Dry Transfer Facility Criticality Safety Calculations Rev 00C, ICN 00 100-00C-WHS0-00100-000-00C May 2005 1. PURPOSE This design calculation updates the previous criticality evaluation for the fuel handling, transfer, and staging operations to be performed in the Dry Transfer Facility (DTF) including the remediation area. The purpose of the calculation is to demonstrate that operations performed in the DTF and RF meet the nuclear criticality safety design criteria specified in the Project Design Criteria (PDC) Document (BSC 2004 [DIRS 171599], Section 4.9.2.2), the nuclear facility safety requirement in Project Requirements Document (Canori and Leitner 2003 [DIRS 166275], p. 4-206), the functional/operational nuclear safety requirement in the Project Functional and Operational Requirements document (Curry 2004 [DIRS 170557], p. 75), and the functional nuclear criticality safety requirements described in the Dry Transfer Facility Description Document (BSC 2005 [DIRS 173737], p. 3-8). A description of the changes is as follows: • Update the supporting calculations for the various Category 1 and 2 event sequences as identified in the Categorization of Event Sequences for License Application (BSC 2005 [DIRS 171429], Section 7). • Update the criticality safety calculations for the DTF staging racks and the remediation pool to reflect the current design. This design calculation focuses on commercial spent nuclear fuel (SNF) assemblies, i.e., pressurized water reactor (PWR) and boiling water reactor (BWR) SNF. U.S. Department of Energy (DOE) Environmental Management (EM) owned SNF is evaluated in depth in the Canister Handling Facility Criticality Safety Calculations (BSC 2005 [DIRS 173284]) and is also applicable to DTF operations. Further, the design and safety analyses of the naval SNF canisters are the responsibility of the U.S. Department of the Navy (Naval Nuclear Propulsion Program) and will not be included in this document. Also, note that the results for the Monitored Geologic Repository (MGR) Site specific Cask (MSC) calculations are limited to the specific design chosen (see Assumption 3.4). A more current design will be included in the next revision of the criticality calculations for the Aging Facility. In addition, this calculation is valid for the current design as provided in Attachment III of the DTF and may not reflect the ongoing design evolution of the facility. However, it is anticipated that design changes to the facility layout will have little or no impact on the criticality results and/or conclusions presented in this document. This calculation is subject to the Quality Assurance Requirements and Description (DOE 2004 [DIRS 171539]) because the DTF and the staging racks/baskets in the remediation pool are included as safety category items in the Q-List (BSC 2005 [DIRS 171190], pp. A-4 and A-7) as items important to safety. This calculation is prepared in accordance with AP-3.12Q, Design Calculations and Analyses [DIRS 168413]. Licensing & Nuclear Safety Calculation Title: Dry Transfer Facility Criticality Safety Calculations Document Identifier: 100-00C-WHS0-00100-000-00C Page 8 of 47 2. METHOD 2.1 CRITICALITY SAFETY ANALYSIS The criticality safety calculations presented in this document evaluate fuel handling, transfer and staging operations in the DTF and remediation area to ensure they meet the criticality safety requirements under normal conditions as well as for Category 1 and 2 event sequences, in accordance with 10 CFR 63 [DIRS 173273]. Minimum spacing for PWR and BWR fuel storage racks in the DTF and remediation pool is determined along with varied neutron poison (Boral) panel thickness. The off-normal and accident conditions are inclusive of Category 1 and 2 event sequences as defined in 10 CFR 63.2 [DIRS 173273]. Moderator and reflector conditions are varied to find the most reactive configuration. The process and methodology for criticality safety analysis given in the Preclosure Criticality Analysis Process Report (BSC 2004 [DIRS 172058], Section 2.2.7) will be implemented in these calculations, and the following method will be followed: • The criticality safety design of the facility will be based on the most reactive fuel. • The effective neutron multiplication factor (keff), including all biases and uncertainties at a 95 % confidence level, will not exceed 0.95 under all credible normal, and Category 1 and 2 event sequences (NRC 2000 [DIRS 149756], Section 8.4.1.1). • Conservative modeling assumptions leading to maximum reactivity for dimensional variables (e.g., pitch and manufacturing tolerances for assemblies) will be used. • Conservative modeling assumptions will also be used regarding materials in the fuel including no accounting for burnable poisons in fuel, no credit for 234U and 236U in fuel, no credit for fission products or transuranic absorbers in fuel, and use of the most reactive fuel stack density. • Fixed neutron absorbers used for criticality control can only take credit for up to 75% of the neutron absorbing content (NRC 2000 [DIRS 149756], Section 8.4.1.1). Note that the terms “model(s)” and “modeling” as used in this calculation document refer to the geometric configurations of the criticality cases analyzed and not scientific models per LP-SIII.10QBSC, Models [DIRS 172972]. These calculations use the qualified software MCNP (CRWMS M&O 1998 [DIRS 154060]). MCNP is a three-dimensional Monte Carlo particle transportation code with the capability to calculate eigenvalues. The Nuclear Regulatory Commission (NRC) accepts MCNP in NUREG-1567 (NRC 2000 [DIRS 149756], 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 Section 5.1.2 of AP-3.13Q, Design Control [DIRS 167460]. 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 II). Licensing & Nuclear Safety Calculation Title: Dry Transfer Facility Criticality Safety Calculations Document Identifier: 100-00C-WHS0-00100-000-00C Page 9 of 47 3. ASSUMPTIONS 3.1 LOADING OF STAGING RACKS AND POOL Assumption: It is assumed that the DTF staging racks and remediation pool are full. Rationale: In reality, the storage racks may not always be full. However, for modeling purposes this should give maximum reactivity. Confirmation Status: This assumption requires no further confirmation based on the stated rationale. Use in the Calculation: Sections 5.1.4.1, 5.1.5.1, 5.2.1, and 5.2.2. 3.2 HYDRAULIC FLUID COMPOSITION Assumption: It is assumed that the hydraulic fluid used as an alternative moderator material was a conventional silicone fluid (polysiloxane fluid) with a degree of polymerization of four (Gelest 2004 [DIRS 169915], p. 11). Rationale: The basis for this assumption is that the DTF design has not identified the hydraulic fluid for lubrication, but the candidate material used for lubrication is expected to be less effective as a moderator than water. The material used for this calculation to demonstrate criticality safety is a common silicone based hydraulic fluid (Gelest 2004 [DIRS 169915], p. 7). It should be recognized that this material might not represent the most reactive condition. Any change in the design selection of the hydraulic fluid would require re-evaluation of criticality safety to ensure that all the criticality design criteria are met. Confirmation Status: This assumption requires no further confirmation based on the stated rationale. Use in the Calculation: Section 5.1.7. 3.3 WP BREACH Assumption: It is assumed that a WP failure caused by drops and/or collisions is a non-credible event (i.e., beyond Category 2). Rationale: The primary mechanisms that can cause the WP to breach (e.g., drops or collisions) have been screened in Categorization of Event Sequences for License Application (BSC 2005 [DIRS 171429]), and show that drops onto sharp objects are beyond Category 2 (BSC 2005 [DIRS 171429], Table III-9), as well as collisions are beyond Category 2 (BSC 2005 [DIRS 171429], Table III-9). Therefore, it is reasonable to assume that a breach of a WP is a beyond Category 2 event. Confirmation Status: This assumption requires no further confirmation based on the stated rationale. Use in the Calculation: Section 6.3 Licensing & Nuclear Safety Calculation Title: Dry Transfer Facility Criticality Safety Calculations Document Identifier: 100-00C-WHS0-00100-000-00C Page 10 of 47 3.4 MSC DESIGN Assumption: The Monitored Geologic Repository (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 BWR fuel that is part of the Holtec HISTORM 100 cask system (Holtec International 2002 [DIRS 168494]). Rationale: Since the MSC is still being developed, the criticality control features will be made similar to the existing NRC-certified storage casks. Confirmation Status: This assumption requires no further confirmation based on the stated rationale. Use in the Calculation: Sections 1, 6.3, and 6.4. 3.5 SELECTION OF REPRESENTATIVE TRANSPORTATION CASK Assumption: It is assumed that the HI-STAR transportation cask is a representative and bounding cask for operations in the DTF. Rationale: Performance requirement 3.1.1.3.1 of the Transportation Cask Receipt/Return Facility Description Document (BSC 2005 [DIRS 171702]) states that no decision has been made as to which transportation casks will be used. However, the following approved casks by the NRC should be bounding for operations of CSNF: (1) GA-4; (2) HI-STAR; (3) NAC-UMS; (4) TS-125; (5) NAC-STC; (6) NUHOMS-MP-187; (7) TN-68; and (8) NUHOMS-MP-197. Tables 3.6-1 and 3.6-2 compare keff values along with enrichment of the various transportation casks. It can be seen that the HI-STAR cask has one of the highest keff values for PWR fuel and the highest keff value for BWR fuel in addition to a relatively high enrichment. Therefore, HI-STAR is an acceptable choice for a representative cask for this calculation. The selection of this cask is bounding for normal conditions of transport under 10 CFR 71 [DIRS 171308]. Further evaluation should be made in future revisions of this calculation to address the applicability to repository conditions including normal operations and Category 1 and Category 2 conditions. Confirmation Status: This assumption requires no further confirmation based on the stated rationale. Use in the Calculation: Section 5.1.8. Licensing & Nuclear Safety Calculation Title: Dry Transfer Facility Criticality Safety Calculations Document Identifier: 100-00C-WHS0-00100-000-00C Page 11 of 47 Table 3.6-1 Comparison of Most Reactive PWR Fuel in Various Transportation Casks PWR Fuel Cask Type Fuel Type Enrichment (wt% 235U) keff St. Dev. Reference GA-4 W 15x15 OFA 3.15 0.9300 0.0011 General Atomics 1998 [DIRS 103042], Table 6.4-2 HI-STAR B&W 15x15 4.1 0.9395 a N/A Holtec International 2003 [DIRS 172633], p. 6.2-27 NAC-UMS W 17x17 OFA 4.2 0.9247 0.0009 NAC International 2002 [DIRS 164612], Section 6.1 & Table 6.1-1 TS-125 (W21 canister) W 17x17 OFA 4.6 0.94025 0.00093 Sisley, S.E. 2002 [DIRS 171545], Table 6.4-6 NAC-STC Framatome- Cogema 17x17 4.5 0.92541 0.00086 Thompson, T.C. 2003 [DIRS 169362], Section 6.1.1 NUHOMS-MP-187 B&W 15x15 3.43 0.9374 b N/A Transnuclear West 2002 [DIRS 161510], Table 6.1-1 & Table 6.2-1 TN-68 Not applicable to PWR fuel NUHOMS-MP-197 Not applicable to PWR fuel a maximum keff. b keff includes 2 standard deviations uncertainty. Table 3.6-2 Comparison of Most Reactive BWR Fuel in Various Transportation Casks BWR Fuel Cask Type Fuel Type Enrichment (wt% 235U) keff St. Dev. Reference GA-4 Not applicable to BWR fuel HI-STAR GE 8x8 a 4.2 0.9384 0.0007 Holtec International 2003 [DIRS 172633], p. 6.2-36 NAC-UMS Exxon/ANF 9x9 4.0 0.9055 0.0008 NAC International 2002 [DIRS 164612], Section 6.1 & Table 6.1-1 TS-125 (W74 canister) Siemens 11x11 4.1 0.93422 0.00089 Sisley, S.E. 2002 [DIRS 171545], Table 6.4-10 NAC-STC Not applicable to BWR fuel NUHOMS-MP-187 Not applicable to BWR fuel TN-68 GE 10x10 3.7 0.9250 0.0008 Transnuclear 2001 [DIRS 167988], Sections 6.1 & 6.4.2 NUHOMS-MP-197 GE 10x10 4.4 0.9349 0.0011 Transnuclear 2002 [DIRS 160736], Sections 6.4.4 & 6.4.2 a Some other fuel assembly types have higher keff values (see Section 5.1.2). Licensing & Nuclear Safety Calculation Title: Dry Transfer Facility Criticality Safety Calculations Document Identifier: 100-00C-WHS0-00100-000-00C Page 12 of 47 3.6 SELECTION OF REPRESENTATIVE FUEL BASKET Assumption: It is assumed that the fuel basket design used in both the HI-STAR and HI-STORM cask systems is a representative fuel basket design for staging operations in the DTF and remediation pool. Rationale: Storage rack cells, in the United States, designed to accommodate spent fuel are generally composed of stainless steel walls with a fixed neutron absorber panel (e.g., Boral) on each side (Wagner and Parks 2000 [DIRS 163140], p.7). This is consistent with the fuel baskets featured in the Holtec cask systems. Therefore, it is acceptable to use the fuel basket design used in the HISTAR/ HI-STORM cask systems as a representative fuel basket for this calculation. Confirmation Status: This assumption requires no further confirmation based on the stated rationale. Use in the Calculation: Section 5.1.6. Licensing & Nuclear Safety Calculation Title: Dry Transfer Facility Criticality Safety Calculations Document Identifier: 100-00C-WHS0-00100-000-00C Page 13 of 47 4. USE OF COMPUTER SOFTWARE 4.1 BASELINED SOFTWARE 4.1.1 MCNP The MCNP code (CRWMS M&O 1998 [DIRS 154060]) was used to calculate keff for all systems presented in this report. The software specifications are as follows: • Program Name: MCNP (CRWMS M&O 1998 [DIRS 154060]) • Version/Revision Number: Version 4B2LV • Status/Operating System: Qualified/HP-UX B.10.20 • Software Tracking Number: 30033 V4B2LV • 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 II) and the files are listed in Attachment I. 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 [DIRS 103897]) and CRWMS M&O (1998 [DIRS 102836]), 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: Microsoft® Excel 97 SR-2 • This version is installed on a PC running Microsoft Windows 2000 with CPU number 503009 The file for the Excel calculation is contained on a CD (Attachment II) and the file is listed in Attachment I. Excel was used to illustrate data in Section 6.3.2. Excel is exempt from qualification per Section 2.1.6 of LP-SI.11Q-BSC, Software Management [DIRS 171923]. Licensing & Nuclear Safety Calculation Title: Dry Transfer Facility Criticality Safety Calculations Document Identifier: 100-00C-WHS0-00100-000-00C Page 14 of 47 5. CALCULATION All technical product inputs and sources of the inputs used in the development of this calculation are documented in this section. Attachment III features General Arrangement Drawings of the DTF and its remediation area as of the date of this calculation, and may not reflect the ongoing design evolution. The purpose of these drawings is to show the functional areas where the SNF will be handled and staged as well as to show the area where moderator control is required. 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 (BSC 2004 [DIRS 171599]) are used in these calculations. The pertinent criteria for DTF criticality include the following (BSC 2004 [DIRS 171599], Section 4.9.2.2): • The multiplication factor, keff, including all biases and uncertainty at a 95 percent confidence level, shall not exceed 0.95 under all normal conditions, and Category 1 and Category 2 event sequences. • For fixed-neutron absorbers used for criticality control such as grid plates or inserts, no more than 75 percent credit of the neutron absorber content is used for preclosure criticality analyses, unless standard acceptance tests verify that the presence and uniformity of the neutron absorber are more effective. • To use moderator control for nuclear criticality safety during preclosure for a facility, the facility shall limit the amount of moderator that may be present in any area where radioactive waste is being handled to show that there is no criticality concern under all normal conditions, Category 1 event sequences, and Category 2 event sequences. The Project Requirements Document seeks to ‘prevent unplanned nuclear criticality events’ through requirement PRD-015/P-096 (Canori and Leitner 2003 [DIRS 166275], p. 4-206). From an operational/performance requirement point of view, ‘the staging racks … shall be designed to maintain criticality safety’ per the Project Functional and Operational Requirements document (Curry 2004 [DIRS 170557], Section 1.1.6-4). The functional requirement 3.2.3.1 of the Dry Transfer Facility Description Document (BSC 2005 [DIRS 173737], p. 3-8) states that the “design of the DTF spacing and neutron absorbers shall ensure that there is no credible criticality event under all normal conditions and Category 1 and 2 event sequences”. 5.1.2 Fuel Assembly Selection The Westinghouse 17x17 optimized fuel assembly (OFA) was chosen as a representative PWR fuel assembly for the DTF and remediation pool staging rack criticality calculations. The basket cell designs utilized in this calculation is that of the HI-STAR 100 cask system (Holtec International 2003 [DIRS 172633]). Studies have shown that either the B&W 15x15 or the Westinghouse 17x17 OFA (optimized fuel assembly) are the most reactive fuel assemblies inside the HI-STAR basket cells (Holtec International 2003 [DIRS 172633], p. 6.2-2). Therefore, the Westinghouse 17x17 OFA Licensing & Nuclear Safety Calculation Title: Dry Transfer Facility Criticality Safety Calculations Document Identifier: 100-00C-WHS0-00100-000-00C Page 15 of 47 was chosen for the DTF and remediation area staging rack calculations. Evaluations of the HI-STAR transportation cask were also considered for Category 2 event sequences in the DTF. The Babcock & Wilcox (B&W) 15x15 fuel assembly (4.1 wt% enriched) was chosen as a representative fuel assembly inside the multi-purpose canister (MPC) for PWR fuel (note that MPC is a name of a component of the HI-STAR cask system). This is consistent with previously performed HI-STAR transportation cask calculations (BSC 2005 [DIRS 173616]). The GE 8x8 fuel assembly was selected as representative of BWR fuel, because it is the most numerous design, inside the HI-STAR 100 basket cell design. Studies have shown that an 8x8 fuel array is one of the more reactive assembly configurations in the design inventory (Holtec International 2003 [DIRS 172633], Tables 6.2.20 – 6.2.36). There is a 10x10 assembly design that produces a slightly higher keff than the 8x8 fuel assembly (Table 6.2.32). However, this 10x10 fuel assembly design is not an actual assembly design, due to artificial bounding parameters for maximizing reactivity (Holtec International 2003 [DIRS 172633], p. 6.2-3), and was therefore not considered for this evaluation. A 9x9 fuel assembly design (Holtec International 2003 [DIRS 172633], Table 6.2.27) also produced a slightly higher keff value than the selected 8x8 design. For completeness, one calculation was performed with this 9x9 fuel design in the staging racks to show the difference in calculated keff values (see Section 6.1). 5.1.3 Upper Subcritical Limit The upper subcritical limit (USL) is the value characterized by statistical tolerance limits that account for biases and uncertainties associated with the criticality code trending process, any uncertainties due to extrapolation outside the range of experimental data, or limitations in the geometrical or material representations used in the computational method, and an administrative margin-to-criticality that can reduce the actual critical limit. The critical limit is a limiting value of keff at which a configuration is considered potentially critical (BSC 2004 [DIRS 172058], Section 3.4.1). The USL for CSNF (both PWR and BWR fuel) used in this calculation is 0.9472 as a limit in order to meet the design criteria that keff shall not exceed 0.95 including bias and uncertainties at 95% confidence level (BSC 2004 [DIRS 171599], Section 4.9.2.2). This number was calculated based on applicable code bias for similar fuel type and enrichment range of this analysis that has been estimated to be 0.0021 (value increased by truncation) with a standard deviation of ± 0.0007 (Holtec International 2003 [DIRS 172633], Appendix 6.A-2). Consequently, the USL provides a margin of 0.0028 (0.95 – (0.0021+0.0007) = 0.9472) to account for code bias and uncertainties at 95% confidence level. Note that the uncertainties associated with the MCNP calculated keff values are not included in the USL (see discussion in the last paragraph of this section). All evaluations utilizing the HI-STAR/HI-STORM 100 cask system design are using worst case combination of manufacturing tolerances with respect to criticality determined by Holtec International (Holtec International 2003 [DIRS 172633], p. 6.3-2). The values presented in Sections 5.1.6.1, 5.1.6.2, and 5.1.8 of this document include the tolerances that maximize criticality potential. Evaluations have been performed to determine the effects of tolerances (Holtec International 2003 [DIRS 172633], Tables 6.3.1 & 6.3.2). It was determined that design parameters important to criticality safety are fuel enrichment, the inherent geometry of the fuel basket structure and the fixed neutron absorbing panels (Boral) (Holtec International 2003 [DIRS 172633], p. 6.3-3). Licensing & Nuclear Safety Calculation Title: Dry Transfer Facility Criticality Safety Calculations Document Identifier: 100-00C-WHS0-00100-000-00C Page 16 of 47 For preclosure, the USL criterion is used for comparison to predicted keff values to determine a configuration’s potential of exceeding the keff limit that ensures subcriticality. In equation notation, the definition of the USL criterion is (BSC 2004 [DIRS 172058], Section 3.5): kS + .kS = USL (1) where kS is the MCNP calculated value for the system, .kS is an allowance for (a) statistical or convergence uncertainties, or both in the computation of ks, (b) material and fabrication tolerances, and (c) uncertainties due to the geometric or material representations used in the computational method. Based on using a bounding representation in this calculation, items (b) and (c) were eliminated. The uncertainty associated with the MCNP calculated value is accounted for by adding 2 times the standard deviation to the calculated keff value and compare against USL. As an example, if the standard deviation associated with the MCNP calculated value for CSNF fuel is 0.00028 (see Section 6), the MCNP calculated value may not exceed 0.94664 (0.94664 + 2x0.00028=0.9472) in order to not violate the USL. For a more detailed description of USL determination and criterion, see BSC 2004 [DIRS 172058] (Sections 3.4.1, 3.4.2, and 3.5). 5.1.4 DTF Staging Rack Calculation Inputs The staging racks in the DTF were modeled as they are currently designed in accordance with Attachment III. Physical inputs for the staging racks are described in the following subsection. 5.1.4.1 DTF Staging Rack Configuration and Physical Dimensions The various layout plans considered for the staging tubes are shown in Attachment IV. The staging rack configuration for the DTF PWR and BWR fuel assemblies for this calculation consists mainly of a 9 x 11 array of 16 in. (outside diameter) single staging tubes. The tubes were modeled with 0.5 in. thickness of stainless steel. A sensitivity study was performed with a 1.0 in. thick stainless steel tube (the distance between assemblies was maintained the same) to ensure that a varied thickness would not impact the keff values (see results in Section 6.1). The spacing between the tubes has not been determined (that is part of this evaluation), and the different spacing distances considered are 10 in., 12 in., and 16 in. of concrete between the tubes. The calculation considers the full length of the spacing between the tubes to consist of concrete. It should be mentioned that the current design plans do not include concrete for the full length (only approximately the top 5 ft). However, including the full length of concrete in the MCNP model will be conservative from a criticality perspective. In addition to the 9 x 11 array configuration, the 10 in. concrete spacing also considers a 10x12 array configuration with staggered lines (triangular pitch). Attachment IV displays the various layout plans and the dimensions. For conservatism in criticality safety analysis, the MCNP models of the DTF fuel staging racks are always modeled as full (Assumption 3.1). In addition, only the active fuel region is modeled in the MCNP models (the active fuel length is extended slightly to cover the end-fittings regions). A Licensing & Nuclear Safety Calculation Title: Dry Transfer Facility Criticality Safety Calculations Document Identifier: 100-00C-WHS0-00100-000-00C Page 17 of 47 previous study has shown that the fuel rod can be modeled either with or without end-regions included for criticality calculations (BSC 2005 [DIRS 173616], Table 5.2-1). The DTF staging tube was modeled with a length of 27 ft and the top of the tube beginning 5 ft below the floor, comprising a total length of 32 ft. The total length was estimated from the General Arrangement Drawing presented in Attachment III, p. III-24. The area below the staging tube, that holds the fuel assembly, was modeled as a stainless steel structure. 5.1.5 Remediation Pool Calculation Inputs The pool in the RF was modeled per the current design as shown in Attachment III. Physical inputs for the RF pool are described in the following subsection. 5.1.5.1 Remediation Pool Configuration and Physical Dimensions The remediation pool allows for 14 staging racks placed in 2 rows in accordance with p. III-13 of Attachment III. Each staging rack is 3’-4”x 3’-4” and can hold a 3x3 array of PWR fuel assemblies and a 4x4 array of BWR fuel assemblies (Cogema 2003 [DIRS 167153]). The width of the pool is 59 ft. and 3 in. and the pool along with the crane maintenance platform is 55 ft. deep (Cogema 2004 [DIRS 173297]). To minimize neutron leakage and account for reflection, the pool itself was modeled as 55 ft deep whereof 5 ft is a bottom steel plate. The maximum spacing of the first to last staging rack in one row is limited to 35’-1” due to travel capacity of the bridge crane (Cogema 2004 [DIRS 173297). In addition, the stainless steel liner surrounding the pool is 5 ft. wide (BSC 2004 [DIRS 170728]). As with the DTF staging rack, the MCNP models of the staging racks in the remediation pool are always modeled as full (Assumption 3.1). In addition, only the active fuel region is modeled in the MCNP models (the active fuel length is extended slightly to cover the endfittings regions). 5.1.6 Basket Cell and Fuel Assembly Calculation Inputs Since a final design of the basket cells contained in the staging racks has not been selected, the HISTAR 100 cask system cell baskets were utilized (Assumption 3.6). The following sub-sections describe the basket cell geometry and the fuel assembly dimensions. 5.1.6.1 PWR Basket Cell and Fuel Assembly Dimensions The PWR fuel cell basket features SS walls with a Boral panel situated on each side. The PWR fuel cell basket contains a Westinghouse 17 x 17 OFA (Section 5.1.2). Figure 5.1-1 displays the cell basket with the Westinghouse 17 x 17 OFA and Table 5.1-1 features the dimensions of the cell basket. Table 5.1-2 displays the specifications of the PWR fuel assembly. Licensing & Nuclear Safety Calculation Title: Dry Transfer Facility Criticality Safety Calculations Document Identifier: 100-00C-WHS0-00100-000-00C Page 18 of 47 NOTE: Dimensions are in inches. Figure is not to scale. Figure 5.1-1 PWR Fuel Basket Cell Containing Westinghouse 17 x 17 OFA (Source: Holtec International 2003 [DIRS 172633], Figure 6.3.1) Table 5.1-1 Radial Dimensions of the Basket Cell Geometry Component Dimension (cm) Reference Cell box inside dimension 22.8092 (8.98 in.) Holtec International 2003 [DIRS 172633], Table 6.3.3 Cell pitch 27.7012 (10.906 in.) Holtec International 2003 [DIRS 172633], Table 6.3.3 Cell wall thickness (SS) 0.79375 (0.3125 in.) Holtec International 2003 [DIRS 172633], Figure 6.3.1 SS sheathing 0.05969 (0.0235 in.) Holtec International 2003 [DIRS 172633], Figure 6.3.1 Boral thickness 0.1397 (0.055 in.) Holtec International 2003 [DIRS 172633], Figure 6.3.1 Al thickness (Clad) 0.0254 (0.010 in.) Holtec International 2003 [DIRS 172633], Figure 6.3.1 Boral width 19.05 (7.50 in.) Holtec International 2003 [DIRS 172633], Figure 6.3.1 Boral clearance gap 0.00889 (0.0035 in.) Holtec International 2003 [DIRS 172633], Figure 6.3.1 Licensing & Nuclear Safety Calculation Title: Dry Transfer Facility Criticality Safety Calculations Document Identifier: 100-00C-WHS0-00100-000-00C Page 19 of 47 Table 5.1-2 Specifications of the PWR W 17 x 17 OFA Parameter Dimension (cm) Reference b Rod pitch 1.2598 (0.496 in.) Holtec International 2002 [DIRS 168494], p. 2.1-11 Active fuel length 381.0 (150.0 in.) Holtec International 2002 [DIRS 168494], p. 2.1-11 Cladding outside diameter 0.9144 (0.360 in.) Holtec International 2002 [DIRS 168494], p. 2.1-11 Cladding inside diameter 0.8001 (0.315 in.) Holtec International 2002 [DIRS 168494], p. 2.1-11 Pellet outside diameter 0.784352 (0.3088 in.) Holtec International 2002 [DIRS 168494], p. 2.1-11 Guide/instrument tube outside diameter 1.204 Sanders and Wagner 2002 [DIRS 163141], p.8 Guide/instrument tube thickness 0.04064 (0.016 in.) Holtec International 2002 [DIRS 168494], p. 2.1-11 Array size 17 x 17 Holtec International 2002 [DIRS 168494], p. 2.1-11 Number of fuel rods 264 Holtec International 2002 [DIRS 168494], p. 2.1-11 Number of guide/instrument tubes a 25 Holtec International 2002 [DIRS 168494], p. 2.1-11 a Locations of guide tubes are shown in Figure 5.1-1. b Holtec International 2002 [DIRS 168494], p. 6.2-37 demonstrates that the dimensions cited are conservative 5.1.6.2 BWR Basket Cell and Fuel Assembly Dimensions The BWR fuel cell basket features SS walls with a Boral panel situated on the top and on the left hand side. The BWR fuel cell basket contains the GE 8x8 assembly (see Section 5.1.2). Figure 5.1-2 displays the basket cell with the GE 8x8 assembly. Table 5.1-3 provides the radial dimensions of the basket cell while Table 5.1-4 displays the specifications of the 8x8 BWR fuel assembly. For completeness, one calculation was also performed for a 9x9 fuel array arrangement to ensure criticality safety and most reactive BWR fuel configuration (see Section 5.1.2). Table 5.1-5 displays the specifications of the 9x9 BWR fuel assembly. Licensing & Nuclear Safety Calculation Title: Dry Transfer Facility Criticality Safety Calculations Document Identifier: 100-00C-WHS0-00100-000-00C Page 20 of 47 NOTE: Dimensions are in inches. Figure not to scale. Figure 5.1-2 BWR Fuel Basket Cell Containing GE 8x8 Assembly (Source: Holtec International 2003 [DIRS 172633], Figure 6.3.3) Table 5.1-3 Radial Dimensions of the BWR Basket Cell Component Dimension (cm) Reference Cell box inside dimension 15.2222 (5.993 in.) Holtec International 2003 [DIRS 172633], Table 6.3.3 Cell pitch 16.3322 (6.43 in.) Holtec International 2003 [DIRS 172633], Table 6.3.3 Cell wall thickness (SS) 0.635 (0.25 in.) Holtec International 2003 [DIRS 172633], Table 6.3.3 SS sheathing 0.1905 (0.075 in.) Holtec International 2003 [DIRS 172633], Figure 6.3.3 Boral thickness 0.2057 (0.081 in.) Holtec International 2003 [DIRS 172633], Figure 6.3.3 Al thickness (Clad) 0.0254 (0.010 in.) Holtec International 2003 [DIRS 172633], Figure 6.3.3 Boral width 12.065 (4.75 in.) Holtec International 2003 [DIRS 172633], Figure 6.3.3 Boral clearance gap 0.01397 (0.0055 in.) Holtec International 2003 [DIRS 172633], Figure 6.3.3 Flow channel thickness 0.254 (0.100 in.) Holtec International 2003 [DIRS 172633], Figure 6.3.3 Licensing & Nuclear Safety Calculation Title: Dry Transfer Facility Criticality Safety Calculations Document Identifier: 100-00C-WHS0-00100-000-00C Page 21 of 47 Table 5.1-4 Specifications of the BWR GE 8 x 8 Standard Assembly Parameter Dimension (cm) Reference Rod pitch 1.6256 (0.640 in.) Holtec International 2003 [DIRS 172633], Table 6.2.22 Active fuel length 381.0 (150 in.) Holtec International 2003 [DIRS 172633], Table 6.2.22 Cladding outside diameter 1.2268 (0.483 in.) Holtec International 2003 [DIRS 172633], Table 6.2.22 Cladding inside diameter 1.0796 (0.425 in.) Holtec International 2003 [DIRS 172633], Table 6.2.22 Pellet outside diameter 1.0566 (0.416 in.) Holtec International 2003 [DIRS 172633], Table 6.2.22 Guide tube (Thimble) outside diameter 1.3488 Holtec International 2003 [DIRS 172633], p. 6.D-15 Guide tube (Thimble) thickness 1.5012 Holtec International 2003 [DIRS 172633], p. 6.D-15 Array size 8 x 8 Holtec International 2003 [DIRS 172633], Table 6.2.22 Number of fuel rods 62 Holtec International 2003 [DIRS 172633], Table 6.2.22 Number of guide/instrument tubes a 2 Holtec International 2003 [DIRS 172633], Table 6.2.22 a Locations of guide tubes are shown in Figure 5.1-2. Table 5.1-5 Specifications of the BWR 9 x 9 Assembly Parameter Dimension (cm) Reference Rod pitch 1.45288 (0.572 in.) Holtec International 2003 [DIRS 172633], Table 6.2.27 Active fuel length 381.0 (150 in.) Holtec International 2003 [DIRS 172633], Table 6.2.27 Cladding outside diameter 1.09982 (0.433 in.) Holtec International 2003 [DIRS 172633], Table 6.2.27 Cladding inside diameter 0.96774 (0.381 in.) Holtec International 2003 [DIRS 172633], Table 6.2.27 Pellet outside diameter 0.94996 (0.374 in.) Holtec International 2003 [DIRS 172633], Table 6.2.27 Array size 9 x 9 Holtec International 2003 [DIRS 172633], Table 6.2.27 Number of fuel rods 72 Holtec International 2003 [DIRS 172633], Table 6.2.27 Number of guide/instrument tubes a 1 Holtec International 2003 [DIRS 172633], Table 6.2.27 a This tube replaces the space of 9 fuel rods. 5.1.7 Material Compositions This section presents the material compositions for the PWR and BWR fuel along with the cell basket, DTF staging racks, and the remediation pool. The calculations were performed with either the isotopic compositions given in weight fraction, weight percent (wt%), or atom densities (atoms/barn-cm) depending on the source of the input. Table 5.1-6 displays the relevant materials used for the staging racks and the fuel. Licensing & Nuclear Safety Calculation Title: Dry Transfer Facility Criticality Safety Calculations Document Identifier: 100-00C-WHS0-00100-000-00C Page 22 of 47 Per Assumption 3.2, Polysiloxane fluid was chosen as an alternate moderator material (in the event hydraulic fluid/oil leak from a handling crane). Polysiloxane fluid was modeled with a density of 0.9 g/cm3 (Gelest 2004 [DIRS 169915], p. 11). The chemical formula for this fluid is CH3 CH3 CH3 CH3 CH3 CH3 CH3 CH3 Si O Si Si O 4 Source: (Gelest 2004 [DIRS 169915], p. 11) Table 5.1-6 Material Properties for the Staging Racks and Fuel Material Density (g/cm3) Element Weight Fraction or Weight Percent (wt %) Atom Fraction or Atom Density (atoms/barn-cm) Reference/ Remark H2O (throughout model) 1.0 HO N/A 6.688E-02 c 3.344E-02 c Holtec International 2003 [DIRS 172633], Table 6.3.4 SS (tube, cell wall & reflector) 7.84 Cr Mn Fe Ni N/A 1.761E-02 1.761E-03 5.977E-02 8.239E-03 Holtec International 2003 [DIRS 172633], Table 6.3.4 Al (Boral panel) 2.7 Al 1.0 N/A Holtec International 2003 [DIRS 172633], Table 6.3.4 Boral (0.02 g 10B/cm2) a 2.66 B-10 B-11 C Al 5.443E-02 2.414E-01 8.210E-02 6.222E-01 N/A Holtec International 2003 [DIRS 172633], Table 6.3.4 UO2 – (fuel) 5.00 % enriched 10.522 U-235 U-238 O-16 4.408 83.74 11.85 N/A Holtec International 2002 [DIRS 168494], p. 6.3-9 UO2 – (fuel) 4.10 % enriched b 10.522 U-235 U-238 O-16 3.6140 84.536 11.850 N/A Average value taken from Holtec International 2003 [DIRS 172633], Table 6.3.4 of 4.0 % and 4.2 % enrichment Zr (Cladding) 6.55 Zr 100 N/A Holtec International 2003 [DIRS 172633], Table 6.3.4 Concrete (floor & reflector material) 2.147 Fe, H, C, Na, O, Mg, Al, Si, S, Cl, K, Ca, Ti, Mn 0.5595, 0.3319, 10.5321, 0.1411, 49.9430, 9.4200, 0.7859, 4.2101, 0.2483, 0.0523 d, 0.9445, 22.6318, 0.1488, 0.0512 d N/A Petrie, L. M. et. al. 2000 [DIRS 170550], p. M8.2.28 a The 10B loading of 0.020 g/cm2 is 75 % of the minimum loading 0.0267 g/cm2 (Holtec International 2003 [DIRS 172633], p. 6.2-2) b Used for HI-STAR transportation cask calculations (Section 5.1.8) c Fractions of 0.6667 and 0.3333 used in MCNP input d 0.0520 used in MCNP input instead of 0.0523 and 0.0510 used instead of 0.0512 Licensing & Nuclear Safety Calculation Title: Dry Transfer Facility Criticality Safety Calculations Document Identifier: 100-00C-WHS0-00100-000-00C Page 23 of 47 5.1.8 Category 1 and 2 Event Sequence Evaluations Input Description Evaluations of Category 2 event sequences used the HI-STAR transportation cask (Assumption 3.5). B&W 15x15 fuel assembly at 4.1 wt% enrichment was used as a representative fuel (see Section 5.1.2) and the transportation cask was modeled in MCNP in accordance with the HI-STAR 100 Transport Cask Safety Analysis Report (Holtec International 2003 [DIRS 172633], Section 6.3). The physical dimensions and material compositions of the transportation cask and fuel are documented in BSC 2005 [DIRS 173616], Section 5.1.6. The MPC-24 for pressurized water reactor (PWR) fuel consists of a stainless steel (SS) cask/overpack and an interior 24 PWR assembly basket. Figure 5.1-3 displays the planar cross-section of the MPC-24 calculational model inside the overpack and Figure 5.1-4 presents the axial view. Table 5.1-7 provides the radial and axial dimensions of the MPC-24 and overpack. The material composition of the B&W 15x15 fuel is shown in Table 5.1-6 and the dimensions are shown in Table 5.1-8. NOTE: Figure not to scale. Figure 5.1-3 Radial View of the MPC-24 and Overpack (Source: Holtec International 2003 [DIRS 172633], Figure 6.3.4) Licensing & Nuclear Safety Calculation Title: Dry Transfer Facility Criticality Safety Calculations Document Identifier: 100-00C-WHS0-00100-000-00C Page 24 of 47 NOTE: Figure is not to scale. Figure 5.1-4 Axial View of the MPC-24 PWR Fuel Transportation Cask (Source: Holtec International 2003 [DIRS 172633], Figure 6.3.7) Table 5.1-7 Dimensions of the MPC-24, Overpack, and Cell Geometry Component b Dimension (cm) Reference Stainless steel (SS) overpack thickness 22.86 (9.00 in.) Holtec International 2003 [DIRS 172633], Figure 6.3.4 SS overpack, i.d. 171.1325 (67.375 in.) Holtec International 2003 [DIRS 172633], Figure 6.3.4 Center column 6.985 (2.75 in.) Holtec International 2003 [DIRS 172633], Drawing 3926 (Sheet 2) Flux trap 2.7686 (1.09 in.) Holtec International 2003 [DIRS 172633], Figure 6.3.1 & Table 6.3.3 Boral width – narrow a 15.875 (6.25 in.) Holtec International 2003 [DIRS 172633], Drawing 3926 (Sheet 2) Lower water thickness (below active fuel region) 10.16 (4.0 in.) Holtec International 2003 [DIRS 172633], Figure 6.3.7 Upper water thickness (above active fuel region) 15.24 (6.0 in.) Holtec International 2003 [DIRS 172633], Figure 6.3.7 Bottom overpack SS plate thickness 21.59 (8.5 in.) Holtec International 2003 [DIRS 172633], Figure 6.3.7 Top overpack SS plate thickness 39.37 (15.5 in.) Holtec International 2003 [DIRS 172633], Figure 6.3.7 a The periphery Boral panels have reduced width. See Table 5.1-1 for regular Boral panel width. b See Table 5.1-1 for additional dimensions. Licensing & Nuclear Safety Calculation Title: Dry Transfer Facility Criticality Safety Calculations Document Identifier: 100-00C-WHS0-00100-000-00C Page 25 of 47 Table 5.1-8 Specifications of the PWR B&W 15 x15 Assembly Parameter Dimension (cm) a Reference Rod pitch 1.4428 (0.568 in.) Holtec International 2003 [DIRS 172633], Table 6.2.2 Active fuel length 381.0 (150 in.) Holtec International 2003 [DIRS 172633], Table 6.2.2 Cladding outside diameter 1.0872 (0.428 in.) Holtec International 2003 [DIRS 172633], Table 6.2.2 Cladding thickness 0.05842 (0.023 in.) Holtec International 2003 [DIRS 172633], Table 6.2.2 Pellet outside diameter 0.9504 (0.3742 in.) Holtec International 2003 [DIRS 172633], Table 6.2.2 Guide/instrument tube outside diameter 1.3412 (0.528 in.) Holtec International 2003 [DIRS 172633], Table 6.2.2 Guide/instrument tube inside diameter 1.27 (0.50 in.) Holtec International 2003 [DIRS 172633], Table 6.2.2 Array size 15 x 15 Holtec International 2003 [DIRS 172633], Table 6.2.2 Number of fuel rods 208 Holtec International 2003 [DIRS 172633], Table 6.2.2 Number of guide/instrument tubes b 17 Holtec International 2003 [DIRS 172633], Table 6.2.2 a Some of the values are rounded up slightly. b Locations of guide tubes are shown in Holtec International 2003 [DIRS 172633], Figure 6.3.2. To evaluate the impact of fuel reconfiguration on keff under dry and fully flooded moderator conditions for Category 1 and 2 event sequences, a simplified pin cell model was utilized. The calculations used 5 wt% fresh fuel initial enrichment with Westinghouse 17x17 OFA and GE 7x7 fuel assembly types for PWR and BWR fuel, respectively. Note that the pin cell model was only used to study trends and not to produce absolute keff values. Therefore, a different BWR fuel type was used for the pin cell model since the pin cell model has already been used in previous calculations. Table 5.1-9 lists the PWR and BWR fuel considered and their fuel parameters. Figure 5.1-5 displays a cross-sectional view of the pin cell model used for this analysis. Table 5.1-9 Fuel Types Used for Fuel Reconfiguration Evaluation and their Fuel Parameters PWR fuel Manufacturer Array and Version Pin pitch a (cm) Clad o.d. a (cm) Clad thickness (cm) Pellet diameter a (cm) U-235 wt % enrichment Westinghouse 17x17 OFA 1.25984 0.91440 0.05715 b 0.784352 5.00 BWR fuel GE 7x7 Std 1.87452 1.43002 0.08128 c 1.23698 5.00 a Source: General Atomics 1993 [DIRS 101831], p. 6.2-2 b Source: DOE 1987 [DIRS 132333], p. 2A-351 c Source: DOE 1992 [DIRS 102812], p. 2A-15 and 2A-16 Licensing & Nuclear Safety Calculation Title: Dry Transfer Facility Criticality Safety Calculations Document Identifier: 100-00C-WHS0-00100-000-00C Page 26 of 47 Figure 5.1-5 Cross-section View of Pin Cell Model 5.2 CRITICALITY CALCULATIONS The process and methodology for criticality safety analysis given in the Preclosure Criticality Analysis Process Report (BSC 2004 [DIRS 172058], Section 2.2.7) was implemented in these calculations. This process and methodology requires, as stated earlier in Section 2, consideration of the most reactive fuel assembly, the effective neutron multiplication factor will not exceed 0.95 including all uncertainties and bias, no burnup credit, and no credit for 234U and 236U. Further, all calculations were performed with flooded fuel pin gaps except for dry moderator conditions, 5 wt% fresh fuel, and only 75 % credit for the fixed neutron absorber was applied. 5.2.1 DTF Staging Rack Calculations Water moderator density, which could vary from dry to fully moderated conditions under accident conditions, was considered dry (0 % density water) and fully moderated (100 % density water) in the DTF staging racks. Hydraulic fluid/oil was also evaluated as an alternate moderator since it could possibly leak from a handling crane. Further, additional steel was included inside the steel tube (surrounding the fuel assembly) to be evaluated as a possible reflector. Different spacings between the tubes were considered to allow room for reinforcing steel in the concrete slab. As mentioned in Section 5.1.4.1, the spacings considered are 10 in., 12 in., and 16 in. of concrete between the 9x11 array of tubes. For the 10 in. spacing scenario, a triangular pitch (staggered rows as opposed to straight rows) was also considered in a 10x12 array (see Attachment IV). Per Assumption 3.1, the tubes in the DTF staging rack are modeled in MCNP as full (i.e., every available tube in the array is filled with an assembly). Licensing & Nuclear Safety Calculation Title: Dry Transfer Facility Criticality Safety Calculations Document Identifier: 100-00C-WHS0-00100-000-00C Page 27 of 47 5.2.2 Remediation Pool The PWR and BWR storage rack configurations for the remediation pool were analyzed for various assembly pitches and 5.0 wt % enrichment in order to ensure that the configuration meets design criteria (BSC 2004 [DIRS 171599], Section 4.9.2.2). The various distances/pitches evaluated consist of a nominal scenario and a close spacing scenario. The nominal scenario spacing is based on the maximum traveling distance of the bridge crane (see Section 5.1.5.1) while the close spacing scenario spacing represents assemblies and staging racks more closely situated. The nominal arrangement distances for the staging racks are approximately 11 in. in the x-direction with the two rows of staging racks at approximately 39 in. apart (in the y-direction). The BWR assemblies inside each staging rack have an equal spacing of 3.2 in. (both in the x and y directions) and the PWR assemblies have an equal spacing of 4.0 in. (both in the x and y directions). The close spacing scenario feature arrangement distances for the staging racks at approximately 11 in. in the xdirection with the two rows of staging racks at approximately 3 in. apart (in the y-direction). The BWR assemblies inside each staging rack, for the close spacing scenario, have an equal spacing of 0.4 in. (both in the x and y directions) and the PWR assemblies have an equal spacing of 2.0 in. (both in the x and y directions). It should be mentioned that the distances in the x-direction between the staging racks vary slightly for both the nominal and close spacing scenario but will not impact the results significantly. The reason for this is that the spacings presented above represents the maximum distance between the staging racks and will ensure that the results presented in Section 6 are conservative. As mentioned in Section 5.1.5.1, the staging rack can hold a 3x3 array for PWR fuel or a 4x4 array for BWR fuel. Per Assumption 3.1, the staging racks in the remediation pool are modeled in MCNP as full. The neutron poison (e.g., Boral) loadings evaluated were for 0 g 10B/cm2 and 0.02 g 10B/cm2. In addition, the thickness of the Boral plate was reduced to half its width in the MCNP model for the 0.02 g 10B/cm2 scenario. 5.2.3 Category 1 and 2 Event Sequences The Category 1 and Category 2 event sequences applicable to the DTF and its remediation area have been identified in the Categorization of Event Sequences for License Application document (BSC 2005 [DIRS 171429], Section 7). Category 1 event sequences are presented in Table 5.2-1 and Category 2 event sequences are presented in Table 5.2-2. The supporting calculations for the event sequences are provided in Section 6.3. Table 5.2-1 Category 1 Event Sequences Event No. Event Description Annual Event Frequency Reference GET- 03D Drop of a CSNF assembly 0.5 drops/y GET- 03B Collision involving a CSNF assembly 0.5 collisions/y BSC 2005 [DIRS 171429], Section 7 Licensing & Nuclear Safety Calculation Title: Dry Transfer Facility Criticality Safety Calculations Document Identifier: 100-00C-WHS0-00100-000-00C Page 28 of 47 Table 5.2-2 Category 2 Event Sequences Event No. Event Description Expected Occurrence Reference GET- 01A Drop of a transportation or transfer cask without impact limiters 5.7E-01 GET- 02B Drop of inner lid of a transportation or transfer cask, MSC, or WP into a transportation cask, MSC, or WP 5.7E-01 GET- 03E Drop of a CSNF assembly during transfer combined with HVAC failure 9.7E-02 GET- 03C Collision involving a CSNF assembly during transfer combined with HVAC failure 9.7E-02 GET- 04B Drop or collision of handling equipment onto a CSNF assembly 9.7E-02 GET- 05B Drop of a canister during transfer by crane 5.7E-01 GET- 06B Drop of handling equipment onto a canister 5.7E-01 GET- 07B Drop of an unsealed WP 8.6E-02 GET- 09C Drop of a WP with a known closure defect 2.6E-02 GET- 10C Drop of a CSNF assembly during dry remediation activities 4.9E-01 GET- 10B Collision involving a CSNF assembly during dry remediation activities 4.9E-01 GET- 11B Drop of a canister from a crane during WP remediation or dry remediation activities 5.7E-02 GET- 12B Drop of CSNF assembly transfer/handling equipment onto CSNF assemblies in the DTF remediation pool 4.9E-03 GET- 13A Drop of a loaded transportation cask or MSC in the DTF wet remediation area 1.1E-01 GET- 15C Drop of an empty or full canister for damaged CSNF assemblies (empty or full) onto or against a cask or basket in the DTF remediation pool 4.9E-02 GET- 15B Collision of an empty or full canister for damaged CSNF assemblies (empty or full) onto or against a cask or basket in the DTF remediation pool 4.9E-02 GET- 16C Drop of a CSNF assembly in the DTF remediation pool 4.9E-01 GET- 16B Collision involving a CSNF assembly in the DTF remediation pool 4.9E-01 GET- 17B Drop of a filled SNF basket from the spent fuel transfer machine onto the DTF remediation pool floor 5.4E-02 GET- 18B Drop of handling equipment into or against an opened WP filled with CSNF during dry remediation activities 4.9E-03 GET- 19B Drop of handling equipment into an open WP loaded with DOE canisters or a naval canister during dry remediation activities 3.6E-02 GET- 20B Drop of a severed DPC lid back onto the DPC 1.0E-01 BSC 2005 [DIRS 171429], Section 7 Licensing & Nuclear Safety Calculation Title: Dry Transfer Facility Criticality Safety Calculations Document Identifier: 100-00C-WHS0-00100-000-00C Page 29 of 47 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 DTF STAGING RACK The concrete spacing between the tubes containing the staging racks with PWR fuel was altered in order to ensure criticality safety under the most reactive configuration. The scenarios considered include a fully flooded 9x11 tube array (straight lines) with 10 in., 12 in., and 16 in. spacing, respectively (see Section 5.2.1). Further, the 10 in. concrete spacing was also repeated for a 10x12 fully flooded tube array with a triangular pitch (staggered lines) as opposed to the regular square pitch. None of these calculations include fixed neutron poison. The results from the calculations are presented in Table 6.1-1. It can be seen from the results that each spacing and pitch scenario meets the criticality safety criteria. In addition, keff is not sensitive to the thickness of the stainless steel tube (see first and second entries in Table 6.1-1). Note that the distance between assemblies was maintained the same for the two calculations where the stainless steel thicknesses were varied. Consequently, there is only 9 in. of concrete between the tubes and 1 in. of stainless steel for the calculation with the thicker tube (second entry in Table 6.1-1). Table 6.1-1 PWR Fuel in Various DTF Staging Tube Configurations (w/o Neutron Poison) DTF Staging Tube Configuration keff St. Dev. MCNP files 10 “ concrete spacing square pitch (9x11 array) 0.91008 0.00028 DTF10pwr, DTF10pwr.out 10 “ spacing square pitch (9x11 array) a 0.90996 0.00029 DTF10p10, DTF10p10.out 10 “ concrete spacing triangular pitch (10x12 array) 0.90958 0.00030 DTF10tri, DTF10tri.out 12 “ concrete spacing square pitch (9x11 array) 0.90920 0.00027 DTF12pwr, DTF12pwr.out 16 “ concrete spacing square pitch (9x11 array) 0.90788 0.00030 DTF-pwr1, DTF-pwr1.out a This calculation has a steel tube thickness of 1.0 in. (nominal thickness is 0.5 in.). Table 6.1-2 shows the keff for a fully flooded 9x11 tube array (straight lines) with 16 in. of concrete spacing for PWR and BWR fuel. The cases containing fixed neutron poison (i.e., Boral) has a loading of 0.02 g 10B/cm2 (regular width). Note that the table contains one BWR calculation using a 9x9 fuel assembly. It can be seen that it is only slightly more reactive than the 8x8 fuel assembly used throughout this calculation. Licensing & Nuclear Safety Calculation Title: Dry Transfer Facility Criticality Safety Calculations Document Identifier: 100-00C-WHS0-00100-000-00C Page 30 of 47 Table 6.1-2 PWR and BWR Fuel in DTF Staging Tubes (16 in. Spacing) DTF Staging Rack Configuration keff St. Dev. MCNP files PWR Fuel 16 “ concrete spacing no neutron poison present 0.90788 0.00030 DTF-pwr1, DTF-pwr1.out 16 “ concrete spacing neutron poison present 0.85366 0.00029 DTF-pwr, DTF-pwr.out BWR Fuel 16 “ concrete spacing no neutron poison present 0.61263 0.00027 DTF1-bwr1, DTF-bwr1.out 16 “ concrete spacing no neutron poison present a 0.61927 0.00027 DTF1-9x9, DTF1-9x9.out 16 “ concrete spacing neutron poison present 0.56718 0.00027 DTF-bwr, DTF-bwr.out a Featuring a 9x9 fuel array configuration Various moderator and reflector conditions were evaluated for PWR and BWR fuel in the DTF staging racks. The 9x11 tube array (straight lines) with 16 in. of concrete spacing model was used with no fixed neutron poison present. Table 6.1-3 PWR and BWR Fuel in DTF Staging Tubes with Varied Moderator and Reflector Conditions Reflector/Moderator Material keff St. Dev. MCNP files PWR Fuel Oil (moderator) 0.80101 0.00028 DTFopwr1, DTFopwr1.out Void (moderator) 0.24831 0.00012 DTF-pwr0, DTF-pwr0.out Steel (reflector) a 0.16869 0.00007 DTFsspwr, DTFsspwr.out BWR Fuel Oil (moderator) 0.52850 0.00024 DTFobwr1, DTFobwr1.out Void (moderator) 0.16886 0.00010 DTF-bwr0, DTF-bwr0.out Steel (reflector) a 0.11067 0.00005 DTFssbwr, DTFssbwr.out a The inside of the assembly is dry. In summary, the calculations for PWR and BWR fuel in the DTF staging racks show that for normal operations that the design is criticality safe. Either spacing between the tubes or line arrangement (straight versus staggered) in the array is acceptable. Further, no fixed neutron poison is required in the cell basket. Also, most of these evaluations were performed under fully flooded moderator conditions for conservatism. 6.2 REMEDIATION POOL The PWR and BWR assembly pitches inside the staging racks as well as the distances between the staging racks were varied from a nominal to a close spacing scenario to evaluate the impact on keff. Table 6.2-1 presents the keff values as a function of spacing and neutron poison loading in a fully flooded remediation pool. The nominal and close spacing inside the staging racks and between the Licensing & Nuclear Safety Calculation Title: Dry Transfer Facility Criticality Safety Calculations Document Identifier: 100-00C-WHS0-00100-000-00C Page 31 of 47 staging racks are defined in Section 5.2.2. The results from Table 6.2-1 indicate that no neutron poison is required for PWR or BWR fuel for a nominal arrangement. For staging racks more closely situated (i.e., the close spacing scenario), no neutron poison is required for BWR fuel but neutron poison should be present for PWR fuel for keff to be below USL. Table 6.2-1 PWR and BWR Fuel in Various Remediation Pool Configurations DTF Staging Rack Configuration keff St. Dev. MCNP files PWR Fuel Nominal spacing neutron poison present 0.86578 0.00029 RB-pwr1, RB-pwr1.out Nominal spacing no neutron poison present 0.93087 0.00027 RB-pwr2, RB-pwr2.out Close spacing neutron poison present a 0.90811 0.00028 RBpw2aB, RBpw2aB.out Close spacing neutron poison present b 0.91726 0.00029 RBpw2aB1, RBpw2aB1.out Close spacing no neutron poison present 0.99742 0.00029 RBpwr2a, RBpwr2a.out BWR Fuel Nominal spacing neutron poison present 0.61817 0.00027 RB-bwr1, RB-bwr1.out Nominal spacing no neutron poison present 0.68446 0.00025 RB-bwr2, RB-bwr2.out Close spacing neutron poison present a 0.80855 0.00030 RB-bw2aB, RB-bw2aB.out Close spacing neutron poison present b 0.82919 0.00027 RBbw2aB1, RBbw2aB1.out Close spacing no neutron poison present 0.93280 0.00026 RB-bwr2a, RBbwr2a. out a Boral loading is 0.02 g 10B/cm2. b Boral panel at half width. 6.3 CATEGORY 1 AND 2 EVENT SEQUENCE EVALUATIONS Table 6.3-1 describes the Category 1 event sequences and the applicable criticality safety evaluation performed for each event. Table 6.3-2 provides the criticality evaluation for the Category 2 event sequences. The supporting calculations for the criticality events and event sequences are provided in the subsections. Table 6.3-1 Criticality Safety Evaluation for Category 1 Event Sequences Event No. Event Description a Annual Event Frequency Criticality Safety Evaluation GET- 03D Drop of a CSNF assembly 0.50 drops/y The fuel assemblies are transferred one at a time and a single flooded fuel assembly has a keff of less than 0.9 (see Section 6.3.1). Therefore, criticality safety is not an issue for this event. In addition, a potential reconfiguration of a fuel assembly (inside the fuel basket dimensions) will not pose a criticality concern (see Section 6.3.1). GET- 03B Collision involving a CSNF assembly 0.50 collisions/y See criticality safety evaluation for event number GET-03D. a BSC 2005 [DIRS 171429], Section 7 Licensing & Nuclear Safety Calculation Title: Dry Transfer Facility Criticality Safety Calculations Document Identifier: 100-00C-WHS0-00100-000-00C Page 32 of 47 Table 6.3-2 Criticality Safety Evaluation for Category 2 Event Sequences Event No. Event Description a Expected Occurrences b Criticality Safety Evaluation GET- 01A Drop of a transportation or transfer cask without impact limiters 5.7E-01 The transportation casks are unopened and dry on the inside. Section 6.3.2 shows that a dry on the inside cask does not pose a criticality concern even if the fuel assemblies were rearranged. In the event the drop would cause the transportation cask to breach and rearrange the fuel, the moderator height must be controlled (see Section 6.3.2) to remain below USL. GET- 02B Drop of inner lid of a transportation or transfer cask, MSC, or WP into a transportation cask, MSC, or WP 5.7E-01 See criticality safety evaluation for event number GET-01A for the transportation cask and MSC (Assumption 3.4). The WP design provides criticality safety for this event per Assumption 3.3. GET- 03E Drop of a CSNF assembly during transfer combined with HVAC failure 9.7E-02 See criticality safety evaluation for event number GET-03D (Table 6.3-1). GET- 03C Collision involving a CSNF assembly during transfer combined with HVAC failure 9.7E-02 See criticality safety evaluation for event number GET-03D (Table 6.3-1). GET- 04B Drop or collision of handling equipment onto a CSNF assembly 9.7E-02 See criticality safety evaluation for event number GET-03D (Table 6.3-1). GET- 05B Drop of a canister during transfer by crane 5.7E-01 Section 6.3 of BSC 2005 [DIRS 173284] demonstrates that a rearrangement of fuel for the DOE canisters does not pose a criticality concern. In addition, Section 5.2.3.2 of BSC 2004 [DIRS 170213] shows that losing the skirt, or impact limiter, of the canister due to a drop does not increase the reactivity. GET- 06B Drop of handling equipment onto a canister 5.7E-01 See criticality safety evaluation for event number GET-05B GET- 07B Drop of an unsealed WP 8.6E-02 The WP design provides criticality safety for this event per Assumption 3.3 (prevents moderator from intruding). Note, however, that moderator cannot intrude into an open/unsealed WP for bounding 5.0 wt% fresh fuel scenarios without exceeding the USL (BSC 2004 [DIRS 172553], p. 30). GET- 09C Drop of a WP with a known closure defect 2.6E-02 See criticality safety evaluation for event number GET-07B GET- 10C Drop of a CSNF assembly during dry remediation activities 4.9E-01 While keff will increase if assemblies were reconfigured due to the drop, this event will still not pose a criticality concern as demonstrated by fuel reconfiguration calculations in Section 6.3.1 and Table 6.3-5 (dry activities/scenarios). GET- 10B Collision involving a CSNF assembly during dry remediation activities 4.9E-01 See criticality safety evaluation for event number GET-10C. GET- 11B Drop of a canister from a crane during WP remediation or dry remediation activities 5.7E-02 See criticality safety evaluation for event number GET-05B GET- 12B Drop of CSNF assembly transfer/handling equipment onto CSNF assemblies in the DTF remediation pool 4.9E-03 See criticality safety evaluation for event number GET-03D (Table 6.3-1). Licensing & Nuclear Safety Calculation Title: Dry Transfer Facility Criticality Safety Calculations Document Identifier: 100-00C-WHS0-00100-000-00C Page 33 of 47 Table 6.3-2 (cont.) Criticality Safety Evaluation for Category 2 Event Sequences Event No. Event Description a Expected Occurrences b Criticality Safety Evaluation GET- 13A Drop of a loaded transportation cask or MSC in the DTF wet remediation area 1.1E-01 Regulatory compliance with 10 CFR 50 [DIRS 165855], 71 [DIRS 171308] and 72 [DIRS 173336] provides assurance of criticality safety (MSC per Assumption 3.4). However, drops causing severe fuel rearrangement in open casks should have moderator control (see Section 6.3.4). GET- 15C Drop of an empty or full canister for damaged CSNF assemblies (empty or full) onto or against a cask or basket in the DTF remediation pool 4.9E-02 Damaged fuel impacts radiological releases for confinement analysis. The criticality evaluations already include gap flooding to cover this condition. GET- 15B Collision of an empty or full canister for damaged CSNF assemblies (empty or full) onto or against a cask or basket in the DTF remediation pool 4.9E-02 See criticality safety evaluation for event number GET-15C. GET- 16C Drop of a CSNF assembly in the DTF remediation pool 4.9E-01 The drop could cause reconfiguration of the CSNF assembly. Section 6.3.1 evaluates the keff of a reconfigured, fully flooded, CSNF and it remains safely below 0.9. GET- 16B Collision involving a CSNF assembly in the DTF remediation pool 4.9E-01 See criticality safety evaluation for event number GET-16C GET- 17B Drop of a filled SNF basket from the spent fuel transfer machine onto the DTF remediation pool floor 5.4E-02 The drop could cause reconfiguration of the CSNF assembly. Section 6.3.3 evaluate the keff of a reconfigured, fully flooded, CSNF in a basket and it remains below USL (with fixed neutron poison present). GET- 18B Drop of handling equipment into or against an opened WP filled with CSNF during dry remediation activities 4.9E-03 Studies show that a WP for maximum dry fuel reactivity (i.e., 5.0 wt% fresh fuel) has a keff below 0.5 USL (BSC 2004 [DIRS 172553], p. I-4). In addition, Section 6.3.2 shows that keff is insensitive to fuel reconfiguration (caused by drop/crushing) in dry conditions. GET- 19B Drop of handling equipment into an open WP loaded with DOE canisters or a naval canister during dry remediation activities 3.6E-02 See criticality safety evaluation for event number GET-05B for the DOE canisters. It is the responsibility of the U.S Department of the Navy to ensure criticality safety for the naval SNF canister. GET- 20B Drop of a severed dualpurpose canister (DPC) lid back onto the DPC 1.0E-01 Regulatory compliance with 10 CFR 50 [DIRS 165855], 71 [DIRS 171308] and 72 [DIRS 173336] provides assurance of criticality safety. a BSC 2005 [DIRS 171429], Section 7 b Before permanent closure Licensing & Nuclear Safety Calculation Title: Dry Transfer Facility Criticality Safety Calculations Document Identifier: 100-00C-WHS0-00100-000-00C Page 34 of 47 6.3.1 Single and Multiple Spent Fuel Assembly Drop Category 1 and 2 event sequences include potential drops of single fuel assemblies on a floor, empty rack, empty cask or empty waste package as well as collision with handling equipment, building wall, etc. For this purpose, a single PWR W 17 x 17 OFA (most reactive fuel type) was modeled in MCNP. The fuel assembly was intact and fully flooded. Compression or compaction of the fuel assembly after the drop would result in a less reactive condition as compared to the intact condition, as demonstrated in Section 6.3.2. However, expansion of the fuel assembly after the drop would result in a higher reactive condition, per Section 6.3.2. Table 6.3-3 lists on the first data row the keff of a single PWR fuel assembly (un-poisoned) and it can be seen that the reactivity is considerably below the USL (Section 5.1.3). In addition, calculations of single PWR fuel assemblies (un-poisoned) with the pin pitch reconfigured (i.e., expanded) were also performed. A maximum expansion of the pin pitch inside the fuel basket (pin pitch of 1.36 cm versus the nominal pin pitch of 1.2598 cm) produces a keff below the USL as shown in Table 6.3-3. Consequently, a drop of a single fuel assembly does not pose a criticality safety concern (even with moderator present). An expanded pin pitch (1.45 cm) beyond the fuel basket structure that produces maximum reactivity (see Section 6.3.2) was also modeled. It can be seen from the table that this case does, however, exceed the USL. This indicates that the amount of moderator present when handling fuel assemblies needs to be controlled. The table further shows multiple (un-poisoned) PWR assemblies stacked together. It can be seen that it takes two PWR fuel assembles in order to pose a criticality safety concern. Table 6.3-3 keff of Single and Multiple PWR W 17 x17 Assemblies Enrichment (wt %) keff St. Dev. MCNP files PWR W 17 x17 OFA (1 assembly - regular 1.2598 cm pitch) 5.0 0.89792 0.00028 pwr-asm1, pwr-asm1.out PWR W 17 x17 OFA (1 assembly - expanded 1.36 cm pitch) 5.0 0.94063 0.00027 pwrBasm1, pwrBasm1.out PWR W 17 x17 OFA (1 assembly – maximum 1.45 cm expanded pitch) 5.0 0.99157 0.00026 pwrEasm1, pwrEasm1.out PWR W 17 x17 OFA (2 assemblies with regular 1.2598 cm pitch) 5.0 0.99999 0.00028 pwr-asm2, pwr-asm2.out PWR W 17 x17 OFA (3 assemblies with regular 1.2598 cm pitch) 5.0 1.06238 0.00028 pwr-asm3, pwr-asm3.out Licensing & Nuclear Safety Calculation Title: Dry Transfer Facility Criticality Safety Calculations Document Identifier: 100-00C-WHS0-00100-000-00C Page 35 of 47 6.3.2 Fuel Reconfiguration In the event of a drop or collision, fuel may be reconfigured into a new geometry, i.e., either a pitch reduction or increase. In case of a pitch reduction, this can be modeled by decreasing the fuel pin pitch in the MCNP model. This scenario was evaluated utilizing the MCNP PWR and BWR pin cell models (Section 5.1.8), and featuring reflective boundary conditions to simulate an infinite array of fuel pins. The pin cell model was fully flooded. Also, note that this study was performed to demonstrate the trends of the infinite multiplication factor (kinf) (i.e., not an absolute value) from a drop event, which would result in a pin pitch reduction (due to impact). Note that kinf values are given instead of keff values since this is an infinite system. Table 6.3-4 shows kinf versus pin pitch for PWR and BWR fuel modeled as an array of infinite pin cells (Section 5.2.1). The table indicates that in case of reduction in fuel pin pitch, the reactivity will decrease. Table 6.3-4 kinf of Pin Pitch Reduction of PWR and BWR Fuel Pin Pitch (cm) kinf St. Dev. MCNP files PWR Fuel (W17 x 17 OFA pin cell model, 5.0 wt % enrichment) 1.25984 (regular) 1.50521 0.00018 w17ofa5, w17ofa5.out 1.20 1.48433 0.00024 w17o12, w17o12.out 1.10 1.42603 0.00020 w17o11, w17o11.out 1.00 1.32650 0.00020 w17o10, w17o10.out 0.9145 (smallest) 1.18456 0.00020 w17o09, w17o09.out BWR Fuel (GE 7 x7 Std pin cell model, 4.5 wt % enrichment) 1.87452 (regular) 1.48306 0.00017 bwr7x7, bwr7x7.out 1.80 1.46316 0.00019 bwr18, bwr18.out 1.70 1.42416 0.00019 bwr17, bwr17.out 1.60 1.36543 0.00019 bwr16, bwr16.out 1.50 1.27844 0.00020 bwr15, bwr15.out 1.43003 (smallest) 1.19414 0.00020 bwr14, bwr14.out The same scenario (i.e., pin pitch reduction) was also studied for PWR and BWR fuel in a completely dry environment. As before, this was evaluated utilizing the MCNP PWR and BWR pin cell models (Section 5.1.8), and featuring reflective boundary conditions to simulate an infinite array of fuel pins. Table 6.3-5 shows kinf versus pin pitch for PWR and BWR fuel. It can be seen that kinf is virtually insensitive to pin pitch alterations under dry conditions. Licensing & Nuclear Safety Calculation Title: Dry Transfer Facility Criticality Safety Calculations Document Identifier: 100-00C-WHS0-00100-000-00C Page 36 of 47 Table 6.3-5 kinf of Pin Pitch Reduction of PWR and BWR Fuel (Dry Conditions) Pin Pitch (cm) kinf St. Dev. MCNP files PWR Fuel (W17 x 17 OFA pin cell model, 5.0 wt % enrichment) 1.4 0.77376 0.00017 w17o14D, w17o14D.out 1.3 0.77377 0.00015 w17o13D, w17o13D.out 1.25984 (regular) 0.77368 0.00016 w17ofa5D, w17ofa5D.out 1.20 0.77394 0.00015 w17o12D, w17o12D.out 1.10 0.77362 0.00016 w17o11D, w17o11D.out 1.00 0.77343 0.00019 w17o10D, w17o10D.out 0.9145 (smallest) 0.77368 0.00016 w17o09D, w17o09D.out BWR Fuel (GE 7 x7 Std pin cell model, 5.0 wt % enrichment) 2.1 0.77856 0.00016 bwr21D, bwr21D.out 1.95 0.77890 0.00015 bwr195D, bwr195D.out 1.87452 (regular) 0.77884 0.00012 bwr7x75D, bwr7x75D.out 1.80 0.77873 0.00017 bwr18D, bwr18D.out 1.70 0.77896 0.00016 bwr17D, bwr17D.out 1.60 0.77897 0.00016 bwr16D, bwr16D.out 1.50 0.77901 0.00023 bwr15D, bwr15D.out 1.43003 (smallest) 0.77899 0.00017 bwr14D, bwr14D.out In case of a pitch increase, this can be modeled by increasing the fuel pin pitch in the MCNP model. This scenario was evaluated utilizing the MCNP PWR and BWR pin cell models (Section 5.1.8), and featuring reflective boundary conditions to simulate an infinite array of fuel pins. The pin cell model was fully flooded. Also, note that this study was performed to demonstrate the trends of kinf (i.e., not an absolute value) from a drop event, which would result in a pin pitch increase (due to impact). Table 6.3-6 shows kinf versus pin pitch for PWR and BWR fuel modeled as an array of infinite pin cells (Section 5.2.1). The table indicates that in case of an increase in fuel pin pitch, the reactivity will increase. Figure 6.3-1 graphically displays the results presented in Table 6.3-6 and shows that the peak in kinf occurs at a pin pitch of 1.45 cm for PWR fuel and 2.15 cm for BWR fuel. The data and figure can be found in Excel file pin_pitch.xls. Licensing & Nuclear Safety Calculation Title: Dry Transfer Facility Criticality Safety Calculations Document Identifier: 100-00C-WHS0-00100-000-00C Page 37 of 47 Table 6.3-6 kinf of Pin Pitch Increase for PWR and BWR Fuel (Flooded Conditions) Pin Pitch (cm) kinf St. Dev. MCNP files PWR Fuel (W17 x 17 OFA pin cell model, 5.0 wt % enrichment) 1.20 1.48433 0.00024 w17o12, w17o12.out 1.30 1.51560 0.00025 w17o13, w17o13.out 1.40 1.52855 0.00025 w17o14, w17o14.out 1.45 1.53041 0.00021 w17o145, w17o145.out 1.50 1.52921 0.00021 w17o15, w17o15.out 1.60 1.52066 0.00022 w17o16, w17o16.out 1.70 1.50493 0.00021 w17o17, w17o17.out 1.80 1.48424 0.00020 w17o18, w17o18.out 1.90 1.45839 0.00022 w17o19, w17o19.out BWR Fuel (GE 7 x7 Std pin cell model, 5.0 wt % enrichment) 1.87452 (regular) 1.50116 0.00017 bwr7x75, bwr7x75.out 1.95 1.51473 0.00023 bwr195, bwr195.out 2.1 1.52811 0.00022 bwr21, bwr21.out 2.15 1.52924 0.00022 bwr215, bwr215.out 2.2 1.52851 0.00023 bwr22, bwr22.out 2.3 1.52309 0.00019 bwr23, bwr23.out 2.5 1.50050 0.00022 bwr25, bwr25.out Figure 6.3-1 PWR and BWR Pin Pitch versus kinf for Various Distances 1.45 1.46 1.47 1.48 1.49 1.5 1.51 1.52 1.53 1.54 1 1.2 1.4 1.6 1.8 2 2.2 2.4 2.6 Pin Pitch (cm) Kinf W 17x17 OFA (Pin Cell) GE 7x7 Std (Pin Cell) PWR BWR Licensing & Nuclear Safety Calculation Title: Dry Transfer Facility Criticality Safety Calculations Document Identifier: 100-00C-WHS0-00100-000-00C Page 38 of 47 6.3.3 Fuel Reconfiguration in DTF Staging Rack and Remediation Pool Fuel reconfiguration of PWR and BWR fuel assemblies were evaluated in case of a drop into the staging racks in the DTF or remediation pool. Since the fuel assemblies will be moved one at a time, only one fuel assembly is modeled as reconfigured (entire fuel rod height is at maximum possible pitch inside fuel basket). The DTF staging rack consists of a 9x11 array with 16 in. concrete spacing with both partially and fully flooded tubes and no neutron poison present. The remediation pool features both the close and regular spacing with 0.02 g 10B/cm2 Boral loading (both regular and half width Boral panels). Table 6.3-7 shows that moderator control is necessary around the DTF staging racks for a PWR fuel assembly reconfiguration to remain below the USL. The DTF staging racks can be filled with water to a height of approximately 190 cm (50 % of fuel height) when a reconfigured PWR fuel assembly is present in fully loaded DTF staging racks and be safely below the USL Table 6.3-7 further shows that regular spacing with a neutron poison loading of 0.02 g 10B/cm2 (regular width) is required in the remediation pool in order to be below the USL in the event of a dropped, reconfigured PWR fuel assembly. BWR fuel poses no criticality concern in the DTF staging racks or the remediation pool. Table 6.3-7 keff of DTF and Remediation Pool Staging Rack with Reconfigured Assembly Case keff Standard Deviation MCNP Files Reconfigured Assembly in DTF Staging Rack PWR fuel –fully flooded, no neutron poison 0.95058 0.00027 DTFdpwr5, DTFdpwr5.out PWR fuel – 75 % of fuel height flooded, no neutron poison 0.94931 0.00027 DTFdpwrP, DTFdpwrP.out PWR fuel – 50 % of fuel height flooded, no neutron poison 0.94557 0.00026 DTFdpwP1, DTFdpwP1.out PWR fuel – 25 % of fuel height flooded, no neutron poison 0.93051 0.00027 DTFdpwP2, DTFdpwP2.out BWR fuel - fully flooded, no neutron poison 0.63474 0.00027 DTFdbwr2, DTFdbwr2.out Reconfigured Assembly in the Remediation Pool PWR fuel – close spacing, half width Boral panel 0.97555 0.00032 RBp2aB1d,RBp2aB1d.out PWR fuel – close spacing, 0.02 g 10B/cm2 0.97144 0.00028 RBp2aB2d,RBp2aB2d.out PWR fuel – nominal spacing, half width Boral panel 0.94508 0.00027 RBp2rB1d,RBp2rB1d.out PWR fuel – nominal spacing, 0.02 g 10B/cm2 0.94389 0.00029 RBp2rB2d,RBp2rB2d.out BWR fuel - close spacing, half width Boral panel 0.85043 0.00028 RBb2aB1d,RBb2aB1d.out Licensing & Nuclear Safety Calculation Title: Dry Transfer Facility Criticality Safety Calculations Document Identifier: 100-00C-WHS0-00100-000-00C Page 39 of 47 6.3.4 Fuel Reconfiguration in HI-STAR Transportation Cask Fuel reconfiguration was studied for PWR fuel in the HI-STAR transportation cask to ensure criticality safety in the event this would occur due to a drop. Both dry, partially and fully flooded cask inside conditions were evaluated. Since the transportation casks will remain unopened, the dry conditions will represent undamaged transportation cask containment. In the event a drop punctures the transportation cask containment, fully flooded conditions are conservatively considered. Table 6.3-8 shows the keff values for uniformly rearranged PWR fuel (largest possible uniform pin pitch inside the fuel basket) at various water heights in the vertically orientated transportation cask. It can also be seen that fuel rearrangement does not impact keff significantly during dry conditions (keff is 0.36610 ± 0.00010 for a dry, nominal pitch configuration (BSC 2005 [DIRS 173616], Table 6.1-1)), but an increase in pin pitch during fully flooded conditions causes keff to exceed the USL under the bounding fresh fuel scenario. The transportation casks can be filled up to about 25 % of the fuel height (about 95 cm above the bottom of the cask) with rearranged fuel (entire fuel rod height) and not be in violation of the USL. Table 6.3-8 also contains a fully flooded transportation cask with only 50 cm of the bottom portion of the active fuel region reconfigured. Spacers, preventing the fuel from bowing out, are located near the end, as well as approximately 28 cm and 50 cm from the ends, of the fuel assembly (DOE 1987 [DIRS 132333], p. 2A-353). In the event the drop or slap down is not of significant height/force, only a portion of the fuel is more likely to be reconfigured than the entire fuel rod. It can be seen that with only 50 cm of the fuel reconfigured, keff remains below the USL for fully flooded conditions. In the event that hydraulic fluid/oil is present in the facility, due to a leak form the handling crane, a calculation was also performed for a cask containing rearranged fuel (from a drop) partially filled with oil. It can be seen that water is the bounding moderator compared to oil and will not violate the USL. Table 6.3-8 Comparison of Various Partial Water Flooding Scenarios Case a keff Standard Deviation MCNP Files Reconfigured PWR fuel in HI-STAR cask 0 % flooding 0.36537 0.00010 star15bD, star15bD.out 25 % flooding 0.93725 0.00030 sta15P25, sta15P25.out 50 % flooding 0.95444 0.00027 sta15P50, sta15P50.out 75 % flooding 0.95692 0.00026 sta15P75, sta15P75.out 75 % flooding (oil) b 0.86759 0.00028 sta15O75, sta15O75.out 100 % flooding 0.95935 0.00029 star15b, star15b.out 100 % flooding (50 cm reconfigured) c 0.93668 0.00027 sta15-50, sta15-50.out a The percent value listed refers to the portion of the fuel height that is flooded b Oil is used instead of water as a moderator c Only 50 cm of the fuel height is expanded/reconfigured Licensing & Nuclear Safety Calculation Title: Dry Transfer Facility Criticality Safety Calculations Document Identifier: 100-00C-WHS0-00100-000-00C Page 40 of 47 6.4 CONCLUSIONS AND RECOMMENDATIONS The processes for the DTF and its remediation area have been evaluated for criticality safety for normal operations, Category 1 and 2 event sequences. The results presented in this document lead to the following conclusions and recommendations: • Fixed neutron absorber is required for PWR fuel in the storage racks in the remediation pool to ensure adequate safety margin for both normal operations and Category 1 and 2 event sequences. The minimum spacing between each staging rack is 11 in. in the x-direction and 39 in. between the two rows (y-direction). The spacing between assemblies inside the staging racks is 4.0 in. for PWR fuel (3x3 array) and 3.2 in. for BWR fuel (4x4 array). These spacings require a neutron poison (i.e., Boral) loading of 0.02 g 10B/cm2 (for PWR fuel). Additional work should be performed if smaller spacing is desired. • Concrete spacings between the DTF staging tubes of 10 in., 12 in., and 16 in. were evaluated with array sizes of 9x11 and 10x12 (square and triangular pitches). Either configuration is criticality safe under normal operations for both dry and fully flooded conditions. Further, no neutron poison is needed for PWR or BWR fuel assemblies. • Category 1 and 2 event sequences potentially occurring in these facilities do not compromise criticality safety. Moderator control is required in the DTF staging rack area in case a drop of a PWR fuel assembly into the DTF staging racks causes fuel reconfiguration. The acceptable water height that can cover the PWR active fuel region in the staging tubes to ensure a keff below the USL is approximately 190 cm. Further, a moderator height of 95 cm above the bottom of a vertically oriented transportation cask cannot be exceeded in order to maintain a keff below the USL. This is in case a drop of a transportation cask causes a breach that can flood the inside of the cask while also reconfiguring the fuel. In summary, normal operations in the DTF and its remediation area prove to be criticality safe for CSNF. It should be mentioned that the results for the Monitored Geologic Repository (MGR) Site specific Cask (MSC) calculations are limited to the specific design chosen (see Assumption 3.4). A more current design will be included in the next revision of the criticality calculations for the Aging Facility. During Category 2 conditions, a controlled moderator (i.e., water) height of less than 190 cm above the bottom of the active PWR fuel region in the DTF staging tube is required in order for keff to not exceed the USL. Further, a maximum moderator height of 95 cm above the bottom of a vertically oriented transportation cask cannot be exceeded in order to maintain a keff below the USL. It should also be mentioned that although it is not explicitly shown in this calculation, moderator control is also needed in areas with open WPs to ensure criticality safety during Category 2 event sequences. Attachment III displays the areas requiring moderator control in the DTF. 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[DIRS 170345] BSC (Bechtel SAIC Company) 2004. Dry Transfer Facility #1 /Remediation Facility General Arrangement Ground Floor Plan C. 110-P10-WHS0-00110-000-00B. Las Vegas, Nevada: Bechtel SAIC Company. ACC: ENG.20040623.0006. [DIRS 170346] BSC (Bechtel SAIC Company) 2004. Dry Transfer Facility #1 /Remediation Facility General Arrangement Ground Floor Plan D. 110-P10-WHS0-00111-000-00B. Las Vegas, Nevada: Bechtel SAIC Company. ACC: ENG.20040623.0007. [DIRS 170347] Licensing & Nuclear Safety Calculation Title: Dry Transfer Facility Criticality Safety Calculations Document Identifier: 100-00C-WHS0-00100-000-00C Page 42 of 47 BSC (Bechtel SAIC Company) 2004. Dry Transfer Facility #1 /Remediation Facility General Arrangement Operating Floor Plan A. 110-P10-WHS0-00112-000-00B. Las Vegas, Nevada: Bechtel SAIC Company. ACC: ENG.20040623.0008. [DIRS 170348] BSC (Bechtel SAIC Company) 2004. Dry Transfer Facility #1 /Remediation Facility General Arrangement Operating Floor Plan B. 110-P10-WHS0-00113-000-00B. Las Vegas, Nevada: Bechtel SAIC Company. ACC: ENG.20040623.0009. [DIRS 170349] BSC (Bechtel SAIC Company) 2004. Dry Transfer Facility #1 /Remediation Facility General Arrangement Operating Floor Plan C. 110-P10-WHS0-00114-000-00B. Las Vegas, Nevada: Bechtel SAIC Company. ACC: ENG.20040623.0010. [DIRS 170350] BSC (Bechtel SAIC Company) 2004. Dry Transfer Facility #1 /Remediation Facility General Arrangement Operating Floor Plan D. 110-P10-WHS0-00115-000-00B. Las Vegas, Nevada: Bechtel SAIC Company. ACC: ENG.20040623.0011. [DIRS 170351] BSC (Bechtel SAIC Company) 2004. Dry Transfer Facility #1 /Remediation Facility General Arrangement Upper Floor Plan A. 110-P10-WHS0-00116-000-00A. Las Vegas, Nevada: Bechtel SAIC Company. ACC: ENG.20040503.0014. [DIRS 169076] BSC (Bechtel SAIC Company) 2004. Dry Transfer Facility #1 /Remediation Facility General Arrangement Upper Floor Plan B. 110-P10-WHS0-00117-000-00B. Las Vegas, Nevada: Bechtel SAIC Company. ACC: ENG.20040623.0012. [DIRS 170380] BSC (Bechtel SAIC Company) 2004. Dry Transfer Facility #1 /Remediation Facility General Arrangement Upper Floor Plan C. 110-P10-WHS0-00118-000-00A. Las Vegas, Nevada: Bechtel SAIC Company. ACC: ENG.20040503.0016. [DIRS 169078] BSC (Bechtel SAIC Company) 2004. Dry Transfer Facility #1 /Remediation Facility General Arrangement Upper Floor Plan D. 110-P10-WHS0-00119-000-00A. Las Vegas, Nevada: Bechtel SAIC Company. ACC: ENG.20040503.0017. [DIRS 169079] BSC (Bechtel SAIC Company) 2004. Dry Transfer Facility #1 /Remediation Facility General Arrangement Pool Floor Plans. 110-P10-WHS0-00124-000-00A. Las Vegas, Nevada: Bechtel SAIC Company. ACC: ENG.20040503.0023. [DIRS 169084] BSC (Bechtel SAIC Company) 2004. Dry Transfer Facility #1 /Remediation Facility General Arrangement Section A. 110-P10-WHS0-00128-000-00B. Las Vegas, Nevada: Bechtel SAIC Company. ACC: ENG.20040623.0013. [DIRS 170381] BSC (Bechtel SAIC Company) 2004. Dry Transfer Facility #1 /Remediation Facility General Arrangement Section B. 110-P10-WHS0-00129-000-00B. Las Vegas, Nevada: Bechtel SAIC Company. ACC: ENG.20040623.0014. [DIRS 170382] BSC (Bechtel SAIC Company) 2004. Dry Transfer Facility #1 /Remediation Facility General Arrangement Section C. 110-P10-WHS0-00130-000-00B. Las Vegas, Nevada: Bechtel SAIC Company. ACC: ENG.20040623.0015. [DIRS 170383] Licensing & Nuclear Safety Calculation Title: Dry Transfer Facility Criticality Safety Calculations Document Identifier: 100-00C-WHS0-00100-000-00C Page 43 of 47 BSC (Bechtel SAIC Company) 2004. Dry Transfer Facility #1 /Remediation Facility General Arrangement Section D. 110-P10-WHS0-00131-000-00B. Las Vegas, Nevada: Bechtel SAIC Company. ACC: ENG.20040623.0016. [DIRS 170384] BSC (Bechtel SAIC Company) 2004. Dry Transfer Facility #1 /Remediation Facility General Arrangement Section E. 110-P10-WHS0-00132-000-00B. Las Vegas, Nevada: Bechtel SAIC Company. ACC: ENG.20040623.0017. [DIRS 170385] BSC (Bechtel SAIC Company) 2004. Dry Transfer Facility #1 /Remediation Facility General Arrangement Section F. 110-P10-WHS0-00133-000-00B. Las Vegas, Nevada: Bechtel SAIC Company. ACC: ENG.20040623.0018. [DIRS 170386] BSC (Bechtel SAIC Company) 2004. Dry Transfer Facility #1 /Remediation Facility General Arrangement Section G. 110-P10-WHS0-00134-000-00B. Las Vegas, Nevada: Bechtel SAIC Company. ACC: ENG.20040623.0019. [DIRS 170387] BSC (Bechtel SAIC Company) 2004. Dry Transfer Facility #1 /Remediation Facility General Arrangement Section J. 110-P10-WHS0-00136-000-00B. Las Vegas, Nevada: Bechtel SAIC Company. ACC: ENG.20040623.0021. [DIRS 170391] BSC (Bechtel SAIC Company) 2004. Dry Transfer Facility #1 /Remediation Facility General Arrangement Section K. 110-P10-WHS0-00137-000-00B. Las Vegas, Nevada: Bechtel SAIC Company. ACC: ENG.20040623.0022. [DIRS 170392] BSC (Bechtel SAIC Company) 2004. Dry Transfer Facility #1 /Remediation Facility General Arrangement Section L. 110-P10-WHS0-00138-000-00B. Las Vegas, Nevada: Bechtel SAIC Company. ACC: ENG.20040623.0023. [DIRS 170393] BSC (Bechtel SAIC Company) 2004. Dry Transfer Facility #1 /Remediation Facility General Arrangement Section M. 110-P10-WHS0-00139-000-00B. Las Vegas, Nevada: Bechtel SAIC Company. ACC: ENG.20040623.0024. [DIRS 170394] BSC (Bechtel SAIC Company) 2004. Dry Transfer Facility #1 /Remediation Facility General Arrangement Section N. 110-P10-WHS0-00140-000-00B. Las Vegas, Nevada: Bechtel SAIC Company. ACC: ENG.20040623.0025. [DIRS 170395] BSC (Bechtel SAIC Company) 2004. Dry Transfer Facility #1 /Remediation Facility General Arrangement Section P. 110-P10-WHS0-00141-000-00B. Las Vegas, Nevada: Bechtel SAIC Company. ACC: ENG.20040623.0026. [DIRS 170396] BSC (Bechtel SAIC Company) 2004. Dry Transfer Facility #1 /Remediation Facility Pool Floor Overall Plan TOC El (–)55'–0". 110-DB0-WHS0-00101-000-00B. Las Vegas, Nevada: Bechtel SAIC Company. ACC: ENG.20040831.0004. [DIRS 170728] Licensing & Nuclear Safety Calculation Title: Dry Transfer Facility Criticality Safety Calculations Document Identifier: 100-00C-WHS0-00100-000-00C Page 44 of 47 BSC (Bechtel SAIC Company) 2004. Fuel Handling Facility Criticality Safety Calculations. 210- 00C-FH00-00400-000-00A. Las Vegas, Nevada: Bechtel SAIC Company. ACC: ENG.20040630.0012. [DIRS 170213] BSC (Bechtel SAIC Company) 2004. Preclosure Criticality Analysis Process Report. TDR-EBSNU- 000004 REV 04. Las Vegas, Nevada: Bechtel SAIC Company. ACC: DOC.20041025.0005. [DIRS 172058] BSC (Bechtel SAIC Company) 2004. Project Design Criteria Document. 000-3DR-MGR0-00100- 000-003. Las Vegas, Nevada: Bechtel SAIC Company. ACC: ENG.20041124.0001. [DIRS 171599] BSC (Bechtel SAIC Company) 2005. Canister Handling Facility Criticality Safety Calculations. 190-00C-CH00-00100-000-00B. Las Vegas, Nevada: Bechtel SAIC Company. ACC: ENG.20050411.0001. [DIRS 173284] BSC (Bechtel SAIC Company) 2005. Categorization of Event Sequences for License Application. 000-00C-MGR0-00800-000-00B. Las Vegas, Nevada: Bechtel SAIC Company. ACC: ENG.20050414.0011. [DIRS 171429] BSC (Bechtel SAIC Company) 2005. Dry Transfer Facility Description Document. 110-3YDCD00- 00100-000-002. Las Vegas, Nevada: Bechtel SAIC Company. ACC: ENG.20050512.0001. [DIRS 173737] BSC (Bechtel SAIC Company) 2005. Q-List. 000-30R-MGR0-00500-000-001. Las Vegas, Nevada: Bechtel SAIC Company. ACC: ENG.20050217.0010. [DIRS 171190] BSC (Bechtel SAIC Company) 2005. Transportation Cask Receipt/Return Facility Criticality Safety Evaluations. 140-00C-HCR0-00300-000-00B. Las Vegas, Nevada: Bechtel SAIC Company. ACC: ENG.20050427.0006. [DIRS 173616] BSC (Bechtel SAIC Company) 2005. Transportation Cask Receipt/Return Facility Description Document. 140-3YD-CC00-00100-000-003. Las Vegas, Nevada: Bechtel SAIC Company. ACC: ENG.20050315.0004. [DIRS 171702] 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.20031222.0006. [DIRS 166275] Cogema. 2003. Remediation System SNF Staging Racks Mechanical Equipment Envelope. YMPC0115- 0738, Rev. A. Las Vegas, Nevada: Cogema. ACC: ENG.20040123.0122. [DIRS 167153] Cogema. 2004. Remediation System Fuel Handling Machine Mechanical Equipment Envelope. YMP-C0115-0730, Rev. D. Las Vegas, Nevada: Cogema. ACC: ENG.20041216.0017. [DIRS 173297] CRWMS M&O 1998. Software Code: MCNP. V4B2LV. HP, HPUX 9.07 and 10.20; PC, Windows 95; Sun, Solaris 2.6. 30033 V4B2LV. [DIRS 154060] Licensing & Nuclear Safety Calculation Title: Dry Transfer Facility Criticality Safety Calculations Document Identifier: 100-00C-WHS0-00100-000-00C Page 45 of 47 CRWMS M&O 1998. Software Qualification Report for 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. [DIRS 102836] Curry, P.M. 2004. Project Functional and Operational Requirements. TDR-MGR-ME-000003 REV 02. Las Vegas, Nevada: Bechtel SAIC Company. ACC: DOC.20040714.0003. [DIRS 170557] DOE (U.S. Department of Energy) 1987. Appendix 2A. Physical Descriptions of LWR Fuel Assemblies. Volume 3 of 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. ACC: HQX.19880405.0024. [DIRS 132333] 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) 2004. Quality Assurance Requirements and Description. DOE/RW-0333P, Rev. 16. Washington, D.C.: U.S. Department of Energy, Office of Civilian Radioactive Waste Management. ACC: DOC.20040907.0002. [DIRS 171539] Gelest. 2004. Gelest Silicone Fluids: Stable, Inert Media. Morrisville, Pennsylvania: Gelest. TIC: 256122. [DIRS 169915] General Atomics. 1993. GA–4 Legal Weight Truck From-Reactor Spent Fuel Shipping Cask, Final Design Report. 910353/0. San Diego, California: General Atomics. ACC: NNA.19940513.0215. [DIRS 101831] General Atomics. 1998. GA-4 Legal Weight Truck Spent Fuel Shipping Cask, Safety Analysis Report for Packaging (SARP). 910469/H. Two volumes. San Diego, California: General Atomics. TIC: 245137. [DIRS 103042] HOLTEC International. 2002. Final Safety Analysis Report for the Holtec International Storage and Transfer Operation Reinforced Module Cask System (Hi-Storm 100 Cask System). HOLTEC Report HI-2002444. Two volumes. NRC Docket No. 72-1014. Marlton, New Jersey: HOLTEC International. TIC: 255899. [DIRS 168494] Holtec International 2003. Storage, Transport, and Repository Cask Systems, (Hi-Star Cask System) Safety Analysis Report, 10 CFR 71, Docket 71-9261. HI-951251, Rev. 10. [Marlton, New Jersey]: Holtec International. ACC: MOL.20050119.0271. [DIRS 172633] LP-SI.11Q-BSC, Rev. 0, ICN 1. Software Management. Washington, D.C.: U.S. Department of Energy, Office of Civilian Radioactive Waste Management. ACC: DOC.20041005.0008. [DIRS 171923] Licensing & Nuclear Safety Calculation Title: Dry Transfer Facility Criticality Safety Calculations Document Identifier: 100-00C-WHS0-00100-000-00C Page 46 of 47 LP-SIII.10Q-BSC, Rev. 0, ICN 0. Models. Washington, D.C.: U.S. Department of Energy, Office of Civilian Radioactive Waste Management. ACC: DOC.20050203.0009. [DIRS 172972] NAC International. 2002. Safety Analysis Report for the UMS® Universal Transport Cask. UMST- 02E. EA790-SAR-001. Docket No. 71-9270. Volume 1. Norcross, Georgia: NAC International. TIC: 254855. [DIRS 164612] 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] Petrie, L.M.; Fox, P.B.; and Lucius, K. 2000. "Standard Composition Library." Volume 3, Section M8 of SCALE: A Modular Code System for Performing Standardized Computer Analyses for Licensing Evaluation. NUREG/CR-0200, Rev. 6. Washington, D.C.: U.S. Nuclear Regulatory Commission. ACC: MOL.20040323.0056. [DIRS 170550] Sanders, C.E. and Wagner, J.C. 2002. Study of the Effect of Integral Burnable Absorbers for PWR Burnup Credit. NUREG/CR-6760. Washington, D.C.: U.S. Nuclear Regulatory Commission. TIC: 254260. [DIRS 163141] Sisley, S.E. 2002. "Transportation License Application for the FuelSolutions™ System (TAC No. L23311), Submittal of the FuelSolution™ Transportation SARs, Revision 3." Letter from S.E. Sisley (BNFL) to the NRC, April 11, 2002, BFS/NRC 02-011, with enclosures. TIC: 255245. [DIRS 171545] Thompson, T.C. 2003. "NAC-STC Safety Analysis Report, Revision 14." Letter from T.C. Thompson (NAC International) to the NRC, June 18, 2003, with enclosures. TIC: 255255. [DIRS 169362] Transnuclear. 2001. TN-68 Transport Packaging Safety Analysis Report. NRC Docket 71-9293. Hawthorne, New York: Transnuclear. ACC: MOL.20040517.0697. [DIRS 167988] Transnuclear 2002. NUHOMS-MP197 Transport Packaging Safety Analysis Report, Revision 3. Hawthorne, New York: Transnuclear. TIC: 254016. [DIRS 160736] Transnuclear West 2002. NUHOMS-MP-187 Transportation Package Safety Analysis Report, Revision 16. NUH-05-151. Fremont, California: Transnuclear West. TIC: 254012. [DIRS 161510] Wagner, J.C. and Parks, C.V. 2000. A Critical Review of the Practice of Equating the Reactivity of Spent Fuel to Fresh Fuel in Burnup Credit Criticality Safety Analyses for PWR Spent Fuel Pool Storage. NUREG/CR-6683. Washington, D.C.: U.S. Nuclear Regulatory Commission. TIC: 254257. [DIRS 163140] Licensing & Nuclear Safety Calculation Title: Dry Transfer Facility Criticality Safety Calculations Document Identifier: 100-00C-WHS0-00100-000-00C Page 47 of 47 8. ATTACHMENTS This calculation document includes four attachments: ATTACHMENT I List of Computer Files ATTACHMENT II One Compact Disc ATTACHMENT III General Arrangement Drawings of the Dry Transfer Facility and its Remediation Area (secondary references on these drawings are not applicable to this calculation) ATTACHMENT IV Sketch of DTF Staging Tube Layouts Licensing & Nuclear Safety Calculation Title: Dry Transfer Facility Criticality Safety Calculations Document Identifier: 100-00C-WHS0-00100-000-00C Page I-1 of I-5 ATTACHMENT I LISTING OF COMPUTER FILES All MCNP input and output files documented in this calculation were stored on an electronic medium (compact disc) as Attachment II. Also, the Microsoft® Excel spreadsheet used to illustrate results are included on the compact disc. Date Time Size File 04/18/2005 01:45p 18,432 pin_pitch.xls 04/13/2005 01:59p 33,927 HI-STAR/sta15-50 04/13/2005 01:59p 604,193 HI-STAR/sta15-50.out 04/13/2005 01:59p 18,396 HI-STAR/star15b 04/13/2005 01:59p 521,411 HI-STAR/star15b.out 04/13/2005 01:59p 34,425 HI-STAR/PARTIAL/sta15O75 04/13/2005 01:59p 605,444 HI-STAR/PARTIAL/sta15O75.out 04/13/2005 01:59p 34,383 HI-STAR/PARTIAL/sta15P25 04/13/2005 01:59p 605,667 HI-STAR/PARTIAL/sta15P25.out 04/13/2005 01:59p 34,384 HI-STAR/PARTIAL/sta15P50 04/13/2005 01:59p 605,667 HI-STAR/PARTIAL/sta15P50.out 04/13/2005 01:59p 34,384 HI-STAR/PARTIAL/sta15P75 04/13/2005 01:59p 605,667 HI-STAR/PARTIAL/sta15P75.out 04/13/2005 01:59p 18,874 HI-STAR/PARTIAL/star15bD 04/13/2005 01:59p 521,710 HI-STAR/PARTIAL/star15bD.out 04/13/2005 02:00p 18,496 DTF-RACK/DTF-pwr 04/13/2005 02:00p 517,094 DTF-RACK/DTF-pwr.out 04/13/2005 02:00p 18,148 DTF-RACK/DTF-pwr0 04/13/2005 02:00p 513,890 DTF-RACK/DTF-pwr0.out 04/13/2005 02:00p 18,496 DTF-RACK/DTF-pwr1 04/13/2005 02:00p 516,856 DTF-RACK/DTF-pwr1.out 04/18/2005 07:09p 23,684 DTF-RACK/DTFdpwr5 04/18/2005 07:09p 537,518 DTF-RACK/ DTFdpwr5.out 04/13/2005 02:00p 18,295 DTF-RACK/DTFopwr1 04/13/2005 02:00p 514,866 DTF-RACK/DTFopwr1.out 04/13/2005 02:00p 18,153 DTF-RACK/DTFsspwr 04/13/2005 02:00p 513,897 DTF-RACK/DTFsspwr.out 05/02/2005 10:58a 40,938 DTF-RACK/DTFdpwP1 05/02/2005 10:58a 627,718 DTF-RACK/DTFdpwP1.out 05/02/2005 10:58a 40,937 DTF-RACK/DTFdpwP2 05/02/2005 10:58a 626,234 DTF-RACK/DTFdpwP2.out 05/02/2005 10:59a 40,938 DTF-RACK/DTFdpwrP 05/02/2005 10:59a 627,767 DTF-RACK/DTFdpwrP.out 05/04/2005 04:21p 18,496 DTF-RACK/10IN/DTF10pwr 05/04/2005 04:21p 514,494 DTF-RACK/10IN/DTF10pwr.out 05/04/2005 04:21p 20,948 DTF-RACK/10IN/DTF10tri 05/04/2005 04:21p 526,794 DTF-RACK/10IN/DTF10tri.out 05/12/2005 11:33a 18,496 DTF-RACK/10IN/DTF10p10 Licensing & Nuclear Safety Calculation Title: Dry Transfer Facility Criticality Safety Calculations Document Identifier: 100-00C-WHS0-00100-000-00C Page I-2 of I-5 Date Time Size File 05/12/2005 11:34a 515,876 DTF-RACK/10IN/DTF10p10.out 04/13/2005 02:01p 18,496 DTF-RACK/12IN/DTF12pwr 04/13/2005 02:01p 515,974 DTF-RACK/12IN/DTF12pwr.out 05/02/2005 11:00a 17,705 DTF-RACK/BWR/DTF-bwr 05/02/2005 11:00a 514,282 DTF-RACK/BWR/DTF-bwr.out 05/02/2005 11:00a 17,625 DTF-RACK/BWR/DTF-bwr0 05/02/2005 11:00a 512,698 DTF-RACK/BWR/DTF-bwr0.out 05/02/2005 11:00a 17,705 DTF-RACK/BWR/DTF-bwr1 05/02/2005 11:00a 515,426 DTF-RACK/BWR/DTF-bwr1.out 05/02/2005 11:02a 17,756 DTF-RACK/BWR/DTF1-9x9 05/02/2005 11:03a 515,374 DTF-RACK/BWR/DTF1-9x9.out 05/02/2005 11:00a 21,525 DTF-RACK/BWR/DTFdbwr2 05/02/2005 11:00a 531,749 DTF-RACK/BWR/DTFdbwr2.out 05/02/2005 11:00a 17,727 DTF-RACK/BWR/DTFobwr1 05/02/2005 11:00a 515,039 DTF-RACK/BWR/DTFobwr1.out 05/02/2005 11:00a 17,631 DTF-RACK/BWR/DTFssbwr 05/02/2005 11:00a 512,904 DTF-RACK/BWR/DTFssbwr.out 04/13/2005 02:03p 12,195 RB-POOL/RB-pwr1 04/13/2005 02:03p 489,675 RB-POOL/RB-pwr1.out 04/13/2005 02:03p 12,207 RB-POOL/RB-pwr2 04/13/2005 02:03p 489,352 RB-POOL/RB-pwr2.out 05/02/2005 10:54a 17,313 RB-POOL/RBp2aB2d 05/02/2005 10:54a 501,075 RB-POOL/RBp2aB2d.out 04/18/2005 12:11p 17,396 RB-POOL/RBp2aB1d 04/18/2005 12:11p 472,077 RB-POOL/RBp2aB1d.out 04/18/2005 12:11p 17,430 RB-POOL/RBp2rB1d 04/18/2005 12:11p 514,601 RB-POOL/RBp2rB1d.out 04/18/2005 12:11p 17,347 RB-POOL/RBp2rB2d 04/18/2005 12:11p 515,657 RB-POOL/RBp2rB2d.out 04/13/2005 02:02p 12,180 RB-POOL/RBpw2aB 04/13/2005 02:02p 476,513 RB-POOL/RBpw2aB.out 04/13/2005 02:02p 11,756 RB-POOL/RBpw2aB1 04/13/2005 02:02p 475,139 RB-POOL/ RBpw2aB1.out 04/13/2005 02:03p 12,191 RB-POOL/RBpwr2a 04/13/2005 02:03p 475,896 RB-POOL/RBpwr2a.out 05/02/2005 10:50a 9,680 RB-POOL/BWR/RB-bw2aB 05/02/2005 10:50a 481,612 RB-POOL/BWR/RB-bw2aB.out 05/02/2005 10:50a 10,699 RB-POOL/BWR/RB-bwr1 05/02/2005 10:50a 483,607 RB-POOL/BWR/RB-bwr1.out 05/02/2005 10:50a 10,699 RB-POOL/BWR/RB-bwr2 05/02/2005 10:50a 469,934 RB-POOL/BWR/RB-bwr2.out 05/02/2005 10:50a 9,680 RB-POOL/BWR/RB-bwr2a 05/02/2005 10:50a 481,191 RB-POOL/BWR/RB-bwr2a.out 05/02/2005 10:50a 14,775 RB-POOL/BWR/RBb2aB1d 05/02/2005 10:50a 504,687 RB-POOL/BWR/RBb2aB1d.out Licensing & Nuclear Safety Calculation Title: Dry Transfer Facility Criticality Safety Calculations Document Identifier: 100-00C-WHS0-00100-000-00C Page I-3 of I-5 Date Time Size File 05/02/2005 10:50a 9,757 RB-POOL/BWR/RBbw2aB1 05/02/2005 10:50a 481,612 RB-POOL/BWR/RBbw2aB1.out 05/02/2005 11:06a 4,999 ASM/pwr-asm1 05/02/2005 11:06a 430,292 ASM/pwr-asm1.out 05/02/2005 11:06a 6,258 ASM/pwr-asm2 05/02/2005 11:06a 440,013 ASM/pwr-asm2.out 05/02/2005 11:06a 6,360 ASM/pwr-asm3 05/02/2005 11:06a 440,537 ASM/pwr-asm3.out 05/02/2005 11:06a 4,995 ASM/pwrBasm1 05/02/2005 11:06a 430,292 ASM/pwrBasm1.out 05/02/2005 11:07a 5,009 ASM/pwrEasm1 05/02/2005 11:07a 430,292 ASM/pwrEasm1.out 02/26/2004 10:05a 2,969 PinCell/bwr195 02/26/2004 10:05a 279,566 PinCell/bwr195.out 02/26/2004 10:05a 2,950 PinCell/bwr195D 02/26/2004 10:05a 278,290 PinCell/bwr195D.out 02/26/2004 10:05a 2,965 PinCell/bwr21 02/26/2004 10:05a 279,360 PinCell/bwr21.out 02/26/2004 10:05a 2,969 PinCell/bwr215 02/26/2004 10:05a 279,360 PinCell/bwr215.out 02/26/2004 10:05a 2,945 PinCell/bwr21D 02/26/2004 10:05a 278,290 PinCell/bwr21D.out 02/26/2004 10:05a 2,965 PinCell/bwr22 02/26/2004 10:05a 279,360 PinCell/bwr22.out 02/26/2004 10:05a 2,965 PinCell/bwr23 02/26/2004 10:05a 279,360 PinCell/bwr23.out 02/26/2004 10:05a 2,965 PinCell/bwr25 02/26/2004 10:05a 279,360 PinCell/bwr25.out 05/06/2003 09:13a 2,980 PinCell/bwr7x7 05/06/2003 10:37a 440,588 PinCell/bwr7x7.out 07/09/2003 02:44p 2,981 PinCell/bwr7x75 07/09/2003 12:16p 440,588 PinCell/bwr7x75.out 02/26/2004 10:05a 2,962 PinCell/bwr7x75D 02/26/2004 10:05a 439,502 PinCell/bwr7x75D.out 07/10/2003 10:31a 2,980 PinCell/bwr14 07/11/2003 05:33a 440,588 PinCell/bwr14.out 02/26/2004 10:05a 2,959 PinCell/bwr14D 02/26/2004 10:05a 278,496 PinCell/bwr14D.out 07/10/2003 10:26a 2,964 PinCell/bwr15 07/11/2003 02:48a 440,588 PinCell/bwr15.out 02/26/2004 10:05a 2,944 PinCell/bwr15D 02/26/2004 10:05a 70,883 PinCell/bwr15D.out 07/10/2003 10:26a 2,964 PinCell/bwr16 07/10/2003 04:26p 440,588 PinCell/bwr16.out 02/26/2004 10:05a 2,942 PinCell/bwr16D Licensing & Nuclear Safety Calculation Title: Dry Transfer Facility Criticality Safety Calculations Document Identifier: 100-00C-WHS0-00100-000-00C Page I-4 of I-5 Date Time Size File 02/26/2004 10:05a 278,496 PinCell/bwr16D.out 07/10/2003 10:25a 2,964 PinCell/bwr17 07/10/2003 02:24p 440,794 PinCell/bwr17.out 02/26/2004 10:05a 2,948 PinCell/bwr17D 02/26/2004 10:05a 278,290 PinCell/bwr17D.out 07/10/2003 10:25a 2,964 PinCell/bwr18 07/10/2003 12:33p 440,794 PinCell/bwr18.out 02/26/2004 10:05a 2,943 PinCell/bwr18D 02/26/2004 10:05a 278,496 PinCell/bwr18D.out 07/08/2003 09:49a 3,229 PinCell/w17o09 07/08/2003 01:30p 440,667 PinCell/w17o09.out 03/02/2004 12:33p 3,209 PinCell/w17o09D 03/02/2004 12:33p 278,307 PinCell/w17o09D.out 07/08/2003 09:50a 3,209 PinCell/w17o10 07/08/2003 04:12p 440,667 PinCell/w17o10.out 03/02/2004 12:33p 3,191 PinCell/w17o10D 03/02/2004 12:33p 197,425 PinCell/w17o10D.out 07/08/2003 09:52a 3,201 PinCell/w17o11 07/09/2003 02:25a 440,461 PinCell/w17o11.out 03/02/2004 12:33p 3,184 PinCell/w17o11D 03/02/2004 12:33p 278,101 PinCell/w17o11D.out 02/26/2004 10:06a 3,128 PinCell/w17o12 02/26/2004 10:06a 279,233 PinCell/w17o12.out 03/02/2004 12:33p 3,181 PinCell/w17o12D 03/02/2004 12:33p 278,101 PinCell/w17o12D.out 02/26/2004 10:06a 3,128 PinCell/w17o13 02/26/2004 10:06a 279,233 PinCell/w17o13.out 02/26/2004 10:05a 3,110 PinCell/w17o13D 02/26/2004 10:05a 278,306 PinCell/w17o13D.out 02/26/2004 10:06a 3,128 PinCell/w17o14 02/26/2004 10:06a 279,233 PinCell/w17o14.out 02/26/2004 10:06a 3,128 PinCell/w17o145 02/26/2004 10:06a 279,233 PinCell/w17o145.out 03/02/2004 12:33p 3,111 PinCell/w17o14D 03/02/2004 12:33p 278,100 PinCell/w17o14D.out 02/26/2004 10:06a 3,128 PinCell/w17o15 02/26/2004 10:06a 279,233 PinCell/w17o15.out 02/26/2004 10:06a 3,128 PinCell/w17o16 02/26/2004 10:06a 279,233 PinCell/w17o16.out 02/26/2004 10:06a 3,128 PinCell/w17o17 02/26/2004 10:06a 279,233 PinCell/w17o17.out 02/26/2004 10:06a 3,128 PinCell/w17o18 02/26/2004 10:06a 279,233 PinCell/w17o18.out 02/26/2004 10:06a 3,128 PinCell/w17o19 02/26/2004 10:06a 279,233 PinCell/w17o19.out Licensing & Nuclear Safety Calculation Title: Dry Transfer Facility Criticality Safety Calculations Document Identifier: 100-00C-WHS0-00100-000-00C Page I-5 of I-5 Date Time Size File 05/07/2003 08:00a 3,277 PinCell/w17ofa 05/07/2003 10:10a 279,280 PinCell/w17ofa.out 04/19/2003 09:31a 3,275 PinCell/w17ofa5 04/19/2003 01:03p 440,461 PinCell/w17ofa5.out 03/02/2004 12:33p 3,231 PinCell/w17ofa5D 03/02/2004 12:34p 278,101 PinCell/w17ofa5D.out