Fuel Handling Facility Criticality Safety Calculations Rev 00A, ICN 00 210-00C-FH00-00400-000-00 June, 2004 1. PURPOSE The purpose of this design calculation is to perforin a criticality evaluation + of the Fuel Handling Facility (FHF) and the operations and processes performed therein. The current + intent of the FHF is to receive transportation casks whose contents will be unloaded and + transferred to waste packages (WP) or MGR Specific Casks (MSC) in the fuel transfer bays. Further, + the WPs will also be prepared in the FHF for transfer to the sub-surface facility (for + disposal). The MSCs will be transferred to the Aging Facility for storage. The criticality evaluation of+ the FHF features the following: 0 Consider the types of waste to be received in the FHF as specified below: I . Uncanistered commercial spent nuclear fuel (CSNF) 2. Canistered CSNF (with the exception of horizontal dual-purpose canister + (DPC) and/or multi-purpose canisters (MPCs) 3. Navy canistered SNF (long and short) 4. Department of Energy (DOE) canistered high-level waste (HLW) 5. DOE canistered SNF (with the exception of MCOs) Evaluate the criticality analyses previously perfortned for the existing + Nuclear Regulatory Commission (NRC)-certified transportation casks (under 10 CFR 71) to be + received in the FHF to ensure that these analyses address all FHF conditions including nonnal + operations, and Category I and 2 event sequences. Evaluate FHF criticality conditions resulting from various Category I and 2 + event sequences. Note that there are currently no Category I and 2 event sequences identified + for FHF. Consequently, potential hazards from a criticality point of view will be + considered as identified in the Internal Hazards Analysis for License Application document + (BSC 2004c, Section 6.6.4). o Assess effects of potential moderator intrusion into the fuel transfer bay + for defense in depth. The SNF/HLW waste transfer activity (i.e., assembly and canister transfer) that+ is being canied out in the FHF has been classified as safety category in the Q-Iist (BSC 2003, + p. A-6). Therefore, this design calculation is subject to the requirements of the Quality Assurance+ Requirements and Description (DOE 2004), even though the FHF itself has not yet been classified + in the Q-Iist. Performance of the work scope as described and development of the associated + technical product conform to the procedure AP-3.12Q, Design Calculations and Analyses. It should also be mentioned that the facility description document or the + system description document for the FHF is not available at this time. Consequently, this + calculation is valid for the current design and may not reflect the ongoing design evolution of the FHF. Nuclear Analysis Design Calculation Title: Fuel Handling Facility Criticality Safety Calculations Document Identifier: 210-OOC-FHOO-00400-000-OOA Page 8 of 40 2. METHOD 2.1 CRITICALITY SAFETY ANALYSIS The criticality safety calculations presented in this document evaluate the + various waste fonns in the transportation casks, WPs, canisters, and MSCs casks in the FHF to ensure + they all meet the criticality safety requirements under normal conditions as well as for Category+ I and 2 events. Moderator conditions are also varied to find the most reactive configuration. + The process and methodology for criticality safety analysis given in the Preclosure Criticality+ Analysis Process Report (BSC 2004e, Section 2.2.7) will be implemented in these calculations. + The following method will be pursued for each waste form and cask/canister configuration (BSC+ 2004e, Section 2.2.7): • The design basis for the FHF relies on the most reactive fuel assemblies • The multiplication factor (kff) will not exceed 0.95, including all + biases and uncertainties in the data and method of the analysis, under all normal, and Category I and 2 + event sequences • Conservative modeling dimensional variables will be used (e.g., assembly+ pitch, manufacturing tolerances for assemblies, etc.) in order to maximize reactivity • Conservative modeling assumptions will also be used regarding materials + in fuel including no accounting for bumable poisons in fuel, no credit for 231U and 231U in fuel,+ and use of the most reactive fuel stack density • Credit can only be taken for up to 75 % of the neutron absorbing + material in criticality controls (e.g., grid plates). Moderator density will be varied over the range of 0.0 through 1.0 in order to + include all possible criticality conditions. hi addition, the effect on criticality due to rearrangement of fuel inside the + casks/canisters from natural events (e.g., earthquakes) or incidents (e.g., dropped assembly) will + be considered (BSC 2004e, Section 2.2.8). These calculations use the qualified software MCNP (Briesmeister 1997 and CRWMS+ M&O 1998a). MCNP is a three-dimensional Monte Carlo particle transportation code + with the capability to calculate eigenvalues for critical systems. The Nuclear + Regulatory Commission (NRC) accepts MCNP in NUREG-1567 (NRC 2000, p. 8-10) for criticality + calculations. 2.2 ELECTRONIC MANAGEMENT OF INFORMATION Electronic management of information generated from these calculations is + controlled in accordance with AP-3.13Q, Design Control. The computer input and output files + generated from this calculation are stored on a Compact Disc (CD), and submitted as an + attachment to this document (Attachment 11). Nuclear Analysis Design Calculation Title: Fuel Handling Facility Cn'ticality Safety Calculations Document Identifier: 210-OOC-FHOO-00400-000-OOA Page 9 of 40 3. ASSUMPTIONS 3.1 The nominal acceptable calculated value of kff is assumed to be 0.925 as a + criticality limit in order to meet the design criten'a specified in the PDC Document [i.e.,+ kff can not exceed 0.95 including uncertainties and bias at 95% confidence level (Minwalla + 2003, Section 4.9.2.2.1)]. In other words, the nominal value provides a margin of + 0.025 (0.95 - 0.925) to account for code bias and uncertainties at 95% confidence level. Rationale: Uncertainties and bias that need to be considered in this analysis + pertain to statistical uncertainties, dimensional uncertainties, code bias, and tolerance + uncertainties. Applicable code bias for the fuel type and enrichment range of this analysis is+ typically less than 0.5 % (CRWMS M&O 1999c, Section 4). An allowance of 2% is + provided to account for the remaining uncertainties associated with statistical variation, + dimensional variables and tolerances. This allowance is similar to, and slightly greater + than (conservative), the value used for the SNF storage and transportation cask + criticality evaluations (General Atomics 1993, p. 6.4-7). Usage: This assumption is used throughout this desigii calculation. 3.2 The naval waste packages will be designed in such a way as to make + criticality not credible. Rationale: Criticality analyses for naval waste packages are the responsibility+ of the U.S Department of the Navy. Per the Naval Spent Nuclear Fuel Waste Package System Description document, the "sealed waste package shall provide criticality + Qontrol" to reduce the probability of a criticality occurring (BSC 2004d, Section 3.1.1.3). Usage: Sections 5.1 and 5.2.3. 3.3 The MGR Site specific Cask (MSC) is assumed to be similar in design, other + than the neutron poison loading/configuration, to the Multi Purpose Canister (MPC)-24 + for PWR fuel and the MPC-68 for BV~R fuel. Rationale: Since the MSC is still being developed, the criticality control + features will be similar to the existing NRC-certified storage casks. Usage: This assumption is used in Sections 5.1, 5.2.3, and 6. L 3.4 The Fort St. Vrain fuel is assumed to have a U-23 5 enrichment of 100%. Rationale: This assumption was used to introduce conservatism into the + calculation. Nuclear Analysis Design Calculation Title: Fuel Handling Facility Criticality Safety Calculations Document Identirier: 210-OOC-FHOO-00400-000-OOA Page 10 of 40 Usage: Section 5.1. 3.5 It is assumed that the fuel basket inside the canister (if present) remains+ intact following a canister drop. Rationale: The canister intemals are designed to remain sufficiently intact + that there would be no criticality concem following any credible drop or handling mishap (Minwalla 2003, Section 4.9.2.2.7). Usage: Section 6.2. 3.6 For damaged fuel calculations, it is assumed that the inside of the + canister is dry. Rationale: The canister will be shipped to the repository dry. There is no + credible mechanism by which the inside of the canister could become flooddd since the + canister is designed to withstand any credible drop without breaching (Canori and Leitner + 2003, p. 3-63). Usage: Section 6.2. Nuclear Analysis Design Calculation Title: Fuel Handling Facility Criticality Safety Calculations Document Identifier: 210-OOC-FHOO-00400-000-OOA Page 11 of 40 4. USE OF COMPUTER SOFTWARE 4.1 BASELINED SOFTWARE 4.1.1 MCNP The MCNP code (CRWMS M&O 1998a) was used to calculate the multiplication + factor, kff, for all systems presented in this report. The software specifications are as + follows: • Program Name: MCNP (CRWMS M&O 1998a) • Version/Revision Number: Version 4B2LV • Status/Operating System: Qualified/HP-UX B. 10.20 • Software Tracking Number: 30033 V4132LV • Computer Type: HP 9000 Series Workstations • CPU Number: 700887 The input and output files for the various MCNP calculations are contained on a+ CD (Attachment 11) and the files are listed in Attachment 1. The MCNP software used was: (1) appropriate for the criticality (keff) + calculations, (2) used only within the range of validation as documented through Briesmeister (1997) and + CRWMS M&O (I 998b, Section 3. 1), and (3) obtained from Software Configuration Management+ in accordance with appropriate procedures. 4.2 COMMERCIAL OFF-THE-SHELF SOFTWARE 4.2.1 MICROSOFT EXCEL 97 SR-2 • Title: Excel • Version/Revision Number: Microsofl@ Excel 97 SR-2 • This version is installed on a PC running Microsofl Windows 2000 with + CPU number 503009 The files for the various Excel calculations are contained on a CD (Attachment + 11) and the.files are listed in Attachment 1. The Excel software was used to illustrate results in Sections 5.2 and 6. Excel + is exempt from qualification per Section 2.1.6 of LP-SI. I IQ, Software Management. Nuclear Calculation Title: Fuel Handling Facility Criticality Safety Calculations Document Identifier: 210-OOC-FHOO-00400-000-OOA Page 12 of 40 5. CALCULATION All technical product inputs and sources of the inputs used in thc: development+ of this calculation are documented in this section. It should also be mentioned that the terrns + "model(s)" and "modeling" as used in this calculation document refer to the + geometnic configurations of the cnticality cases analyzed. 5.1 CALCULATIONAL INPUTS 5.1.1 Design Requirements and Criteria The design criteria for criticality safety analysis provided in Section 4.9.2.2+ of the Project Design Criteria document (Minwalla 2003) are used in these calculations. The + pertinent criteria for surface facility criticality include the following (Minwalla 2003, Section + 4.9.2.2): 0 The multiplication factor (keff) will not exceed 0.95, including all biases + and uncertainties in the data and method of the analysis, under all nortnal and off-normal event + sequences. This criterion satisfies Requirement Number PRD-013/T-022 in the Project + Requirements Document (Canori and Leitner 2003, p. 3-76). The facility design will utilize a favorable geometry and/or fixed neutron + absorbers without the use of bumup credit. No moderator shall be present in any area where radioactive waste is being + handled (fuel transfer rooms, WP closure room etc.). Attachment III features draft sketches + of the FHF as of the date of this calculation (the general arrangement drawings have not been+ finalized yet), and may not reflect the ongoing design evolution. The purpose of the sketches + is to show functional areas where moderator control is engineered in the design for + criticality safety. These functional areas will remain with moderator control, even if design + changes are made to the FHF with respect to the layout. 5.1.2 Most Reactive Fuel Selection In accordance with the requirements given in Preclosure Criticality Analysis + Process Report (BSC 2004e, Section 2.2.7), the design of the facility should be based on most + reactive fuel assemblies. The following evaluations were performed with MCNP to detennine the+ most reactive fuel assembly. Commercial Fuel The evaluations perforined in the Surface Facility Criticality Safety + Calculations document show that the W 17 x 17 OFA was found to be the most reactive PWR fuel assembly when+ modeled as a single fuel assembly and in a storage cask configuration (BSC 2004f, Section + 5.2.1). In addition, an evaluation to detennine the most reactive fuel assembly was + perfon-ned in the Final Safety and Analysis Report for the Holtec International Storage and Transfer + Operation Nuclear Analysis Calculation Title: Fuel Handling Facility Criticality Safety Calculations Document Identirier: 210-OOC-FHOO-00400-000-OOA Page 13 of 40 Reinforced Module Cask System (HI-STORM 100 Cask Systeni) and the Westinghouse + 17xl7 OFA was selected (Holtec Intemational 2002, Section 6.2-2). The GE 8 x 8 array + was selected for the BWR fuel (Holtec Intemational 2002, Section 6.2-3 as featured in the + Aging Facility Criticality Safety Calculations document (BSC 2004a, Section 5.1.3). The + results from the Surface Facility Criticality Safety Calculations and Aging Facility Criticality+ Safety Calculation documents will be utilized later in this FHF evaluation where applicable. DOE Fuel DOE SNF has been categorized into nine fuel groups (Mecharn, D.C. 2004, Section+ 4.2.4.1): I . Uranium Metal fuels (N-Reactor) 2. Uranium-Zirconium/Uranium-Molybdenum fuels (Enrico Fei-mi Liquid Metal + Reactor) 3. Uranium Oxide fuels (high enriched uranium - Shippingport PWR) 4. Uranium Oxide fuels (low enriched uranium - Three Mile Island (TMI)-2 PVvR) 5. Uranium-Aluminum fuels (foreign research reactor - Melt & Dilute) 6. Uranium/Thorium/Plutonium Carbide fuels (Ft. St. Vrain Gas Cooled Reactor) 7. Mixed Oxide fuels (Fast Flux Test Facility (FFTF) Reactor) 8. Uranium/Thorium Oxide fuels (Shippingport Light Water Breeder Reactor + (LWBR)) 9. Uranium-Zirconium-Hydride fuels (Training Research Isotopes General Atomics + (TRIGA)). The Canister Handling Criticality Safety Calculations document evaluates the + DOE fuel types listed above (both Mark IA and Mark IV type fuel are considered for N Reactor + and type "D" and type "K" canister are evaluated for TMI-2 fuel) (BSC 2004b, + Section 5.1.2). The FFTF fuel type was shown to be the most reactive in a single flooded canister under + normal conditions (BSC 2004b, Section 6.1). However, during off-normal conditions the TRIGA fuel + is the more reactive fuel type (BSC 2004b, Section 6.2). Additional calculations will be + presented later in this document for all DOE fuel types to show the effect of varying moderator + density (Section 6.2.1) and effects of neutron poison variation for applicable DOE fuel types + (Section 6.2.2). An off-norrnal condition with TRIGA fuel will also be shown later in this document+ (Section 5.2). Naval Fuel Canisters The design and safety analyses of the naval spent nuclear fuel VVT are the + responsibility of the U. S. Department of the Navy. No data on the naval waste package, including + criticality analyses, are publicly available. Section 3.1.1.3 of Naval Spent Nuclear Fuel + Waste Package System Description Document (BSC 2004d) states "The sealed waste package + shall provide criticality control ... The tenn 'control' used in this requirement is intended+ to mean that neutron absorber materials are added and the waste package loading configuration desip + reduces the probability of a criticality occurring." From this description, it is + assumed that the naval canister and intemals will be designed to make criticality not credible (Assumption 3.2)+ and no evaluations will be featured in this document. Nuclear Analysis Design Calculation Title: Fuel Handling Facility Criticality Safety Calculations Document Identifier: 210-OOC-FHOO-00400-000-OOA Page 14 of 40 5.1.3 Upper Suberitical Limit In accordance with the requirements given in Preclosure Criticality Analysis + Process Report (BSC 2004e, Section 2.2.7), kff should not exceed 0.95, including all biases + and uncertainties in the data and method of the analysis. All evaluations featuring a cask and the + MSC are performed for the worst case combination of manufacturing tolerances with respect to + criticality (Holtec Intemational 2002, p.6.3-2). Evaluations were perforrned to deterTnine the + effects of tolerances (Holtec Intemational 2002, Tables 6.3-1 & 6.3-2). It was determined that + design pararneters important to criticality safety are fuel enrichment, the inherent geometry of + the fuel basket structure and the fixed neutron absorbing panels (Boral) (Holtec Intemational + 2002, p. 6.3-3). Further, the results referred to in Section 6 of this report are within the + bounds of the kff values demonstrated in the Final Safety and Analysis Reportfor the Holtec + International Storage and Transfer Operation Reinforced Module Cask System (HI-STORM 100 Cask System) to + cover uncertainties and bias. The remainder of the results either presented or + referred to in this document is designed to meet an upper subcritical limit of 0.925 per Assumption+ 3. 1. 5.1.4 Geometry Calculation Inputs Physical inputs for the van'ous casks/canisters and waste fonns are described + in this section. Since criticality evaluations have been perfonned for some of the + casks/canisters and waste forms previously, the inputs will be referenced from these previous studies. Commercial Fuel (Transportation Casks and MSCs) A representative vertical cask is selected here for criticality calculations to+ demonstrate compliance with the cn'ticality safety requirements. The selected cask is + HI-STORM 100, as this system is currently qualified for high seismic requirements to ensure that the + YMP seismic spectrum will be enveloped (Cogema 2004, p.5). Since the MSC is still being + developed, it was assumed that it is similar in design to the MPC-24 of the HI-STORM 100 cask + system for PWR fuel and MPC-68 of the HI-STORM 100 cask system for BWR fuel (Assumption 3.3). + A criticality evaluation for these casks has already been perfon-ned in the Aging+ Facility Criticality Safety Calculations document, where the configuration and physical dimensions + for the MPC-24 as well as fuel specifications (BSC 2004a, Section 5.1.5. 1) and the MPC-68 + (BSC 2004a, Section 5.1.5.3) are given. Figure 5.1-1 illustrates the radial view of the MPC-24 cask+ inside the HI-STORM 100 overpack and Figure 5.1-2 illustrates the radial view of the + MPC-68 cask inside the HI-STORM 100 overpack. Figure 5.1-2 also displays the axial view of the MPC+ inside the overpack, which configuration is the same for both the MPC-24 and MPC-68 casks. It should be mentioned that calculations of the MSCs only exist for commercial + SNF, which was valid at the time of the criticality safety calculation perforined for the + Aging Facility. This criticality safety calculation is in the process of being revised to include + other waste types for aging. Nuclear ign Calculation Title: Fuel Handling Facility Cniticality Safety Calculations Document Identifier: 210-OOC-FHOO-00400-000-OOA Page 15 of 40 Radial Overpack Figure 5.1 -1 Radial View of the MPC-24 Cask Inside the H [-STORM 100 Overpack Nuclear Aiialysis Design Calculation 'I'Itle- Fuel f faiidliiig Facility Criticality Safety Calculations Document ldentifier: 2 1 O-OOC-FHOO-00400-000-OOA Page 16 of 40 MPC MPC ),erpack X O,e,pack Active F,pl TRIM, "f, A Figure 5.1-2 Radial and Axtal View of the MPC-68 Cask Inside the HI-STORM + 1000verpack DOE FLICI (Canisters I'able 5.1-1 presetits the pliysical dimensioiis of ilie caiiisters atid Table + 5.1-2 stiows the DOE Ctiel pariiiiieters. Figtire 5.1-3 displays the DOE caiiisters considered in + this cvaltiation, as describcd in Section5.1.2, inthera(lial %iew. ATiaxial represeTitation of the + DOE SNFcaiiisters is iilso iiiclticled in Figtirc 5.1-3. It should be iiiciitioiied iliat the + MCNP iiiptit 1-ites f'toiii t[ic Safeli CalculationsdOCUIrCut (BSC2004b) were used asastitlliiig point for the ciilctilatiotis presciited in this doctiniciit. For inore details+ regardiiig caiiister physical (Iiiiiezisioiis, sec Section 5.1.4 (BSC 2004b) aiid Sectioii 5.1.2 + (BSC 2004b) for more specifics re.prding DOE fuel parameters. Nuclear Calculation Title: Fuel Handling Facility Criticality Safety Calculations Document Identiirier: 210-OOC-FHOO-00400-000-OOA Page 17 of 40 Table 5.1 -1 Physical Dimensions of DOE Canisters DOE Fuel Type Canister o.d. Canister length Canister Capacity Reference (cm) (cm) 3360 fuel pins (2 sets of CRWMS M&O Enrico Fermi 45.72 12 tubes each 2000a p. 13 containing 140 pins) , 45.72(0.95 456.90 (414.50 cm 1302 fuel pins (6 CRWMS M&O FFTF cm wall internal length) assemblies with each 1999a, Figures 5-3 & thickness) 217 fuel pins) 5-4 Fort St. Vrain 45.72(0.95 cm wall 457.0 (411.71 cm 5 fuel elements stacked BSC 2001 a p. 15 thickness) internal length) vertically , 45.72(0.95 299.90 (254.0 cm 3 ingots stacked Melt & Dilute cm wall internal length) vertically BSC 2001 b, p. 11 thickness) 270 fuel elements (54 CRWMS M&O 2001, N Reactor 64.29 419.84 fuel elements stacked 5 p. 14 high) DOE 2000, pp. 23-25 (canister capacity) 45.72(0.95 7428 fuel rods (12 CRWMS M&O Shippingport LWBR cm wall 457.0 (411.71 cm assemblies with each 2000b, p. 18 thickness) internal length) 619 fuel rods) DOE 1999b, p. 16 (canister capacity) Shippingport PWR 45.72(0.95 cm wall 268.09 (internal 1 fuel cluster CRWMS M&O thickness) length) 2000c, p. 15 b 35.56(0.64 380 37 (346 55 cm 1 fuel assembly (1 5xl 5 DOE 2003, pp. 21 TMI-2 (D canister cm wall . . internal length) array having 204 fuel (canister capacity), thickness) rods) 25 & 26 45.72(0.95 254.70 (internal 111 fuel elements (37 CRWMS M&O TRIGA cm wall length) fuel elements stacked 3 1999d p. 13 thickness) I high) , a Mark 1A contains 48 fuel elements stacked 5 high, comprising a total of 240 + fuel elements (DOE 2000,Fig. 4-2~ b The K canister has a large internal diameter over which fuel matrix material + is not constrained (see Fig. 5.1-3) Nuclear Amalysis Design Calculation Title: Fuel Handling Facility Criticality Safety Calculations Document Identifier: 210-OOC-FHOO-00400-000-OOA Page 18 of 40 Table 5.1-2 DOE Fuel Parameters Max. fissile Fuel Clad Clad Pin Pitch Fuel DOE Fuel Type enrichment a o.d. b i.d. o.d. (cm) c length Reference (%j (CM) (CM) (cm) (cm) DOE 1999a, p.8 Enrico Fermi 25.69 0.376 0.376 0.401 0.44 77.47 CRWMS M&O 2000a, p. 12 (clad) INEEL 2002, FFTF 25.95 0.495 0.508 0.584 0.726 237.24 p.15,17 (pin pitch) & Fig. 3 (fuel o. .) 100.0 Taylor 2001, p. 21 Fort St. Vrain (Assumption 1.245 - 1.880 & Fig. 2-3 (pin 3.4) itch) Melt & Dilute 20.0 41.91 - - - 76.2 BSC 2001c, p.3 N Reactor - outer e 1 25 6.096 6.096 6.223 DOff-2000, fuel tube d . 4.496 f 4.496 4.607 6.096 53.0 Tables 3-1 & 3-2 (clad) N Reactor - inner e 1 25 3.175 3.175 3.378 DOE 2000, fuel tube d . 1.118' 1.118 1.245 6.096 53.0 Tables 3-1 & 3-2 (clad) DOE 1999b, p. 16 (enr.), Fig. 3-3 Shippingport LWBR 4.90 0.640 0.734 0.778 0.937 (pin pitch), Table 3-5 (fuel o.d.) & Table 3-8 (clad) Shippingport PWR 93.2 - - DOE 1999c, Table 3-1 DOE 2003, p. 19 TMI-2 2.96 0.936 0.958 1.092 1.443 360.12 (enr,), p. 21, p. 22 (fuel length) & p.23 (fuel o.d.) TRIGA 70 0 3.480 3 490 592 3 3 480 38 10 CRWMS M&O . 0.635 g , . . . . 1999b, p. 19 b This is the total fissile content divided by the total heavy metal mass x + 100. For fuel in the form of cylindrical rods, this is the fuel outside diameter c For fuel in the form of cylindrical rods, this is the nominal pin pitch in + the canister d See Figure 5.1-3 for locations of outer and inner fuel tubes ' The enrichment for Mark IV (case B) is 0.95 % f Inside diameters of fuel tubes 9 Inside diameters of fuel tube NUClcar Analysis Dcsigii C'alculatioii Title: Fuel Handling Facility Criticality Safety Calculations Document Identifier: 210-OOC-FHOO-00400-000-OOA Page 19 of 40 Radial V. of N Phil or Mark I A Crosier i Voee or N Pvactur Mark V Canister + R,d.dJ Vd. of Pool, & Dial, soon, It I r P a dial Vowo of Fng St Vrain Canistpr i View, of FFTF Cents, Roodial Vew of Enrico Fermi Camotor Rad,il View of Shippingport LWBR Collate, Fqadial View of 7Mli2 7ype K C a lilsier Pedial Vipw of Shippingpnq PWP Canisier i \/,. of TRIGA C col Radial View or TMI I Typ, D C,fi Anal View or DIE Cedeller Figure 5.1-3 Radial and Axial View of the DOE Fuel Canisters Navy Fuel Per Assumption 3.2, no evaluations will be perfortned of navy fuel since it is + the responsibility of the U. S. Department of the Navy. 5.1.5 Material Compositions Material compositions for the various casks/canisters and waste foryns are + described in this section. Since criticality evaluations have been performed for some of the + casks/canisters and waste forrns previously, the inputs will be referenced from these previous + studies. Commercial Fuel (PVVR & BWR) Since no additional calculations were perfonned in this document for commercial+ fuel, the material compositions can be found in the Aging Facility Criticality Safety + Calculations document for PWR (BSC 2004a, Section 5.1.5.2) and BWR (BSC 2004a, Section + 5.1.5.4) fuel. DOE Fuel Table 5.1-3 display's the relevant material properties for DOE non-fuel + materials used in the MCNP models. Table 5.1-4 presents the isotopic content of the fuel materials + for each DOE type fuel considered in this calculation. Table 5.1-3 Material Properties for DOE Non-Fuel Materials Material Density 3 Weight Percent (wt %) Reference/ (g/CM ) Remark H20 1 0 a H - 0.6666667 (throughout model) . 0 - 0.3333333 0:49.94 Ca:22.63 C:10.53 Mg:9.42 Magnuson Concrete 2.147 SiA.21 K0.9445 A1:0.7859 Fe:0.559,5 NRC 1997, Volume 3, Ti:0.148 Na:0.1411 H:0.3319 S:0.2483 p. M8.2.4 C1:0.0523 Mn 0.0512 Type 304L 7 94 Fe:68.045 Cr:19.0 Ni:10.0 Mn:2.0 ASTM A 276-91 a 199 1, p. 2 Stainless Steel . Si:0.75 N:0.1 13:0.045 S:0.03 C:0.03 ASTM GI-90 1999, Table + Xl Type 316L Fe:65.295 COTO Ni:12.0 Mn-2 * 0 ASTM A 276-91 a 199 1 p 2 Stainless Steel 7.98 Mo:2.5 Si:1.0 N:0.1 P:0.045 S:0.03 , . ASTM G 1 -90 1999 Table Xi C:0.03 , Type 516 Fe:98.33 Mn:1 025 Si:0 275 P:0 035 ASME 2001, Sec IIA, SA- Carbon Steel 7.85 . . . S:0.035 C:0.3 516/SA-516M & Sec IIA, SA- 20/SA-20M, item 14 - I he moderator density was varied between 0.0 - 1 .0 glcm' to study moderator+ density variations in Section 6 b Values given in atom fraction and not wt % Nuclear Table 5.1-4 Material Properties for Each DOE Fuel Type DOE Fuel Type Dens Weight Percent (wt%) Neutron Absorber (kg) (g/c ) Enrico Fermi 17.424 U7235:22.96 U-238:66-41 1.2-3.03 Mo:10.63 0:11.63 U-235:0.13 FFTF 10 02 U-238:62.37 Pu-239:22.54 0.4-19.26 b . Pu-240:3.01 Pu-241:0.26 Pu-242:0.06 Th-232:25.69 C:64.81 Fort St. Vrain 1.991 U-235:3.54 Si:5.96 ------ U-235:3.64 U-238:14.56 Melt and Dilute 3.00 A1:77.97 Gd:0.50 H:0.37 0.0009-4.73 c 0:2.96 N Reactor 18.39 U-235:1.25 U-238:98.75 ------ 0:12.12 U-233:4.57 Shippingport LWBR 9.71 U-234:0.06 U-238:0.02 ------ Th-232:83.23 U-235:45.04 U-238:3.29 Shippingport PWR - zone 1 6.36 Ca:3.72 Zr:29.54 0:18.41 ------ Shippingport PWR -zone 2 6.36 U-235:32.98 U-238:2.41 ------ Ca:4.15 Zr:39.98 0:20.48 U-235:21.74 U-238:1.59 Shippingport PWR - zone 3 6.36 Ca:4.57 Zr:49.67 0:22.43 ------ U-235:2.61 U-238:85.53 TMI-2 10.42 0:11.86 ------ U-235:5.94 U-238:2.56 TRIGA 6.58 Zr:89.91 H:1.59 SOURCE: BSC 2004b, Table 5-3. Also, see BSC 2004b, Section 5.1.2 for fuel + description. 8 Neutron absorber (Gd) contents in canister were varied. 1 vol% corresponds to+ 3 kg (CRWMS M&O 2000a, p. 12) b Neutron absorber (Gd) contents in canister were varied. 5 wt% corresponds to + 19.26 kg, which is the maximum amount of gadolinium (CRWMS M&O 1999a, p.21) c Neutron absorber (Gd) contents in ingots were varied. 0.5 wt% corresponds to + 4.73 kg (BSC 2001c, p.3) Naval Fuel Canisters Per Assumption 3.2, no evaluations will be perfonned of naval fuel canisters + since it is the responsibility of the U. S. Department of the Navy. 5.2 CRITICALITY CALCULATIONS The process and methodology for criticality safety analysis given in the + Preclosure Criticality Analysis Process Report (BSC 2004e, Section 2.2.7) were implemented in these + calculations. The sections below feature calculations perfonned in this document. As + mentioned earlier, results from previously perfonned applicable criticality evaluations are + utilized in Section 6 and are not included in sections below. 5.2.1 DOE Fuel Moderator Density Variations Moderator density, which could vary from dry to fully moderated conditions + under accident conditions, was varied in the MCNP model over the range of 0.0 to 1.0 g/cm 3 . + The calculations were perforTned for all DOE ftiels considered in this document. They were each + modeled as single ca"nisters with their outsides surrounded by concrete. 5.2.2 DOE Fuel Neutron Poison Variations Variations in neutron poison loading were studied to cover any errors of its + integration in the basket structure or fuel mixture. The DOE fuels that have neutron poisons as + part of their interior basket structure are Enrico Fenni and FFTF. Melt & Dilute fuel has+ integrated neutron poison in its fuel. Each of these DOE fuels were modeled as single canisters + with their outsides surrounded by concrete. The mass of the neutron absorbers was varied in + accordance with the range shown in Table 5.1-4. 5.2.3 Category 1 and 2 Event Sequences At the present time, no Category I and 2 events have been identified for the + Fuel Handling Facility. The Internal Hazards Analysis for License Application document (BSC + 2004c) identifies potential events that could lead to a criticality accident. While + these potential events have not yet been categorized into Category I and 2 or beyond Category 2 + events, all of the identified potential events will be considered in this document for + conservatism. Table 5.2-1 describes the potential criticality events for the FHF entrance + vestibule and the applicable criticality safety evaluation performed for each event. Table 5.2-2 + describes the potential criticality events for the FHF preparation room and the applicable + criticality safety evaluation performed for each event. Table 5.2-3 refers to potential + criticality events associated with CSNF assembly transfer and the FHF main transfer room, the fuel transfer + bay, and the fuel transfer room. Table 5.2-4 refers to potential criticality events for canister + transfer in the FHF main transfer room. Table 5.2-5 describes the potential criticality events for + FHF WP closure. Table 5.2-6 describes the potential criticality events for WP loadout in the + FHF main transfer room, preparation room, and entrance vestibule. Table 5.2-7 refers to potential+ criticality events for loaded MSC removal operations in the FHF main transfer room, preparation + room, and entrance vestibule. The supporting calculations for the potential criticality + events are provided in the subsections. Table 5.2-1 Potential Criticality Events for the FHF Entrance Vestibule Potential Event a Criticality Safety Evaluation Criticality associated with a raifcar (holding a loaded cask) derailment or collision Regulatory compliance with 10 CFR 50, 71+ and 72 provides followed by a load tipover or fall and assurance of criticality safety for this+ event. rearrangement of the cask internals. Criticality associated with an overturning or collision involving an LWT or an OWT Regulatory compliance with 10 CFR 50, 71 + and 72 provides holding a loaded cask and rearrangement assurance of criticality safety for + this event. of cask internals. Criticality associated with a drop or Regulatory compliance with 10 CFR 50, 71 + and 72 provides slapdown of a cask and a rearrangement assurance of criticaliiy safety for this+ event. of the container internals. Per Assumption 3.3, the MSC is similar in design to a NRC-certified Criticality associated with a drop or cask. There is no effect on the + criticality control features of the slapdown of an MSC and a rearrangement system as a result of this event shown + by the cask handling of the container internals. accident evaluation in BSC 2004a, Table 5.2-1. + Furthermore, there is no moderator intrusion to make the configuration more reactive. Criticality associated with a gantry crane Per Assumption 3.3, the MSC is + similar in design to a NRC-certified holding an MSC collision followed by a cask. There is no effect on the + criticality control features of the load drop or tipover and a rearrangement system as a result of this event shown+ by the cask tip-over of the MSC internals evaluation in BSC 2004a, Table 5.2-1. Furthermore, there + is no . moderator intrusion to make the configuration more reactive. " BSC 2004c, Table 23b Table 5.2-2 Potential Criticality Events for the FHF Preparation Room Potential Event a Criticality Safety Evaluation Regulatory compliance with 10 CFR 50, ~1 and 72 provides Criticality associated with a loaded cask or assurance of criticality safety + for this event. Also, there is no effect loaded MSC collision or trolley derailment on the criticality control features + of the system as a result of this followed by a load tipover or fall and a event shown by the tip-over evaluation+ in BSC 2004a, Table 5.2-1. rearrangement of the cask internals. Furthermore, there is no moderator + intrusion to make the configurati n more reactive. ' BSC 2004c, Table 24b Nuclear im Calculation Title: Fuel Handling Facility Criticality Safety Calculations Document Identifier: 210-OOC-FHOO-00400-000-OOA Page 24 of 40 Table 5.2-3 Potential Criticality Events for the FHF Transfer RooMS b Potential Event a Criticality Safety Evaluation Since the transportation cask, for the SNF assembly drop into a transportation cask scenado, is NRC-certified to be 10 CFR 71 Criticality associated with a drop of an compliant, the criticality evaluation + performed for the cask SN F assembly from the spent fuel transfer certification is adequate to cover + this event. Per Assumption 3.3, machine into a cask, MSC, or WP and a the MSC is similar in design to a NRC + certified cask, and will also rearrangement of the cask, MSC, or WP be in compliance with 10 CFR 71. The WP + is designed to internals. withstand credible hazards without significant rearrangement of the iuel (Minwalia 2003, Section 4.9.2.2.7) for the potential event involving the WP. Criticality associated with a drop of an SNF assembly from the spent fuel transfer The drop could cause reconfiguration + of the CSNF assembly. machine and a rearrangement of the fuel Section 5.2.3.1 of BSC 2004f evaluates + the keff of a reconfigured, rods that comprise the assembly due to fully flooded, CSNF and it remains + safely below 0.9. impact. Regulatory compliance with 10 CFR 50, 71 and 72 provides assurance of criticality safety for the event with the transportation Criticality associated with the drop of cask. Per Assumption 3.3, the MSC is + designed to the same heavy equipment onto a loaded, open standards as the cask and consequently + regulatory compliance cask, MSC, or WP and a rearrangement of provides assurance of criticality + safety. The WP is designed to the container internals. withstand credible hazards without significant + rearrangement of the fuel (Minwalia 2003, Section 4.9.2.2.7) for the potential event involving the WP. Fuel assembly misloading is not an issue for "out-of-package" Criticality associated with a misload of a criticality, as the criticality + evaluations (BSC 2004a & BSC 2004f) WP or and MSC. are based on 5% maximum enrichment and no burnup credit is taken. Criticality evaluations for fully flooded conditions of a MSC (BSC Docking dng leaking water into an 2004a, Section 6.3) and 5% maximum enrichment+ with no burnup unsealed loaded cask or MSC leads to a credit taken shows that there is no + criticality concern. Further, criticality. moderator intrusion studies shows (BSC 2004f, Section 6.1) that fully flooded conditions are bounding. a BSC 2004c, Table 26b b The events are associated with CSNF assembly transfer and applies to all FHF + transfer rooms (i.e., main transfer room, fuel transfer bay, and fuel transfer room) Table 5.2-4 Potential Criticality Events for the FHF Main Transfer ROOM b Potential Event a Criticality Safety Evaluation Criticality associated with a drop of a loaded cask or MSC from the main Per Assumption 3.3, the MSC is designed to the+ same standards as transfer room overhead crane and a the cask and consequently regulatory + compliance with 10 CFR 50, rearrangement of the cask or MSC 71 and 72 provides a.ssurance of criticality + safety. internals. Per Assumption 3.2 it is the responsibility of the U.S Department of the Navy to ensure criticality safety for the naval SNF canister. Criticality associated with a drop of a DPC, Regulatory compliance with 10 CFR + 50, 71 and 72 provides a DOE SNF canister, a naval SNF assurance of criticality safety for the event + with the DPC. Section canister, or a DOE HLW canister and a 5.2.3.1 of this document demonstrates + that a rearrangement of fuel rearrangement of canister internals. for the DOE canisters does not increase + the reactivity. In addition, Section 5.2.3.2 shows that losing the skirt, or impact limiter, of the canister due to a drop also not increase the reactivity. Regulatory compliance with 10 CFR 50, 71 and 72 provides assurance of criticality safety for the event with the transportation Criticality associated with the drop of cask. Per Assumption 3.3, the MSC is + designed to the same heavy equipment onto a loaded, open standards as the cask and consequently + regulatory compliance cask, MSC, or WP and a rearrangement of provides assurance of criticality + safety. The WP is designed to the container internals. withstand credible hazards without significant + rearrangement of the fuel (Minwalia 2003, Section 4.9.2.2.7) for the potential event involving the WP. Fuel assembly misloading is not an issue for"out-of-package" Criticality associated with a misload of a criticality (BSC 2004e, p. 1), as + the criticality evaluations (BSC WP or an MSC. 2004a & BSC 2004f) are based on 5% maximum enrichment and no bumup credit is taken. ' BSC 2004c, Table 27b b The potential events are associated with canister transfer Table 5.2-5 Potential Criticality Events for FHF Waste Package Closure Potential Event a Criticality Safety Evaluation Criticality associated with a trolley holding The WP is designed to withstand + credible hazards without a sealed or unsealed WP derailment significant rearrangement of the fuel + (Minwalla 2003, Section followed by a load tipover or fall and 4.9.2.2.7). rearrangement of the container intemals. Criticality associated with a drop of a The WP is designed to withstand + credible hazards without loaded, unsealed WP from the main significant rearrangement of the fuel + (Minwalla 2003, Section transfer room overhead crane and a 4.9.2.2.7). rearrangement of the container intemals. Criticality associated with the drop of The WP is designed to withstand + credible hazards without heavy equipment onto a loaded, unsealed significant rearrangement of the fuel + (Minwalla 2003, Section WP and a rearrangement of the container 4.9.2.2.7). internals. a BSC 2004c, Table 28b Table 5.2-6 Potential Criticality Events for Waste Package Loadout Potential Event Criticality Safety Evaluation Criticality associated with a trolley holding The WP is designed to withstand + credible hazards without a sealed WP derailment followed by a load significant rearrangement of the fuel+ (Minwalla 2003 Section tipover or fall and rearrangement of the , 4.9.2.2.7). WP internals. Criticality associated with a drop or The WP is designed to withstand credible + hazards without collision of a sealed WP and a significant rearrangement of the fuel (Minwalia + 2003, Section rearrangement of the container internals. 4.9.2.2.7). Criticality associated with a slapdown of a The WP is designed to withstand + credible hazards without sealed WP and a rearrangement of the significant rearrangement of the fuel + (Minwalla 2003, Section container internals. 4.9.2.2.7). oou zuu4c, i aDie zvD b The events applies to the FHF main transfer room, preparation room, and + entrance vestibule Table 5.2-7 Potential Criticality Events for Loaded MSC Removal OperationS b Potential Event a Criticality Safety Evaluation Criticality associated with an MSC trolley Per Assumption 3.3, the MSC is + similar in design to a NRC-certified collision or trolley derailment followed by a cask. There is no effect on the + criticality control features of the load tipover or fall and a rearrangement of system as a result of this event + shown by the cask tip-over the MSC internals. evaluation in BSC 2004a, Table 5.2-1. Furthermore, there is + no moderator intrusion to make the configuration more reactive. Criticality associated with a drop or Per Assumption 3.2, the MSC is similar in+ design to a NRC-certified slapdown of a loaded MSC from an cask. There is no effect on the criticality + control features of the overhead crane and a rearrangement of system as a result of this event shown by+ the cask handling cask internals accident evaluation in BSC 2004a, Table 5.2-1. Furthermore, + there . is no moderator intrusion to make the configuration more reactive. Regulatory compliance with 10 CFR 50, 71 and 72 provides assurance of cdticality safety for the event with the transportation Criticality associated with the drop of cask. Per Assumption 3.3, the MSC is + designed to the same heavy equipment onto an unsealed MSC standards as the cask and consequently + regulatory compliance and a rearrangement of the container provides assurance of criticality safety. + The WP is designed to internals. withstand credible hazards without significant rearrangement of the fuel (Minwalla 2003, Section 4.9.2.2.7) for the potential event involving the WP. - t3,-jU ZUU4C, i anie 3ob b The events applies to the FHF main transfer room, preparation room, and + entrance vestibule 5.2.3.1 Rearrangement of DOE Canister Iniernals Potential Category I and 2 event sequences includes drops of DOE canisters + causing a rearrangement of canister intemals (see Table 5.2-4). For this purpose fuel pin+ pitch variations (i.e., increased and deceased pin pitch) were modeled for the applicable DOE + canisters in MCNP to simulate this potential event. The DOE canisters were modeled as single + canisters with flooded inside conditions and concrete reflection outside, representing the + most reactive configuration. Table 6-1 of BSC 2004b demonstrates that for nearly all DOE fuel+ types, k,ff increases when the outside is surrounded by concrete instead of water. The + reason for this is that concrete is a better reflector than water in the enviromnent that is being + modeled. Table 5.2-8 lists the keff as a function of pin pitch for each applicable DOE fuel types in+ their respective canisters. It can be seen that varying the pin pitch does not cause a + criticality concem. Further, the pin pitches that produce the maximum keff are used for criticality + evaluations under nonnal conditions (BSC 2004b, Section 6.1). The results presented in Table 5.2-8 are + also illustrated in Figure 5.2-1, which is generated in D0E_fue1.x1s. Note that the N Reactor fuel + denoted "A" in the figure refers to Mark I A fuel and "B" refers to Mark IV fuel. Table 5.2-8 DOE Fuel Pin Pitch Variations Pitch (cm) Koff I St. Dev I MCNP files a Pitch (cm) Keff I St. Dev MCNP files' Enrico Fermi FFTF 0.01 0.81681 0.00077 efwdsOl.out 0.6 0.83516 0.00094 ffwds06.out 0.02 0.83621 0.00076 efwds02.out 0.8 0.82977 0.00097 ffwds08.out 0.03 0.85551 0.00076 efwds03.out 1.0 0.8512 0.00095 ffwdsl O.out 0.04 0.87361 0.00074 efwds04.out 1.2 0.86941 0.00091 ffwdsl 2.out 0.05 0.88478 0.00077 efwds05.out 1.4 0.87562 0.00102 ffwdsl4.out 0.06 0.88616 0.00078 efwds06.out 1.5 0.87592 0.00076 ffwdsl5.out 0.07 0.88564 0.00077 efwds07.out 1.6 0.87644 0.00103 ffwds16.out 0.08 0.88028 0.00079 efwds08.out 1.7 0.87629 0.0008 ffwdsl7.out 0.09 0.87757 0.00077 efwds09.out 1.8 0.87231 0.00094 ffwdsl8.out 0.1 0.87865 0.00085 efwdsl O.out 2.0 0.86517 0.00106 ffwds20.out 0.11 0.88474 0.00081 efwdsl 1.out TRIGA 0.12 0.88278 0.0008 efwdsl2.out 6.03 0.83818 0.00112 trwds60.out 0.13 0.87078 0.0008 efwdsl 3.out 6.5 0.79473 0.00110 trwds65.out 0.14 0.85951 0.00083 efwds14.out 7.0 0.74618 0.00980 trwds70.out 0.15 0.85611 0.00075 efwdsl 5.out 7.5 0.69476 0.00100 trwds75.out N Reactor Mark IA Fuel N R eactor Mar k IV Fuel 6.2 0.72258 0.00063 nrwds62a.out 6.2 0.71364 0.00058 nrwds62b.out 6.6 0.78310 0.00064 nrwds66a.out 6.6 0.77336 0,00060 nrwds66b.out 7.0 0.82639 0.00063 nrwds70a.out 7.0 0.81915 0.00058 nrwds70b.out 7.4 0.84729 0.00064 nrwds74a.out 7.4 0.84396 0.00059 nrwds74b.out 7.8 0.84929 0.00067 nrwds78a.out 7.8 0.85334 0.00056 nrwds78b.out 8.2 0.83357 0.00067 nrwds82a.out 8.2 0.84373 0.00058 nrwds82b.out 8.6 0.80056 0.00065 nrwds86a.out 8.6 0.82394 0.00057 nrwds86b.out 9.0 0.76490 1 0.00064 1 nrwds90a.out 9.0 0.80148 1 0.00055 1 nrwds90b.out TM I-2 Type D Canister TM I-2 Type K Canister 1.1 0.61646 0.0009 tmwds 11 d.out 1.1 0.41605 0.0007 tmwdsl 1 k.out 1.3 0.77349 0.00097 tmwdsl3d.out 1.3 0.57354 0.00087 tmwdsl3k.out 1.5 0.85634 0.00097 tmwdsl5d.out 1.5 0.70058 0.00087 tmwdsl5k.out 1.7 0.85296 0.00091 tmwdsl 7d.out 1.7 0.77586 0.00099 tmwdsl7k.out 1.9 0.83542 0.00097 tmwdsl 9d.out 1.9 0.80811 0.00089 tmwdsl9k.out 2.1 0.78778 0.00096 tmwds2l d.out 2.1 0.78911 0.00097 tmwds2l Kout 2.3 0.74795 0.00086 tmwds23d.out 2.3 0.73864 0.00088 tmwds23k.out 2.5 0.69909 , 0.00084 , tmwds25d.out 2.5 0.68338 , 0.0009 tmwds25k.out a The input files to each run have the same name as the corresponding output + file but without the out b extension (e.g., the input file matching output file tmwdsl 1 d.out is tmwdsl+ 1 d). Smallest possible physical spacing. 0.9 0.85 0.8 0.75 0.7 0.65 0.6 5.5 0.9 0.85 0.8 0.75 0.7 0.65 0.6 0.55 0.5 0.45 0.4 0 0.95 0.9 0.85 0.8 0.75 0.7 0.65 0.6 0 Figure 5.2-1 DOE Fuel Pin Pitches versus keff 6 6.5 7 7.5 8 8.5 9 9.5 Pitch (cm ) 0.5 1 1.5 2 2.5 3 Pitch (cm) 0.05 0.1 0.15 0.2 Pitch (cm) 5.2.3.2 DOE Fuel Drop or Slapdown In the event of a drop or slapdown, as discussed earlier in Section 5.2.3. 1, + the lower skirts of the canister might get damaged causing the interior fuel basket to sit directly on + the bottom of the canister. A calculation was perfonned with a single flooded TRIGA fuel canister+ featuring a neglected lower skirt (the area below the skirt was exchanged from water to + concrete). Table 5.2- 9 shows the results and it can be seen by comparing this calculation to the + calculation of an intact TRIGA fuel canister that a lost or damaged lower skirt does not impact kff. Table 5.2-9 TRIGA Fuel Canister With and Without Skirt MCNP Model Description Kff ~ Standard Deviation -MCNP input & output files TRIGA Intact Lower Skirt 0.83804 0.00115 trwds60, trwds60.out Damaged Lower Skirt 0.83804 0.00115 trwds60s, trwds6ft.out 6. RESULTS AND CONCLUSIONS This section presents the results of the criticality calculations and makes + recommendations for additional criticality safety design features as appropriate. The outputs + presented in this document are all reasonable compared to the inputs and the results are suitable+ for the intended use. The uncertainties are taken into account by consistently using a + conservative approach, which is the result of the methods and assumptions described in Sections 2 and + 3, respectively. 6.1 CASK AND MSC CRITIALITY EVALUATION Criticality evaluations performed in the Aging Facility Criticality Safety + Calculations document (BSC, 2004a) feature the Holtec HI-STORM 100 cask system as a representative + cask. Per Assumption 3.3, the MSC is similar in design to this NRC-licensed cask. The + results presented in Section 6 of BSC 2004a shows that there are no criticality concems + associated with the Hi- STORM cask or the MSC. The results lead to the following main conclusions (BSC + 2004a, Section 6.5): Both the PWR and BWR results consistently demonstrate that the conditions + outside the Hl- STORM overpack (e.g., spacing, moderation, and reflection) have no discemable + impact on the reactivity of the cask. This indicates that the casks are neutronically + isolated and consequently the cask orientation (e.g., vertical versus horizontal) will not + matter. 9 Reactivity of the loaded casks decreases with reduction in moderator density. Maximum reactivity is reached when the casks are fully flooded with water at + full density (1.0 g/CM3). The scenarios considered in BSC 2004a covers the conditions featured in the + FHF. For this purpose, no additional criticality calculations were performed for a cask or + MSC in the FHF. 6.2 DOE FUEL CANISTER CRITICALITY EVALUATIONS Criticality evaluations were performed fo'r DOE fuel in the Canister Handling + Facility Criticality Safety Calculations document (BSC 2004b) were the results lead to the following+ conclusions applicable to FHF operations (BSC 2004b, Section 6.3): Criticality is not a concem for any single DOE fuel canister under both nonnal + and flooded conditions. Cn'ticality is not a concem for darnaged fuel resulting from a catastrophic + drop of a single fuel canister for any DOE type. This is based on the assumptions that interior + basket, if present, remains intact following a canister drop (Assumption 3.5) and that the+ damaged fuel inside the canister is dry (Assumption 3.6). To verify that fully flooded conditions are the most reactive, additional + calculations were perforrned were the fuel moderator density was varied. The results from the + calculations are presented below in Section 6.2. L Further, the Melt & Dilute has integrated+ neutron poison (Gd) in its ftiel while the Enrico Ferrni and FFTF fiiels have neutron absorbers in + their interior baskets. Variations in neutron poison loading were also studied to cover any errors of + its integration in the basket structure or fuel mixture. The results from the calculations are + presented below in Section 6.2.2. 6.2.1 DOE Fuel Moderator Density Variations Moderator density, which could vary from dry to fully moderated conditions + under accident conditions, has been varied over the range of 0.0 to 1.0 g/cm 3 for all DOE + fuel types considered in this document. Table 6.2-1 displays kff as a function of moderator density + for the various DOE fuel. It can be seen that the reactivity of the DOE fuel decreases with + reduction in moderator density. Consequently, it can be concluded that fully flooded + conditions are the most reactive. Figure 6.2- 1, generated in D0E_fue1.x1s, illustrates the results + presented in Table 6.2- 1. Table 6.2-1 K,.ff as a Function of Moderator Density Variations Density (g/cc) K -off I St. Dev MCNP fileS a I Keff St. Dev MCNP files a I Enrico Fermi FFTF 1.0 0.91063 0.00078 efwdsOft.out 0.92084 0.00074 ffwdsl5y.out 0.98 0.90684 0.00078 efwds698.out 0.91541 0.00078 ftdsWout 0.95 0.90194 0.00079 efwds695.out 0.90891 0.00082 ffwds95.out 0.90 0.88938 0.00079 efwds069.out 0.89624 0.00079 ffwds159.out 0.80 0.86651 0.00075 efwds068.out 0.87326 0.00075 ffwds158.out 0.50 0.78822 0.00076 efwds650.out 0.79055 0.00073 ffwds50.out 0.0 0.49484 0.00041 1 efddsOU.out 0.56406 0.00054 , ffwdsNy.out Fort St. Vrain Melt & Di lute 1.0 0.84467 0.00093 fswdsOO.out 0.83129 0.00268 mdwdsOOz.out 0.98 0.84357 0.00092 fswds09B.out 0.82582 0.00253 mdwdsO98.out 0.95 0.84218 0.00095 fswds095.out 0.82558 0.00243 mdwds095.out 0.90 0.83603 0.00100 fswds009.out 0.81020 0.00261 mdwds009.out 0.80 0.82287 0.00101 fswds008.out 0.80444 0.00268 mdwds008.out 0.50 0.76258 0.00104 fswds05O.out 0.75754 0.0022 rndwds050.out 0.0 0.51526 0.00099 fsdds00.0ut 0.65896 0.00227 rnddds00z.out Table 6.2-1 (cont.) K,ff as a Function of Moderator Density Variations Density (g/cc) Keff St. Dev MCNP files' K,.ff St. Dev T MCNP fileS a N Reactor "A" N Reactor "B" 1.0 0.84929 0.00067 nrwds78a.out 0.85334 0.00056 nrwds78b.out 0.98 0.84736 0.00062 nrwd98a.out 0.84978 0.00058 nrwd98b.out 0.95 0.84189 0.00066 nrwd95a.out 0.84547 0.00056 nrwd95b.out 0.90 0.83229 0.00064 nrwd789a.out 0.83613 0.00059 nrwd789b.out 0.80 0.81078 0.00060 nrwd788a.out 0.81506 0.00060 nrwd788b.out 0.50 0.70305 0.00056 nrwd50a.out 0.70516 0.00058 nrwd50b.out 0.0 0.34216 , 0.00038 , nrdds62a.out 0.34489 0.00034 nrdds62b.out Shippingpor t LWBR Shippingport PWR 1.0 0.87003 0.00101 slwds94.out 0.87972 0.00100 spwdsOO.out 0.98 0.85963 0.00112 slwds98.out 0.87777 0.00111 spwds98.out 0.95 0.84579 0.00110 slwds95.out 0.86895 0.00102 spwds95.out 0.90 0.82493 0.00105 slwds949.out 0.86277 0.00103 spwds009.out 0.80 0.77979 0.00104 slwds948.out 0.84687 0.00107 spwds008.out 0.50 0.61967 0.00096 slwds50.out 0.78996 0.00105 spwds50.out 0.0 0.22755 0.00045 , sldds78.out 0.23594 0.00044 , spddsOO.out TMI-2 " D" TMI-2 " K" 1.0 0.85634 0.00097 tmwdsl5d.out 0.80811 0.00089 trnwdsl9k.out 0.98 0.84905 0.00091 tmwd98d.out 0.80082 0.00098 tmwd98k.out 0.95 0.83840 0.00094 tmwd95d.out 0.79464 0.00089 tmwd95k.out 0.90 0.81916 0.00106 tmwdl59d.out 0.78005 0.00093 tmwdl99k.out 0.80 0.77684 0.00100 tmwdl58d.out 0.74678 0.00097 tmwdl98k.out 0.50 0.60730 0.00094 tmwd50d.out 0.60890 0.00094 tmwd50k.out 0.0 0.21237 0.00035 , tmddsl 1 d.out 0.18203 , 0.00032 trnddsl5k.out TRIGA 1.0 0.83818 0.00112 trwds60.out 0.98 0.83600 0.00105 trwds98.out a The input files to each run have the 0.95 0.83750 0.00106 trwds95.out same name as the corresponding fil i h h 0.90 0.83361 0.00104 trwds609.0ut Output e but w t out t e out extension (e g the input file matching 0.80 0.82977 0.00112 trwds608.out . ., output file trwds6O.out is trwds60). E5 0::::: F 0.80371 0.00104 trwds50.out 0.0 0.70702 0.00093 trdds60.out 0.9 0.8 0.7 0.6 0.5 0.4 0.3 0.2 0.1 0 Figure 6.2-1 DOE Fuel Moderator Density Variations versus k,,ff 0.2 0.4 0.6 0.8 1 Density (g/CM3) 6.2.2 DOE Fuel Neutron Poison Variations Variations in neutron poison loading were studied, to cover any errors of its + integration in the basket structure or fuel mixture, and Table 6.2-2 shows the variations in k,ff + as a function of neutron poison loading. Note that the Enrico Fenni and FFTF fuels have + non-integrated neutron poison (i.e., the neutron poison is fixed and a part of the basket structure) + while the Melt & Dilute fuel has integrated neutron poison in its fuel. It can be seen ftom + Table 6.2-2 that a large decrease in neutron poison will not cause kff of a single canister to exceed + the upper subcritical limit. Table 6.2-2 DOE Fuel Neutron Poison Variations Neutron Neutron Poison Keff St. Dev MCNP files a Poison Keff St. Dev MCNP files' (Vol%) (wt%) Enrico Fermi (non-integrated FFTF (non-integrated neutron neutron poison) poison) 1.0 0.88616 0.00078 efwds06 out 5.0 0 87592 0 00076 ffwdS15 out (3 kg) . (19 .26 kg) . . . 0.4 091063 0.00078 efwdsOft.out 0.1 0.92084 0.00074 ffwdsl5y.out (1.2 kg) (0-39 kg) Melt & Dilute (integrated neutron poison) 'The input files to each run have the same 0 * 5 0.39017 0.00124 mdwdsOO.out name as the corresponding output file but (4.73 kg) without the out extension (e.g., the input file 0.001 0 80902 00256 0 mdwdsOOx out matching output file mdwdsOO.out is mdwdsOO). _(0.009 kg) . . . 0 , 0001 3129 0.00268 mdwdsOOz out (9E-4 kg) I . 6.3 CATEGORY 1 AND 2 EVENT SEQUENCES , Category 1 and 2 event sequences were evaluated as presented in Section 5.2.3 + and were found to be within the criticality safety design limits. 6.4 CONCLUSIONS AND RECOMMENDATIONS The FHF and its processes have been evaluated for criticality safety for normal+ operations, Category I and 2 event sequences. The results presented in this document lead + to the following conclusions and recommendations: 9 The criticality evaluations of the cask and MSC demonstrate that the + conditions outside the overpack (e.g., spacing, moderation, reflection) have no discemable impact on + the reactivity of the cask for both PWR and BVVR fuel (also noted in Section 6. 1). Single DOE canisters are subcritical under both norinal and flooded conditions.+ Concrete surrounding the DOE canister is a better reflector than water and consequently + produces a higher keff- The DOE fuel containing neutron poison (ie., Enrico Fermi, FFTF, and Melt &+ Dilute) showed that losing a significant amount of neutron poison (due to manufacturing+ errors etc.) still promotes subcriticality. Reactivity of the loaded cask, MSC or DOE canisters decreases with reduction in+ moderator density. Maximum reactivity is reached when the cask, MSC or DOE canisters are fully + flooded with water at full density (1.0 g/cm 3). Criticality events potentially occuning in the FHF do not compromise + criticality safety. It should be recognized that the potential criticality events have not yet been + categon'zed into Category I and 2 event sequences. 7. REFERENCES 10 CFR 50. 2002. Energy: Domestic Licensing of Production and Utilization + Facilities. Readily available. 10 CFR 71. 1999. Energy: Packaging and Transportation of Radioactive Material. + Readily available. 10 CFR 72. 1999. Energy: Licensing Requirements for the Independent Storage of + Spent Nuclear Fuel and High-Level Radioactive Waste. Readily available. 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. 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. 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. TIC: 2 5 142 5. ASTM A 276-9 1 a. 199 1. Standard Specification for Stainless and + Heat-Resisting Steel Bars and Shapes. Philadelphia, Pennsylvania: American Society for Testing and Materials.+ TIC: 240022. ASTM G 1-90 (Reapproved 1999). 1999. Standord Practice for Preparing, Cleaning,+ and Evaluating Corrosion Test Specimens. West Conshohocken, Pennsylvania: American + Society for Testing and Maten'als. TIC: 238771. Briesmeister, J.F., ed. 1997. MCNP-A General Monte Carlo N-Particle Transport + Code. LA- 12625-M, Version 4B. Los Alamos, New Mexico: Los Alamos National Laboratory. + ACC: MOL. 19980624.0328. BSC (Bechtel SAIC Company) 2001 a. Intact and Degraded Mode Criticality + Calculations for the Codisposal of Fort Saint Vrain HTGR Spent Nuclear Fuel in a Waste Package. + CAL-EDC- NU-000007 REV 00. Las Vegas, Nevada: Bechtel SAIC Company. ACC: + MOL.20011015.0020. BSC (Bechtel SAIC Company) 200 lb. Intact and Degraded Mode Criticality + Calculations for the Codisposal of Melt and Dilute Ingots in a Waste Package. CAL-EDC-NU-000006 + REV 00. Las Vegas, Nevada: Bechtel SAIC Company. ACC: MOL.20010730.0062. BSC (Bechtel SAIC Company) 2001c. Statement o Work for DOE-Office of Civilian f Radioactive Waste Management, Technical Assistance on Melt-Dilute Criticality + and Shielding Analyses, Revision Z May 30, 2001. Las Vegas, Nevada: Bechtel SAIC Company. + ACC: MOL.20010619.0626. BSC (Bechtel SAIC Company) 2003. Q-List. TDR-MGR-RL-000005 REV 00. Las Vegas, Nevada: Bechtel SAIC Company. ACC: DOC.20030930.0002. BSC (Bechtel SAIC Company) 2004a. Aging Facility Criticality Safety + Calculations. 170-OOC- HAOO-00100-000-OOA. Las Vegas, Nevada: Bechtel SAIC Company. ACC: ENG.20040415.0012. BSC (Bechtel SAIC Company) 2004b. Canister Handling Facility Criticality Safety Calculations. 190-OOC-CHOO-00100-000-OOA. Las Vegas, Nevada: Bechtel SAIC + Company. ACC: ENG.20040302.0017. BSC (Bechtel SAIC Company) 2004c. Internal Hazards Analysisfor License + Application. 000- OOC-MGRO-00600-000-OOA. Las Vegas, Nevada: Bechtel SAIC Company. ACC: ENG.20040623.0032. BSC (Bechtel SAIC Company) 2004d. Naval Spent Nuclear Fuel Waste Package System Description Document. 000-3YD-DNOO-00100-0.00-003. Las Vegas, Nevada: Bechtel + SAIC Company. ACC: ENG.20040322.0006. BSC (Bechtel SAIC Company) 2004e. Preclosure Criticality Analysis Process + Report. TDR- EBS-NU-000004 REV 03. Las Vegas, Nevada: Bechtel SAIC Company. ACC: DOC.20040524.0001. BSC (Bechtel SAIC Company) 2004f. Surface Facility Criticality Safety + Calculations. 1 00-OOC- WHSO-00100-000-OOB. Las Vegas, Nevada: Bechtel SAIC Company. ACC: ENG.20040309.0001. Canori, G.F. and Leitner, M.M. 2003. Project Requirements Document. + TER-MGR-MD-000001 REV 02. Las Vegas, Nevada: Bechtel SAIC Company. ACC: DOC.20031.222.0006. Cogema. 2004. Evaluation of I OCFR72 - Licensed Casks for Use in Aging Spent + Nuclear FueL COGEMA-COI 15-RP-04-001, Rev. 0. Las Vegas, Nevada: Cogema. ACC: ENG.20040405.0025. CRWMS M&O 1998a. Software Code: MCNP. V4B2LV. HP, HPUX 9.07 and 10.20; PC, Windows 95; Sun, Solaris 2.6. 30033 V4B2LV. CRWMS M&O 1998b. Software Qualification Reportfor MCNP Version 4B2, A + General Monte Carlo N-Particle Transport Code. CSCI: 30033 V4B2LV. DI: 30033-2003, Rev. 01. + Las Vegas, Nevada: CRWMS M&O. ACC: MOL.19980622.0637. CRWMS M&O 1999a. Fast Flux Test Facility (FFTF) Reactor Fuel Criticality + Calculations. BBAOOOOOO-01717-0210-00016 REV 00. Las Vegas, Nevada: CRWMS M&O. ACC: MOL. 19990426.0142. CRWMS M&O 1999b. Special Process Disk (CD) and Supporting Documentation for+ TRIGA Fuel Phase I and II Criticality Calculation. CAL-MGR-NU-000001 REV 00. Las + Vegas, Nevada: CRWMS M&O. ACC: MOL. 19990902.0464. CRWMS M&O 1999c. Summary Report of Laboratory Critical Experiment Analyses + Performed for the Disposal Criticality Analysis Methodology. BOOOOOOOO-0 1717-5705-00076 + REV 02. Las Vegas, Nevada: CRWMS M&O. ACC: MOL. 19990920.0167. CRWMS M&O 1999d. TRIGA Fuel Phase I and 11 Criticality Calculation. + CAL-MGR-NU-00000 I REV 00. Las Vegas, Nevada: CRWMS M&O. ACC: MOL. 19991209.0195. CRWMS M&O 2000a. Enrico Fermi Fast Reactor Spent Nuclear Fuel Criticality + Calculations: Degraded Mode. CAL-EDC-N-U-000001 REV 00. Las Vegas, Nevada: CRWMS M&O. + ACC: MOL.20000802.0002. CRWMS M&O 2000b. Intact and Degraded Criticality Calculations for the + Codisposal of Shippingport LWBR Spent Nuclear Fuel in a Waste Package. CAL-EDC-NU-000004 REV + 00. Las Vegas, Nevada: CRWMS M&O. ACC: MOL.20000922.0093. CRWMS M&O 2000c. Intact and Degraded Criticality Calculations for the + Codisposal of Shippingport PWR Fuel in a Waste Package. CAL-EDC-NU-000002 REV 00. Las Vegas, Nevada: CRWMS M&O. ACC: MOL.20000209.0233. CRWMS M&O 2001. Intact and Degraded Component Criticality Calculations of N+ Reactor Spent Nuclear Fuel. CAIL-EDC-NU-000003 REV 00. Las Vegas, Nevada: CRWMS + M&O. ACC: MOL.20010223.0060. DOE (U.S. Department of Energy) 1999a. Fermi (U-Mo) Fuel Characteristics for + Disposal Criticality Analysis. DOE/SNF/REP-03 5, Rev. 0. Washington, D.C.: U.S. + Department of Energy. TIC:242461. DOE (U.S. Department of Energy) 1999b. Shippingport LWBR (ThlU Oxide) Fuel Characteristics for Disposal Criticality Analysis. DOE/SNF/REP-051, Rev. 0. + Washington, D.C.: U.S. Department of Energy, Office of Enviromnental Management. TIC: + 245631. DOE (U.S. Department of Energy) 1999c. Shippingport PWR (HEU Oxide) Fuel + Characteristics for Disposal Criticality Analysis. DOE/SNF/REP-040, Rev. 0. Washington, D.C.: + U.S. Department of Energy. TIC: 243528. DOE (U.S. Department of Energy) 2000. N Reactor (U-Metal) Fuel Characteristics + for Disposal Criticality Analysis. DOE/SNF/REP-056, Rev. 0. [Washington, D.C.]: U.S. + Department of Energy, Office of Environmental Management. TIC: 247956. DOE (U.S. Department of Energy) 2003. TMI Fuel Characteristics for Disposal + Criticality Analysis. DOE/SNF/REP-084, Rev. 0. Idaho Falls, Idaho: U.S. Department of + Energy, Idaho Operations Office. ACC: MOL.20031013.0388. 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. General Atomics. 1993. GA-4 Legal Weight Truck From-Reactor Spent Fuel Shipping+ Cask, Final Design Report. 910353/0. San Diego, Califomia: General Atomics. ACC: NNA. 19940513.0215. HOLTEC Intemational. 2002. Final Safety Analysis Reportfor the Holtec + International Storage and Transfer Operation Reinforced Module Cask System (Hi-Storm 100 Cask + System). HOLTEC Repoft HI-2002444. Two volumes. NRC Docket No. 72-1014. Marlton, New Jersey: + HOLTEC Intemational. TIC: 255899. INEEL (Idaho National Engineering and Environmental Laboratory) 2002. FFTF + (MOX) Fuel Characteristics for Disposal Criticality Analysis. DOE/SNF/REP-032, Rev. 1. + Idaho Falls, Idaho: U.S. Department of Energy, Idaho National Operations Office. TIC: + 252933. LP-SI. I I Q-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. Mecham, D.C., ed. 2004. Waste Package Component Design Methodology Report. + 000-30R- WISO-00100-000-001. Las Vegas, Nevada: Bechtel SAIC Company. ACC: ENG.20040617.0012. Minwalla, H.J. 2003. Project Design Criteria Document. + 000-3DR-MGRO-00100-000-001. Las Vegas, Nevada: Bechtel SAIC Company. ACC: ENG.20030402.0001. NRC (U.S. Nuclear Regulatory Commission) 1997. SCALE, RSIC Computer Code + Collection (CCC-545). NUREG/CR-0200, Rev. 5. Washington, D.C.: U.S. Nuclear Regulatory Commission. TIC: 235920. 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. Taylor, L.L. 2001. Fort Saint Vrain HTGR (ThlU Carbide) Fuel Characteristics + for Disposal Criticality Analysis. DOE/SNF/REP-060, Rev. 0. [Washington, D.C.]: U.S. + Department of Energy, Office of Enviromnental Management. TIC: 249783. 8. ATTACHMENTS This calculation document includes three attachments: ATTACHMENT I Listing of Computer Files (7 pages) ATTACHMENT 11 One Compact Disk Containing All Files Listed in Attachment I (I + of 1) (0 pages) ATTACHMENTIII Draft Sketches of FHF Moderator Control Areas (10 pages) ATTACHMENT I LISTING OF COMPUTER FILES This attachment lists the input and output file names for the MCNP and Excel + calculations. All input and output are stored on an electronic medium (compact disc) in ASCII + fonnat as part of this attachment. Date Time File Size File Nwne 06/07/2004 04:23p 78,848 DOE - fuel.xls 06/08/2004 12:1 lp 5,961 trwds60 06/08/2004 12:1 lp 474,856 trwds60.out 06/08/2004 12:1 lp 5,999 trwds60s 06/08/2004 12:1 lp 476,188 trwds60s.out 02/26/2004 04:54p 11,90S efwds068 02/26/2004 10:23p 654,776 efwds068.out 02/26/2004 04:54p 11,905 efwds069 02/26/2004 05:50p 654,776 efwds069.out 06/08/2004 12:09p 11,918 efwds650 06/08/2004 12:09p 654,956 efwds650.out 06/08/2004 12:09p 11,918 efwds695 06/08/2004 12:09p 654,642 efwds695.out 06/08/2004 12:09p 11,918 efwds698 06/08/2004 12:09p 654,743 efwds698.out 02/26/2004 04:55p 15,398 ffwds 15 8 02/26/2004 11:26p 676,100 ffwdsl58.out 02/26/2004 04:55p 15,398 ffwds I S 9 02/26/2004 06:51p 676,100 ffwdsl59.out 06/08/2004 12:10p 15,402 ffwds50 06/08/2004 12:10p 676,601 ffwds50.out 06/08/2004 12:10p 15,402 ffwds95 06/08/2004 12:10p 676,178 ffwds95.out 06/08/2004 12:10p 15,402 ffwds98 06/08/2004 12:10p 676,279 ffwds98.out 02/26/2004 04:57p 11,053 fswds008 02/26/2004 11:57p 520,020 fswds008.out 02/26/2004 04:57p 11,053 fswds009 02/26/2004 07:23p 520,020 fswds009.out 06/08/2004 12:10p 11,083 fswds05O 06/08/2004 12:10p 519,950 fswds05O.out 06/08/2004 12:10p 11,069 fswds095 06/08/2004 12:10p 520,066 fswds095.out 06/08/2004 12:10p 11,083 fswds098 06/08/2004 12:10p 519,950 fswds098.out 02/26/2004 04:59p 5,540 mdwds008 02/27/2004 12:00a 342,086 mdwdsOO8.out Date Time File Size File Name 02/26/2004 04:59p 5,540 mdwds009 02/26/2004 07:26p 342,086 mdwds009.out 06/08/2004 12: 1 Op 5,539 mdwds05O 06/08/2004 12: 1 Op 342,085 mdwds05O.out 06/08/2004 12: 1 Op 5,539 mdwds095 06/08/2004 12: 1 Op 342,085 mdwds095.out 06/08/2004 12: 1 Op 5,539 mdwds098 06/08/2004 12: 1 Op 342,220 mdwds098.out 06/08/2004 12: 1 Op 7,302 nrwd50a 06/08/2004 12: 1 Op 413,605 nrwd50a.out 06/08/2004 12: 1 Op 7,324 nrwd50b 06/08/2004 12: 1 Op 414,896 nrwd50b.out 02/26/2004 05:03p 7,302 nrwd788a 02/27/2004 12:20a 415,067 nrwd788a.out 02/26/2004 05:04p 7,324 nrwd788b 02/27/2004 12:39a 413,820 nrwd788b.out 02/26/2004 05:03p 7,302 nrwd789a 02/26/2004 07:45p 415,067 nrwd789a.out 02/26/2004 05:04p 7,324 nrwd789b 02/26/2004 08:04p 414,972 nrwd789b.out 06/08/2004 12: 1 Op 7,302 nrwd95a 06/08/2004 12: 1 Op 414,439 nrwd95a.out 06/08/2004 12: 1 Op 7,324 nrwd95b 06/08/2004 12: 1 Op 414,112 nrwd95b.out 06/08/2004 12: 1 Op 7,302 nrwd98a 06/08/2004 12: 1 Op 414,475 nrwd98a.out 06/08/2004 12: 1 Op 7,324 nrwd98b 06/08/2004 12: 1 Op 414,896 nrwd98b.out 06/08/2004 12: 1 Op 11,758 s1wds50 06/08/2004 12: 1 Op 491,718 s1wds50.out 02/26/2004 05:05p 11,741 slwds948 02/27/2004 01:07a 491,504 slwds948.out 02/26/2004 05:05p 11,741 s1wds949 02/26/2004 08:3 lp 491,504 slwds949.out 06/08/2004 12: 1 Op 11,758 slwds95 06/08/2004 12: 1 Op 491,765 slwds95.out 06/08/2004 12: 1 Op 11,758 slwds98 06/08/2004 12: 1 Op 491,718 slwds98.out 02/26/2004 05:06p 27,582 spwds008 02/27/2004 01:50a 796,106 spwdsOO8.out 02/26/2004 05:05p 27,582 spwds009 02/26/2004 09:15p 796,106 spwds009.out 06/08/2004 12: 1 Op 27,587 spwds05O 06/08/2004 12: 1 Op 796,371 spwds05O.out Date Time File Size File Name 06/08/2004 12: 1 Op 27,587 spwds095 06/08/2004 12: 1 Op 796,573 spwds095.out 06/08/2004 12: 1 Op 27,587 spwds098 06/08/2004 12: 1 Op 796,674 spwds098.out 02/26/2004 05:12p 6,457 tmwdl58d 02/27/2004 02:07a 445,923 tmwdl58d.out 02/26/2004 05:12p 6,457 tmwdl59d 02/26/2004 09:31p 445,916 trnwd159d.out 02/27/2004 06:54a 7,618 trnwd198k 02/27/2004 07:20a 467,153 tmwdl98k.out 02/27/2004 06:54a 7,618 tmwdl99k 02/27/2004 07:45a 467,153 tmwdl99k.out 06/08/2004 12: 1 Op 6,457 tmwd50d 06/08/2004 12: 1 Op 445,868 tmwd50d.out 06/08/2004 12: 1 Op 7,630 tmwd50k 06/08/2004 12: 1 Op 467,423 tmwd50k.out 06/08/2004 12: 1 Op 6,457 tmwd95d 06/08/2004 12: 1 Op 445,969 tmwd95d.out 06/08/2004 12: 1 Op 7,630 tmwd95k 06/08/2004 12: 1 Op 467,524 tmwd95k.out 06/08/2004 12: 1 Op 6,457 tmwd98d 06/08/2004 12: 1 Op 445,969 tmwd98d.out 06/08/2004 12: 1 Op 7,630 tmwd98k 06/08/2004 12: 1 Op 467,421 tmwd98k.out 06/08/2004 12: 1 Op 5,980 trwds50 06/08/2004 12: 1 Op 464,230 trwds50.out 02/26/2004 05:07p 5,989 trwds608 02/27/2004 02:22a 463,651 trwds608.out 02/26/2004 05:07p 5,989 trwds609 02/26/2004 09:46p 463,333 trwds609.out 06/08/2004 12: 1 Op 5,981 trwds95 06/08/2004 12: 1 Op 464,064 trwds95.out 06/08/2004 12: 1 Op 5,982 trwds98 06/08/2004 12: 1 Op 464,162 trwds98.out 06/08/2004 01:09P 12,243 efddsOIx 06/08/2004 01:09P 655,227 efddsOlx.out 06/08/2004 01:09P 11,878 efwdsOl 06/08/2004 01:09P 654,644 efwdsOl.out 06/08/2004 01:10P 11,878 efwds02 06/08/2004 01:10P 654,157 efwds02.out 06/08/2004 01:10P 11,878 efwds03 06/08/2004 01:10P 655,167 efwds03.out 06/08/2004 01:10P 11,878 efwds04 06/08/2004 01:10P 654,776 efwds04.out Date Time File Size File Name 06/08/2004 01:10P 11,878 efwds05 06/08/2004 01:10P 654,776 efwds05.out 06/08/2004 01:10P 11,878 efwds06 06/08/2004 01:10P 654,776 efwds06.out 06/08/2004 01:10P 11,879 efwds06x 06/08/2004 01:10P 654,776 efwds06x.out 06/08/2004 01:10P 11,878 efwds07 06/08/2004 01:10P 655,167 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