12,13,14,15 is based on 7925 1451 statistical theory, hauser-feshbach, preequilibrium calculations 7925 1451 with the comnuc and gnash codes (ref1,2). deformed optical poten-7925 1451 tial of delaroche and ecis coupled-channel code were used to cal-7925 1451 culate neutron transmission coefficients and total and elastic 7925 1451 elastic cross sections (ref3,4). gamma-ray strength functions 7925 1451 were obtained by fitting morgan n,xg data (ref5) at 0.4 and 6.5 7925 1451 mev. calculated results were used for all major reactions except 7925 1451 total cross section. for total, the theoretical cross section 7925 1451 was used as prior in covariance analysis of experimental data 7925 1451 using glucs code (ref6). more details on experimental data used 7925 1451 are given below and in main reference for evaluation (ref 7). 7925 1451 7925 1451 **************************************************************** 7925 1451 7925 1451 STANDARDS COVARIANCES 7925 1451 7925 1451 Phase 1 reviewers of the ENDF/B-VI standards cross sections have 7925 1451 expressed the concern that the uncertainties resulting from the 7925 1451 combination of R-matrix and simultaneous evaluations might have 7925 1451 led to uncertainties that are too small. As a result, the 7925 1451 Standards Subcommittee produced (at the May, 1990 CSEWG meeting) 7925 1451 a set of expanded covariance estimates for the standard cross 7925 1451 section reactions. These uncertainties are estimates such that 7925 1451 if a modern day experiment were performed on a given standard 7925 1451 cross section using the best techniques, approximately 2/3 of 7925 1451 the results should fall within these expanded uncertainties. The 7925 1451 expanded uncertainties for the Au-197(n,gamma) cross section are 7925 1451 given in the following table and are compared to values from the 7925 1451 combined output of the standards covariance analysis: 7925 1451 7925 1451 Energy Range Estimated Uncertainty Combined Analysis 7925 1451 (keV) (percent) (percent) 7925 1451 7925 1451 2.53E-05 0.14 0.14 7925 1451 200 - 500 3.0 1.31 7925 1451 500 - 1000 3.5 2.1 7925 1451 1000 - 2500 4.5 2.0 7925 1451 7925 1451 7925 1451 *************** mf=2 resonance parameters ************************7925 1451 7925 1451 mt=151 resolved resonance parameters given from 1.0e-05ev 7925 1451 to 2 kev based on ref8 and references therein 7925 1451 and a bound level. some of the reson. spin assignments 7925 1451 from ref9. from 2 to 4.827 kev the parameters are based7925 1451 on macklin et al and hoffman et al normalized data. 7925 1451 see refs 10 and 11. 7925 1451 thermal cross sections are as follows: 7925 1451 capture = 98.71 b 7925 1451 scattering = 6.84 b 7925 1451 total = 105.55 b 7925 1451 the absorption resonance integral is 1559 b 7925 1451 7925 1451 *************** mf=3 smooth neutron cross sections ***************7925 1451 7925 1451 mt= 1 total cross section. based on glucs covariance analysis 7925 1451 using deformed optical model calculation as the prior and 7925 1451 experimental data from refs 12-22, 29 for fitting. 7925 1451 mt= 2 elastic cross section. difference of mt=1 and sum of 7925 1451 all nonelastic cross sections. closely approximates theore- 7925 1451 tical results. 7925 1451 mt=102 (n,gamma) cross section. below 2.5 MeV, adopted the 7925 1451 ENDF/B-VI standard cross section (Ref.30,31) down to the 7925 1451 resonance region. At higher energies, the theoretical cal- 7925 1451 culations were adjusted to agree with experimental data. A 7925 1451 semi-direct component normalized to an average of experimental7925 1451 data at 14 MeV was included above En = 6 MeV. 7925 1451 at higher energies, use theoretical calculations, which agree 7925 1451 reasonably with available exp. data. above 5 mev, calculation7925 1451 includes semi-direct component normalized to average of 7925 1451 14 mev data. 7925 1451 7925 1451 *************** mf=33 neutron cross section covariances **********7925 1451 7925 1451 mt= 1 total cross section covariance from glucs analysis. 7925 1451 7925 1451 *************** references ***************************************7925 1451 7925 1451 1. c.l.dunford. ai-aec-12931(1970) 7925 1451 2. p.g.young, e.d.arthur, la-6947 (1977). 7925 1451 3. j.p.delaroche, harwell conference (1978)p.366. 7925 1451 4. j.raynal, iaea smr-9/8 (1970). 7925 1451 5. g.l.morgan, e.newman, ornl-tm-4973 (1975). 7925 1451 6. d.m.hetrick, c.y.fu, ornl/tm-7341 (1980). 7925 1451 7. p.g.young, e.d.arthur, in la-10069-pr (1984)p.12. 7925 1451 8. s.f.mughabghab and d.i.garber bnl-325,3rd edn,vol i(1973). 7925 1451 9. a.lottin and a.jain conf on nuclear structure study with 7925 1451 neutrons,budapest,1972 p34 and private communication. 7925 1451 10. r.macklin et al. phys. rev/c 11,1270(1975) and private 7925 1451 communication. 7925 1451 11. m.m. hoffman et al. 71knoxville conf., 2, 868(1971) 7925 1451 12. w.poenitz et al., nuc.sci.eng. 78, 333(1981). 7925 1451 13. d.g.foster jr., d.glasgow, phys.rev. c3, 576(1971). 7925 1451 14. k.k.seth,phys.letters,16,306(1965). 7925 1451 15. s.c.snowdon, phys.rev. 90, 615(1953). 7925 1451 16. j.f.whalen,anl-7210,16(1966). 7925 1451 17. n.nereson, phys.rev. 94, 1678(1954). 7925 1451 18. a.bratenahl et al., phys.rev. 110, 927(1958). 7925 1451 19. j.p.conner,phys.rev.109,1268(1958). 7925 1451 20. j.h.coon,phys.rev.88,562(1952). 7925 1451 21. j.m.peterson,phys.rev.120,521(1960). 7925 1451 22. e.g.bilpuch,private communication(1959). 7925 1451 23. j.frehaut et al, proc. 10-50 mev conf, bnl-ncs-51245 (1980) 7925 1451 page 399. 7925 1451 24. l.r.veeser et al, phys.rev. c16, 1792(1977). 7925 1451 25. b.p.bayhurst et al, phys.rev. c12, 451(1975). 7925 1451 26. r.j.prestwood and b.p.bayhurst,phys.rev.121,1438(1961). 7925 1451 27. c.kalbach and f.mann, bnl-ncs-5/245,p.689 (1980). 7925 1451 28. v.j.orphan et al, ga-10248 (1970). 7925 1451 29. d.c.larson, proc. 10-50 mev conf, bnl-ncs-51245 (1980) p.277.7925 1451 30. A.Carlson et al., Nuc.Data for Basic & Applied Science, 7925 1451 Santa Fe, NM (1985) p.1429. 7925 1451 31. W.Poenitz, ANL-West, personnal communication (1989). 7925 1451 7925 1451 ******************************************************************7925 1451 7925 1451 7925 1451 7925 1451 7925 1451 ***************** PROGRAM FIXUP (VERSION 86-2) ******************7925 1451 *RECONSTRUCTED MT NUMBERS 7925 1451 4 =+( 51, 91) 7925 1451 103 =+(700,718) 7925 1451 104 =+(720,738) 7925 1451 105 =+(740,758) 7925 1451 106 =+(760,778) 7925 1451 107 =+(780,798) 7925 1451 101 =+(102,114) 7925 1451 27 =+( 18, 18)+(101,101) 7925 1451 3 =+( 4, 4)+( 6, 9)+( 16, 17)+( 22, 37) 7925 1451 19 =+( 18, 18)-( 20, 21)-( 38, 38) 7925 1451 1 =+( 2, 3) 7925 1451 ***************** Program LINEAR (VERSION 2002-1) ***************7925 1451 For All Data Greater than 1.0000E-10 barns in Absolute Value 7925 1451 Data Linearized to Within an Accuracy of .100000000 per-cent 7925 1451 ***************** Program RECENT (VERSION 2002-1) ***************7925 1451 for All Data Greater than 1.0000E-10 barns in Absolute Value 7925 1451 Data Linearized to within an Accuracy of .100000000 per-cent 7925 1451 ***************** Program SIGMA1 (VERSION 2002-1) ***************7925 1451 Data Doppler Broadened to 300.000000 Kelvin 7925 1451 for All Data Greater than 1.0000E-10 barns in Absolute Value 7925 1451 Data Linearized to Within an Accuracy pf .100000000 per-cent 7925 1451 ***************** Program FIXUP (Version 2002-1) ****************7925 1451 Corrected ZA/AWR in All Sections-----------------------------Yes 7925 1451 Corrected Thresholds-----------------------------------------Yes 7925 1451 Extended Cross Sections to 20 MeV----------------------------No 7925 1451 Allow Cross Section Deletion---------------------------------No 7925 1451 Allow Cross Section Reconstruction---------------------------No 7925 1451 Make All Cross Sections Non-Negative-------------------------Yes 7925 1451 Delete Energies Not in Ascending Order-----------------------Yes 7925 1451 Deleted Duplicate Points-------------------------------------Yes 7925 1451 Check for Ascending MAT/MF/MT Order--------------------------Yes 7925 1451 Check for Legal MF/MT Numbers--------------------------------Yes 7925 1451 Allow Creation of Missing Sections---------------------------No 7925 1451 Allow Insertion of Energy Points-----------------------------No 7925 1451 Create Uniform Energy Grid-----------------------------------No 7925 1451 Delete Section if Cross Section =0 at All Energies-----------Yes 7925 1451 ***************** Program GROUPIE (VERSION 2002-1) **************7925 1451 Unshielded Group Averages Using 640 Groups 7925 1451 Weighting Spectrum: Flat (Constant) Spectrum 7925 1451 1 451 211 27925 1451 3 16 43 17925 1451 3 102 217 27925 1451 33 16 71 17925 1451 33 102 26 17925 1451 7925 1 0 7925 0 0 7.91970E+4 1.95274E+2 0 0 0 07925 3 16 -8.08000E+6-8.08000E+6 0 0 1 1207925 3 16 120 1 7925 3 16 8100000.00 .005579896 8200000.00 .023230150 8300000.00 .0417860347925 3 16 8400000.00 .071166000 8500000.00 .104974000 8600000.00 .1404245007925 3 16 8700000.00 .179890000 8800000.00 .220456917 8900000.00 .2812400007925 3 16 9000000.00 .350293333 9100000.00 .419346667 9200000.00 .4888333337925 3 16 9300000.00 .560920000 9400000.00 .633440000 9500000.00 .7059600007925 3 16 9600000.00 .778480000 9700000.00 .851136750 9800000.00 .9246140007925 3 16 9900000.00 .998228000 10000000.0 1.07184200 10100000.0 1.145456007925 3 16 10200000.0 1.21760825 10300000.0 1.28099000 10400000.0 1.342910007925 3 16 10500000.0 1.40483000 10600000.0 1.46675000 10700000.0 1.524579007925 3 16 10800000.0 1.55786200 10900000.0 1.58705400 11000000.0 1.616246007925 3 16 11100000.0 1.64543800 11200000.0 1.67711725 11300000.0 1.723720007925 3 16 11400000.0 1.77281000 11500000.0 1.82190000 11600000.0 1.870990007925 3 16 11700000.0 1.91616500 11800000.0 1.93785000 11900000.0 1.955620007925 3 16 12000000.0 1.97339000 12100000.0 1.99116000 12200000.0 2.008416687925 3 16 12300000.0 2.02259345 12400000.0 2.03625690 12500000.0 2.049920347925 3 16 12600000.0 2.06358379 12700000.0 2.07724724 12800000.0 2.090910697925 3 16 12900000.0 2.10411356 13000000.0 2.10718360 13100000.0 2.106108407925 3 16 13200000.0 2.10503320 13300000.0 2.10395800 13400000.0 2.102882807925 3 16 13500000.0 2.10180760 13600000.0 2.10313545 13700000.0 2.106866367925 3 16 13800000.0 2.11048693 13900000.0 2.11168000 14000000.0 2.111880007925 3 16 14100000.0 2.11336363 14200000.0 2.11798500 14300000.0 2.122594637925 3 16 14400000.0 2.12380800 14500000.0 2.12363200 14600000.0 2.124854387925 3 16 14700000.0 2.12949500 14800000.0 2.13413834 14900000.0 2.135423147925 3 16 15000000.0 2.13533400 15100000.0 2.13524486 15200000.0 2.135155717925 3 16 15300000.0 2.13506657 15400000.0 2.13497743 15500000.0 2.134888297925 3 16 15600000.0 2.13479914 15700000.0 2.13272231 15800000.0 2.118719337925 3 16 15900000.0 2.10272867 16000000.0 2.08673800 16100000.0 2.070747337925 3 16 16200000.0 2.05475667 16300000.0 2.03876600 16400000.0 2.022775337925 3 16 16500000.0 2.00678467 16600000.0 1.99079400 16700000.0 1.974803337925 3 16 16800000.0 1.95881267 16900000.0 1.94282200 17000000.0 1.926831337925 3 16 17100000.0 1.91084067 17200000.0 1.89294833 17300000.0 1.863646007925 3 16 17400000.0 1.83244200 17500000.0 1.80123800 17600000.0 1.770034007925 3 16 17700000.0 1.73883000 17800000.0 1.70762600 17900000.0 1.676422007925 3 16 18000000.0 1.64521800 18100000.0 1.61401400 18200000.0 1.582810007925 3 16 18300000.0 1.55160600 18400000.0 1.52040200 18500000.0 1.483501767925 3 16 18600000.0 1.44090529 18700000.0 1.39830882 18800000.0 1.355712357925 3 16 18900000.0 1.31311588 19000000.0 1.27051941 19100000.0 1.227922947925 3 16 19200000.0 1.18532647 19300000.0 1.14579943 19400000.0 1.124689007925 3 16 19500000.0 1.10664800 19600000.0 1.08860700 19700000.0 1.070566007925 3 16 19800000.0 1.05252500 19900000.0 1.03448400 20000000.0 0.0 7925 3 16 7925 3 0 7.91970E+4 1.95274E+2 0 0 0 07925 3102 6.51238E+6 6.51238E+6 0 0 1 6417925 3102 641 1 7925 3102 .000100000 1537.29252 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1.337000-5 3.556000-5 1.379000-4 1.431000-4792533 16 1.085000-4 3.331000-4 0.000000+0 0.000000+0 0.000000+0 0.000000+0792533 16 0.000000+0 2.957000-3 5.334000-4 3.684000-4 1.255000-5 7.771000-6792533 16 1.161000-5 1.992000-5 1.165000-5 2.903000-5 1.202000-4 1.247000-4792533 16 9.458000-5 2.902000-4 0.000000+0 0.000000+0 0.000000+0 0.000000+0792533 16 0.000000+0 9.967000-4 1.245000-4 0.000000+0 3.713000-6 0.000000+0792533 16 4.314000-6 0.000000+0 4.769000-6 0.000000+0 1.268000-4 0.000000+0792533 16 0.000000+0 2.199000-4 0.000000+0 0.000000+0 0.000000+0 0.000000+0792533 16 5.889000-4 7.578000-6 1.814000-6 7.933000-6 6.584000-6 7.627000-6792533 16 1.587000-5 1.079000-4 2.119000-4 2.050000-4 7.207000-5 2.340000-4792533 16 2.849000-4 3.354000-4 3.826000-4 0.000000+0 1.742000-4 1.488000-4792533 16 1.488000-4 1.488000-4 1.488000-4 1.488000-4 6.333000-6 6.335000-6792533 16 6.361000-6 8.390000-6 0.000000+0 0.000000+0 0.000000+0 0.000000+0792533 16 0.000000+0 1.664000-4 1.488000-4 1.488000-4 1.488000-4 1.488000-4792533 16 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0792533 16 0.000000+0 0.000000+0 0.000000+0 1.716000-4 1.488000-4 1.488000-4792533 16 1.488000-4 3.214000-6 6.714000-6 5.151000-6 7.761000-6 7.435000-6792533 16 9.052000-6 1.066000-5 1.216000-5 0.000000+0 1.690000-4 1.488000-4792533 16 1.488000-4 3.014000-6 3.127000-6 2.372000-6 7.278000-6 0.000000+0792533 16 0.000000+0 0.000000+0 0.000000+0 0.000000+0 1.690000-4 1.488000-4792533 16 6.608000-6 6.578000-6 6.832000-6 7.785000-6 0.000000+0 0.000000+0792533 16 0.000000+0 0.000000+0 0.000000+0 1.904000-4 5.274000-6 1.145000-5792533 16 8.791000-6 1.273000-5 1.316000-5 1.603000-5 1.886000-5 2.152000-5792533 16 0.000000+0 6.799000-4 1.240000-4 1.451000-4 8.034000-5 0.000000+0792533 16 0.000000+0 0.000000+0 0.000000+0 2.339000-4 7.292000-4 2.290000-4792533 16 8.333000-5 3.463000-4 3.044000-4 3.583000-4 4.087000-4 0.000000+0792533 16 5.244000-4 6.321000-5 1.939000-4 2.361000-4 2.779000-4 3.171000-4792533 16 0.000000+0 1.660000-3 0.000000+0 0.000000+0 0.000000+0 0.000000+0792533 16 0.000000+0 1.451000-3 6.695000-4 7.881000-4 8.991000-4 0.000000+0792533 16 2.126000-3 9.595000-4 1.095000-3 0.000000+0 2.659000-3 1.288000-3792533 16 0.000000+0 3.057000-3 0.000000+0 1.717000-3 792533 16 792533 0 7.91970E+4 1.95274E+2 0 0 0 1792533102 0.000000+0 0.000000+0 0 102 0 2792533102 0.000000+0 0.000000+0 0 1 8 4792533102 1.000000-5 0.000000+0 2.900000+6 5.000000-2 1.000000+7 6.250000-2792533102 2.000000+7 0.000000+0 792533102 0.000000+0 0.000000+0 1 5 120 15792533102 1.000000-5 3.000000-1 2.000000+3 1.000000+4 3.750000+4 9.000000+4792533102 1.350000+5 2.050000+5 2.750000+5 4.125000+5 6.250000+5 9.500000+5792533102 1.325000+6 1.900000+6 2.900000+6 1.968800-6 1.968800-6 1.197701-7792533102 1.277322-7 1.199737-7 1.180243-7 1.210327-7 1.175025-7 1.184462-7792533102 1.362774-7 1.098021-7 1.079868-7 1.089638-7 9.914542-8 2.968800-6792533102 1.197701-7 1.277322-7 1.199737-7 1.180243-7 1.210327-7 1.175025-7792533102 1.184462-7 1.362774-7 1.098021-7 1.079868-7 1.089638-7 9.914542-8792533102 1.094560-4 4.577940-5 4.677039-5 3.948904-5 3.768945-5 3.747630-5792533102 3.690735-5 3.428047-5 3.338854-5 3.229846-5 3.154064-5 2.637712-5792533102 6.056941-5 4.202671-5 3.750888-5 3.574985-5 3.503839-5 3.439512-5792533102 3.204558-5 3.144094-5 3.085597-5 2.984776-5 2.493463-5 6.285997-5792533102 4.547523-5 4.201196-5 4.049420-5 3.979974-5 3.715363-5 3.659218-5792533102 3.509315-5 3.465730-5 2.895877-5 8.021999-5 4.906673-5 4.591913-5792533102 4.475850-5 4.261771-5 4.283861-5 4.025727-5 4.041204-5 3.382565-5792533102 8.028292-5 5.304576-5 4.689840-5 4.483827-5 4.540301-5 4.299516-5792533102 4.362940-5 3.699681-5 9.958098-5 5.534255-5 4.876851-5 5.031469-5792533102 4.706249-5 4.804336-5 4.072953-5 9.094943-5 5.454792-5 5.269018-5792533102 4.994396-5 5.121352-5 4.343583-5 8.936425-5 5.920198-5 5.242733-5792533102 5.357873-5 4.570034-5 1.241536-4 6.756325-5 6.367727-5 5.555147-5792533102 1.336240-4 7.281097-5 5.539927-5 1.846612-4 8.162500-5 2.399631-4792533102 792533 0 7925 0 0 0 0 0 8.01990E+4 1.97259E+2 0 0 34 108034 1451 0.0 0.0 0 0 0 68034 1451 1.00000E+0 2.00000E+7 0 0 10 20028034 1451 3.00000E+2 0.0 1 0 78 48034 1451 80-Hg-199 JAERI EVAL-MAR90 K.SAKURAI 8034 1451 DIST-Feb2004 8034 1451 ----IRDF-2002 MATERIAL 8034 8034 1451 -----INCIDENT NEUTRON DATA 8034 1451 ------ENDF-6 FORMAT 8034 1451 *************************************************************** 8034 1451 80-HG-199 JAERI EVAL-MAR90 K.SAKURAI 8034 1451 DIST-JUL91 8034 1451 ----JENDL/D-99 MATERIAL 8034 8034 1451 *************************************************************** 8034 1451 HISTORY 8034 1451 90-03 EVALUATION WAS MADE BY K. SAKURA(JAERI). 8034 1451 90-10 COMPILED TO JENDL DOSIMETRY FILE VERSION 1. 8034 1451 91-07 DESCRIPTIVE DATA WERE ADDED. 8034 1451 8034 1451 ===== POINT-WISE DATA FILE ===== 8034 1451 8034 1451 8034 1451 HG-199 (N,N')HG-199M (HALF-LIFE = 42.6 M) 8034 1451 8034 1451 MF=1 GENERAL INFORMATION 8034 1451 MT=451 DESCRIPTIVE DATA AND DICTIONARY 8034 1451 8034 1451 MF=2 RESONANCE PARAMETERS 8034 1451 MT=151 PARAMETERS 8034 1451 ONLY SPIN AND SCATTERING RADIUS ARE GIVEN. 8034 1451 8034 1451 MF=3 NEUTRON CROSS SECTIONS 8034 1451 MT=51 HG-199(N,N')HG-199M CROSS SECTION 8034 1451 ISOMERIC STATE = 0.531 MEV 13/2+, HALF-LIFE = 42.6 M 8034 1451 EVALUATION WAS BASED ON THE EXPERIMENTAL DATA REPORTED IN 8034 1451 REF./1/. 8034 1451 8034 1451 MF=33 COVARIANCES OF NEUTRON CROSS SECTIONS 8034 1451 MT=51 8034 1451 ESTIMATED FROM THE EXPERIMENTAL DATA. 8034 1451 8034 1451 REFERENCE 8034 1451 1) SAKURAI K., ET AL.: J. NUCL. SCI. TECHNOL., 19, 775 (1982). 8034 1451 8034 1451 *************************************************************** 8034 1451 ----- MF=3 MT=51 ----- 8034 1451 For the IRDF-2002 file this reaction was converted at 8034 1451 IAEA/NDS. 8034 1451 The reaction MF/MT=3/51 was converted to MF/MT=10/ 4 8034 1451 The corresponding co-variance files were also converted 8034 1451 The reaction MF/MT=33/51 was converted to MF/MT=40/ 4 8034 1451 *************************************************************** 8034 1451 The threshold energy was recalculated from 5.25000+5 to 8034 1451 5.33691+5 to satisfy code FIZCON criteria 8034 1451 8034 1451 8034 1451 *************************************************************** 8034 1451 ***************** Program LINEAR (VERSION 2002-1) ***************8034 1451 For All Data Greater than 1.0000E-10 barns in Absolute Value 8034 1451 Data Linearized to Within an Accuracy of .100000000 per-cent 8034 1451 ***************** Program SIGMA1 (VERSION 2002-1) ***************8034 1451 Data Doppler Broadened to 300.000000 Kelvin 8034 1451 for All Data Greater than 1.0000E-10 barns in Absolute Value 8034 1451 Data Linearized to Within an Accuracy pf .100000000 per-cent 8034 1451 ***************** Program FIXUP (Version 2002-1) ****************8034 1451 Corrected ZA/AWR in All Sections-----------------------------Yes 8034 1451 Corrected Thresholds-----------------------------------------Yes 8034 1451 Extended Cross Sections to 20 MeV----------------------------No 8034 1451 Allow Cross Section Deletion---------------------------------No 8034 1451 Allow Cross Section Reconstruction---------------------------No 8034 1451 Make All Cross Sections Non-Negative-------------------------Yes 8034 1451 Delete Energies Not in Ascending Order-----------------------Yes 8034 1451 Deleted Duplicate Points-------------------------------------Yes 8034 1451 Check for Ascending MAT/MF/MT Order--------------------------Yes 8034 1451 Check for Legal MF/MT Numbers--------------------------------Yes 8034 1451 Allow Creation of Missing Sections---------------------------No 8034 1451 Allow Insertion of Energy Points-----------------------------No 8034 1451 Create Uniform Energy Grid-----------------------------------No 8034 1451 Delete Section if Cross Section =0 at All Energies-----------Yes 8034 1451 ***************** Program GROUPIE (VERSION 2002-1) **************8034 1451 Unshielded Group Averages Using 640 Groups 8034 1451 Weighting Spectrum: Flat (Constant) Spectrum 8034 1451 1 451 86 08034 1451 8 4 2 08034 1451 10 4 72 08034 1451 40 4 17 08034 1451 8034 1 0 8034 0 0 8.01990E+4 1.97259E+2 0 0 1 08034 3291 0.0 -5.30999E+5 80199 0 1 2058034 3291 205 1 8034 3291 525000.000 .000774677 550000.000 .004195311 575000.000 .0078359368034 3291 600000.000 .011840624 630000.000 .016209374 660000.000 .0205781248034 3291 690000.000 .024946875 720000.000 .030024219 760000.000 .0354000008034 3291 800000.000 .040600000 840000.000 .045800000 880000.000 .0510000008034 3291 920000.000 .056200000 960000.000 .061400000 1000000.00 .0714000008034 3291 1100000.00 .086200000 1200000.00 .102100000 1300000.00 .1246000008034 3291 1400000.00 .148200000 1500000.00 .174000000 1600000.00 .2020000008034 3291 1700000.00 .229000000 1800000.00 .250000000 1900000.00 .2700000008034 3291 2000000.00 .290000000 2100000.00 .310000000 2200000.00 .3292500008034 3291 2300000.00 .344000000 2400000.00 .358000000 2500000.00 .3750000008034 3291 2600000.00 .395000000 2700000.00 .414250000 2800000.00 .4290000008034 3291 2900000.00 .443000000 3000000.00 .457000000 3100000.00 .4710000008034 3291 3200000.00 .484000000 3300000.00 .491000000 3400000.00 .4970000008034 3291 3500000.00 .506000000 3600000.00 .518000000 3700000.00 .5295000008034 3291 3800000.00 .538000000 3900000.00 .546000000 4000000.00 .5540000008034 3291 4100000.00 .562000000 4200000.00 .569500000 4300000.00 .5740000008034 3291 4400000.00 .578000000 4500000.00 .582000000 4600000.00 .5860000008034 3291 4700000.00 .590000000 4800000.00 .594000000 4900000.00 .5980000008034 3291 5000000.00 .600400000 5100000.00 .601200000 5200000.00 .6023000008034 3291 5300000.00 .605200000 5400000.00 .608400000 5500000.00 .6120000008034 3291 5600000.00 .616000000 5700000.00 .619550000 5800000.00 .6204000008034 3291 5900000.00 .620800000 6000000.00 .622900000 6100000.00 .6267000008034 3291 6200000.00 .630500000 6300000.00 .634300000 6400000.00 .6381000008034 3291 6500000.00 .640000000 6600000.00 .640000000 6700000.00 .6400000008034 3291 6800000.00 .640000000 6900000.00 .640000000 7000000.00 .6400000008034 3291 7100000.00 .640000000 7200000.00 .640000000 7300000.00 .6400000008034 3291 7400000.00 .640000000 7500000.00 .640000000 7600000.00 .6400000008034 3291 7700000.00 .640000000 7800000.00 .640000000 7900000.00 .6400000008034 3291 8000000.00 .637000000 8100000.00 .631000000 8200000.00 .6256250008034 3291 8300000.00 .624000000 8400000.00 .623000000 8500000.00 .6220000008034 3291 8600000.00 .621000000 8700000.00 .619625000 8800000.00 .6160000008034 3291 8900000.00 .612000000 9000000.00 .606000000 9100000.00 .5980000008034 3291 9200000.00 .590500000 9300000.00 .586000000 9400000.00 .5820000008034 3291 9500000.00 .576000000 9600000.00 .568000000 9700000.00 .5595000008034 3291 9800000.00 .548000000 9900000.00 .536000000 10000000.0 .5260000008034 3291 10100000.0 .518000000 10200000.0 .510500000 10300000.0 .5060000008034 3291 10400000.0 .502000000 10500000.0 .494000000 10600000.0 .4820000008034 3291 10700000.0 .470500000 10800000.0 .462000000 10900000.0 .4540000008034 3291 11000000.0 .442000000 11100000.0 .426000000 11200000.0 .4105000008034 3291 11300000.0 .398000000 11400000.0 .386000000 11500000.0 .3740000008034 3291 11600000.0 .362000000 11700000.0 .349000000 11800000.0 .3300000008034 3291 11900000.0 .310000000 12000000.0 .296000000 12100000.0 .2880000008034 3291 12200000.0 .279500000 12300000.0 .268000000 12400000.0 .2560000008034 3291 12500000.0 .244000000 12600000.0 .232000000 12700000.0 .2205000008034 3291 12800000.0 .212000000 12900000.0 .204000000 13000000.0 .1960000008034 3291 13100000.0 .188000000 13200000.0 .180500000 13300000.0 .1760000008034 3291 13400000.0 .172000000 13500000.0 .168000000 13600000.0 .1640000008034 3291 13700000.0 .160000000 13800000.0 .156000000 13900000.0 .1520000008034 3291 14000000.0 .148000000 14100000.0 .144000000 14200000.0 .1400000008034 3291 14300000.0 .136000000 14400000.0 .132000000 14500000.0 .1280000008034 3291 14600000.0 .124000000 14700000.0 .120250000 14800000.0 .1180000008034 3291 14900000.0 .116000000 15000000.0 .114000000 15100000.0 .1120000008034 3291 15200000.0 .110000000 15300000.0 .108000000 15400000.0 .1060000008034 3291 15500000.0 .104600000 15600000.0 .103800000 15700000.0 .1029500008034 3291 15800000.0 .101800000 15900000.0 .100600000 16000000.0 .0996000008034 3291 16100000.0 .098800000 16200000.0 .098050000 16300000.0 .0976000008034 3291 16400000.0 .097200000 16500000.0 .096400000 16600000.0 .0952000008034 3291 16700000.0 .093950000 16800000.0 .092400000 16900000.0 .0908000008034 3291 17000000.0 .089800000 17100000.0 .089400000 17200000.0 .0890000008034 3291 17300000.0 .088600000 17400000.0 .088200000 17500000.0 .0878000008034 3291 17600000.0 .087400000 17700000.0 .087000000 17800000.0 .0866000008034 3291 17900000.0 .086200000 18000000.0 .085800000 18100000.0 .0854000008034 3291 18200000.0 .085000000 18300000.0 .084600000 18400000.0 .0842000008034 3291 18500000.0 .083800000 18600000.0 .083400000 18700000.0 .0830000008034 3291 18800000.0 .082600000 18900000.0 .082200000 19000000.0 .0818000008034 3291 19100000.0 .081400000 19200000.0 .081050000 19300000.0 .0810000008034 3291 19400000.0 .081000000 19500000.0 .080800000 19600000.0 .0804000008034 3291 19700000.0 .080050000 19800000.0 .080000000 19900000.0 .0800000008034 3291 20000000.0 0.0 8034 3291 8034 3 0 8034 0 0 8.01990E+4 1.97259E+2 0 0 0 1803433291 0.0000E+00 0.0000E+00 0 291 0 2803433291 0.0000E+00 0.0000E+00 0 1 62 31803433291 1.000000-5 0.000000+0 5.250000+5 2.464000-1 7.500000+5 8.640000-2803433291 1.000000+6 1.600000-2 2.000000+6 1.600000-2 2.250000+6 2.200000-2803433291 3.000000+6 2.200000-2 3.500000+6 1.600000-2 4.000000+6 1.600000-2803433291 5.000000+6 1.600000-2 6.000000+6 1.600000-2 7.000000+6 1.600000-2803433291 7.250000+6 8.640000-2 8.000000+6 8.640000-2 9.000000+6 8.640000-2803433291 1.000000+7 8.640000-2 1.100000+7 8.640000-2 1.200000+7 8.640000-2803433291 1.300000+7 8.640000-2 1.325000+7 3.640000-2 1.375000+7 6.400000-3803433291 1.400000+7 6.400000-3 1.450000+7 3.640000-2 1.500000+7 3.640000-2803433291 1.525000+7 8.640000-2 1.600000+7 8.640000-2 1.650000+7 2.464000-1803433291 1.700000+7 2.464000-1 1.800000+7 2.464000-1 1.900000+7 2.464000-1803433291 2.000000+7 0.000000+0 803433291 0.000000+0 0.000000+0 0 1 6 3803433291 1.000000-5 0.000000+0 5.250000+5 3.600000-3 2.000000+7 0.000000+0803433291 803433 0 8034 0 0 0 0 0 8.22040E+4 2.02220E+2 0 0 34 108225 1451 0.0 0.0 0 0 0 68225 1451 1.00000E+0 2.00000E+7 0 0 10 20028225 1451 3.00000E+2 0.0 1 0 147 48225 1451 82-Pb-204 FEI EVAL-Oct01 K.I.Zolotarev 8225 1451 DIST-Feb2004 8225 1451 ----IRDF-2002 MATERIAL 8225 8225 1451 -----INCIDENT NEUTRON DATA 8225 1451 ------ENDF-6 FORMAT 8225 1451 ***************************************************************** 8225 1451 82-Pb-204 FEI EVAL-Oct01 K.I.Zolotarev 8225 1451 DIST-Oct01 20011014 8225 1451 ----BROND-2 MATERIAL 8225 8225 1451 ------Russian Reactor Dosimetry File RRDF-2002 8225 1451 ***************************************************************** 8225 1451 ----- MF=3 MT=71 ----- 8225 1451 For the IRDF-2002 file this reaction was converted at IAEA/NDS 8225 1451 The reaction MF/MT=3/71 was converted to MF/MT=10/ 4 8225 1451 The corresponding co-variance files were also converted 8225 1451 The reaction MF/MT=33/71 was converted to MF/MT=40/ 4 8225 1451 ***************************************************************** 8225 1451 Author of evaluation: K.I.Zolotarev 8225 1451 ***************************************************************** 8225 1451 8225 1451 ----- MF=3 MT=71 ----- 8225 1451 8225 1451 In the section MT= 71 are given evaluated cross section data 8225 1451 of neutron inelastic excitation of the 67.2-min isomeric state in 8225 1451 in the lead-204 in the energy range from threshold to 20 MeV. 8225 1451 The isomeric state is the 21-excited level in Pb-204. This level 8225 1451 has energy 2185.79 keV with spin and parity 9- [1]. 8225 1451 Data base for the evaluation Pb-204(n,n')Pb-204m excitation 8225 1451 function was formed from microscopic experimental data [2-9] and 8225 1451 data received from theoretical model calculations. Experimental 8225 1451 data [3-9] were corrected to the new standards. Data of Decowski 8225 1451 et al. [6] were renormalized also to the result of precise measu- 8225 1451 rements of Ryves et al. at 14.3 MeV [8]. Uncertainty for cross 8225 1451 section value measured by Bornemisza-Pauspertl et al. at 2.80 MeV 8225 1451 [2] was evaluated as +-35 %. Theoretical model calculations was 8225 1451 done by means of GNASH code [10]. New modified version of GNASH 8225 1451 was used for calculations [11]. 8225 1451 The evaluation Pb-204(n,n')Pb-204m excitation function from 8225 1451 threshold to 20 MeV has been carried out within the framework of 8225 1451 generalized least squares method , rational function was used as 8225 1451 model function [12]. Procedure of calculation recommended cross 8225 1451 section data was performed by means of PADE-2 code [13]. 8225 1451 Integral experimental data for U-235 neutron fission spectrum 8225 1451 [14,15] and Cf-252 spontaneous fission neutron spectrum [16,17] 8225 1451 was used for testing evaluated Pb-204(n,n')Pb-204m excitation fun-8225 1451 ction. Data for U-235 thermal fission neutron spectrum and Cf-252 8225 1451 spontaneous fission neutron spectrum were taken from ref.[18] and 8225 1451 [19], respectively. The results of testing are given in the 8225 1451 table 1. 8225 1451 Table 1. 8225 1451 --------------------------------------------------------------- 8225 1451 TYPE OF SPECTRUM I ,MB (calc.) I , MB (measured) 8225 1451 ----------------------I-----------------I---------------------- 8225 1451 I I 8225 1451 U-235 neutron fission I 17.770 I 18.900 +- 2.000 [14] 8225 1451 I I 19.010 +- 1.536 [15] 8225 1451 I I 8225 1451 ----------------------I-----------------I---------------------- 8225 1451 I I 8225 1451 CF-252 spont. fission I 20.373 I 20.900 +- 1.202 [16] 8225 1451 I I 20.850 +- 0.920 [17] 8225 1451 --------------------------------------------------------------- 8225 1451 8225 1451 ----- MF=33 MT=71 ----- 8225 1451 Uncertainties in the evaluated excitation function for the 8225 1451 reaction Pb-204(n,n')Pb-204m are given in the form of relative 8225 1451 covariance matrix for the 32-neutron energy groups (LB=5). Cova- 8225 1451 riance matrix of uncertainties was calculated simultaneously with 8225 1451 recommended cross section data by means of PADE-2 code. 8225 1451 Eigenvalues of the 6-th digits relative covariance matrix 8225 1451 given in the 33-file are the following: 8225 1451 2.56972E-09 3.22598E-09 3.49414E-09 4.00985E-09 8225 1451 4.91484E-09 5.78671E-09 6.70005E-09 8.71427E-09 8225 1451 1.03160E-08 1.33154E-08 1.55381E-08 2.07583E-08 8225 1451 2.40281E-08 3.05913E-08 4.05801E-08 4.76532E-08 8225 1451 5.93691E-08 9.33863E-08 1.58183E-07 2.72581E-06 8225 1451 1.76176E-04 7.87730E-04 1.22783E-03 2.11803E-03 8225 1451 3.22408E-03 4.16999E-03 4.89344E-03 1.08231E-02 8225 1451 2.01870E-02 3.34051E-02 7.41942E-02 8225 1451 8225 1451 References : 8225 1451 1. R.B.Firestone Table of Isotopes, Eighth edition, Vol. 1, 8225 1451 John Wiley & Sons, Inc., New York, 1995 8225 1451 2. P.Bornemisza-Pauspertl, J.Karolyi, G.Peto Atomki Kozl., v.10, 8225 1451 no.2, p.112, July 1968 8225 1451 3. J.Csikai, G.Peto Acta Phys. Hung., v.23, p.87, May 1967 8225 1451 4. A.K.Hankla, R.W.Fink, J.H.Hamilton Nucl. Phys., v.A180, 8225 1451 p.157, January 1972 8225 1451 5. G.N.Maslov, F.Nasyrov, N.F.Pashkin Yadernye Konstanty, v.9, 8225 1451 p.50, Obninsk, 1972 8225 1451 6. P.Decowski et al. Nucl. Phys., v.A204, p.121, April 1973 8225 1451 7. D.L.Smith, J.W.Meadows Report ANL-NDM-37, December 1977 8225 1451 8. T.B.Ryves, P.Kolkowski, A.C.Hooley Annals Nucl. Energy, 8225 1451 v.17, p.107, 1990 8225 1451 9. I.Kimura, K.Kobayashi Nucl. Sci. Eng., v.106, p.332, 1990 8225 1451 10. P.G.Young, E.D.Arthur A Preequilibrium Statistical Nuclear 8225 1451 Model Code for Calculation of Cross Section and Emission 8225 1451 Spectra. Report LA-6947, Los Alamos, 1977 8225 1451 11. E.L.Trykov, G.Ya.Tertychnyi Private communication, IPPE, 8225 1451 Obninsk, May 1999 8225 1451 12. S.Badikov,N.Rabotnov,K.Zolotarev Proc. of NEANSC Specialist's 8225 1451 Meeting on Evaluation and Processing of Covariance Data, Oak 8225 1451 Ridge , USA, 7-9 September 1992, OECD, Paris, 1993, p.105 8225 1451 13. S.A.Badikov et al. Preprint FEI-1686, Obninsk, 1985 8225 1451 14. I.Kimura, K.Kobayashi, T.Shibata Nucl. Sci. Technol., v.8, 8225 1451 p.59, February 1971 8225 1451 15. A.K.Brodskaja et al. Jadernyje Konstanty, no.23, v.4, 8225 1451 October 1976 8225 1451 16. J.Csikai, Z.Dezso Proc. Educational Seminar on the use of 8225 1451 Cf-252, Karlsruhe, 14-18 April 1975, p.29 8225 1451 17. K.Kobayashi et al. Report KURRI-AR-17, p.15, 1984 ; 8225 1451 K.Kobayashi et al. Progress Report NEANDC(J)-106/U, p.41, 8225 1451 September 1984 8225 1451 18. L.W.Weston et al. Evaluated Neutron Data File for U-235, 8225 1451 ENDF/B-VI Library, MAT=9228, MF=5, MT=18, eval. April 1989 8225 1451 19. W.Mannhart Report IAEA-TECDOC-410, IAEA, Vienna, 1987, p158 8225 1451 ***************************************************************** 8225 1451 File 2 added to the pointwise file containing only the effective 8225 1451 scattering radius with no resonance parameters given. 8225 1451 Taken from JENDL-3.2 8225 1451 8225 1451 ***************************************************************** 8225 1451 ***************** Program LINEAR (VERSION 2002-1) ***************8225 1451 For All Data Greater than 1.0000E-10 barns in Absolute Value 8225 1451 Data Linearized to Within an Accuracy of .100000000 per-cent 8225 1451 ***************** Program SIGMA1 (VERSION 2002-1) ***************8225 1451 Data Doppler Broadened to 300.000000 Kelvin 8225 1451 for All Data Greater than 1.0000E-10 barns in Absolute Value 8225 1451 Data Linearized to Within an Accuracy pf .100000000 per-cent 8225 1451 ***************** Program FIXUP (Version 2002-1) ****************8225 1451 Corrected ZA/AWR in All Sections-----------------------------Yes 8225 1451 Corrected Thresholds-----------------------------------------Yes 8225 1451 Extended Cross Sections to 20 MeV----------------------------No 8225 1451 Allow Cross Section Deletion---------------------------------No 8225 1451 Allow Cross Section Reconstruction---------------------------No 8225 1451 Make All Cross Sections Non-Negative-------------------------Yes 8225 1451 Delete Energies Not in Ascending Order-----------------------Yes 8225 1451 Deleted Duplicate Points-------------------------------------Yes 8225 1451 Check for Ascending MAT/MF/MT Order--------------------------Yes 8225 1451 Check for Legal MF/MT Numbers--------------------------------Yes 8225 1451 Allow Creation of Missing Sections---------------------------No 8225 1451 Allow Insertion of Energy Points-----------------------------No 8225 1451 Create Uniform Energy Grid-----------------------------------No 8225 1451 Delete Section if Cross Section =0 at All Energies-----------Yes 8225 1451 ***************** Program GROUPIE (VERSION 2002-1) **************8225 1451 Unshielded Group Averages Using 640 Groups 8225 1451 Weighting Spectrum: Flat (Constant) Spectrum 8225 1451 1 451 155 18225 1451 8 4 2 18225 1451 10 4 63 18225 1451 40 4 98 18225 1451 8225 1 0 8225 0 0 8.22040E+4 2.02220E+2 0 0 1 08225 3291 0.0 -2.18579E+6 82204 0 1 1808225 3291 180 1 8225 3291 2100000.00 1.34936E-8 2200000.00 3.51923E-5 2300000.00 .0001891968225 3291 2400000.00 .001283871 2500000.00 .004581835 2600000.00 .0080065758225 3291 2700000.00 .011515626 2800000.00 .015279183 2900000.00 .0192366028225 3291 3000000.00 .023323413 3100000.00 .027458638 3200000.00 .0315938638225 3291 3300000.00 .035729088 3400000.00 .039819900 3500000.00 .0438663008225 3291 3600000.00 .047857600 3700000.00 .051793800 3800000.00 .0556669508225 3291 3900000.00 .059477050 4000000.00 .063287150 4100000.00 .0670321088225 3291 4200000.00 .070711925 4300000.00 .074391742 4400000.00 .0780715588225 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5.494970-4 3.256820-4822533291 2.785150-4 3.407000-4 4.403740-4 5.251840-4 5.676180-4 5.611320-4822533291 5.135370-4 4.034660-4 2.809920-4 3.093840-4 5.829890-4 1.137240-3822533291 2.000580-3 1.826280-3 1.520870-3 1.195250-3 9.411980-4 7.950340-4822533291 7.457830-4 7.605310-4 8.046270-4 8.515680-4 8.855910-4 9.006540-4822533291 8.918920-4 8.697480-4 9.247820-4 1.198590-3 1.930540-3 2.069600-3822533291 2.170730-3 2.127140-3 1.982240-3 1.792810-3 1.603980-3 1.442320-3822533291 1.319490-3 1.238160-3 1.196730-3 1.192270-3 1.252010-3 1.444090-3822533291 1.791820-3 2.368020-3 3.354430-3 2.757910-3 3.136490-3 3.244060-3822533291 3.107020-3 2.801640-3 2.414680-3 2.021720-3 1.679890-3 1.427830-3822533291 1.288530-3 1.324970-3 1.778140-3 2.659800-3 3.857740-3 5.178410-3822533291 3.951380-3 4.381340-3 4.382000-3 4.025400-3 3.439110-3 2.759100-3822533291 2.104450-3 1.567490-3 1.212250-3 1.125220-3 1.782580-3 3.259190-3822533291 5.214970-3 6.909040-3 5.098540-3 5.270890-3 4.939470-3 4.244030-3822533291 3.358470-3 2.449770-3 1.657290-3 1.085100-3 8.215140-4 1.550350-3822533291 3.460510-3 6.039040-3 8.070540-3 5.590640-3 5.343050-3 4.654550-3822533291 3.706320-3 2.684930-3 1.754210-3 1.042510-3 6.194140-4 1.274620-3822533291 3.314340-3 6.172660-3 8.424730-3 5.207660-3 4.630850-3 3.773000-3822533291 2.808770-3 1.897690-3 1.169280-3 6.629810-4 1.141720-3 2.986270-3822533291 5.708020-3 8.000890-3 4.237350-3 3.594190-3 2.838430-3 2.099130-3822533291 1.484160-3 1.003900-3 1.272530-3 2.672740-3 4.889680-3 7.008540-3822533291 3.241590-3 2.798940-3 2.346960-3 1.954760-3 1.616460-3 1.713150-3822533291 2.545590-3 4.012270-3 5.735180-3 2.716800-3 2.616640-3 2.520100-3822533291 2.426860-3 2.449610-3 2.727590-3 3.350830-3 4.472360-3 2.879730-3822533291 3.109940-3 3.340090-3 3.428280-3 3.287320-3 3.123280-3 3.473200-3822533291 3.658080-3 4.259390-3 4.573780-3 4.243340-3 3.475180-3 2.932480-3822533291 5.424620-3 6.383490-3 6.350240-3 5.263230-3 3.318230-3 8.625060-3822533291 1.000150-2 9.656750-3 6.041350-3 1.365320-2 1.538780-2 1.111050-2822533291 1.989260-2 1.732600-2 2.174600-2 822533291 822533 0 8225 0 0 0 0 0 9.02320E+4 2.30044E+2 0 1 34 109040 1451 0.0 1.00000E+0 0 0 0 69040 1451 1.00000E+0 2.00000E+7 0 0 10 20029040 1451 3.00000E+2 0.0 1 0 208 59040 1451 90-Th-232 BNL,ANL+ EVAL-DEC77 BHAT,SMITH,LEONARD,DESAUSSURE+ 9040 1451 DIST-Feb2004 9040 1451 ----IRDF-2002 MATERIAL 9040 9040 1451 -----INCIDENT NEUTRON DATA 9040 1451 ------ENDF-6 FORMAT 9040 1451 ******************************************************************9040 1451 90-TH-232 BNL,ANL+ EVAL-DEC77 BHAT,SMITH,LEONARD,DESAUSSURE+ 9040 1451 DIST-FEB90 19900205 9040 1451 ----ENDF/B-VI MATERIAL 9040 9040 1451 *****************EXTRACT FOR SPECIAL PURPOSE FILE*****************9040 1451 DOSIMETRY 9040 1451 ******************************************************************9040 1451 ENDF/B-V MATERIAL CONVERTED TO ENDF-6 FORMAT BY NNDC 9040 1451 * * * * * * *9040 1451 9040 1451 PRINCIPAL EVALUATORS- J.W.MEADOWS,W.P.POENITZ,A.B.SMITH,D.L.SMITH9040 1451 AND J.F.WHALEN(ANL) 9040 1451 R.J.HOWERTON(LLL) 9040 1451 B.R.LEONARD ET.AL.(BNW) 9040 1451 G.DE SAUSSURE,R.L.MACKLIN AND R.GWIN(ORNL) 9040 1451 D.K. OLSEN (ORNL) 9040 1451 THIS EVALUATION ASSEMBLED BY M.R.BHAT(BNL) USING THE FOLLOWING 9040 1451 EVALUATIONS 9040 1451 9040 1451 NU-BAR(PROMPT) - R.GWIN(ORNL)(REF 4) 9040 1451 DELAYED NEUTRONS -M.BRADY AND T.ENGLAND (REF 5) 9040 1451 ENERGY RELEASE IN FISSION - R.SHER ET.AL.(STANFORD)(REF 6) 9040 1451 THERMAL RANGE - B.R.LEONARD JR.ET.AL.(REF 2) WITH CHANGES BY9040 1451 M.R.BHAT(REF 31) 9040 1451 RESOLVED RESONANCE REGION - D.K. OLSEN (REF 37) 9040 1451 9040 1451 UNRESOLVED RESONANCE REGION - G.DE SAUSSURE AND R.L.MACKLIN 9040 1451 (REF 3)WITH CHANGES BY M.R.BHAT(REF 31) 9040 1451 SMOOTH CROSS-SECTIONS 50KEV TO 20MEV - 9040 1451 TOTAL - A.B.SMITH(ANL)(REF1) 9040 1451 FISSION - W.P.POENITZ(ANL)(REF1) 9040 1451 CAPTURE - W.P.POENITZ(ANL)(REF1) 9040 1451 INELASTIC AND OTHER CROSS-SECTIONS - J.W.MEADOWS ET.AL.(ANL)9040 1451 (REF1)9040 1451 FISSION NEUTRON SPECTRUM - M.R.BHAT(BNL) (REF 31) 9040 1451 GAMMA-RAY PRODUCTION - R.J.HOWERTON(LLL) 9040 1451 9040 1451 MF = 1 9040 1451 9040 1451 MT=452(NU-BAR TOTAL) CONSISTENT WITH MT=455 AND MT=456 9040 1451 MT=455(DELAYED NEUTRON YIELDS) (REF 5) 9040 1451 MT=456(PROMPT NU-BAR) BASED ON CF-252 NU-BAR(PROMPT)=3.757 9040 1451 EVALUATION BY GWIN(ORNL) BASED ON A STRAIGHT LINE LEAST- 9040 1451 SQUARES FIT TO THE AVAILABLE DATA LISTED IN REF.4. 9040 1451 MT=458(ENERGY RELEASE IN FISSION)EVALUATION BY R.SHER ET.AL(STAN-9040 1451 FORD)(REF 6) 9040 1451 9040 1451 MF = 2 9040 1451 9040 1451 RESOLVED RESONANCE REGION - D.K. OLSEN (REF 37) RESONANCE 9040 1451 PARAMETERS ARE UNCERTAINTY WEIGHTED AVERAGE OF AVAILABLE 9040 1451 DATA AND THE TWO BOUND RESONANCES OF REF 2 WITH MF=3 FILES 9040 1451 FOR TRUNCATION, MISSED P-WAVES, 1/V CAPTURE, AND SMOOTH 9040 1451 CONNECTION TO THE END OF THE THERMAL REGION CROSS SECTIONS 9040 1451 AT 5 EV. AVERAGE CAPTURE WIDTH = 2.44 X 10-2 EV, S-WAVE 9040 1451 STRENGTH FUNCTION = 0.826 X 10-4, P-WAVE STRENGTH FUNCTION 9040 1451 ASSUMED = 1.6 X 10-4, 5 TO 4000 EV INFINITLY DILUTE CAPTURE 9040 1451 RESONANCE INTEGRAL = 81.58 B. 9040 1451 9040 1451 UNRESOLVED RESONANCE REGION - G.DE SAUSSURE AND R.L.MACKLIN(REF3)9040 1451 EVALUATION OF CAPTURE CROSS-SECTION USED FROM 4 TO 25KEV. 9040 1451 ABOVE 25KEV THIS WAS JOINED SMOOTHLY TO THE POENITZ EVALUAT-9040 1451 ION FROM 40KEV TO 50KEV. THESE WERE THEN FITTED USING THE 9040 1451 CODE UR(REF30) TO GIVE THE UNRESOLVED RESONANCE 9040 1451 PARAMETERS. IN THIS FIT THE S- AND D-WAVE PARAMETERS WERE 9040 1451 KEPT CONSTANT AT THE VALUES GIVEN BY DE SAUSSURE AND MACKLIN9040 1451 THE P-WAVE GAMMA WIDTH WAS INCREASED TO 2.52E-02EV AND THE 9040 1451 P-WAVE REDUCED NEUTRON WIDTH VARIED TO FIT THE CAPTURE 9040 1451 CROSS-SECTION. 9040 1451 THE CAPTURE CROSS-SECTION INPUT USED TO EXTRACT THE 9040 1451 UNRESOLVED RESONANCE PARAMETERS IN FILE 2/151 ARE GIVEN HERE9040 1451 FROM 4KEV TO 50KEV AT 1KEV INTERVAL. THESE ARE - 9040 1451 1.130,0.982,0.908,0.850,0.803,0.764,0.730,0.702,0.677,0.655,9040 1451 0.636,0.618,0.601,0.585,0.571,0.557,0.545,0.533,0.521,0.511,9040 1451 0.500,0.491,0.482,0.473,0.465,0.459,0.452,0.448,0.443,0.439,9040 1451 0.435,0.431,0.429,0.426,0.424,0.422,0.420,0.420,0.420,0.420,9040 1451 0.417,0.414,0.411,0.407,0.401,0.395,0.390. 9040 1451 9040 1451 MF = 3 9040 1451 9040 1451 THERMAL REGION(1.0E-05 TO 5.0EV) - MAINLY BASED ON LEONARD(REF 2)9040 1451 EVALUATION WITH THE FOLLOWING CHANGES. THE CAPTURE CROSS- 9040 1451 SECTION IN REF 2 NORMALISED TO 7.40B AT 2.53E-02 EV AND USED9040 1451 FROM 1.0E-05EV TO 2.53E-02EV. FROM 2.6E-02EV TO 14EV THE FIT9040 1451 TO THEIR CAPTURE DATA BY CHRIEN AND LIOU(REF 32) USED. THE 9040 1451 SCATTERING CROSS-SECTION WAS CHANGED TO ACCOMMODATE CHANGES 9040 1451 IN CAPTURE USING TOTAL SLIGHTLY MODIFIED FROM THAT GIVEN IN 9040 1451 REF 2. THESE SMALL CHANGES ARE DESCRIBED IN REF 31. 9040 1451 2.53E-02EV CROSS-SECTION CAPTURE 7.4 B 9040 1451 DILUTE CAPTURE RESONANCE INTEGRAL 86.1 B 9040 1451 (0.5 TO 2.0E+07 EV) 9040 1451 TOTAL CROSS-SECTION FROM 50KEV TO 20MEV - A.B.SMITH(ANL)(REF1) 9040 1451 MAINLY BASED ON REF9-13 FROM 0.025 TO 1 MEV,FROM 1 TO 5MEV 9040 1451 BASED ON REF 9-11,14-16.FROM 5 TO 15MEV EVALUATION BASED ON 9040 1451 REF11,15,17.FROM 15 TO 20MEV EVALUATION BASED ON NUCLEAR 9040 1451 MODEL CALCULATIONS DESCRIBED IN REF18 AND 19.TOTAL CROSS- 9040 1451 SECTION MODIFIED BY ABOUT 3PER-CENT BETWEEN 50 TO 150KEV 9040 1451 JOIN SMOOTHLY WITH THAT CALCULATED FROM THE UNRESOLVED 9040 1451 RESONANCE PARAMETERS BELOW 50KEV. 9040 1451 ELASTIC SCATTERING CROSS-SECTION FROM 50KEV TO 20MEV OBTAINED AS 9040 1451 THE DIFFERENCE BETWEEN TOTAL AND OTHER PARTIAL CROSSSECTIONS9040 1451 FISSION CROSS-SECTION UP TO 20MEV - W.P.POENITZ(REF1) DETAILS OF 9040 1451 THE EVALUATION AND THE DATA SETS USED ARE IN REF1. 9040 1451 FISSION CROSS-SECTION CHANGED TO BE CONSISTENT WITH 9040 1451 U-235(N,F) EVALUATION GIVEN IN REF 33 9040 1451 CAPTURE CROSS-SECTION 50KEV -20MEV - W.P.POENITZ. DETAILS OF THE 9040 1451 EVALUATION AND THE DATA SETS USED ARE GIVEN IN REF1. 9040 1451 CAPTURE CROSS-SECTION CHANGED TO BE CONSISTENT WITH 9040 1451 U-235(N,F) EVALUATION GIVEN IN REF 33 9040 1451 INELASTIC SCATTERING CROSS-SECTIONS - J.W.MEADOWS ET.AL.(REF 1) 9040 1451 EVALUATION BASED MAINLY ON THE DATA IN REF 20-25. INTER- 9040 1451 POLATED BETWEEN EXPERIMENTAL DATA AND EXTRAPOLATED TO 20MEV 9040 1451 USING COUPLED CHANNEL NUCLEAR MODEL CALCULATIONS OF REF26. 9040 1451 (N,2N) AND (N,3N) CROSS-SECTIONS - J.W.MEADOWS ET.AL.(REF1) 9040 1451 EXPERIMENTAL DATA USED(REF1) RENORMALISED TO A COMMON SET OF9040 1451 STANDARDS BEFORE EVALUATION. 9040 1451 9040 1451 MF = 33 9040 1451 MT = 18,102 BASED ON ERROR ESTIMATES IN THESE CROSS SECTIONS IN 9040 1451 REF. 32,1 9040 1451 9040 1451 PRINCIPAL REFERENCES 9040 1451 9040 1451 1 J.W.MEADOWS,W.P.POENITZ,A.B.SMITH,D.L.SMITH,J.F.WHALEN AND 9040 1451 R.J.HOWERTON ANL/NDM-35(FEB.1978) 9040 1451 2 D.F.NEWMAN,B.R.LEONARD JR.ET.AL. EPRI NP-222,MAY1977 9040 1451 3 G.DE SAUSSURE AND R.L.MACKLIN ORNL/TM-6161(ENDF-255)DEC.1977 9040 1451 4 R.GWIN ORNL/TM-6245(ENDF-262)MAY1978. 9040 1451 5 M.C.BRADY AND T.R.ENGLAND NSE 103,129(1989) 9040 1451 **DATA INSERTED INTO FILE AT BNL** 9040 1451 6. R.SHER + C.BECK EPRI 1771/81 + REV. 1/83 + PC TO MAGURNO 2/83 9040 1451 7 S.F.MUGHABGHAB AND D.I.GARBER,BNL-325(3RD EDN)1973 9040 1451 8 H.DERRIEN,NEANDC(E)163U(1975) 9040 1451 9 J.F.WHALEN AND A.B.SMITH,NUC.SCI.AND ENG.67,129(1978) 9040 1451 10J.MEADOWS,A.SMITH AND J.WHALEN,PRIV.COMM.1977 9040 1451 11U.FASOLI ET.AL.NUCL.PHYS.A151,369(1970) 9040 1451 12M.DIVADEENAM ET.AL. DISS.ABS.28,3834(1968) 9040 1451 13C.UTTLEY ET.AL.66PARIS,1,165,1966 9040 1451 14L.GREEN ET.AL.PROC.CONF.ON NUCL.CROSS-SECTIONS AND TECHNOLOGY, 9040 1451 KNOXVILLE,1,325(1971) 9040 1451 15D.FOSTER ET.AL.PRIV.COMM(1967) 9040 1451 16R.BATCHELOR ET.AL.NUCL.PHYS.65,236(1956) 9040 1451 17J.COON ET.AL.PHYS.REV.88,562(1952) 9040 1451 18P.GUENTHER ET.AL.ACCEPTED FOR PUBLICATION NUC.SCI.AND ENG.(19779040 1451 19C.PHILIS ET.AL. ANL/NDM-28(1977) 9040 1451 20A.SMITH PHYS.REV.126,718(1962) 9040 1451 21W.MCMURRAY ET.AL.SOUTHERN UNIV.NUCLEAR INST.REPORT,SUNI-41(19759040 1451 22J.HAOUAT ETAL.PROC.INER.CONF.ON INTERACTIONS OF NEUTRONS WITH 9040 1451 NUCLEI,CONF-760715(1976) 9040 1451 23R.BATCHELOR AND J.TOWLE,NUCL.PHYS.65,236(1965) 9040 1451 24R.BATCHELOR AND J.TOWLE,PROC.PHYS.SOC.73,193(1959) 9040 1451 25A.SMITH,PRIV.COMM,1970 AND 1977 9040 1451 26A.SMITH ET.AL.ACCEPTED FOR PUBLICATION NUC.SCI.AND ENG.(1977) 9040 1451 27M.SEGEV.ET.AL.TRANS.AM.NUCL.SOC,22,679(1975)AND PRIV.COMM(1976)9040 1451 28R.J.HOWERTON AND R.J.DOYAS,NUCL.SCI AND ENG,46,414(1971) 9040 1451 29YU.A.VASIL'EV ET.AL.PHYSICS OF NUCLEAR FISSION,N.A.PERFILOV AND9040 1451 V.P.EISMONT(EDS),ISRAEL PROGRAM OF SCI.TRANS.JERUSALEM(1964) 9040 1451 30E.M.PENNINGTON,PRIVATE COMMUNICATION(1973) 9040 1451 31M.R.BHAT,ENDF-268(TO BE PUBLISHED) 9040 1451 32R.E.CHRIEN AND H.LIOU(1978) TO BE PUBLISHED 9040 1451 33W.P.POENITZ,ANL/NDM-45(1978) TO BE PUBLISHED 9040 1451 34R.W.PEELE AND F.C.MAIENSCHEIN, NUCL.SCI.ENG. 40, 485 (1970) 9040 1451 35R.J. HOWERTON, D.E. CULLEN, R.C. HAIGHT, M.H. MACGREGOR, S.T. 9040 1451 PERKINS, AND E.F. PLECHARTY, "THE LLL EVALUATED NUCLEAR DATA 9040 1451 LIBRARY (ENDL): EVALUATION TECHNIQUES, REACTION INDEX, AND 9040 1451 DESCRIPTIONS OF INDIVIDUAL EVALUATIONS," UCRL-50400, VOL. 15, 9040 1451 PART A, LAWERENCE LIVERMORE LABORATORY (1975). 9040 1451 36S.T.PERKINS,R.C.HAIGHT AND R.J.HOWERTON,NUCL.SCI.ENG.57,1(1975)9040 1451 37D.K.OLSEN, ORNL/TM-8056(1982), ENDF-319. 9040 1451 ******************************************************************9040 1451 9040 1451 9040 1451 9040 1451 9040 1451 ******************************************************************9040 1451 ***************** Program LINEAR (VERSION 2002-1) ***************9040 1451 For All Data Greater than 1.0000E-10 barns in Absolute Value 9040 1451 Data Linearized to Within an Accuracy of .100000000 per-cent 9040 1451 ***************** Program RECENT (VERSION 2002-1) ***************9040 1451 for All Data Greater than 1.0000E-10 barns in Absolute Value 9040 1451 Data Linearized to within an Accuracy of .100000000 per-cent 9040 1451 ***************** Program SIGMA1 (VERSION 2002-1) ***************9040 1451 Data Doppler Broadened to 300.000000 Kelvin 9040 1451 for All Data Greater than 1.0000E-10 barns in Absolute Value 9040 1451 Data Linearized to Within an Accuracy pf .100000000 per-cent 9040 1451 ***************** Program FIXUP (Version 2002-1) ****************9040 1451 Corrected ZA/AWR in All Sections-----------------------------Yes 9040 1451 Corrected Thresholds-----------------------------------------Yes 9040 1451 Extended Cross Sections to 20 MeV----------------------------No 9040 1451 Allow Cross Section Deletion---------------------------------No 9040 1451 Allow Cross Section Reconstruction---------------------------No 9040 1451 Make All Cross Sections Non-Negative-------------------------Yes 9040 1451 Delete Energies Not in Ascending Order-----------------------Yes 9040 1451 Deleted Duplicate Points-------------------------------------Yes 9040 1451 Check for Ascending MAT/MF/MT Order--------------------------Yes 9040 1451 Check for Legal MF/MT Numbers--------------------------------Yes 9040 1451 Allow Creation of Missing Sections---------------------------No 9040 1451 Allow Insertion of Energy Points-----------------------------No 9040 1451 Create Uniform Energy Grid-----------------------------------No 9040 1451 Delete Section if Cross Section =0 at All Energies-----------Yes 9040 1451 ***************** Program GROUPIE (VERSION 2002-1) **************9040 1451 Unshielded Group Averages Using 640 Groups 9040 1451 Weighting Spectrum: Flat (Constant) Spectrum 9040 1451 1 451 217 19040 1451 3 18 72 09040 1451 3 102 217 09040 1451 33 18 9 09040 1451 33 102 7 09040 1451 9040 1 0 9040 0 0 9.02320E+4 2.30044E+2 0 0 0 09040 3 18 1.88470E+8 1.88470E+8 0 0 1 2069040 3 18 206 1 9040 3 18 500000.000 1.37500E-6 525000.000 4.12499E-6 550000.000 6.87498E-69040 3 18 575000.000 9.62500E-6 600000.000 1.44501E-5 630000.000 2.13500E-59040 3 18 660000.000 2.82500E-5 690000.000 4.12173E-5 720000.000 8.03773E-59040 3 18 760000.000 .000130050 800000.000 .000154600 840000.000 .0002595459040 3 18 880000.000 .000567766 920000.000 .001115025 960000.000 .0011350869040 3 18 1000000.00 .001805998 1100000.00 .003257634 1200000.00 .0100931179040 3 18 1300000.00 .038089744 1400000.00 .064670533 1500000.00 .0994199279040 3 18 1600000.00 .097810452 1700000.00 .084170147 1800000.00 .0952105919040 3 18 1900000.00 .116499192 2000000.00 .118500645 2100000.00 .1289995989040 3 18 2200000.00 .130499193 2300000.00 .117750737 2400000.00 .1157501309040 3 18 2500000.00 .118800036 2600000.00 .122399807 2700000.00 .1254663819040 3 18 2800000.00 .125333781 2900000.00 .129000649 3000000.00 .1374367209040 3 18 3100000.00 .138499460 3200000.00 .138262606 3300000.00 .1396000289040 3 18 3400000.00 .141199830 3500000.00 .141599831 3600000.00 .1408000329040 3 18 3700000.00 .140250187 3800000.00 .141200072 3900000.00 .1423999139040 3 18 4000000.00 .142899857 4100000.00 .142699906 4200000.00 .1424999559040 3 18 4300000.00 .142300003 4400000.00 .142100052 4500000.00 .1422143549040 3 18 4600000.00 .142642911 4700000.00 .143071468 4800000.00 .1435000259040 3 18 4900000.00 .143928582 5000000.00 .144517803 5100000.00 .1454997569040 3 18 5200000.00 .145249710 5300000.00 .143749811 5400000.00 .1417500209040 3 18 5500000.00 .139250337 5600000.00 .138500176 5700000.00 .1390002899040 3 18 5800000.00 .141500578 5900000.00 .148001097 6000000.00 .1620014499040 3 18 6100000.00 .182000830 6200000.00 .203750523 6300000.00 .2272505289040 3 18 6400000.00 .252499823 6500000.00 .276599003 6600000.00 .2977987789040 3 18 6700000.00 .318998553 6800000.00 .340198328 6900000.00 .3613981039040 3 18 7000000.00 .376248272 7100000.00 .384748835 7200000.00 .3904989219040 3 18 7300000.00 .393498528 7400000.00 .391373544 7500000.00 .3841239679040 3 18 7600000.00 .376874391 7700000.00 .369631916 7800000.00 .3644204229040 3 18 7900000.00 .360806729 8000000.00 .356899918 8100000.00 .3526999919040 3 18 8200000.00 .348500063 8300000.00 .344300135 8400000.00 .3401002089040 3 18 8500000.00 .336600256 8600000.00 .333800279 8700000.00 .3310003029040 3 18 8800000.00 .328200325 8900000.00 .325400348 9000000.00 .3236003459040 3 18 9100000.00 .322800318 9200000.00 .322000290 9300000.00 .3212002629040 3 18 9400000.00 .320400235 9500000.00 .320200157 9600000.00 .3206000309040 3 18 9700000.00 .320999903 9800000.00 .321399775 9900000.00 .3217996489040 3 18 10000000.0 .321099623 10100000.0 .319299700 10200000.0 .3174997779040 3 18 10300000.0 .315699855 10400000.0 .313899932 10500000.0 .3121000099040 3 18 10600000.0 .310300086 10700000.0 .308500164 10800000.0 .3067002419040 3 18 10900000.0 .304900318 11000000.0 .304000321 11100000.0 .3040002509040 3 18 11200000.0 .304000179 11300000.0 .304000107 11400000.0 .3040000369040 3 18 11500000.0 .304000033 11600000.0 .304000099 11700000.0 .3040001669040 3 18 11800000.0 .304000232 11900000.0 .304000298 12000000.0 .3048003069040 3 18 12100000.0 .306400257 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.001409169 19500000.0 .0014075029040 3102 19600000.0 .001405835 19700000.0 .001404169 19800000.0 .0014025029040 3102 19900000.0 .001400835 20000000.0 0.0 9040 3102 9040 3 0 9040 0 0 9.02320E+4 2.30044E+2 0 0 0 1904033 18 0.000000+0 0.000000+0 0 18 0 2904033 18 0.000000+0 0.000000+0 0 1 6 3904033 18 1.000000-5 0.000000+0 5.000000+5 1.600000-3 2.000000+7 0.000000+0904033 18 0.000000+0 0.000000+0 0 1 24 12904033 18 1.000000-5 0.000000+0 5.000000+5 1.024000-1 8.000000+5 1.440000-2904033 18 9.000000+5 1.000000-2 1.000000+6 4.900000-3 1.200000+6 3.600000-3904033 18 1.400000+6 1.600000-3 1.600000+6 3.600000-3 3.000000+6 1.600000-3904033 18 4.000000+6 3.600000-3 6.000000+6 1.000000-2 2.000000+7 0.000000+0904033 18 904033 0 9.02320E+4 2.30044E+2 0 0 0 1904033102 0.000000+0 0.000000+0 0 102 0 2904033102 0.000000+0 0.000000+0 0 1 6 3904033102 1.000000-5 0.000000+0 1.500000+1 2.500000-3 2.000000+7 0.000000+0904033102 0.000000+0 0.000000+0 0 1 10 5904033102 1.000000-5 2.000000-4 3.000000-2 6.400000-3 1.500000+1 1.000000-2904033102 1.000000+6 4.000000-2 2.000000+7 0.000000+0 904033102 904033 0 9040 0 0 0 0 0 9.22350E+4 2.33024E+2 0 1 34 109228 1451 0.0 1.00000E+0 0 0 0 69228 1451 1.00000E+0 2.00000E+7 0 0 10 20029228 1451 3.00000E+2 0.0 1 0 414 39228 1451 92-U -235 ORNL,LANL,+EVAL-NOV89 WESTON, YOUNG, POENITZ, LUBITZ 9228 1451 DIST-Feb2004 9228 1451 ----IRDF-2002 MATERIAL 9228 9228 1451 -----INCIDENT NEUTRON DATA 9228 1451 ------ENDF-6 FORMAT 9228 1451 **************************************************************** 9228 1451 92-U -235 ORNL,LANL,+EVAL-NOV89 WESTON, YOUNG, POENITZ, LUBITZ 9228 1451 DIST-NOV98 REV5-OCT97 19981109 9228 1451 ----ENDF/B-VI MATERIAL 9228 REVISION 5 9228 1451 *****************EXTRACT FOR SPECIAL PURPOSE FILE*****************9228 1451 DOSIMETRY 9228 1451 ******************************************************************9228 1451 9228 1451 ENDF/B-VI MOD 6 Revision, October 1997, L.C. Leal, H. Derrien, 9228 1451 N.M. Larson, and R.Q. Wright (ORNL) 9228 1451 9228 1451 1. New resonance parameter analysis; File 2, MT=151. 9228 1451 2. Total and prompt-nubar revised below 2 eV; File 1, MT=451, 9228 1451 452,456. 9228 1451 9228 1451 ---------------------------------------------------------------- 9228 1451 File 2 9228 1451 MT=151 Resonance parameters, from a new analysis by Leal 9228 1451 et al. [LE97], using the multilevel R-matrix analysis code 9228 1451 SAMMY [LA96]. Energy range for U235 is 0 to 2.25 keV. 9228 1451 For the first time, integral data were fitted during the 9228 1451 analysis process: Thermal cross sections (fission, 9228 1451 capture, and elastic), Westcott g-factors (fission and 9228 1451 absorption) are from the ENDF/B-6 standards [CA93], and 9228 1451 the K1 value is from Hardy [HA79]. 9228 1451 Thermal parameters obtained in the present evaluation, 9228 1451 first using the microscopic experimental data only, and 9228 1451 second including the integral data as well, are compared 9228 1451 to the SAMMY input in the following Table: 9228 1451 9228 1451 Parameter SAMMY input Fit to Fit to diff. 9228 1451 value diff data & integ. 9228 1451 alone data 9228 1451 ----------- ----------------- -------- ------------ 9228 1451 Fission 584.25 +/- 1.11 582.28 584.88 9228 1451 Capture 98.96 +/- 0.74 99.18 98.66 9228 1451 Scattering 15.46 +/- 1.06 15.44 15.12 9228 1451 Westcott gf 0.9771 +/- 0.0008 0.9743 0.9764 9228 1451 Westcott ga 0.9790 +/- 0.0008 0.9774 0.9785 9228 1451 Westcott gg 0.9956 0.9910 9228 1451 K1(barn) 722.70 +/- 3.90 717.48 722.43 9228 1451 9228 1451 The final adjustment of nu by SAMMY to the recommended K1 9228 1451 value of 722.7 gave nu = 2.4367 +/- 0.0005, with fission 9228 1451 and absorption cross sections calculated from the final 9228 1451 resonance parameters. 9228 1451 In the following Tables, the fission and capture cross 9228 1451 sections calculated in this evaluation with the code 9228 1451 SAMMY are compared with experimental data. 9228 1451 9228 1451 Experimental and calculated total cross sections. 9228 1451 9228 1451 Energy Range Calculated Schrack Weston Weston 9228 1451 (eV) (b.eV) (b.eV) (b.eV) (b.eV) 9228 1451 --------------- ---------- ------- ------ ------ 9228 1451 0.5 - 20.0 910.4 929.9 9228 1451 20.0 - 60.0 1867.8 1882.8 1869.9 9228 1451 60.0 - 100.0 954.0 968.0 954.2 9228 1451 100.0 - 200.0 2032.7 2092.7 2089.5 2073.9 9228 1451 200.0 - 300.0 2062.2 2007.0 2060.0 2054.6 9228 1451 300.0 - 400.0 1280.8 1321.6 1297.1 1292.9 9228 1451 400.0 - 500.0 1333.2 1391.5 1351.8 1347.9 9228 1451 500.0 - 600.0 1489.2 1467.9 1499.2 1494.3 9228 1451 600.0 - 700.0 1126.6 1156.4 1134.1 1132.6 9228 1451 700.0 - 800.0 1088.7 1085.8 1093.3 1075.7 9228 1451 800.0 - 900.0 797.6 784.0 813.0 804.9 9228 1451 900.0 - 1000.0 724.4 723.9 738.2 721.4 9228 1451 1000.0 - 2000.0 7036.1 7054.2 9228 1451 9228 1451 Experimental and calculated capture cross sections. 9228 1451 9228 1451 Energy Range Calculated De Saussure Perez 9228 1451 (eV) (b.eV) (b.eV) (b.eV) 9228 1451 --------------- ---------- ----------- ------ 9228 1451 0.5 - 20.0 653.5 647 9228 1451 20.0 - 60.0 1066.1 1084 1057 9228 1451 60.0 - 100.0 490.2 477 504 9228 1451 100.0 - 200.0 1158.8 1148 1138 9228 1451 200.0 - 300.0 907.8 904 940 9228 1451 300.0 - 400.0 660.2 658 642 9228 1451 400.0 - 500.0 495.9 506 478 9228 1451 500.0 - 600.0 533.3 506 562 9228 1451 600.0 - 700.0 494.8 481 449 9228 1451 700.0 - 800.0 490.1 513 475 9228 1451 800.0 - 900.0 439.8 444 397 9228 1451 900.0 - 1000.0 504.2 542 482 9228 1451 1000.0 - 1100.0 509.6 522 463 9228 1451 1100.0 - 1200.0 413.7 395 332 9228 1451 1200.0 - 1300.0 340.4 372 267 9228 1451 1300.0 - 1400.0 304.1 304 225 9228 1451 1400.0 - 1500.0 355.7 301 254 9228 1451 --------------- ---------- ----------- ------ 9228 1451 20.0 - 1500.0 9164.7 9046 8665 9228 1451 9228 1451 The fission and capture resonance integral calculated from 9228 1451 the present evaluation are 276.04 b and 140.49 b, respectively, 9228 1451 giving a capture-to-fission ratio (alpha value) of 0.509 in 9228 1451 excellent agreement with the value obtained from integral 9228 1451 measurements. 9228 1451 The following energy-differential data were included in the 9228 1451 analysis: 9228 1451 (1) Transmission data of Harvey et al. [HA86] on the ORELA 9228 1451 18-meter flight path, with sample thickness of 0.03269 9228 1451 atoms/barn, cooled to 77 K (0.4 to 68 eV). 9228 1451 (2) Transmission data of Harvey et al. [HA86] on the ORELA 9228 1451 80-meter flight path, with sample thickness of 0.00233 9228 1451 atoms/barn, cooled to 77 K (4 to 2250 eV). 9228 1451 (3) Transmission data of Harvey et al. [HA86] on the ORELA 9228 1451 80-meter flight path, with sample thickness of 0.03269 9228 1451 atoms/barn, cooled to 77 K (4 to 2250 eV). 9228 1451 (4) Fission data of Schrack [SC88] on the RPI Linac at 8.4 9228 1451 meter flight path (0.02 to 20 eV). 9228 1451 (5,6) Fission and capture data of de Saussure et al. [DE67] 9228 1451 on the ORELA 25.2-meter flight path (0.01 to 2250 eV). 9228 1451 (7,8) Fission and capture data of Perez et al. [PE73] on the 9228 1451 ORELA 39-meter flight path (0.01 to 100 eV). 9228 1451 (9) Fission data of Gwin et al. [GW84] on the ORELA 25.6-meter 9228 1451 flight path (0.01 to 20 eV). 9228 1451 (10) Transmission data of Spencer et al. [SP84] on the ORELA 9228 1451 ORELA 18-meter flight path, sample thickness of 0.001468 9228 1451 atom/barn (0.01 to 1.0 eV). 9228 1451 (11) Fission data of Wagemans et al. [WA88] on the Geel 18- 9228 1451 meter flight path (0.001 to 1.0 eV) 9228 1451 (12,13) Absorption and fission data of Gwin [GW96] at ORELA 9228 1451 (0.01 to 4.0 eV). 9228 1451 (14) Fission data of Weston and Todd [WE84] on the ORELA 9228 1451 18.9-meter flight path (14 to 2250 eV). 9228 1451 (15) Eta data of Wartena et al. [WA87] at 8 meters (0.0018 to 9228 1451 1.0 eV). 9228 1451 (16) Eta (chopper) data of Weigmann et al [WE90] (0.0015 to 9228 1451 0.15 eV). 9228 1451 (17) Fission data of Weston and Todd [WE92] on the ORELA 9228 1451 86.5-meter flight path (100 to 2000 eV). 9228 1451 (18) Fission yield data of Moxon et al. [MO92] at ORELA 9228 1451 (0.01 to 50.0 eV). 9228 1451 9228 1451 ---------------------------------------------------------------- 9228 1451 REFERENCES FOR RESOLVED RESONANCE REGION 9228 1451 9228 1451 [CA93] A. Carlson, W.P. Poenitz, G.M. Hale et al., "The ENDF/B-6 9228 1451 Neutron Cross Section Measurements Standards," National 9228 1451 Institute of Standards and Technology report NISTIR-5177 9228 1451 (1993) 9228 1451 [DE67] G. de Saussure, R. Gwin, L.W. Weston, and R.W. Ingle, 9228 1451 "Simultaneous Measurements of the Neutron Fission and Capture 9228 1451 Cross Section for 235U for Incident Neutron Energy from 9228 1451 0.04 eV to 3 keV," Oak Ridge National Laboratory report 9228 1451 ORNL/TM-1804 (1967) 9228 1451 [GW84] R. Gwin, R.R. Spencer, R.W. Ingle, J.H. Todd, and S.W. 9228 1451 Scoles, Nuc.Sci.Eng. 88, 37 (1984) 9228 1451 [GW96] R. Gwin, To be published in Nuclear Science Engineering 9228 1451 [HA79] J. Hardy, Brookhaven National Laboratory, report 9228 1451 BNL-NCS-51123 [ENDF-300] (1979) Sec. B.1 9228 1451 [HA86] J.A. Harvey, N.W. Hill, F.G. Perey et al., Nuclear Data 9228 1451 for Science and Technology, Proc. Int. Conf. May 30-June 3, 9228 1451 1988, Mito, Japan. (Saikon Publishing, 1988) p. 115 9228 1451 [LA96] N.M. Larson, "Updated Users' Guide to SAMMY" report 9228 1451 ORNL/TM-9179/R3 (1996) 9228 1451 [LE97] L.C. Leal, H. Derrien, N.M. Larson, R.Q. Wright, 9228 1451 "R-Matrix Analysis of 235U Neutron Transmission and Cross 9228 1451 Sections in the Energy Range 0 eV to 2.25 keV," Oak Ridge 9228 1451 National Laboratory report ORNL/TM-13516 (1997). 9228 1451 [MO92] M.C. Moxon, J.A. Harvey, and N.W. Hill, private 9228 1451 communication, Oak Ridge National Laboratory (1992). 9228 1451 [PE73] R.B. Perez, G. de Saussure, and E.G. Silver, Nucl.Sci. 9228 1451 Eng. 52, 46 (1973) 9228 1451 [SC88] R.A. Schrack, "Measurement of the 235U(n,f) Reaction from 9228 1451 Thermal to 1 keV," Nuclear Data for Science and Technology, 9228 1451 Proc. Int. Conf. May 30-June 3, Mito, Japan (Saikon 9228 1451 Publishing, 1988) p. 101 9228 1451 [SP84] R.R. Spencer, J.A. Harvey, N.W. Hill, and L. Weston, 9228 1451 Nucl.Sci.Eng. 96, 318 (1987) 9228 1451 [WA87] J.A. Wartena, H. Weigmann, and C. Burkholz, report IAEA 9228 1451 Tecdoc 491 (1987) p.123 9228 1451 [WA88] C. Wagemans, P. Schillebeeckx, A.J. Deruyter, and R. 9228 1451 Barthelemy, "Subthermal Fission Cross Section Measurements 9228 1451 for 233U and 239Pu," Nuclear Data for Science and Technology, 9228 1451 Proc. Int. Conf. May 30-June 3, Mito, Japan (Saikon 9228 1451 Publishing, 1988) p. 91 9228 1451 [WE84] L.W. Weston and J.H. Todd, Nucl.Sci.Eng. 88, 567 (1984) 9228 1451 [WE90] H. Weigmann, P. Geltenbort, B. Keck, K. Shrenckenbach, 9228 1451 and J.A. Wartena, The Physics of Reactors, Proc. Int. Conf., 9228 1451 Marseille, 1990, Vol.1 (1990) p. 133 9228 1451 [WE92] L.W. Weston and J.H. Todd, Nucl.Sci.Eng. 111, 415 (1992) 9228 1451 9228 1451 **************************************************************** 9228 1451 9228 1451 ENDF/B-VI MOD 5 Revision 4 (R.Q.Wright, ORNL, V. McLane, NNDC, 9228 1451 October, 1996) 9228 1451 9228 1451 File 1, MT=451: Update of comments. 9228 1451 File 1, MT=456: Update of prompt nubar. 9228 1451 9228 1451 *****************************************************************9228 1451 9228 1451 ENDF/B-VI MOD 4 Revision, February 1995, C. Lubitz (KAPL) 9228 1451 9228 1451 The resonance parameters below 900 eV were adjusted as described 9228 1451 in C.R. Lubitz, "A modification to ENDF6 U235 to increase 9228 1451 epithermal alpha and K1", Proceedings of the Int. Conf. on 9228 1451 Nuclear Data for Science and Technology, Gatlinburg, TN, May 9228 1451 9-13, Vol. 2 (American Nuclear Soc., 1994) page 646. 9228 1451 9228 1451 *****************************************************************9228 1451 9228 1451 ENDF/B-VI MOD 3 Revision 2 9228 1451 9228 1451 Missing File 1 section 458 added. 9228 1451 9228 1451 *****************************************************************9228 1451 9228 1451 ENDF/B-VI MOD 2 Revision, L. Weston (ORNL) 9228 1451 9228 1451 FILE 1: The uncertainties on the fission cross section 9228 1451 recommended by the Standards Committee of CSEWG are listed. The 9228 1451 updated description of File 2, the resolved resonance parameters,9228 1451 is given. 9228 1451 9228 1451 FILE 2: The resolved resonance parameters have been changed 9228 1451 extensively. The neutron energy region below 4 eV has been made 9228 1451 a separate group. Eta decreases with decreasing neutron energy 9228 1451 below 0.1 eV. At the higher neutron energies, the resonance 9228 1451 parameters are more refined. 9228 1451 9228 1451 FILE 3: Smooth cross sections above 100 keV have minor 9228 1451 corrections. 9228 1451 9228 1451 FILE 31: Nubar covariance files were updated. 9228 1451 9228 1451 FILE 33: All covariance files were removed as correct new files 9228 1451 are not yet available. 9228 1451 *****************************************************************9228 1451 9228 1451 ENDF/B-VI, MOD 1 Evaluation, April 1989 9228 1451 9228 1451 Principal evaluators: 9228 1451 Thermal parameters: Standards Committee of CSEWG. 9228 1451 Resolved resonance region: (0-2250 eV) L.C. Leal (U.Tenn.), 9228 1451 G. deSaussure (ORNL), R.B. Perez (U.Tenn.), N.M. Larson 9228 1451 (ORNL), M.S.Moore (LANL), and R.Q. Wright (ORNL). 9228 1451 Replaced by revision 6. 9228 1451 Unresolved resonance region and File 3 below 100 keV and the 9228 1451 capture cross section above 100 keV: W.P. Poenitz (ANL) and 9228 1451 L.W. Weston (ORNL) 9228 1451 Fission cross section above 100 keV: W.P. Poenitz (ANL) and the 9228 1451 Standards Committee of CSEWG. 9228 1451 Model calculations and fits above 100 keV: P.G. Young (LANL), 9228 1451 R.E. MacFarlane (LANL), and E.D. Arthur (LANL). 9228 1451 Covariance files: R.W. Peelle (ORNL). All files reoved in 9228 1451 revision 1, except MT=456. 9228 1451 9228 1451 File 1 Descriptive and Nubar Information ----------------------- 9228 1451 MT=452 Total nubar. Replaced in revision 6. 9228 1451 MT=455 Delayed neutron yields from England [EN89]. 9228 1451 MT=456 Prompt neutron yields. Replaced in revision 6. 9228 1451 9228 1451 FISSION CROSS SECTION UNCERTAINTIES: The CSEWG Standards 9228 1451 Committee supplied the following uncertainties with the these 9228 1451 statements. "These uncertainties are estimates such that if a 9228 1451 modern day experiment were performed today on a given standard 9228 1451 using the best techniques, those results should fall within these9228 1451 expanded uncertainties (2/3 of the time). They take into account9228 1451 inconsistencies and concerns about R-matrix parameters. Note 9228 1451 that it is not assumed that the uncertainties are totally 9228 1451 correlated within the energy ranges given." 9228 1451 9228 1451 Energy (keV) Estimated Comb Result (%) 9228 1451 Uncertainty (%) 9228 1451 2.53E-05 0.2 (0.19) 9228 1451 150-600 1.5 9228 1451 600-1000 1.6 (0.60) 9228 1451 1000-3000 1.8 9228 1451 3000-6000 2.3 (0.69) 9228 1451 6000-10000 2.2 9228 1451 10000-12000 1.8 (1.14) 9228 1451 12000-14000 1.2 9228 1451 14000-14500 0.8 (0.55) 9228 1451 14500-15000 1.5 9228 1451 15000-16000 2.0 (0.97) 9228 1451 16000-17000 2.5 9228 1451 17000-19000 3.0 (1.26) 9228 1451 19000-20000 4.0 9228 1451 9228 1451 File 2. RESOLVED AND UNRESOLVED RESONANCE PARAMETERS ---------- 9228 1451 MT=151 9228 1451 Resolved resonance region: replaced in revision 6. 9228 1451 Unresolved resonance region: 9228 1451 The unresolved resonance region was derived by a FITACS 9228 1451 (Fritz Froehner code) fit by L.W. Weston to the Standards 9228 1451 Committee recommendation for the fission cross section and 9228 1451 new capture evaluation based on newer alpha measurements 9228 1451 (see ANL-83-4 supplement). These results were then fit 9228 1451 with URES (Ed Pennington code) so ENDF would reproduce. 9228 1451 unresolved resonance region extends from 2.25 to 25 keV 9228 1451 and is used only for self shielding calculations. Dilute 9228 1451 cross sections are taken from File 3 which shows 9228 1451 experimentally observed structure carried over from 9228 1451 version 5 up to 100 keV. 9228 1451 9228 1451 File 3. SMOOTH CROSS SECTIONS ---------------------------------- 9228 1451 Model calculations from LANL: P.G. Young, R.E.MacFarlane, 9228 1451 E.D.Arthur. The evaluation above 100 keV is based on a detailed 9228 1451 theoretical analysis utilizing the available experimental data. 9228 1451 Coupled channel optical model calculations with the ECIS code 9228 1451 [Ra70] were used to provide the total, elastic, and inelastic 9228 1451 cross sections to the first 3 members of the ground state 9228 1451 rotational band, as well as neutron elastic and inelastic 9228 1451 angular distributions to the rotational levels. The ECIS code 9228 1451 was also used to calculate neutron transmission coefficients. 9228 1451 Hauser-Feshbach statistical theory calculations were carried out 9228 1451 with the GNASH [Ar88,Yo77] and COMNUC [Du70] code systems, 9228 1451 including preequilibrium and fission. DWBA calculations were 9228 1451 performed with the DWUCK code [Ku70] for several vibrational 9228 1451 levels, using B(El) values inferred from (d,d') data on U234, 9228 1451 U235, U238, as well as Coulomb excitation measurements. A weak 9228 1451 coupling model [Pe69] was used to apply the U234 and U238 9228 1451 results to states in U235. 9228 1451 A preliminary description of the analysis was given at the 9228 1451 Mito conference [Yo88]. 9228 1451 9228 1451 MT=1 SUM OF PARTIAL CROSS SECTIONS FROM 2.25 TO 100 KEV. 9228 1451 0.10 to 20 Mev, obtained from a covariance analysis of 9228 1451 available experimental data, using an initial or prior cross 9228 1451 section from the coupled-channel optical model analysis. 9228 1451 Experimental data used include [Fo71], [Ve80], [Bo72], [Po81], 9228 1451 [Gr73], [Sc74], [Po83], [Pe60], [Wh65], [Ca73], and [Br58]. 9228 1451 The GLUCS code was used for analysis [He80]. 9228 1451 MT=2 UNCHANGED FROM VERSION 5 FROM 2.25 TO 100 KEV. 9228 1451 0.12 to 20 MeV, based on subtraction of (MT=4,16,17,18,37,102) 9228 1451 from MT=1. 9228 1451 MT=18 2.2 to 120 keV: average values from simultaneous 9228 1451 evaluation of W.P. Poenitz with structure carried over from 9228 1451 version 5. Renormalizaion was over 3 ranges per decade. 9228 1451 0.10 to 20 MeV: standards evaluation by CSEWG Standards 9228 1451 Committee. 9228 1451 9228 1451 ---------------------------------------------------------------- 9228 1451 REFERENCES 9228 1451 9228 1451 [Ar88] E.D. Arthur, Los Alamos National Laboratory report 9228 1451 LA-UR-88-382 (1988) 9228 1451 [Bo72] K. Boeckhoff et al., J.Nuc.En. 26, 91 (1972) 9228 1451 [Br58] A. Bratenahl et al., Phys.Rev. 110, 927 (1958) 9228 1451 [Ca73] J. Cabe et al., report CEA-R-4524 (1973) 9228 1451 [Du70] C.L. Dunford, report AI-AEC-12931 (1970) 9228 1451 [En89] T.R. England et al, Los Alamos National Laboratory 9228 1451 reports LA-11151-MS(88), LA-11534T(89); LAUR-88-4118 to be 9228 1451 published in Nucl.Sci.Eng. (1989) 9228 1451 [Fo71] D. Foster and D. Glasgow, Phys.Rev. C 3, 576 (1971) 9228 1451 [Fr86] J. Frehaut, report NEANDC(E) 238/L (1986) 9228 1451 [Gr73] L. Green et al., report USNDC-9 (1973) p.170 9228 1451 [He80] D. Hetrick and C.Y. Fu, Oak Ridge National Laboratory 9228 1451 report ORNL/TM-7341 (1980) 9228 1451 [Ku70] P.D. Kunz, "DWUCK: A Distorted-Wave Born Approximation 9228 1451 Program," unpublished report 9228 1451 [Ma88] D.G.Madland, Nuclear Data for Science and Technology, 9228 1451 Proc. Int. Conf., Mito, Japan, May 30-June 3, 1988, Mito, 9228 1451 Japan. (Saikon Publishing, 1988) p.759 9228 1451 [Pe60] J. Peterson et al., Phys.Rev. 120, 521 (1960) 9228 1451 [Pe69] R.J. Peterson, Ann.Phys. 53, 40 (1069) 9228 1451 [Po81] W. Poenitz et al., Nuc.Sci.Eng. 78, 333 (1981) 9228 1451 [Po83] W. Poenitz et al., Argonne National Laboratory report 9228 1451 ANL-NDM-80 (1983) 9228 1451 [Ra70] J. Raynal, report IAEA SMR-9/8 (1972) p.281 9228 1451 [Sc74] R. Schwartz et al., Nuc.Sci.Eng. 54, 322 (1974) 9228 1451 [Ut66] C. Uttley et al., Paris Conf. (1966) v1, p165 9228 1451 [Ve80] V. Vertebnyj et al., report YFI-16,8(1973) 9228 1451 [Wh65] W. Whalen et al., Argonne National Laboratory report 9228 1451 ANL-7110 (1965) p.15 9228 1451 [Yo77] P.G. Young and E.D. Arthur, Los Alamos National 9228 1451 Laboratory report LA-6947 (1977). 9228 1451 [Yo88] P.G. Young and E.D. Arthur, Nuclear Data for Science and 9228 1451 Technology, Proc. Int. Conf., Mito, Japan, May 30-June 3, 9228 1451 1988, Mito, Japan. (Saikon Publishing, 1988) p.603 9228 1451 9228 1451 *****************************************************************9228 1451 9228 1451 9228 1451 9228 1451 9228 1451 ************************ C O N T E N T S *********************** 9228 1451 ***************** Program LINEAR (VERSION 2002-1) ***************9228 1451 For All Data Greater than 1.0000E-10 barns in Absolute Value 9228 1451 Data Linearized to Within an Accuracy of .100000000 per-cent 9228 1451 ***************** Program RECENT (VERSION 2002-1) ***************9228 1451 for All Data Greater than 1.0000E-10 barns in Absolute Value 9228 1451 Data Linearized to within an Accuracy of .100000000 per-cent 9228 1451 ***************** Program SIGMA1 (VERSION 2002-1) ***************9228 1451 Data Doppler Broadened to 300.000000 Kelvin 9228 1451 for All Data Greater than 1.0000E-10 barns in Absolute Value 9228 1451 Data Linearized to Within an Accuracy pf .100000000 per-cent 9228 1451 ***************** Program FIXUP (Version 2002-1) ****************9228 1451 Corrected ZA/AWR in All Sections-----------------------------Yes 9228 1451 Corrected Thresholds-----------------------------------------Yes 9228 1451 Extended Cross Sections to 20 MeV----------------------------No 9228 1451 Allow Cross Section Deletion---------------------------------No 9228 1451 Allow Cross Section Reconstruction---------------------------No 9228 1451 Make All Cross Sections Non-Negative-------------------------Yes 9228 1451 Delete Energies Not in Ascending Order-----------------------Yes 9228 1451 Deleted Duplicate Points-------------------------------------Yes 9228 1451 Check for Ascending MAT/MF/MT Order--------------------------Yes 9228 1451 Check for Legal MF/MT Numbers--------------------------------Yes 9228 1451 Allow Creation of Missing Sections---------------------------No 9228 1451 Allow Insertion of Energy Points-----------------------------No 9228 1451 Create Uniform Energy Grid-----------------------------------No 9228 1451 Delete Section if Cross Section =0 at All Energies-----------Yes 9228 1451 ***************** Program GROUPIE (VERSION 2002-1) **************9228 1451 Unshielded Group Averages Using 640 Groups 9228 1451 Weighting Spectrum: Flat (Constant) Spectrum 9228 1451 1 451 421 69228 1451 3 18 217 29228 1451 33 18 53 19228 1451 9228 1 0 9228 0 0 9.22350E+4 2.33024E+2 0 0 0 09228 3 18 1.93720E+8 1.93720E+8 0 0 1 6419228 3 18 641 1 9228 3 18 .000100000 9805.44565 .000105000 9576.58959 .000110000 9359.172989228 3 18 .000115000 9158.81971 .000120000 8924.60065 .000127500 8664.732699228 3 18 .000135000 8427.28509 .000142500 8209.06581 .000150000 7972.884059228 3 18 .000160000 7726.83530 .000170000 7502.95387 .000180000 7297.437119228 3 18 .000190000 7106.20951 .000200000 6931.59545 .000210000 6767.315239228 3 18 .000220000 6614.73347 .000230000 6472.96795 .000240000 6306.559359228 3 18 .000255000 6123.66483 .000270000 5982.52302 .000280000 5825.396019228 3 18 .000300000 5633.76253 .000320000 5460.23314 .000340000 5301.297839228 3 18 .000360000 5154.89776 .000380000 5021.23541 .000400000 4881.864359228 3 18 .000425000 4739.23564 .000450000 4608.94667 .000475000 4489.456319228 3 18 .000500000 4377.06284 .000525000 4274.34196 .000550000 4177.297809228 3 18 .000575000 4087.03116 .000600000 3994.50545 .000630000 3899.312979228 3 18 .000660000 3812.19928 .000690000 3728.62825 .000720000 3639.695339228 3 18 .000760000 3544.32250 .000800000 3455.75135 .000840000 3374.564279228 3 18 .000880000 3297.39250 .000920000 3226.63610 .000960000 3158.862739228 3 18 .001000000 3088.58871 .001050000 3015.21513 .001100000 2946.416769228 3 18 .001150000 2883.00805 .001200000 2808.19450 .001275000 2725.942999228 3 18 .001350000 2650.71743 .001425000 2580.49435 .001500000 2505.864519228 3 18 .001600000 2427.81125 .001700000 2356.53647 .001800000 2290.600769228 3 18 .001900000 2230.35443 .002000000 2174.48226 .002100000 2122.043409228 3 18 .002200000 2074.01496 .002300000 2028.08069 .002400000 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1.972685649228 3 18 18100000.0 1.96845569 18200000.0 1.96422574 18300000.0 1.959995809228 3 18 18400000.0 1.95576585 18500000.0 1.95153590 18600000.0 1.947305959228 3 18 18700000.0 1.94307601 18800000.0 1.93884606 18900000.0 1.934616119228 3 18 19000000.0 1.93259121 19100000.0 1.93277135 19200000.0 1.932951499228 3 18 19300000.0 1.93313162 19400000.0 1.93331176 19500000.0 1.933491909228 3 18 19600000.0 1.93367204 19700000.0 1.93385218 19800000.0 1.934032319228 3 18 19900000.0 1.93421245 20000000.0 0.0 9228 3 18 9228 3 0 9228 0 0 9.22350E+4 2.33024E+2 0 0 0 1922833 18 0.000000+0 0.000000+0 0 18 0 1922833 18 0.000000+0 0.000000+0 1 5 300 24922833 18 1.000000-5 3.000000-1 7.800000+0 1.100000+1 1.000000+2 1.000000+3922833 18 1.000000+4 3.750000+4 9.000000+4 1.350000+5 2.050000+5 2.750000+5922833 18 4.125000+5 6.250000+5 9.500000+5 1.325000+6 1.900000+6 2.900000+6922833 18 4.600000+6 6.750000+6 9.500000+6 1.350000+7 1.550000+7 2.000000+7922833 18 3.634398-6 3.634398-6 3.101778-6 3.101778-6 2.475285-6 2.322641-6922833 18 1.959580-6 1.703434-6 1.466791-6 1.351542-6 1.325030-6 1.266146-6922833 18 1.194608-6 1.079757-6 9.347829-7 9.056430-7 8.066544-7 7.272780-7922833 18 7.101887-7 6.671650-7 5.787847-7 3.145285-7 5.273911-7 4.634398-6922833 18 3.101778-6 3.101778-6 2.475285-6 2.322641-6 1.959580-6 1.703434-6922833 18 1.466791-6 1.351542-6 1.325030-6 1.266146-6 1.194608-6 1.079757-6922833 18 9.347829-7 9.056430-7 8.066544-7 7.272780-7 7.101887-7 6.671650-7922833 18 5.787847-7 3.145285-7 5.273911-7 1.789157-5 1.789157-5 7.247852-6922833 18 6.727942-6 5.540369-6 4.708277-6 3.992362-6 3.642562-6 3.577145-6922833 18 3.415922-6 3.207211-6 2.857612-6 2.437250-6 2.338613-6 2.053255-6922833 18 1.758301-6 1.688479-6 1.584746-6 1.370763-6 7.259742-7 1.252452-6922833 18 1.889157-5 7.247852-6 6.727942-6 5.540369-6 4.708277-6 3.992362-6922833 18 3.642562-6 3.577145-6 3.415922-6 3.207211-6 2.857612-6 2.437250-6922833 18 2.338613-6 2.053255-6 1.758301-6 1.688479-6 1.584746-6 1.370763-6922833 18 7.259742-7 1.252452-6 1.515003-5 1.175065-5 9.613425-6 8.353013-6922833 18 7.031353-6 6.360433-6 6.235565-6 5.958995-6 5.586026-6 4.961029-6922833 18 4.215194-6 4.034481-6 3.527568-6 2.977505-6 2.845209-6 2.669809-6922833 18 2.307214-6 1.212660-6 2.109852-6 1.425651-5 1.027029-5 9.264273-6922833 18 8.092281-6 7.193645-6 6.909848-6 6.597376-6 6.189397-6 5.529027-6922833 18 4.687975-6 4.474730-6 3.910263-6 3.294562-6 3.145358-6 2.949774-6922833 18 2.547051-6 1.336222-6 2.329611-6 1.828855-5 1.084003-5 1.010942-5922833 18 8.788468-6 8.118550-6 7.624354-6 7.217124-6 6.596855-6 5.637136-6922833 18 5.393782-6 4.707867-6 3.960688-6 3.778518-6 3.541295-6 3.057533-6922833 18 1.601948-6 2.796258-6 1.713836-5 1.346428-5 1.135534-5 9.803183-6922833 18 8.931132-6 8.471377-6 7.776282-6 6.667301-6 6.315108-6 5.495624-6922833 18 4.604988-6 4.386879-6 4.109636-6 3.535872-6 1.845778-6 3.233329-6922833 18 2.472297-5 1.420676-5 1.121323-5 1.012406-5 9.380244-6 8.650187-6922833 18 7.491573-6 7.060471-6 6.124241-6 5.104539-6 4.857984-6 4.560243-6922833 18 3.917419-6 2.041280-6 3.581805-6 2.452771-5 1.363176-5 1.080702-5922833 18 9.786191-6 8.890658-6 7.810343-6 7.390222-6 6.436816-6 5.432394-6922833 18 5.172414-6 4.876118-6 4.149269-6 2.158182-6 3.787016-6 2.330751-5922833 18 1.243220-5 1.034842-5 9.414536-6 8.300378-6 7.862886-6 6.840450-6922833 18 5.730159-6 5.447138-6 5.150638-6 4.365020-6 2.258554-6 3.983342-6922833 18 2.215062-5 1.231017-5 9.937610-6 8.819333-6 8.360300-6 7.232673-6922833 18 5.890169-6 5.557986-6 5.194965-6 4.487461-6 2.331275-6 4.112111-6922833 18 2.101510-5 1.116621-5 9.236808-6 8.618871-6 7.453539-6 6.004822-6922833 18 5.674326-6 5.321985-6 4.594350-6 2.389630-6 4.213657-6 1.824382-5922833 18 1.102998-5 9.652663-6 8.263210-6 6.798812-6 6.440576-6 6.060138-6922833 18 5.216436-6 2.681023-6 4.772513-6 1.857572-5 1.154913-5 9.302521-6922833 18 7.887074-6 7.561684-6 7.140802-6 6.174039-6 3.135542-6 5.679011-6922833 18 1.951436-5 1.048593-5 8.568757-6 8.226505-6 7.723123-6 6.680752-6922833 18 3.392445-6 6.159440-6 1.662942-5 1.010621-5 9.422602-6 9.006112-6922833 18 7.813030-6 4.480511-6 7.256347-6 2.313803-5 1.605371-5 1.369189-5922833 18 1.141800-5 5.553546-6 9.669227-6 3.160813-5 1.991877-5 1.359895-5922833 18 6.513847-6 1.172085-5 5.091886-5 2.121602-5 9.574444-6 1.525340-5922833 18 5.533352-5 1.220843-5 1.470634-5 2.737248-5 1.530720-5 8.062025-5922833 18 922833 0 9228 0 0 0 0 0 9.22380E+4 2.36005E+2 0 1 34 49237 1451 0.0 1.00000E+0 0 0 0 69237 1451 1.00000E+0 2.00000E+7 0 0 10 20029237 1451 3.00000E+2 0.0 1 0 423 59237 1451 92-U -238 ORNL+,JAER EVAL-NOV89 L.W.WESTON ET AL.,Y.KANDA ET AL. 9237 1451 DIST-OCT98 REV3-FEB97 19981007 9237 1451 ----IRDF-2002 MATERIAL 9237 REVISION 3 9237 1451 -----INCIDENT NEUTRON DATA 9237 1451 ------ENDF-6 FORMAT 9237 1451 Relative covariances MF33 MT102 in the energy range up to 100 eV 9237 1451 were increased to account for the systematic difference between 9237 1451 the present ENDF/B-VI value of 2.718 barns and the newly 9237 1451 recommended value of 2.683 barns for the thermal capture cross 9237 1451 section. The correction amounts to ((2.718-2.683)/2.683)**2. 9237 1451 Ref.: A.Trkov et al: Re-visiting the U-238 Thermal Capture Cross 9237 1451 Section and Gamma-ray Emission Probabilities of Np-239 Decay, 9237 1451 to be published in Nuclear Science and Engineering, 2005. 9237 1451 ******************************************************************9237 1451 92-U -238 ORNL,LANL+ EVAL-NOV89 L.W.WESTON,P.G.YOUNG,W.POENITZ 9237 1451 DIST-OCT98 REV3-FEB97 19981007 9237 1451 ----ENDF/B-VI MATERIAL 9237 REVISION 3 9237 1451 *****************EXTRACT FOR SPECIAL PURPOSE FILE*****************9237 1451 DOSIMETRY 9237 1451 ******************************************************************9237 1451 92-U -238 KYU,JAERI+ EVAL-MAR94 Y.KANDA ET AL. 9237 1451 DIST-JUL98 9237 1451 ----JENDL/D-99 MATERIAL 9237 9237 1451 9237 1451 9237 1451 U -238 FISSION 9237 1451 **************************************************************** 9237 1451 9237 1451 ENDF/B-VI MOD 4 Revision, February 1997, V.McLane (NNDC) 9237 1451 9237 1451 MT=1 changes 9237 1451 MF=451: Updated comments, references. 9237 1451 MF=455: Corrected TAB1 bookkeeping (minor revision). 9237 1451 MT=2 changes 9237 1451 MF=151 9237 1451 Added l-dependent scattering radius equal to scattering 9237 1451 radius for resolved region. 9237 1451 Corrected energy upper limit in unresolved region. 9237 1451 9237 1451 **************************************************************** 9237 1451 9237 1451 ENDF/B-VI MOD 3 Revision, January 1993, L. Weston (ORNL) 9237 1451 9237 1451 MF=2 changes 9237 1451 Resolved region (to 10 keV): Replaces the preliminary 9237 1451 evaluation of Moxon and Sowerby with the final version 9237 1451 [Mo??]. 9237 1451 Unresolved region: Uses Froehner's evaluation. 9237 1451 9237 1451 **************************************************************** 9237 1451 9237 1451 ENDF/B-VI MOD 2 Revision 1, January 1991 (NNDC) 9237 1451 9237 1451 Covariance files for total, elastic scattering, fission and 9237 1451 capture cross sections removed, since the correct files were and 9237 1451 still are not yet available. 9237 1451 9237 1451 **************************************************************** 9237 1451 9237 1451 ENDF/B-VI MOD 1 Evaluation, November 1989 (ORNL, LANL, ANL) 9237 1451 9237 1451 Principal evaluators 9237 1451 -------------------- 9237 1451 RESOLVED RESONANCE REGION: (10-5 to 10,000 eV) 9237 1451 M.G.Sowerby and M. C. Moxon (Harwell) 9237 1451 UNRESOLVED RESONANCE REGION: (10 TO 149 keV) 9237 1451 F.H.Froehner(KFK) and W.P.Poenitz(ANL) 9237 1451 REACTIONS ABOVE 149 keV: 9237 1451 MODEL CODE CALCULATIONS: (LANL)-P.G.Young and R.E.MacFarlane 9237 1451 FISSION: W.P.Poenitz(ANL) 9237 1451 NU-BAR DELAYED AND DELAYED NEUTRON SPECTRA: 9237 1451 KAISER AND CARPENTER (Ka78) 9237 1451 GAMMA PRODUCTION FILES: R. HOWERTON (LLL) 9237 1451 UNCERTAINTY FILES: NOT YET AVAILABLE 9237 1451 9237 1451 ---------------------------------------------------------------- 9237 1451 9237 1451 ENDF/B-VI EVALUATION ABOVE THE UNRESOLVED RESONANCE REGION 9237 1451 P.G.Young and R.E.MacFarlane (LANL) 9237 1451 Updated by R.Q. Wright (ORNL) 9237 1451 9237 1451 Above the unresolved resonance region, new evaluations were 9237 1451 performed of the neutron total, (n,2n), (,n,3n), (n,4n), (n,f), 9237 1451 (n,nf), (n,2nf), (n,3nf), and (n,gamma) cross sections as well as9237 1451 prompt nubar. The elastic and inelastic data from ENDF/B-V were 9237 1451 carried over for Version VI. 9237 1451 To provide the new data, coupled channel optical model calcu- 9237 1451 lations were performed with the ECIS code (Ra70) for the lowest 9237 1451 3 members of the U238 ground state rotational band. These calcu- 9237 1451 lations were used to provide initial (prior) values for a covar- 9237 1451 iance analysis of the total cross section and to provide neutron 9237 1451 transmission coefficients for nuclear reaction theory 9237 1451 calculations with the GNASH (Ar88,Yo77) and COMNUC (Du70) Hauser-9237 1451 Feshbach statistical/fission/preequilibrium codes. These theory 9237 1451 calculations were used to provide the MF=6 neutron distributions 9237 1451 from the (n,2n), (n,3n), and (n,4n) reactions as well as prior 9237 1451 values for covariance analyses of the cross sections for those 9237 1451 reactions. Additionally, the above analyses plus DWBA calcu- 9237 1451 lations were used to check the ENDF/B-V evaluation of elastic 9237 1451 and inelastic scattering. While some differences were found, 9237 1451 the earlier work was generally found to be reliable, and we 9237 1451 decided to carry over the ENDF/B-V data because of the effort 9237 1451 taken to match experimental data, both at lower energies and at 9237 1451 14 MeV. 9237 1451 9237 1451 MF=1 DESCRIPTIVE INFORMATION 9237 1451 9237 1451 MF=2 RESONANCE PARAMETERS (Updated for MOD 3) 9237 1451 MT=151 RESOLVED AND UNRESOLVED RESONANCE PARAMETERS 9237 1451 RESOLVED RESONANCE PARAMETERS (1.0E-5 eV to 10.0 keV) 9237 1451 Reich-Moore analysis from new evaluation by Moxon and 9237 1451 Sowerby [1]. Present parameters are a preliminary set 9237 1451 due to use of preliminary capture yield data of 9237 1451 Macklin et al. [2]. Parameters above 4.0 kev corrected9237 1451 for background and normalisation errors in [2]. 9237 1451 Parameters 0f the bound resonances evaluated so as to 9237 1451 reproduce the following values at 2200m/s 9237 1451 9237 1451 Reaction Cross section (barns) Reference 9237 1451 Elastic 9.38 [3] 9237 1451 Fission 11.0E-6 [4] 9237 1451 Capture 2.708 [5],[6] 9237 1451 RESOLVED RESONANCE RANGE (10.0 to 149 keV) 9237 1451 Coherent calculation of point cross sections and of 9237 1451 energy-dependent average resonance parameters from 9237 1451 evaluated average resonance parameters for E = 0 with 9237 1451 Hauser-Feshbach code FITACS [7] 9237 1451 In the unresolved region the point cross sections given 9237 1451 in MF=3 are considered as correct. The unresolved 9237 1451 parameters in MF=2 are only to be used for self- 9237 1451 shielding calculations. The use of the parameters with 9237 1451 ENDF processing codes does not lead to cross sections 9237 1451 consistent with MF=3 as the codes use a more primative 9237 1451 version of the formalism. 9237 1451 Cross-sections used to determine the parameters 9237 1451 TOTAL: determined mainly by the following data. 9237 1451 below 45 keV: Uttley et al. [19], Byoun et al. [20], 9237 1451 Kononov and Poletaev [9], Tsubone et al. [8], 9237 1451 above 45 keV: Smith et al. [10] 9237 1451 ELASTIC SCATTERING: 9237 1451 Calculated as (total)-(partial cross-sections) 9237 1451 TOTAL INELASTIC AND INELASTIC LEVELS MT=51 AND 52 9237 1451 Determined mainly by the following data: 9237 1451 Winters et al. [11], Murzin et al. [12], Tsang and 9237 1451 Brugger [13], Smith [14], with level scheme from 9237 1451 Table of Isotopes [15] 9237 1451 Level No. Energy(MeV) Spin-parity 9237 1451 G.S. 0.0 0+ 9237 1451 1 0.0448 2+ 9237 1451 2 0.1484 4+ 9237 1451 3 0.3072 6+ 9237 1451 FISSION 9237 1451 Extrapolated downward from JEF-1. 9237 1451 CAPTURE CROSS-SECTION 9237 1451 Determined mainly by the post-1966 data compiled by 9237 1451 Poenitz [16], by the measurement of Kazakov [17] and 9237 1451 by the mean parameters (radiation width, level 9237 1451 density) extracted from resolved resonance 9237 1451 parameters [7]. 9237 1451 9237 1451 2200 M/S VALUES AND RESONANCE INTEGRALS GIVEN BY EVALUATION 9237 1451 9237 1451 2200m/s values Resonance Integral 9237 1451 (barns) (barns) 9237 1451 Total 12.32 597.4 9237 1451 Elastic 9.60 319.4 9237 1451 Fission 6.46E-05 1.7E-03 9237 1451 Capture 2.72 278.0 9237 1451 9237 1451 ---------------------------------------------------------------- 9237 1451 REFERENCES FOR MF=2 9237 1451 9237 1451 1. M.C. Moxon and M.G. Sowerby, Pr0c 1988 Int.Reac.Phys.Conf. 9237 1451 Jackson Hole (1988) Vol.1, p,281 9237 1451 2. R.L. Macklin et al, Conf. 1988 Mito, Japan, (1988) p.71 9237 1451 3. S.F. Mughabghab, Neutron Cross Sections, Vol.1, Part B, 9237 1451 Z=61-100 (Academic Press, 1984) 9237 1451 4. P. d'Hondt et al, Ann.Nucl.EN., 11, 485 (1984) 9237 1451 5. W.P. Poenitz, report IAEA-TECDOC-335 (1985) p.426 and 9237 1451 private communication (1987) 9237 1451 6. A.D. Carlson et al, Proc.Conf. 1985 Santa Fe, Vol.2, p.1429 9237 1451 (1985) 9237 1451 7. F.H. Froehner, PROC 1988 INT.REAC.PHYS. CONF., Jackson Hole 9237 1451 (1988) Vol.III, p 171 9237 1451 8. I. Tsubone et al, Nucl.Sci.Eng. 88, 579 (1984) 9237 1451 9. V.N. Kononov and E.D. Poletaev, 1973 KIEV CONF 2, 199 (1984)9237 1451 10. A.B. Smith et al, private communication (1988) 9237 1451 11. R.R. Winters et al, Nucl.Sci.Eng. 78, 147 (1981) 9237 1451 12. A.W. Murzin et al, Yad.Konst. 4, 30 (1986) 9237 1451 13. F.Y. Tsang and R.M. Brugger, Nucl.Sci.Eng. 65, 70 (1978) 9237 1451 14. A.B. Smith, report INDC/P(84)-28, unpublished (1984) 9237 1451 15. C.M. Lederer, Ed., Table of Isotopes, 7th edition, (1978) 9237 1451 16. W.P. Poenitz et al, Argonne National Lab. report ANL-83-4 9237 1451 suppl. (1983) 9237 1451 17. L.E. Kazakov et al, Yad.Konst. 3, 37 (1986) 9237 1451 18. F.C. Difilippo et al, Phys.Rev.C 21, 1400 (1980) 9237 1451 19. C.A. Uttley et al., Nucl.Data for Reactors, Proc.Int.Conf. 9237 1451 Paris 1966 (IAEA, 1966) Vol.I, p.165 9237 1451 20. T.Y. Byoun et al., PROC.MEET. REACTOR PHYS. AND SHIELD CALC.,9237 1451 Kiamesha Lake, NY, 1972 (U.S.Nucl.Reg.Comm., 1972) Vol.1, 9237 1451 p.1115 9237 1451 9237 1451 MF=3 SMOOTH CROSS SECTIONS 9237 1451 MT=102 Radiative Capture Cross Section. 10 to 149 keV, 9237 1451 evaluation by F.H. Froehner (KFK) which is based on FITACS 9237 1451 fit to the experimental data. 0.15 to 20 MeV, taken 9237 1451 directly from simultaneous standards analysis [Ca85], 9237 1451 [Po89]. 9237 1451 9237 1451 ---------------------------------------------------------------- 9237 1451 REFERENCES 9237 1451 9237 1451 [Al61] K. Allen et al., J.Nuc.En. 14, 100 (1961) 9237 1451 [An58] G. Antropov et al., At.En. 5, 456 (1958) 9237 1451 [Ar88] E.D. Arthur, Los Alamos National Lab. report LA-UR-88-382 9237 1451 (1988) 9237 1451 [Ba66] D. Barr, Los Aalamos National Lab., private communication 9237 1451 to R.Howerton (1966) 9237 1451 [Ba65] R. Batchelor et al., Nuc.Phys. 65, 236 (1965) 9237 1451 [Br58] A. Bratenahl et al., Phys.Rev. 110, 927 (1958) 9237 1451 [Ca73] J. Cabe et al., CEN Saclay report CEA-R-4524 (1973) 9237 1451 [Ca85] A. Carlson et al., Nucl. Data for Basic & Applied Science,9237 1451 Proc.Conf., Santa Fe, NM 1985 [] p.1429. 9237 1451 [Di80] F.C. Difillippo et al., Phys.Rev.C 21, 1400 (1980) 9237 1451 [Du70] C.L. Dunford, Atomics Int. report AI-AEC-12931 (1970) 9237 1451 [Fo71] D. Foster and D. Glasgow, Phys.Rev.C 3, 576 (1971) 9237 1451 [Fr80a] J. Frehaut et al., Nucl.Sci.Eng. 74, 29 (1980) 9237 1451 [Fr80b] J. Frehaut et al., Brookhaven National Lab. report 9237 1451 BNL-NCS-512457 (1980) p.399 9237 1451 [Fr86] J. Frehaut, NEANDC(E) 238/L (1986) 9237 1451 [Ha73] S. Hayes et al., Nuc.Sci.Eng. 50, 243 (1973) 9237 1451 [He80] D. Hetrick and C.Y. Fu, Oak Ridge National Lab. report 9237 1451 ORNL/TM-7341 (1980) 9237 1451 [Ka78] R. Kaiser and S. Carpenter (ANL-West) private 9237 1451 communication (1978). 9237 1451 [Ko80] N. Kornilov et al., report ZFK-410 (1980) p.68 9237 1451 [La73] J. Landrum et al., Phys.Rev.C 8, 1938 (1973) 9237 1451 [Li79] P. Lisowski et al., (LANL WNR measurement) private 9237 1451 communication (1979) 9237 1451 [Ma69] D. Mather, report AWRE-O-47 (1969) 9237 1451 [Ma72] D. Mather et al., report AWRE-O-72 (1972). 9237 1451 [Pe60] J. Peterson et al., Phys.Rev. 120, 521 (1960) 9237 1451 [Pe61] J. Perkin, J.Nuc.En. 14, 69 (1961) 9237 1451 [Ph56] J. Phillips, report AERE-NP/R-2033 (1956) 9237 1451 [Po81] W. Poenitz et al., Nuc.Sci.Eng. 78, 333 (1981) 9237 1451 [Po83] W.Poenitz et al., Argonne National Lab. report 9237 1451 ANL-NDM-80 (1983) 9237 1451 [Po89] W. Poenitz, ANL-West, private communication (1989) 9237 1451 [Ra70] J. Raynal, report IAEA SMR-9/8 (1970) 9237 1451 [Ro57] L. Rosen et al., Los Alamos National Lab. report LA-2111 9237 1451 (1958) 9237 1451 [Ry80] T. Ryves, J.Phys.G 6, 771 (1980) 9237 1451 [Sc74] R. Schwartz et al., Nuc.Sci.Eng. 54, 322 (1974) 9237 1451 [Sh78] R. Shamu et al., private communication (1978) 9237 1451 [Sl77] R. Slovacek et al., Nuc.Sci.Eng. 62, 455 (1977) 9237 1451 [Sm82] A. Smith et al., Argonne National Lab. report ANL/NDM-74 9237 1451 (1982) 9237 1451 [Ut66] C. Uttley et al., Paris Conf. (1966) vol.1, p.165 9237 1451 [Ve78] L. Veeser and E. Arthur, Harwell Nuc.Data Conf. (1978) 9237 1451 p 1054 9237 1451 [Wh62] P. White et al., J.Nuc.En.A/B 16, 261 (1962) 9237 1451 [Wh71] J. Whalen et al., Argonne National Lab. report ANL-7710 9237 1451 (1971) p.9 9237 1451 [Yo77] P.G. Young and E.D. Arthur, Los Alamos National Lab. 9237 1451 report LA-6947 (1977) 9237 1451 [Yo78] Chou You-Pu, report HSJ-77091 (1978) 9237 1451 9237 1451 **************************************************************** 9237 1451 Modifications applied by the IRDF evaluators 9237 1451 H.Vonach and S.Tagesen, IRK Vienna 9237 1451 MT=102 Capture Cross Section 9237 1451 Uncertainties from simultaneous standards analysis 9237 1451 used up to En=2.3+6 ev 9237 1451 Uncertainties for En=2.3+6 to 2.0+7 ev estimated from 9237 1451 existing experimental data (1) 9237 1451 REFERENCES: 9237 1451 1.Neutron Cross Sections, vol2, V.McLane, C.L.Dunford, 9237 1451 P.F.Rose, Academic Press(1988) 9237 1451 9237 1451 **************************************************************** 9237 1451 9237 1451 ENDF/B-V MAT 1393 Evaluation, November 1981 (ANL, LLNL) 9237 1451 9237 1451 Description of data carried over from ENDF/B-V 9237 1451 9237 1451 Principal Evaluators: 9237 1451 E.M.Pennington (ANL), A.B.Smith (ANL), W.P.Poenitz (ANL) 9237 1451 R.Howerton (LLL) - gamma production files 9237 1451 Contributing evaluators: 9237 1451 R.Kaiser and S.Carpenter (ANL-west) - nu-bar delayed and 9237 1451 delayed neutron spectra 9237 1451 9237 1451 MF = 3 smooth cross sections 9237 1451 Cross sections above 45 kev were evaluated by A.Smith. 9237 1451 more details are provided under each reaction type below. Most 9237 1451 of the evaluation above 45 kev is described in ANL/NDM-32. 9237 1451 ---------------------------------------------------------------- 9237 1451 REFERENCES 9237 1451 1. R.E. Kaiser and S.G. Carpenter (ANL-west) private comm. 9237 1451 (March 1978) 9237 1451 2. S.A. Cox, Argonne National Lab. report ANL/NDM-5 (1974) 9237 1451 3. C. Besant et al, Sem. Fast Pulsed Reactors CONF-760111 (1976)9237 1451 4. F. Manero and V. Konshin, At.En.Review 10, #4 (1972) 9237 1451 5. M. Soleilhac, revised data received from NNCSC 9237 1451 6. B. Nurpeisov et al., Sov.At.En. translation 807 (Mar 1976) 9237 1451 7. G. deSaussure et al., Prog.Nucl.En. 3, 87 (1979) 9237 1451 8. D. Olsen et al., Nuc.Sci.Eng. 62, 479 (1977) 9237 1451 9. P. Guenther, D. Havel, A. Smith Argonne National Lab. report 9237 1451 ANL-NDM-22 (1976) 9237 1451 10. E. Pennington, W. Poenitz, A. Smith, Trans.Am.Nucl.Soc. 26, 9237 1451 591 (1977) 9237 1451 11. W.P. Poenitz Argonne National Lab. report ANL/NDM-45 (1979) 9237 1451 12. W.P. Poenitz and A.B. Smith, ed. Argonne National Lab. report9237 1451 ANL-76-90 (1976) 9237 1451 13. R. Slovacek et al., Nucl.Sci.Eng. 62, 455 (1977) 9237 1451 14. M. Caner, M. Segev, W. Yiftah, Nucl.Sci.Eng. 59, 395 (1976) 9237 1451 15. E. Kujawski and L. Stewart, Trans.Am.Nucl.Soc. 24, 453 (1976)9237 1451 16. L.W. Weston private comm. to B.A. Magurno Nov.12, 1982 9237 1451 17. R. Sher and ?. Beck report EPRI NP-1771/81 and rev. 1/83, 9237 1451 and private communication to Magurno, February 1983 9237 1451 9237 1451 **************************************************************** 9237 1451 92-U -238 KYU,JAERI+ EVAL-MAR94 Y.KANDA ET AL. 9237 1451 DIST-JUL98 9237 1451 ----JENDL/D-99 MATERIAL 9237 9237 1451 -----INCIDENT NEUTRON DATA 9237 1451 ------ENDF-6 FORMAT 9237 1451 HISTORY 9237 1451 87-01 SIMULTANEOUS EVALUATION FOR FISSION AND CAPTURE CROSS 9237 1451 SECTIONS WAS COMPLETED IN THE ENERGY RANGE ABOVE 50 KEV. 9237 1451 93-03 FISSION SPECTRA CALCULATED BY T.OHSAWA(KINKI UNIV.) 9237 1451 COMPILED BY T.NAKAGAWA (NDC/JAERI) 9237 1451 98-04 COMPILED TO JENDL DOSIMETRY FILE 99. 9237 1451 9237 1451 ===== POINT-WISE DATA FILE ===== 9237 1451 9237 1451 9237 1451 U -238 FISSION 9237 1451 9237 1451 MF=1 GENERAL INFORMATION 9237 1451 MT=451 DESCRIPTIVE DATA AND DIRECTORY RECORDS 9237 1451 9237 1451 9237 1451 2200-M/S CROSS SECTIONS AND CALCULATED RESONANCE INTEGRALS. 9237 1451 2200 M/S(B) RES. INTEG.(B) 9237 1451 FISSION 11.8E-6 1.72 9237 1451 9237 1451 ***************** PROGRAM LINEAR (VERSION 87-1) *****************9237 1451 DATA LINEARIZED TO WITHIN AN ACCURACY OF 0.800 PER-CENT 9237 1451 ***************** PROGRAM RECENT (VERSION 89-1) *****************9237 1451 RESONANCE CONTRIBUTION RECONSTRUCTED TO WITHIN 1.000 PER-CENT 9237 1451 COMBINED DATA NOT THINNED (ALL RESONANCE + BACKGROUND DATA KEPT) 9237 1451 ***************** PROGRAM LINEAR (VERSION 87-1) *****************9237 1451 DATA LINEARIZED TO WITHIN AN ACCURACY OF 0.800 PER-CENT 9237 1451 9237 1451 MF=3 NEUTRON CROSS SECTIONS 9237 1451 BELOW 10 KEV, NO BACKGROUND CROSS SECTIONS WERE GIVEN. 9237 1451 ABOVE 10 KEV, CROSS SECTIONS WERE EVALUATED AS FOLLOWS, AND 9237 1451 THEY WERE REPRESENTED WITH THE UNRESOLVED RESONANCE 9237 1451 PARAMETERS EXCEPT THE FISSION CROSS SECTION. 9237 1451 9237 1451 MT=18 FISSION 9237 1451 BELOW 100 KEV : TAKEN FROM EXPERIMENTAL DATA /12/. 9237 1451 100 - 600 KEV : EVALUATED ON THE BASIS OF THE DATA OF 9237 1451 DIFFILIPPO ET AL. /13/, BEHRENS AND CARLSON /14/, 9237 1451 NORDBORG ET AL. /15/ AND MEADOWS /16,17/. 9237 1451 ABOVE 600 KEV : RESULTS OF SIMULTANEOUS EVALUATION /18/ MADE 9237 1451 BY CONSIDERING THE EXPERIMENTAL DATA OF REFS./14,15, 9237 1451 16,17,19,20,21,22,23,24,25,26,27,28/. 9237 1451 9237 1451 MF=33 9237 1451 MT=18 Taken from JENDL/D-99 9237 1451 9237 1451 9237 1451 REFERENCES 9237 1451 12) DIFILLIPO F.C. ET AL.: NUCL. SCI. ENG., 63, 153 (1977). 9237 1451 13) DIFILLIPO F.C. ET AL.: NUCL. SCI. ENG., 68, 43 (1978). 9237 1451 14) BEHRENS J.W. AND CARLSON G.W.: NUCL. SCI. ENG., 63, 250 9237 1451 (1977). 9237 1451 15) NORDBORG C. ET AL: ANL-76-90, 128 (1976). 9237 1451 16) MEADOWS J.W.: NUCL. SCI. ENG., 58, 255 (1975). 9237 1451 17) MEADOWS J.W.: NUCL. SCI. ENG., 49, 310 (1972). 9237 1451 18) KANDA Y. ET AL.: 1985 SANTA FE, 2, 1567 (1986). 9237 1451 19) CANCE M. AND GRENIER G.:NUCL. SCI. ENG., 68, 197 (1978). 9237 1451 20) BILAUD P. ET AL.:1958 GENEVA,16, 106, 5809 (1958). 9237 1451 21) ADAMOV V.M. ET AL.: 1977 NBS, 313 (1977). 9237 1451 22) ARLT R. ET AL.: KE, 24, 48, 8102 (1981). 9237 1451 23) CIERJACKS S. ET AL.:1976 ANL, 94 (1976). 9237 1451 24) GOVERDOVSKII A.A. ET AL.: 1983 KIEV, 2, 159 (1983). 9237 1451 25) ANDROSENKO S.D. ET AL.: 1983 KIEV, 2, 153 (1983). 9237 1451 26) FURSOV B.I. ET AL.: SOV. ATOM. ENERG., 43, 808 (1978). 9237 1451 27) POENITZ W.P. AND ARMANI R.J.: J. NUCL. ENEG., 26, 483 (1972). 9237 1451 28) POENITZ W.P.: NUCL. SCI. ENG., 57, 300 (1975). 9237 1451 9237 1451 ************************ C O N T E N T S *********************** 9237 1451 ***************** Program LINEAR (VERSION 2002-1) ***************9237 1451 For All Data Greater than 1.0000E-10 barns in Absolute Value 9237 1451 Data Linearized to Within an Accuracy of .100000000 per-cent 9237 1451 ***************** Program RECENT (VERSION 2002-1) ***************9237 1451 for All Data Greater than 1.0000E-10 barns in Absolute Value 9237 1451 Data Linearized to within an Accuracy of .100000000 per-cent 9237 1451 ***************** Program SIGMA1 (VERSION 2002-1) ***************9237 1451 Data Doppler Broadened to 300.000000 Kelvin 9237 1451 for All Data Greater than 1.0000E-10 barns in Absolute Value 9237 1451 Data Linearized to Within an Accuracy pf .100000000 per-cent 9237 1451 ***************** Program FIXUP (Version 2002-1) ****************9237 1451 Corrected ZA/AWR in All Sections-----------------------------Yes 9237 1451 Corrected Thresholds-----------------------------------------Yes 9237 1451 Extended Cross Sections to 20 MeV----------------------------No 9237 1451 Allow Cross Section Deletion---------------------------------No 9237 1451 Allow Cross Section Reconstruction---------------------------No 9237 1451 Make All Cross Sections Non-Negative-------------------------Yes 9237 1451 Delete Energies Not in Ascending Order-----------------------Yes 9237 1451 Deleted Duplicate Points-------------------------------------Yes 9237 1451 Check for Ascending MAT/MF/MT Order--------------------------Yes 9237 1451 Check for Legal MF/MT Numbers--------------------------------Yes 9237 1451 Allow Creation of Missing Sections---------------------------No 9237 1451 Allow Insertion of Energy Points-----------------------------No 9237 1451 Create Uniform Energy Grid-----------------------------------No 9237 1451 Delete Section if Cross Section =0 at All Energies-----------Yes 9237 1451 ***************** Program GROUPIE (VERSION 2002-1) **************9237 1451 Unshielded Group Averages Using 640 Groups 9237 1451 Weighting Spectrum: Flat (Constant) Spectrum 9237 1451 1 451 432 49237 1451 3 18 217 19237 1451 3 102 217 39237 1451 33 18 11 09237 1451 33 102 335 09237 1451 9237 1 0 9237 0 0 9.22380E+4 2.36005E+2 0 0 0 09237 3 18 1.89629E+8 1.89629E+8 0 0 1 6419237 3 18 641 1 9237 3 18 .000100000 .000184668 .000105000 .000180362 .000110000 .0001762729237 3 18 .000115000 .000172502 .000120000 .000168096 .000127500 .0001632069237 3 18 .000135000 .000158740 .000142500 .000154635 .000150000 .0001501939237 3 18 .000160000 .000145564 .000170000 .000141353 .000180000 .0001374889237 3 18 .000190000 .000133892 .000200000 .000130609 .000210000 .0001275209237 3 18 .000220000 .000124650 .000230000 .000121984 .000240000 .0001188559237 3 18 .000255000 .000115416 .000270000 .000112762 .000280000 .0001098089237 3 18 .000300000 .000106207 .000320000 .000102945 .000340000 9.99571E-59237 3 18 .000360000 9.72059E-5 .000380000 9.46943E-5 .000400000 9.20757E-59237 3 18 .000425000 8.93960E-5 .000450000 8.69475E-5 .000475000 8.47028E-59237 3 18 .000500000 8.25920E-5 .000525000 8.06627E-5 .000550000 7.88401E-59237 3 18 .000575000 7.71450E-5 .000600000 7.54077E-5 .000630000 7.36205E-59237 3 18 .000660000 7.19851E-5 .000690000 7.04164E-5 .000720000 6.87471E-59237 3 18 .000760000 6.69572E-5 .000800000 6.52954E-5 .000840000 6.37724E-59237 3 18 .000880000 6.23249E-5 .000920000 6.09980E-5 .000960000 5.97271E-59237 3 18 .001000000 5.84097E-5 .001050000 5.70345E-5 .001100000 5.57454E-59237 3 18 .001150000 5.45576E-5 .001200000 5.31564E-5 .001275000 5.16164E-59237 3 18 .001350000 5.02084E-5 .001425000 4.88945E-5 .001500000 4.74987E-59237 3 18 .001600000 4.60394E-5 .001700000 4.47074E-5 .001800000 4.34754E-59237 3 18 .001900000 4.23502E-5 .002000000 4.13069E-5 .002100000 4.03278E-59237 3 18 .002200000 3.94312E-5 .002300000 3.85737E-5 .002400000 3.75953E-59237 3 18 .002550000 3.65078E-5 .002700000 3.56570E-5 .002800000 3.47323E-59237 3 18 .003000000 3.35937E-5 .003200000 3.25576E-5 .003400000 3.16104E-59237 3 18 .003600000 3.07471E-5 .003800000 2.99506E-5 .004000000 2.91208E-59237 3 18 .004250000 2.82791E-5 .004500000 2.75109E-5 .004750000 2.67928E-59237 3 18 .005000000 2.61334E-5 .005250000 2.55216E-5 .005500000 2.49428E-59237 3 18 .005750000 2.44116E-5 .006000000 2.38534E-5 .006300000 2.32950E-59237 3 18 .006600000 2.27684E-5 .006900000 2.22829E-5 .007200000 2.17535E-59237 3 18 .007600000 2.11862E-5 .008000000 2.06673E-5 .008400000 2.01803E-59237 3 18 .008800000 1.97270E-5 .009200000 1.93056E-5 .009600000 1.89046E-59237 3 18 .010000000 1.84894E-5 .010500000 1.80505E-5 .011000000 1.76491E-59237 3 18 .011500000 1.72663E-5 .012000000 1.68277E-5 .012750000 1.63424E-59237 3 18 .013500000 1.58931E-5 .014250000 1.54812E-5 .015000000 1.50421E-59237 3 18 .016000001 1.45812E-5 .017000001 1.41583E-5 .017999999 1.37719E-59237 3 18 .018999999 1.34166E-5 .020000000 1.30832E-5 .021000000 1.27781E-59237 3 18 .022000000 1.24907E-5 .023000000 1.22230E-5 .024000000 1.19084E-59237 3 18 .025500000 1.15692E-5 .027000001 1.13006E-5 .028000001 1.10089E-59237 3 18 .029999999 1.06500E-5 .032000002 1.03234E-5 .034000002 1.00257E-59237 3 18 .035999998 9.75425E-6 .037999999 9.50355E-6 .039999999 9.24161E-69237 3 18 .042500000 8.97539E-6 .045000002 8.73211E-6 .047499999 8.50551E-69237 3 18 .050000001 8.29718E-6 .052499998 8.10449E-6 .055000000 7.92212E-69237 3 18 .057500001 7.75531E-6 .059999999 7.58005E-6 .063000001 7.40522E-69237 3 18 .066000000 7.24071E-6 .068999998 7.08773E-6 .071999997 6.92131E-69237 3 18 .075999998 6.74264E-6 .079999998 6.57928E-6 .083999999 6.42576E-69237 3 18 .088000000 6.28307E-6 .092000000 6.15099E-6 .096000001 6.02488E-69237 3 18 .100000001 5.89475E-6 .104999997 5.75707E-6 .109999999 5.63137E-69237 3 18 .115000002 5.51135E-6 .119999997 5.37402E-6 .127499998 5.22263E-69237 3 18 .135000005 5.08266E-6 .142499998 4.95465E-6 .150000006 4.81743E-69237 3 18 .159999996 4.67296E-6 .170000002 4.54092E-6 .180000007 4.42044E-69237 3 18 .189999998 4.30995E-6 .200000003 4.20631E-6 .209999993 4.11165E-69237 3 18 .219999999 4.02241E-6 .230000004 3.93941E-6 .239999995 3.84384E-69237 3 18 .254999995 3.73714E-6 .270000011 3.65553E-6 .280000001 3.56497E-69237 3 18 .300000012 3.45451E-6 .319999993 3.35535E-6 .340000004 3.26599E-69237 3 18 .360000014 3.18067E-6 .379999995 3.10393E-6 .400000006 3.02593E-69237 3 18 .425000012 2.94428E-6 .449999988 2.87043E-6 .474999994 2.80166E-69237 3 18 .500000000 2.73988E-6 .524999976 2.68321E-6 .550000012 2.62772E-69237 3 18 .574999988 2.57815E-6 .600000024 2.52624E-6 .629999995 2.47453E-69237 3 18 .660000026 2.42564E-6 .689999998 2.38131E-6 .720000029 2.33401E-69237 3 18 .759999990 2.28178E-6 .800000012 2.23587E-6 .839999974 2.19123E-69237 3 18 .879999995 2.15172E-6 .920000017 2.11434E-6 .959999979 2.08108E-69237 3 18 1.00000000 2.04424E-6 1.04999995 2.00718E-6 1.10000002 1.97269E-69237 3 18 1.14999998 1.94246E-6 1.20000005 1.90602E-6 1.27499998 1.86651E-69237 3 18 1.35000002 1.83288E-6 1.42499995 1.80086E-6 1.50000000 1.76880E-69237 3 18 1.60000002 1.73647E-6 1.70000005 1.70830E-6 1.79999995 1.68472E-69237 3 18 1.89999998 1.66301E-6 2.00000000 1.64531E-6 2.09999990 1.63059E-69237 3 18 2.20000005 1.61794E-6 2.29999995 1.60871E-6 2.40000010 1.59963E-69237 3 18 2.54999995 1.59324E-6 2.70000005 1.59020E-6 2.79999995 1.59329E-69237 3 18 3.00000000 1.60347E-6 3.20000005 1.62120E-6 3.40000010 1.64925E-69237 3 18 3.59999990 1.68712E-6 3.79999995 1.73754E-6 4.00000000 1.81176E-69237 3 18 4.25000000 1.92264E-6 4.50000000 2.07439E-6 4.75000000 2.28731E-69237 3 18 5.00000000 2.59737E-6 5.25000000 3.07418E-6 5.50000000 3.87484E-69237 3 18 5.75000000 5.42717E-6 6.00000000 1.02457E-5 6.30000019 .0001078319237 3 18 6.59999990 .001115810 6.90000010 8.67297E-6 7.19999981 1.30114E-69237 3 18 7.59999990 3.80596E-7 8.00000000 2.50553E-7 8.39999962 2.58209E-79237 3 18 8.80000019 2.97648E-7 9.19999981 3.43688E-7 9.60000038 3.90050E-79237 3 18 10.0000000 4.40501E-7 10.5000000 4.95398E-7 11.0000000 5.49559E-79237 3 18 11.5000000 6.04128E-7 12.0000000 6.76534E-7 12.7500000 7.73992E-79237 3 18 13.5000000 8.90463E-7 14.2500000 1.03734E-6 15.0000000 1.27641E-69237 3 18 16.0000000 1.70901E-6 17.0000000 2.53130E-6 18.0000000 4.53423E-69237 3 18 19.0000000 1.29072E-5 20.0000000 .003044283 21.0000000 .0002335999237 3 18 22.0000000 6.21566E-6 23.0000000 2.11257E-6 24.0000000 1.00282E-69237 3 18 25.5000000 6.22895E-7 27.0000000 5.34668E-7 28.0000000 5.45691E-79237 3 18 30.0000000 6.90426E-7 32.0000000 1.16027E-6 34.0000000 4.80306E-69237 3 18 36.0000000 .000321611 38.0000000 6.81239E-7 40.0000000 8.15589E-89237 3 18 42.5000000 7.06497E-8 45.0000000 1.10208E-7 47.5000000 1.58009E-79237 3 18 50.0000000 2.13391E-7 52.5000000 2.84426E-7 55.0000000 3.90089E-79237 3 18 57.5000000 5.84208E-7 60.0000000 1.21992E-6 63.0000000 .0002314589237 3 18 66.0000000 .000328269 69.0000000 4.80554E-7 72.0000000 2.11524E-79237 3 18 76.0000000 5.97862E-7 80.0000000 5.74436E-5 84.0000000 5.16331E-89237 3 18 88.0000000 1.16773E-7 92.0000000 2.17638E-7 96.0000000 5.95206E-79237 3 18 100.000000 7.71072E-5 105.000000 2.38574E-7 110.000000 8.34158E-89237 3 18 115.000000 8.57761E-8 120.000000 1.03492E-7 127.500000 1.24924E-79237 3 18 135.000000 1.47168E-7 142.500000 1.71946E-7 150.000000 2.09340E-79237 3 18 160.000000 2.77760E-7 170.000000 4.55506E-7 180.000000 5.95252E-59237 3 18 190.000000 9.40205E-6 200.000000 .000111774 210.000000 4.93613E-79237 3 18 220.000000 6.74250E-8 230.000000 3.22793E-5 240.000000 6.33722E-89237 3 18 255.000000 8.06979E-8 270.000000 1.08109E-7 280.000000 1.43942E-79237 3 18 300.000000 2.15402E-7 320.000000 5.05216E-7 340.000000 .0001076249237 3 18 360.000000 3.66765E-6 380.000000 8.82725E-8 400.000000 1.65882E-79237 3 18 425.000000 2.95672E-7 450.000000 9.49531E-5 475.000000 1.21865E-59237 3 18 500.000000 5.03088E-5 525.000000 7.41218E-5 550.000000 3.70682E-79237 3 18 575.000000 .000220720 600.000000 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7.59153E-99237 3 18 5000.00000 1.09465E-8 5250.00000 1.87454E-8 5500.00000 9.98495E-59237 3 18 5750.00000 2.07209E-8 6000.00000 1.34150E-8 6300.00000 3.06706E-89237 3 18 6600.00000 7.54352E-8 6900.00000 .000249709 7200.00000 .0004390949237 3 18 7600.00000 .000403901 8000.00000 1.69908E-7 8400.00000 3.85076E-89237 3 18 8800.00000 9.14997E-9 9200.00000 .000117323 9600.00000 2.39228E-89237 3 18 10000.0000 6.11736E-5 10500.0000 6.87096E-5 11000.0000 7.63109E-59237 3 18 11500.0000 8.38604E-5 12000.0000 9.33259E-5 12750.0000 .0001046049237 3 18 13500.0000 .000115961 14250.0000 .000127320 15000.0000 .0001306509237 3 18 16000.0000 .000125950 17000.0000 .000121250 18000.0000 .0001165509237 3 18 19000.0000 .000111850 20000.0000 .000107150 21000.0000 .0001024509237 3 18 22000.0000 9.77501E-5 23000.0000 9.30500E-5 24000.0000 8.70085E-59237 3 18 25500.0000 7.76953E-5 27000.0000 6.94596E-5 28000.0000 5.95473E-59237 3 18 30000.0000 4.60400E-5 32000.0000 3.27058E-5 34000.0000 2.31997E-59237 3 18 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5.028297-6923733102 4.732302-6 4.281294-6 4.081069-6 3.583898-6 3.108881-6 1.671810-6923733102 2.921791-6 5.179343-7 5.179343-7 3.228833-6 3.502100-6 4.376715-6923733102 5.429168-6 6.657624-6 8.720918-6 1.189946-5 7.362235-6 6.140660-6923733102 5.733672-6 5.281981-6 5.181792-6 4.906823-6 4.452004-6 4.239362-6923733102 3.730595-6 3.251817-6 1.751799-6 3.052687-6 4.618064-7 4.618064-7923733102 3.101498-6 3.554217-6 4.180992-6 4.990470-6 5.929547-6 6.708322-6923733102 7.279768-6 1.250699-5 7.214284-6 6.057901-6 5.665926-6 5.554190-6923733102 5.173895-6 4.593276-6 4.355488-6 3.817554-6 3.360511-6 1.809253-6923733102 3.170271-6 4.124357-7 4.124357-7 2.762236-6 3.088725-6 3.571202-6923733102 4.445007-6 5.195247-6 6.309300-6 6.113009-6 7.239937-6 1.216836-5923733102 6.836683-6 5.833379-6 5.606253-6 5.194231-6 4.520603-6 4.285438-6923733102 3.770336-6 3.302310-6 1.771589-6 3.122687-6 2.810233-7 2.810233-7923733102 2.155689-6 2.505228-6 2.936489-6 4.045053-6 4.836048-6 5.855992-6923733102 5.702776-6 5.664378-6 6.012916-6 1.167280-5 7.220186-6 6.264512-6923733102 5.695446-6 4.919647-6 4.676586-6 4.184389-6 3.636112-6 1.925731-6923733102 3.444378-6 1.203301-7 1.203301-7 1.262618-6 1.521675-6 1.824114-6923733102 2.555311-6 2.803257-6 3.856687-6 3.718516-6 4.195658-6 4.391636-6923733102 6.740464-6 1.247719-5 8.932404-6 6.456683-6 5.685773-6 5.465747-6923733102 4.982233-6 4.350095-6 2.231388-6 4.170753-6 6.444943-8 6.444943-8923733102 9.519869-7 1.166006-6 1.423474-6 2.034209-6 2.151504-6 3.242044-6923733102 3.021133-6 3.535045-6 3.502486-6 5.328298-6 6.687729-6 1.212462-5923733102 9.427675-6 6.431210-6 6.282109-6 5.700808-6 4.973386-6 2.534591-6923733102 4.757867-6 923733102 923733 0 9237 0 0 0 0 0 9.32370E+4 2.35011E+2 0 1 34 109346 1451 0.0 1.00000E+0 0 0 0 69346 1451 1.00000E+0 2.00000E+7 0 0 10 20029346 1451 3.00000E+2 0.0 1 0 345 39346 1451 93-Np-237 FEI/LANL EVAL-May96 K.Zolotarev et al.,P.Young et al.9346 1451 DIST-Feb2004 9346 1451 ----IRDF-2002 MATERIAL 9346 9346 1451 -----INCIDENT NEUTRON DATA 9346 1451 ------ENDF-6 FORMAT 9346 1451 ***************************************************************** 9346 1451 93-Np-237 FEI EVAL-May96 K.Zolotarev et al. 9346 1451 DIST-Sep03 9346 1451 ----BROND-3 MATERIAL 9346 Revision 2, Oct. 2003 9346 1451 ***************************************************************** 9346 1451 ------Russian Reactor Dosimetry File RRDF-2002 9346 1451 ***************************************************************** 9346 1451 Authors of evaluation: K.I.Zolotarev, S.A.Badikov, A.I.Blokhin, 9346 1451 A.V.Ignatyuk 9346 1451 ***************************************************************** 9346 1451 9346 1451 ----- MF=2 MT=151 ----- 9346 1451 9346 1451 RESONANCE PARAMETERS 9346 1451 -------------------- 9346 1451 Resolved Resonance Region : 1.0E-5 - 600 eV 9346 1451 Resolved MLBW resonance parameters up to 600 eV were based on 9346 1451 on data from ref. [1]. 9346 1451 A statistical method of the resonance analysis showed that 9346 1451 the missing of the resonances becomes essential above the energy 9346 1451 of 150 eV. However only weak resonances are missed with the neut- 9346 1451 ron widths less than the average width by the factor 5-10. 9346 1451 The resonances with widths close to average or above it were 9346 1451 identified without the noticeable missing up to the energy of 9346 1451 600 eV. So as these resonances give dominant contribution into 9346 1451 the neutron cross sections the resolved resonances region was ex- 9346 1451 panded up to 600 eV and average contribution of the missed weak 9346 1451 resonances was taken into account by addition of relevant cross 9346 1451 section background in the file MF=3. (For fission cross sections 9346 1451 this background is equal to zero). 9346 1451 9346 1451 Unresolved Resonance Region : 600 eV - 6 keV 9346 1451 For the optical-statistical calculation of the cross sections 9346 1451 in the unresolved resonance region were used the following para- 9346 1451 meters: 9346 1451 Average resonance spacing - D0 = 0.57+-0.03 eV 9346 1451 Neutron strengh function - S0 = (0.97+-0.07)*104 9346 1451 Average radiation width - Gg = 40.0+-1.2 meV 9346 1451 This values were obtained as a result of the analysis resolwed 9346 1451 MLBW resonance parameters taking into account the missed resonan- 9346 1451 ces correction. 9346 1451 Considerable attention was paid to the analysis of fission 9346 1451 widths and elimination of contradictions in the description of 9346 1451 intermediate structure of averaged fission cross sections for the 9346 1451 neutron energies above 100 eV. The new experimental data [2-4] on 9346 1451 fission cross sections were included in the analysis. They were 9346 1451 obtained by LANL and JAERI physicists on their neutron spectrome- 9346 1451 ter [2,3] and by Dubna-Obninsk collaboration on the pulsed reac- 9346 1451 tor of JINR [4]. 9346 1451 The resulting set of resonance parameters reproduces well the 9346 1451 observed intermediate structure of fission cross sections (100 eV-9346 1451 6 keV). Corresponding group fission cross sections are 30-50 % 9346 1451 lower than JENDL-3.2 evaluation [5] and approximately twice 9346 1451 higher than ENDF/B-VI ones [6]. Corrected fission widths of the 9346 1451 neutron resonances below 100 eV also result in fission cross sec- 9346 1451 tions which are lower than JENDL-3.2 values [5]. 9346 1451 Calculated from the adopted resonance parameters 2200 M/S 9346 1451 cross sections and resonance integrals are agree rather well with 9346 1451 the recommended evaluated experimental data [1] : 9346 1451 9346 1451 Value This evaluation Experimental data [1] 9346 1451 Total 190.23 barn 9346 1451 Elastic 14.17 barn 9346 1451 Fission 0.0215 barn 0.0215+-0.0024 barn 9346 1451 Capture 176.04 barn 175.9+-2.9 barn 9346 1451 Fission res. int. 6.930 barn 6.9+-1.0 barn 9346 1451 Capture res. int. 642.30 barn 640+-50 barn 9346 1451 9346 1451 ----- MF=3 MT= 18 ----- 9346 1451 9346 1451 FISSION CROSS SECTIONS 9346 1451 ---------------------- 9346 1451 Np-237 fission cross sections from 1.0E-5 eV to 6 keV are re- 9346 1451 constructed from evaluated MLBW resolved and unresolved resonance 9346 1451 parameters. 9346 1451 For the evaluation of the Np-237(n,f) cross sections in the 9346 1451 neutron energy region of 6 keV - 20 MeV available experimental 9346 1451 data [7- 46] were analyzed. During this procedure experimental 9346 1451 data if it was possible were corrected to the new recommended 9346 1451 cross section data for monitor reactions used in the measurements 9346 1451 and to the new recommended decay data. 9346 1451 The top priority was given to absolute measurements where no 9346 1451 reference cross sections were used to determine the neutron flux 9346 1451 and to time-of-flight experiments with simultaneous registration 9346 1451 of the fission and monitoring reaction events. 9346 1451 In ten experiments the Np-237(n,f) cross sections were measu- 9346 1451 red relatively to the fission of U-235. The use as the standard 9346 1451 of the U-235 fission cross section from ENDF/B-VI [47] instead 9346 1451 of the old one (ENDF/B-V) results in decrease of the Np-237(n,f) 9346 1451 cross sections in average by 2 % in 0.1 - 2.0 MeV range and by 9346 1451 1.5 % in 2.0 - 3.0 MeV range. 9346 1451 In the work of Kupriyanov et al. [20] the Np-237(n,f) cross 9346 1451 sections were measured against the fission cross sections of 9346 1451 Pu-239. At the present time there are no recommended Pu-239 cross 9346 1451 sections as a standard data. The data from two libraries were 9346 1451 used to get absolute values: JENDL-3.2 [48] and ENDF/B-VI [49]. 9346 1451 Below 1.6 MeV the results of Ref. [20] disagree with the integral 9346 1451 experiments no matter which data are used for Pu-239(n,f) . The 9346 1451 analysis of Ref. [20] data demonstrated that the good agreement 9346 1451 with the rest of data may be obtained for the energy region of 9346 1451 1.6-7.0 MeV if Pu-239(n,f) reaction cross section data are taken 9346 1451 from JENDL-3.2 library. 9346 1451 The ratio of Np-237 to U-235 fission cross sections measured 9346 1451 by Behrens et al. [23] in 0.11 - 18.89 MeV energy range was multi-9346 1451 plied by factor of 1.051. This normalization factor was obtained 9346 1451 from the values of the functional / in 9346 1451 the 1 - 5 MeV energy range evaluated before. There is a lot of 9346 1451 experimental data which are in good agreement in this interval. 9346 1451 The ratio of the Np-237(n,f) evaluated averaged cross section for 9346 1451 the neutron spectrum of Cf-252 spontaneous fission, known from 9346 1451 many works, to the averaged data of Behrens et al. [23] is equal 9346 1451 to 1.055 that confirms our renormalization of these data. 9346 1451 The analysis of experimental data on the ratio of Np-237 and 9346 1451 U-235 fission cross sections indicates that the relative energy 9346 1451 trends of / measured by Terayama et al.9346 1451 [36] in the energy range of 4.19 - 6.99 MeV and by Goverdovskiy 9346 1451 et al. [32] in the 5.66 - 10.06 MeV energy range coincides with 9346 1451 the results of other authors. To make them agree in absolute 9346 1451 values Terayama et al. and Goverdovskiy et al. results were multi-9346 1451 plied by 0.96 and 1.079, respectively. The data on the Np-237 9346 1451 fission cross sections obtained by Terayama et al. on T(p,n)He-3 9346 1451 neutron source for the 0.70 - 2.99 MeV energy range are in good 9346 1451 agreement with the results of Refs. [38-40]. So they were correc- 9346 1451 ted only according to the new data on the monitoring reaction 9346 1451 U-235(n,f) [46]. 9346 1451 Experimental data of Meadows et al. [26] as well as that of 9346 1451 Kupriyanov et al. [20] are systematically too low below 1 MeV 9346 1451 that contradicts to the evaluated integral experiments 9346 1451 available. Above 1 MeV Meadows et al. [26] data well agree with 9346 1451 the results of other authors so they were included in the final 9346 1451 evaluation only above this energy. Data of Kupriyanov et al. [20] 9346 1451 were taken into account in the evaluation above 1.6 MeV. 9346 1451 Statistical analysis of the input experimental cross section 9346 1451 data for Np-237(n,f) reaction in 0.1-20 MeV neutron energy range 9346 1451 was carried out by means of PADE-2 code [50]. Rational function 9346 1451 was used as the model function [51]. 9346 1451 Averaged cross sections for U-235 thermal fission [52] and 9346 1451 Cf-252 spontaneous fission neutron spectra [53] calculated from 9346 1451 the evaluated Np-237(n,f) excitation function are the following: 9346 1451 9346 1451 ÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÂÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÂÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄ 9346 1451 TYPE OF SPECTRUM ³ ,mb (calc.) ³ , mb (measured) 9346 1451 ÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÅÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÅÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄ 9346 1451 U-235 neutron fission ³ 1356.2 ³ 1353.0 +- 24.0 [54] 9346 1451 ³ ³ 1350.0 +- 24.0 [55] 9346 1451 ÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÅÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÅÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄ 9346 1451 CF-252 spont. fission ³ 1359.9 ³ 1361.0 +- 21.6 [55] 9346 1451 ÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÁÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÁÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄ 9346 1451 9346 1451 Evaluated Np-237 fission cross section is in a good agreement 9346 1451 with the integral experiment data both for U-235 thermal fission 9346 1451 neutron spectrum and for Cf-252 spontaneous fission neutron spec- 9346 1451 trum. 9346 1451 The evaluated Np-237(n,f) cross section agrees within the ac- 9346 1451 curacy with integral experimental data for SIGNA-SIGMA field. 9346 1451 The Np-237 fission cross sections measured in CFRMF facility 9346 1451 are lower than calculated ones for any fission cross section eva- 9346 1451 luation. This apparently indicates the necessity of more detailed 9346 1451 and careful analysis of the CFRMF experimental data accuracy. 9346 1451 9346 1451 ----- MF=33 MT= 18----- 9346 1451 Uncertainties in the evaluated excitation function for the 9346 1451 Np-237(n,f) reaction are given in the two independent matrixes. 9346 1451 In the energy range 1.000E-05 eV - 0.1 MeV uncertainties are 9346 1451 given in the form of diagonal matrix of uncertainties for 6-th 9346 1451 neutron energy intervals (LB=1) 9346 1451 In the energy range 0.1 - 20 MeV uncertainties are presented 9346 1451 in the form of relative covariance matrix for the 48 -neutron 9346 1451 energy groups (LB=5). Eigenvalues of this 6-th digits relative 9346 1451 covariance matrix are the following: 9346 1451 9346 1451 1.49939E-07 1.52531E-07 1.53109E-07 1.53318E-07 9346 1451 1.54049E-07 1.55336E-07 1.56641E-07 1.58051E-07 9346 1451 1.58347E-07 1.60763E-07 1.63743E-07 1.64776E-07 9346 1451 1.67513E-07 1.69820E-07 1.73330E-07 1.75751E-07 9346 1451 1.77557E-07 2.05370E-07 2.66364E-07 2.70615E-07 9346 1451 2.75409E-07 2.80956E-07 2.93138E-07 3.15582E-07 9346 1451 3.35128E-07 4.86134E-07 7.89097E-07 3.07453E-06 9346 1451 1.65989E-05 9.67332E-05 1.24983E-04 1.51845E-04 9346 1451 4.05040E-04 4.14092E-04 5.11273E-04 5.63190E-04 9346 1451 7.24776E-04 8.29295E-04 9.68770E-04 9.77438E-04 9346 1451 1.07709E-03 1.13969E-03 1.57856E-03 1.84372E-03 9346 1451 2.04068E-03 2.65362E-03 3.89983E-03 1.60049E-02 9346 1451 9346 1451 References: 9346 1451 1. S.F.Mughabghab Neutron Cross Sections. N.Y.-London, Academic 9346 1451 Press, 1984, v.1, part B. 9346 1451 2. J.Kimura Nucl. Sci. 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Progress Report NEANDC(J)-122, Sep. 1986 9346 1451 37. I.D.Alkhazov et al. VANT, Serija: Yadernye Konstanty, v.4, 9346 1451 p.19, December 1986 9346 1451 38. V.A.Kalinin et al. VANT, Serija: Yadernye Konstanty, v.4, 9346 1451 p.3, December 1987 9346 1451 39. P.W.Lisowski et al. Proc. of an Int. Conf. on Nuclear Data 9346 1451 for Science and Technology, Mito, Japan, 30 May-3 June 1988, 9346 1451 pp. 97-99 9346 1451 40. J.W.Meadows Annals of Nucl. Energy, v.15, p.421, August 1988 9346 1451 41. J.W.Meadows, D.L.Smith, L.P.Geraldo Annals of Nucl. Energy, 9346 1451 v.16, p.471, September 1989 9346 1451 36. L.Desdin, S.Szegedi, J.Csikai Acta Physica Hungaria, v.65, 9346 1451 p.271, 1989 9346 1451 43. P.W.Lisowski et al. Proc. of the Conference on Fifty Years 9346 1451 with Nuclear Fission, NIST, Gaithersburg, MD, 1989, p.443 9346 1451 44. K.Merla et al. Proc. of an Int. Conf. on Nuclear Data for 9346 1451 Science and Technology, Julich, FRG, 13-17 May 1991. Springer 9346 1451 Verlag, Berlin - Heidelberg, 1992, p.510-513 9346 1451 45. A.D.Carlson et al. Proc. of an Int. Conf. on Nuclear Data for 9346 1451 Science and Technology, Gatlinburg, Tennessee, USA, May 9-13, 9346 1451 1994, Vol.1, pp. 40-42 9346 1451 46. T.Iwasaki et al. Nucl. Sci. Technology, v.36, p.127, 1999 9346 1451 47. L.Weston et al. ORNL,LANL eval. April 1989, ENDF/B-VI Library 9346 1451 MAT=9228; 9346 1451 H.Conde ed. Nuclear Data Standards for Nuclear Measurements, 9346 1451 Report NEANDC-311 U, p.51-58, OECD, Paris, 1992. 9346 1451 48. M.Kawai et al. eval. NAIG, JENDL-3.2 library, MAT=9437, 9346 1451 Rev. 2, February 1993 9346 1451 49. P.Young et al. eval. ORNL, LANL , ENDF/B-VI library, MAT=9437,9346 1451 Rev.1, January 1993 9346 1451 50. S.A.Badikov et al. Preprint FEI-1686, Obninsk, 1985 9346 1451 51. S.Badikov, N.Rabotnov, K.Zolotarev Proc. of NEANSC Speciali- 9346 1451 st's Meeting on Evaluation and Processing of Covariance Data, 9346 1451 Oak Ridge, USA, 1992, OECD, Paris, 1993, p.105 9346 1451 52. L.W.Weston et al. Evaluated Neutron Data for Uranium-235, 9346 1451 ENDF/B-VI Library, MAT=9228, MF=5, MT=18, eval. April 1989 9346 1451 53. W.Mannhart IAEA-TECDOC-410, p.158, IAEA, Vienna, 1987 9346 1451 54. W.Mannhart Prog. Report INDC(Ger)-045, pp.40-43, June 1999 9346 1451 55. W.Mannhart Validation of Differential Cross Sections with 9346 1451 Integral Data , Report INDC(NDS)-435, pp.59-64, IAEA, Vienna, 9346 1451 September 2002 9346 1451 ******************************************************************9346 1451 ----- MF=3 MT= 1 ----- 9346 1451 ----- MF=3 MT= 2 ----- 9346 1451 93-NP-237 LANL EVAL-APR90 P.YOUNG, E.ARTHUR, F.MANN 9346 1451 DIST-SEP91 REV1-JUL91 19910806 9346 1451 ----ENDF/B-VI MATERIAL 9346 REVISION 1 9346 1451 MT=1 Neutron Total Cross Section. Derrien evaluation (De80, 9346 1451 De89) used from 1.0E-5 eV to 0.008 MeV. Above 0.008 MeV,9346 1451 coupled channel optical model calculations (ECIS code) 9346 1451 with modified Lagrange (Ha80) potential used as prior 9346 1451 in a covariance analysis of measurement by Lisowski et 9346 1451 al. from 3 to 20 MeV (Li90). 9346 1451 MT=2 All Energies, based on subtraction of MT=4,16,17,18,37, 9346 1451 and 102 from MT=1. 9346 1451 ******************************************************************9346 1451 The cross sections for total and elastic were modified between 9346 1451 6 and 10 Kev for continuity with the data of Zolatarev file 18. 9346 1451 The number of interpolation schemes is set to 1. 9346 1451 9346 1451 ******************************************************************9346 1451 ***************** Program LINEAR (VERSION 2002-1) ***************9346 1451 For All Data Greater than 1.0000E-10 barns in Absolute Value 9346 1451 Data Linearized to Within an Accuracy of .100000000 per-cent 9346 1451 ***************** Program RECENT (VERSION 2002-1) ***************9346 1451 for All Data Greater than 1.0000E-10 barns in Absolute Value 9346 1451 Data Linearized to within an Accuracy of .100000000 per-cent 9346 1451 ***************** Program SIGMA1 (VERSION 2002-1) ***************9346 1451 Data Doppler Broadened to 300.000000 Kelvin 9346 1451 for All Data Greater than 1.0000E-10 barns in Absolute Value 9346 1451 Data Linearized to Within an Accuracy pf .100000000 per-cent 9346 1451 ***************** Program FIXUP (Version 2002-1) ****************9346 1451 Corrected ZA/AWR in All Sections-----------------------------Yes 9346 1451 Corrected Thresholds-----------------------------------------Yes 9346 1451 Extended Cross Sections to 20 MeV----------------------------No 9346 1451 Allow Cross Section Deletion---------------------------------No 9346 1451 Allow Cross Section Reconstruction---------------------------No 9346 1451 Make All Cross Sections Non-Negative-------------------------Yes 9346 1451 Delete Energies Not in Ascending Order-----------------------Yes 9346 1451 Deleted Duplicate Points-------------------------------------Yes 9346 1451 Check for Ascending MAT/MF/MT Order--------------------------Yes 9346 1451 Check for Legal MF/MT Numbers--------------------------------Yes 9346 1451 Allow Creation of Missing Sections---------------------------No 9346 1451 Allow Insertion of Energy Points-----------------------------No 9346 1451 Create Uniform Energy Grid-----------------------------------No 9346 1451 Delete Section if Cross Section =0 at All Energies-----------Yes 9346 1451 ***************** Program GROUPIE (VERSION 2002-1) **************9346 1451 Unshielded Group Averages Using 640 Groups 9346 1451 Weighting Spectrum: Flat (Constant) Spectrum 9346 1451 1 451 352 19346 1451 3 18 217 19346 1451 33 18 220 19346 1451 9346 1 0 9346 0 0 9.32370E+4 2.35011E+2 0 0 0 09346 3 18 1.96370E+8 1.96370E+8 0 0 1 6419346 3 18 641 1 9346 3 18 .000100000 .350176171 .000105000 .342008374 .000110000 .3342488439346 3 18 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4934633 18 3.56628- 4 3.08074- 4 2.00343- 4 1.00419- 4 5.98365- 5 8.23513- 5934633 18 1.41452- 4 2.08688- 4 2.66476- 4 3.07850- 4 3.32256- 4 3.41989- 4934633 18 3.34886- 4 3.02380- 4 5.57432- 4 4.93285- 4 3.90762- 4 2.86104- 4934633 18 2.07425- 4 1.75261- 4 1.96933- 4 2.55178- 4 3.04268- 4 2.95024- 4934633 18 2.20720- 4 1.28196- 4 7.22802- 5 7.26645- 5 1.15428- 4 1.76292- 4934633 18 2.36288- 4 2.85266- 4 3.19732- 4 3.39905- 4 3.46390- 4 3.26467- 4934633 18 5.20814- 4 4.90773- 4 4.23620- 4 3.40991- 4 2.65742- 4 2.19526- 4934633 18 2.13609- 4 2.35429- 4 2.48687- 4 2.22352- 4 1.62363- 4 1.03524- 4934633 18 7.51805- 5 8.40554- 5 1.20915- 4 1.72526- 4 2.28244- 4 2.81467- 4934633 18 3.28873- 4 3.85829- 4 4.39684- 4 5.44209- 4 5.44137- 4 4.94518- 4934633 18 4.07637- 4 3.07249- 4 2.25930- 4 1.89121- 4 1.91727- 4 1.97704- 4934633 18 1.76050- 4 1.30261- 4 8.65519- 5 6.68813- 5 7.80470- 5 1.15832- 4934633 18 1.72057- 4 2.38893- 4 3.10384- 4 4.17586- 4 5.51601- 4 6.17393- 4934633 18 6.25532- 4 5.63821- 4 4.44969- 4 3.05856- 4 1.97860- 4 1.52979- 4934633 18 1.54588- 4 1.57496- 4 1.36506- 4 1.01383- 4 7.50123- 5 7.28036- 5934633 18 9.84158- 5 1.48103- 4 2.15526- 4 2.94497- 4 4.24091- 4 6.00908- 4934633 18 7.01848- 4 6.99402- 4 6.10456- 4 4.54759- 4 2.85268- 4 1.63429- 4934633 18 1.13789- 4 1.10696- 4 1.15827- 4 1.13531- 4 1.09834- 4 1.15820- 4934633 18 1.37931- 4 1.77038- 4 2.30735- 4 2.95477- 4 4.06168- 4 5.64219- 4934633 18 7.79050- 4 7.69067- 4 6.54107- 4 4.58342- 4 2.51297- 4 1.08549- 4934633 18 5.80237- 5 7.41570- 5 1.16655- 4 1.60384- 4 1.97857- 4 2.30396- 4934633 18 2.61252- 4 2.92868- 4 3.26450- 4 3.81206- 4 4.60124- 4 8.69495- 4934633 18 8.52683- 4 6.90019- 4 4.27097- 4 1.77648- 4 4.11395- 5 3.38196- 5934633 18 1.06327- 4 2.00867- 4 2.82548- 4 3.39481- 4 3.72530- 4 3.87081- 4934633 18 3.88907- 4 3.77309- 4 3.51727- 4 9.64193- 4 9.01880- 4 6.57344- 4934633 18 3.41072- 4 1.07066- 4 3.11372- 5 8.47740- 5 1.96925- 4 3.10075- 4934633 18 3.95536- 4 4.45941- 4 4.64516- 4 4.58009- 4 4.14746- 4 3.25813- 4934633 18 9.79517- 4 8.48395- 4 5.60783- 4 2.68204- 4 9.54045- 5 6.53598- 5934633 18 1.30517- 4 2.32221- 4 3.30598- 4 4.06862- 4 4.56004- 4 4.79820- 4934633 18 4.75834- 4 4.25947- 4 8.85772- 4 7.28861- 4 4.60726- 4 2.13428- 4934633 18 6.81451- 5 3.16448- 5 7.13005- 5 1.50030- 4 2.40988- 4 3.28775- 4934633 18 4.06173- 4 4.96655- 4 5.78294- 4 7.35670- 4 5.68859- 4 3.21582- 4934633 18 1.03067- 4-2.84224- 5-6.78220- 5-3.62829- 5 4.05124- 5 1.41735- 4934633 18 2.53158- 4 4.20532- 4 6.26698- 4 5.22812- 4 3.55656- 4 1.57839- 4934633 18 2.11845- 6-8.22425- 5-9.78357- 5-6.04168- 5 1.27671- 5 1.07588- 4934633 18 2.69172- 4 4.90480- 4 3.05310- 4 2.08198- 4 1.10496- 4 3.95566- 5934633 18 3.42473- 6-1.04504- 6 1.90253- 5 5.63655- 5 1.32607- 4 2.49668- 4934633 18 2.37424- 4 2.42348- 4 2.29411- 4 2.06211- 4 1.78849- 4 1.51149- 4934633 18 1.25140- 4 9.14376- 5 5.55298- 5 3.49455- 4 4.09754- 4 4.23490- 4934633 18 4.00001- 4 3.50544- 4 2.84715- 4 1.70068- 4 9.99255- 6 5.36145- 4934633 18 5.97068- 4 6.00550- 4 5.60150- 4 4.89289- 4 3.47930- 4 1.33595- 4934633 18 7.05793- 4 7.51272- 4 7.45658- 4 7.02138- 4 5.88980- 4 3.96187- 4934633 18 8.46993- 4 8.93652- 4 9.01817- 4 8.60734- 4 7.52757- 4 1.00335- 3934633 18 1.07891- 3 1.14035- 3 1.16290- 3 1.23137- 3 1.41398- 3 1.59674- 3934633 18 1.79587- 3 2.24847- 3 3.07560- 3 934633 18 934633 0 9346 0 0 0 0 0 9.42390E+4 2.36999E+2 0 1 34 109437 1451 0.0 1.00000E+0 0 0 0 69437 1451 1.00000E+0 2.00000E+7 0 0 10 20029437 1451 3.00000E+2 0.0 1 0 564 39437 1451 94-Pu-239 NAIG EVAL-MAR87 M.KAWAI, T.YOSHIDA, K.HIDA 9437 1451 DIST-Feb2004 9437 1451 ----IRDF-2002 MATERIAL 9437 9437 1451 -----INCIDENT NEUTRON DATA 9437 1451 ------ENDF-6 FORMAT 9437 1451 ================================================================= 9437 1451 94-PU-239 NAIG EVAL-MAR87 M.KAWAI, T.YOSHIDA, K.HIDA 9437 1451 DIST-SEP89 REV2-FEB93 9437 1451 ----JENDL-3.2 MATERIAL 9437 9437 1451 ================================================================= 9437 1451 HISTORY 9437 1451 87-03 EVALUATION WAS MADE BY 9437 1451 M.KAWAI AND K.HIDA(NAIG) : CROSS SECTIONS ABOVE 9437 1451 RESONANCE REGION AND OTHER QUANTITIES, 9437 1451 T.YOSHIDA(NAIG) : RESONANCE PARAMETERS AND BACKGROUND 9437 1451 CROSS SECTIONS, 9437 1451 DATA WERE COMPILED BY T.NAKAGAWA (JAERI). 9437 1451 88-08 PARTLY MODIFIED. 9437 1451 NU-BAR, RESOLVED RESONS., (N,2N). 9437 1451 89-02 FP YIELDS WERE TAKEN FROM JNDC FP DECAY DATA FILE VERSION-2.9437 1451 89-03 UNRESOLVED RESONANCE PARAMETERS WERE SLIGHTLY MODIFIED. 9437 1451 93-02 JENDL-3.2. 9437 1451 RESONANCE PARAMETERS EVALUATED BY H.DERRIEN (JAERI)/1/. 9437 1451 FISSION SPECTRA CALCULATED BY T.OHSAWA (KINKI UNIV.) 9437 1451 COMPILED BY T.NAKAGAWA (NDC/JAERI) 9437 1451 9437 1451 ***** MODIFIED PARTS FOR JENDL-3.2 ******************** 9437 1451 (2,151) RESOLVED RESONANCE PARAMETERS UP TO 2.5 KEV 9437 1451 *********************************************************** 9437 1451 9437 1451 9437 1451 MF=1 GENERAL INFORMATION 9437 1451 MT=451 DESCRIPTIVE DATA AND DICTIONARY 9437 1451 9437 1451 MF=2 RESONANCE PARAMETERS 9437 1451 MT=151 RESOLVED AND UNRESOLVED RESONANCE PARAMETERS 9437 1451 RESOLVED RES. PARAMETERS FOR REICH-MOORE FORMULA: UP TO 2.5 KEV9437 1451 PARAMETERS WERE TAKEN FROM REF./1/ DETAILS OF EVALUATION 9437 1451 ARE GIVEN IN APPENDIX. 9437 1451 UNRESOLVED RESONANCES : FROM 2.5 TO 30 KEV. 9437 1451 THE ENERGY DEPENDENT S0, S1 AND FISSION WIDTH WERE DETER- 9437 1451 MINED SO AS TO REPRODUCE THE EVALUATED TOTAL, CAPTURE AND 9437 1451 FISSION CROSS SECTIONS. 9437 1451 9437 1451 2200-M/SEC CROSS SECTIONS AND CALCULATED RESONANCE INTEGRALS. 9437 1451 2200 M/S RES. INTEG. 9437 1451 TOTAL 1025.7 B - 9437 1451 ELASTIC 7.970 B - 9437 1451 FISSION 747.4 B 302.6 B 9437 1451 CAPTURE 270.3 B 181.6 B 9437 1451 9437 1451 MF=3 NEUTRON CROSS SECTIONS 9437 1451 BELOW 2.5 KEV, CROSS SECTIONS WERE REPRESENTED WITH THE 9437 1451 RESOLVED RESONANCE PARAMETERS. BETWEEN 2.5 AND 30 KEV, CROSS9437 1451 SECTIONS WERE REPLACED WITH UNRESOLVED RESONANCE PARAMETERS. 9437 1451 9437 1451 MT=1 TOTAL 9437 1451 BELOW 7 MEV, JENDL-2 EVALUATION WHICH WERE BASED ON THE 9437 1451 EXPERIMENTS OF REFS./10,11,12,13,14/ WAS ADOPTED. ABOVE 9437 1451 7 MEV, EXPERIMENTAL DATA BY POENITZ /15/ WERE ADOPTED. 9437 1451 9437 1451 MT=2 ELASTIC SCATTERING 9437 1451 CALCULATED AS (TOTAL) - (PARTIAL CROSS SECTIONS). 9437 1451 9437 1451 MT=18 FISSION 9437 1451 BELOW 50 KEV 9437 1451 BASED ON MEASUREMENTS OF REF./31/ AND REF./32/. 9437 1451 ABOVE 50 KEV 9437 1451 SIMULTANEOUS EVALUATION WAS PERFORMED BY KANDA ET AL./33/ 9437 1451 9437 1451 MF=33 COVARIANCES 9437 1451 MT=18 9437 1451 TAKEN FROM JENDL/D-99. 9437 1451 9437 1451 REFERENCES 9437 1451 1) DERRIEN H.: J. NUCL. SCI. TECHNOL., 30, 845 (1993). 9437 1451 10) UTTELY C.A.: EANDC(UK)-40 (1964). 9437 1451 11) SCHWARTZ R.B. ET AL.: NUCL. SCI. ENG., 54, 322 (1974). 9437 1451 12) FOSTER D.G.JR. AND GLAGOW D.W.: PHYS. REV., C3, 576 (1971). 9437 1451 13) SMITH A.B. ET AL.: J. NUCL. ENERGY, 27, 317 (1973). 9437 1451 14) NADOLNY ET AL.: C00-3058-39, 33 (1973). 9437 1451 15) POENITZ W.P. ET AL.: NUCL. SCI. ENG., 78, 333 (1981). 9437 1451 31) GAYTHER D.B.: 1975 WASHINGTON, 2, 560 (1975). 9437 1451 32) WAGEMANS C. ET AL.: ANN. NUCL. ENERGY, 7, 495 (1980). 9437 1451 9437 1451 9437 1451 ================================================================= 9437 1451 APPENDIX RESONANCE DATA 9437 1451 ================================================================= 9437 1451 9437 1451 THE PRESENT FILE CONTAINS THE RESONANCE PARAMETERS OBTAINED 9437 1451 FROM A SAMMY FIT ANALYSIS OF HIGH RESOLUTION EXPERIMENTAL DATA, 9437 1451 PERFORMED AT ORNL(OAK RIDGE NATIONNAL LABORATORY,USA) BY H.DERRIEN9437 1451 AND G.DE SAUSSURE AND AT JAERI(TOKAI-MURA RESEARCH ESTABLISHMENT, 9437 1451 JAPAN) BY H.DERRIEN. 9437 1451 THE FILE CONTAINS THREE INDEPENDANT SECTIONS: 9437 1451 1/ THE FIRST CORRESPONDS TO THE ENERGY RANGE 0 KEV TO 1 KEV. 9437 1451 THE CORRESPONDING SET OF RESONANCE PARAMETRES CONTAINS 398 RESO- 9437 1451 NANCES IN THE ENERGY RANGE 0 KEV TO 1 KEV, 4 FICTICIOUS NEGATIVE 9437 1451 ENERGY RESONANCES AND 3 FICTICIOUS RESONANCES ABOVE 1 KEV; 9437 1451 2/ THE SECOND CORRESPONDS TO THE ENERGY RANGE 1 KEV TO 2 KEV. 9437 1451 THE CORRESPONDING SET OF RESONANCE PARAMETERS CONTAINS 435 RESON- 9437 1451 ANCES IN THE ENERGY RANGE 0.980 KEV TO 2.02 KEV, 3 FICTICIOUS 9437 1451 RESONANCES BELOW 0.9 KEV AND 3 FICTICIOUS RESONANCES ABOVE 2.02 9437 1451 KEV; 9437 1451 3/ THE THIRD CORRESPONDS TO THE ENERGY RANGE 2 KEV TO 2.5 KEV. 9437 1451 THE CORRESPONDING SET OF RESONANCE PARAMETERS CONTAINS 218 RESO- 9437 1451 NANCES IN THE ENERGY RANGE 1.98 KEV TO 2.53 KEV, 3 FICTICIOUS 9437 1451 RESONANCES BELOW 1.98 KEV AND 3 FICTICIOUS RESONANCES ABOVE 2.53 9437 1451 KEV. 9437 1451 IN ALL SECTIONS THE FICTICIOUS RESONANCE PARAMETERS TAKE INTO 9437 1451 ACCOUNT THE CONTRIBUTION OF ALL THE EXTERNAL TRUNCATED RESONANCES 9437 1451 IN SUCH A WAY THAT NO TOTAL, SCATTERING, FISSION AND CAPTURE 9437 1451 SMOOTH FILES ARE NEEDED IN THE CORRESPONDING ENERGY RANGES FOR THE9437 1451 REPRODUCTION OF THE CROSS SECTIONS WITHIN THE EXPERIMENTAL ERRORS.9437 1451 THE FOLLOWING EXPERIMENTAL DATA BASE HAS BEEN USED IN THE SAMMY9437 1451 FITS: 9437 1451 -ABSORPTION AND FISSION FROM R.GWIN ET AL./1,2/; 9437 1451 -FISSION FROM R.GWIN ET AL./3,4/, J.BLONS/5/, L.W.WESTON ET 9437 1451 AL./6,7/; 9437 1451 -TRANSMISSION FROM R.R.SPENCER ET AL./8/, J.A.HARVEY ET AL./9/ 9437 1451 PRIOR TO THE FITS THE EXPERIMENTAL FISSION AND ABSORPTION CROSS 9437 1451 SECTIONS WERE NORMALISED, DIRECTLY OR INDIRECTLY TO THE 0.0253 EV 9437 1451 VALUES OBTAINED BY THE ENDF/B-VI STANDARD EVALUATION GROUP/10/. 9437 1451 THE TRANSMISSION DATA WERE CONSIDERED AS ACCURATE ABSOLUTE MEASU- 9437 1451 REMENTS(R.R.SPENCER TOTAL CROSS SECTION AT 0.0253 EV IS 1025.0 B 9437 1451 IN EXCELLENT AGREEMENT WITH THE 1027.3 B STANDARD VALUE). 9437 1451 DETAILS ON THE ANALYSIS ARE FOUND IN REFERENCES/11,12,13/ 9437 1451 9437 1451 ***************************************************************** 9437 1451 COMMENTS ON THE THERMAL AND LOW ENERGY RANGES 9437 1451 ***************************************************************** 9437 1451 9437 1451 THE THERMAL CROSS SECTION VALUES CALCULATED AT 293 K BY THE 9437 1451 RESONANCE PARAMETERS OF THE FIRST SECTION ARE GIVEN IN THE FOLLO- 9437 1451 WING TABLE: 9437 1451 9437 1451 9437 1451 SAMMY RESENDD PROPOSED 9437 1451 293 K (BARN) STANDARD VALUES(BARN)/10/ 9437 1451 **************************************************************** 9437 1451 FISSION 747.64 747.90 747.99+-1.87 9437 1451 CAPTURE 271.10 270.73 271.43+-2.14 9437 1451 SCATTERING 7.97 7.99 7.88+-0.97 9437 1451 TOTAL 1026.71 1026.62 1027.30+-5.00 9437 1451 **************************************************************** 9437 1451 9437 1451 ONE SHOULD NOTE THAT THE 293 K CROSS SECTIONS CALCULATED AT 9437 1451 0.0253 EV DEPEND ON THE WAY THE DOPPLER BROADENING CALCULATION IS 9437 1451 PERFORMED. FOR INSTANCE USING A GAUSSIAN BROADENING FUNCTION WILL 9437 1451 GIVE A FISSION CROSS SECTION ABOUT 2.5 BARNS LARGER THAN THE ONE 9437 1451 OBTAINED FROM THE ACCURATE CALCULATION WHICH CONSERVES THE 1/V 9437 1451 SHAPE OF THE THERMAL CROSS SECTION. THE VALUES GIVEN IN THE TABLE 9437 1451 ABOVE WERE OBTAINED FROM SAMMY (LEAL-HWANG METHOD)/14,15/ AND FROM9437 1451 RESENDD WITH 0.1% FOR THE INTERPOLATION ACCURACY/16/. 9437 1451 THE FOLLOWING TABLE SHOWS EXPERIMENTAL CROSS SECTIONS AVE- 9437 1451 RAGED OVER THE ENERGY RANGES 0.02 EV TO 0.06 EV AND 0.02 EV TO 9437 1451 0.65 EV, COMPARED TO THE CALCULATED VALUES: 9437 1451 9437 1451 9437 1451 AVERAGE CROSS-SECTIONS(BARN) 9437 1451 **************************************************************** 9437 1451 REFERENCES(1-10) 0.02 TO 0.06 EV 0.02 TO 0.65 EV 9437 1451 **************************************************************** 9437 1451 EXP CALC (293K) EXP CALC (293K) 9437 1451 GWIN71 FISS 631.41 843.71 9437 1451 GWIN76 FISS 631.41 838.39 9437 1451 GWIN84 FISS(*) 631.41 631.75(+0.05%) 837.18 838.69(+0.18%) 9437 1451 DERUYTER70 FISS 631.41 859.43 9437 1451 WAGEMANS80 FISS 631.41 862.56 9437 1451 WAGEMANS88 FISS 631.41 841.80 9437 1451 GWIN71 CAPTURE 243.84 243.22(-0.25%) 524.75 518.13(-1.26%) 9437 1451 GWIN76 ABSORPT(*) 875.90 874.29(-0.18%) 1359.96 1357.14(-0.21%) 9437 1451 SPENCER84 TOT(*) 883.20 882.86(-0.04%) 1361.69 1367.6 (+0.43%) 9437 1451 **************************************************************** 9437 1451 (*)THESE DATA HAD THE LARGIEST WEIGHT IN THE THERMAL FIT. THE VA- 9437 1451 LUES BETWEEN THE PARENTHESES GIVE THE PERCENTAGE DEVIATION BETWEEN9437 1451 THE CALCULATED DATA AND THE EXPERIMENTAL DATA. 9437 1451 ***************************************************************** 9437 1451 9437 1451 THE VALUE OF 631.4 BARNS FOR ALL THE AVERAGED EXPERIMENTAL 9437 1451 FISSION CROSS SECTIONS IN THE ENERGY RANGE 0.02 EV TO 0.06 EV 9437 1451 CORRESPONDS TO THE RENORMALISATION OF THE FISSION EXPERIMENTS TO 9437 1451 748.0+-1. BARNS AT 0.0253 EV. ORNL DATA ARE CONSISTENT WITHIN 0.8%9437 1451 OVER THE ENERGY RANGE 0.02 EV TO 0.65 EV (I.E. OVER THE 0.3 EV 9437 1451 RESONANCE). DERUYTER 1970 AND WAGEMANS 1980 DATA ARE ABOUT 2.5% 9437 1451 LARGER AND WERE NOT INCLUDED IN THE SAMMY FIT. WHEN NORMALIZED ON 9437 1451 THE STANDARD VALUE AT 0.0253 EV, GWIN 76 ABSORPTION AGREES WITH 9437 1451 THE ABSORPTION OBTAINED FROM SPENCER TOTAL CROSS SECTION WITHIN 9437 1451 0.7% OVER THE 0.3 EV RESONANCE. THE PRESENT EVALUATION IS 9437 1451 ESSENTIALLY THE RESULT OF A CONSISTENT SAMMY ANALYSIS OF ALL THE 9437 1451 AVAILABLE ORNL DATA WITH A LARGER WEIGHT ON GWIN 1984 FISSION, 9437 1451 GWIN 1976 ABSORPTION AND SPENCER TRANSMISSION DATA. 9437 1451 AFTER RENORMALIZATION OF THE CALCULATED FISSION CROSS SECTION9437 1451 ON THE PRELIMINARY 1991 WESTON AND TODD FISSION DATA(SEE NEXT 9437 1451 SECTION) A SLIGHT ADJUSTMENT OF THE NEGATIVE RESONANCE PARAMETERS 9437 1451 WAS PERFORMED TO KEEP THE VALUES CALCULATED AT 0.0253 EV IN CLOSE 9437 1451 AGREEMENT WITH THE STANDARD VALUES. THE 1988 DATA OF WAGEMANS ET 9437 1451 AL./17/ AGREE WITHIN 0.4% WITH THE CALCULATED VALUES OVER THE 9437 1451 ENERGY RANGE FROM 0.02 EV TO 0.65 EV AFTER ADJUSTMENT OF THE 9437 1451 ENERGY SCALE TO THE ORNL SCALE (THE DIFFERENCE WAS 0.27 EV AT 20 9437 1451 EV BETWEEN 1988 WAGEMANS AND ORNL SAMMY FIT ENERGY SCALES). 9437 1451 9437 1451 ***************************************************************** 9437 1451 COMMENTS ON THE 0 KEV TO 1 KEV ENERGY RANGE. 9437 1451 ***************************************************************** 9437 1451 9437 1451 AT THE END OF 1987, AN ANALYSIS WAS COMPLETED UP TO 1 KEV. IN A9437 1451 PRELIMINARY STEP, A CORRELATED FIT OF HARVEY TRANSMISSION DATA, 9437 1451 WESTON 84 FISSION DATA AND BLONS FISSION DATA WAS PERFORMED, WITH 9437 1451 POSSIBLE ADJUSTMENT OF THE NORMALIZATION COEFFICIENTS AND OF THE 9437 1451 BACKGROUND CORRECTIONS. THIS PRELIMINARY STEP HAS SHOWN THAT THIS 9437 1451 ADJUSTMENT WAS NOT NECESSARY TO HAVE CONSISTENCY BETWEEN HARVEY 9437 1451 DATA AND WESTON DATA. BLONS DATA NEEDED A LARGE READJUSTMENT OF 9437 1451 THE BACKGROUND AND OF THE NORMALIZATION. THEREFORE, THE FINAL FIT 9437 1451 WAS PERFORMED ONLY ON HARVEY TRANSMISSION DATA, GWIN 84 FISSION 9437 1451 DATA (BELOW 30 EV) AND WESTON 84 FISSION DATA, WITH NO BACKGROUND 9437 1451 AND NORMALISATION ADJUSTMENT. BLONS DATA, WHICH HAVE BETTER 9437 1451 RESOLUTION THAN WESTON 84 DATA, WERE USED ONLY TO OBTAIN MORE 9437 1451 ACCURATE FISSION WIDTHS OF SOME NARROW RESONANCES IN THE HIGH 9437 1451 ENERGY RANGE. 9437 1451 IN 1989, PRELIMINARY RESULTS OF THE 1988 WESTON FISSION 9437 1451 MEASUREMENT/7/ WERE INCLUDED IN THE SAMMY EXPERIMENTAL DATA 9437 1451 BASE. ONE EXPECTED FROM THIS MEASUREMENT, WHICH WAS PERFORMED BY 9437 1451 USING A 86 M FLIGHT PATH WITH A RESOLUTION COMPARABLE TO THE 9437 1451 RESOLUTION OF HARVEY TRANSMISSION, A CONFIRMATION OF THE EXCELLENT9437 1451 QUALITY OF THE 1984 MEASUREMENT. A CONSISTENT SAMMY FIT OF HARVEY 9437 1451 TRANSMISSION, WESTON 84 FISSION AND PRELIMINARY WESTON 88 FISSION 9437 1451 WAS RESTARTED FROM THE PARAMETER AND COVARIANCE FILES OBTAINED IN 9437 1451 1987. IT APPEARED THAT LARGE BACKGROUND AND NORMALISATION 9437 1451 CORRECTIONS WERE NEEDED ON THE NEW WESTON FISSION TO OBTAIN 9437 1451 CONSISTENCY WITH HARVEY TRANSMISSION DATA. THESE CORRECTIONS WERE 9437 1451 COMPARABLE TO THOSE FOUND ON BLONS DATA AND WERE NOT UNDERSTOOD BY9437 1451 THE AUTHORS OF THE EXPERIMENT. THE LAST SAMMY RUNS WERE PERFORMED 9437 1451 BY NOT ALLOWING BACKGROUND AND NORMALIZATION VARIATIONS ON HARVEY 9437 1451 TRANSMISSION AND WESTON 84 FISSION(VERY SMALL ERROR BARS WERE 9437 1451 ASSIGNED TO THE CORRESPONDING PARAMETERS IN THE COVARIANCE MATRIX)9437 1451 AND BY ALLOWING THESE VARIATIONS ON WESTON 88 DATA. A NEW SET OF 9437 1451 RESONANCE PARAMETERS WAS OBTAINED,WHICH WAS IMPROVED COMPARED TO 9437 1451 THE PREVIOUS SET DUE TO THE VERY HIGH RESOLUTION OF THE NEW WESTON9437 1451 FISSION MEASUREMENT. 9437 1451 THE CALCULATED AVERAGE FISSION CROSS SECTION IN THE ENERGY 9437 1451 RANGE FROM 0.1 KEV TO 1.0 KEV WAS 3.7% SMALLER THAN THE VALUES 9437 1451 OBTAINED BY THE ENDF/B-VI STANDARD EVALUATION GROUP, DUE TO THE 9437 1451 FACT THAT WESTON 84 DATA WERE 3.1% LOWER THAN THE AVERAGE STANDARD9437 1451 VALUE. A NEW MEASUREMENT WAS PERFORMED BY WESTON AND TODD IN 1991 9437 1451 /18/ IN ORDER TO CHECK THEIR 1984 DATA. A CAREFUL NORMALIZATION 9437 1451 OF THE DATA IN THE THERMAL ENERGY RANGE SHOWED THAT THE 1984 DATA 9437 1451 SHOULD BE RENORMALIZED BY ABOUT +3%. TO TAKE INTO ACCOUNT THIS 9437 1451 RENORMALIZATION, THE 1989 RESONANCE PARAMETERS WERE MODIFIED AT 9437 1451 JAERI/13/ IN THE FOLLOWING WAY: 9437 1451 1/ INCREASE OF THE FISSION WIDTH BY 3% AND DECREASE OF THE 9437 1451 CAPTURE WIDTH BY A QUANTITY EQUAL TO THE VARIATION OF THE FISSION 9437 1451 WIDTH, IN THE NARROW RESONANCES(MAINLY 1+ RESONANCES); THAT DOES 9437 1451 NOT MODIFY THE TOTAL CROSS SECTION IN THE CORRESPONDING 9437 1451 RESONANCES; 9437 1451 2/ ADJUSTMENT OF THE NEUTRON WIDTH OF THE 0+ RESONANCES BY A 9437 1451 REFIT OF THE TRANSMISSION DATA AND OF THE RENORMALIZED WESTON AND 9437 1451 TODD 1984 DATA IN ENERGY RANGES WHERE THE CONTRIBUTION OF THE 0+ 9437 1451 RESONANCES IS DOMINANT, AND INCREASE OF THE OTHER(SMALL) 0+ 9437 1451 NEUTRON WIDTHS BY 3%. NO SEVERE INCONSISTENCY WAS OBSERVED BETWEEN9437 1451 THE TRANSMISSION DATA AND THE NEW FISSION DATA OVER THE DOMINANT 9437 1451 0+ RESONANCES;THE DIFFERENCES BETWEEN THE 1989 FITS OF THE TRANS- 9437 1451 MISSION AND THE NEW FITS WERE CONSISTENT WITHIN THE EXPERIMENTAL 9437 1451 ERROR BARS. 9437 1451 THE FOLLOWING TABLE SHOWS THE FISSION CROSS SECTIONS CALCULA- 9437 1451 TED FROM THE RESONANCE PARAMETERS, THE EXPERIMENTAL VALUES AND THE9437 1451 RESULTS OF THE ENDF/B-VI STANDARD EVALUATION GROUP AVERAGED IN THE9437 1451 SAME ENERGY INTERVALS. WESTON 1991 DATA ARE PRELIMINARY. WESTON 9437 1451 1984 DATA ARE NORMALIZED ON PRELIMINARY WESTON 1991: 9437 1451 9437 1451 CROSS-SECTIONS(BARN) 9437 1451 ************************************************* 9437 1451 ENERGY CALCUL WESTON WESTON STANDARD 9437 1451 (EV) 1991 1984 9437 1451 ************************************************* 9437 1451 0.010-10. 80.12 79.98 9437 1451 9-20 94.74 94.91 9437 1451 20-40 17.52 17.76 17.97 9437 1451 40-60 50.64 50.90 50.87 9437 1451 60-100 54.42 54.38 54.33 9437 1451 100-200 18.63 18.59 18.56 18.66 9437 1451 200-300 17.85 17.89 17.88 9437 1451 300-400 8.31 8.34 8.43 9437 1451 400-500 9.59 9.58 9.57 9437 1451 ------------------------------------------------- 9437 1451 200-500 11.92 11.93 11.93 11.96 9437 1451 ------------------------------------------------- 9437 1451 500-600 15.39 15.57 15.86 9437 1451 600-700 4.37 4.30 4.46 9437 1451 700-800 5.51 5.53 5.63 9437 1451 800-900 4.84 4.89 4.98 9437 1451 900-1000 8.33 8.38 8.30 9437 1451 ------------------------------------------------- 9437 1451 500-1000 7.69 7.73 7.73 7.79 9437 1451 ------------------------------------------------- 9437 1451 20-1000 13.09 13.11 13.11 9437 1451 ************************************************** 9437 1451 9437 1451 9437 1451 GWIN 1971 AND 1976 ABSORPTION DATA WERE NOT INCLUDED IN THE 9437 1451 SAMMY FIT IN THE ENERGY RANGE ABOVE 1 EV. ACCURATE ABSORPTION 9437 1451 CROSS SECTIONS SHOULD BE CALCULATED FROM THE PARAMETERS OBTAINED 9437 1451 FROM THE ANALYSIS OF THE TRANSMISSION AND FISSION DATA. THE FOLLO-9437 1451 WING TABLE SHOWS THE CALCULATED AVERAGE VALUES OF THE CAPTURE, AB-9437 1451 SORPTION AND ALPHA COMPARED TO GWIN 1971 AND GWIN 1976 DATA. THE 9437 1451 CALCULATIONS WERE PERFORMED WITH RESENDD, 1.0 % ACCURACY: 9437 1451 9437 1451 CROSS-SECTIONS(BARN) 9437 1451 ******************************************************* 9437 1451 ENERGY(EV) CALC. VALUES (293K) GWIN DATA 9437 1451 ******************************************************* 9437 1451 CAPT ABSORP ALPHA ABSORP ALPHA 9437 1451 7.3- 16.0 76.61 196.04 0.64 208.00 0.74(*) 9437 1451 16.0- 37.5 20.51 44.55 0.85 46.50 0.89(*) 9437 1451 37.5- 50.0 48.72 70.00 2.29 83.15 2.96(*) 9437 1451 50.0-100.0 33.60 92.13 0.57 92.84 0.63 9437 1451 100.0-200.0 15.58 34.29 0.83 33.66 0.87 9437 1451 200.0-300.0 15.85 33.68 0.89 34.69 0.94 9437 1451 300.0-400.0 9.69 18.01 1.16 18.31 1.16 9437 1451 400.0-500.0 3.96 13.56 0.41 13.56 0.44 9437 1451 500.0-600.0 10.87 26.30 0.70 26.54 0.72 9437 1451 600.0-700.0 6.53 10.90 1.49 11.57 1.54 9437 1451 700.0-800.0 4.95 10.47 0.90 10.52 0.97 9437 1451 800.0-900.0 3.65 8.50 0.75 9.30 0.82 9437 1451 900.0-999.9 5.06 13.51 0.60 13.23 0.70 9437 1451 ****************************************************** 9437 1451 (*) GWIN 1971 DATA 9437 1451 9437 1451 IF ONE EXCEPTS THE ENERGY RANGE 37.5-50 EV, THE CALCULATED AB- 9437 1451 SORPTION VALUES AGREE WELL WITH GWIN EXPERIMENTAL DATA; THEY ARE 9437 1451 ON AVERAGE 1.2% LOWER IN THE ENERGY RANGE FROM 50 EV TO 1000 EV. 9437 1451 9437 1451 ***************************************************************** 9437 1451 COMMENTS ON THE 1 KEV TO 2 KEV ENERGY RANGE 9437 1451 ***************************************************************** 9437 1451 9437 1451 PRELIMINARY RESONANCE PARAMETERS WERE OBTAINED IN 1989 FROM THE9437 1451 ANALYSIS OF THE HARVEY THICK SAMPLE TRANSMISSION DATA AND OF THE 9437 1451 PRELIMINARY RESULTS OF WESTON 88 FISSION MEASUREMENT. DUE TO LACK9437 1451 OF TIME, THE MEDIUM AND THIN SAMPLE TRANSMISSION DATA WERE NOT 9437 1451 INCLUDED IN THE SAMMY DATA BASE, AND THE CONTRIBUTION OF THE 9437 1451 TRUNCATED EXTERNAL RESONANCES WAS NOT CAREFULLY INVESTIGATED. 9437 1451 NEVERTHELESS, THE RESULTS WERE USED IN THE ENDF/B-VI FILE, ALONG 9437 1451 WITH A SMOOTH FILE IN ORDER TO AGREE WITH THE AVERAGE VALUES OF A 9437 1451 PREVIOUS ENDF/B-VI EVALUATION (THIS PRELIMINARY SET OF PARAMETERS 9437 1451 WAS CONSIDERED AS MORE USEFUL THAN THE STATISTICAL PARAMETERS IN 9437 1451 THE ENERGY RANGE 1 KEV TO 2 KEV FOR THE CALCULATION OF THE SELF- 9437 1451 SHIELDING FACTORS). 9437 1451 THE ANALYSIS WAS RESTARTED IN APRIL 1991 AT JAERI(TOKAI 9437 1451 RESEARCH ESTABLISHMENT) WITH AN UPDATED VERSION OF SAMMY ADAPTED 9437 1451 BY T.NAKAGAWA TO THE FACOM 780. THE PRELIMINARY SET OF PARAMETERS 9437 1451 OBTAINED AT OAK RIDGE IN 1989 WAS USED AS PRIOR INFORMATIONS TO 9437 1451 START THE SAMMY CALCULATIONS. ALSO PRIOR TO THE ANALYSIS, THE CON-9437 1451 TRIBUTION OF THE EXTERNAL RESONANCES WAS CALCULATED BY USING THE 9437 1451 SET OF THE 0 KEV TO 1 KEV KNOWN RESONANCES, SHIFTED IN THE ENERGY 9437 1451 RANGES -1 KEV TO 0 KEV, 2 KEV TO 3KEV AND 3 KEV TO 4 KEV; EQUIVA- 9437 1451 LENT CONTRIBUTION WAS OBTAINED BY USING 3 FICTICIOUS RESONANCES 9437 1451 BELOW 1 KEV AND 3 FICTICIOUS RESONANCES ABOVE 2 KEV(SEE DETAILS IN9437 1451 REF./13/). THE ANALYSIS WAS PERFORMED ON THE THICK AND MEDIUM 9437 1451 SAMPLE TRANSMISSIONS OF HARVEY DATA (THE THIN SAMPLE DATA WAS NOT 9437 1451 USEFUL IN THE HIGH ENERGY RANGE) AND ON THE 1988 FISSION DATA RE- 9437 1451 LEASED BY WESTON AT THE BEGINNING OF 1991/7/. THE DEFINITIVE 9437 1451 SAMMY FITS WERE PERFORMED IN APRIL 1992 AFTER RENORMALIZATION OF 9437 1451 THE 1988 DATA OF WESTON ON THE ENDF/B-VI STANDARD VALUES BETWEEN 19437 1451 KEV AND 2 KEV, IN AGREEMENT WITH THE 1991 NEW MEASUREMENTS OF 9437 1451 WESTON AND TODD. 9437 1451 THE AVERAGE CROSS SECTIONS CALCULATED FROM THE RESONANCE 9437 1451 PARAMETERS ARE COMPARED TO THE EXPERIMENTAL VALUES IN THE FOLLO- 9437 1451 WING TABLE: 9437 1451 9437 1451 9437 1451 ************************************************************** 9437 1451 CROSS-SECTIONS(BARN) 9437 1451 --------------------------------------------------- 9437 1451 TOTAL FISSION CAPTURE 9437 1451 ENERGY -------------- ------------- ------------- 9437 1451 KEV CALC(A) EXP(B) CALC(A) EXP(C) CALC(A) EXP(D) 9437 1451 ************************************************************** 9437 1451 1.0-1.1 24.47 24.95 5.549 5.581 4.728 5.04 9437 1451 1.1-1.2 22.82 23.10 5.985 6.017 3.757 2.95 9437 1451 1.2-1.3 22.29 22.90 4.601 4.501 4.287 4.00 9437 1451 1.3-1.4 22.63 22.85 6.997 6.997 3.012 2.52 9437 1451 1.4-1.5 20.42 20.95 4.041 4.059 3.450 3.57 9437 1451 1.5-1.6 18.30 18.95 2.564 2.613 3.521 3.89 9437 1451 1.6-1.7 21.82 21.90 3.952 3.955 3.833 4.36 9437 1451 1.7-1.8 21.26 21.35 3.400 3.425 4.091 4.37 9437 1451 1.8-1.9 23.76 23.30 5.178 5.187 3.639 3.14 9437 1451 1.9-2.0 18.48 18.90 2.152 2.180 3.205 4.06 9437 1451 ************************************************************** 9437 1451 1.0-2.0 21.63 21.92 4.442 4.446 3.752 3.79 9437 1451 ************************************************************** 9437 1451 (A) TOTAL, FISSION AND CAPTURE CROSS SECTIONS CALCULATED BY 9437 1451 RESENDD FROM THE RESONANCE PARAMETERS. 9437 1451 (B) EXPERIMENTAL TOTAL CROSS SECTIONS FROM REFERENCE/19/. 9437 1451 (C) WESTON AND TODD 1988 HIGH RESOLUTION FISSION CROSS SECTIONS 9437 1451 FROM REFERENCE/7/ NORMALIZED TO ENDF/B-VI STANDARD IN THE 9437 1451 ENERGY RANGE FROM 1.0 KEV TO 2.0 KEV. 9437 1451 (D) GWIN 1971 EXPERIMENTAL DATA NORMALIZED TO GWIN 1976 DATA. 9437 1451 ************************************************************** 9437 1451 9437 1451 THE DIFFERENCE OF 1.3% BETWEEN THE AVERAGE CALCULATED TOTAL 9437 1451 CROSS SECTION AND THE AVERAGE EXPERIMENTAL CROSS SECTION IN THE 9437 1451 ENERGY RANGE FROM 1.0 KEV AND 2.0 KEV IS MAINLY DUE TO THE METHOD 9437 1451 OF EVALUATING THE TOTAL CROSS SECTION FROM THE EFFECTIVE CROSS 9437 1451 SECTION IN REFERENCE/19/. THE ACCURACY OF THE SAMMY FIT OF THE 9437 1451 EXPERIMENTAL TRANSMISSION DATA IS BETTER THAN 0.5% ON THE CROSS 9437 1451 SECTION. THE CALCULATED FISSION CROSS SECTIONS ARE IN VERY GOOD 9437 1451 AGREEMENT WITH THE EXPERIMENTAL DATA. THE CAPTURE DATA /1/ ARE 9437 1451 AVERAGE VALUES OBTAINED FROM THE DATA AVAILABLE IN THE EXFOR FILE 9437 1451 AND NORMALIZED TO GWIN 1976 AVERAGE VALUES; THERE ARE LARGE 9437 1451 DIFFERENCES BETWEEN THE CALCULATED DATA AND THE EXPERIMENTAL DATA 9437 1451 AVERAGED OVER 0.1 KEV INTERVALS; BUT ON THE INTERVAL FROM 1.0 KEV 9437 1451 TO 2.0 KEV THE AVERAGE VALUES ARE CONSISTENT WITHIN 1.0%. 9437 1451 9437 1451 ***************************************************************** 9437 1451 COMMENTS ON THE 2.0 KEV TO 2.5 KEV REGION 9437 1451 ***************************************************************** 9437 1451 9437 1451 THIS ENERGY RANGE WAS ALSO ANALYSED AT JAERI /13/. NO 9437 1451 PRELIMINARY SET OF RESONANCE PARAMETERS WAS AVAILABLE PRIOR TO THE9437 1451 ANALYSIS. MORE THAN 90% OF THE RESONANCES, COMPARED TO THE LOW 9437 1451 ENERGY RANGE, COULD STILL BE IDENTIFIED IN THE TRANSMISSION DATA 9437 1451 BETWEEN 2 KEV AND 2.5 KEV. THEREFORE THE CORRELATED SAMMY ANALYSIS9437 1451 OF HARVEY TRANSMISSIONS AND WESTON FISSION WAS STILL FEASIBLE IN 9437 1451 THIS ENERGY RANGE. THE RESONANCE PARAMETERS OBTAINED ARE 9437 1451 CONSISTENT AND HAS NEARLY THE SAME STATISTICAL PROPERTIES AS THOSE9437 1451 OF THE RESONANCES IN THE 0 TO 2 KEV ENERGY RANGE. A QUITE GOOD FIT9437 1451 OF THE TRANSMISSION AND FISSION DATA WAS OBTAINED WITHOUT 9437 1451 BACKGROUND AND NORMALISATION ADJUSTMENT. HOWEVER, THE CALCULATED 9437 1451 FISSION CROSS SECTIONS ARE, ON AVERAGE, 1.4% LOWER THAN THE 9437 1451 EXPERIMENTAL VALUES. THIS DIFFERENCE,WHICH HOWEVER IS NOT LARGER 9437 1451 THAN THE SYSTEMATIC ERRORS ON THE EXPERIMENTAL DATA, COULD BE DUE 9437 1451 TO THE DIFFICULTIES OF IDENTIFYING THE WIDE J=0+ RESONANCES IN THE9437 1451 EXPERIMENTAL DATA, BECAUSE THE EFFECTS OF THE INCREASING 9437 1451 RESOLUTION AND DOPPLER WIDTHS. PRIOR TO THE SAMMY FITS, THE 9437 1451 FISSION DATA OF WESTON AND TODD (1988 HIGH RESOLUTION DATA) WERE 9437 1451 NORMALIZED TO THE ENDF/B-VI STANDARD IN THE ENERGY RANGE FROM 1 9437 1451 KEV TO 2 KEV. 9437 1451 THE CROSS SECTIONS,CALCULATED FROM THE RESONANCE PARAMETERS 9437 1451 AND AVERAGED OVER 0.1 KEV INTERVALS,ARE GIVEN IN THE FOLLOWING 9437 1451 TABLE: 9437 1451 9437 1451 ************************************************************** 9437 1451 CR0SS-SECTIONS(BARN) 9437 1451 ------------------------------------------ 9437 1451 TOTAL FISSION CAPTURE 9437 1451 ENERGY -------------- -------------- ------- 9437 1451 (KEV) CALC(A) EXP(B) CALC(A) EXP(C) CALC(A) 9437 1451 ************************************************************** 9437 1451 2.0-2.1 17.34 17.30 2.034 2.062 3.223 9437 1451 2.1-2.2 20.27 19.80 2.949 2.999 4.051 9437 1451 2.2-2.3 19.34 19.10 2.357 2.393 3.324 9437 1451 2.3-2.4 21.28 21.20 3.646 3.679 3.640 9437 1451 2.4-2.5 20.03 20.60 3.956 4.024 3.128 9437 1451 ************************************************************** 9437 1451 2.0-2.5 19.65 19.60 2.989 3.031 3.473 9437 1451 ************************************************************** 9437 1451 (A) TOTAL, FISSION AND CAPTURE CROSS SECTIONS CALCULATED BY 9437 1451 RESENDD, 1% ACCURACY, AT 300 K,FROM THE RESONANCE PARAME- 9437 1451 TERS. 9437 1451 (B) AVERAGE TOTAL CROSS SECTIONS OBTAINED FROM THE AVERAGE 9437 1451 EXPERIMENTAL EFFECTIVE CROSS SECTIONS IN REFERENCE/19/. 9437 1451 (C) 1988 HIGH RESOLUTION DATA OF WESTON AND TODD NORMALIZED 9437 1451 TO ENDF/B-VI STANDARD IN THE ENERGY RANGE FROM 1 KEV TO 9437 1451 2 KEV. 9437 1451 ***************************************************************** 9437 1451 9437 1451 9437 1451 ***************************************************************** 9437 1451 FISSION AND CAPTURE RESONANCE INTEGRALS 9437 1451 ***************************************************************** 9437 1451 9437 1451 THE FISSION AND CAPTURE RESONANCE INTEGRALS ARE COMPARED TO 9437 1451 JENDL-3 DATA IN THE FOLLOWING TABLE: 9437 1451 9437 1451 ********************************************************** 9437 1451 ENERGY RANGE(EV) FISSION(BARN) CAPTURE(BARN) 9437 1451 ********************************************************** 9437 1451 JENDL-3 PRESENT JENDL-3 PRESENT 9437 1451 0.5 - 5.0 85.725 84.879 28.651 28.723 9437 1451 5.0 - 10.0 25.081 25.147 19.059 18.950 9437 1451 10.0 - 50.0 96.856 99.715 77.181 74.686 9437 1451 50.0 - 100.0 40.479 41.552 25.930 25.376 9437 1451 100.0 - 301.0 19.677 20.252 17.952 17.729 9437 1451 301.0 -1000.0 10.047 10.317 8.348 8.418 9437 1451 1000.0 -2000.0 3.484 3.206 2.840 2.634 9437 1451 2000.0 -2.E+07 17.783 (17.783) 5.224 (5.224) 9437 1451 ********************************************************** 9437 1451 TOTAL 299.132 302.851 185.185 181.739 9437 1451 ********************************************************** 9437 1451 9437 1451 THE JENDL-3 RESONANCE PARAMETERS ARE THOSE OBTAINED IN 1987 IN 9437 1451 THE ENERGY RANGE 0 KEV TO 1 KEV. THEY ARE SLIGTHLY DIFFERENT FROM 9437 1451 THOSE PUBLISHED IN 1989. THAT EXPLAINS THE SMALL DIFFERENCES OB- 9437 1451 SERVED BETWEEN JENDL-3 AND THE PRESENT RESULTS IN THIS ENERGY RAN-9437 1451 GE. IN THE ENERGY RANGE 1 KEV TO 2 KEV JENDL-3 IS UNRESOLVED 9437 1451 RANGE. THE FISSION AND CAPTURE RESONANCE INTEGRALS CALCULATED 9437 1451 FROM ENDF/B-V AND THOSE FOUND IN BNL-325 ARE THE FOLLOWING: 9437 1451 9437 1451 ENDF/B-V FISSION: 302.13 B CAPTURE: 194.10 B 9437 1451 BNL-325 FISSION: 310+-10 B CAPTURE: 200+-20 B 9437 1451 9437 1451 THE CONSEQUENCE OF CHANGING FROM THE OLD SETS OF RESONANCE 9437 1451 PARAMETERS(ENDF/B-V AND PREVIOUS SETS) TO THE NEW SET IS THAT 9437 1451 THE CAPTURE RESONANCE INTEGRAL WILL DECREASE BY 6.7% COMPARED 9437 1451 WITH ENDF/B-V VALUE. 9437 1451 9437 1451 REFERENCES OF APPENDIX 9437 1451 1) R.GWIN ET AL.,NUCL.SCI.ENG.,45,25(1971) 9437 1451 2) R.GWIN ET AL.,NUCL.SCI.ENG.,59,79(1976) 9437 1451 3) R.GWIN ET AL.,NUCL.SCI.ENG.,61,116(1976) 9437 1451 4) R.GWIN ET AL.,NUCL.SCI.ENG.,88,37(1984) 9437 1451 5) J.BLONS, NUCL.SCI.ENG.,51,130(1973) 9437 1451 6) L.W.WESTON ET AL.,NUCL.SCI.ENG.88,567(1984) 9437 1451 7) L.W.WESTON ET AL.,TO BE PUBLISHED(HIGH RESOLUTION 1988 DATA) 9437 1451 8) R.R.SPENCER ET AL.,NUCL.SCI.ENG.,96,318(1987) 9437 1451 9) J.A.HARVEY ,MITO 1988,PAGE 115 9437 1451 10) A.CARLSON ET AL.,PRELIMINARY RESULTS OF THE ENDF/B-6 STANDARD 9437 1451 EVALUATION(SEPT 8 1987) 9437 1451 11) H.DERRIEN AND G. DE SAUSSURE,ORNL-TM-10986(1988) 9437 1451 12) H.DERRIEN ET AL.,NUCL.SCI.ENG.,106,434(1990) 9437 1451 13) H.DERRIEN, J.NUCL>SCI.TECHNOL.,30,845(1993). 9437 1451 14) N.M.LARSON ET AL.,ORNL/TM-7485,ORNL/TM-9179,ORNL/TM-9719/R1 9437 1451 15) L.LEAL AND R.N.HWANG,TRANS.AM.NUC.SOC.,55,340(1987) 9437 1451 16) T.NAKAGAWA,RESENDD A JAERI VERSION OF RESEND,JAER=-M 84-192 9437 1451 (1984). 9437 1451 17) C.WAGEMANS ET AL.,MITO 1988,PAGE 91 9437 1451 18) L.W.WESTON,PRIVATE COMMUNICATION(1992) 9437 1451 19) H.DERRIEN,J.NUCL.SCI.TECHNOL.,29,794(1992). 9437 1451 ********************************************************** 9437 1451 9437 1451 9437 1451 9437 1451 9437 1451 ********************************************************** 9437 1451 ***************** Program LINEAR (VERSION 2002-1) ***************9437 1451 For All Data Greater than 1.0000E-10 barns in Absolute Value 9437 1451 Data Linearized to Within an Accuracy of .100000000 per-cent 9437 1451 ***************** Program RECENT (VERSION 2002-1) ***************9437 1451 for All Data Greater than 1.0000E-10 barns in Absolute Value 9437 1451 Data Linearized to within an Accuracy of .100000000 per-cent 9437 1451 ***************** Program SIGMA1 (VERSION 2002-1) ***************9437 1451 Data Doppler Broadened to 300.000000 Kelvin 9437 1451 for All Data Greater than 1.0000E-10 barns in Absolute Value 9437 1451 Data Linearized to Within an Accuracy pf .100000000 per-cent 9437 1451 ***************** Program FIXUP (Version 2002-1) ****************9437 1451 Corrected ZA/AWR in All Sections-----------------------------Yes 9437 1451 Corrected Thresholds-----------------------------------------Yes 9437 1451 Extended Cross Sections to 20 MeV----------------------------No 9437 1451 Allow Cross Section Deletion---------------------------------No 9437 1451 Allow Cross Section Reconstruction---------------------------No 9437 1451 Make All Cross Sections Non-Negative-------------------------Yes 9437 1451 Delete Energies Not in Ascending Order-----------------------Yes 9437 1451 Deleted Duplicate Points-------------------------------------Yes 9437 1451 Check for Ascending MAT/MF/MT Order--------------------------Yes 9437 1451 Check for Legal MF/MT Numbers--------------------------------Yes 9437 1451 Allow Creation of Missing Sections---------------------------No 9437 1451 Allow Insertion of Energy Points-----------------------------No 9437 1451 Create Uniform Energy Grid-----------------------------------No 9437 1451 Delete Section if Cross Section =0 at All Energies-----------Yes 9437 1451 ***************** Program GROUPIE (VERSION 2002-1) **************9437 1451 Unshielded Group Averages Using 640 Groups 9437 1451 Weighting Spectrum: Flat (Constant) Spectrum 9437 1451 1 451 571 19437 1451 3 18 217 19437 1451 33 18 18 09437 1451 9437 1 0 9437 0 0 9.42390E+4 2.36999E+2 0 0 0 09437 3 18 1.99219E+8 1.99219E+8 0 0 1 6419437 3 18 641 1 9437 3 18 .000100000 11417.9620 .000105000 11151.7657 .000110000 10898.87749437 3 18 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0.000000+0943733 18 2.560000-4 0.000000+0 0.000000+0 943733 18 943733 0 9437 0 0 0 0 0 9.52410E+4 2.38986E+2 0 1 34 109543 1451 0.0 1.00000E+0 0 0 0 69543 1451 1.00000E+0 2.00000E+7 0 0 10 20029543 1451 3.00000E+2 0.0 1 0 110 39543 1451 95-Am-241 JAERI EVAL-MAR88 T.NAKAGAWA 9543 1451 DIST-Feb2004 9543 1451 ----IRDF-2002 MATERIAL 9543 9543 1451 -----INCIDENT NEUTRON DATA 9543 1451 ------ENDF-6 FORMAT 9543 1451 **************************************************************** 9543 1451 95-AM-241 JAERI EVAL-MAR88 T.NAKAGAWA 9543 1451 JAERI-M 89-008 DIST-SEP89 9543 1451 ----JENDL-3.2 MATERIAL 9543 9543 1451 9543 1451 AM-241 FISSION 9543 1451 **************************************************************** 9543 1451 HISTORY 9543 1451 82-03 EVALUATION FOR JENDL-2 WAS MADE BY Y.KIKUCHI (JAERI) /1/. 9543 1451 88-03 RE-EVALUATION FOR JENDL-3 WAS MADE BY T.NAKAGAWA (JAERI) 9543 1451 /2/. 9543 1451 9543 1451 MF=1 GENERAL INFORMATION 9543 1451 MT=451 COMMENT AND DICTIONARY 9543 1451 9543 1451 MF=2,MT=151 RESONANCE PARAMETERS 9543 1451 RESOLVED RESONANCES FOR MLBW FORMULA : 1.0E-5 - 150 EV 9543 1451 DATA OF DERRIEN AND LUCAS /5/ WERE ADOPTED AND 5 NEGATIVE 9543 1451 RESONANCES WERE ADDED. VALUES OF TOTAL SPIN J WERE 9543 1451 REPLACED WITH ARBITRARILY ASSUMED VALUES. 9543 1451 9543 1451 UNRESOLVED RESONANCES : 150 EV - 30 KEV 9543 1451 PARAMETERS WERE DETERMINED BY USING ASREP/6/ SO AS TO 9543 1451 REPRODUCE THE CAPTURE CROSS SECTION MEASURED BY VANPRAET 9543 1451 ET AL. /7/ AND THE FISSION CROSS SECTION BY DABBS ET AL. 9543 1451 /8/. 9543 1451 ENERGY INDEPENDENT PARAMETERS: 9543 1451 R=9.37 FM, GAM-G=0.044 EV, DOBS=0.4 EV 9543 1451 ENERGY DEPENDENT PARAMETERS: 9543 1451 AT 150 EV: S0= 1.08E-4, S1=2.72E-4, WF=0.24 MILLI-EV 9543 1451 AT 30 KEV: S0= 0.79E-4, S1=1.99E-4, WF=0.30 MILLI-EV 9543 1451 9543 1451 CALCULATED 2200-M/S CROSS SECTIONS AND RESONANCE INTEGRALS 9543 1451 2200 M/S VALUE RES. INT. 9543 1451 TOTAL 614.6 B - 9543 1451 ELASTIC 11.13 B - 9543 1451 FISSION 3.018 B 13.9 B 9543 1451 CAPTURE 600.4 B 1300 B 9543 1451 9543 1451 MF=3 NEUTRON CROSS SECTIONS 9543 1451 9543 1451 MT=1,2 TOTAL AND ELASTIC SCATTERING CROSS SECTIONS 9543 1451 CALCULATED WITH OPTICAL AND STATISTICAL MODELS BY USING 9543 1451 CASTHY/9/. OPTICAL POTENTIAL PARAMETERS/10/ WERE OBTAINED9543 1451 BY FITTING THE DATA OF PHILLIPS AND HOWE /11/ : 9543 1451 V = 43.4 - 0.107*EN (MEV) 9543 1451 WS= 6.95 - 0.339*EN + 0.0531*EN**2 (MEV) 9543 1451 WV= 0 , VSO = 7.0 (MEV) 9543 1451 R = RSO = 1.282 , RS = 1.29 (FM) 9543 1451 A = ASO = 0.60 , B = 0.5 (FM) 9543 1451 9543 1451 9543 1451 MT=18 FISSION CROSS SECTION 9543 1451 EVALUATED ON THE BASIS OF THE DATA BY DABBS ET AL./8/ 9543 1451 9543 1451 MF=33 COVARIANCES OF NEUTRON CROSS SECTIONS FROM JENDL/D-99 9543 1451 MT=18 9543 1451 ADOPTED FROM ENDF/B-VI/12/. 9543 1451 9543 1451 9543 1451 REFERENCES 9543 1451 1) KIKUCHI Y.: JAERI-M 82-096 (1982). 9543 1451 2) NAKAGAWA T.: JAERI-M 88-008 (1989). 9543 1451 5) DERRIEN H. AND LUCAS B.: 1975 WASHINGTON, P.637, 9543 1451 NBS-SP-425 (1975). 9543 1451 6) KIKUCHI Y.: PRIVATE COMMUNICATION. 9543 1451 7) VANPRAET G. ET AL.: 1985 SANTA FE, 1, 493 (1986). 9543 1451 8) DABBS J.W.T. ET AL.: NUCL. SCI. ENG., 83, 22 (1983). 9543 1451 9) IGARASI S.: J.NUCL.SCI.TECHNOL.,12,67 (1975). 9543 1451 10) IGARASI S. AND NAKAGAWA T.: JAERI-M 8342 (1979). 9543 1451 11) PHILLIPS T.W. AND HOWE R.E.: NUCL. SCI. ENG., 69, 375 (1979). 9543 1451 12) ZHOU DELIN, GU FUHUA ET AL.: ENDF/B-VI (MAT=9543) (1988). 9543 1451 **************************************************************** 9543 1451 9543 1451 9543 1451 9543 1451 9543 1451 ***************** Program LINEAR (VERSION 2002-1) ***************9543 1451 For All Data Greater than 1.0000E-10 barns in Absolute Value 9543 1451 Data Linearized to Within an Accuracy of .100000000 per-cent 9543 1451 ***************** Program RECENT (VERSION 2002-1) ***************9543 1451 for All Data Greater than 1.0000E-10 barns in Absolute Value 9543 1451 Data Linearized to within an Accuracy of .100000000 per-cent 9543 1451 ***************** Program SIGMA1 (VERSION 2002-1) ***************9543 1451 Data Doppler Broadened to 300.000000 Kelvin 9543 1451 for All Data Greater than 1.0000E-10 barns in Absolute Value 9543 1451 Data Linearized to Within an Accuracy pf .100000000 per-cent 9543 1451 ***************** Program FIXUP (Version 2002-1) ****************9543 1451 Corrected ZA/AWR in All Sections-----------------------------Yes 9543 1451 Corrected Thresholds-----------------------------------------Yes 9543 1451 Extended Cross Sections to 20 MeV----------------------------No 9543 1451 Allow Cross Section Deletion---------------------------------No 9543 1451 Allow Cross Section Reconstruction---------------------------No 9543 1451 Make All Cross Sections Non-Negative-------------------------Yes 9543 1451 Delete Energies Not in Ascending Order-----------------------Yes 9543 1451 Deleted Duplicate Points-------------------------------------Yes 9543 1451 Check for Ascending MAT/MF/MT Order--------------------------Yes 9543 1451 Check for Legal MF/MT Numbers--------------------------------Yes 9543 1451 Allow Creation of Missing Sections---------------------------No 9543 1451 Allow Insertion of Energy Points-----------------------------No 9543 1451 Create Uniform Energy Grid-----------------------------------No 9543 1451 Delete Section if Cross Section =0 at All Energies-----------Yes 9543 1451 ***************** Program GROUPIE (VERSION 2002-1) **************9543 1451 Unshielded Group Averages Using 640 Groups 9543 1451 Weighting Spectrum: Flat (Constant) Spectrum 9543 1451 1 451 117 19543 1451 3 18 217 19543 1451 33 18 16 09543 1451 9543 1 0 9543 0 0 9.52410E+4 2.38986E+2 0 0 0 09543 3 18 1.99999E+8 1.99999E+8 0 0 1 6419543 3 18 641 1 9543 3 18 .000100000 51.3052773 .000105000 50.1079487 .000110000 48.97045709543 3 18 .000115000 47.9222736 .000120000 46.6968760 .000127500 45.33727459543 3 18 .000135000 44.0950790 .000142500 42.9534963 .000150000 41.71795199543 3 18 .000160000 40.4307858 .000170000 39.2595742 .000180000 38.18445049543 3 18 .000190000 37.1841024 .000200000 36.2707160 .000210000 35.41134409543 3 18 .000220000 34.6130933 .000230000 33.8713750 .000240000 33.00070879543 3 18 .000255000 32.0439328 .000270000 31.3056049 .000280000 30.48366259543 3 18 .000300000 29.4812057 .000320000 28.5734378 .000340000 27.74200399543 3 18 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