PWR Source Term Generation and Evaluation Rev 00B, ICN 00 000-00C-MGR0-00100-000-00B May 2004 1. PURPOSE This calculation is a revision of a previous calculation (BSC 2003 [DIRS 162471]) that bears the same title and has the Document Identifier 000-00C-MGR0-00100-000-00A. The purpose of this calculation is to revise the burnup value of the maximum pressurized water reactor (PWR) spent nuclear fuel (SNF) assembly to be consistent with Licensing Position-009 (Williams 2003, Attachment, p. 1 [DIRS 166132]). No new computer runs were performed for this revision. The source terms for the maximum PWR assembly were taken from the results of computer runs performed for the previous revision - BSC 2003 [DIRS 162471]. The scope of work includes the following: • Generate PWR SNF assembly source terms as a function of initial enrichment, burnup, and cooling time using an appropriate and defensible methodology. • Provide the average and maximum PWR fuel assembly specifications. • Calculate crud source term deposited on the surfaces of the assembly. This calculation establishes PWR SNF assembly source terms. The results of this calculation are intended for use in evaluation of shielding requirement or other follow-on analysis. They may also be used as input for Preclosure Safety Analysis and Total System Performance Assessment analyses. The results are provided on a per-assembly basis for the representative PWR fuel assembly used in this calculation. Limitation on the use of the results for other fuel assembly types should be evaluated on a case-by-case basis. The source terms of the representative PWR SNF assembly are generated for the first one million years after the SNF is discharged from the reactors. These source terms provide data characterizing the neutron and gamma spectra in particles per second, the decay heat in watts, and radionuclide inventories in curies. Conservative source terms are generated for a wide range of burnups and enrichments (see Tables 8 and 12) that are representative of the waste stream, stainless steel (SS) clad, and South Texas PWR assemblies. The source term due to the activation of corrosion products deposited on the surfaces of the assembly from the coolant is also calculated. The results of this calculation support many areas of the Monitored Geologic Repository (MGR), which include thermal evaluation, radiation dose determination, radiological safety analyses, surface and subsurface facility designs, preclosure safety analysis, and total system performance assessment. Therefore, it is subject to the requirements of the Quality Assurance Requirements and Description (DOE 2004 [DIRS 168669]). This includes MGR items classified as Safety Category, for example, the Commercial Waste Package (BSC 2003, p. A-3 [DIRS 165179]). Development, performance, and documentation of this calculation conform to the administrative procedure AP-3.12Q, Design Calculation and Analyses [DIRS 168413]. Environmental and Nuclear Engineering Design Calculation Title: PWR Source Term Generation and Evaluation Document Identifier: 000-00C-MGR0-00100-000-00B Page 6 of 33 2. METHOD The method used to evaluate the radioactive nuclides inventory and source terms at a specific time involves the simulation of the burnup and decay of fuel assemblies. For the previous revision of this calculation, the SAS2H/ORIGEN-S sequence of SCALE 4.3 computer code system was used to calculate the source terms for the selected fuel assemblies as function of assembly average burnup and cooling time. No new calculations were performed with SCALE 4.3 for this revision. Uncertainties and limitations in the computed source terms by the SAS2H/ORIGEN-S sequence are discussed in the SAS2H manual (CRWMS M&O 2000, Section S2.3 [DIRS 153872]) and in the NUREG/CR-5625 report (Hermann et al. 1994, Section 6.6.2 [DIRS 154045]). For each time-dependent fuel composition, SAS2H performs 1D neutron transport analysis of the reactor fuel assembly using a two-part procedure with two separate lattice-models. Basically, the model represents the fuel by an infinite lattice of fuel pins (path-A model) and the cell-weighted cross sections produced with the path-A model are then applied to the fuel region of a larger unit-cell model (path B) used to represent a fuel assembly. The concept of using cell-weighted data in the 1D transport analysis of the path-B model is an approximate method for including the 2D assembly effects. The path-B model is used by SAS2H to calculate an "assembly-averaged" fuel-region flux spectrum that includes the effects of the path-A model and other components in the assembly (e.g., guide tubes, burnable poison rods, etc.). The cross sections derived from a transport analysis at each time step are used in a point-depletion computation via ORIGEN-S that produces the burnup-dependent fuel composition to be used in the next spectrum calculation. This sequence is repeated over many cycles over the power history). Geometrically, SAS2H models the actual assembly by defining an equivalent cylindrical representation. Besides the geometric simplification, other approximations within SAS2H are: • Axial uniformity, • Densities within the fuel pin are considered uniform with space, • A constant temperature is applied in each zone of the unit cell, • A single fuel-pin unit-cell type is assumed for the entire assembly (i.e., all fuel pins are assumed to have the same enrichment), and • Fuel loading is proportional to the fuel length. The resulting source terms are extracted from the SAS2H output to create data files for future analyses. Electronic management of information generated from this calculation is controlled in accordance with AP-3.13Q, Design Control, [DIRS 167460]. The computer input and output files generated from this calculation are stored on four compact disks (CDs), and submitted as an attachment to this document (Attachment X). Environmental and Nuclear Engineering Design Calculation Title: PWR Source Term Generation and Evaluation Document Identifier: 000-00C-MGR0-00100-000-00B Page 7 of 33 3. ASSUMPTIONS 3.1 It is assumed that various PWR assembly types can be approximated by a single assembly and the source terms generated will not be greatly affected by using this common geometry. This assumption results in the path A and B representations in SAS2H for unit cells of the South Texas and SS assemblies being identical to those of the Babcock & Wilcox (B&W) Mark B assembly used to represent the waste stream cases. Different initial heavy metal loadings (IHMLs) are accounted for by adjusting the fuel length, which is calculated from the IHML and the fuel density. Rationale: A sensitivity study (BSC 2004, Section 6.4 [DIRS 167058]) determined that the selection of a particular assembly is not sensitive to the resulting source terms. Usage: This assumption is used in the entire calculation. 3.2 It is assumed that the neutron activation factors for the assembly hardware regions (bottom end-fitting, plenum, and top end-fitting) are 1.5 time of the values recommended by the U.S. Nuclear Regulatory Commission (NRC) and provided in Luksic 1989, Table S.1 [DIRS 120506]. Rationale: This assumption provides 50% conservative margin for the neutron-activated sources in the hardware regions. Usage: This assumption is used in Section 5.3. 3.3 It is assumed that the fuel rod length is 153.68 in. or 390.3472 cm (DOE 1988, p. 2A-33 [DIRS 100979]) and that the plenum region height is the difference between the fuel rod length and the fuel stack height (30.1752 cm = 390.3472 cm – 360.172 cm). Rationale: Fuel rod data are proprietary information and. DOE 1988 [DIRS 100979] is the best available technical information. Usage: This assumption is used in Section 5. 3.4 The hardware parts and materials of the B&W Mark B PWR assembly described in DOE 1988, p. 2A-32 and p. 2A-34 [DIRS 100979] are assumed. Rationale: Fuel assembly data are proprietary information and the information in DOE 1988 [DIRS 100979] is the best available. Usage: This assumption is used in Section 5.3. 3.5 It is assumed that the South Texas assembly will generate the bounding crud source term and that the physical characteristics of the South Texas assembly can be obtained from DOE 1988, p. 2A-361 and p. 2A-363 [DIRS 100979]. Rationale: Again, DOE 1988 [DIRS 100979] is the best available technical information on PWR fuel. The crud source is proportional to the surface area that is exposed to coolant; since the South Texas fuel assembly has larger surface area than the B&W Mark B assembly, it will generate a more conservative (higher) crud source that will serve as the bounding crud source for the representative PWR assembly. Usage: This assumption is used in Section 5.3. Environmental and Nuclear Engineering Design Calculation Title: PWR Source Term Generation and Evaluation Document Identifier: 000-00C-MGR0-00100-000-00B Page 8 of 33 4. USE OF COMPUTER SOFTWARE AND MODELS 4.1 SOFTWARE APPROVED FOR QA WORK The SAS2H/ORIGEN-S sequence of SCALE Version 4.3, modular code system (CRWMS M&O 1997, [DIRS 154059]) was used to perform source terms (decay thermal power, gamma spectrum, neutron spectrum) and radionuclide inventory calculations. The software specifications are as follows: • Program Name: SCALE • Version/Revision Number: 4.3 (Addendum) HP 9000 Version • CSCI Number: 30011 V4.3 (Addendum) (CRWMS M&O 1997 [DIRS 154059]) • Computer Type: Hewlett Packard 9000 Series • Status/Operating System: qualified1/HP-UX 10.20 • Computer Processing Unit Names and Civilian Radioactive Waste Management System (CRWMS) Management and Operating (M&O) Contractor Tag Number: ‘Bloom,’ CRWMS-M&O Tag 700887. The SAS2H/ORIGEN-S sequence of SCALE Version 4.3 software was: (a) appropriate for source term calculations, (b) used only within the range of validation as documented in CRWMS M&O 1997. Software Qualification Report for THE SCALE Modular Code System Version 4.3. [DIRS 106016], and (c) obtained from Software Configuration Management in accordance with the appropriate procedure. The “*.cut” files, which consist of the input echoes and the final ORIGEN-S outputs, are listed in Attachment VIII and their electronic versions are stored in Attachment X. The “*.cut” files are sufficient to independently repeat the calculation. 4.2 EXEMPT SOFTWARE 4.2.1 Excel • Title: EXCEL • Version/Revision Number: Microsoft® Excel 2000 SR-1 • This software is installed on a personal computers running Microsoft Windows 2000 with CPU Number 501661. Standard functions of Microsoft Excel for Windows, Version 2000 SR-1, are used in this calculation to display results graphically or in tabular form and to perform simple calculations as documented in Section 5 of this calculation. The user-defined formulas, inputs, and results are documented in sufficient detail in Section 5 to allow an independent repetition of the various computations without recourse to the originator. Microsoft Excel is an exempt software product 1 The modular code system SCALE Version 4.3 results were produced in prior revisions of this calculation. This software version has since been retired. Environmental and Nuclear Engineering Design Calculation Title: PWR Source Term Generation and Evaluation Document Identifier: 000-00C-MGR0-00100-000-00B Page 9 of 33 according to the procedure LP-SI.11Q-BSC, Software Management, Section 2.1.6 [DIRS 168412]. The Excel files are stored on CDs (Attachment X) and documented in Attachment VIII. 4.2.2 UNIX Script Files • Titles: cut-script, neutrons, gammas, curies, watts • Version/Revision Number: All are version 00 • Computer Type: Hewlett Packard 9000 Series The UNIX script files are listed in Table 1. The “cut-script” file executes SAS2H and generates a “*.cut” file from the SAS2H output file. The other four files perform simple editorial tasks that extract lines of information from a single “*.cut” file. They are executed by typing the command “awk –f (script file) (input *.cut file name) > (output file name)”. These files are intended for use only with the SAS2H/ORIGEN-S output files for the waste stream, SS clad, and South Texas assemblies stored in Attachment X. The output of the script files has been verified by visual inspection. The specific task of each script file is noted in Table 1. The script files are listed in Attachment V. This software is considered exempt from the requirements of LP-SI.11Q-BSC, Software Management, Section 2.1.2 [DIRS 168412]. Table 1. UNIX Script Files File Name Function neutrons Extracts the total (alpha-n plus spontaneous fission) neutron source table from a “*.cut” file gammas Extracts the gamma source from the light element, actinide, and fission product contributions from a “*.cut” file watts Extracts the total thermal output from the light element, actinide, and fission product contributions from a “*.cut” file curies Extracts the tables of nuclide curies from a “*.cut” file for the light element, actinide, and fission product contributions cut-script Generates “*.cut” file from each SAS2H output file Environmental and Nuclear Engineering Design Calculation Title: PWR Source Term Generation and Evaluation Document Identifier: 000-00C-MGR0-00100-000-00B Page 10 of 33 5. CALCULATION 5.1 REVISION HISTORY The purpose of this revision is to update the burnup value for the maximum PWR SNF assembly to 80 GWd/MTU in conformance with the Licensing Position-009 (Williams 2003, Attachment, p. 1 [DIRS 166132]). At the same time, based on the conclusion of BSC 2004 [DIRS 167058], the rationale of the Assumption 3.1 was modified since a sensitivity study (BSC 2004, Section 6.4 [DIRS 167058]) determined that the selection of a particular assembly is not sensitive to the resulting source terms. Also, an “Acronyms and Abbreviation” section and Attachment XI were added. Several calculations have been performed to provide PWR SNF source terms for shielding calculations. CRWMS M&O 1997 [DIRS 136439] was the first in a series of calculations to provide source terms, but was limited to a South Texas assembly and a handful of burnups and enrichments that represented the anticipated average and maximum waste stream assemblies. In addition, the source terms represented only the first twenty-five years after discharge from the reactor. CRWMS M&O 1998 [DIRS 124815] built on this calculation to extend the source terms to 30,000 years. CRWMS M&O 1999 [DIRS 105912] extends the calculation done in CRWMS M&O 1998 [DIRS 124815] to include a wide range of burnups and enrichments, and the inclusion of a crud source. The sources were also provided out to 1 million years after discharge. CRWMS M&O 1999 [DIRS 136429] provides a complete set of conservative source terms for the SS clad, South Texas, and PWR assemblies in the expected waste stream. BSC 2003 [DIRS 162471] expands the source terms database of the representative PWR SNF in CRWMS M&O 1999 [DIRS 136429] to include burnups of 80 and 85 GWd/MTHM. The following modifications are included in this revision: • Figure 3, Table 14, Attachment VII and IX were updated to include the values calculated in the last revision (BSC 2003 [DIRS 162471]) for the PWR SNF assembly maximum burnup of 80 GWd/MTU. The cooling times selected to present the curies inventories correspond with those used in SAS2H calculations. • The description of the design methodology was expanded in Section 2. • “Acronyms and Abbreviations” section was added. • Sections 1, 3, 5.1, 5.5, and 6 were modified. • Attachment VIII was modified to reflect the current content of the CDs. Since this revision retains all the results of BSC 2003 [DIRS 162471], the methodology and calculation method of the last revision and this revision are the same. The fuel densities for the PWR assemblies are kept constant, and the lengths of the assemblies are varied to reflect the heavy metal loading in the SAS2H calculations (see Section 5.4.2). 5.2 SELECTION OF CONSERVATIVE PARAMETERS The inputs for this calculation are specifically chosen to lead to conservative source terms. This section discusses the main inputs and the reasons they are used. It covers several different Environmental and Nuclear Engineering Design Calculation Title: PWR Source Term Generation and Evaluation Document Identifier: 000-00C-MGR0-00100-000-00B Page 11 of 33 parameters. The first of these is the geometry for the South Texas, waste stream, and SS clad assemblies. In this calculation, the geometry used to model the assemblies in SAS2H corresponds to a typical B&W Mark B PWR fuel assembly. The Mark B has a high IHML and a large amount of assembly hardware, supporting more fission product generation and hardware activation. Therefore, this assembly provides a conservative basis for the PWR waste stream. While a real Mark B assembly has an IHML of 464 kg, this is increased to 475 kg for the waste stream and the SS clad assemblies to provide slightly higher source intensities. The South Texas assembly calculations use a 550 kg IHML. As mentioned in Section 5.1, additional heavy metal mass is accounted for by increasing the fuel length, rather than the fuel density. A longer active fuel length and a lower density rather than a shorter fuel length and a higher density decrease the fuel self-shielding. This results in a higher flux and consequently higher source intensities. In this calculation the higher IHML of the waste stream, SS clad, and South Texas assemblies is accounted for by the fuel length rather than the fuel density (see Section 5.4.2). The fuel density also affects the source from the hardware regions of the assembly. The irradiation of the assembly hardware is calculated by inputting the desired amount of light element material into the fuel region in SAS2H. When the sources are calculated, only the light elements are included. For the South Texas assembly, since the same power is generated by a longer fuel rod than by the shorter Mark B, the result is a lower power density. This lower flux per unit height for the South Texas assembly provides a lower source for the hardware regions. For this reason, the South Texas assemblies use the hardware files for the waste stream (the SS assemblies also use waste stream files for the bottom and top assembly regions). In addition, since these hardware sources are calculated by irradiating the material in the fuel region, scaling factors (SF) are used to account for the lower flux seen in the non-fuel regions of the reactor. The SFs for the neutron flux, provided in Luksic 1989, Table S.1 [DIRS 120506], have an uncertainty of ± 50%. To generate conservative source terms for the non-fuel regions of the PWR assembly, the scaling factors used in this calculation represent 150% of those listed in Luksic 1989 [DIRS 120506]. Consideration has also been given to the material definitions. The compositions of Zircaloy-4 (ASTM B 811-90 1991 [DIRS 131753]), Stainless Steel 304, Stainless Steel 304L (ASME 2001 [DIRS 158115]), Stainless Steel 302 (ASTM A 240/A 240M-97a 1997 [DIRS 102769]), Inconel-718 (Inco Alloys International 1988 [DIRS 130835]), and Stainless Steel CF3M (ASM 1980 [DIRS 104317]) are representative of materials used in the manufacture of nuclear fuel assemblies. These compositions use the maximum amounts of cobalt given by the references and a 0.08 wt% cobalt impurity (Ludwig and Renier, p. 45 [DIRS 146398]) for the stainless steels. While none is indicated in ASME 2001 [DIRS 158115] and ASTM A 240/A 240M-97a 1997 [DIRS 102769], it is a common practice to include a cobalt impurity in stainless steels (due to nickel previously alloyed with cobalt). The balance of the remaining elements are representative of the material compositions for each material, but are biased towards the maximum amount of Sn, Ni, and Nb. Impurities are also included in the fuel itself, and are given in Table 4. Environmental and Nuclear Engineering Design Calculation Title: PWR Source Term Generation and Evaluation Document Identifier: 000-00C-MGR0-00100-000-00B Page 12 of 33 Not all the activation sources can be calculated with SAS2H. The source due to corrosion material that accumulates on the surfaces of the assembly from the flow of coolant (crud) is also calculated. A bounding crud source term is based on the South Texas assembly. This representation of the South Texas does not use Mark B dimensions. The South Texas has a greater surface area exposed to coolant, and the crud source is strongly dependent on the available surface area. This dependence comes from the source being calculated with a radioisotope activity density (Ci/cm2). Two estimates for radioisotope densities are used, from two sources (NRC 2000 [DIRS 149756] and Jones 1992 [DIRS 146405]). This calculation is discussed in detail in Section 5.3.2. The enrichments calculated range from 0.711 wt% (natural) to 5.5 wt% (the complete list is shown in Section 5.4.2). This is done to cover the wide range seen in the waste stream and avoid the need to extrapolate for information on assemblies currently being developed. Also, the number of time steps has been increased to 180, in order to provide more detailed information, particularly for the first 100 years the fuel is out of the reactors. 5.3 INPUT 5.3.1 Input for SAS2H Source Calculations The general assembly description for a B&W Mark B PWR fuel assembly is provided in Table 2. Table 3 presents the PWR assembly hardware components and flux SF used to determine the masses of the light elements that are entered in SAS2H data Block 10. The concentrations of element impurities in fresh light water reactor fuel, in parts of element per million parts of heavy metal (ppm), are listed in Table 4. These values are provided by Ludwig and Renier [DIRS 146398] and reflect measured concentrations instead of the maximum allowable concentrations given in purity specifications. Table 5 provides the chemical compositions of the hardware materials used. Ranges are provided for some of the elements, and the values used in this calculation are provided. Environmental and Nuclear Engineering Design Calculation Title: PWR Source Term Generation and Evaluation Document Identifier: 000-00C-MGR0-00100-000-00B Page 13 of 33 Table 2. B&W Mark B PWR Fuel Assembly Description and Operating Parameters Assembly Parameter Value Units Metric Units Reference Average core exit moderator temperature 612 °F 595.4 K Average core moderator pressure 2200 Psia - - Maximum beginning of cycle boron concentration - - 1050 ppm Core thermal power - - 2568 MW Pellet average temperature (K) 1200 °F 922 K Fuel cladding to moderator temperature differential 50-75 °F 28-42 K Average core moderator temperature rise 59.4 °F 33 K Framatome Cogema 1999, p 3 [DIRS 146419] In association with BSC 2003, p. 2 [DIRS 165684] Number of guide tubes 16 /assembly NA NA Number of instrument tubes 1 /assembly NA NA Clad/tube material Zircaloy-4a NA NA NA Number of assemblies 177 In core NA NA Total number of fuel rods 208 /assembly NA NA Number of rods on a lattice side 15 /side NA NA Fuel pellet outer diameter (OD) 0.3686 inches 0.93624 cm Fuel stack height 141.8 inches 360.172 cm Fuel clad OD 0.430 inches 1.0922 cm Clad thickness 0.0265 inches 0.06731 cm Fuel rod pitch 0.568 inches 1.44272 cm Guide tube OD 0.530 inches 1.3462 cm Guide tube ID 0.498 inches 1.26492 cm Fuel clad inner diameter (ID) 0.377 inches 0.95758 cm Table 2-2 Fuel pellet fraction of theoretical density - - 0.95 NA Table 2-9 Mass of U - - 463.63 kg Table 3-1 Fuel assembly envelope 8.536 inches 21.6814 cm Punatar 2001b [DIRS 155635] p. 2-3 Fuel rod length 153.68 inches 390.3472 cm Plenum region height - - 30.1752 cm Assumption 3.3 Specific volume of steam at 2200 psi, 580°F 0.02275 ft3/lbm - - Specific volume of steam at 2200 psi, 590°F 0.02235 ft3/lbm - - ASME 1993, p. 281 [DIRS 108050] NOTE: a Stainless steel clad calculations use SS-304 as the clad material. b The sources of the fuel assembly physical data in Punatar 2001 [DIRS 155635] are from [DIRS 154999], [DIRS 155000], [DIRS 155001], and [DIRS 155002]. Environmental and Nuclear Engineering Design Calculation Title: PWR Source Term Generation and Evaluation Document Identifier: 000-00C-MGR0-00100-000-00B Page 14 of 33 Table 3. Assembly Hardware and Flux Scaling Factors Region Flux Scaling Factorsa Part Nameb kg/Assemblyb Materialb Top nozzle 7.48 Stainless Steel CF3M Spring retainer 0.91 Stainless Steel CF3M Hold down spring 1.8 Inconel-718 Upper end plug 0.06 Stainless Steel 304 Top end fitting 0.15 Upper nuts 0.51 Stainless Steel 304L Bottom nozzle 8.16 Stainless Steel CF3M Spacer-bottom 1.3 Inconel-718 Bottom end fitting 0.30 Lower nuts 0.15 Stainless Steel 304c Guide tubes 8.0 Zircaloy-4d Instrument tube 0.64 Zircaloy-4 Spacers-incore 4.9 Inconel-718 Grid supports 0.64 Zircaloy-4 Spacer-plenum 1.04 Inconel-718 Fuel/Plenum fuel region - 1.00 plenum region - 0.30 Plenum spring 0.042e Stainless Steel 302e NOTE: a Assumption 3.2. b Assumption 3.4. c Zircaloy-4 was used instead of SS 304, but the effect on 60Co activated source is negligible because of the much larger cobalt content in the bottom nozzle and bottom spacer. d Stainless steel clad calculations use SS-304 as the clad material. e Pounds per assembly, Assumption 3.4. Table 4. Non-actinide Composition of UO2 Element Concentration (ppm) Element Concentration (ppm) Element Concentration (ppm) Li 1.0 Mn 1.7 V 3.0 B 1.0 Fe 18.0 Cr 4.0 C 89.4 Co 1.0 Bi 0.4 N 25.0 Ni 24.0 Pb 1.0 F 10.7 Cu 1.0 Ti 1.0 Na 15.0 Zn 40.3 Ca 2.0 Mg 2.0 Mo 10.0 W 2.0 Al 16.7 Ag 0.1 Cl 5.3 Si 12.1 Cd 25.0 Sn 4.0 P 35.0 In 2.0 - - Source: Ludwig and Renier, Table 5.4 [DIRS 146398]. Environmental and Nuclear Engineering Design Calculation Title: PWR Source Term Generation and Evaluation Document Identifier: 000-00C-MGR0-00100-000-00B Page 15 of 33 Table 5. Chemical Composition for Hardware Materials Element Wt% Range Value Used Chemical Composition for Stainless Steel 304La C 0.03 (max) 0.03 Mn 2.00 (max) 2.00 Si 0.75 (max) 0.75 Cr 18.00-20.00 19.00 Ni 8.00-12.00 11.92 Co - 0.08 P 0.045 (max) 0.045 S 0.03 (max) 0.03 N 0.1 (max) 0.1 Fe Balance 66.045 Chemical Composition for Stainless Steel 304a C 0.08 (max) 0.08 Mn 2.00 (max) 2.00 Si 1.00 (max) 0.75 Cr 18.00-20.00 19.00 Ni 8.00-10.50 10.42 Co - 0.08 P 0.045 (max) 0.045 S 0.03 (max) 0.03 N 0.1 (max) 0.1 Fe Balance 67.495 Chemical Composition for Stainless Steel 302e C 0.15 0.15 Mn 2.00 2.00 Si 0.75 0.75 Cr 17.00-19.00 18.00 Ni 8.00-10.00 9.92 Co - 0.08 P 0.045 0.045 S 0.03 0.03 N 0.10 (max) 0.1 Fe Balance 68.925 Chemical Composition for Stainless Steel CF3Mb C 0.03 (max) 0.03 Mn 1.50 (max) 1.50 Si 2.00 (max) 2.00 Cr 17.00-21.00 19.00 Ni 8.00-12.00 11.92 Co - 0.08 Mo 2.00-3.00 2.50 Fe Balance 62.97 Environmental and Nuclear Engineering Design Calculation Title: PWR Source Term Generation and Evaluation Document Identifier: 000-00C-MGR0-00100-000-00B Page 16 of 33 Table 5. Chemical Composition for Hardware Materials (Continued) Element Wt% Range Value Used Chemical Composition for Inconel-718c Ni 50.00-55.00 54.00 Cr 17.00-21.00 19.00 Fe Balance 14.934 Nb/Ta 4.75-5.50 5.5 Mo 2.80-3.30 3.05 Ti 0.65-1.15 0.90 Al 0.20-0.80 0.50 Co 1.00 (max) 1.00 Mn 0.35 (max) 0.35 Si 0.35 (max) 0.35 Cu 0.30 (max) 0.30 C 0.08 (max) 0.08 S 0.015 (max) 0.015 P 0.015 (max) 0.015 B 0.006 (max) 0.006 Chemical Composition for Zircaloy-4d O 0.09-0.16 0.12 Cr 0.07-0.13 0.10 Fe+Cr 0.28-0.37 0.20 (Fe) Sn 1.20-1.70 1.70 Zr Balance 97.88 NOTE: a ASME 2001, SEC II A SA-240, Table 1 [DIRS 158115]. b ASM 1980, p. 95 [DIRS 104317]. c Inco Alloys International 1988, p. 11 [DIRS 130835]. d ASTM B 811-90 1991, Table 2 [DIRS 131753]. e ASTM A 240/A 240M-97a 1997, Table 1 [DIRS 102769]. 5.3.2 Input for Crud Calculations In addition to the source terms for an assembly based on the irradiated assembly, it is also necessary to estimate the radiation source due to the activated corrosion products from the coolant deposited on the surfaces of the assembly (crud). These surfaces include all the areas of the assembly exposed to the flow of coolant. A bounding estimate of the PWR assembly surface area is based on a South Texas assembly (Assumption 3.5), which has a longer length, 264 fuel rods, and 25 guide or instrumentation tubes. The fuel rods for this assembly have a 0.374-inch (0.95-cm) outer diameter (OD) and a length of 176.642 inches (448.67 cm) (DOE 1988, p. 2A-363 [DIRS 100979]). Both the inner and outer surface areas of the guide and instrument tubes are included in the estimation of the surface area of the assembly. The tube OD is taken to be the pitch of the fuel pins for a conservative estimate and is 0.496 inches (1.2598 cm) (DOE 1988, p. 2A-361 [DIRS 100979]). The cladding thickness is used to determine the inside diameter and surface area and is 0.0225 inches (0.05715 cm) (DOE 1988, p. 2A-363 [DIRS 100979]). The overall assembly length is 199 inches (505.46 cm) (DOE 1988, p. 2A-361 [DIRS 100979]). In addition to the calculation of a bounding crud source based on the South Texas assembly parameters, a source for a regular B&W Mark B assembly is also calculated. Environmental and Nuclear Engineering Design Calculation Title: PWR Source Term Generation and Evaluation Document Identifier: 000-00C-MGR0-00100-000-00B Page 17 of 33 A radionuclide activity density (Ci/cm2) is required to calculate the crud source. Two estimates for the radionuclide activity density are used in this calculation. The first estimate for the activity density is provided by the NRC NUREG-1567 (NRC 2000, Table 9.2 [DIRS 149756]), and has a value 140 µCi/cm2, which is entirely due to 60Co. The second estimate of crud activity density is provided by Jones 1992 [DIRS 146405]. This estimate has activities for eight radionuclides and is included in this calculation for information purposes only. The data from NRC 2000 [DIRS 149756] and Jones 1992 [DIRS 146405] are summarized in Table 6. Although the data in Jones 1992 [DIRS 146405] contain activities from multiple radioactive isotopes, it is recommended that the NRC value be used for crud activity. The reason is that the 60Co activity from NRC is nearly three times the 60Co activity from Jones 1992 [DIRS 146405]. Emitting two very energetic photons and some beta particles, 60Co is the most dominant isotope for crud. Since other radionuclides in Table 6 either have shorter half-lives or emit much lower intensity radiation, they become insignificant by the time the SNF arrives the potential repository. Table 6. Radionuclide Activity Densities Used in Crud Source Calculation Radionuclide Activity Density (Ci/cm2) at Discharge Half Lifea Reference 60C0 1.40x10-4 5.27 years NRC 2000, Table 9.2 [DIRS 149756] 51Cr 1.89x10-4 27.70 days 54Mn 7.40x10-5 312.10 days 55Fe 5.90x10-3 2.73 years 58Co 4.03x10-4 70.88 days 59Fe 7.60x10-5 44.51 days 60Co 5.10x10-5 5.27 years 63Ni 2.00x10-6 100.00 years 95Zr 1.90x10-5 64.02 days Jones 1992, Table 1 [DIRS 146405] NOTE: a Parrington et.al. 1996, pp. 24-28 [DIRS 103896]. 5.4 SAS2H INPUT PARAMETER CALCULATION This section describes the calculation of the required input for SAS2H. The SAS2H sequence calculates source terms for the four axial regions: bottom end fitting, active fuel, plenum, and top end fitting. Every calculation requires the input of assembly parameters, burnup history, decay time steps, the pin cell description (path A), and the larger unit cell description for the whole assembly (path B). The path A input is presented in Section 5.4.1. Section 5.4.2 describes the calculation of uranium isotopic composition, density, and length for the active fuel region. The calculation of the path B representation is provided in Section 5.4.3. Section 5.4.4 provides the masses of the light elements used in SAS2H source calculation of the activation products. Section 5.4.5 calculates the SAS2H input data for moderator temperature and density and boron concentration. Fuel burnup and decay calculations are provided in Section 5.4.6. Environmental and Nuclear Engineering Design Calculation Title: PWR Source Term Generation and Evaluation Document Identifier: 000-00C-MGR0-00100-000-00B Page 18 of 33 5.4.1 Input for the Path A Calculation in SAS2H The first information in the SAS2H input files specifies the type of calculation and the cross section library to be used. The “latticecell” type of calculation and the “44groupndf5” cross section library are used. Because no dose calculation is performed in this calculation, the parameters for the shipping cask “skipcellwt” and “skipshipdata” are specified. SAS2H then uses this information to generate an infinite lattice of unit pin cells and calculate cell-weighted cross sections. These cross sections are then applied to the assembly described by the path B representation. Figure 1 presents the path A unit pin cell representation for SAS2H and Table 7 presents the input parameters for cell calculations. fuel pin fuel cell moderator cladding Figure 1. Unit Pin Cell Representation for Path A Table 7. SAS2H Input for the Path A Calculation Variable Name Value Used Lattice type Square pitch Pitch 1.44272 cm Fuel OD 0.93624 cm Mfuel 1 Mmod 3 Clad OD 1.0922 cm Mclad 2 Clad ID 0.95758 cm Mgap 0 Environmental and Nuclear Engineering Design Calculation Title: PWR Source Term Generation and Evaluation Document Identifier: 000-00C-MGR0-00100-000-00B Page 19 of 33 5.4.2 Fuel Uranium Isotopic Composition, Density, and Calculated Length The isotopic composition of uranium for commercially available enriched uranium is determined by the given initial 235U enrichment and the following formulae (Bowman et.al. 1995, p. 20 [DIRS 123796]): batch fuel associated the of enrichment U U wt _ _ _ _ _ 235 235 % = 0837 . 1 235 234 ) % ( * 007731 . 0 % U wt U wt = ) % ( * 0046 . 0 % 235 236 U wt U wt = ) % % % ( 100 % 236 235 234 238 U wt U wt U wt U wt + + - = The values for initial uranium isotope composition, presented in Table 8, are calculated in Attachment I. Table 8. Initial Uranium Isotope Compositiona “*.cut” Files E Code 235U (wt%) 236U (wt%) 234U (wt%) 238U (wt%) E12 0.711 0.00327 0.00534 99.28039 E11 1.0 0.00460 0.00773 98.98767 E10 1.5 0.00690 0.01200 98.48110 E9 2.0 0.00920 0.01639 97.97441 E8 2.5 0.01150 0.02087 97.46763 E7 3.0 0.01380 0.02543 96.96077 E6 3.5 0.01610 0.03005 96.45385 E5 4.0 0.01840 0.03473 95.94687 E4 4.2 0.01932 0.03661 95.74407 E3 4.5 0.02070 0.03946 95.43984 E2 5.0 0.02300 0.04423 94.93277 E1 5.5 0.02530 0.04904 94.42566 NOTE: a Only enrichment 1.5 wt% through 4.0 wt% are used for the SS clad calculations. A smeared fuel density is used in SAS2H. This is a common approach because fuel pellets expand and fill the gap during and after reactor operation. The smeared density is calculated from the following equation: ( ) . .. . . .. . . = . 2 2 l theoretica smeared diameter inside clad diameter pellet * * density l theoretica % Environmental and Nuclear Engineering Design Calculation Title: PWR Source Term Generation and Evaluation Document Identifier: 000-00C-MGR0-00100-000-00B Page 20 of 33 With a UO2 density of 10.96 g/cm3 (ORNL 1997, p. M8.2.3 [DIRS 135808]) and the pellet and clad inside diameters from Table 2, the smeared fuel density is: 3 2 2 3 953 . 9 95758 . 0 93624 . 0 * ) 96 . 10 ( * ) 95 . 0 ( cm g cm g smeared = . .. . . .. . = . A smeared fuel density of 9.96 g/cm3 is used in SAS2H. With the initial heavy metal loadings of 464 kg, 475 kg, and 550 kg respectively for the waste stream assembly, the SS clad assembly, and the South Texas assembly, the active fuel lengths of 369.699 cm, 378.463 cm, and 438.221 cm are obtained (see Attachment I). 5.4.3 Input for the Path B Calculation The larger representation of the assembly used in the path B calculation by SAS2H is shown in Figure 2. Four regions are used in describing this larger unit cell. The four regions are concentric rings, similar to the unit pin cell representation in path A. The inner region is the moderator inside the guide tube, which is surrounded by the guide tube itself (region 2). The third region is the moderator in the guide tube cell surrounding the outside of the guide tube. The last region represents the homogenized fuel, cladding, and moderator of the assembly. Figure 2. Larger Unit Cell Representation for Path B The first radius, r1, corresponds to the inner radius of the guide tube (1.26492/2 = 0.63246 cm). The second radius, r2, corresponds to the outer radius of the guide tube (1.3462/2 = 0.6731 cm). The third radius, r3, corresponds to an area equivalent to rod pitch squared. Thus, cm 814 . 0 44272 . 1 pitch r 2 2 3 = p = p = r3-outside of guide tube cell moderator r4-outside of homogenized fuel and moderator region r2-outside of guide tube r1-inside of guide tube Environmental and Nuclear Engineering Design Calculation Title: PWR Source Term Generation and Evaluation Document Identifier: 000-00C-MGR0-00100-000-00B Page 21 of 33 The fourth radius is calculated to preserve the guide tube-to-fuel volume ratio. The material for the fourth region is defined as mixture 500, which is the homogenized fuel, cladding and moderator determined from the path A calculation by SAS2H. Thus, the fourth radius is determined by taking a ratio of the number of available positions for fuel pins in the assembly to the number of positions used by the guide tubes and instrumentation tubes. This ratio is set equal to the ratio of the areas of the intended fuel region and the moderator (i.e., the guide tube pin cell). This allocates the number of fuel pins equally to each guide tube, and then sets the total volume of the assembly cell to be equal to the volume of the number of pin cells per guide tube. The instrument tube is taken to be equivalent to a guide tube for the path B calculation. This approach preserves the guide tube to fuel ratio and provides a unit cell that is representative of a core average composition. Therefore, the r4 radius is derived according to the following equations: r1 through r4 are shown in Table 9 with the mixtures used in each radius. Table 10 lists the rest of the input for the fuel assembly required by SAS2H. Table 9. Dimensions of the Larger Unit Cell for Path B Variable Name Value Used Comment Mixes 3 Material No. for the moderator Radius 0.63246 cm, region 1 Mixes 2 Material No. for the Zircaloy-4 Radius 0.67310 cm, region 2 Mixes 3 Material No. for the moderator Radius 0.814 cm, region 3 Mixes 500 Special material No. for fuel in the larger unit cell Radius 2.961 cm, region 4 cm 961 . 2 17 15 * 15 * 814 . 0 * : Yielding * as written be can which , 2 2 3 4 2 3 2 4 2 3 2 4 2 3 2 4 = = = = = = N N r N N r r r r r r N N tubes guide positions tubes guide positions tubes guide positions r p p Environmental and Nuclear Engineering Design Calculation Title: PWR Source Term Generation and Evaluation Document Identifier: 000-00C-MGR0-00100-000-00B Page 22 of 33 Table 10. Input for Fuel Assembly in SAS2H Calculations Variable Name Value Used Npin/assm 208 pins per assembly Fuelngth 360.172a Ncycles 1 cycle Nlib/cycb See Table 12 Lightel 33 Printlevel 5 Inplevel 2 Numholes 17 Numinstr 1 Mxtube 2 Ortube 0.6731 cm Srtube 0.63246 cm Asmpitch 21.6814 cm Numztotal 4 Mxrepeats 1 Mixmod 3 Facmesh 1.0 NOTE: a See Section 5.4.2 for the calculation of the fuel lengths to account for the different IHMLs. b Number of libraries made per cycle. 5.4.4 Light Element Mass Calculations The masses of the light elements for each axial fuel region are calculated in Attachment III and are reported in Table 11. The light element masses are determined by multiplying the weight of the hardware by the wt%s of the elements that make up the hardware and then adjusting by scaling factors to account for hardware's location in the reactor. For the active fuel region, the impurities in the fuel are also included. SAS2H is then used to simulate the irradiation of the fuel and the light elements and to decay the radiation source. Environmental and Nuclear Engineering Design Calculation Title: PWR Source Term Generation and Evaluation Document Identifier: 000-00C-MGR0-00100-000-00B Page 23 of 33 Table 11. Light Element Masses per Assembly (kg) Region: Top (SF: 0.15) Bottom (SF: 0.30) Fuel (SF: 1) Plenum (SF: 0.30) Element Waste Stream Waste Stream Waste Stream South Texas Steel Clad Waste Stream Steel Clad Ag - - 0.0000 0.0001 0.0000 - - Al 0.0014 0.0020 0.0324 0.0337 0.0324 0.0016 0.0016 B 0.0000 0.0000 0.0008 0.0008 0.0008 0.0000 0.0000 Bi - - 0.0002 0.0002 0.0002 - - C 0.0006 0.0010 0.0464 0.0531 0.1567 0.0003 0.0030 Ca - - 0.0010 0.0011 0.0010 - - Cd - - 0.0119 0.0138 0.0119 - - Cl - - 0.0025 0.0029 0.0025 - - Co 0.0038 0.0059 0.0495 0.0496 0.1598 0.0031 0.0059 Cr 0.3067 0.5393 1.0480 1.0764 27.1279 0.0632 0.7187 Cu 0.0008 0.0012 0.0152 0.0153 0.0152 0.0009 0.0009 F - - 0.0051 0.0059 0.0051 - - Fe 0.8894 1.5998 0.9706 1.0282 93.7937 0.0563 2.3893 In - - 0.0010 0.0011 0.0010 - - Li - - 0.0005 0.0006 0.0005 - - Mg - - 0.0010 0.0011 0.0010 - - Mn 0.0215 0.0381 0.0180 0.0181 2.7753 0.0012 0.0705 Mo 0.0397 0.0731 0.1542 0.1550 0.1542 0.0095 0.0095 N 0.0001 0.0000 0.0119 0.0138 0.1497 0.0000 0.0035 Na - - 0.0071 0.0083 0.0071 - - Nb 0.0149 0.0215 0.2695 0.2695 0.2695 0.0172 0.0172 Ni 0.3059 0.5024 2.6574 2.6592 17.0229 0.1690 0.5301 O 0.0000 0.0001 64.9109 75.1719 64.7735 0.0035 0.0000 P 0.0001 0.0001 0.0173 0.0200 0.0794 0.0000 0.0061 Pb - - 0.0005 0.0006 0.0005 - - S 0.0001 0.0001 0.0007 0.0007 0.0421 0.0000 0.0011 Si 0.0268 0.0503 0.0229 0.0238 1.0569 0.0011 0.0271 Sn 0.0000 0.0008 1.9589 2.4374 0.0128 0.0489 0.0000 Ti 0.0024 0.0035 0.0446 0.0447 0.0446 0.0028 0.0028 V - - 0.0014 0.0017 0.0014 - - W - - 0.0010 0.0011 0.0010 - - Zn - - 0.0191 0.0222 0.0191 - - Zr 0.0000 0.0440 112.6799 140.2127 0.6264 2.8164 0.0000 Environmental and Nuclear Engineering Design Calculation Title: PWR Source Term Generation and Evaluation Document Identifier: 000-00C-MGR0-00100-000-00B Page 24 of 33 5.4.5 Assembly Operating Parameters To specify the reactor moderator condition used in the pin cell calculations, the moderator temperature and density and the boron concentration are required. The moderator temperature is determined from the average core exit temperature, 612°F, and the average moderator temperature rise across the core, 59.4°F, (see Table 2) as follows: K F F F 9 . 578 3 . 582 2 4 . 59 612 . ° = .. . .. . ° - ° The density of the moderator for this temperature is then determined from steam tables by linear interpolation. Using the values corresponding to steam temperatures of 580°F and 590°F at a steam pressure of 2200 psia, (see Table 2) the specific volume is: ( ) lbm ft 022446 . 0 4 . 582 590 * 580 590 02235 . 0 02275 . 0 02275 . 0 3 = - .. . .. . - - - Converting this to density yields: 3 3 3 cm g 7136 . 0 ft lbm 551 . 44 lbm ft 022446 . 0 1 = = If more than one library per cycle is required in SAS2H calculations, the boron concentration is automatically changed during the passes of a cycle and is assumed to vary linearly from 1.9 to 0.1 times its average (input) density during the time interval of the cycle (ORNL 1997, p. S2.2.10 [DIRS 135808]). Therefore, the required input average boron concentration is calculated by dividing the beginning of cycle value provided in Table 2 by 1.9, e.g., 1050 ppm/1.9 = 552.6 ppm. 5.4.6 Power History Input Also required by SAS2H is the operational history of the assembly in the reactor. Table 12 presents the values required in Block 9 of SAS2H describing the assembly's power history. Since the histories can be very complicated, this calculation uses the simple approach of one cycle and new libraries at least every 100 days. The assemblies are irradiated at a power of 14.5085 MW (2568 MWt/ 177 assemblies in a core). These parameters are calculated in Attachment II. 5.4.7 ORIGEN-S Input The inputs for ORIGEN-S are the decay time steps shown in Table 13. Other inputs for ORIGEN-S are the specification of information tables to be printed, and how the gamma sources are generated. The gamma spectra can be calculated from the light elements, fission products and actinides of an assembly, or just from the light elements. For the hardware regions, only the light elements are used to generate the gamma source. In this calculation, the tables of curies and watts for each isotope are printed, as are the neutron and gamma spectra. Environmental and Nuclear Engineering Design Calculation Title: PWR Source Term Generation and Evaluation Document Identifier: 000-00C-MGR0-00100-000-00B Page 25 of 33 Table 12. SAS2H Input Data for Assembly Depletion/Decay ParametersX “*.cut” Files B Code Final Burnup (GWd/MTU) Effective Full Power Days U Mass (kg /assembly) Number of Libraries per Cycle B1 0.001 0.03274 475 1 B1 0.001 0.03791 550 1 B2 0.01 0.3274 475 1 B2 0.01 0.3791 550 1 B3 0.1 3.274 475 1 B3 0.1 3.791 550 1 B4 1 32.739 475 1 B4 1 37.909 550 1 B5 10 327.39 475 4 B5 10 379.09 550 4 B6 20 654.79 475 7 B6 20 758.18 550 8 B7 30 982.18 475 10 B7 30 1137.26 550 12 B8 40 1309.58 475 14 B8 40 1516.35 550 16 B9 48.086 1574.31 475 16 B9 48.086 1822.88 550 19 B10 50 1636.97 475 17 B10 50 1895.44 550 19 B11 60 1964.37 475 20 B11 60 2274.53 550 23 B12 70 2291.76 475 23 B12 70 2653.62 550 27 B13 75 2455.46 475 25 B13 75 2843.16 550 29 B14 80 2619.15 475 27 B15 85 2782.85 475 28 Environmental and Nuclear Engineering Design Calculation Title: PWR Source Term Generation and Evaluation Document Identifier: 000-00C-MGR0-00100-000-00B Page 26 of 33 Table 13. Time Steps Used for Decay Calculation Decay time steps (years) 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 16 17 18 19 20 21 22 23 24 25 26 27 28 29 30 31 32 33 34 35 36 37 38 39 40 41 42 43 44 45 46 47 48 49 50 51 52 53 54 55 56 57 58 59 60 61 62 63 64 65 66 67 68 69 70 71 72 73 74 75 76 77 78 79 80 81 82 83 84 85 86 87 88 89 90 91 92 93 94 95 96 97 98 99 100 110 120 130 140 150 160 170 180 190 200 250 300 350 400 450 500 550 600 650 700 750 800 850 900 950 1000 1500 2000 2500 3000 3500 4000 4500 5000 5500 6000 6500 7000 7500 8000 8500 9000 9500 1.0x104 1.5x104 2.0x104 2.5x104 3.0x104 3.5x104 4.0x104 4.5x104 5.0x104 5.5x104 6.0x104 6.5x104 7.0x104 7.5x104 8.0x104 8.5x104 9.0x104 9.5x104 1.0x105 1.5x105 2.0x105 2.5x105 3.0x105 3.5x105 4.0x105 4.5x105 5.0x105 5.5x105 6.0x105 6.5x105 7.0x105 7.5x105 8.0x105 8.5x105 9.0x105 9.505 1.0x106 5.5 RADIONUCLIDE INVENTORIES FOR PERFORMANCE ASSESSMENT Several separate SAS2H/ORIGEN-S cases are provided in this calculation to determine average and maximum radionuclide inventories for specific years. The average and maximum PWR assemblies are derived from the results of CRWMS M&O 2000 [DIRS 138239] and listed below. The characteristics of the average PWR assembly are estimated based on the average PWR assembly of Case A with full inventory (83,800 MTU) in CRWMS M&O 2000, Table 5 [DIRS 138239]. For this case, the characteristics of the average PWR assembly are 3.75wt%, 41.70 GWd/MTU, and 25.3 years old with an initial uranium loading of 434 kg. A comparison study, based on the data base in DOE 1992, Appendix 1C [DIRS 102812], indicates that per initial MTU loading the average PWR assembly selected in this calculation is more conservative than any average PWR assembly for the scenarios in CRWMS M&O 2000, Table 5 [DIRS 138239] (See Attachment IV). The characteristics of the maximum PWR assembly are also derived from CRWMS M&O 2000, Attachment III, preblend files [DIRS 138239]. From these files the following maximum PWR SNF characteristics are noted: initial uranium loading of 477 kg, burnup of 69 GWd/MTU, initial uranium enrichment of 5.0 wt%, and cooling time of 5 years. It should be noted that there is no single assembly in the waste stream with these combined characteristics. Rather, these are the maximum characteristics of each parameter in the entire waste stream. Compared to these SNF characteristics, the maximum PWR assembly selected here is more conservative because the effect of the higher burnup (80 GWd/MTU versus 69 GWd/MTU) exceed that of the two kilogram (475 versus 477 kg) difference in the initial uranium loading. Environmental and Nuclear Engineering Design Calculation Title: PWR Source Term Generation and Evaluation Document Identifier: 000-00C-MGR0-00100-000-00B Page 27 of 33 To provide margin for conservatism and sufficiently cover future projected commercial high burnup fuel, the maximum burnup of 69 GWd/MTU based on the Waste Packages and Source Terms for the Commercial 1999 Design Basis Waste Streams (CRWMS M&O 2000, Attachment III, preblend files [DIRS 138239]) was increased to 80 GWd/MTU as the bounding burnup value. This selection is consistent with the value used in Licensing Position-009, Waste Stream Parameters (Williams 2003, Attachment, p. 1 [DIRS 166132]). The characteristics of the average and maximum PWR SNF assemblies used in this calculation are: • Average PWR assembly: 4.0%, 48 GWd/MTU, 25 years cooling time • Maximum PWR assembly: 5.0%, 80 GWd/MTU, 5 years cooling time The radionuclide inventories (curies) for both average and maximum PWR assemblies are provided in Attachment IX for different cooling times. 5.6 CRUD SOURCE TERM The activity of the crud on the surface of the PWR assemblies at time zero is determined by multiplying the calculated assembly surface area by the 60Co or other corrosion product activity per unit area of surface. A time dependent crud activity is derived using the following equation: where t½ is the radionuclide half-life and t is the decay time in years. A bounding estimate of the PWR assembly surface area is 449,003 cm2, based on a South Texas assembly. This value and the crud source term are calculated in Attachment VI. 2 / 1 t 2 ln t e ) 0 ( N ) t ( N - = Environmental and Nuclear Engineering Design Calculation Title: PWR Source Term Generation and Evaluation Document Identifier: 000-00C-MGR0-00100-000-00B Page 28 of 33 6. RESULTS The section presents the results of this calculation. The outputs of this calculation are reasonable compared to the inputs, and the results are suitable for the intended use. The uncertainties are taken into account by consistently using the most conservative approach; the calculations, therefore, yield a conservatively bounding set of results. 6.1 EFFECTS OF FUEL IMPURITIES The effects of impurities in the fuel are examined using a waste stream assembly with a 4.2 wt% 235U initial enrichment. Negligible differences in radiation spectra and thermal powers have been observed. However, the results for 36Cl and 14C are summarized in this section. Attention is given to 36Cl and 14C (with half-lives of 3.01x105 and 5715 years (Parrington et.al. 1996, p. 19 and p. 22 [DIRS 103896]), respectively) because of their very high solubility-limits in aqueous concentrations (CRWMS M&O 1995, p. 6-7 [DIRS 100198]). 36Cl and 14C activities at discharge as function of burnup are presented in Table 14. Two mass concentrations for the chlorine impurity have been used in these calculations: 5.3 ppm (a measured concentration from Ludwig and Renier, Table 5.4 [DIRS 146398]) and 10 ppm. Figure 3 presents a plot of 36Cl and 14C activity per ppm versus burnup. This conclusion can be used to estimate the resulting radioactivity from activation of a certain chlorine impurity concentration. Table 14. 36Cl and 14C Activities for the Waste Stream Assembly with 4.2 wt% 235U Enrichment at Time of Discharge 36Cl Activity (Ci/assembly) 36Cl Activity (Ci/assembly)/Initial Impurity (ppm) Burnup (GWd/MTU) 5.3 ppm 10 ppm 5.3 ppm 10 ppm 14C Activity (Ci/assembly) 14C Activity (Ci/assembly)/ Initial Impurity (ppm) 1 1.24x10-4 2.35x10-4 2.34x10-5 2.35x10-5 5.67x10-3 6.34x10-5 10 1.24x10-3 2.33x10-3 2.34x10-4 2.33x10-4 5.69x10-2 6.36x10-4 20 2.51x10-3 4.74x10-3 4.74x10-4 4.74x10-4 1.17x10-1 1.31x10-3 30 3.87x10-3 7.32x10-3 7.30x10-4 7.32x10-4 1.87x10-1 2.09x10-3 40 5.32x10-3 1.01x10-2 1.00x10-3 1.01x10-3 2.54x10-1 2.84x10-3 50 6.89x10-3 1.31x10-2 1.30x10-3 1.31x10-3 3.33x10-1 3.72x10-3 60 8.55x10-3 1.62x10-2 1.61x10-3 1.62x10-3 4.20x10-1 4.70x10-3 70 1.03x10-2 1.95x10-2 1.94x10-3 1.95x10-3 5.13x10-1 5.74x10-3 75 1.11x10-2 2.12x10-2 2.09x10-3 2.12x10-3 5.62x10-1 6.29x10-3 80 1.19x10-2 - 2.25x10-3 - 6.11x10-1 6.83x10-3 Environmental and Nuclear Engineering Design Calculation Title: PWR Source Term Generation and Evaluation Document Identifier: 000-00C-MGR0-00100-000-00B Page 29 of 33 1.E-05 1.E-04 1.E-03 1.E-02 0 10 20 30 40 50 60 70 80 90 Burnup (GWd/MTU) Activity per Unit Impurity (Ci/ppm) chlorine 36 (5.3 ppm) chlorine 36 (10 ppm) carbon 14 Figure 3. 36Cl and 14C Activity per Unit Impurity at Discharge as a Function of Burnup for a Waste Stream Assembly with 4.2 wt% 235U Enrichment 6.2 WASTE STREAM, SS CLAD, AND SOUTH TEXAS SOURCE TERMS Due to large amount of information generated by this calculation, the results are provided as electronic files on four CDs (Attachment X). However, a summary of the results are presented also in Section 6 and Attachments IV, VI, VII, and IX. Attachment XI presents information to be used in License Application – Safety Analysis Report, Section 1.5.1. 6.3 STUDY LIMITATIONS Data evaluated in this calculation are the assembly average source terms and it will be left to the subsequent analysis to account for any axial distribution of the radionuclide inventories and consequently radiation and heat source terms. 6.4 CONCLUSIONS The objectives of this calculation are to generate PWR SNF assembly source terms that bound selected groupings of PWR assemblies, with regard to assembly burnup and cooling time. Bounding PWR SNF assembly source terms to be used for evaluation of shielding requirements or other follow-on analysis have been calculated in this study. The results of this calculation may also be used as input for Preclosure Safety Analysis and Total System Performance Assessment analyses. Environmental and Nuclear Engineering Design Calculation Title: PWR Source Term Generation and Evaluation Document Identifier: 000-00C-MGR0-00100-000-00B Page 30 of 33 7. REFERENCES AP-3.12Q, Rev. 2, ICN 2. Design Calculations and Analyses. Washington, D.C.: U.S. Department of Energy, Office of Civilian Radioactive Waste Management. ACC: DOC.20040318.0002. [DIRS 168413] AP-3.13Q, Rev. 3, ICN 3. Design Control. Washington, D.C.: U.S. Department of Energy, Office of Civilian Radioactive Waste Management. ACC: DOC.20040202.0006. [DIRS 167460] ASM (American Society for Metals) 1980. Properties and Selection: Stainless Steels, Tool Materials and Special-Purpose Metals. Volume 3 of Metals Handbook. 9th Edition. Benjamin, D., ed. Metals Park, Ohio: American Society for Metals. TIC: 209801. [DIRS 104317] ASME (American Society of Mechanical Engineers) 2001. 2001 ASME Boiler and Pressure Vessel Code (includes 2002 addenda). New York, New York: American Society of Mechanical Engineers. 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[DIRS 123796] BSC (Bechtel SAIC Company) 2003. BWR Source Term Generation and Evaluation. 000-00CMGR0- 00200-000-00A. Las Vegas, Nevada: Bechtel SAIC Company. ACC: ENG.20030723.0001. [DIRS 164364] BSC (Bechtel SAIC Company) 2003. PWR Source Term Generation and Evaluation. 000-00CMGR0- 00100-000-00A. Las Vegas, Nevada: Bechtel SAIC Company. ACC: ENG.20030324.0004. [DIRS 162471] BSC (Bechtel SAIC Company) 2003. Q-List. TDR-MGR-RL-000005 REV 00. Las Vegas, Nevada: Bechtel SAIC Company. ACC: DOC.20030930.0002. [DIRS 165179] BSC (Bechtel SAIC Company) 2003. Technical Product Review Summary for PWR Source Term Generation and Evaluation. 000-00C-MGR0-00100-000-00A. Las Vegas, Nevada: Bechtel SAIC Company. ACC: MOL.20030916.0015. [DIRS 165684] Environmental and Nuclear Engineering Design Calculation Title: PWR Source Term Generation and Evaluation Document Identifier: 000-00C-MGR0-00100-000-00B Page 31 of 33 BSC (Bechtel SAIC Company) 2004. PWR and BWR Source Term Sensitivity Study. 000-00CMGR0- 00300-000-00A. Las Vegas, Nevada: Bechtel SAIC Company. ACC: ENG.20040114.0003. [DIRS 167058] CRWMS M&O 1995. Total System Performance Assessment - 1995: An Evaluation of the Potential Yucca Mountain Repository. B00000000-01717-2200-00136 REV 01. Las Vegas, Nevada: CRWMS M&O. ACC: MOL.19960724.0188. [DIRS 100198] CRWMS M&O 1997. PWR Assembly Source Terms for Waste Package Design. BBA000000- 01717-0200-00023 REV 00. Las Vegas, Nevada: CRWMS M&O. ACC: MOL.19970708.0431. [DIRS 136439] CRWMS M&O 1997. Software Code: SCALE. V4.3. HP. 30011 V4.3. [DIRS 154059] CRWMS M&O 1997. Software Qualification Report for THE SCALE Modular Code System Version 4.3. CSCI: 30011 V4.3. DI: 30011-2002, Rev. 01. Las Vegas, Nevada: CRWMS M&O. ACC: MOL.19970731.0884. [DIRS 106016] CRWMS M&O 1998. Source Term Generation and Shielding Analysis for the 21 and 12 PWR Waste Packages. BBAC00000-01717-0210-00001 REV 00. Las Vegas, Nevada: CRWMS M&O. ACC: MOL.19980824.0256. [DIRS 124815] CRWMS M&O 1999. PWR Source Term Generation and Evaluation. BBAC00000-01717- 0210-00010 REV 00. Las Vegas, Nevada: CRWMS M&O. ACC: MOL.19990215.0395. [DIRS 105912] CRWMS M&O 1999. PWR Source Term Generation and Evaluation. BBAC00000-01717- 0210-00010 REV 01. Las Vegas, Nevada: CRWMS M&O. ACC: MOL.20000113.0333. [DIRS 136429] CRWMS M&O 2000. Users Manual for SCALE-4.4A. 10129-UM-4.4A-00. Las Vegas, Nevada: CRWMS M&O. ACC: MOL.20001130.0136; MOL.20010822.0084; MOL.20010822.0085; MOL.20010822.0086; MOL.20010822.0087. [DIRS 153872] CRWMS M&O 2000. Waste Packages and Source Terms for the Commercial 1999 Design Basis Waste Streams. CAL-MGR-MD-000001 REV 00. Las Vegas, Nevada: CRWMS M&O. ACC: MOL.20000214.0479. [DIRS 138239] DOE (U.S. Department of Energy) 1992. Characteristics of Potential Repository Wastes. DOE/RW-0184-R1. Volume 1. Washington, D.C.: U.S. Department of Energy, Office of Civilian Radioactive Waste Management. ACC: HQO.19920827.0001. [DIRS 102812] DOE (U.S. Department of Energy) 1988. Characteristics of Spent Fuel, High-Level Waste, and Other Radioactive Wastes Which May Require Long-Term Isolation. DOE/RW-0184. Washington, D.C.: U.S. Department of Energy, Office of Civilian Radioactive Waste Management. TIC: 202243. [DIRS 100979] DOE (U.S. Department of Energy) 2004. Quality Assurance Requirements and Description. DOE/RW-0333P, Rev. 14. Washington, D.C.: U.S. Department of Energy, Office of Civilian Radioactive Waste Management. ACC: DOC.20040331.0004. [DIRS 168669] Framatome Cogema Fuels. 1999. Operational Data - B&W NSS. 51-5004349-01. Lynchburg, Virginia: Framatome Cogema Fuels. ACC: MOL.19990610.0201. [DIRS 146419] Environmental and Nuclear Engineering Design Calculation Title: PWR Source Term Generation and Evaluation Document Identifier: 000-00C-MGR0-00100-000-00B Page 32 of 33 Hermann, O.W.; Parks, C.V.; and Renier, J.P. 1994. Technical Support for a Proposed Decay Heat Guide Using SAS2H/ORIGEN-S Data. NUREG/CR-5625. Washington, D.C.: U.S. Nuclear Regulatory Commission. TIC: 212756. [DIRS 154045] Inco Alloys International. 1988. Product Handbook. Huntington, West Virginia: Inco Alloys International. TIC: 239397. [DIRS 130835] Jones, R.H. 1992. Spent Fuel Corrosion Product and Fuel Cleaning Assessment. Los Gatos, California: Robert H. Jones, P.E., Consultant. ACC: HQX.19920825.0007. [DIRS 146405] LP-SI.11Q-BSC, Rev. 0, ICN 0. Software Management. Washington, D.C.: U.S. Department of Energy, Office of Civilian Radioactive Waste Management. ACC: DOC.20040225.0007. [DIRS 168412] Ludwig, S.B. and Renier, J.P. 1989. Standard- and Extended-Burnup PWR and BWR Reactor Models for the ORIGEN2 Computer Code. ORNL/TM-11018. Oak Ridge, Tennessee: Oak Ridge National Laboratory. TIC: 203557. [DIRS 146398] Luksic, A. 1989. Activation Measurements and Comparison with Calculations for Spent Fuel Assembly Hardware. Volume 1 of Spent Fuel Assembly Hardware: Characterization and 10 CFR 61 Classification for Waste Disposal. PNL-6906. Richland, Washington: Pacific Northwest Laboratory. ACC: NNA.19890926.0124. [DIRS 120506] MO0105SPACR300.002. CR3 NEMO Depl and Statepoints. Submittal date: 05/22/2001. [DIRS 154999] MO0105SPACR301.003. CR3 NEMO Depl and Sp. Submittal date: 05/22/2001. [DIRS 155000] MO0105SPACR302.004. CR3 NEMO Depletion and State Points. Submittal date: 05/22/2001. [DIRS 155001] MO0105SPACR303.001. CR3 NEMO Depletion and Statepoints. Submittal date: 05/22/2001. [DIRS 155002] NRC (U.S. Nuclear Regulatory Commission) 2000. Standard Review Plan for Spent Fuel Dry Storage Facilities. NUREG-1567. Washington, D.C.: U.S. Nuclear Regulatory Commission. TIC: 247929. [DIRS 149756] ORNL (Oak Ridge National Laboratory) 1997. SCALE: A Modular Code System for Performing Standardized Computer Analyses for Licensing Evaluation. NUREG/CR-0200, Rev. 5. Washington, D.C.: U.S. Nuclear Regulatory Commission. TIC: 235920. [DIRS 135808] Parrington, J.R.; Knox, H.D.; Breneman, S.L.; Baum, E.M.; and Feiner, F. 1996. Nuclides and Isotopes, Chart of the Nuclides. 15th Edition. San Jose, California: General Electric Company and KAPL, Inc. TIC: 233705. [DIRS 103896] Punatar, M.K. 2001. Summary Report of Commercial Reactor Criticality Data for Crystal River Unit 3. TDR-UDC-NU-000001 REV 02. Las Vegas, Nevada: Bechtel SAIC Company. ACC: MOL.20010702.0087. [DIRS 155635] Williams, N.H. 2003. "Contract No. DE-AC28-01RW12101 - Licensing Position-009, Waste Stream Parameters." Letter from N.H. Williams (BSC) to J.D. Ziegler (DOE/ORD), November 13, 2003, 1105039412, with enclosure. ACC: MOL.20031215.0076. [DIRS 166132] Environmental and Nuclear Engineering Design Calculation Title: PWR Source Term Generation and Evaluation Document Identifier: 000-00C-MGR0-00100-000-00B Page 33 of 33 8. ATTACHMENTS The ten attachments for this calculation are listed in Table 15. Attachments I to IX are included with this document. Attachment X consists of four compact discs, which content is listed in Attachment VIII. Two new worksheets “PWR.bou.2.curies” and “PWR.bou.3.curies” were added to the spreadsheet “ATTACH_IX.XLS” to include the new burnup value (80 GWd/MTU) for the maximum PWR assembly. Table 15. Attachments Description Attachment Number No. of Pages Fuel density and composition calculations I 1 Burn history calculations II 2 Light element mass calculations III 12 Comparison of source terms per MTU of 4 average PWR SNF assemblies IV 1 Listing of UNIX script files V 1 Calculation of crud source VI 4 36Cl and 14C calculations VII 1 List of *.cut files on compact discs VIII 22 Radionuclide inventories for Performance Assessment (all nuclides included) IX 10 Electronic copies versions of *.cut files X 4 CDs Evolution in Time of Thermal Power and Total Radioactivity XI 1 Environmental and Nuclear Engineering Design Calculation Title: PWR Source Term Generation and Evaluation Document Identifier: 000-00C-MGR0-00100-000-00B Page I-1 of I-1 Attachment I Determination of fuel compositions and burn histories Initial Heavy Metal Mass # of pins in assembly Pin length pellet outer radius 464 208 360.172 0.46812 475 550 KG UO2/Assembly densities (g/cc) cross sect. area (cm^2) volume needed (cm^3) density (g/cc) new fuel length (cm) 527.27 10.22 143.1949164 52939.03 9.96 369.699 539.77 10.47 54194.05 378.463 625 12.12 62751.00 438.221 Burn histories Burnup,GWd/MTU days for 464 assembly to achieve burnup Required libraries/cycle for 100 day cycles days for 475 assembly to achieve burnup Required libraries/cycle for 100 day cycles days for 550 assembly to achieve burnup Required libraries/cycle for 100 day cycles 0.001 0.03 0.0003 0.0327 0.0003 0.0379 0.0004 0.01 0.32 0.0032 0.3274 0.0033 0.3791 0.0038 0.1 3.20 0.0320 3.2739 0.0327 3.7909 0.0379 1 31.98 0.3198 32.7394 0.3274 37.9088 0.3791 10 319.81 3.1981 327.39 3.3 379.09 3.8 20 639.63 6.4 654.79 6.5 758.18 7.6 30 959.44 9.6 982.18 9.8 1137.26 11.4 40 1279.25 12.8 1309.58 13.1 1516.35 15.2 50 1599.06 16.0 1636.97 16.4 1895.44 19.0 60 1918.88 19.2 1964.37 19.6 2274.53 22.7 70 2238.69 22.4 2291.76 22.9 2653.62 26.5 75 2398.59 24.0 2455.46 24.6 2843.16 28.4 80 2558.50 25.6 2619.15 26.2 3032.70 30.3 85 2718.41 27.2 2782.85 27.8 3222.25 32.2 235 Enrichment 236 enrichment 234 enrichment 238 0.711 0.00327 0.00534 99.28039 1 0.00460 0.00773 98.98767 1.5 0.00690 0.01200 98.48110 2 0.00920 0.01639 97.97441 2.5 0.01150 0.02087 97.46763 3 0.01380 0.02543 96.96077 3.5 0.01610 0.03005 96.45385 4 0.01840 0.03473 95.94687 4.2 0.01932 0.03661 95.74407 4.5 0.02070 0.03946 95.43984 5 0.02300 0.04423 94.93277 5.5 0.02530 0.04904 94.42566 Environmental and Nuclear Engineering Design Calculation Title: PWR Source Term Generation and Evaluation Document Identifier: 000-00C-MGR0-00100-000-00B Page II-1 of II-2 Attachment II PWR Source Term Generation and Evaluation Final Burnup (GWd/MTU) Effective Full Power Days MW /assembly kg /assembly Number of libraries for 1 library every 100 days #days covered by last library Nlib/cycle required 0.001 0.03274 14.5085 475 0 0.03274 1 0.001 0.03791 14.5085 550 0 0.03791 1 0.01 0.3274 14.5085 475 0 0.32739 1 0.01 0.3791 14.5085 550 0 0.37909 1 0.1 3.274 14.5085 475 0 3.27394 1 0.1 3.791 14.5085 550 0 3.79088 1 1 32.739 14.5085 475 0 32.73943 1 1 37.909 14.5085 550 0 37.90881 1 10 319.81 14.5085 464 3 19.8 4 10 327.39 14.5085 475 3 27.4 4 10 379.09 14.5085 550 3 79.1 4 20 639.63 14.5085 464 6 39.6 7 20 654.79 14.5085 475 6 54.8 7 20 758.18 14.5085 550 7 58.2 8 30 959.44 14.5085 464 9 59.4 10 30 982.18 14.5085 475 9 82.2 10 30 1137.26 14.5085 550 11 37.3 12 Environmental and Nuclear Engineering Design Calculation Title: PWR Source Term Generation and Evaluation Document Identifier: 000-00C-MGR0-00100-000-00B Page II-2 of II-2 Attachment II Final Burnup (GWd/MTU) Effective Full Power Days MW /assembly kg /assembly Number of libraries for 1 library every 100 days #days covered by last library Nlib/cycle required 40 1279.25 14.5085 464 12 79.3 13 40 1309.58 14.5085 475 13 9.6 14 40 1516.35 14.5085 550 15 16.4 16 48.086 1537.85 14.5085 464 15 37.9 16 48.086 1574.31 14.5085 475 15 74.3 16 48.086 1822.88 14.5085 550 18 22.9 19 50 1599.06 14.5085 464 15 99.1 16 50 1636.97 14.5085 475 16 37.0 17 50 1895.44 14.5085 550 18 95.4 19 60 1918.88 14.5085 464 19 18.9 20 60 1964.37 14.5085 475 19 64.4 20 60 2274.53 14.5085 550 22 74.5 23 70 2238.69 14.5085 464 22 38.7 23 70 2291.76 14.5085 475 22 91.8 23 70 2653.62 14.5085 550 26 53.6 27 75 2398.59 14.5085 464 23 98.6 24 75 2455.46 14.5085 475 24 55.5 25 75 2843.16 14.5085 550 28 43.2 29 80 2558.50 14.5085 464 25 58.5 26 80 2619.15 14.5085 475 26 19.2 27 80 3032.70 14.5085 550 30 32.7 31 85 2718.41 14.5085 464 27 18.4 28 85 2782.85 14.5085 475 27 82.9 28 85 3222.25 14.5085 550 32 22.2 33 Environmental and Nuclear Engineering Design Calculation Title: PWR Source Term Generation and Evaluation Document Identifier: 000-00C-MGR0-00100-000-00B Page III-1 of III-12 Attachment III - Worksheet "Data" Babcock and Wilcox Mark B fuel assembly Hardware Region Part Name kg/assembly Material top nozzle 7.48 SS CF3M ***Mass of oxygen to be added to fuel region light elements to acount for the oxygen in the fuel: spring retainer 0.91 SS CF3M KG U KG UO2 KG O Top hold down spring 1.8 Inconel-718 475 539.77 64.77 upper end plug 0.06 SS 304 550 625 75.00 upper nuts 0.51 SS 304L (KG U)/0.88 (KG UO2)- (KG U) bottom nozzle 8.16 SS CF3M Bottom bottom spacer 1.3 Inconel-718 lower nuts 0.15 Zircaloy-4 Fuel/plenum guide tubes 8 Zircaloy-4 Fuel/plenum instrument tube 0.64 Zircaloy-4 *** Cladding mass is calculated on next worksheet Fuel/plenum incore spacers 4.9 Zircaloy-4 Inconel 718 used in calculation since many have this material for spacers Fuel grid supports 0.64 Zircaloy-4 plenum plenum spacers 1.04 Inconel-718 plenum springs 0.019 SS 302 0.042 (lbs) RAW MATERIAL DATA: Updated 8/17 Material: SS 304 SS304L Zircaloy-4 Inconel-718 SS CF3M ss302 Element Range Used Range Used Range Used Element Range Used Element Range Used Element Range Used c 0.08 (max) 0.08 0.03 (max) 0.03 0 0 ni+co 50-55 54 c 0.03 0.03 c 0.15 0.15 mn 2 (max) 2 2 (max) 2 0 0 cr 17-21 19 mn 1.5 1.5 mn 2 2 p 0.045 (max) 0.045 0.045 (max) 0.045 0 0 fe bal 14.934 si 2 2 p 0.045 0.045 s 0.03 (max) 0.03 0.03 (max) 0.03 0 0 co (max) 1 1 cr 17-21 19 s 0.03 0.03 si 1.0 (max) 0.75 0.75 (max) 0.75 0 0 mo 2.8-3.3 3.05 ni 8.-12 11.92 si 0.75 0.75 cr 18-20 19 18-20 19 0.07-0.13 0.1 w 0 0 mo 2.-3 2.5 cr 17-19 18 ni 8-10.5 10.42 8.0-12.0 11.92 0 0 nb+ta 4.75-5.5 5.5 fe balance 62.97 ni 8.-10 9.92 n 0.1 (max) 0.1 0.1 0.1 0 0 ti .65-1.15 0.9 99.92 mo 0 0 fe Balance 67.495 Balance 66.045 0.18-0.24 0.2 al .2-.8 0.5 co 0.08 n 0.1 0.1 o 0 0 0 0 0.09-0.16 0.12 c 0.08 0.08 cu 0 0 sn 0 0 0 0 1.2-1.7 1.7 mn 0.35 0.35 fe balance 68.925 zr 0 0 0 0 Balance 97.88 si 0.35 0.35 99.92 99.92 99.92 100 b 0.006 0.006 co - 0.08 - 0.08 cu 0.3 0.3 co 0.08 Material: SS 304 SS304L Zircaloy-4 p 0.015 0.015 s 0.015 0.015 100 REORGANIZED DATA BY ELEMENT: Material: SS 304 SS304L Zircaloy-4 Inconel-718 SS CF3M ss302 ***Here the assumption of a 0.08wt% Cobalt impurity in the steels. Element al 0 0 0 0.5 0 0 b 0 0 0 0.006 0 0 c 0.08 0.03 0 0.08 0.03 0.15 co 0.08 0.08 0 1 0.08 0.08 cr 19 19 0.1 19 19 18 cu 0 0 0 0.3 0 0 fe 67.495 66.045 0.2 14.934 62.97 68.925 mn 2 2 0 0.35 1.5 2 mo 0 0 0 3.05 2.5 0 n 0.1 0.1 0 0 0 0.1 nb+ta 0 0 0 5.5 0 0 ni 10.42 11.92 0 54 11.92 9.92 o 0 0 0.12 0 0 0 p 0.045 0.045 0 0.015 0 0.045 s 0.03 0.03 0 0.015 0 0.03 si 0.75 0.75 0 0.35 2 0.75 sn 0 0 1.7 0 0 0 ti 0 0 0 0.9 0 0 zr 0 0 97.88 0 0 0 100 100 100 100 100 100 Environmental and Nuclear Engineering Design Calculation Title: PWR Source Term Generation and Evaluation Document Identifier: 000-00C-MGR0-00100-000-00B Page III-2 of III-12 Attachment III – Worksheet “impurities.test.cases” PWR U mass (g) 475000 PWR mass (kg) 539.7727273 PWR assembly volume (cc) 54194.05 PWR Impurities atomic mass (g) ppm/U mass (kg) at dens Li 6.941 1 0.000475 7.6043E-07 B 10.811 1 0.000475 4.8822E-07 C 12.0107 89.4 0.042465 3.9287E-05 N 14.00674 25 0.011875 9.4208E-06 F 18.9984032 10.7 0.0050825 2.9727E-06 Na 22.98977 15 0.007125 3.4438E-06 Mg 24.305 2 0.00095 4.3433E-07 Al 26.981538 16.7 0.0079325 3.2669E-06 Si 28.0855 12.1 0.0057475 2.2740E-06 P 30.973761 35 0.016625 5.9643E-06 Cl 35.4527 5.3 0.0025175 7.8906E-07 Ca 40.078 2 0.00095 2.6339E-07 Ti 47.867 1 0.000475 1.1027E-07 V 50.9415 3 0.001425 3.1084E-07 Cr 51.9961 4 0.0019 4.0604E-07 Mn 54.938049 1.7 0.0008075 1.6333E-07 Fe 55.845 18 0.00855 1.7013E-06 Co 58.9332 1 0.000475 8.9562E-08 Ni 58.6934 24 0.0114 2.1583E-06 Cu 63.546 1 0.000475 8.3060E-08 VF Zn 65.39 40.3 0.0191425 3.2529E-06 0.000035464 Mo 95.94 10 0.00475 5.5015E-07 Ag 107.8682 0.1 0.0000475 4.8932E-09 Cd 112.411 25 0.011875 1.1739E-06 In 114.818 2 0.00095 9.1940E-08 Sn 118.71 4 0.0019 1.7785E-07 W 183.84 2 0.00095 5.7421E-08 Pb 207.2 1 0.000475 2.5474E-08 Bi 208.98038 0.4 0.00019 1.0103E-08 Total 0.1680075 mass (g) at. Dens. cl 10 4.75 1.48879E-06 Environmental and Nuclear Engineering Design Calculation Title: PWR Source Term Generation and Evaluation Document Identifier: 000-00C-MGR0-00100-000-00B Page III-3 of III-12 Attachment III - Worksheet “impurities.test.cases” Light Elements (PWR) From hardware Add impurities Zn mass (g) 19.1425 element mass (kg) mass (kg) at mass (g) wt fr at dens al 0.0245 0.0324 Zn-64 63.929146 0.4751 1.5808E-06 b 0.0003 0.0008 Zn-66 65.926036 0.28126 9.0748E-07 c 0.0039 0.0464 Zn-67 66.927131 0.04196 1.3336E-07 co 0.0490 0.0495 Zn-68 67.924847 0.19527 6.1150E-07 cr 1.0461 1.0480 Zn-70 69.925325 0.00642 1.9529E-08 cu 0.0147 0.0152 1.00001 3.2527E-06 fe 0.9819 0.9904 Ag mass (g) 0.0475 mn 0.0172 0.0180 at mass (g) wt fr mo 0.1495 0.1542 Ag-107 106.90509 0.51377 2.5366E-09 n 0.0000 0.0119 Ag-109 108.904756 0.48623 2.3566E-09 nb 0.2511 0.2511 1 4.8932E-09 ni 2.6460 2.6574 In mass (g) 0.95 o 0.1381 64.9081 at mass (g) wt fr p 0.0000 0.0166 In-113 112.904062 0.04228 3.9531E-09 s 0.0000 0.0000 In-115 114.903879 0.95772 8.7987E-08 si 0.0172 0.0229 1 9.1940E-08 sn 1.6117 1.6136 Sn mass (g) 1.9 ti 0.0441 0.0446 at mass (g) wt fr zr 113.0252 113.0252 Sn-112 111.90482 0.00914 1.7244E-09 li 0.000475 Sn-114 113.902782 0.00624 1.1566E-09 f 0.0050825 Sn-115 114.903347 0.00348 6.3942E-10 na 0.007125 Sn-116 115.901745 0.14186 2.5841E-08 mg 0.00095 Sn-117 116.902955 0.07563 1.3659E-08 cl 0.0025175 Sn-118 117.901608 0.24055 4.3075E-08 ca 0.00095 Sn-119 118.903311 0.08594 1.5260E-08 v 0.001425 Sn-120 119.902198 0.32917 5.7961E-08 zn 0.0191425 Sn-122 121.903441 0.04755 8.2353E-09 ag 0.0000475 Sn-124 123.905274 0.06043 1.0297E-08 cd 0.011875 0.99999 1.7785E-07 in 0.00095 w 0.00095 B mass (g) 0.475 pb 0.000475 at mass (g) wt fr bi 0.00019 b-10 10.0129371 0.18431 9.71561E-08 b-11 11.0093055 0.81569 3.91064E-07 4.8822E-07 Li mass (g) 0.475 at mass (g) wt fr li-6 6.015122 0.065 5.70363E-08 li-7 7.016004 0.935 7.03404E-07 7.6044E-07 Environmental and Nuclear Engineering Design Calculation Title: PWR Source Term Generation and Evaluation Document Identifier: 000-00C-MGR0-00100-000-00B Page III-4 of III-12 Attachment III - Worksheet “South.Texas.impurities” South Texas impurities and number densities 539.7727273 PWR U mass (g) 550000 PWR mass (kg) 625 PWR assembly volume (cc) 62751.00402 PWR Impurities atomic mass (g) ppm/U mass (kg) at dens Li 6.941 1 0.00055 7.6043E-07 B 10.811 1 0.00055 4.8822E-07 C 12.0107 89.4 0.04917 3.9287E-05 N 14.00674 25 0.01375 9.4208E-06 F 18.9984032 10.7 0.005885 2.9727E-06 Na 22.98977 15 0.00825 3.4438E-06 Mg 24.305 2 0.0011 4.3433E-07 Al 26.981538 16.7 0.009185 3.2669E-06 Si 28.0855 12.1 0.006655 2.2740E-06 P 30.973761 35 0.01925 5.9643E-06 Cl 35.4527 5.3 0.002915 7.8906E-07 Ca 40.078 2 0.0011 2.6339E-07 Ti 47.867 1 0.00055 1.1027E-07 V 50.9415 3 0.00165 3.1084E-07 Cr 51.9961 4 0.0022 4.0604E-07 Mn 54.938049 1.7 0.000935 1.6333E-07 Fe 55.845 18 0.0099 1.7013E-06 VF Co 58.9332 1 0.00055 8.9562E-08 0.000035464 Ni 58.6934 24 0.0132 2.1583E-06 Cu 63.546 1 0.00055 8.3061E-08 Zn 65.39 40.3 0.022165 3.2529E-06 Mo 95.94 10 0.0055 5.5015E-07 Ag 107.8682 0.1 0.000055 4.8932E-09 Cd 112.411 25 0.01375 1.1739E-06 In 114.818 2 0.0011 9.1940E-08 Sn 118.71 4 0.0022 1.7785E-07 W 183.84 2 0.0011 5.7421E-08 Pb 207.2 1 0.00055 2.5474E-08 Bi 208.98038 0.4 0.00022 1.0103E-08 Total 0.194535 mass (g) at. Dens. cl 10 5.5 1.72386E-06 Environmental and Nuclear Engineering Design Calculation Title: PWR Source Term Generation and Evaluation Document Identifier: 000-00C-MGR0-00100-000-00B Page III-5 of III-12 Attachment III - Worksheet ‘South.Texas.impurities’ Light Elements (PWR) From hardware Add impurities Zn mass (g) 22.165 element mass (kg) mass (kg) at mass (g) wt fr at dens al 0.0245 0.0337 Zn-64 63.929146 0.4751 1.5808E-06 b 0.0003 0.0008 Zn-66 65.926036 0.28126 9.0748E-07 c 0.0039 0.0531 Zn-67 66.927131 0.04196 1.3336E-07 co 0.0490 0.0496 Zn-68 67.924847 0.19527 6.1150E-07 cr 1.0461 1.0483 Zn-70 69.925325 0.00642 1.9529E-08 cu 0.0147 0.0153 1.00001 3.2527E-06 fe 0.9819 0.9918 Ag mass (g) 0.055 mn 0.0172 0.0181 at mass (g) wt fr mo 0.1495 0.1550 Ag-107 106.90509 0.51377 2.5366E-09 n 0.0000 0.0138 Ag-109 108.904756 0.48623 2.3566E-09 nb 0.2511 0.2511 1 4.8932E-09 ni 2.6460 2.6592 In mass (g) 1.1 o 0.1381 64.9081 at mass (g) wt fr p 0.0000 0.0193 In-113 112.904062 0.04228 3.9531E-09 s 0.0000 0.0000 In-115 114.903879 0.95772 8.7987E-08 si 0.0172 0.0238 1 9.1940E-08 sn 1.6117 1.6139 Sn mass (g) 2.2 ti 0.0441 0.0447 at mass (g) wt fr zr 113.0252 113.0252 Sn-112 111.90482 0.00914 1.7244E-09 li 0.000475 Sn-114 113.902782 0.00624 1.1566E-09 f 0.005885 Sn-115 114.903347 0.00348 6.3942E-10 na 0.00825 Sn-116 115.901745 0.14186 2.5841E-08 mg 0.0011 Sn-117 116.902955 0.07563 1.3659E-08 cl 0.002915 Sn-118 117.901608 0.24055 4.3075E-08 ca 0.0011 Sn-119 118.903311 0.08594 1.5260E-08 v 0.00165 Sn-120 119.902198 0.32917 5.7961E-08 zn 0.022165 Sn-122 121.903441 0.04755 8.2353E-09 ag 0.000055 Sn-124 123.905274 0.06043 1.0297E-08 cd 0.01375 0.99999 1.7785E-07 in 0.0011 w 0.0011 B mass (g) 0.55 pb 0.00055 at mass (g) wt fr bi 0.00022 b-10 10.0129371 0.18431 9.7156E-08 b-11 11.0093055 0.81569 3.9106E-07 4.8822E-07 Li mass (g) 0.55 at mass (g) wt fr li-6 6.015122 0.065 5.7036E-08 li-7 7.016004 0.935 7.034E-07 7.6044E-07 Environmental and Nuclear Engineering Design Calculation Title: PWR Source Term Generation and Evaluation Document Identifier: 000-00C-MGR0-00100-000-00B Page III-6 of III-12 Attachment III - Worksheet ‘impurities’ For volumes and weights, see Attachment I PWR U mass (g) 475000 PWR mass (kg) 539.7727273 South Texas U mass (g) 550000 South Texas mass (kg) 625 PWR assembly volume (cc) 54194.05 South Texas assembly volume (cc) 62751 PWR South Texas Impurities atomic mass (g) ppm/U mass (kg) at dens mass (kg) at dens Li 6.941 1 0.000475 7.6043E-07 0.00055 7.6043E-07 B 10.811 1 0.000475 4.8822E-07 0.00055 4.8822E-07 C 12.0107 89.4 0.042465 3.9287E-05 0.04917 3.9287E-05 N 14.00674 25 0.011875 9.4208E-06 0.01375 9.4208E-06 F 18.9984032 10.7 0.005083 2.9727E-06 0.005885 2.9727E-06 Na 22.98977 15 0.007125 3.4438E-06 0.00825 3.4438E-06 Mg 24.305 2 0.00095 4.3433E-07 0.0011 4.3433E-07 Al 26.981538 16.7 0.007933 3.2669E-06 0.009185 3.2669E-06 Si 28.0855 12.1 0.005748 2.2740E-06 0.006655 2.2740E-06 P 30.973761 35 0.016625 5.9643E-06 0.01925 5.9643E-06 Cl 35.4527 5.3 0.002518 7.8906E-07 0.002915 7.8906E-07 Ca 40.078 2 0.00095 2.6339E-07 0.0011 2.6339E-07 Ti 47.867 1 0.000475 1.1027E-07 0.00055 1.1027E-07 V 50.9415 3 0.001425 3.1084E-07 0.00165 3.1084E-07 Cr 51.9961 4 0.0019 4.0604E-07 0.0022 4.0604E-07 Mn 54.938049 1.7 0.000808 1.6333E-07 0.000935 1.6333E-07 Fe 55.845 18 0.00855 1.7013E-06 0.0099 1.7013E-06 VF Co 58.9332 1 0.000475 8.9562E-08 0.00055 8.9562E-08 3.5464E-05 Ni 58.6934 24 0.0114 2.1583E-06 0.0132 2.1583E-06 Cu 63.546 1 0.000475 8.3060E-08 0.00055 8.3061E-08 Zn 65.39 40.3 0.019143 3.2529E-06 0.022165 3.2529E-06 Mo 95.94 10 0.00475 5.5015E-07 0.0055 5.5015E-07 Ag 107.8682 0.1 4.75E-05 4.8932E-09 0.000055 4.8932E-09 Cd 112.411 25 0.011875 1.1739E-06 0.01375 1.1739E-06 In 114.818 2 0.00095 9.1940E-08 0.0011 9.1940E-08 Sn 118.71 4 0.0019 1.7785E-07 0.0022 1.7785E-07 W 183.84 2 0.00095 5.7421E-08 0.0011 5.7421E-08 Pb 207.2 1 0.000475 2.5474E-08 0.00055 2.5474E-08 Bi 208.98038 0.4 0.00019 1.0103E-08 0.00022 1.0103E-08 Total 0.168008 0.194535 Environmental and Nuclear Engineering Design Calculation Title: PWR Source Term Generation and Evaluation Document Identifier: 000-00C-MGR0-00100-000-00B Page III-7 of III-12 Attachment III - Worksheet “tables” element Waste Stream Composition (kg/assembly) fuel impurities fuel oxygen Total bottom plenum top Ag 4.75E-05 0.000048 Al 0.0245 0.007933 0.032433 0.0020 0.0016 0.0014 B 0.00029 0.000475 0.000769 0.0000 0.0000 0.0000 Bi 0.00019 0.000190 C 0.00392 0.042465 0.046385 0.0010 0.0003 0.0006 Ca 0.00095 0.000950 Cd 0.011875 0.011875 Cl 0.002518 0.002518 Co 0.04900 0.000475 0.049475 0.0059 0.0031 0.0038 Cr 1.04612 0.0019 1.048020 0.5393 0.0632 0.3067 Cu 0.01470 0.000475 0.015175 0.0012 0.0009 0.0008 F 0.005083 0.005083 Fe 0.96200 0.00855 0.970550 1.5998 0.0563 0.8894 In 0.00095 0.000950 Li 0.000475 0.000475 Mg 0.00095 0.000950 Mn 0.01715 0.000808 0.017958 0.0381 0.0012 0.0215 Mo 0.14945 0.00475 0.154200 0.0731 0.0095 0.0397 N 0.00000 0.011875 0.011875 0.0000 0.0000 0.0001 Na 0.007125 0.007125 Nb 0.26950 0.269500 0.0215 0.0172 0.0149 Ni 2.64600 0.0114 2.657400 0.5024 0.1690 0.3059 O 0.13814 64.77273 64.910872 0.0001 0.0035 0.0000 P 0.00070 0.016625 0.017325 0.0001 0.0000 0.0001 Pb 0.000475 0.000475 S 0.00070 0 0.0001 0.0000 0.0001 Si 0.01715 0.005748 0.022898 0.0503 0.0011 0.0268 Sn 1.95705 0.0019 1.958947 0.0008 0.0489 0.0000 Ti 0.04410 0.000475 0.044575 0.0035 0.0028 0.0024 V 0.001425 0.001425 W 0.00095 0.000950 Zn 0.019143 0.019143 Zr 112.67989 112.679885 0.0440 2.8164 0.0000 Environmental and Nuclear Engineering Design Calculation Title: PWR Source Term Generation and Evaluation Document Identifier: 000-00C-MGR0-00100-000-00B Page III-8 of III-12 Attachment III - Worksheet “tables” element South Texas Composition (kg/assembly) Steel Clad Composition (kg/assembly) fuel fuel oxygen impurities Total fuel hardware fuel oxygen impurities Total plenum Ag 0.000055 0.00006 4.75E-05 4.75E-05 Al 0.0245 0.009185 0.03369 0.0245 0.00793 0.03243 0.00156 B 0.0003 0.00055 0.00084 0.00029 0.00048 0.00077 0.00002 Bi 0.00022 0.00022 0.00019 0.00019 C 0.0039 0.04917 0.05309 0.11421 0.04247 0.15668 0.00303 Ca 0.0011 0.00110 0.00095 0.00095 Cd 0.01375 0.01375 0.01188 0.01188 Cl 0.002915 0.00292 0.00252 0.00252 Co 0.0490 0.00055 0.04955 0.15929 0.00048 0.15977 0.00590 Cr 1.0742 0.0022 1.07645 27.12602 0.00190 27.1279 0.71868 Cu 0.0147 0.00055 0.01525 0.01470 0.00048 0.01518 0.00094 F 0.005885 0.00589 0.00508 0.00508 Fe 1.0183 0.0099 1.02820 93.78510 0.00855 93.7937 2.38930 In 0.0011 0.00110 0.00095 0.00095 Li 0.00055 0.00055 0.00048 0.00048 Mg 0.0011 0.00110 0.00095 0.00095 Mn 0.0172 0.000935 0.01809 2.77445 0.00081 2.77526 0.07051 Mo 0.1495 0.0055 0.15495 0.14945 0.00475 0.1542 0.00952 N 0.0000 0.01375 0.01375 0.13787 0.01188 0.14974 0.00347 Na 0.00825 0.00825 0.00713 0.00713 Nb 0.2695 0.26950 0.26950 0.2695 0.01720 Ni 2.6460 0.0132 2.65920 17.01150 0.01140 17.0229 0.53010 O 0.1719 75.00000 75.17190 0.00077 64.77273 64.7735 0.00000 P 0.0007 0.01925 0.01995 0.06280 0.01663 0.07943 0.00610 Pb 0.00055 0.00055 0.00048 0.00048 S 0.0007 0 0.04210 0.0421 0.00110 Si 0.0172 0.006655 0.02381 1.05114 0.00575 1.05689 0.02712 Sn 2.4352 0.0022 2.43744 0.01088 0.00190 0.01278 0.00000 Ti 0.0441 0.00055 0.04465 0.04410 0.00048 0.04458 0.00281 V 0.00165 0.00165 0.00143 0.00143 W 0.0011 0.00110 0.00095 0.00095 Zn 0.022165 0.02217 0.01914 0.01914 Zr 140.2127 140.21265 0.62643 0.62643 0.00000 Environmental and Nuclear Engineering Design Calculation Title: PWR Source Term Generation and Evaluation Document Identifier: 000-00C-MGR0-00100-000-00B Page III-9 of III-12 Attachment III - Worksheet “Steel clad” cladding volume (cm3) 17596.28 Guide tube mass (kg) 9.634146 cladding mass (g) 139010.6 Instrument tube mass (kg) 0.770732 Materials and cladding mass updated 8/16 fuel: 92% plenum 8% Oxygen from fuel not included in these tables. See final worksheet 'tables' Region Fuel Plenum guide tubes instrument tube incore spacers cladding grid supports guide tubes instrument tube cladding spacers springs 8.889393 0.711151 4.9 128.2646 0.64 0.744753 0.05958 10.74601 1.04 0.019 SS 304 SS304 Inconel- 718 SS304 Zircaloy-4 SS 304 SS 304 SS 304 Inconel- 718 SS 302 al 0 0 0.5 0 0 0 0 0 0.5 0 b 0 0 0.006 0 0 0 0 0 0.006 0 c 0.08 0.08 0.08 0.08 0 0.08 0.08 0.08 0.08 0.15 co 0.08 0.08 1 0.08 0 0.08 0.08 0.08 1 0.08 cr 19 19 19 19 0.1 19 19 19 19 18 cu 0 0 0.3 0 0 0 0 0 0.3 0 fe 67.495 67.495 14.934 67.495 0.2 67.495 67.495 67.495 14.934 68.925 mn 2 2 0.35 2 0 2 2 2 0.35 2 mo 0 0 3.05 0 0 0 0 0 3.05 0 n 0.1 0.1 0 0.1 0 0.1 0.1 0.1 0 0.1 nb+ta 0 0 5.5 0 0 0 0 0 5.5 0 ni 10.42 10.42 54 10.42 0 10.42 10.42 10.42 54 9.92 o 0 0 0 0 0.12 0 0 0 0 0 p 0.045 0.045 0.015 0.045 0 0.045 0.045 0.045 0.015 0.045 s 0.03 0.03 0.015 0.03 0 0.03 0.03 0.03 0.015 0.03 si 0.75 0.75 0.35 0.75 0 0.75 0.75 0.75 0.35 0.75 sn 0 0 0 0 1.7 0 0 0 0 0 ti 0 0 0.9 0 0 0 0 0 0.9 0 zr 0 0 0 0 97.88 0 0 0 0 0 Fuel Plenum guide tubes instrument tube incore spacers cladding grid supports guide tubes instrument tube cladding spacers springs Element Mass Calculations 8.889393 0.711151 4.9 128.2646 0.64 0.744753 0.05958 10.74601 1.04 0.019 SS 304 SS 304 Inconel- 718 SS 304 Zircaloy-4 TOTALS SS 304 SS 304 SS 304 Inconel- 718 SS 302 TOTALS al 0.0000 0.0000 0.0245 0.0000 0.0000 0.0245 0.0000 0.0000 0.0000 0.0016 0.0000 0.0016 b 0.0000 0.0000 0.0003 0.0000 0.0000 0.0003 0.0000 0.0000 0.0000 0.0000 0.0000 0.0000 c 0.0071 0.0006 0.0039 0.1026 0.0000 0.1142 0.0002 0.0000 0.0026 0.0002 0.0000 0.0030 co 0.0071 0.0006 0.0490 0.1026 0.0000 0.1593 0.0002 0.0000 0.0026 0.0031 0.0000 0.0059 cr 1.6890 0.1351 0.9310 24.3703 0.0006 27.1260 0.0425 0.0034 0.6125 0.0593 0.0010 0.7187 cu 0.0000 0.0000 0.0147 0.0000 0.0000 0.0147 0.0000 0.0000 0.0000 0.0009 0.0000 0.0009 fe 5.9999 0.4800 0.7318 86.5722 0.0013 93.7851 0.1508 0.0121 2.1759 0.0466 0.0039 2.3893 mn 0.1778 0.0142 0.0172 2.5653 0.0000 2.7745 0.0045 0.0004 0.0645 0.0011 0.0001 0.0705 mo 0.0000 0.0000 0.1495 0.0000 0.0000 0.1495 0.0000 0.0000 0.0000 0.0095 0.0000 0.0095 n 0.0089 0.0007 0.0000 0.1283 0.0000 0.1379 0.0002 0.0000 0.0032 0.0000 0.0000 0.0035 nb+ta 0.0000 0.0000 0.2695 0.0000 0.0000 0.2695 0.0000 0.0000 0.0000 0.0172 0.0000 0.0172 ni 0.9263 0.0741 2.6460 13.3652 0.0000 17.0115 0.0233 0.0019 0.3359 0.1685 0.0006 0.5301 o 0.0000 0.0000 0.0000 0.0000 0.0008 0.0008 0.0000 0.0000 0.0000 0.0000 0.0000 0.0000 p 0.0040 0.0003 0.0007 0.0577 0.0000 0.0628 0.0001 0.0000 0.0015 0.0000 0.0000 0.0016 s 0.0027 0.0002 0.0007 0.0385 0.0000 0.0421 0.0001 0.0000 0.0010 0.0000 0.0000 0.0011 si 0.0667 0.0053 0.0172 0.9620 0.0000 1.0511 0.0017 0.0001 0.0242 0.0011 0.0000 0.0271 sn 0.0000 0.0000 0.0000 0.0000 0.0109 0.0109 0.0000 0.0000 0.0000 0.0000 0.0000 0.0000 ti 0.0000 0.0000 0.0441 0.0000 0.0000 0.0441 0.0000 0.0000 0.0000 0.0028 0.0000 0.0028 zr 0.0000 0.0000 0.0000 0.0000 0.6264 0.6264 0.0000 0.0000 0.0000 0.0000 0.0000 0.0000 TOTALS 8.889 0.711 4.900 128.265 0.640 143.405 0.223 0.018 3.224 0.312 0.006 3.783 kg*SCALING FACTOR 8.889 0.711 4.900 128.265 0.640 0.223 0.018 3.224 0.312 0.006 Environmental and Nuclear Engineering Design Calculation Title: PWR Source Term Generation and Evaluation Document Identifier: 000-00C-MGR0-00100-000-00B Page III-10 of III-12 Attachment III - Worksheet “South Texas” fuel plus plenum length 478.8452 Fractions in fuel and plenum cladding volume (CM^3) 21585.65305 fuel: 94% plenum 6% cladding mass (g) 141601.88 ****Oxygen from fuel not included in these tables. See final worksheet 'tables' South Texas/Mark B 1.22671674 Part Name guide tubes instrument tube incore spacers cladding grid supports kg/assembly 9.195305715 0.73562446 4.9 132.6786137 0.64 Material Zircaloy-4 Zircaloy-4 Inconel-718 Zircaloy-4 Zircaloy-4 al 0 0 0.5 0 0 b 0 0 0.006 0 0 c 0 0 0.08 0 0 co 0 0 1 0 0 Wt % of the constituent isotopes cr 0.1 0.1 19 0.1 0.1 cu 0 0 0.3 0 0 fe 0.2 0.2 14.934 0.2 0.2 mn 0 0 0.35 0 0 mo 0 0 3.05 0 0 n 0 0 0 0 0 nb+ta 0 0 5.5 0 0 ni 0 0 54 0 0 o 0.12 0.12 0 0.12 0.12 p 0 0 0.015 0 0 s 0 0 0.015 0 0 si 0 0 0.35 0 0 sn 1.7 1.7 0 1.7 1.7 ti 0 0 0.9 0 0 zr 97.88 97.88 0 97.88 97.88 Region Part Name guide tubes instrument tube incore spacers cladding grid supports Element Mass Calculations kg/assembly 9.195305715 0.73562446 4.9 132.6786137 0.64 Material Zircaloy-4 Inconel-718 Inconel-718 Zircaloy-4 Zircaloy-4 TOTALS al 0.0000 0.0000 0.0245 0.0000 0.0000 0.0245 b 0.0000 0.0000 0.0003 0.0000 0.0000 0.0003 c 0.0000 0.0000 0.0039 0.0000 0.0000 0.0039 co 0.0000 0.0000 0.0490 0.0000 0.0000 0.0490 cr 0.0092 0.0007 0.9310 0.1327 0.0006 1.0742 cu 0.0000 0.0000 0.0147 0.0000 0.0000 0.0147 fe 0.0184 0.0015 0.7318 0.2654 0.0013 1.0183 mn 0.0000 0.0000 0.0172 0.0000 0.0000 0.0172 mo 0.0000 0.0000 0.1495 0.0000 0.0000 0.1495 n 0.0000 0.0000 0.0000 0.0000 0.0000 0.0000 nb+ta 0.0000 0.0000 0.2695 0.0000 0.0000 0.2695 ni 0.0000 0.0000 2.6460 0.0000 0.0000 2.6460 o 0.0110 0.0009 0.0000 0.1592 0.0008 0.1719 p 0.0000 0.0000 0.0007 0.0000 0.0000 0.0007 s 0.0000 0.0000 0.0007 0.0000 0.0000 0.0007 si 0.0000 0.0000 0.0172 0.0000 0.0000 0.0172 sn 0.1563 0.0125 0.0000 2.2555 0.0109 2.4352 ti 0.0000 0.0000 0.0441 0.0000 0.0000 0.0441 zr 9.0004 0.7200 0.0000 129.8658 0.6264 140.2127 TOTALS 9.195 0.736 4.900 132.679 0.640 148.150 kg*SCALING FACTOR 9.195 0.736 4.900 0.638 0.640 16.109 Environmental and Nuclear Engineering Design Calculation Title: PWR Source Term Generation and Evaluation Document Identifier: 000-00C-MGR0-00100-000-00B Page III-11 of III-12 Attachment III - Worksheet “Waste Stream” Materials and cladding mass updated 8/16 ****Oxygen from fuel not included in these tables. See final worksheet 'tables' Region Top Bottom Part Name top nozzle spring retainer hold down spring upper end plug upper nuts bottom nozzle bottom spacer lower nuts kg/assembly 7.48 0.91 1.8 0.06 0.51 8.16 1.3 0.15 Material SS CF3M SS CF3M Inconel-718 SS 304 SS 304L SS CF3M Inconel- 718 Zircaloy-4 al 0 0 0.5 0 0 0 0.5 0 b 0 0 0.006 0 0 0 0.006 0 c 0.03 0.03 0.08 0.08 0.03 0.03 0.08 0 co 0.08 0.08 1 0.08 0.08 0.08 1 0 cr 19 19 19 19 19 19 19 0.1 cu 0 0 0.3 0 0 0 0.3 0 fe 62.97 62.97 14.934 67.495 66.045 62.97 14.934 0.2 mn 1.5 1.5 0.35 2 2 1.5 0.35 0 mo 2.5 2.5 3.05 0 0 2.5 3.05 0 n 0 0 0 0.1 0.1 0 0 0 nb+ta 0 0 5.5 0 0 0 5.5 0 ni 11.92 11.92 54 10.42 11.92 11.92 54 0 o 0 0 0 0 0 0 0 0.12 p 0 0 0.015 0.045 0.045 0 0.015 0 s 0 0 0.015 0.03 0.03 0 0.015 0 si 2 2 0.35 0.75 0.75 2 0.35 0 sn 0 0 0 0 0 0 0 1.7 ti 0 0 0.9 0 0 0 0.9 0 Wt % of the constituent isotopes zr 0 0 0 0 0 0 0 97.88 Region Top Bottom Element Mass Calculations Part Name top nozzle spring retainer hold down spring upper end plug upper nuts bottom nozzle bottom spacer lower nuts kg/assembly 7.48 0.91 1.8 0.06 0.51 8.16 1.3 0.15 Material SS CF3M SS CF3M Inconel- 718 SS 304 SS 304L TOTALS SS CF3M Inconel- 718 Zircaloy- 4 TOTALS al 0.0000 0.0000 0.0014 0.0000 0.0000 0.0014 0.0000 0.0020 0.0000 0.0020 b 0.0000 0.0000 0.0000 0.0000 0.0000 0.0000 0.0000 0.0000 0.0000 0.0000 c 0.0003 0.0000 0.0002 0.0000 0.0000 0.0006 0.0007 0.0003 0.0000 0.0010 co 0.0009 0.0001 0.0027 0.0000 0.0001 0.0038 0.0020 0.0039 0.0000 0.0059 cr 0.2132 0.0259 0.0513 0.0017 0.0145 0.3067 0.4651 0.0741 0.0000 0.5393 cu 0.0000 0.0000 0.0008 0.0000 0.0000 0.0008 0.0000 0.0012 0.0000 0.0012 fe 0.7065 0.0860 0.0403 0.0061 0.0505 0.8894 1.5415 0.0582 0.0001 1.5998 mn 0.0168 0.0020 0.0009 0.0002 0.0015 0.0215 0.0367 0.0014 0.0000 0.0381 mo 0.0281 0.0034 0.0082 0.0000 0.0000 0.0397 0.0612 0.0119 0.0000 0.0731 n 0.0000 0.0000 0.0000 0.0000 0.0001 0.0001 0.0000 0.0000 0.0000 0.0000 nb+ta 0.0000 0.0000 0.0149 0.0000 0.0000 0.0149 0.0000 0.0215 0.0000 0.0215 ni 0.1337 0.0163 0.1458 0.0009 0.0091 0.3059 0.2918 0.2106 0.0000 0.5024 o 0.0000 0.0000 0.0000 0.0000 0.0000 0.0000 0.0000 0.0000 0.0001 0.0001 p 0.0000 0.0000 0.0000 0.0000 0.0000 0.0001 0.0000 0.0001 0.0000 0.0001 s 0.0000 0.0000 0.0000 0.0000 0.0000 0.0001 0.0000 0.0001 0.0000 0.0001 si 0.0224 0.0027 0.0009 0.0001 0.0006 0.0268 0.0490 0.0014 0.0000 0.0503 sn 0.0000 0.0000 0.0000 0.0000 0.0000 0.0000 0.0000 0.0000 0.0008 0.0008 ti 0.0000 0.0000 0.0024 0.0000 0.0000 0.0024 0.0000 0.0035 0.0000 0.0035 zr 0.0000 0.0000 0.0000 0.0000 0.0000 0.0000 0.0000 0.0000 0.0440 0.0440 TOTALS 1.122 0.137 0.270 0.009 0.077 1.614 2.448 0.390 0.045 2.883 kg*SCALING FACTOR 1.122 0.137 0.270 0.009 0.077 1.614 2.448 0.390 0.045 2.883 Environmental and Nuclear Engineering Design Calculation Title: PWR Source Term Generation and Evaluation Document Identifier: 000-00C-MGR0-00100-000-00B Page III-12 of III-12 Attachment III - Worksheet “Waste Stream” cladding volume (CM^3) 17596.28 cladding mass (g) 115431.61 Fractions in fuel and plenum fuel: 92% plenum 8% Region FUEL PLENUM guide tubes instrument tube incore spacers cladding grid supports guide tubes instrument tube cladding spacers springs 7.381572 0.590526 4.9 106.50834 0.64 0.618428 0.049474 8.923266 1.04 0.019 Zircaloy-4 Zircaloy-4 Inconel-718 Zircaloy-4 Zircaloy-4 Zircaloy-4 Zircaloy-4 Zircaloy-4 Inconel-718 SS 302 0 0 0.5 0 0 0 0 0 0.5 0 0 0 0.006 0 0 0 0 0 0.006 0 0 0 0.08 0 0 0 0 0 0.08 0.15 0 0 1 0 0 0 0 0 1 0.08 0.1 0.1 19 0.1 0.1 0.1 0.1 0.1 19 18 0 0 0.3 0 0 0 0 0 0.3 0 0.2 0.2 14.934 0.2 0.2 0.2 0.2 0.2 14.934 68.925 0 0 0.35 0 0 0 0 0 0.35 2 0 0 3.05 0 0 0 0 0 3.05 0 0 0 0 0 0 0 0 0 0 0.1 0 0 5.5 0 0 0 0 0 5.5 0 0 0 54 0 0 0 0 0 54 9.92 0.12 0.12 0 0.12 0.12 0.12 0.12 0.12 0 0 0 0 0.015 0 0 0 0 0 0.015 0.045 0 0 0.015 0 0 0 0 0 0.015 0.03 0 0 0.35 0 0 0 0 0 0.35 0.75 1.7 1.7 0 1.7 1.7 1.7 1.7 1.7 0 0 0 0 0.9 0 0 0 0 0 0.9 0 Wt % of the constituent isotopes 97.88 97.88 0 97.88 97.88 97.88 97.88 97.88 0 0 FUEL PLENUM Element Mass Calculations guide tubes instrumen t tube incore spacers cladding grid supports guide tubes instrumen t tube cladding spacers springs 7.381572 0.590526 4.9 106.50834 0.64 0.618428 0.049474 8.923266 1.04 0.019 Zircaloy-4 Zircaloy-4 Inconel- 718 Zircaloy-4 Zircaloy-4 TOTALS Zircaloy-4 Zircaloy-4 Zircaloy-4 Inconel- 718 SS 302 TOTALS 0.0000 0.0000 0.0245 0.0000 0.0000 0.0245 0.0000 0.0000 0.0000 0.0016 0.0000 0.0016 0.0000 0.0000 0.0003 0.0000 0.0000 0.0003 0.0000 0.0000 0.0000 0.0000 0.0000 0.0000 0.0000 0.0000 0.0039 0.0000 0.0000 0.0039 0.0000 0.0000 0.0000 0.0002 0.0000 0.0003 0.0000 0.0000 0.0490 0.0000 0.0000 0.0490 0.0000 0.0000 0.0000 0.0031 0.0000 0.0031 0.0074 0.0006 0.9310 0.1065 0.0006 1.0461 0.0002 0.0000 0.0027 0.0593 0.0010 0.0632 0.0000 0.0000 0.0147 0.0000 0.0000 0.0147 0.0000 0.0000 0.0000 0.0009 0.0000 0.0009 0.0148 0.0012 0.7318 0.2130 0.0013 0.9620 0.0004 0.0000 0.0054 0.0466 0.0039 0.0563 0.0000 0.0000 0.0172 0.0000 0.0000 0.0172 0.0000 0.0000 0.0000 0.0011 0.0001 0.0012 0.0000 0.0000 0.1495 0.0000 0.0000 0.1495 0.0000 0.0000 0.0000 0.0095 0.0000 0.0095 0.0000 0.0000 0.0000 0.0000 0.0000 0.0000 0.0000 0.0000 0.0000 0.0000 0.0000 0.0000 0.0000 0.0000 0.2695 0.0000 0.0000 0.2695 0.0000 0.0000 0.0000 0.0172 0.0000 0.0172 0.0000 0.0000 2.6460 0.0000 0.0000 2.6460 0.0000 0.0000 0.0000 0.1685 0.0006 0.1690 0.0089 0.0007 0.0000 0.1278 0.0008 0.1381 0.0002 0.0000 0.0032 0.0000 0.0000 0.0035 0.0000 0.0000 0.0007 0.0000 0.0000 0.0007 0.0000 0.0000 0.0000 0.0000 0.0000 0.0000 0.0000 0.0000 0.0007 0.0000 0.0000 0.0007 0.0000 0.0000 0.0000 0.0000 0.0000 0.0000 0.0000 0.0000 0.0172 0.0000 0.0000 0.0172 0.0000 0.0000 0.0000 0.0011 0.0000 0.0011 0.1255 0.0100 0.0000 1.8106 0.0109 1.9570 0.0032 0.0003 0.0455 0.0000 0.0000 0.0489 0.0000 0.0000 0.0441 0.0000 0.0000 0.0441 0.0000 0.0000 0.0000 0.0028 0.0000 0.0028 7.2251 0.5780 0.0000 104.2504 0.6264 112.6799 0.1816 0.0145 2.6202 0.0000 0.0000 2.8164 TOTALS 7.382 0.591 4.900 106.508 0.640 120.020 0.186 0.015 2.677 0.312 0.006 3.195 kg*SCALING FACTOR 7.382 0.591 4.900 0.512 0.640 14.024 0.186 0.015 2.677 0.312 0.006 3.195 Environmental and Nuclear Engineering Design Calculation Title: PWR Source Term Generation and Evaluation Document Identifier: 000-00C-MGR0-00100-000-00B Page IV-1 of IV-1 Attachment IV Comparison of Source Terms per MTU of 4 Average PWR SNF Assemblies 4%-48GWd-25y 3.75%-41.2GWD-23y 3.75%-41.7GWd-25y 3.82%-42.6GWd-22y (475 MTU/Assembly) (432 MTU/Assembly) (434 MTU/Assembly) (432 MTU/Assembly) Curies Activation Products 9.6490E+02 9.7950E+02 8.6260E+02 1.0830E+03 Actinides and daughters 6.3270E+04 6.2700E+04 5.8460E+04 6.6380E+04 Fission products 2.7740E+05 2.5350E+05 2.4360E+05 2.6820E+05 Total 3.4180E+05 3.1720E+05 3.0290E+05 3.3580E+05 Watts Activation Products 6.4270E+00 7.4370E+00 5.8040E+00 8.6610E+00 Actinides and daughters 4.6310E+02 3.6000E+02 3.6600E+02 3.8180E+02 Fission products 7.9390E+02 7.2600E+02 6.9640E+02 7.6900E+02 Total 1.2650E+03 1.0940E+03 1.0690E+03 1.1590E+03 Grams Activation Products 4.4030E+05 4.4030E+05 4.4030E+05 4.4030E+05 Actinides and daughters 9.5050E+05 9.5740E+05 9.5690E+05 9.5590E+05 Fission products 4.9330E+04 4.2400E+04 4.2870E+04 4.3830E+04 Total 1.4400E+06 1.4400E+06 1.4400E+06 1.4400E+06 Neutrons/s Alpha,n 1.3490E+07 1.0420E+07 1.0580E+07 1.1070E+07 Sponteneous fission 4.2280E+08 2.7610E+08 2.6950E+08 3.1690E+08 Total neutrons 4.3650E+08 2.8650E+08 2.8030E+08 3.2810E+08 Photons/s Total intensity 6.7570E+15 6.1720E+15 5.9220E+15 6.5400E+15 Environmental and Nuclear Engineering Design Calculation Title: PWR Source Term Generation and Evaluation Document Identifier: 000-00C-MGR0-00100-000-00B Page V-1 of V-1 Attachment V - Listing of UNIX Script Files cut-script batch43 INPUTFILE csplit -f cut INPUTFILE.output "/module origens is finished/"\ "/halt feature/" cat cut00 cut02 > INPUTFILE.cut rm cut0* INPUTFILE.output curies BEGIN {intable=0 && insas=0 } /halt feature/ {insas=1} /nuclide radioactivity/{if (insas) print $0; intable=1} /initial/ {if (insas && intable) print $0} /charge/ {if (insas && intable) print $0} /E/ {if (insas && intable) print $0} /total/ {intable=0} gammas BEGIN {insas=0 && intable=0} /halt feature/{insas=1} /gamma source spectrum/ {if (insas) intable=1; print $0} / to / {if (insas && intable) print $0} /totals/ {intable=0} neutrons BEGIN { intable=0} /alpha-n plus/ {intable=1; print $0} /yr/{if (intable) print $0} /E/ {if (intable) print $0} / gamma sources determined / {intable=0} watts BEGIN {intable=0 && insas=0 } /halt feature/ {insas=1} /nuclide thermal power, watts/{if (insas) print $0; intable=1} /charge/ {if (insas && intable) print $0} /initial/ {if (insas && intable) print $0} /E/ {if (insas && intable) print $0} /total/ {intable=0} Environmental and Nuclear Engineering Design Calculation Title: PWR Source Term Generation and Evaluation Document Identifier: 000-00C-MGR0-00100-000-00B Page VI-1 of VI-4 Attachment VI - Calculation of Crud Source South Texas information taken from 2A-361-2A-363 PWR assembly parameters to maximize surface area Mark B South Texas Rod OD, cm 1.0922 0.95 Instrument tube OD 1.3462 1.25984 # of Rods 208 264 Rod Length 390.347 448.67 All dimensions in cm! # Guide and instrument tubes 17 25 Assembly Length 420.688 505.46 Instr. tube outside area+ inside area 3380.439626 3819.621782 PWR Assembly Surf. Area, cm^2 336057.9243 449002.7692 (Rod surface +instrument tubes inside and outside surfaces) Crud activities per unit area Co-60 1.40E-04 Ci/cm^2 NRC recommended value: 1.40E-04 Ci/cm^2 From Jones report, Table 1, p. 7 Cr 51 1.89E-04 Ci/cm^2 Mn 54 7.40E-05 Ci/cm^2 Fe 55 5.90E-03 Ci/cm^2 Co 58 4.03E-04 Ci/cm^2 Fe 59 7.60E-05 Ci/cm^2 Co 60 5.10E-05 Ci/cm^2 Ni 63 2.00E-06 Ci/cm^2 Zn 65 0.00E+00 Ci/cm^2 Zr 95 1.90E-05 Ci/cm^2 Half Life information Cr 51 27.70 days 0.0758 years Mn 54 312.10 days 0.8545 years Fe 55 2.73 years 2.73 years Co 58 70.88 days 0.1941 years Fe 59 44.51 days 0.1219 years Co 60 5.27 years 5.27 years Ni 63 100.00 years 100.00 years Zn 65 243.80 days 0.6675 years Zr 95 64.02 days 0.1753 years Environmental and Nuclear Engineering Design Calculation Title: PWR Source Term Generation and Evaluation Document Identifier: 000-00C-MGR0-00100-000-00B Page VI-2 of VI-4 Attachment VI - Calculation of Crud Source Fuel Age, years since reactor discharge Crud (Ci) for regular assembly, using NRC Co 60 values For South Texas Assembly, using NRC Co 60 values 0 47.05 62.86 5 24.38 32.57 6 21.37 28.56 7 18.74 25.04 8 16.43 21.95 9 14.41 19.25 10 12.63 16.88 11 11.07 14.80 15 6.54 8.74 20 3.39 4.53 25 1.76 2.35 30 0.91 1.22 35 0.47 0.63 40 0.24 0.33 45 0.13 0.17 50 0.07 0.09 55 0.03 0.05 60 0.02 0.02 65 0.01 0.01 70 0.00 0.01 75 0.00 0.00 80 0.00 0.00 85 0.00 0.00 90 0.00 0.00 95 0.00 0.00 100 0.00 0.00 200 0.00 0.00 300 0.00 0.00 Environmental and Nuclear Engineering Design Calculation Title: PWR Source Term Generation and Evaluation Document Identifier: 000-00C-MGR0-00100-000-00B Page VI-3 of VI-4 Attachment VI - Calculation of Crud Source Crud (Ci) for regular assembly, using Jones values Fuel Age, years since reactor discharge Cr 51 Mn 54 Fe 55 Co 58 Fe 59 Co 60 Ni 63 Zn 65 Zr 95 0 63.51 24.87 1983.41 135.43 25.54 17.14 0.67 0.00 6.39 5 0.00 0.43 557.28 0.00 0.00 8.88 0.65 0.00 0.00 6 0.00 0.19 432.32 0.00 0.00 7.79 0.64 0.00 0.00 7 0.00 0.09 335.38 0.00 0.00 6.83 0.64 0.00 0.00 8 0.00 0.04 260.18 0.00 0.00 5.99 0.64 0.00 0.00 9 0.00 0.02 201.84 0.00 0.00 5.25 0.63 0.00 0.00 10 0.00 0.01 156.58 0.00 0.00 4.60 0.63 0.00 0.00 11 0.00 0.00 121.47 0.00 0.00 4.03 0.62 0.00 0.00 15 0.00 0.00 43.99 0.00 0.00 2.38 0.61 0.00 0.00 20 0.00 0.00 12.36 0.00 0.00 1.24 0.59 0.00 0.00 25 0.00 0.00 3.47 0.00 0.00 0.64 0.57 0.00 0.00 30 0.00 0.00 0.98 0.00 0.00 0.33 0.55 0.00 0.00 35 0.00 0.00 0.27 0.00 0.00 0.17 0.53 0.00 0.00 40 0.00 0.00 0.08 0.00 0.00 0.09 0.51 0.00 0.00 45 0.00 0.00 0.02 0.00 0.00 0.05 0.49 0.00 0.00 50 0.00 0.00 0.01 0.00 0.00 0.02 0.48 0.00 0.00 55 0.00 0.00 0.00 0.00 0.00 0.01 0.46 0.00 0.00 60 0.00 0.00 0.00 0.00 0.00 0.01 0.44 0.00 0.00 65 0.00 0.00 0.00 0.00 0.00 0.00 0.43 0.00 0.00 70 0.00 0.00 0.00 0.00 0.00 0.00 0.41 0.00 0.00 75 0.00 0.00 0.00 0.00 0.00 0.00 0.40 0.00 0.00 80 0.00 0.00 0.00 0.00 0.00 0.00 0.39 0.00 0.00 85 0.00 0.00 0.00 0.00 0.00 0.00 0.37 0.00 0.00 90 0.00 0.00 0.00 0.00 0.00 0.00 0.36 0.00 0.00 95 0.00 0.00 0.00 0.00 0.00 0.00 0.35 0.00 0.00 100 0.00 0.00 0.00 0.00 0.00 0.00 0.34 0.00 0.00 200 0.00 0.00 0.00 0.00 0.00 0.00 0.17 0.00 0.00 300 0.00 0.00 0.00 0.00 0.00 0.00 0.08 0.00 0.00 Environmental and Nuclear Engineering Design Calculation Title: PWR Source Term Generation and Evaluation Document Identifier: 000-00C-MGR0-00100-000-00B Page VI-4 of VI-4 Attachment VI - Calculation of Crud Source Crud (Ci) for South Texas assembly, using Jones values Fuel Age, years since reactor discharge Cr 51 Mn 54 Fe 55 Co 58 Fe 59 Co 60 Ni 63 Zn 65 Zr 95 0 84.86 33.23 2650.01 180.95 34.12 22.90 0.90 0.00 8.53 5 0.00 0.58 744.58 0.00 0.00 11.86 0.87 0.00 0.00 6 0.00 0.26 577.62 0.00 0.00 10.40 0.86 0.00 0.00 7 0.00 0.11 448.10 0.00 0.00 9.12 0.86 0.00 0.00 8 0.00 0.05 347.62 0.00 0.00 8.00 0.85 0.00 0.00 9 0.00 0.02 269.68 0.00 0.00 7.01 0.84 0.00 0.00 10 0.00 0.01 209.21 0.00 0.00 6.15 0.84 0.00 0.00 11 0.00 0.00 162.30 0.00 0.00 5.39 0.83 0.00 0.00 15 0.00 0.00 58.78 0.00 0.00 3.19 0.81 0.00 0.00 20 0.00 0.00 16.52 0.00 0.00 1.65 0.78 0.00 0.00 25 0.00 0.00 4.64 0.00 0.00 0.86 0.76 0.00 0.00 30 0.00 0.00 1.30 0.00 0.00 0.44 0.73 0.00 0.00 35 0.00 0.00 0.37 0.00 0.00 0.23 0.70 0.00 0.00 40 0.00 0.00 0.10 0.00 0.00 0.12 0.68 0.00 0.00 45 0.00 0.00 0.03 0.00 0.00 0.06 0.66 0.00 0.00 50 0.00 0.00 0.01 0.00 0.00 0.03 0.63 0.00 0.00 55 0.00 0.00 0.00 0.00 0.00 0.02 0.61 0.00 0.00 60 0.00 0.00 0.00 0.00 0.00 0.01 0.59 0.00 0.00 65 0.00 0.00 0.00 0.00 0.00 0.00 0.57 0.00 0.00 70 0.00 0.00 0.00 0.00 0.00 0.00 0.55 0.00 0.00 75 0.00 0.00 0.00 0.00 0.00 0.00 0.53 0.00 0.00 80 0.00 0.00 0.00 0.00 0.00 0.00 0.52 0.00 0.00 85 0.00 0.00 0.00 0.00 0.00 0.00 0.50 0.00 0.00 90 0.00 0.00 0.00 0.00 0.00 0.00 0.48 0.00 0.00 95 0.00 0.00 0.00 0.00 0.00 0.00 0.46 0.00 0.00 100 0.00 0.00 0.00 0.00 0.00 0.00 0.45 0.00 0.00 200 0.00 0.00 0.00 0.00 0.00 0.00 0.22 0.00 0.00 300 0.00 0.00 0.00 0.00 0.00 0.00 0.11 0.00 0.00 Environmental and Nuclear Engineering Design Calculation Title: PWR Source Term Generation and Evaluation Document Identifier: 000-00C-MGR0-00100-000-00B Page VII-1 of VII-1 Attachment VII - 36Cl and 14C calculations PWR curies per ppm Burnup (GWd/MTU) CL36 (5.3ppm) Cl36(10ppm) Ratio of 10 ppm to 5.3 ppm values for Cl36 Cl36 5.3 ppm value/5.3 Cl36 10 ppm value/10 Ratio of 10 ppm to 5.3 ppm values for Cl36 for curies per ppm 1 1.24E-04 2.35E-04 8.952E-01 2.34E-05 2.35E-05 1.0044E+00 10 1.23E-03 2.33E-03 8.943E-01 2.32E-04 2.33E-04 1.0040E+00 20 2.51E-03 4.74E-03 8.884E-01 4.74E-04 4.74E-04 1.0009E+00 30 3.87E-03 7.32E-03 8.915E-01 7.30E-04 7.32E-04 1.0025E+00 40 5.32E-03 1.01E-02 8.985E-01 1.00E-03 1.01E-03 1.0062E+00 50 6.89E-03 1.31E-02 9.013E-01 1.30E-03 1.31E-03 1.0077E+00 60 8.55E-03 1.62E-02 8.947E-01 1.61E-03 1.62E-03 1.0042E+00 70 1.03E-02 1.95E-02 8.932E-01 1.94E-03 1.95E-03 1.0034E+00 75 1.11E-02 2.12E-02 9.099E-01 2.09E-03 2.12E-03 1.0123E+00 80 1.19E-02 2.25E-03 1.886792453 Burnup (GWd/MTU) C-14 (Ci/assembly) 1 5.67E-03 6.34E-05 10 5.69E-02 6.36E-04 20 1.17E-01 1.31E-03 30 1.87E-01 2.09E-03 40 2.54E-01 2.84E-03 50 3.33E-01 3.72E-03 60 4.20E-01 4.70E-03 70 5.13E-01 5.74E-03 75 5.62E-01 6.29E-03 80 6.11E-01 6.83E-03 1.E-05 1.E-04 1.E-03 1.E-02 0 10 20 30 40 50 60 70 80 90 Burnup (GWd/MTU) Activity per Unit Impurity (Ci/ppm) chlorine 36 (5.3 ppm) chlorine 36 (10 ppm) carbon 14 Environmental and Nuclear Engineering Design Calculation Title: PWR Source Term Generation and Evaluation Document Identifier: 000-00C-MGR0-00100-000-00B Page VIII-1 of VIII-22 Attachment VIII - List of Files on Compact Disks This attachment lists the files stored on Attachment X (four CDs). These files include: “*.cut” files - these files are the sections of the SAS2H/ORIGEN-S output files that contain the input echoes and the final ORIGEN-S output. Other intermediate calculations and information generated by SAS2H and included in the output files have been removed. In the “*.cut” file names, the E, R, and B codes explanations are listed in Table VIII-1. The files listed as "PWR.max.2.cut”, “PWR.ave.2.cut”, “PWR.max.3.cut”, and “PWR.min.3.cut” are used to generate radionuclide inventories for specific years (see Section 5.5). The “*.2.cut” files generate the information for cooling times up to 10000 years. The “*.3.cut” files generate the information for cooling times over 10000 years. The files listed as “pwr_imp_BU#.cut” and “pwr_cl2_BU#.cut” are the SAS2H/ORIGEN-S output files used to demonstrate the effect of fuel impurities on the source terms. The “BU#” indicates the burnup for that particular calculation, and follows the same convention as the rest of the files. Table VIII-1. “*.cut” files Codification “*.cut” Files E Code Enrichment (% 235U) “*.cut” Files R Code Light Elements Source Term Region “*.cut” Files B Code Burnup (MWd/MTU) E1 5.5 R1 assembly fuel B1 0.001 E2 5.0 R2 bottom end fitting B2 0.01 E3 4.5 R3 plenum B3 0.1 E4 4.2 R4 top end fitting B4 1. E5 4.0 - - B5 10. E6 3.5 - - B6 20. E7 3.0 - - B7 30. E8 2.5 - - B8 40. E9 2.0 - - B9 48.086 E10 1.5 - - B10 50. E11 1.0 - - B11 60. E12 0.711 - - B12 70. - - - - B13 75. - - - - B14 80. - - - - B15 85. UNIX script-files – these files are used to generate “*.cut” files from SAS2H/ORIGEN-S output files and to extract different parameters of interest from the “*.cut” files. They are described in Table 1. Microsoft® EXCEL spreadsheets – these files were used to calculate and display different parameters, presented in the Attachments I, II, III, IV, VI, VII, and IX, in a tabular and/or graphical form. Microsoft® WORD file – This file presents the UNIX script-files (Attachment V). Environmental and Nuclear Engineering Design Calculation Title: PWR Source Term Generation and Evaluation Document Identifier: 000-00C-MGR0-00100-000-00B Page VIII-2 of VIII-22 Attachment VIII - List of Files on Disk 1 Volume in drive D is PWR ST G&E 1 Volume Serial Number is 5F87-42C7 Directory of D:\ 05/17/2004 10:17a