ML15336A087: Difference between revisions
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==SUMMARY== | ==SUMMARY== | ||
DESCRIPTION | DESCRIPTION 2.0 DETAILED DESCRIPTION 2.1 Proposed Changes to the Technical Specifications 2.2 Need for Proposed Changes 3.0 TECHNICAL EVALUATION 3.1 Spent Fuel Pool Analysis 3.2 New Fuel Storage and Fuel Transfer Equipment Analysis 3.3 Spent Fuel Pool Transition Plan | ||
DESCRIPTION | |||
Changes to the Technical Specifications 2.2 Need for Proposed Changes 3.0 TECHNICAL EVALUATION | |||
Fuel Pool Analysis 3.2 New Fuel Storage and Fuel Transfer Equipment Analysis 3.3 Spent Fuel Pool Transition Plan | |||
==4.0 REGULATORY EVALUATION== | ==4.0 REGULATORY EVALUATION== | ||
4.1 Applicable Regulatory Requirements 4.2 Precedent 4.3 Significant Hazards Consideration 4.4 Conclusion 5.0 ENVIRONMENTAL CONSIDERATION | |||
Regulatory Requirements | |||
4.3 Significant Hazards Consideration | |||
5.0 ENVIRONMENTAL CONSIDERATION | |||
==6.0 REFERENCES== | ==6.0 REFERENCES== | ||
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Attachment 1 provides the marked-up existing TS pages. Attachment 2 provides the revised (clean) TS pages. Attachment 3 provides the marked-up TS Bases pages to show the proposed changes.This submittal contains new regulatory commitments (as defined by NEI 99-04, Guidelines for Managing NRC Commitment Changes, Revision 0) to be implemented, which are identified in Attachment | Attachment 1 provides the marked-up existing TS pages. Attachment 2 provides the revised (clean) TS pages. Attachment 3 provides the marked-up TS Bases pages to show the proposed changes.This submittal contains new regulatory commitments (as defined by NEI 99-04, Guidelines for Managing NRC Commitment Changes, Revision 0) to be implemented, which are identified in Attachment | ||
: 4. Attachment 5 provides a non-proprietary version of the criticality safety analysis.Attachment 6 provides a material qualification report for NETCO-SNAP-IN neutron absorbing spent fuel pool rack inserts.Attachment 7 is an affidavit signed by Westinghouse Electric Company LLC that sets forth the basis on which the proprietary information in Attachment 8 may be withheld from public disclosure by the Commission and addresses with specificity the considerations listed in 10 CFR 2.390(b)(4). | : 4. Attachment 5 provides a non-proprietary version of the criticality safety analysis.Attachment 6 provides a material qualification report for NETCO-SNAP-IN neutron absorbing spent fuel pool rack inserts.Attachment 7 is an affidavit signed by Westinghouse Electric Company LLC that sets forth the basis on which the proprietary information in Attachment 8 may be withheld from public disclosure by the Commission and addresses with specificity the considerations listed in 10 CFR 2.390(b)(4). | ||
Attachment 8 is the Criticality Safety Analysis for Palo Verde Nuclear Generating Station Units 1, 2, and 3, WCAP-1 8030-P (Proprietary), which contains information proprietary to Westinghouse Electric Company LLC.2.0 DETAILED DESCRIPTION | Attachment 8 is the Criticality Safety Analysis for Palo Verde Nuclear Generating Station Units 1, 2, and 3, WCAP-1 8030-P (Proprietary), which contains information proprietary to Westinghouse Electric Company LLC.2.0 DETAILED DESCRIPTION 2.1 Proposed Changes to the Technical Specifications The following specific TS changes are proposed as part of the updated criticality safety analysis.Marked-up TS pages are provided in Attachment 1 and revised (clean) TS pages are provided in Attachment 2.TS 3.7.17, Spent Fuel Assembly Storage-Revise the LCO statement and surveillance requirement statement to reflect the updated spent fuel assembly storage requirements resulting from WCAP-18030-P | ||
Changes to the Technical Specifications The following specific TS changes are proposed as part of the updated criticality safety analysis.Marked-up TS pages are provided in Attachment 1 and revised (clean) TS pages are provided in Attachment 2.TS 3.7.17, Spent Fuel Assembly Storage-Revise the LCO statement and surveillance requirement statement to reflect the updated spent fuel assembly storage requirements resulting from WCAP-18030-P | |||
-Add Tables 3.7.17-1 through 3.7.17-5 to define the Fuel Regions-Replace Figure 3.7.17-1 to define the Allowable Storage Arrays-Delete Figures 3.7.17-2 and 3.7.17-3 (information is replaced by Tables 3.7.17-1 through 3.7.17-5)I Enclosure Description and Assessment of Proposed License Amendment*TS 4.3.1, Criticality | -Add Tables 3.7.17-1 through 3.7.17-5 to define the Fuel Regions-Replace Figure 3.7.17-1 to define the Allowable Storage Arrays-Delete Figures 3.7.17-2 and 3.7.17-3 (information is replaced by Tables 3.7.17-1 through 3.7.17-5)I Enclosure Description and Assessment of Proposed License Amendment*TS 4.3.1, Criticality | ||
-Change TS 4.3.1.1 .a to read "4.65 weight percent" instead of "4.80 weight percent"-Change TS 4.3.1.1.c to read "1460 ppm" [parts per million] instead of "900 ppm"-Change TS 4.3.1.1.e to refer to Fuel Regions 1 -6as shown in Tables 3.7.17-1 through 3.7.17-5-Delete TS 4.3.1.1.f through 4.3.1.1.h (Fuel Regions are defined in Tables 3.7.17-1 through 3.7.17-5)-Change TS 4.3.1 .2.a to read "4.65 weight percent" instead of "4.80 weight percent"-Change TS 4.3.1 .2.d to replace "A nominal 17 inch center to center..." with"A nominal 18 inch (east-west) and 31 inch (north-south) center-to-center..."*Add new program TS 5.5.21, Spent Fuel Storage Rack Neutron Absorber Monitoring Program (Proposed TS 5.5.20, Risk In formed Completion Time Program, was submitted on July 31, 2015 [Agency Document Access and Management System (ADAMS)accession number ML15218A300]) | -Change TS 4.3.1.1 .a to read "4.65 weight percent" instead of "4.80 weight percent"-Change TS 4.3.1.1.c to read "1460 ppm" [parts per million] instead of "900 ppm"-Change TS 4.3.1.1.e to refer to Fuel Regions 1 -6as shown in Tables 3.7.17-1 through 3.7.17-5-Delete TS 4.3.1.1.f through 4.3.1.1.h (Fuel Regions are defined in Tables 3.7.17-1 through 3.7.17-5)-Change TS 4.3.1 .2.a to read "4.65 weight percent" instead of "4.80 weight percent"-Change TS 4.3.1 .2.d to replace "A nominal 17 inch center to center..." with"A nominal 18 inch (east-west) and 31 inch (north-south) center-to-center..."*Add new program TS 5.5.21, Spent Fuel Storage Rack Neutron Absorber Monitoring Program (Proposed TS 5.5.20, Risk In formed Completion Time Program, was submitted on July 31, 2015 [Agency Document Access and Management System (ADAMS)accession number ML15218A300]) | ||
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Therefore, TS 4.3.1 .2'.d must accurately reflect the as-built dimensions that must be maintained throughout the life of the plant. The racks have a nominal 18-inch center-to-center pitch on the short axis (east-west) and a 31-inch center-to-center pitch on the long axis (north-south). | Therefore, TS 4.3.1 .2'.d must accurately reflect the as-built dimensions that must be maintained throughout the life of the plant. The racks have a nominal 18-inch center-to-center pitch on the short axis (east-west) and a 31-inch center-to-center pitch on the long axis (north-south). | ||
According to both the existing NFS and updated criticality safety analyses, this rack design maintains keff <0.95 during all normal and accident conditions. | According to both the existing NFS and updated criticality safety analyses, this rack design maintains keff <0.95 during all normal and accident conditions. | ||
3.1 Spent Fuel Pool Analysis Design Approach The existing SEP storage racks are evaluated for the placement of fuel within the storage arrays described in the proposed TS changes. Credit is taken for the negative reactivity associated with burnup and post-irradiation cooling time (decay time). Additionally, some SEP storage arrays credit the presence of the neutron poison in the NETCO-SNAP-IN inserts. Finally, credit is taken for the presence of soluble boron in the SFP for specific conditions. | |||
Fuel Pool Analysis Design Approach The existing SEP storage racks are evaluated for the placement of fuel within the storage arrays described in the proposed TS changes. Credit is taken for the negative reactivity associated with burnup and post-irradiation cooling time (decay time). Additionally, some SEP storage arrays credit the presence of the neutron poison in the NETCO-SNAP-IN inserts. Finally, credit is taken for the presence of soluble boron in the SFP for specific conditions. | |||
Compliance for the SFP is demonstrated by establishing limits on the minimum allowable burnup as a function of initial enrichment and decay time for each fuel storage array. A conservative combination of best estimate and bounding values has been selected to model the fuel in the analysis to ensure that fuel represented by the proposed TS is less reactive than the fuel modeled in the analysis. | Compliance for the SFP is demonstrated by establishing limits on the minimum allowable burnup as a function of initial enrichment and decay time for each fuel storage array. A conservative combination of best estimate and bounding values has been selected to model the fuel in the analysis to ensure that fuel represented by the proposed TS is less reactive than the fuel modeled in the analysis. | ||
Therefore, burnup limits will conservatively bound fuel to be stored in the SFP.3 Enclosure Description and Assessment of Proposed License Amendment Acceptance Criteria The objective of the SEP criticality safety analysis is to ensure that the SEP operates within the bounds of 10 CFR 50.68(b)(4): | Therefore, burnup limits will conservatively bound fuel to be stored in the SFP.3 Enclosure Description and Assessment of Proposed License Amendment Acceptance Criteria The objective of the SEP criticality safety analysis is to ensure that the SEP operates within the bounds of 10 CFR 50.68(b)(4): | ||
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* Uncertainty in the predictive capability of SCALE 6.1.2 and the associated cross-section library* Temperature bias for operating temperature range* Planar enrichment bias 20 Enclosure Description and Assessment of Proposed License Amendment New Fuel Storagqe Rack Criticality Safety Analysis The criticality safety analysis for the NFS rack consists of determining the limiting fuel design under both the fully flooded and optimum moderation condition. | * Uncertainty in the predictive capability of SCALE 6.1.2 and the associated cross-section library* Temperature bias for operating temperature range* Planar enrichment bias 20 Enclosure Description and Assessment of Proposed License Amendment New Fuel Storagqe Rack Criticality Safety Analysis The criticality safety analysis for the NFS rack consists of determining the limiting fuel design under both the fully flooded and optimum moderation condition. | ||
Biases and uncertainties for both fully flooded and optimum moderation conditions are calculated using the limiting fuel design. The best estimate keff of the NES rack under both full density water and optimum moderation conditions is less than the target keff. This demonstrates that the NFS rack complies with the requirements of 10 CFR 50.68. Additional information is provided in Section B.4.6 of Attachment 8.The analysis of the NES rack has demonstrated that it can be operated in its design capacity without risk of exceeding the maximum reactivity imposed by regulation. | Biases and uncertainties for both fully flooded and optimum moderation conditions are calculated using the limiting fuel design. The best estimate keff of the NES rack under both full density water and optimum moderation conditions is less than the target keff. This demonstrates that the NFS rack complies with the requirements of 10 CFR 50.68. Additional information is provided in Section B.4.6 of Attachment 8.The analysis of the NES rack has demonstrated that it can be operated in its design capacity without risk of exceeding the maximum reactivity imposed by regulation. | ||
The analysis supports use of these components up to a maximum radially averaged enrichment of 4.65 weight percent U-235. All fuel used to date at PVNGS has an initial enrichment of < 4.55 weight percent U-235.Intermediate Fuel Storaqe Rack Criticality Safety Analysis The criticality safety analysis for the IFSR consists of determining the target keff for the IFSR, then confirming that the best estimate system keff (plus 2 o) is below the target keff with the limiting fuel design. The analysis uses the NGF design because the NGF is more reactive with full density water than VAP fuel. The biases and uncertainties are also calculated using the NGF design. The best estimate keff of the IFSR is less than the target keff, which demonstrates that the IFSR complies with the requirements of 10 CFR 50.68. Additional information is provided in Section B.4.7 of Attachment 8.Fuel Upender. Transfer Machine, and New Fuel Elevator Criticality Safety Analysis The criticality safety analysis for the fuel upender and transfer machine, and the analysis for the new fuel elevator, demonstrate that they can be used with fresh 4.65 weight percent U-235 fuel without exceeding a keff of 0.95 at a 95 percent probability, 95 percent confidence level. The design basis fuel is the NGF design. The best estimate keff of the fuel upender and transfer machine, and the new fuel elevator, is less than the target which demonstrates compliance with the requirements of 10 CFR 50.68. Additional information is provided in Section B.4.8 of Attachment | The analysis supports use of these components up to a maximum radially averaged enrichment of 4.65 weight percent U-235. All fuel used to date at PVNGS has an initial enrichment of < 4.55 weight percent U-235.Intermediate Fuel Storaqe Rack Criticality Safety Analysis The criticality safety analysis for the IFSR consists of determining the target keff for the IFSR, then confirming that the best estimate system keff (plus 2 o) is below the target keff with the limiting fuel design. The analysis uses the NGF design because the NGF is more reactive with full density water than VAP fuel. The biases and uncertainties are also calculated using the NGF design. The best estimate keff of the IFSR is less than the target keff, which demonstrates that the IFSR complies with the requirements of 10 CFR 50.68. Additional information is provided in Section B.4.7 of Attachment 8.Fuel Upender. Transfer Machine, and New Fuel Elevator Criticality Safety Analysis The criticality safety analysis for the fuel upender and transfer machine, and the analysis for the new fuel elevator, demonstrate that they can be used with fresh 4.65 weight percent U-235 fuel without exceeding a keff of 0.95 at a 95 percent probability, 95 percent confidence level. The design basis fuel is the NGF design. The best estimate keff of the fuel upender and transfer machine, and the new fuel elevator, is less than the target which demonstrates compliance with the requirements of 10 CFR 50.68. Additional information is provided in Section B.4.8 of Attachment 8.3.3 Spent Fuel Pool Transition Plan The SEP transition will be conducted over a total lapsed time of approximately 24 months with a schedule based on the unit refueling outages. Therefore, APS will insert two sets of TS and TS Bases pages during implementation. | ||
Fuel Pool Transition Plan The SEP transition will be conducted over a total lapsed time of approximately 24 months with a schedule based on the unit refueling outages. Therefore, APS will insert two sets of TS and TS Bases pages during implementation. | |||
One set will be labeled "Before SEP transition" and the other set will be labeled "After SFP transition." The Spent Fuel Pool Transition Plan is based on a SEP module-by-module transition scheme.Each SEP module that has not been transitioned is governed by the 'Before SEP transition" pages. As each module in a SEP is transitioned to the new configuration, the Shift Manager will make an entry in the control room log and declare that module as "transitioned." That particular module is then governed by the "After SEP transition" pages. When all three units have been transitioned, APS will submit an administrative TS change to remove the "before" and "after" pages, and insert the final pages.21 Enclosure Description and Assessment of Proposed License Amendment APS will transition to the proposed TS 3.7.17 in each of the three units in the following manner: 1. Move fuel assemblies as needed in order to neutronically decouple one module from the balance of the SEP. Analysis demonstrates that one row of empty cells is enough to decouple modules in the SEP.2. Perform a shuffle of the decoupled module, including installation of NETCO-SNAP-IN rack inserts.a. Some modules may need to be completely emptied of fuel assemblies. | One set will be labeled "Before SEP transition" and the other set will be labeled "After SFP transition." The Spent Fuel Pool Transition Plan is based on a SEP module-by-module transition scheme.Each SEP module that has not been transitioned is governed by the 'Before SEP transition" pages. As each module in a SEP is transitioned to the new configuration, the Shift Manager will make an entry in the control room log and declare that module as "transitioned." That particular module is then governed by the "After SEP transition" pages. When all three units have been transitioned, APS will submit an administrative TS change to remove the "before" and "after" pages, and insert the final pages.21 Enclosure Description and Assessment of Proposed License Amendment APS will transition to the proposed TS 3.7.17 in each of the three units in the following manner: 1. Move fuel assemblies as needed in order to neutronically decouple one module from the balance of the SEP. Analysis demonstrates that one row of empty cells is enough to decouple modules in the SEP.2. Perform a shuffle of the decoupled module, including installation of NETCO-SNAP-IN rack inserts.a. Some modules may need to be completely emptied of fuel assemblies. | ||
The assemblies that must be moved may be stored in the appropriate regions of the rest of the SEP.b. Fuel assemblies that already meet the new TS 3.17.17 requirements may not need to be shuffled.3. Upon completion of the fuel shuffle and NETCO-SNAP-IN rack insert installation, the Shift Manager will declare the module has transitioned to the new TS 3.7.17 and enter this information into the control room log.4. Perform additional fuel shuffles, as needed, in order to move fuel from other parts of the SEP to the recently transitioned module.5. Repeat steps 1 through 4 for all 17 modules.6. Perform a 100 percent pool verification to confirm the following a. Fuel has been properly moved as confirmed by a 100 percent serial number check b. Stainless steel L-Jnserts are in the locations assumed in the analysis of record c. NETCO-SNAP-IN rack inserts are in their assumed locations d. Blocking devices are in their assumed locations A generic Westinghouse study (Reference 6.6) investigated the adequacy of assuming a distance of one cell pitch for neutronic isolation. | The assemblies that must be moved may be stored in the appropriate regions of the rest of the SEP.b. Fuel assemblies that already meet the new TS 3.17.17 requirements may not need to be shuffled.3. Upon completion of the fuel shuffle and NETCO-SNAP-IN rack insert installation, the Shift Manager will declare the module has transitioned to the new TS 3.7.17 and enter this information into the control room log.4. Perform additional fuel shuffles, as needed, in order to move fuel from other parts of the SEP to the recently transitioned module.5. Repeat steps 1 through 4 for all 17 modules.6. Perform a 100 percent pool verification to confirm the following a. Fuel has been properly moved as confirmed by a 100 percent serial number check b. Stainless steel L-Jnserts are in the locations assumed in the analysis of record c. NETCO-SNAP-IN rack inserts are in their assumed locations d. Blocking devices are in their assumed locations A generic Westinghouse study (Reference 6.6) investigated the adequacy of assuming a distance of one cell pitch for neutronic isolation. | ||
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==4.0 REGULATORY EVALUATION== | ==4.0 REGULATORY EVALUATION== | ||
4.1 Applicable Regulatory Requirements The regulations in 10 CFR 50.36(c)(2)(ii)(B), Limiting conditions for operation, state: 22 Enclosure Description and Assessment of Proposed License Amendment Criterion | |||
Regulatory Requirements The regulations in 10 CFR 50.36(c)(2)(ii)(B), Limiting conditions for operation, state: 22 Enclosure Description and Assessment of Proposed License Amendment Criterion | |||
: 2. A process variable, design feature, or operating restriction that is an initial condition of a design basis accident or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.Technical Specification (TS) 3. 7.17 currently meets this requirement and will continue to meet this requirement after the proposed changes are approved and implemented. | : 2. A process variable, design feature, or operating restriction that is an initial condition of a design basis accident or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.Technical Specification (TS) 3. 7.17 currently meets this requirement and will continue to meet this requirement after the proposed changes are approved and implemented. | ||
The regulations in 10 CFR 50.36(c)(4), Design features, state: Design features to be included are those features of the facility such as materials of construction and geometric arrangements, which, if altered or modified, would have a significant effect on safety and are not covered in categories described in paragraphs (c)(1), (2), and (3) of this section.TS 4.3.1 currently meets this requirement and will continue to meet this requirement after the proposed changes are approved and implemented. | The regulations in 10 CFR 50.36(c)(4), Design features, state: Design features to be included are those features of the facility such as materials of construction and geometric arrangements, which, if altered or modified, would have a significant effect on safety and are not covered in categories described in paragraphs (c)(1), (2), and (3) of this section.TS 4.3.1 currently meets this requirement and will continue to meet this requirement after the proposed changes are approved and implemented. | ||
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TS 3. 7.17 currently meets this requirement and will continue to meet this requirement after the proposed changes are approved and implemented. | TS 3. 7.17 currently meets this requirement and will continue to meet this requirement after the proposed changes are approved and implemented. | ||
The guidance in DSS-ISG-2010-01 is to be used by NRC staff to review nuclear criticality safety analyses for the storage of new and spent nuclear fuel as they apply to applications for license amendments submitted after September 29, 2011.This license amendment request and WCAP-18030-P were developed using the guidance of DSS-ISG-2010-01. | The guidance in DSS-ISG-2010-01 is to be used by NRC staff to review nuclear criticality safety analyses for the storage of new and spent nuclear fuel as they apply to applications for license amendments submitted after September 29, 2011.This license amendment request and WCAP-18030-P were developed using the guidance of DSS-ISG-2010-01. | ||
4.2 Precedent The analysis methodology for the site-specific criticality analysis employs the PARAGON code, which is approved for use by the NRC (Reference 6.4).4.:3 Significant Hazards Consideration As required by 10 CFR 50.91(a), Notice for Public Comment, an analysis of the issue of no significant hazards consideration using the standards in 10 CFR 50.92, Issuance of Amendment, is presented below: 1. Does the proposed amendment involve a significant increase in the probability or consequences of an accident previously evaluated? | |||
The analysis methodology for the site-specific criticality analysis employs the PARAGON code, which is approved for use by the NRC (Reference 6.4).4.:3 Significant Hazards Consideration As required by 10 CFR 50.91(a), Notice for Public Comment, an analysis of the issue of no significant hazards consideration using the standards in 10 CFR 50.92, Issuance of Amendment, is presented below: 1. Does the proposed amendment involve a significant increase in the probability or consequences of an accident previously evaluated? | |||
Response: | Response: | ||
No.24 Enclosure Description and Assessment of Proposed License Amendment The proposed amendment would modify the Palo Verde Nuclear Generating Station (PVNGS) Technical Specifications (TS) to incorporate the results of an updated criticality safety analysis for both new fuel and spent fuel storage. The revised criticality safety analysis provides an updated methodology that allows credit for neutron absorbing NETCO-SNAP-IN rack inserts and corrects non-conservative input assumptions in the previous criticality safety analysis.The proposed amendment does not change or modify the fuel, fuel handling processes, number of fuel assemblies that may be stored in the spent fuel pool (SFP), decay heat generation rate, or the SFP cooling and cleanup system. The proposed amendment was evaluated for impact on the following previously evaluated events and accidents: | No.24 Enclosure Description and Assessment of Proposed License Amendment The proposed amendment would modify the Palo Verde Nuclear Generating Station (PVNGS) Technical Specifications (TS) to incorporate the results of an updated criticality safety analysis for both new fuel and spent fuel storage. The revised criticality safety analysis provides an updated methodology that allows credit for neutron absorbing NETCO-SNAP-IN rack inserts and corrects non-conservative input assumptions in the previous criticality safety analysis.The proposed amendment does not change or modify the fuel, fuel handling processes, number of fuel assemblies that may be stored in the spent fuel pool (SFP), decay heat generation rate, or the SFP cooling and cleanup system. The proposed amendment was evaluated for impact on the following previously evaluated events and accidents: | ||
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==6.0 REFERENCES== | ==6.0 REFERENCES== | ||
6.1 Criticality Safety Analysis for Palo Verde Nuclear Generating Station Units 1, 2, and 3 (Proprietary), WCAP-1 8030-P, Revision 0, September 2015.6.2 Staff Guidance Regarding the Nuclear Criticality Safety Analysis for Spent Fuel Pools, DSS-ISG-201 0-01, Revision 0, Nuclear Regulatory Commission Division of Safety Systems, Rockville, MD, September 29, 2011. (ML1 10620086)6.3 Scale: A Comprehensive Modeling and Simulation Suite for Nuclear Safety Analysis and Design, ORNL/TM-2005/39, Version 6.1, Oak Ridge National Laboratory, Oak Ridge, TN, June 2011.6.4 M. Ouisloumen, H. Huria, et al, Qualification of the Two-Dimensional Transport Code PARAGON, WCAP-16045-P-A, Revision 0, Westinghouse Electric Company LLC, Monroeville, PA, August 2004.28 Enclosure Description and Assessment of Proposed License Amendment 6.5 C. V. Parks, et al, Review and Prioritization of Technical Issues Related to Bumup Credit for LWR Fuel, NUREG/CR-6665, Oak Ridge National Laboratory, Oak Ridge, TN, February 2000.6,6 Letter, J. Gresham (WEC) to NRC, Responses to Requests for Additional Information from the Review of WCAP- 1 7483-PA/WCAP-1 7483-NP, Revision 0, 'Westinghouse Methodology for Spent Fuel Pool and New Fuel Rack Criticality Safety Analysis,' | |||
Safety Analysis for Palo Verde Nuclear Generating Station Units 1, 2, and 3 (Proprietary), WCAP-1 8030-P, Revision 0, September 2015.6.2 Staff Guidance Regarding the Nuclear Criticality Safety Analysis for Spent Fuel Pools, DSS-ISG-201 0-01, Revision 0, Nuclear Regulatory Commission Division of Safety Systems, Rockville, MD, September 29, 2011. (ML1 10620086)6.3 Scale: A Comprehensive Modeling and Simulation Suite for Nuclear Safety Analysis and Design, ORNL/TM-2005/39, Version 6.1, Oak Ridge National Laboratory, Oak Ridge, TN, June 2011.6.4 M. Ouisloumen, H. Huria, et al, Qualification of the Two-Dimensional Transport Code PARAGON, WCAP-16045-P-A, Revision 0, Westinghouse Electric Company LLC, Monroeville, PA, August 2004.28 Enclosure Description and Assessment of Proposed License Amendment 6.5 C. V. Parks, et al, Review and Prioritization of Technical Issues Related to Bumup Credit for LWR Fuel, NUREG/CR-6665, Oak Ridge National Laboratory, Oak Ridge, TN, February 2000.6,6 Letter, J. Gresham (WEC) to NRC, Responses to Requests for Additional Information from the Review of WCAP- 1 7483-PA/WCAP-1 7483-NP, Revision 0, 'Westinghouse Methodology for Spent Fuel Pool and New Fuel Rack Criticality Safety Analysis,' | |||
LTR-NRC-15-60, dated July 20, 2015.29 Enclosure ,Description and Assessment of Proposed License Amendment ATTACHMENT 1 Marked-up Technical Specifications Pages (Pages Provided for Before and After SEP Transition) 3.7.17-1-3.7.17-2 3.7.17-3 3.7.17-4 4.0-2 4.0-3 5.5-19 SBefore SFP transitionI Spent Fuel Assembly Storage 3.7.17 3.7 PLANT SYSTEMS 3.7.17 Spent Fuel Assembly Storage LCO 3.7.17 APPLICABILITY: | LTR-NRC-15-60, dated July 20, 2015.29 Enclosure ,Description and Assessment of Proposed License Amendment ATTACHMENT 1 Marked-up Technical Specifications Pages (Pages Provided for Before and After SEP Transition) 3.7.17-1-3.7.17-2 3.7.17-3 3.7.17-4 4.0-2 4.0-3 5.5-19 SBefore SFP transitionI Spent Fuel Assembly Storage 3.7.17 3.7 PLANT SYSTEMS 3.7.17 Spent Fuel Assembly Storage LCO 3.7.17 APPLICABILITY: | ||
The combination of initial enrichment, burnup, and decay time of each fuel assembly stored in each of the four regions of the fuel storage pool shall be within the acceptable burnup domain for each region as shown in Figures 3.7.17-1, 3.7.17-2, or 3.7.17-3, and described in Specification 4.3.1.1.Whenever any fuel assembly is stored in the fuel storage pool.ACTIONS__________________________ | The combination of initial enrichment, burnup, and decay time of each fuel assembly stored in each of the four regions of the fuel storage pool shall be within the acceptable burnup domain for each region as shown in Figures 3.7.17-1, 3.7.17-2, or 3.7.17-3, and described in Specification 4.3.1.1.Whenever any fuel assembly is stored in the fuel storage pool.ACTIONS__________________________ | ||
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==SUMMARY== | ==SUMMARY== | ||
DESCRIPTION | DESCRIPTION 2.0 DETAILED DESCRIPTION 2.1 Proposed Changes to the Technical Specifications 2.2 Need for Proposed Changes 3.0 TECHNICAL EVALUATION 3.1 Spent Fuel Pool Analysis 3.2 New Fuel Storage and Fuel Transfer Equipment Analysis 3.3 Spent Fuel Pool Transition Plan | ||
DESCRIPTION | |||
Changes to the Technical Specifications 2.2 Need for Proposed Changes 3.0 TECHNICAL EVALUATION | |||
Fuel Pool Analysis 3.2 New Fuel Storage and Fuel Transfer Equipment Analysis 3.3 Spent Fuel Pool Transition Plan | |||
==4.0 REGULATORY EVALUATION== | ==4.0 REGULATORY EVALUATION== | ||
4.1 Applicable Regulatory Requirements 4.2 Precedent 4.3 Significant Hazards Consideration 4.4 Conclusion 5.0 ENVIRONMENTAL CONSIDERATION | |||
Regulatory Requirements | |||
4.3 Significant Hazards Consideration | |||
5.0 ENVIRONMENTAL CONSIDERATION | |||
==6.0 REFERENCES== | ==6.0 REFERENCES== | ||
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Attachment 1 provides the marked-up existing TS pages. Attachment 2 provides the revised (clean) TS pages. Attachment 3 provides the marked-up TS Bases pages to show the proposed changes.This submittal contains new regulatory commitments (as defined by NEI 99-04, Guidelines for Managing NRC Commitment Changes, Revision 0) to be implemented, which are identified in Attachment | Attachment 1 provides the marked-up existing TS pages. Attachment 2 provides the revised (clean) TS pages. Attachment 3 provides the marked-up TS Bases pages to show the proposed changes.This submittal contains new regulatory commitments (as defined by NEI 99-04, Guidelines for Managing NRC Commitment Changes, Revision 0) to be implemented, which are identified in Attachment | ||
: 4. Attachment 5 provides a non-proprietary version of the criticality safety analysis.Attachment 6 provides a material qualification report for NETCO-SNAP-IN neutron absorbing spent fuel pool rack inserts.Attachment 7 is an affidavit signed by Westinghouse Electric Company LLC that sets forth the basis on which the proprietary information in Attachment 8 may be withheld from public disclosure by the Commission and addresses with specificity the considerations listed in 10 CFR 2.390(b)(4). | : 4. Attachment 5 provides a non-proprietary version of the criticality safety analysis.Attachment 6 provides a material qualification report for NETCO-SNAP-IN neutron absorbing spent fuel pool rack inserts.Attachment 7 is an affidavit signed by Westinghouse Electric Company LLC that sets forth the basis on which the proprietary information in Attachment 8 may be withheld from public disclosure by the Commission and addresses with specificity the considerations listed in 10 CFR 2.390(b)(4). | ||
Attachment 8 is the Criticality Safety Analysis for Palo Verde Nuclear Generating Station Units 1, 2, and 3, WCAP-1 8030-P (Proprietary), which contains information proprietary to Westinghouse Electric Company LLC.2.0 DETAILED DESCRIPTION | Attachment 8 is the Criticality Safety Analysis for Palo Verde Nuclear Generating Station Units 1, 2, and 3, WCAP-1 8030-P (Proprietary), which contains information proprietary to Westinghouse Electric Company LLC.2.0 DETAILED DESCRIPTION 2.1 Proposed Changes to the Technical Specifications The following specific TS changes are proposed as part of the updated criticality safety analysis.Marked-up TS pages are provided in Attachment 1 and revised (clean) TS pages are provided in Attachment 2.TS 3.7.17, Spent Fuel Assembly Storage-Revise the LCO statement and surveillance requirement statement to reflect the updated spent fuel assembly storage requirements resulting from WCAP-18030-P | ||
Changes to the Technical Specifications The following specific TS changes are proposed as part of the updated criticality safety analysis.Marked-up TS pages are provided in Attachment 1 and revised (clean) TS pages are provided in Attachment 2.TS 3.7.17, Spent Fuel Assembly Storage-Revise the LCO statement and surveillance requirement statement to reflect the updated spent fuel assembly storage requirements resulting from WCAP-18030-P | |||
-Add Tables 3.7.17-1 through 3.7.17-5 to define the Fuel Regions-Replace Figure 3.7.17-1 to define the Allowable Storage Arrays-Delete Figures 3.7.17-2 and 3.7.17-3 (information is replaced by Tables 3.7.17-1 through 3.7.17-5)I Enclosure Description and Assessment of Proposed License Amendment*TS 4.3.1, Criticality | -Add Tables 3.7.17-1 through 3.7.17-5 to define the Fuel Regions-Replace Figure 3.7.17-1 to define the Allowable Storage Arrays-Delete Figures 3.7.17-2 and 3.7.17-3 (information is replaced by Tables 3.7.17-1 through 3.7.17-5)I Enclosure Description and Assessment of Proposed License Amendment*TS 4.3.1, Criticality | ||
-Change TS 4.3.1.1 .a to read "4.65 weight percent" instead of "4.80 weight percent"-Change TS 4.3.1.1.c to read "1460 ppm" [parts per million] instead of "900 ppm"-Change TS 4.3.1.1.e to refer to Fuel Regions 1 -6as shown in Tables 3.7.17-1 through 3.7.17-5-Delete TS 4.3.1.1.f through 4.3.1.1.h (Fuel Regions are defined in Tables 3.7.17-1 through 3.7.17-5)-Change TS 4.3.1 .2.a to read "4.65 weight percent" instead of "4.80 weight percent"-Change TS 4.3.1 .2.d to replace "A nominal 17 inch center to center..." with"A nominal 18 inch (east-west) and 31 inch (north-south) center-to-center..."*Add new program TS 5.5.21, Spent Fuel Storage Rack Neutron Absorber Monitoring Program (Proposed TS 5.5.20, Risk In formed Completion Time Program, was submitted on July 31, 2015 [Agency Document Access and Management System (ADAMS)accession number ML15218A300]) | -Change TS 4.3.1.1 .a to read "4.65 weight percent" instead of "4.80 weight percent"-Change TS 4.3.1.1.c to read "1460 ppm" [parts per million] instead of "900 ppm"-Change TS 4.3.1.1.e to refer to Fuel Regions 1 -6as shown in Tables 3.7.17-1 through 3.7.17-5-Delete TS 4.3.1.1.f through 4.3.1.1.h (Fuel Regions are defined in Tables 3.7.17-1 through 3.7.17-5)-Change TS 4.3.1 .2.a to read "4.65 weight percent" instead of "4.80 weight percent"-Change TS 4.3.1 .2.d to replace "A nominal 17 inch center to center..." with"A nominal 18 inch (east-west) and 31 inch (north-south) center-to-center..."*Add new program TS 5.5.21, Spent Fuel Storage Rack Neutron Absorber Monitoring Program (Proposed TS 5.5.20, Risk In formed Completion Time Program, was submitted on July 31, 2015 [Agency Document Access and Management System (ADAMS)accession number ML15218A300]) | ||
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Therefore, TS 4.3.1 .2'.d must accurately reflect the as-built dimensions that must be maintained throughout the life of the plant. The racks have a nominal 18-inch center-to-center pitch on the short axis (east-west) and a 31-inch center-to-center pitch on the long axis (north-south). | Therefore, TS 4.3.1 .2'.d must accurately reflect the as-built dimensions that must be maintained throughout the life of the plant. The racks have a nominal 18-inch center-to-center pitch on the short axis (east-west) and a 31-inch center-to-center pitch on the long axis (north-south). | ||
According to both the existing NFS and updated criticality safety analyses, this rack design maintains keff <0.95 during all normal and accident conditions. | According to both the existing NFS and updated criticality safety analyses, this rack design maintains keff <0.95 during all normal and accident conditions. | ||
3.1 Spent Fuel Pool Analysis Design Approach The existing SEP storage racks are evaluated for the placement of fuel within the storage arrays described in the proposed TS changes. Credit is taken for the negative reactivity associated with burnup and post-irradiation cooling time (decay time). Additionally, some SEP storage arrays credit the presence of the neutron poison in the NETCO-SNAP-IN inserts. Finally, credit is taken for the presence of soluble boron in the SFP for specific conditions. | |||
Fuel Pool Analysis Design Approach The existing SEP storage racks are evaluated for the placement of fuel within the storage arrays described in the proposed TS changes. Credit is taken for the negative reactivity associated with burnup and post-irradiation cooling time (decay time). Additionally, some SEP storage arrays credit the presence of the neutron poison in the NETCO-SNAP-IN inserts. Finally, credit is taken for the presence of soluble boron in the SFP for specific conditions. | |||
Compliance for the SFP is demonstrated by establishing limits on the minimum allowable burnup as a function of initial enrichment and decay time for each fuel storage array. A conservative combination of best estimate and bounding values has been selected to model the fuel in the analysis to ensure that fuel represented by the proposed TS is less reactive than the fuel modeled in the analysis. | Compliance for the SFP is demonstrated by establishing limits on the minimum allowable burnup as a function of initial enrichment and decay time for each fuel storage array. A conservative combination of best estimate and bounding values has been selected to model the fuel in the analysis to ensure that fuel represented by the proposed TS is less reactive than the fuel modeled in the analysis. | ||
Therefore, burnup limits will conservatively bound fuel to be stored in the SFP.3 Enclosure Description and Assessment of Proposed License Amendment Acceptance Criteria The objective of the SEP criticality safety analysis is to ensure that the SEP operates within the bounds of 10 CFR 50.68(b)(4): | Therefore, burnup limits will conservatively bound fuel to be stored in the SFP.3 Enclosure Description and Assessment of Proposed License Amendment Acceptance Criteria The objective of the SEP criticality safety analysis is to ensure that the SEP operates within the bounds of 10 CFR 50.68(b)(4): | ||
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* Uncertainty in the predictive capability of SCALE 6.1.2 and the associated cross-section library* Temperature bias for operating temperature range* Planar enrichment bias 20 Enclosure Description and Assessment of Proposed License Amendment New Fuel Storagqe Rack Criticality Safety Analysis The criticality safety analysis for the NFS rack consists of determining the limiting fuel design under both the fully flooded and optimum moderation condition. | * Uncertainty in the predictive capability of SCALE 6.1.2 and the associated cross-section library* Temperature bias for operating temperature range* Planar enrichment bias 20 Enclosure Description and Assessment of Proposed License Amendment New Fuel Storagqe Rack Criticality Safety Analysis The criticality safety analysis for the NFS rack consists of determining the limiting fuel design under both the fully flooded and optimum moderation condition. | ||
Biases and uncertainties for both fully flooded and optimum moderation conditions are calculated using the limiting fuel design. The best estimate keff of the NES rack under both full density water and optimum moderation conditions is less than the target keff. This demonstrates that the NFS rack complies with the requirements of 10 CFR 50.68. Additional information is provided in Section B.4.6 of Attachment 8.The analysis of the NES rack has demonstrated that it can be operated in its design capacity without risk of exceeding the maximum reactivity imposed by regulation. | Biases and uncertainties for both fully flooded and optimum moderation conditions are calculated using the limiting fuel design. The best estimate keff of the NES rack under both full density water and optimum moderation conditions is less than the target keff. This demonstrates that the NFS rack complies with the requirements of 10 CFR 50.68. Additional information is provided in Section B.4.6 of Attachment 8.The analysis of the NES rack has demonstrated that it can be operated in its design capacity without risk of exceeding the maximum reactivity imposed by regulation. | ||
The analysis supports use of these components up to a maximum radially averaged enrichment of 4.65 weight percent U-235. All fuel used to date at PVNGS has an initial enrichment of < 4.55 weight percent U-235.Intermediate Fuel Storaqe Rack Criticality Safety Analysis The criticality safety analysis for the IFSR consists of determining the target keff for the IFSR, then confirming that the best estimate system keff (plus 2 o) is below the target keff with the limiting fuel design. The analysis uses the NGF design because the NGF is more reactive with full density water than VAP fuel. The biases and uncertainties are also calculated using the NGF design. The best estimate keff of the IFSR is less than the target keff, which demonstrates that the IFSR complies with the requirements of 10 CFR 50.68. Additional information is provided in Section B.4.7 of Attachment 8.Fuel Upender. Transfer Machine, and New Fuel Elevator Criticality Safety Analysis The criticality safety analysis for the fuel upender and transfer machine, and the analysis for the new fuel elevator, demonstrate that they can be used with fresh 4.65 weight percent U-235 fuel without exceeding a keff of 0.95 at a 95 percent probability, 95 percent confidence level. The design basis fuel is the NGF design. The best estimate keff of the fuel upender and transfer machine, and the new fuel elevator, is less than the target which demonstrates compliance with the requirements of 10 CFR 50.68. Additional information is provided in Section B.4.8 of Attachment | The analysis supports use of these components up to a maximum radially averaged enrichment of 4.65 weight percent U-235. All fuel used to date at PVNGS has an initial enrichment of < 4.55 weight percent U-235.Intermediate Fuel Storaqe Rack Criticality Safety Analysis The criticality safety analysis for the IFSR consists of determining the target keff for the IFSR, then confirming that the best estimate system keff (plus 2 o) is below the target keff with the limiting fuel design. The analysis uses the NGF design because the NGF is more reactive with full density water than VAP fuel. The biases and uncertainties are also calculated using the NGF design. The best estimate keff of the IFSR is less than the target keff, which demonstrates that the IFSR complies with the requirements of 10 CFR 50.68. Additional information is provided in Section B.4.7 of Attachment 8.Fuel Upender. Transfer Machine, and New Fuel Elevator Criticality Safety Analysis The criticality safety analysis for the fuel upender and transfer machine, and the analysis for the new fuel elevator, demonstrate that they can be used with fresh 4.65 weight percent U-235 fuel without exceeding a keff of 0.95 at a 95 percent probability, 95 percent confidence level. The design basis fuel is the NGF design. The best estimate keff of the fuel upender and transfer machine, and the new fuel elevator, is less than the target which demonstrates compliance with the requirements of 10 CFR 50.68. Additional information is provided in Section B.4.8 of Attachment 8.3.3 Spent Fuel Pool Transition Plan The SEP transition will be conducted over a total lapsed time of approximately 24 months with a schedule based on the unit refueling outages. Therefore, APS will insert two sets of TS and TS Bases pages during implementation. | ||
Fuel Pool Transition Plan The SEP transition will be conducted over a total lapsed time of approximately 24 months with a schedule based on the unit refueling outages. Therefore, APS will insert two sets of TS and TS Bases pages during implementation. | |||
One set will be labeled "Before SEP transition" and the other set will be labeled "After SFP transition." The Spent Fuel Pool Transition Plan is based on a SEP module-by-module transition scheme.Each SEP module that has not been transitioned is governed by the 'Before SEP transition" pages. As each module in a SEP is transitioned to the new configuration, the Shift Manager will make an entry in the control room log and declare that module as "transitioned." That particular module is then governed by the "After SEP transition" pages. When all three units have been transitioned, APS will submit an administrative TS change to remove the "before" and "after" pages, and insert the final pages.21 Enclosure Description and Assessment of Proposed License Amendment APS will transition to the proposed TS 3.7.17 in each of the three units in the following manner: 1. Move fuel assemblies as needed in order to neutronically decouple one module from the balance of the SEP. Analysis demonstrates that one row of empty cells is enough to decouple modules in the SEP.2. Perform a shuffle of the decoupled module, including installation of NETCO-SNAP-IN rack inserts.a. Some modules may need to be completely emptied of fuel assemblies. | One set will be labeled "Before SEP transition" and the other set will be labeled "After SFP transition." The Spent Fuel Pool Transition Plan is based on a SEP module-by-module transition scheme.Each SEP module that has not been transitioned is governed by the 'Before SEP transition" pages. As each module in a SEP is transitioned to the new configuration, the Shift Manager will make an entry in the control room log and declare that module as "transitioned." That particular module is then governed by the "After SEP transition" pages. When all three units have been transitioned, APS will submit an administrative TS change to remove the "before" and "after" pages, and insert the final pages.21 Enclosure Description and Assessment of Proposed License Amendment APS will transition to the proposed TS 3.7.17 in each of the three units in the following manner: 1. Move fuel assemblies as needed in order to neutronically decouple one module from the balance of the SEP. Analysis demonstrates that one row of empty cells is enough to decouple modules in the SEP.2. Perform a shuffle of the decoupled module, including installation of NETCO-SNAP-IN rack inserts.a. Some modules may need to be completely emptied of fuel assemblies. | ||
The assemblies that must be moved may be stored in the appropriate regions of the rest of the SEP.b. Fuel assemblies that already meet the new TS 3.17.17 requirements may not need to be shuffled.3. Upon completion of the fuel shuffle and NETCO-SNAP-IN rack insert installation, the Shift Manager will declare the module has transitioned to the new TS 3.7.17 and enter this information into the control room log.4. Perform additional fuel shuffles, as needed, in order to move fuel from other parts of the SEP to the recently transitioned module.5. Repeat steps 1 through 4 for all 17 modules.6. Perform a 100 percent pool verification to confirm the following a. Fuel has been properly moved as confirmed by a 100 percent serial number check b. Stainless steel L-Jnserts are in the locations assumed in the analysis of record c. NETCO-SNAP-IN rack inserts are in their assumed locations d. Blocking devices are in their assumed locations A generic Westinghouse study (Reference 6.6) investigated the adequacy of assuming a distance of one cell pitch for neutronic isolation. | The assemblies that must be moved may be stored in the appropriate regions of the rest of the SEP.b. Fuel assemblies that already meet the new TS 3.17.17 requirements may not need to be shuffled.3. Upon completion of the fuel shuffle and NETCO-SNAP-IN rack insert installation, the Shift Manager will declare the module has transitioned to the new TS 3.7.17 and enter this information into the control room log.4. Perform additional fuel shuffles, as needed, in order to move fuel from other parts of the SEP to the recently transitioned module.5. Repeat steps 1 through 4 for all 17 modules.6. Perform a 100 percent pool verification to confirm the following a. Fuel has been properly moved as confirmed by a 100 percent serial number check b. Stainless steel L-Jnserts are in the locations assumed in the analysis of record c. NETCO-SNAP-IN rack inserts are in their assumed locations d. Blocking devices are in their assumed locations A generic Westinghouse study (Reference 6.6) investigated the adequacy of assuming a distance of one cell pitch for neutronic isolation. | ||
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==4.0 REGULATORY EVALUATION== | ==4.0 REGULATORY EVALUATION== | ||
4.1 Applicable Regulatory Requirements The regulations in 10 CFR 50.36(c)(2)(ii)(B), Limiting conditions for operation, state: 22 Enclosure Description and Assessment of Proposed License Amendment Criterion | |||
Regulatory Requirements The regulations in 10 CFR 50.36(c)(2)(ii)(B), Limiting conditions for operation, state: 22 Enclosure Description and Assessment of Proposed License Amendment Criterion | |||
: 2. A process variable, design feature, or operating restriction that is an initial condition of a design basis accident or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.Technical Specification (TS) 3. 7.17 currently meets this requirement and will continue to meet this requirement after the proposed changes are approved and implemented. | : 2. A process variable, design feature, or operating restriction that is an initial condition of a design basis accident or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.Technical Specification (TS) 3. 7.17 currently meets this requirement and will continue to meet this requirement after the proposed changes are approved and implemented. | ||
The regulations in 10 CFR 50.36(c)(4), Design features, state: Design features to be included are those features of the facility such as materials of construction and geometric arrangements, which, if altered or modified, would have a significant effect on safety and are not covered in categories described in paragraphs (c)(1), (2), and (3) of this section.TS 4.3.1 currently meets this requirement and will continue to meet this requirement after the proposed changes are approved and implemented. | The regulations in 10 CFR 50.36(c)(4), Design features, state: Design features to be included are those features of the facility such as materials of construction and geometric arrangements, which, if altered or modified, would have a significant effect on safety and are not covered in categories described in paragraphs (c)(1), (2), and (3) of this section.TS 4.3.1 currently meets this requirement and will continue to meet this requirement after the proposed changes are approved and implemented. | ||
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TS 3. 7.17 currently meets this requirement and will continue to meet this requirement after the proposed changes are approved and implemented. | TS 3. 7.17 currently meets this requirement and will continue to meet this requirement after the proposed changes are approved and implemented. | ||
The guidance in DSS-ISG-2010-01 is to be used by NRC staff to review nuclear criticality safety analyses for the storage of new and spent nuclear fuel as they apply to applications for license amendments submitted after September 29, 2011.This license amendment request and WCAP-18030-P were developed using the guidance of DSS-ISG-2010-01. | The guidance in DSS-ISG-2010-01 is to be used by NRC staff to review nuclear criticality safety analyses for the storage of new and spent nuclear fuel as they apply to applications for license amendments submitted after September 29, 2011.This license amendment request and WCAP-18030-P were developed using the guidance of DSS-ISG-2010-01. | ||
4.2 Precedent The analysis methodology for the site-specific criticality analysis employs the PARAGON code, which is approved for use by the NRC (Reference 6.4).4.:3 Significant Hazards Consideration As required by 10 CFR 50.91(a), Notice for Public Comment, an analysis of the issue of no significant hazards consideration using the standards in 10 CFR 50.92, Issuance of Amendment, is presented below: 1. Does the proposed amendment involve a significant increase in the probability or consequences of an accident previously evaluated? | |||
The analysis methodology for the site-specific criticality analysis employs the PARAGON code, which is approved for use by the NRC (Reference 6.4).4.:3 Significant Hazards Consideration As required by 10 CFR 50.91(a), Notice for Public Comment, an analysis of the issue of no significant hazards consideration using the standards in 10 CFR 50.92, Issuance of Amendment, is presented below: 1. Does the proposed amendment involve a significant increase in the probability or consequences of an accident previously evaluated? | |||
Response: | Response: | ||
No.24 Enclosure Description and Assessment of Proposed License Amendment The proposed amendment would modify the Palo Verde Nuclear Generating Station (PVNGS) Technical Specifications (TS) to incorporate the results of an updated criticality safety analysis for both new fuel and spent fuel storage. The revised criticality safety analysis provides an updated methodology that allows credit for neutron absorbing NETCO-SNAP-IN rack inserts and corrects non-conservative input assumptions in the previous criticality safety analysis.The proposed amendment does not change or modify the fuel, fuel handling processes, number of fuel assemblies that may be stored in the spent fuel pool (SFP), decay heat generation rate, or the SFP cooling and cleanup system. The proposed amendment was evaluated for impact on the following previously evaluated events and accidents: | No.24 Enclosure Description and Assessment of Proposed License Amendment The proposed amendment would modify the Palo Verde Nuclear Generating Station (PVNGS) Technical Specifications (TS) to incorporate the results of an updated criticality safety analysis for both new fuel and spent fuel storage. The revised criticality safety analysis provides an updated methodology that allows credit for neutron absorbing NETCO-SNAP-IN rack inserts and corrects non-conservative input assumptions in the previous criticality safety analysis.The proposed amendment does not change or modify the fuel, fuel handling processes, number of fuel assemblies that may be stored in the spent fuel pool (SFP), decay heat generation rate, or the SFP cooling and cleanup system. The proposed amendment was evaluated for impact on the following previously evaluated events and accidents: | ||
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==6.0 REFERENCES== | ==6.0 REFERENCES== | ||
6.1 Criticality Safety Analysis for Palo Verde Nuclear Generating Station Units 1, 2, and 3 (Proprietary), WCAP-1 8030-P, Revision 0, September 2015.6.2 Staff Guidance Regarding the Nuclear Criticality Safety Analysis for Spent Fuel Pools, DSS-ISG-201 0-01, Revision 0, Nuclear Regulatory Commission Division of Safety Systems, Rockville, MD, September 29, 2011. (ML1 10620086)6.3 Scale: A Comprehensive Modeling and Simulation Suite for Nuclear Safety Analysis and Design, ORNL/TM-2005/39, Version 6.1, Oak Ridge National Laboratory, Oak Ridge, TN, June 2011.6.4 M. Ouisloumen, H. Huria, et al, Qualification of the Two-Dimensional Transport Code PARAGON, WCAP-16045-P-A, Revision 0, Westinghouse Electric Company LLC, Monroeville, PA, August 2004.28 Enclosure Description and Assessment of Proposed License Amendment 6.5 C. V. Parks, et al, Review and Prioritization of Technical Issues Related to Bumup Credit for LWR Fuel, NUREG/CR-6665, Oak Ridge National Laboratory, Oak Ridge, TN, February 2000.6,6 Letter, J. Gresham (WEC) to NRC, Responses to Requests for Additional Information from the Review of WCAP- 1 7483-PA/WCAP-1 7483-NP, Revision 0, 'Westinghouse Methodology for Spent Fuel Pool and New Fuel Rack Criticality Safety Analysis,' | |||
Safety Analysis for Palo Verde Nuclear Generating Station Units 1, 2, and 3 (Proprietary), WCAP-1 8030-P, Revision 0, September 2015.6.2 Staff Guidance Regarding the Nuclear Criticality Safety Analysis for Spent Fuel Pools, DSS-ISG-201 0-01, Revision 0, Nuclear Regulatory Commission Division of Safety Systems, Rockville, MD, September 29, 2011. (ML1 10620086)6.3 Scale: A Comprehensive Modeling and Simulation Suite for Nuclear Safety Analysis and Design, ORNL/TM-2005/39, Version 6.1, Oak Ridge National Laboratory, Oak Ridge, TN, June 2011.6.4 M. Ouisloumen, H. Huria, et al, Qualification of the Two-Dimensional Transport Code PARAGON, WCAP-16045-P-A, Revision 0, Westinghouse Electric Company LLC, Monroeville, PA, August 2004.28 Enclosure Description and Assessment of Proposed License Amendment 6.5 C. V. Parks, et al, Review and Prioritization of Technical Issues Related to Bumup Credit for LWR Fuel, NUREG/CR-6665, Oak Ridge National Laboratory, Oak Ridge, TN, February 2000.6,6 Letter, J. Gresham (WEC) to NRC, Responses to Requests for Additional Information from the Review of WCAP- 1 7483-PA/WCAP-1 7483-NP, Revision 0, 'Westinghouse Methodology for Spent Fuel Pool and New Fuel Rack Criticality Safety Analysis,' | |||
LTR-NRC-15-60, dated July 20, 2015.29 Enclosure ,Description and Assessment of Proposed License Amendment ATTACHMENT 1 Marked-up Technical Specifications Pages (Pages Provided for Before and After SEP Transition) 3.7.17-1-3.7.17-2 3.7.17-3 3.7.17-4 4.0-2 4.0-3 5.5-19 SBefore SFP transitionI Spent Fuel Assembly Storage 3.7.17 3.7 PLANT SYSTEMS 3.7.17 Spent Fuel Assembly Storage LCO 3.7.17 APPLICABILITY: | LTR-NRC-15-60, dated July 20, 2015.29 Enclosure ,Description and Assessment of Proposed License Amendment ATTACHMENT 1 Marked-up Technical Specifications Pages (Pages Provided for Before and After SEP Transition) 3.7.17-1-3.7.17-2 3.7.17-3 3.7.17-4 4.0-2 4.0-3 5.5-19 SBefore SFP transitionI Spent Fuel Assembly Storage 3.7.17 3.7 PLANT SYSTEMS 3.7.17 Spent Fuel Assembly Storage LCO 3.7.17 APPLICABILITY: | ||
The combination of initial enrichment, burnup, and decay time of each fuel assembly stored in each of the four regions of the fuel storage pool shall be within the acceptable burnup domain for each region as shown in Figures 3.7.17-1, 3.7.17-2, or 3.7.17-3, and described in Specification 4.3.1.1.Whenever any fuel assembly is stored in the fuel storage pool.ACTIONS__________________________ | The combination of initial enrichment, burnup, and decay time of each fuel assembly stored in each of the four regions of the fuel storage pool shall be within the acceptable burnup domain for each region as shown in Figures 3.7.17-1, 3.7.17-2, or 3.7.17-3, and described in Specification 4.3.1.1.Whenever any fuel assembly is stored in the fuel storage pool.ACTIONS__________________________ | ||
Revision as of 06:55, 7 May 2019
| ML15336A087 | |
| Person / Time | |
|---|---|
| Site: | Palo Verde  |
| Issue date: | 11/25/2015 |
| From: | Lacal M L Arizona Public Service Co |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML15336A251 | List: |
| References | |
| 102-07149-MLL/TNW | |
| Download: ML15336A087 (96) | |
Text
Enclosure Attachment 8 contains PROPRIETARY information to be withheld under 10 CFR 2.390 10 CFR 50.90 Maria L. Lacal Vice President, Nuclear s~ale Regulatory
&OversightVed Nuclear Generating Station P.O. Box 52034 Phoenix, AZ 85072/° Mail Station 7605 Tel 623.393.6491 102-07149-M LL/TNW November 25, 2015 U. S. Nuclear Regulatory Commission ATI-N: Document Control Desk Washington, DC 20555-0001
Dear Sirs:
Subject:
Palo Verde Nuclear Generating Station (PVNGS)Units 1, 2, and 3 Docket Nos. STN 50-528, 50-529, and 50-530 License Amendment Request to Revise Technical Specifications to Incorporate Updated Criticality Safety Analysis In accordance with the provisions of Section 50.90 of Title 10 of the Code of Federal Regulations (10 CFR), Arizona Public Service Company (APS) is submitting a request for a license amendment to revise the Technical Specifications (TS) for Palo Verde Nuclear Generating Station Units 1, 2, and 3. The proposed amendment would modify TS requirements to incorporate the results of an updated criticality safety analysis for both new and spent fuel storage.The enclosure to this letter provides a description and assessment of the proposed changes including a technical evaluation, a regulatory evaluation, a significant hazards consideration, and an environmental consideration.
The enclosure also contains eight attachments.
Attachment 1 provides the marked-up existing TS pages. Attachment 2 provides the revised (clean) TS pages. Attachment 3 provides the marked-up TS Bases pages to show the proposed changes.This submittal contains new regulatory commitments (as defined by NEI 99-04, Guidelines for Managing NRC Commitment Changes, Revision 0) to be implemented, which are identified in Attachment
- 4. Attachment 5 provides a non-proprietary version of the criticality safety analysis.
Attachment 6 provides a material qualification report for NETCO-SNAP-IN neutron absorbing spent fuel pool rack inserts.Attachment 7 is an affidavit signed by Westinghouse Electric Company LLC that sets forth the basis on which the proprietary information in Attachment 8 may be withheld from public disclosure by the Commission and addresses with specificity the considerations listed in 10 CFR 2.390(b)(4).
Correspondence with respect to the proprietary aspects of Attachment 8 A member of the STAR!S (Strategic Teaming and Resource Sharing) Alliance Callaway *Diablo Canyon " Palo Verde
- Wolf Creek Attachment 8 transmitted herewith contains PROPRIETARY information.
When separated from Attachment 8, this transmittal document is decontrolled.
102-07149-M LL/TNW A-TEN: Document Control Desk U. S. Nuclear Regulatory Commission LAR to Incorporate Updated Criticality Safety Analysis in TS Page 2 or the supporting Westinghouse affidavit should reference Westinghouse letter number CAW-15-4271 and be addressed to James A. Gresham, Manager, Regulatory Compliance, Westinghouse Electric Company, 1000 Westinghouse Drive, Building 3 Suite 310, Cranberry Township, Pennsylvania 16066.Attachment 8 is the Criticality Safety Analysis for Palo Verde Nuclear Generating Station Units 1, 2, and 3, WCAP-18030-P (Proprietary), which contains information proprietary to Westinghouse Electric Company LLC.A public pre-submittal meeting was held with the NRC on May 11, 2015 (Agency Document Access and Management System [ADAMS] accession number ML15140A314) to discuss the criticality safety analysis performed in support of this license amendment request. A follow-up public conference call to address action items from the May 11, 2015, pre-submittal meeting was held on September 1, 2015 (ADAMS accession number ML15286A028).
In accordance with the PVNGS Quality Assurance Program, the Plant Review Board and the Offsite Safety Review Committee have reviewed and approved the proposed amendment.
By copy of this letter, this license amendment request is being forwarded to the Arizona Radiation Regulatory Agency in accordance with 10 CFR 50.91(b)(1).
APS requests approval of the proposed license amendment by October 1, 2017, and will implement the TS amendment within 90 days following NRC approval.
This request is necessary to complete the Spent Fuel Pool Transition Plan by the end of 2019.Should you have any questions concerning the content of this letter, please contact Thomas Weber, Department Leader, Nuclear Regulatory Affairs, at (623) 393-5764.I declare under penalty of perjury that the foregoing is true and correct.Executed on z 2-< .(Date)Sincerely, M LL/TN W/J R/af
Enclosure:
Description and Assessment of Proposed License Amendment cc: M. L. Dapas NRC Region IV Regional Administrator M. M. Watford NRC NRR Project Manager for PVNGS L. J. KIoss NRC NRR Project Manager C. A. Peabody NRC Senior Resident Inspector for PVNGS A. V. Godwin Arizona Radiation Regulatory'Agency (ARRA)T. Morales Arizona Radiation Regulatory Agency (ARRA)
Enclosure Description and Assessment of Proposed License Amendment TABLE OF CONTENTS 1.0
SUMMARY
DESCRIPTION 2.0 DETAILED DESCRIPTION 2.1 Proposed Changes to the Technical Specifications 2.2 Need for Proposed Changes 3.0 TECHNICAL EVALUATION 3.1 Spent Fuel Pool Analysis 3.2 New Fuel Storage and Fuel Transfer Equipment Analysis 3.3 Spent Fuel Pool Transition Plan
4.0 REGULATORY EVALUATION
4.1 Applicable Regulatory Requirements 4.2 Precedent 4.3 Significant Hazards Consideration 4.4 Conclusion 5.0 ENVIRONMENTAL CONSIDERATION
6.0 REFERENCES
ATTACHMENTS
- 1. Marked-up Technical Specifications Pages 2. Revised Technical Specifications Pages (Clean Copy)3. Marked-up Technical Specifications Bases Pages 4. List of Regulatory Commitments
- 5. Criticality Safety Analysis for Palo Verde Nuclear Generating Station Units 1, 2, and 3 (Non-proprietary), WCAP-1 8030-NP, Revision 0, September 2015 6. Material Qualification Report of MAXUS for Spent Fuel Storage, NET-300047-07 Rev 1, November 2015 7. Westinghouse Application for Withholding Proprietary Information from Public Disclosure, CAW-15-4271, September 3, 2015 8. Criticality Safety Analysis for Palo Verde Nuclear Generating Station Units 1, 2, and 3 (Proprietary), WCAP-18030-P, Revision 0, September 2015 LIST OF ACRONYMS ANP AREVA PVNGS Lead Test Assembly Combustion Engineering 16x16 Fuel APS Arizona Public Service Company ENDF Evaluated Nuclear Data File FHE Fuel Handling Equipment IFSR Intermediate Fuel Storage Rack LAR License Amendment Request LER Licensee Event Report NFS New Fuel Storage NGF Combustion Engineering 16x16 Next Generation Fuel PVNGS Palo Verde Nuclear Generating Station SFP Spent Fuel Pool STD Standard Combustion Engineering 16x16 Fuel TS Technical Specification(s)
VAP Value Added Pellet Combustion Engineering 16x16 Fuel ii Enclosure Description and Assessment of Proposed License Amendment 1.0
SUMMARY
DESCRIPTION The proposed amendment would revise Palo Verde Nuclear Generating Station (PVNGS)Renewed Operating License Nos. NPF-41, NPF-51, and NPF-74 to amend the Technical Specifications (TS) to incorporate the results of the updated criticality safety analysis, WCAP-1 8030-P (Reference 6.1). The proposed amendment will correct a non-conservative TS regarding NRC approved License Amendment Number 125, which describes the current licensing bases for the criticality safety analysis for PVNGS. This is further discussed in Section 2.2 of this proposed amendment.
This enclosure provides a description and assessment of the proposed changes including a technical evaluation, a regulatory evaluation, a significant hazards consideration, and an environmental consideration.
The enclosure also contains eight attachments.
Attachment 1 provides the marked-up existing TS pages. Attachment 2 provides the revised (clean) TS pages. Attachment 3 provides the marked-up TS Bases pages to show the proposed changes.This submittal contains new regulatory commitments (as defined by NEI 99-04, Guidelines for Managing NRC Commitment Changes, Revision 0) to be implemented, which are identified in Attachment
- 4. Attachment 5 provides a non-proprietary version of the criticality safety analysis.Attachment 6 provides a material qualification report for NETCO-SNAP-IN neutron absorbing spent fuel pool rack inserts.Attachment 7 is an affidavit signed by Westinghouse Electric Company LLC that sets forth the basis on which the proprietary information in Attachment 8 may be withheld from public disclosure by the Commission and addresses with specificity the considerations listed in 10 CFR 2.390(b)(4).
Attachment 8 is the Criticality Safety Analysis for Palo Verde Nuclear Generating Station Units 1, 2, and 3, WCAP-1 8030-P (Proprietary), which contains information proprietary to Westinghouse Electric Company LLC.2.0 DETAILED DESCRIPTION 2.1 Proposed Changes to the Technical Specifications The following specific TS changes are proposed as part of the updated criticality safety analysis.Marked-up TS pages are provided in Attachment 1 and revised (clean) TS pages are provided in Attachment 2.TS 3.7.17, Spent Fuel Assembly Storage-Revise the LCO statement and surveillance requirement statement to reflect the updated spent fuel assembly storage requirements resulting from WCAP-18030-P
-Add Tables 3.7.17-1 through 3.7.17-5 to define the Fuel Regions-Replace Figure 3.7.17-1 to define the Allowable Storage Arrays-Delete Figures 3.7.17-2 and 3.7.17-3 (information is replaced by Tables 3.7.17-1 through 3.7.17-5)I Enclosure Description and Assessment of Proposed License Amendment*TS 4.3.1, Criticality
-Change TS 4.3.1.1 .a to read "4.65 weight percent" instead of "4.80 weight percent"-Change TS 4.3.1.1.c to read "1460 ppm" [parts per million] instead of "900 ppm"-Change TS 4.3.1.1.e to refer to Fuel Regions 1 -6as shown in Tables 3.7.17-1 through 3.7.17-5-Delete TS 4.3.1.1.f through 4.3.1.1.h (Fuel Regions are defined in Tables 3.7.17-1 through 3.7.17-5)-Change TS 4.3.1 .2.a to read "4.65 weight percent" instead of "4.80 weight percent"-Change TS 4.3.1 .2.d to replace "A nominal 17 inch center to center..." with"A nominal 18 inch (east-west) and 31 inch (north-south) center-to-center..."*Add new program TS 5.5.21, Spent Fuel Storage Rack Neutron Absorber Monitoring Program (Proposed TS 5.5.20, Risk In formed Completion Time Program, was submitted on July 31, 2015 [Agency Document Access and Management System (ADAMS)accession number ML15218A300])
The TS Bases will also be revised for consistency with the proposed TS changes and with WCAP-1 8030-P. A markup of the TS Bases pages reflecting these changes is provided in Attachment 3 for information.
The proposed TS Bases changes will be implemented in accordance with TS 5.5.14, Technical Specifications (TS) Bases Control Program, at the same time that the TS changes in the approved license amendment request (LAR) are implemented.
2.2 Need for Proposed Changes In March of 2000, the NRC approved License Amendment Number 125, which describes the current licensing bases for the criticality safety analysis for PVNGS. That amendment increased the storage capacity of the spent fuel pools (SFPs) by allowing credit for soluble boron and decay time in the criticality safety analysis.
The amendment also increased the maximum radially averaged fuel enrichment from 4.3 weight percent U-235 to 4.8 weight percent U-235.The methodology that was the basis for that amendment was analogous to that developed in WCAP-14416-P-A, Westinghouse Spent Fuel Rack Criticality Analysis Methodology, which was reviewed and approved by the NRC for use [NRC Letter, T. E. Collins (NRC) to T. Greene (WOG), "Acceptance for Referencing of Licensing Topical Report WCAP-14416-P, Westinghouse Spent Fuel Rack Methodology (TAC No. M93254)," dated October 25, 1996]. In 2001 and 2004, Arizona Public Service Company (APS) submitted license amendment requests (LARs) to the NRC that would support replacement of the steam generators and authorize subsequent operation at an increased maximum power level of 3990 Megawatts thermal (a 2.94 percent increase).
The NRC approved the amendments for Unit 2 in 2003 (ADAMS accession number ML032720538) and for Units I and 3 in 2005 (ADAMS accession number ML0531 30275).In May of 2013, APS submitted Licensee Event Report (LER) 2013-001-00 (ADAMS accession number ML13133A002), which reported that certain impacts to the SFP criticality safety analysis approved in License Amendment 125 had not been considered by APS during the increase in the maximum power level to 3990 MWt. One of the corrective actions in the LER was to revise the SFP criticality safety analysis using updated methodology and input parameters and to 2 Enciosure Description and Assessment of Proposed License Amendment submit a LAR to correct the non-conservative TS. The criticality safety analysis methodology included in this LAR is based upon the most recent NRC approved guidance of Staff Guidance Regarding the Nuclear Criticality Safety Analysis for Spent Fuel Pools, DSS-ISG-201 0-01 (Reference 6.2) and satisfies the actions stipulated in the PVNGS Corrective Action Program and PVNGS LER 201 3-001-00.3.0 TECHNICAL EVALUATION This LAR documents an updated criticality safety analysis for the PVNGS SFPs, new fuel storage (NFS) racks, interim fuel storage rack (IFSR) and fuel handling equipment (FHE).Attachment 8 is the plant-specific Westinghouse WCAP-18030-P, Revision 0, Criticality Safety Analysis for Palo Verde Nuclear Generating Station Units 1, 2 and 3. A plant-specific NETCO-SNAP-IN 0 material qualification report is also included (Attachment
- 6) since the LAR credits the presence of neutron poisons in the NETCO-SNAP-IN neutron absorber inserts. The main body of the LAR includes descriptions and summary evaluations, while the attached WCAP and NETCO report provide additional details on topics that include computer codes, fuel design history, depletion analysis, criticality analysis, as well as the interface, normal, and accident conditions for PVNGS.The change to TS 4.3.1 .2.d regarding NFS rack spacing is proposed to more accurately reflect the NFS as-built drawings and the existing NFS criticality safety analysis of record. The text of TS 4.3.1.2 states "The new fuel storage racks are designed and shall be maintained with...", which refers to physical dimensions.
Therefore, TS 4.3.1 .2'.d must accurately reflect the as-built dimensions that must be maintained throughout the life of the plant. The racks have a nominal 18-inch center-to-center pitch on the short axis (east-west) and a 31-inch center-to-center pitch on the long axis (north-south).
According to both the existing NFS and updated criticality safety analyses, this rack design maintains keff <0.95 during all normal and accident conditions.
3.1 Spent Fuel Pool Analysis Design Approach The existing SEP storage racks are evaluated for the placement of fuel within the storage arrays described in the proposed TS changes. Credit is taken for the negative reactivity associated with burnup and post-irradiation cooling time (decay time). Additionally, some SEP storage arrays credit the presence of the neutron poison in the NETCO-SNAP-IN inserts. Finally, credit is taken for the presence of soluble boron in the SFP for specific conditions.
Compliance for the SFP is demonstrated by establishing limits on the minimum allowable burnup as a function of initial enrichment and decay time for each fuel storage array. A conservative combination of best estimate and bounding values has been selected to model the fuel in the analysis to ensure that fuel represented by the proposed TS is less reactive than the fuel modeled in the analysis.
Therefore, burnup limits will conservatively bound fuel to be stored in the SFP.3 Enclosure Description and Assessment of Proposed License Amendment Acceptance Criteria The objective of the SEP criticality safety analysis is to ensure that the SEP operates within the bounds of 10 CFR 50.68(b)(4):
- If no credit for soluble boron is taken, the keff of the spent fuel storage racks loaded with fuel of the maximum fuel assembly reactivity must not exceed 0.95, at a 95 percent probability, 95 percent confidence level, if flooded with unborated water.* If credit is taken for soluble boron, the keff of the spent fuel storage racks loaded with fuel of the maximum fuel assembly reactivity must not exceed 0.95, at a 95 percent probability, 95 percent confidence level, if flooded with borated water, and the keff must remain below 1 .0 (subcritical), at a 95 percent probability, 95 percent confidence level, if flooded with unborated water.Computer Codes The SEP criticality safety analysis methodology employs the following computer codes and cross-section libraries:
- SCALE 6.1.2 (Reference 6.3) with the 238-group cross-section library based on Evaluated Nuclear Data File (ENDF)/B-VII.
- The two-dimensional transport lattice code PARAGON (Reference 6.4) and its 70-group cross-section library based on ENDF/B-VI.3 PARAGON is used for simulation of in-reactor fuel assembly depletion to generate isotopics for burnup credit applications in the SEP. PARAGON is generically approved for depletion calculations (Reference 6.4) and has been chosen for this spent fuel criticality safety analysis because it has all the attributes needed for burnup credit applications.
There are no Safety Evaluation Report limitations for the use of PARAGON in UO 2 criticality analyses.Additional discussion of the computer codes is provided in Section 2.3 of Attachment 8.Code Validation Process The validation of the ENDF/B-VII library with the SCALE 6.1.2 CSAS5 module is documented in Appendix A of Attachment
- 8. The code validation shows that SCALE 6.1.2 is an accurate tool for calculation of kefffor the applications in this LAR. The benchmark calculations utilize the same computer platform and cross-section libraries that are used for the design basis calculations.
3.1.1 Fuel and Fuel Storage Descriptions Four fuel designs were considered for the criticality safety analysis:* Standard Combustion Engineering 16x16 (STD)* Value Added Pellet Combustion Engineering 16x16 (VAP)* Next Generation Fuel Combustion Engineering 16x16 (NGF)* AREVA PVNGS Lead Test Assembly Combustion Engineering 16x16 (ANP)4 Enclosure Description and Assessment of Proposed License Amendment Section 3.1.2 of Attachment 8 discusses the non-mechanical fuel features which are important to criticality safety and how they impact the number of distinct fuel designs to be considered in the analysis.PVNGS has three SFPs, one in each unit, that are identical in layout. Each SFP contains a single rack design and each SEP is surrounded by a concrete wall with a stainless steel liner.The presence of neither the SEP concrete wall nor liner is credited in the criticality safety analysis.
All storage arrays are conservatively assumed to be radially infinite.The SEP storage racks are made up of individual modules, each of which is an array of fuel storage cells. The storage racks are comprised of 17 modules: twelve 8x9 arrays, four 8x12 arrays, and one 9x9 array. The storage racks are stainless steel honeycomb structures with rectangular fuel storage cells compatible with fuel assembly materials and the spent fuel borated water environment.
The fuel assembly spacing of a nominal 9.5 inches center-to-center distance between adjacent storage cell locations is a minimum value after allowances are made for rack fabrication tolerances and predicted deflections resulting from a safe shutdown earthquake.
The storage racks are designed to maximize the number of storage cells available (minimize storage cell pitch) to be used in a checkerboard pattern of fresh (i.e., new) fuel and empty storage locations.
Trashcans in the SEP store various non-fissile materials, such as discarded control element assemblies and in-core instrumentation tubes, filters, and reconstitution materials.
The SEP racks contain a stainless steel L-insert in every other cell location as shown in PVNGS Updated Final Safety Analysis Report (UFSAR) Figure 9.1-5 to help center the new fuel assemblies within this space. The stainless steel L-inserts are offset 11/16 inch from the cell wall. This reduces the positional uncertainty or the eccentric loading positions of fuel in the racks. There is not currently a surveillance program for the stainless steel L-inserts and one will not be included as part of this submittal for the following reasons:* The industry has a large body of operating experience with stainless steel in SEP environments in which nothing suggests the stainless steel L-inserts would become incapable of performing their function.* The stainless steel L-inserts are analogous to the SEP racks in their function and there is no precedent for requiring a surveillance program of the racks themselves.
A NETCO-SNAP-IN rack insert will not be installed in a cell that contains a stainless steel L-insert.
The NETCO-SNAP-IN rack inserts are described in Attachment 6.3.1.2 Depletion Analysis The depletion analysis is a vital part of any SEP criticality safety analysis which uses burnup credit. The isotopic inventory of the fuel as a function of burnup is generated through the depletion analysis, thus the inputs used need to be carefully considered.
Section 4 of Attachment 8 describes the methods used to determine the appropriate inputs for the generation of isotopic number densities to conservatively bound fuel depletion and storage.Some of the salient points regarding the approach to the depletion analysis include: 5 Enclosure Description and Assessment of Proposed License Amendment* The isotopic number densities generated by the fuel depletion calculations are differentiated by fuel enrichment and decay time after discharge.
The fuel has isotopic number densities which are calculated at enrichments of 3.0, 4.0, and 5.0 weight percent U-235 and decay times of 0, 5, 10, 15, and 20 years.* The soluble boron concentration in the reactor during operation impacts the reactivity of fuel being discharged to the SEP. Boron is a strong thermal neutron absorber and its presence hardens the neutron energy spectrum in the core, creating more plutonium.
It is important to account for the presence of soluble boron during reactor operation to ensure this impact is adequately accounted for in the isotopic generation.
- The fuel temperature during operation impacts the reactivity of fuel being discharged to the SFP. Increasing fuel temperature increases resonance absorption in U-238 due to Doppler broadening which leads to increased plutonium production, increasing the reactivity of the fuel.* The limiting distributed axial burnup profiles are used with the uniform axial burnup profile to calculate the burnup limits.* The limiting axial moderator temperature profiles are used with axially distributed and uniform burnup profiles to calculate the isotopics used in generating the burnup limits.Selecting an appropriate moderator temperature profile is important as it impacts the moderator density and the neutron spectrum during depletion, as discussed in NUREG/CR-6665 (Reference 6.5). An appropriate moderator temperature ensures the impact of moderator density on the neutron spectral effects is bounded, conservatively biasing the isotopic inventory of the fuel.* Burnable absorber usage has been considered for the analysis and conservative assumptions have been utilized to bound the effects of burnable absorbers on fuel isotopics.
The burnable absorbers that have been used include both discrete and integral burnable absorbers.
- The PVNGS fuel management strategy uses radial enrichment zoning to control fuel rod power peaking. Individual assemblies may contain two or three different fuel rod enrichments which are used to control peaking factors. A study was performed to determine the reactivity impact of operating with radial enrichment zoning instead of uniform radial zoning.* All four of the different fuel designs listed in Section 3.1.1 and the conditions in which those designs were operated, or are planned to be operated, were considered in the depletion analysis.
It became clear that the NGF design would be limiting throughout life.Therefore, the NGF design was used to develop the isotopics used in the spent fuel reactivity calculations.
- The parameters used in the final depletion calculations include core operation parameters, fuel assembly dimensions, axial burnup profiles, and moderator temperature profiles.3.1.3 Spent Fuel Pool Criticality Analysis KENO is the criticality code used to determine the absolute reactivity of burned and fresh fuel assemblies loaded in storage arrays. The dimensions and tolerances of the design basis fuel assembly and the fuel storage racks are the basis for the KENO models used to determine the burnup requirements for each fuel storage array, and to confirm the safe operation of the SEP 6 Enclosure Description and Assessment of Proposed License Amendment under normal and accident conditions.
The trashcan characteristics are also modeled in the criticality analysis.Differences between fuel types include changes in fuel rod dimensions, such as pellet and cladding dimensions, and structural components, such as grid material and volumes. Each of the fuel types which have been, or are planned to be, operated at the plant were considered.
The bounding fuel assembly design for the analysis has been determined as described in Section 4.3 of Attachment 8.Burnup Limit Generation To ensure safe operation of the PVNGS SFPs, the analysis defines fuel storage arrays which dictate where assemblies can be placed in the SFPs based on enrichment (weight percent U-235), average burnup (GWd/MTU), and decay time (years) since discharge.
Each assembly in the reactor core depletes under slightly different conditions and can have a different reactivity at the same burnup. This is accounted for in the analysis by using a combination of depletion parameters that together produce a bounding isotopic inventory throughout life. Additionally, while fuel manufacturing is a very tightly controlled process, assemblies are not identical.
Reactivity margin is added to the KENO reactivity calculations for the generation of burnup limits to account for manufacturing deviations.
Assembly storage is controlled by defining allowable storage arrays. An array can only be populated by assemblies of the fuel region defined in the array definition or a lower reactivity fuel region. Fuel regions are defined by assembly burnup, initial enrichment, and decay time.Reactivity biases are known variations between the real and analyzed system, and their reactivity impact is added directly to the calculated keff. Uncertainties are random dispersions around a nominal, measured quantity.
Their impact is added to the calculated keff as the square root of the sum of the squares of the uncertainties.
The following biases and uncertainties are accounted for in the analysis.
A detailed discussion of biases and uncertainties is provided in Section 5.2.3 of Attachment 8.* Reactivity effect of manufacturing tolerances
- Burnup measurement uncertainty
- Depletion uncertainty
- Fission product and minor actinide worth bias* An operational uncertainty of 0.002 Ak is an additional conservativism which is added to the conservatism inherent in the specific power histories from reactor operation* Eccentric fuel assembly positioning
- Uncertainty in the predictive capability of SCALE 6.1.2 and the associated cross-section library* SFP temperature bias within the allowable operating range* Borated and unborated biases and uncertainties 7
Enclosure Description and Assessment of Proposed License Amendment Interface Modelingq Interfaces are the locations where there is a change in either the storage racks or the storage requirements of the fuel in question.
At PVNGS, each SFP has a single storage rack design.Therefore, the only interfaces that exist are those between arrays within the single storage rack design and the only interface conditions that need to be addressed in the analysis are those between different fuel storage arrays. Additional details are provided in Section 5.3 of Attachment 8.Normal Conditions Considered in the Criticality Safety Analysis There are five major types of normal conditions beyond the storage of fuel assemblies that are addressed in the criticality safety analysis.Type 1 conditions involve placement of components in or near the intact fuel assemblies while normally stored in the storage racks. This also includes removal and reinsertion of these components into the fuel when stored in the rack positions using specifically designed tooling.Examples include control element assemblies and guide tube inserts, such as in-core instrumentation tubes. The calculation results show that any components designed to be inserted into an assembly may be stored in a fuel assembly guide tube in the SFP.Type 2 conditions involve evolutions where the fuel assembly is removed from the normal storage rack location for a specific procedure and returned to an allowable cell after completion of the procedure, such as fuel assembly cleaning, inspection, reconstitution, or sipping. These are bounded by the criticality analysis.Fuel assembly reconstitution is a normal condition defined as either pulling damaged fuel pins out of an assembly and reinserting intact pins with less reactivity than the damaged pin, or as removing undamaged pins from a damaged assembly for insertion in a new assembly.Damaged pins will be replaced with stainless steel pins or natural uranium pins. Additional information is provided in Section 5.4.2 of Attachment 8.Type 3 conditions involve inserting components that are not intact fuel assemblies into the fuel storage rack cells. Examples include failed fuel rod baskets and miscellaneous maintenance equipment.
Any components that do not contain fissile materials can replace a fuel assembly of any fuel region in one of the approved storage configurations.
Type 4 conditions include temporary installation of non-fissile components on the rack periphery.
Analyses of the storage arrays contained in the criticality analysis assume an infinite array of storage cells. This assumption bounds the installation of any non-fissile components on the periphery of racks.Type 5 conditions involve miscellaneous conditions that do not fit into the first four normal condition types. Examples include usage of fuel handling tools for their intended purpose, miscellaneous debris under the storage racks, and damaged storage cells.Section 5.4 of Attachment 8 provides further details about normal conditions within the SEP.8 Enclosure Description and Assessment of Proposed License Amendment Soluble Boron Credit In accordance with 10 CFR 50.68, the criticality safety analysis ensures that the maximum calculated ke., including all biases and uncertainties, meet the kef limit of less than 1.0 (subcritical) if flooded with unborated water at a 95 percent probability, 95 percent confidence level. Additionally, the criticality safety analysis demonstrates that if the SFP is flooded with borated water, keff does not exceed 0.95, at a 95 percent probability, 95 percent confidence level.The minimum soluble boron concentration in the SEP to maintain keff < 0.95 for the limiting normal condition including biases, uncertainties, and administrative margin is 450 ppm. During normal operation, TS 3.7.15 requires a soluble boron concentration of> 2150 ppm, but the SFP boron concentration is maintained between 4000 and 4400 ppm in accordance with Technical Requirements Manual T3. 1.104, Borated Sources -Shutdown, and T3. 1.105, Borated Sources-Operating.
Consideration of Criticality Accidents in the SEP The following reactivity-increasing accidents are considered and the analysis results are provided in Section 5.6 of Attachment 8.SAssembly misload into the storage racks -this is the limiting accident which addresses both multiple assemblies being misloaded in series into unacceptable storage locations and the misload of a single assembly into an unacceptable storage location.
A multiple assembly misload is a hypothetical accident where assemblies are misloaded in series due to a common cause. A single assembly misload requires 1100 ppm of boron to maintain keff -< 0.95. A multiple assembly misload requires 1460 ppm of boron to maintain kef<0.9 5.* Spent fuel temperature outside operating range -the SEP is to be operated between 60°F and 180°F, but under accident conditions this temperature could be higher.* Dropped and misplaced fresh assembly -the analysis considers the dropping of the fuel assembly from the fuel handling machine during placement of the fuel assemblies in the racks. The dropped assembly could land horizontally on top of the other fuel assemblies in the rack. Additionally, the analysis considers the possibility to misplace a fuel assembly in a location not intended for fuel.* Seismic event -the SEP racks are seismic category I, designed and built to withstand the maximum potential earthquake stresses in this geographic area. Section 5.6.4 of Attachment 8 provides additional details.o The spent fuel pool racks were originally designed to contain a 188 lb. neutron poison insert in every cell. These neutron poison inserts were never installed.
As the mass of the original design is greater than the mass of the NETCO-SNAP-IN, the original analysis bounds the proposed change.* Inadvertent removal of a NETCO-SNAP-IN rack insert -this is a potential reactivity-increasing accident added by the incorporation of NETCO-SNAP-IN rack inserts. The absence of an insert will cause a reactivity increase due to the loss of neutron absorbing material from the storage array.9 Enclosure Description and Assessment of Proposed License Amendment Fuel used to date at PVNGS has an initial radially averaged enrichment of < 4.55 weight percent. Limiting the maximum radially averaged enrichment to 4.65 weight percent mitigates the consequences of a multiple fuel assembly misload event without impacting operational flexibility.
There is no source of water within the fuel building that could reduce the boron concentration of the spent fuel pool from the value of 2150 ppm (Technical Specification LCO 3.7.15) to 1460 ppm. A fire in the fuel building at elevation 140-ft. is the limiting event for boron dilution and it bounds all normal, seismic, and pipe break scenarios.
The current boron dilution analysis demonstrates that the limiting boron dilution event, which is fighting a hypothetical fire on the 140-ft level of the fuel building, will reduce the boron concentration from the TS limit of 2150 ppm to 1900 ppm. This leaves adequate margin to the 1460 ppm credited by the SEP criticality safety analysis.3.1.4 NETCO-SNAP-IN Rack Inserts This proposed change would credit NETCO-SNAP-IN rack inserts for criticality control in individual SEP storage rack cells to ensure that the requirements of TS 3.7.17 and the associated WCAP-1 8030-P are maintained.
The NETCO-SNAP-IN rack inserts are credited in both the borated and unborated conditions.
The installation of the NETCO-SNAP-IN rack inserts will be controlled as a design change implemented under the provisions of 10 CFR 50.59, Changes, Tests and Experiments, from a structural, seismic, and thermal-hydraulic perspective.
Attachment 6 describes the NETCO-SNAP-IN rack inserts, including their manufacture, an engineering evaluation, and corrosion testing information.
Neutron Absorber Monitoring Program (TS 5.5.21)Arizona Public Service will institute a performance-based long-term surveillance program for the NETCO-SNAP-IN inserts based on manufacturer recommendations, current industry operating experience, NEI guidance, and NRC safety evaluations for other plants that are using neutron absorbing inserts. The long-term surveillance program will evolve as information from these sources changes and as the data from the PVNGS-specific inspections accumulate.
The surveillance program consists of periodic inspections of MAXUS material coupons from surveillance assemblies located in the SEPs and periodic inspection of full length inserts.Coupon Inspections Coupons will be selected from MAXUS production material, identical to the material used to manufacture the inserts, for periodic inspection.
Individual coupons will be subjected to pre-test and post-test characterizations.
As appropriate for each coupon type, coupon characterizations may include visual inspection, high resolution photography, neutron attenuation, stress relaxation, blister and pit characterizations, as well as measurement of thickness, length, width, dry weight, and density.A surveillance assembly to which surveillance coupons are attached, also referred to as a coupon tree, will be placed in the related SEP prior to the first installation campaign of NETCO-SNAP-IN inserts and will reside there to support the monitoring program. Periodically, coupons 10 Enclosure Description and Assessment of Proposed License Amendment will be removed and sent to a qualified laboratory for testing. The coupon trees in the related SFP will contain 48 general coupons, 24 galvanic couple coupons, and 24 bend coupons as described below. They will be situated near the center of the active fuel region to maximize exposure from the surrounding fuel. These coupons will be monitored for changes to their physical properties and for changes to their effective areal density or signs of corrosion, which could indicate neutron absorber material degradation.
The frequency for coupon removal and inspection is shown in Table 1.Table I -Frequency for Coupon Removal from Racks After 10 Years with Coupon Type First Ten Years Acpal efrac General 2 coupons every 2 years 2 coupons every 4 years Bend I coupon every 2 years 1 coupon every 4 years Galvanic couples -304L stainless I couple every 6 years Zircaloy 1 couple every 6 years Inconel 718 1 couple every 6 years General and Galvanic Coupons The general coupons in each SFP are designed to carry the largest number of performance indicators for the insert material.
They will be subject to pre-examination, post-examination, and acceptance testing in accordance with Table 2.The galvanic couple coupons are composed of a MAXUS material coupon placed in contact with Zircaloy, Inconel 718, or 304 stainless steel. Eight of each type will be used to produce a total of 24 galvanic couples per coupon tree assembly in each SEP that will be subject to the inspections listed in Table 2.Bend Coupons Once installed, the NETCO-SNAP-IN rack inserts assume a constant strain condition within the SEP storage rack cell. This compression leads to internal stresses, especially at the bend, that might make the rack inserts susceptible to stress corrosion cracking.
An examination of the literature on the subject indicates in general, that high-purity aluminum and low-strength aluminum alloys are not susceptible to stress corrosion cracking.
However, the surveillance bend coupons placed in the related SEP will be maintained under the same strain conditions as the inserts to provide an indication of unexpected crack phenomena.
These coupons will be held in capsules that compress them from their initial manufactured bend angle to an angle of approximately 90 degrees. Table 3 provides the inspection details.11 Enclosure Description and Assessment of Proposed License Amendment Over time, the MAXUS material is expected to release some of the strain built up during the installation process. The material has a metal matrix core made from 1000 series aluminum and boron carbide powder with an outer clad made from 5052 aluminum.
Existing literature for 1100 series aluminum shows a stress relaxation rate of 58 percent over a period of 20 years. Given that 1100 series aluminum is a softer metal than 5052 aluminum, this rate is considered conservative for the MAXUS material due to the 5052 cladding.
The acceptance criterion for stress relaxation is 60 percent over a 20-year period. This rate will be used when determining minimum retention force requirements for the inserts during installation that will still hold the inserts in place during a seismic event after relaxation has occurred.The bend coupon capsules will be removed from the coupon tree and sent to a qualified laboratory for testing where the coupons will be removed, thus relieving the strain on the coupons and allowing them to return to an angle greater than 90 degrees. The change in internal stress can be correlated to the change in bend angle the coupon forms once it is removed from the strained condition.
Deviation from the pre-characterized value will determine the amount of stress relaxation over the life of the coupon.The stress relaxation rate is not linear, rather it tends to follow a logarithmic pattern. Therefore, a more significant loss of stress is expected in the first few years of exposure, but the relaxation rate becomes asymptotic over a longer period of time.*Table 2 -General and Galvanic Coupon Characterizations TetPre- Post- Acceptance I Rejection Characterization Characterization Criteria Visual (high / Evidence of visual resolution digital indications of photo) performance inhibitors.
Dimension
/Min. thickness:
0.005 inch less than nominal thickness (excluding pit locations).
Thickness change: any change of +0.010 inch I-0.004 inch (excluding pit locations).
- Length change: any change of +/- 0.02 inch* Width change: any change of +/- 0.02 inch Density /Any change of +/- 5%Areal density g/cm 2 Boron-10 minimum loading 12 Enclosure Description and Assessment of Proposed License Amendment Weight loss as determined by dry change of +/- 5%weight Corrosion rate /< 0.05 mil/yr Anomaly characterization
/** To be determined at the time of analysis.* Length and width changes are not applicable for galvanic coupons.** At the presence of anomalies Table 3 -Bend Coupon Characterizations TetPre- Post- Acceptance
/Characterization Characterization Rejection Criteria Visual (high Evidence of visual resolution indications of digital photo) performance inhibitors.
Thickness
' V Min. thickness:
0.005 inch less than nominal thickness (excluding pit locations).
Thickness change: any change of +0.010 inch /-0.004 inch (excluding pit locations).
Bending V /Change in stress greater stress than a rate of 60% over 20 years *Weight loss as determined by dry change of +/- 5%weight Anomaly characterization
- To be determined at the time of analysis.Stress relaxation rate is not linear. Stress relaxation will be re-evaluated if 60% is exceeded.**At the presence of anomalies Full-Length Insert Inspections The combined effects of adequate clearance and infrequent fuel assembly movement will preclude significant wear of the rack insert. However, to verify NETCO-SNAP-IN material performance, a portion of the installed inserts will be subject to in-situ visual inspection and removal for detailed inspection of wear performance.
For in-situ inspections at the frequency described in Table 4, rack inserts will be visually inspected by camera (while remaining in the storage racks) to monitor for physical deformities such as bubbling, blistering, corrosion pitting, 13 Enclosure Description and Assessment of Proposed License Amendment cracking, or flaking. Special attention shall be paid to the development of edge or corner defects.A region of high duty spent fuel storage rack cell locations shall be identified for full insert removal and inspection.
These locations will be monitored for fuel insertion and removal events to ensure that their service bounds that of the general population of storage locations.
Once every 10 years, an insert will be fully removed from this region and will be inspected in accordance with Table 5. The thickness measurements at several locations along the full insert length will be compared with the as-built thickness measurements of the removed insert to verify it has sustained uniform wear over its service life. A visual inspection of the removed insert will also be performed.
Table 4 -Frequency for Full Insert Inspections Inspection Type First Ten Years Acpal efrac In-situ 2 inserts every 2 years 2 inserts every 4 years Removal 1 insert every 10 years Table 5 -Full Insert Removal Inspection Characterizations TetPre- Post- Acceptance I Characterization Characterization Rejection Criteria Visual (high of visual resolution indications of digital photo) performance inhibitors.
Thickness thickness:
0.005 inch less than nominal thickness (excluding pit locations).
Thickness change: any change of +0.010 inch I-0.004 inch (excluding pit locations).
Retention force Retention force less than 50 lbs 14 Enclosure Description and Assessment of Proposed License Amendment 3.1.5 Spent Fuel Pool Configuration Control (Human Performance Enhancements)
APS has a multi-tier defense-in-depth program to prevent and mitigate the severity of a scenario in which multiple fuel assemblies are located in the wrong storage locations.
Specific aspects of this program are described below.Control of Move Sheet Generation
- Detailed administrative procedures for the generation of move sheets and revision of move sheets* Training and qualification of individuals responsible for generation and revision of move sheets for use and implementation of the new TS proposed in this LAR.* Graphical representation of approved arrays in TS 3.7.17 to minimize the probability of misinterpretations
- In accordance with the plant special nuclear materials procedures, APS maintains a Region Specification Document for each SEP to aid the move sheet preparer and verifier in selecting and verifying proper placement of fuel assemblies.
This document tracks the following for every fuel assembly at PVNGS:-Fuel assembly initial enrichment
-Fuel assembly burnup-Limiting Fuel Region in which the fuel assembly can be stored* Every fuel move is checked against the availability of the space and the eligibility of fuel to be stored there* A member of management confirms that each move sheet was generated in accordance with PVNGS procedures.
There are at least three signatures on each move sheet sent to the field, including the move sheet preparer, the verifier, and management.
- A move sheet package is a change document that is used to specify and record changes to plant configuration as it relates to special nuclear material.
A move sheet contains as a minimum, the item to be moved, the "from" location, and the "to" location.
Any number of fuel assemblies can be moved using a single move sheet package.Control of Fuel Movement* The spent fuel handling machine is only operated using approved procedures
- All individuals operating the spent fuel handling machine and acting as independent verifiers are trained in their position, including training on industry operating experience pertaining to fuel misload events* Fuel is moved only as directed by approved move sheets* The correct location of the spent fuel handling machine is independently verified before a fuel move begins* The correct location of the spent fuel handing machine is independently verified before the fuel is placed in the SEP racks* Continuous communication is maintained between the fuel mover and verifier 15 Enclosure Description and Assessment of Proposed License Amendment* The fuel handlers visually confirm that fresh fuel is not placed in locations that are "face-adjacent" to other fresh fuel assemblies Use of Blockingq Devices The assumed limiting misload event at PVNGS involves placing a fuel assembly in a location that is required to be empty per TS 3.7.17. APS uses blocking devices to minimize the probability of a fuel assembly being placed in one of these limiting locations.
Each blockintg device meets the following criteria:* Physically configured to prevent insertion of a fuel assembly in a fuel storage location* Requires special tools to install or remove the blocking device from a storage location* The tool used to grapple a fuel assembly is physically incapable of grappling a blocking device* Designed to preclude falling into a storage location or becoming dislodged during normal operation* Will support the full load of a fuel assembly and the fuel assembly grappling tool* Allows continuous water flow through the storage cell 0 S Is easy to distinguish visually from a fuel assembly Blocking devices are administratively controlled with the same level of rigor as fuel assemblies
- A blocking device move sheet package is a change document used to specify and record changes to plant configuration as it relates to blocking devices. A blocking device move sheet contains, as a minimum, the "from" location and the "to" location of the blocking device. Any number of blocking devices may be moved using a single blocking device move sheet package. Blocking devices will not be moved using the same move sheet package as fuel assemblies.
This restriction prevents a single error from removing a blocking device and placing a fuel assembly in a location that is required to be empty.Confirmation of Configquration Control In accordance with existing procedures, a 100 percent serial number check is performed once per calendar year to ensure that every fuel assembly stored in the SFP matches the SFP maps.This provision limits the amount of time that a misload condition could potentially exist.Mitigqation of a Misload Event If the controls discussed above are insufficient to prevent a fuel assembly from being misloaded, the following will mitigate the consequences of such an event:* Misload events, including misload events involving multiple fuel assemblies, have been analyzed-An adequate soluble boron margin mitigates the misload event. TS 3.7.15 requires 2150 ppm of soluble boron* The limiting misload of a single fuel assembly requires 1100 ppm of soluble boron to maintain keff <0.95 16 Enclosure Description and Assessment of Proposed License Amendment* The limiting, credible, multiple misload of placing a fresh fuel assembly into every blocked location in the most limiting array (Array C) requires 1460 ppm of soluble boron to maintain keff <0.95 Placing fresh fuel assemblies face-adjacent to one another is not credible.
A defense-in-depth approach provides multiple, independent barriers to this event. These barriers include:-Move sheets are generated, independently verified, and approved by qualified individuals
-Blocking devices or trash cans are placed in locations that are face-adjacent to locations approved for the storage of fresh fuel-The fuel movers will verify that fresh fuel is not placed in face-adjacent locations prior to completing each fuel move Sufficient rigor is placed into the generation of move sheets, execution of fuel movement, and maintenance of SEP maps that the likelihood of an assembly being misplaced in the SEP is small. The misplacement of multiple fuel assemblies is less probable.
Therefore, the approach used in the analysis is appropriately conservative.
The impact of placing multiple fresh fuel assemblies in face-adjacent locations is not evaluated in the analysis because this event is not considered credible.A regulatory commitment regarding the implementation of procedural controls to require verification that fresh fuel assemblies are not placed face-adjacent to one another before completing a fuel move is provided in Attachment 4.3.2 New Fuel Storage and Fuel Transfer Equipment Analysis A criticality safety analysis was performed to support operation of the NFS racks, the IFSR, the new fuel elevator, and the fuel upender and transfer machine. When discussing the new fuel elevator, and the fuel upender and transfer machine together, they are referred to as the fuel handling equipment (FHE). The existing NFS racks, IFSR, and FHE were evaluated to confirm that each system maintains subcriticality while performing its designed purposes.3.2.1 Storage and Equipment Description N.ew Fuel Storagqe Desigqn The NFS rack assemblies are made up of individual racks similar to those shown in UFSAR Figure 9.1-1. A minimum edge-to-edge spacing between fuel assemblies is maintained in adjacent rows. This spacing is the minimum value after allowances are made for rack fabrication tolerances and the predicted deflections resulting from postulated accident conditions.
The stainless steel construction of the storage racks is compatible with the water and the zirconium-clad fuel. The top structure of the racks is designed such that there is no opening between adjacent fuel cavities that is as large as the cross-section of the fuel bundle. In addition, the outer structure of the racks precludes the inadvertent placement of a bundle against the rack closer than the prescribed edge-to-edge spacing.17 Enclosure Description and Assessment of Proposed License Amendment Two concrete storage cavities are utilized for NFS. Each cavity is approximately 8 feet by 23 feet and contains 45 fuel assemblies in stainless steel racks. Three racks are installed in each cavity, forming a 3x15 array of fuel assemblies.
The rack structure provides at least 10 inches between the top of the active fuel and the top of the rack to preclude criticality in case a fuel assembly is dropped into a horizontal position on the top of the rack. The NFS racks and facilities are qualified as Seismic Category I and will survive a safe shutdown earthquake without loss of safety function.The following postulated accidents were considered in the design of the NFS racks:* Flooding -complete immersion of the entire storage array in pure, unborated, room temperature water* Envelopment of the entire array in a uniform density aqueous foam or mist of optimum density that maximizes the reactivity of the finite array (a condition that could result from firefighting)
- A fuel assembly dropped from a height of 4.5 ft onto the rack that falls horizontally across the top of the rack* Tensile load of 5000 lbs on the rack Although the above accident conditions have been postulated, the FHE, NFS racks, and the building arrangement are designed to minimize the possibility of these accidents and the effects resulting from these accidents.
Intermediate Fuel Storagqe Rack The IFSR is a four-cavity fuel storage rack in a lx4 array designed as an intermediate storage location for fuel bundles during refueling.
The rack is located in the containment adjacent to the core support barrel laydown area, which provides access to the refueling machine for insertion and removal of fuel bundles.Each cavity in the IFSR is a stainless steel can 8.69 inches on a side. The cavities are separated by a fuel center-to-center pitch of 18.56 inches. Each of the cavities is open at the bottom to provide thermal cooling for the worst case fuel bundle. The rack structure is designed to maintain keff < 0.95 by assuring under all normal and accident conditions, which includes SSE, that the minimum edge distance is not violated and that a fuel bundle cannot violate the 12-inch minimum stand-off distance around the cavities.New fuel may be stored in the IFSR before being moved into the core. Partially spent fuel may be moved out of the core and stored temporarily in the IFSR to provide spaces for fuel shuffling.
Spent fuel may be stored in the IFSR before being sent to the SFP.Fuel Upender and Transfer Machine The transfer machine, or carriage, conveys the fuel assemblies through the transfer tube. Two fuel assembly cavities are provided in the fuel carriage to reduce overall fuel handling time. After the refueling machine deposits a spent fuel bundle in the open cavity, it only has to move approximately one foot to pick up the new fuel assembly, which was brought from the fuel building in the other cavity. The handling operation in the fuel building is similar. The dual cavity 18 Enclosure Description and Assessment of Proposed License Amendment arrangement permits both fuel handling machines to travel fully loaded at all times. Fuel assemblies are placed on the transfer carriage in a vertical position, lowered to the horizontal position, moved through the fuel transfer tube on the transfer carriage, and then restored to the vertical position.
Wheels support the carriage and allow it to roll on tracks within the transfer*tube. The track sections at both ends of the transfer tube are mounted on the upending machines to permit the carriage to be properly positioned at the limits of its travel.An upending machine is provided at each end of the transfer tube. Each machine consists of a Structural support base from which is pivoted an upending straddle frame that engages the two-cavity fuel carrier. Hydraulic cylinders attached to the upending frame that engages the support base rotate the fuel carrier between the vertical and horizontal position.
A third fuel assembly was modeled five inches from the transfer carriage to allow for the presence of an additional fuel assembly no closer than five inches from the carriage. (Appendix B of Attachment 8)New Fuel Elevator The new fuel elevator is utilized to lower new fuel from the operating floor to the bottom of the pool where it is grappled by the spent fuel handling tool. The elevator is powered by a cable winch and fuel is contained in a simple support structure whose wheels are captured in two rails.3.2.2 New Fuel Criticality Safety Analysis Acceptance Criteria The objective of the criticality safety analysis is to ensure that the fuel storage operations are within the bounds 10 CFR 50.68(b)(2) and 50.68(b)(3):
- The estimated ratio of neutron production to neutron absorption and leakage (keff) of the fresh fuel in the storage racks shall be calculated assuming the racks are loaded with fuel of the maximum fuel assembly reactivity and flooded with unborated water and must not exceed 0.95, at a 95 percent probability, 95 percent confidence level. This evaluation need not be performed if administrative controls and/or design features prevent such flooding or if fresh fuel storage racks are not used.* If optimum moderation of fresh fuel in the fresh fuel storage racks occurs when the racks are assumed to be loaded with fuel of the maximum fuel assembly reactivity and filled with low-density hydrogenous fluid, the keff corresponding to this optimum moderation must not exceed 0.98, at a 95 percent probability, 95 percent confidence level. This evaluation need not be performed if administrative controls and/or design features prevent such moderation or if fresh fuel storage racks are not used.Design Approach Compliance is shown for the NFS racks, IFSR, and EHE by demonstrating that the system does not exceed 0.95 at a 95 percent probability with a 95 percent confidence level. A conservative combination of best estimate and bounding values has been selected to model the fuel in the analysis to ensure that fuel represented by the proposed TS is less reactive than the fuel modeled in the analysis.19 Enclosure Description and Assessment of Proposed License Amendment Computer Codes The analysis methodology employs SCALE 6.1.2 (Reference 6.3) with the 238-group cross-section library based on ENDF/B-VII.
All analyses performed used the "Fresh fuel without Absorber' validation suite. KENO-Va is used to determine the absolute reactivity of fresh fuel assemblies in the NFS.The validation of the ENDF/B-VII library with the SCALE 6.1.2 CSAS5 module is documented in Appendix A of Attachment
- 8. The code validation shows that SCALE 6.1.2 is an accurate tool for calculation of keff for the applications in this LAR. The benchmark calculations utilize the same computer plafform and cross-section libraries that are used for the design basis calculations.
Limitingq Fuel Desigqn Selection There are four potentially limiting fuel designs that have been used at PVNGS. The VAP design is currently in use on site and NGF is planned for use in the future. Therefore, both VAP and NGF designs are considered the two potentially limiting fuel designs. The STD and ANP fuel designs do not need to be addressed for these calculations because they are bounded by the VAP fuel design. Additional information is provided in Section B.4.2 of Attachment 8.Treatment of Concrete Concrete is a material which has a large variety of different potential compositions, all of which can be labeled as "concrete." Attachment 8 references the most limiting design for each situation.
The analysis has used a bounding treatment for concrete and the methodology remains conservative throughout the life of the concrete.
Additional information is provided in Section B.4.3 of Attachment 8.Biases and Uncertainties Reactivity biases are known variations between the real and analyzed system and their reactivity impact is added directly to the calculated keff. Uncertainties are random dispersions around a nominal, measured quantity.
Their impact is added to the calculated keff as the square root of the sum of the squares of the uncertainties.
The following biases and uncertainties are accounted for in the analysis.
A detailed discussion of biases and uncertainties is provided in Section B.4.5 of Attachment 8.* Reactivity effect of manufacturing tolerances
- Structural material presence* Eccentric fuel assembly positioning
- Uncertainty in the predictive capability of SCALE 6.1.2 and the associated cross-section library* Temperature bias for operating temperature range* Planar enrichment bias 20 Enclosure Description and Assessment of Proposed License Amendment New Fuel Storagqe Rack Criticality Safety Analysis The criticality safety analysis for the NFS rack consists of determining the limiting fuel design under both the fully flooded and optimum moderation condition.
Biases and uncertainties for both fully flooded and optimum moderation conditions are calculated using the limiting fuel design. The best estimate keff of the NES rack under both full density water and optimum moderation conditions is less than the target keff. This demonstrates that the NFS rack complies with the requirements of 10 CFR 50.68. Additional information is provided in Section B.4.6 of Attachment 8.The analysis of the NES rack has demonstrated that it can be operated in its design capacity without risk of exceeding the maximum reactivity imposed by regulation.
The analysis supports use of these components up to a maximum radially averaged enrichment of 4.65 weight percent U-235. All fuel used to date at PVNGS has an initial enrichment of < 4.55 weight percent U-235.Intermediate Fuel Storaqe Rack Criticality Safety Analysis The criticality safety analysis for the IFSR consists of determining the target keff for the IFSR, then confirming that the best estimate system keff (plus 2 o) is below the target keff with the limiting fuel design. The analysis uses the NGF design because the NGF is more reactive with full density water than VAP fuel. The biases and uncertainties are also calculated using the NGF design. The best estimate keff of the IFSR is less than the target keff, which demonstrates that the IFSR complies with the requirements of 10 CFR 50.68. Additional information is provided in Section B.4.7 of Attachment 8.Fuel Upender. Transfer Machine, and New Fuel Elevator Criticality Safety Analysis The criticality safety analysis for the fuel upender and transfer machine, and the analysis for the new fuel elevator, demonstrate that they can be used with fresh 4.65 weight percent U-235 fuel without exceeding a keff of 0.95 at a 95 percent probability, 95 percent confidence level. The design basis fuel is the NGF design. The best estimate keff of the fuel upender and transfer machine, and the new fuel elevator, is less than the target which demonstrates compliance with the requirements of 10 CFR 50.68. Additional information is provided in Section B.4.8 of Attachment 8.3.3 Spent Fuel Pool Transition Plan The SEP transition will be conducted over a total lapsed time of approximately 24 months with a schedule based on the unit refueling outages. Therefore, APS will insert two sets of TS and TS Bases pages during implementation.
One set will be labeled "Before SEP transition" and the other set will be labeled "After SFP transition." The Spent Fuel Pool Transition Plan is based on a SEP module-by-module transition scheme.Each SEP module that has not been transitioned is governed by the 'Before SEP transition" pages. As each module in a SEP is transitioned to the new configuration, the Shift Manager will make an entry in the control room log and declare that module as "transitioned." That particular module is then governed by the "After SEP transition" pages. When all three units have been transitioned, APS will submit an administrative TS change to remove the "before" and "after" pages, and insert the final pages.21 Enclosure Description and Assessment of Proposed License Amendment APS will transition to the proposed TS 3.7.17 in each of the three units in the following manner: 1. Move fuel assemblies as needed in order to neutronically decouple one module from the balance of the SEP. Analysis demonstrates that one row of empty cells is enough to decouple modules in the SEP.2. Perform a shuffle of the decoupled module, including installation of NETCO-SNAP-IN rack inserts.a. Some modules may need to be completely emptied of fuel assemblies.
The assemblies that must be moved may be stored in the appropriate regions of the rest of the SEP.b. Fuel assemblies that already meet the new TS 3.17.17 requirements may not need to be shuffled.3. Upon completion of the fuel shuffle and NETCO-SNAP-IN rack insert installation, the Shift Manager will declare the module has transitioned to the new TS 3.7.17 and enter this information into the control room log.4. Perform additional fuel shuffles, as needed, in order to move fuel from other parts of the SEP to the recently transitioned module.5. Repeat steps 1 through 4 for all 17 modules.6. Perform a 100 percent pool verification to confirm the following a. Fuel has been properly moved as confirmed by a 100 percent serial number check b. Stainless steel L-Jnserts are in the locations assumed in the analysis of record c. NETCO-SNAP-IN rack inserts are in their assumed locations d. Blocking devices are in their assumed locations A generic Westinghouse study (Reference 6.6) investigated the adequacy of assuming a distance of one cell pitch for neutronic isolation.
The study included NGF, which was determined to be the limiting fuel type used in the PVNGS SEP criticality safety analysis.
The results of the study concluded that a distance of approximately 10 cm (3.94 inches) of water was adequate for neutronic decoupling.
Given the required separation distance for neutronic decoupling, fuel assemblies separated by a single cell pitch or more at PVNGS are neutronically decoupled.
Once the Spent Fuel Pool Transition Plan has been started in a particular unit, the insert installation and SFP transition shall be executed in a deliberate, safe, and controlled manner until complete in that unit. The transition to the new SEP configuration will be completed in all three units in accordance with the Spent Fuel Pool Transition Plan within two years of the NRC approval date of the amendment or by December 31, 2019, whichever is later. A regulatory commitment regarding transition plan implementation is provided in Attachment 4.
4.0 REGULATORY EVALUATION
4.1 Applicable Regulatory Requirements The regulations in 10 CFR 50.36(c)(2)(ii)(B), Limiting conditions for operation, state: 22 Enclosure Description and Assessment of Proposed License Amendment Criterion
- 2. A process variable, design feature, or operating restriction that is an initial condition of a design basis accident or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.Technical Specification (TS) 3. 7.17 currently meets this requirement and will continue to meet this requirement after the proposed changes are approved and implemented.
The regulations in 10 CFR 50.36(c)(4), Design features, state: Design features to be included are those features of the facility such as materials of construction and geometric arrangements, which, if altered or modified, would have a significant effect on safety and are not covered in categories described in paragraphs (c)(1), (2), and (3) of this section.TS 4.3.1 currently meets this requirement and will continue to meet this requirement after the proposed changes are approved and implemented.
The regulations in 10 CFR 50.68, Criticality accident requirements, specifically 10 CFR 50.68(b)(1) state: Plant procedures shall prohibit the handling and storage at any one time of more fuel assemblies than have been determined to be safely subcritical under the most adverse moderation conditions feasible by unborated water.This requirement is currently met by existing PVNGS fuel handling procedures and will continue to be met by the same procedures after the proposed changes are approved and implemented.
The regulations in 10 CER 50.68(b)(2) state: The estimated ratio of neutron production to neutron absorption and leakage (k-effective) of the fresh fuel in the fresh fuel storage racks shall be calculated assuming the racks are loaded with fuel of the maximum fuel assembly reactivity and flooded with unborated water and must not exceed 0.95, at a 95 percent probability, 95 percent confidence level. This evaluation need not be performed if administrative controls and/or design features prevent such flooding or if fresh fuel storage racks are not used.The regulations in 10 CFR 50.68(b)(3) state: If optimum moderation of fresh fuel in the fresh fuel storage racks occurs when the racks are assumed to be loaded with fuel of the maximum fuel assembly reactivity and filled with low-density hydrogenous fluid, the k-effective corresponding to this optimum moderation must not exceed 0.98, at a 95 percent probability, 95 percent confidence level. This evaluation need not be performed if administrative controls and/or design features prevent such moderation or if fresh fuel storage racks are not used.The regulations in 10 CFR 50.68(b)(4) state: If no credit for soluble boron is taken, the k-effective of the spent fuel storage racks loaded with fuel of the maximum fuel assembly reactivity must not exceed 0.95, at a 95 percent probability, 95 percent confidence level, if flooded with unborated water. If credit is taken for soluble boron, the k-effective of the spent 23 Enclosure Description and Assessment of Proposed License Amendment fuel storage racks loaded with fuel of the maximum fuel assembly reactivity must not exceed 0.95, at a 95 percent probability, 95 percent confidence level, if flooded with borated water, and the k-effective must remain below 1.0 (subcritical), at a 95 percent probability, 95 percent confidence level, if flooded with unborated water.The requirements in 10 CFR 50.68(b) cited above are met by the nuclear criticality safety analyses provided in WCAP- 18030-P. The results of the criticality analysis form the basis of the proposed TS 3.7.17 changes. TS 3.7.17 currently meets these requirements and will continue to meet these requirements after the proposed changes are approved and implemented.
The regulations in 10 CFR 50.68(b)(7) state: The maximum nominal U-235 enrichment of the fresh fuel assemblies is limited to five (5.0) percent by weight.TS 4.3.1.2 currently meets this requirement and will continue to meet this requirement after the proposed changes are approved and implemented.
The regulations in 10 CFR Part 50, Appendix A, General Design Criteria for Nuclear Power Plants, Criterion 62, Prevention of criticality in fuel storage and handling, state: Criticality in the fuel storage and handling system shall be prevented by physical systems or processes, preferably by use of geometrically safe configurations.
TS 3. 7.17 currently meets this requirement and will continue to meet this requirement after the proposed changes are approved and implemented.
The guidance in DSS-ISG-2010-01 is to be used by NRC staff to review nuclear criticality safety analyses for the storage of new and spent nuclear fuel as they apply to applications for license amendments submitted after September 29, 2011.This license amendment request and WCAP-18030-P were developed using the guidance of DSS-ISG-2010-01.
4.2 Precedent The analysis methodology for the site-specific criticality analysis employs the PARAGON code, which is approved for use by the NRC (Reference 6.4).4.:3 Significant Hazards Consideration As required by 10 CFR 50.91(a), Notice for Public Comment, an analysis of the issue of no significant hazards consideration using the standards in 10 CFR 50.92, Issuance of Amendment, is presented below: 1. Does the proposed amendment involve a significant increase in the probability or consequences of an accident previously evaluated?
Response:
No.24 Enclosure Description and Assessment of Proposed License Amendment The proposed amendment would modify the Palo Verde Nuclear Generating Station (PVNGS) Technical Specifications (TS) to incorporate the results of an updated criticality safety analysis for both new fuel and spent fuel storage. The revised criticality safety analysis provides an updated methodology that allows credit for neutron absorbing NETCO-SNAP-IN rack inserts and corrects non-conservative input assumptions in the previous criticality safety analysis.The proposed amendment does not change or modify the fuel, fuel handling processes, number of fuel assemblies that may be stored in the spent fuel pool (SFP), decay heat generation rate, or the SFP cooling and cleanup system. The proposed amendment was evaluated for impact on the following previously evaluated events and accidents:
- fuel handling accident (FHA)* fuel misload event* SEP boron dilution event* seismic event* loss of SEP cooling event Implementation of the proposed amendment will be accomplished in accordance with the Spent Fuel Pool Transition Plan and does not involve new fuel handling equipment or processes.
The radiological source term of the fuel assemblies is not affected by the proposed amendment request. The EHA radiological dose consequences associated with fuel enrichment at this level are addressed in the PVNGS Updated Final Safety Analysis Report (UFSAR) Section 15.7.4 and remain unchanged.
Therefore, the proposed amendments do not significantly increase the probability or consequences of a FHA.Operation in accordance with the proposed amendment will not change the probability of a fuel misload event because fuel movement will continue to be controlled by approved fuel handling procedures.
Although there will be additional allowable storage arrays defined by the amendment, the fuel handling procedures will continue to require identification of the initial and target locations for each fuel assembly that is moved. The consequences of a fuel misload event are not changed because the reactivity analysis demonstrates that the same subcriticality criteria and requirements continue to be met for the limiting fuel misload event.Operation in accordance with the proposed amendment will not change the probability or consequences of a boron dilution event because the systems and events that could affect SFP soluble boron concentration are unchanged.
The current boron dilution analysis demonstrates that the limiting boron dilution event will reduce the boron concentration from the TS limit of 2150 ppm to 1900 ppm. This leaves sufficient margin to the 1460 ppm credited by the SFP criticality safety analysis.
The analysis confirms that the time needed for dilution to reduce the soluble boron concentration is greater than the time needed for actions to be taken to prevent further dilution.Operation in accordance with the proposed amendment will not change the probability of a seismic event since there are no elements of the updated criticality analysis that influence the occurrence of a seismic event. The consequences of a seismic event are not significantly increased because the forcing functions for seismic excitation are not increased and because the mass of storage racks with NETCO-SNAP-IN inserts is not appreciably 25 Enclosure Description and Assessment of Proposed License Amendment increased.
Seismic analyses demonstrate adequate stress levels in the storage racks when inserts are installed.
Operation in accordance with the proposed amendment will not change the probability of a loss of SEP cooling event because the systems and events that could affect SEP cooling are unchanged.
The consequences are not significantly increased because there are no changes in the SFP heat load or SEP cooling systems, structures, or components.
Furthermore, conservative analyses indicate that the current design requirements and criteria continue to be met with the NETCO-S NAP-IN inserts installed.
Therefore, the proposed amendment does not involve a significant increase in the probability or consequences of an accident previously evaluated.
- 2. Does the proposed amendment create the possibility of a new or different kind of accident from any accident previously evaluated?
Response:
No.The proposed amendment would modify the PVNGS TS to incorporate the results of an updated criticality safety analysis for both new fuel and spent fuel storage. The revised criticality safety analysis provides an updated methodology that allows credit for neutron absorbing NETCO-SNAP-IN rack inserts and corrects non-conservative input assumptions in the previous criticality safety analysis.The proposed amendment does not change or modify the fuel, fuel handling processes, number of fuel assemblies that may be stored in the pool, decay heat generation rate, or the SEP cooling and cleanup system. The effects of operating with the proposed amendment are listed below. The proposed amendment was evaluated for the potential of each effect to create the possibility of a new or different kind of accident:* addition of inserts to the SEP storage racks* additional weight from the inserts* new storage patterns* displacement of SEP water by the inserts, Each NETCO-SNAP-IN insert will be placed between a fuel assembly and the storage cell wall, taking up some of the space available on two sides of the fuel assembly.
Analyses demonstrate that the presence of the inserts does not adversely affect spent fuel cooling, seismic capability, or subcriticality.
The aluminum and boron carbide materials of construction have been shown to be compatible with nuclear fuel, storage racks, and SEP environments, and generate no adverse material interactions.
Therefore, placing the inserts into the SEP storage racks cannot cause a new or different kind of accident.Operation with the added weight of the NETCO-SNAP-IN inserts will not create a new or different accident.
The analyses of the racks with NETCO-SNAP-IN inserts installed demonstrate that the stress levels in the rack modules continue to be considerably less than allowable stress limits. Therefore, the added weight from the inserts cannot cause a new or different kind of accident.26 Enclosure Description and Assessment of Proposed License Amendment Operation with the proposed fuel storage patterns will not create a new or different kind of accident because fuel movement will continue to be controlled by approved fuel handling procedures.
These procedures continue to require identification of the initial and target locations for each fuel assembly that is moved. There are no changes in the criteria or design requirements pertaining to fuel storage safety, including subcriticality requirements.
Analyses demonstrate that the proposed storage patterns meet these requirements and criteria with adequate margins. Therefore, the proposed storage patterns cannot cause a new or different kind of accident.Operation with insert movement above stored fuel will not create a new or different kind of accident.
The insert with its handling tool weighs less than the weight of a single fuel assembly.
Single fuel assemblies are routinely moved safely over fuel assemblies and the same level of safety in design and operation will be maintained when moving the inserts.The installed rack inserts will displace a negligible quantity of the SEP water volume and therefore will not reduce operator response time to previously-evaluated SFP accidents.
The accidents and events previously analyzed remain bounding.
Therefore, the proposed amendment does not create the possibility of a new or different kind of accident from any accident previously evaluated.
- 3. Does the proposed amendment involve a significant reduction in a margin of safety?Response:
No.The proposed amendment would modify the TS to incorporate the results of an updated criticality safety analysis for both new fuel and spent fuel storage. The revised criticality safety analysis provides an updated methodology that allows credit for neutron absorbing NETCO-SNAP-IN rack inserts and corrects non-conservative input assumptions in the previous criticality safety analysis.
It was evaluated for its effect on current margins of safety as they relate to criticality, structural integrity, and spent fuel heat removal capability.
The margin of safety for subcriticality required by 10 CFR 50.68(b)(4) is unchanged.
New criticality analyses confirm that operation in accordance with the proposed amendment continues to meet the required subcriticality margins.The structural evaluations for the racks and spent fuel pool with NETCO-SNAP-IN inserts installed show that the rack and SEP are unimpaired by loading combinations during seismic motion, and there is no adverse seismic-induced interaction between the rack and NETCO-SNAP-IN inserts.The proposed amendment does not affect spent fuel heat generation, heat removal from the fuel assembly, or the SEP cooling systems. The effects of the NETCO-SNAP-IN inserts are negligible with regards to volume of water in the pool, flow in the SEP rack cells, and heat removal system performance.
The addition of a Spent Fuel Pool Rack Neutron Absorber Monitoring program (proposed TS 5.5.21) provides a method to identify potential degradation in the neutron absorber material prior to challenging the assumptions of the criticality safety analysis related to the material.Therefore, the addition of this monitoring program does not reduce the margin of safety;27 Enclosure Description and Assessment of Proposed License Amendment rather it ensures th'e margin of safety is maintained for the planned life of the spent fuel storage racks.Therefore, the proposed amendment does not involve a significant reduction in the margin of safety.4.4 Conclusion APS concludes that operation of the facility in accordance with the proposed amendment does not involve a significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and, accordingly, a finding of "no significant hazards consideration" is justified.
Based on the considerations discussed above, (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or the health and safety of the public.5.0 ENVIRONMENTAL CONSIDERATION A review has determined that the proposed amendment would change a requirement with respect to installation or use of a facility component located within the restricted area, as defined in 10 CFR 20, Standards for Protection Against Radiation.
However, the proposed amendment does not involve (i) a significant hazards consideration, (ii) a significant change in the types or a significant increase in the amounts of any effluents that may be released offsite, or (iii) a significant increase in individual or cumulative occupational radiation exposure.
Accordingly, the proposed amendment meets the eligibility criterion for categorical exclusion set forth in 10 CFR 51.22(c)(9).
Therefore, pursuant to 10 CFR 51 .22(b), no environmental impact statement or environmental assessment need be prepared in connection with the proposed amendment.
6.0 REFERENCES
6.1 Criticality Safety Analysis for Palo Verde Nuclear Generating Station Units 1, 2, and 3 (Proprietary), WCAP-1 8030-P, Revision 0, September 2015.6.2 Staff Guidance Regarding the Nuclear Criticality Safety Analysis for Spent Fuel Pools, DSS-ISG-201 0-01, Revision 0, Nuclear Regulatory Commission Division of Safety Systems, Rockville, MD, September 29, 2011. (ML1 10620086)6.3 Scale: A Comprehensive Modeling and Simulation Suite for Nuclear Safety Analysis and Design, ORNL/TM-2005/39, Version 6.1, Oak Ridge National Laboratory, Oak Ridge, TN, June 2011.6.4 M. Ouisloumen, H. Huria, et al, Qualification of the Two-Dimensional Transport Code PARAGON, WCAP-16045-P-A, Revision 0, Westinghouse Electric Company LLC, Monroeville, PA, August 2004.28 Enclosure Description and Assessment of Proposed License Amendment 6.5 C. V. Parks, et al, Review and Prioritization of Technical Issues Related to Bumup Credit for LWR Fuel, NUREG/CR-6665, Oak Ridge National Laboratory, Oak Ridge, TN, February 2000.6,6 Letter, J. Gresham (WEC) to NRC, Responses to Requests for Additional Information from the Review of WCAP- 1 7483-PA/WCAP-1 7483-NP, Revision 0, 'Westinghouse Methodology for Spent Fuel Pool and New Fuel Rack Criticality Safety Analysis,'
LTR-NRC-15-60, dated July 20, 2015.29 Enclosure ,Description and Assessment of Proposed License Amendment ATTACHMENT 1 Marked-up Technical Specifications Pages (Pages Provided for Before and After SEP Transition) 3.7.17-1-3.7.17-2 3.7.17-3 3.7.17-4 4.0-2 4.0-3 5.5-19 SBefore SFP transitionI Spent Fuel Assembly Storage 3.7.17 3.7 PLANT SYSTEMS 3.7.17 Spent Fuel Assembly Storage LCO 3.7.17 APPLICABILITY:
The combination of initial enrichment, burnup, and decay time of each fuel assembly stored in each of the four regions of the fuel storage pool shall be within the acceptable burnup domain for each region as shown in Figures 3.7.17-1, 3.7.17-2, or 3.7.17-3, and described in Specification 4.3.1.1.Whenever any fuel assembly is stored in the fuel storage pool.ACTIONS__________________________
CONDITION REQUIRED ACTION COMPLETION TIME A. Requirements of the A.1------NOTE----
LCO not met. LCD 3.0.3 is not applicable.
Initiate action to Immediately move the noncomplying fuel assembly into an appropri ate region.SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.17.1 Verify by administrative means the initial Prior to enrichment, burnup, and decay time of the storing the fuel assembly is in accordance with Figures fuel assembly 3.7.17-1, 3.7.17-2, or 3.7.17-3, and in the fuel Specification 4.3.1.1. storage pool.PALO VERDE UNITS 1,2,3 3.7.17-I PAL VEDE NIT 1,,3 .7.7-1AMENDMENT NO. 117, 1£ IBefore SFP transition]I Spent Fuel Assembly Storage 3.7.17 Figure 3.7.17-1 ASSEMBLY BURNUP VERSUS INITIAL ENRICHMENT for Region 2 0 6 0 S S S S 0 ACC EPTABL[" for Reg on 2 15000 500 I-NOT A CEPTA LE for R gion 2 S S S S S S S S S S 0 S S S S S S S_________________
_________________
9 S S S S S S S S S Note: This curve assumes ero decay time.____________
~1~S (1 1.5 2.0 2.5 3.0 3.5 Initial Enrichment, weight %4.0 4.5
- 5.0 4.80%limiting enrichment PALO VERDE UNITS 1,2,3 3.7.17-2 AMENDMENT NO. 117, !2
[Before SFP transition Spent Fuel Assembly Storage 3.7.17 Figure 3.7.17-2 ASSEMBLY BURNUP VERSUS INITIAL ENRICHMENT for Region 3 (at decay times from 0 to 20 years)45000 40000 35000 30000 25000>20000 t'0 E ci 0 ACCEP' ABLE for Region 3 "o ,.S//NOT ACq;EPTABI.
E for Region S 0 ThOUU+J +
I S 0 S S(S-a---S S 5000~~.2 Noe Asn nent and current di S 0 S cay lime.: ly e/igible for Regi jm 3 if actualIBU
> 3U requirement for given initial endch F i, , , , .....m 1.5 2.0 2.5 3.0 3.5 4.0 4.5
- 5.0 Initial Enrichment, weight % 4.80%limiting Decaylime[ -U-1-5years
--4--2Oyears enrichment PALO VERDE UNITS 1,2,33..73AMNETNO 3.7.17-3 AMENDMENT NO.
I Before SFP transition Spent Fuel Assembly Storage 3.7.17 Figure 3.7.17-3 ASSEMBLY BURNUP VERSUS INITIAL ENRICHMENT
~for Region 4 (at decay times from C to 20 years)I-0 E~0 E a)0)3.0 3.5 Initial Enrichment, weight %limiting DecaylimeI
--0years 11-years
--&.-lOyears
-~-'-15 years --o-20years enrdchment PALO VERDE UNITS 1,2,3 371- MNMN O 3.7.17-4 AMENDMENT NO.
SAfter SFP transition Spent Fuel Assembly Storage 3.7.17 3.7 PLANT SYSTEMS 13.7.17-1 through 3.7.17-5.I 3.7.17 Spent Fuel Assembly Storage LCO 3.7.11 The combination of initialnichment, burnup, and decay time of each fuel assemblytreine=hftef'-
acceptable burnup domain for each region as shown in Figures') 7 17 1 ') 7 17 ') ,-~-i Q 7 17 '2 -~A A-,-.~k,-A -vs Specification 1.3.1.1.APPLICABILITY:
Whenever any fuel pool.assembly is stored in the fuel storage ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Requirements of the A.1------NOTE--
--LCO not met. LCO 3.0.3 is not applicable.
Initiate action to Immediately move the noncomplying fuel assembly into an appropriate region.SURVEILLANCE REQUIREMENTS_________
SURVEILLANCE FREQUENCY SR 3.7.17.1 Verify by administrative means the initial Prior to enrichment, burnup, and decay time of the storing the fuel assembly is in accordance with fuel assembly...and in the fuel Specification 4mi.3.1.1
... storage pool.ITables 3.7.17-1 through 3.7.17-5, Figure 3.7.17-PALO VERDE UNITS 1,2,3 3.7.17-1 PAL VEDE NIT 1,,3 .7.7-1AMENDMENT NO. I147, 2 Ilnsert new Tables 3.7.17-1 through 3.7.17-5 and jFigure 3.7.17-1 (total 6 pages) here.I IAfter SFP transitionI Spent Fuel Assembly Storage 3.7.17 FiguJren 3.7.17 1 Acc(rkArnl
\J l \lII-1CI IC TkITTT Al FI-dITriUMCIdT PALO VERDE UNITS 1,2,3 97179 PAL VRD UIT 12, 37.72AMENDMENT NO. 11.7, 2 SAfter SFP transition Spent Fuel Assembly Storage 3.7.17 limiting--me '-O years --U-5 years -* 10 years -U-I--l5 years --4-20 years enrichment PALO VERDE UNITS 1,2,33. 13AMNETNO 15 97]79 AMENDMENT NO.
SAfter SFP transitionI Spent Fuel Assembly Storage 3.7.17 Figure,3.7.17 3-ACSE'MILY VU NUP VERSUS TIT-I-AI ENRT/ICHMEITT
'4,~uuuU 45000 S S 40000 __ __ _ r S ACCEI Region 4 35000 250000 0..~2O000 15000 10000 I 0 S S_________________
_________
I S S S S__________________
___________________
S 0 S S NOT AC E ABLE or Region ____ __S S S S S_________________
_________________
_________________
_________
S______/S S S S 0/ ______ + 4-~---f 5000//N______ 4- + 4-S S 0 S 0 0 0 caYthne. 0 mq~kemwl for tyB~efor Reg M 4&fachiBU
>-hiaI enddi rdw caidwmt 1.2.0 2.5 3.0 3.5 4.0 4.5
- 5 Initial Enrichment, weight % 4.80%-ieI-4-0 years --11-5 years 15 years --4-20Oyears lenicmetnt PALO VERDE UNITS 1,2,33..7AMN ETN. §Q717A AMENDMENT NO.
[After SFP transition]
Spent Fuel Assembly Storage 3.7.17 Table 3.7.17-1 Fuel Regions Ranked by Reactivity Fuel Region 1 Highest Reactivity (See Note 2)Fuel Region 2 Fuel Region 3 Fuel Region 4 Fuel Region 5 Fuel Region 6 Lowest Reactivity Notes: 1. Fuel Regions are defined by assembly average burnup, initial enrichment' and decay time as provided by Table 3.7.17-2 through Table 3.7.17-5.2. Fuel Regions are ranked in order of decreasing reactivity, e.g., Fuel Region 2 is less reactive than Fuel Region 1, etc.3. Fuel Region 1 contains fuel with an initial maximum radially averaged enrichment up to 4.65 wt% 2 3 5 U. No burnup is required.4. Fuel Region 2 contains fuel with an initial maximum radially averaged enrichment up to 4.65 wt% 2 3 5 U with at least 16.0 GWd/MTU of bumup.5. Fuel Regions 3 through 6 are determined from the minimum burnup (BU) equation and coefficients provided in Tables 3.7.17-2 through 3.7.17-5.6. Assembly storage is controlled through the storage arrays defined in Figure 3.7.17-1.7. Each storage cell in an array can only be populated with assemblies of the Fuel Region defined in the array definition or a lower reactivity Fuel Region.SInitial Enrichment is the nominal 2 3 5 U enrichment of the central zone region of fuel, excluding axial blankets, prior to reduction in 2 3 5 U content due to fuel depletion.
If the fuel assembly contains axial regions of different 2 3 5 Uj enrichment values, such as axial blankets, the maximum initial enrichment value is to be utilized.
[After SFP transition[
Spent Fuel Assembly Storage 3.7.17 Table 3.7.17-2 Fuel Region 3: Burnup Requirement Coefficients Coefficients Decay Time (yr.) A 1 A 2 A 3 A 4 0 -1.5473 15.5395 -39.0197 24.1121 5 -1.4149 13.9760 -33.6287 18.3369 10 -1.3012 12.6854 -29.2539 13.6879 15 -1.0850 10.4694 -22.1380 6.3673 20 -0.9568 9.1487 -17.9045 2.0337 Notes: 1. Relevant uncertainties are explicitly included in the criticality analysis.
For instance, no additional allowance for burnup uncertainty or enrichment uncertainty is required.
For a fuel assembly to meet the requirements of a Fuel Region, the assembly burnup must exceed the "minimum burnup" (GWd/MTU) given by the curve fit for the assembly "decay time" and "initial enrichment." The specific minimum burnup (BU) required for each fuel assembly is calculated from the following equation: BU =Al
- En 3 + A 2
- En 2 + A 3
- En + A 4 2. Initial enrichment, En, is the maximum radial average 2 3 5 U enrichment.
Any En value between 2.55 wt% 2 3 5 U and 4.65 wt% 2 3 5 U may be used. Burnup credit is not required for an En below 2.55 wt% 2 3 5 U.3. It is acceptable to linearly interpolate between calculated BU limits based on decay time.4. The 20-year coefficients must be used to calculate the minimum BU for an assembly with a decay time of greater than 20 years.
[After SFP transition Spent Fuel Assembly Storage 3.7.17 Table 3.7.17-3 Fuel Region 4: Burnup Requirement Coefficients Coefficients Decay Time (yr.) A 1 A 2 A 3 A 4 0 0.4260 -6.2766 40.9264 -54.6813 5 0.2333 -4.1!545 32.9080 -46.1161 10 0.4257 -6.2064 39.0371 -51.5889 15 0.53 15 -7.3777 42.5706 -54.7524 20 0.5222 -7.3897 42.6587 -54.8201 Notes: 1. Relevant uncertainties are explicitly included in the criticality analysis.
For instance, no additional allowance for burnup uncertainty or enrichment uncertainty is required.
For a fuel assembly to meet the requirements of a Fuel Region, the assembly bumup must exceed the "minimum burnup" (GWd/MTU) given by the curve fit for the assembly "decay time" and "initial enrichment." The specific minimum burnup (BU) required for each fuel assembly is calculated from the following equation: BU=AI
- En 3 +A 2 *En 2+-bA 3 *En +A 4 2. Initial enrichment, En, is the maximum radial average 2 3 5 U enrichment.
Any En value between 1.75 wt% 2 3 5 U and 4.65 wt% 2 3 5 U may be used. Bumnup credit is not required for an En below 1.75 wt% 235U.3. It is acceptable to linearly interpolate between calculated BU limits based on decay time.4. The 20-year coefficients must be used to calculate the minimum BU for an assembly with a decay time of greater than 20 years.
[After SEP transition Spent Fuel Assembly Storage 3.7.17 Table 3.7.17-4 Fuel Region 5: Burnup Requirement Coefficients Decay Coefficients Time (yr.) A 1 A 2 A 3 A 4 0 -0.1114 -0.4230 20.9136 -32.8551 5 -0.1232 -0.4463 20.8337 -32.6068 10 -0.2357 0.4892 18.0192 -30.0042 15 -0.1402 -0.4523 20.3745 -31.7565 20 -0.0999 -0.8152 21.0059 -31.9911 Notes: 1. Relevant uncertainties are explicitly included in the criticality analysis.
For instance, no additional allowance for burnup uncertainty or enrichment uncertainty is required.
For a fuel assembly to meet the requirements of a Fuel Region, the assembly burnup must exceed the "minimum bumup" (GWd/MTU) given by the curve fit for the assembly "decay time" and "initial enrichment." The specific minimum burnup (BU) required for each fuel assembly is calculated from the following equation: BU =A 1
- En 3 + A 2
- En 2 + A 3
- En + A 4 2. Initial enrichment, En, is the maximum radial average 2 3 5 U enrichment.
Any En value between 1.65 wt% 2 3 5 U and 4.65 wt% 2 3 5 U may be used. Burnup credit is not required for an En below 1.65 wt% 2 3 5 U.3. It is acceptable to linearly interpolate between calculated BU limits based on decay time.4. The 20-year coefficients must be used to calculate the minimum BU for an assembly with a decay time of greater than 20 years.
[After SFP transition Spent Fuel Assembly Storage 3.7.17 Table 3.7.17-5 Fuel Region 6: Burnup Requirement Coefficients Decay Coefficients Time (yr.) A 1 A 2 A 3 A 4 0 0.7732 -9.3583 49.6577 -54.6847 5 0.7117 -8.4920 45.1124 -49.7282 10 0.6002 -7.2638 40.2603 -44.9348 15 0.5027 -6.2842 36.6715 -41.4934 20 0.2483 -3.7639 28.8269 -34.6419 Notes: 1. Relevant uncertainties are explicitly included in the criticality analysis.
For instance, no additional allowance for bumnup uncertainty or enrichment uncertainty is required.
For a fuel assembly to meet the requirements of a Fuel Region, the assembly bumup must exceed the "minimum burnup" (GWd/MTU) given by the curve fit for the assembly "decay time" and "initial enrichment." The specific minimum burnup (BU) required for each fuel assembly is calculated from the following equation: BU =A 1
- En 3 + A 2
- En 2 + A 3
- En + A 4 2. Initial enrichment, En, is the maximum radial average 2 3 5 U enrichment.
Any En value between 1.45 wt% Z3U and 4.65 wt% 23U may be used. Burnup credit is not required for an En below 1.45 wt% 2 3 5 U.3. It is acceptable to linearly interpolate between calculated BU limits based on decay time.4. The 20-year coefficients must be used to calculate the minimum BU for an assembly with a decay time of greater than 20 years.
SAfter SFP trransition Spent Fuel Assembly Storage 3.7.17 Figure 3.7.17-1 Allowable Storage Arrays Fwou Region 6 assemblies (6) Tw steckrbagded cells cotain abtlocess stells L(ise). TheRgo 1. Thsembishaded loathionsa indicatehelhcoti a stainless steel L-insert.NoETOSA-N 2.o Ae block1asedbis()cekrbaddwt w cells(X contains lcing dvcanolyw terainshe actiefulrein 3TC. NTheRgo-n1assemlNinets must bc oientedl winthe saedrcinaah stainless steel L-inserts.Eer 4 NTC -N PN isrsaeolloaeincells without a stainless steel L-inetms oti EC -N P1insert.
5-nsr. Anyhel egontann3 afe assemblyyr iCsa iseaiennemt (aerflld cell intinn aNTCall-Niset stoageayrys 6.e AnR traearagoaion deintdfrafe assembly may cbeckreplaced with noren-gon4fssssiele).Th mein2atserial.
n h ignlylctdRein4asml r ahi I Before SFP transitionI Design Features 4.0 4.0 DESIGN FEATURES (continued) 4.3 Fuel Storage 4.3.1 Criticality 4.3.1.1 The spent fuel storage racks are designed and shall be maintained with: a. Fuel assemblies having a maximum radially averaged U-235 enrichment of 4.80 weight percent;b. keff < 1.0 if fully flooded with unborated water, which includes an allowance for biases and uncertainties as described in Section 9.1 of the UFSAR;c. keff 0.95 if fully flooded with water borated to 900 ppm, which includes an allowance for biases and uncertainties as described in Section 9.1 of the UFSAR.d. A nominal 9.5 inch center-to-center distance between adjacent storage cell locations.
- e. Region 1: Fuel shall be stored in a checkerboard (two-out-of-four) storage pattern. Fuel that qualifies to be stored in Regions 1, 2, 3, or 4 in accordance with Figures 3.7.17-1, 3.7.17-2, or 3.7.17-3, may be stored in Region 1.f. Region 2: Fuel shall be stored in a repeating 3-by-4 storage pattern in which Region 2 (two-out-of-twelve) assemblies and Region 4 (ten-out-of-twelve) assemblies are mixed as shown in Section 9.1 of the UFSAR. Only fuel that qualifies to be stored in Regions 2, 3, or 4, in accordance with Figures 3.7.17-1, 3.1.17-2, or 3.7.17-3, may be stored in Region 2.g. Region 3: Fuel shall be stored in a four-out-of-four storage pattern. Only fuel that qualifies to be stored in Regions 3 or 4, in accordance with Figures 3.7.17-2 or 3.7.17-3, may be stored in Region 3.(conti nued)PALO VERDE UNITS 1,2,3 4.0-2 PALOVERE UITS1.23 40-2AMENDMENT NO. 47~
I Before SEP transition I Design Features 4.0 4.0 DESIGN FEATURES (continued)
- h. Region 4: Fuel shall be stored in a repeating 3-by-4 storage pattern in which Region 2 (two-out-of-twelve) assemblies and Region 4 (ten-out-of-twelve) assemblies are mixed as shown in Section 9.1 of the UFSAR. Only fuel that qualifies to be stored in Region 4 in accordance with Figure 3.7.17-3 shall be stored in Region 4.4.3.1.2 The new fuel storage racks are designed and shall be maintained with: a. Fuel assemblies having a maximum radially averaged U-235 enrichment of 4.80 weight percent;b. keff 0.95 if fully flooded with unborated water, which includes an allowance for biases and uncertainties as described in Section 9.1 of the UFSAR: c. keff 0.98 if moderated by aqueous foam, which includes an allowance for biases and uncertainties as described in Section 9.1 of the UFSAR; and d. A nominal 17 inch center to center distance between fuel assemblies placed in the storage racks.4.3.2 Drainage The spent fuel storage pool is designed and shall be maintained to prevent inadvertent draining of the pool below elevation 137 feet -6 inches.4.3.3 Capacity The spent fuel storage pool is designed and shall be maintained with a storage capacity limited to no more than 1329 fuel assemblies.
PALO VERDE UNITS 1,2,3 4.0-3 PAO EDEUNT 12, .03AMENDMENT NO. 11 o ...
After SFP transition Design Features 4.0 4.0 DESIGN FEATURES (continued) 4.3 Fuel Storage 4.3.1 Criticality 4.3.1.1 The spent fuel storage racks are designed and shall be maintained with:46 a. Fuel assemblies having/ maximum radially averaged U-235 enrichment .weight percent: b. keff < 1.0 if fully flooded with unborated water.which includes an allowance for biases and uncertainties as described in Section 9.1 of the UFSAR-c. keff -< 0.95 if fully flooded with water borated to~ppm, which includes an allowance for biases and unertainties as described in Section 9.1 of the 1460 UFSAR.d. A nominal 9.5 inch center-to-center distance between adjacent storage cell locations.
- e. R~on !:Fuel shall bc stored in a checkerboa=rd Fuel assemblies are .. ..... --.classified in Fuel Regions qualifie to be tore in Regon 1 2 3, or in 1-6 as shown in Tables a ccordance w.i th Figu=res 3.7.17 1, 3.7.17 2, or 3.7.17-1 through 3. 7.17 3 .. mayb in Rcgn 1.3.7.17-5. P Fe,,l shall be a; .......tin-3 byIstrg pattern in ..hich Regon 2 ( .. = atenrutontele seble arfel mixed as. shown accordance wit FiguQres 3.7.17 1, 3.7/.17% 2,-n or Regionnu3.
PALO VERDE UNITS 1,2,340-AMNETNO
.&4.0-2 AMENDMENT NO. 117, 125 After SFP transition Design Featuresi 4.0 4.0 DESIGN FEATURES (continued) 4.3.1.2 Th enewfuel Fuorag rhalls bre dstored ind sa llcat be 3ai bta ine witorgh ateni-hch go (t.oF uteflwlc assemblies hvn m a ndmu ra gionl avrae (t35enrouioctele)t afswembieht aerenmxdt shw incerectiones9.1 dof cthe edUF n S. tion 9.1& of th qualifies to blowne store binaRegio and uccordaincies ait Fiuesrib3 hllb soed in Regtion 9.Ifte FA n m.Ainta ined with inhc, r 465t ene itnc ewe a.fuel assemblies hlavin maximumoradill raverage U-235raiagientrichmenteor eih eret Thesp bn kefue strg 0.95 if fullyflooed withalunboaed waitaiert preveninawhichn drincldeing allowanepfor biaswesevatind u3 et nchertitessdsrbd nScin91o h Thespent includstoane allowanceforsbiase and uncl ermitaintes withstoasgescrpaibyliied iSetiono 9.1ofe then UF3AR andl fuesaseblesplce hisorgeraks PALO VERDE UNITS 1,2,3 4.0-3 PAL VRD UITS1,.34.-3AMENDMENT NO. 1-17,12 ..
IBefore SFP transitionI Programs and Manuals 5.5 5.5 Programs and Manuals (continued) 5.5.19 Battery Monitoring and Maintenance Program (continued)
- 4. In Regulatory Guide 1.129, Regulatory Position 3, Subsection 5.4.1, "State of Charge Indicator," the following statements in paragraph (d) may be omitted: "When it has been recorded that the charging current has stabilized at the charging voltage for three consecutive hourly measurements, the battery is near full charge.These measurements shall be made after the initially high charging current decreases sharply and the battery voltage rises to approach the charger output voltage." 5. In lieu of RG 1.129. Regulatory Position 7, Subsection 7.6, "Restoration," the following may be used: "Following the test, record the float voltage of each cell of the string." b. The program shall include the following provisions:
- 1. Actions to restore battery cells with float voltage<2.13 V;2. Actions to determine whether the float voltage of the remaining battery cells is 2.13 V when the float voltage of a battery cell has been found to be<2.13 V: 3. Actions to equalize and test battery cells that had been discovered with electrolyte level below the top of the plates: 4. Limits on average electrolyte temperature, battery connection resistance, and battery terminal voltage;and 5. A requirement to obtain specific gravity readings of all cells at each discharge test, consistent with manufacturer recommendations.
PALO VERDE UNITS 1,2,3 551 MNMN O 9 5.5-19 AMENDMENT NO.
IAfter SFP TransitionI Programs and Manuals 5.5 5.5 Programs and Manuals (continued) 5.5.19 Battery Monitoring and Maintenance Program (continued)
- 4. In Regulatory Guide 1.129. Regulatory Position 3, Subsection 5.4.1, "State of Charge Indicator." the following statements in paragraph (d) may be omitted: "When it has been recorded that the charging current has stabilized at the charging voltage for three consecutive hourly measurements, the battery is near full charge.These measurements shall be made after the initially high charging current decreases sharply and the battery voltage rises to approach the charger output voltage." 5. In lieu of RG 1.129, Regulatory Position 7, Subsection 1.6. "Restoration." the following may be used: "Following the test, record the float voltage of each cell of the string." b. The program shall include the following provisions:
- 1. Actions to restore battery cells with float voltage<2.13 V: 2. Actions to determine whether the float voltage of the remaining battery cells is 2.13 V when the float voltage of a battery cell has been found to be<2.13 V: 3. Actions to equalize and test battery cells that had been discovered with electrolyte level below the top of the plates: 4. Limits on average electrolyte temperature, battery connection resistance, and battery terminal voltage: and 5. A requirement to obtain specific gravity readings of all cells at each discharge test, consistent with manufacturer recommendations.
page 5.5-19 PALO VERDE UNITS 1,2.3 551 MNMN O 9 5.5-19 AMENDMENT NO.
[After SFP Transition]
Insert for page 5.5-19 5.5.21 Spent Fuel Storagqe Rack Neutron Absorber Monitoring Progqram Certain storage cells in the spent fuel storage racks utilize neutron absorbing material that is credited in the spent fuel storage rack criticality safety analysis to ensure the limitations of Technical Specifications 3.7.17 and 4.3.1.1 are maintained.
In order to ensure the reliability of the neutron absorber material, a monitoring program is provided to confirm the assumptions in the spent fuel pool criticality safety analysis.The Spent Fuel Storage Rack Neutron Absorber Monitoring Program shall require periodic inspection and monitoring of spent fuel pool test coupons and neutron absorber inserts on a performance-based frequency, not to exceed 10 years.Test coupons shall be inspected as part of the monitoring program. These inspections shall include visual, B-10 areal density and corrosion rate.Visual in-situ inspections of inserts shall also be part of the program to monitor for signs of degradation.
In addition, an insert shall be removed periodically for visual inspection, thickness measurements, and determination of retention force.
Enclosure Description and Assessment of Proposed License Amendment ATTACHMENT 2 Revised Technical Specifications Pages (Clean Copy)(Pages Provided for Before and After SEP Transition) 3.7.17-1 3.7.17-2 3.7.17-3 3.7.17-4 3.7.17-5 3.7.17-6 3.7.17-7 4.0-2 4.0-3 5.5-19 5.5-20 Before SFP transition Spent Fuel Assembly Storage 3.7.17 3.7 PLANT SYSTEMS 3.7.17 Spent Fuel Assembly Storage LCO 3.7.17 The combination of initial enrichment, burnup, and decay time of each fuel assembly stored in each of the four regions of the fuel storage pool shall be within the acceptable burnup domain for each region as shown in Figures 3.7.17-1, 3.7.17-2, or 3.7.17-3, and described in Specification 4.3.1.1.APPLICABILITY:
Whenever any fuel pool.assembly is stored in the fuel storage ACTIONS ________________
CONDITION REQUIRED ACTION COMPLETION TIME A. Requirements of the A.1------NOTE----
LCO not met. LCO 3.0.3 is not applicable.
Initiate action to Immediately move the noncomplying fuel assembly into an appropriate region.SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.17.1 Verify by administrative means the initial Prior to enrichment, burnup, and decay time of the storing the fuel assembly is in accordance with Figures fuel assembly 3.7.17-1, 3.7.17-2, or 3.7.17-3, and in the fuel Specification 4.3.1.1. storage pool.PALO VERDE UNITS 1,2,3 371- MNMN O -~3.7.17-1 AMENDMENT NO.
Before SFP transition Spent Fuel Assembly Storage 3.7.17 Figure 3.7.17-1 ASSEMBLY BURNUP VERSUS INITIAL ENRICHMENT for Regi on 2 20000 E E 1.5 2.0 2.5 3.0 3.5 4.0 Initial Enrichment, weight %4.5
- 5.0 4.80%limiting enrichment PALO VERDE UNITS 1,2,3 371- MNMN O 3.7.17-2 AMENDMENT NO.
Before SFP transition Spent Fuel Assembly Storage 3.7.17 Figure 3.7.17-2 ASSEMBLY BURNUP VERSUS INITIAL ENRICHMENT for Regi on 3 (at decay times from 0 to 20 years)E ci 1.5 2.0 2.5 3.0 3.5 4.0 4.5
- 5.0 Initial Enrichment, weight % 4.80%--q Oyers --m-5yars l~ears --X-15ear -- --2yeas ....limiting flmj --Oyars -U-yers ~-1yeas -E-l~yars oyers enrichment Decayl PALO VERDE UNITS 1,2,3 371- MNMN O 2 3.7.17- 3 AMENDMENT NO.
Before SFP transition Spent Fuel Assembly Storage 3.7.17 Figure 3.7.17-3 ASSEMBLY BURNUP VERSUS INITIAL ENRICHMENT for Regi on 4 (at decay times from 0 to 20 years)50000 45000 B S A 1 4 + -~ -S'IvuuU ______________
ACCE TABLE fo Region 4 35000 ____________
30000 B B S~.nnn ___________
I-E 20000 (I)(0J ,I)15000 10000 5000 0-1.5 J S S S S S NOT ACC EPTABLE ~or Region j4/ -U--B S ii~-i I SBB___ I ___ ___ ___ ___ __ B-#A~-- 4 + + +/.J S S S B S---S 0 B S B S S B B caytime.:+ t A------N+ote: Assern4Iy eflgbJ for Regi in 4 facua IBU>* U requiefent forjiI ghen irai ech ent and currnt d 2-.0 2.5 3.0 3.5 4.0 4.5 a 5.0 Initial Enrichment, weight % 4.80%limiting a -U-!--5 years -~-k--l years --UP-15 years --O-20 years .Jenrichment Decaylimej ye.PALO VERDE UNITS 1,2,3 371- MNMN O 2 3.7.17-4 AMENDMENT NO.
After SFP transition I Spent Fuel Assembly Storage 3.7.17 3.7 PLANT SYSTEMS 3.7.17 Spent Fuel Assembly Storage LCO 3.7.17 The combination of i niti al enrichment, burnup, and decay time of each fuel assembly shall be in compliance with the requirements specified in Tables 3.7.17-1 through 3.7.17-5.APPLICABILITY:
Whenever any fuel pool.assembly is stored in the fuel storage ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Requirements of the A.1------NOTE----
LCO not met. LCO 3.0.3 is not applicable.
Initiate action to Immediately move the noncomplying fuel assembly into an appropriate region.SURVEILLANCEREQUIREMENTS__________
SURVEILLANCE FREQUENCY SR 3.7.17.1 Verify by administrative means the initial Prior to enrichment, burnup, and decay time of the storing the fuel assembly is in accordance with Tables fuel assembly 3.7.17-1 through 3.7.17-5, Figure 3.7.17-1, in the fuel and Specification 4.3.1.1. storage pool.PALO VERDE UNITS 1,2,337171AEDNT O.I, 3.7.17-1 AMENDMENT NO.
After SFP transition Spent Fuel Assembly Storage 3.7.17 Table 3.7.17-1 Fuel Regions Ranked by Reactivity Fuel Region 1 Highest Reactivity (See Note 2)Fuel Region 2 Fuel Region 3 Fuel Region 4 Fuel Region 5 Fuel Region 6 Lowest Reactivity Notes: 1. Fuel Regions are defined by assembly average burnup. initial enrichment' and decay time as provided by Table 3.7.11-2 through Table 3.7.17-5.2. Fuel Regions are ranked in order of decreasing reactivity, e.g.. Fuel Region 2 is less reactive than Fuel Region 1. etc.3. Fuel Region 1 contains fuel with an initial maximum radially averaged enrichment up to 4.65 wt% 235U. No burnup is required.4. Fuel Region 2 contains fuel with an initial maximum radially averaged enrichment up to 4.65 wt% 235U with at least 16.0 GWd/MTU of burnup.5. Fuel Regions 3 through 6 are determined from the minimum burnup (BU) equation and coefficients provided in Tables 3.7.17-2 through 3.7.17-5.6. Assembly storage is controlled through the storage arrays defined in Figure 3.7.17-1.7. Each storage cell in an array can only be populated with assemblies of the Fuel Region defined in the array definition or a lower reactivity Fuel Region.'Initial Enrichment is the nominal 2 3 5 U enrichment of the central zone region of fuel, excluding axial blankets, prior to reduction in 2350 content due to fuel depletion.
If the fuel assembly contains axial regions of different 235U enrichment values, such as axial blankets, the maximum initial enrichment value is to be utilized.PALO VERDE UNITS 1,2,337.72AEDNTO.1, 3.7.17-2 AMENDMENT NO.
After SEP transition I Spent Fuel Assembly Storage 3.7.17 Table 3.7.17-2 Fue] Region 3: Burnup Requi rement Coefficients Decay Coefficients Time (yr.) Ai NmA A4 0 -1.5473 15.5395 -39.0197 24.1121 5 -1.4149 13.9760 -33.6287 18.3369 10 -1.3012 12.6854 -29.2539 13.6879 15 -1.0850 10.4694 -22.1380 6.3673 20 -0.9568 9.1487 -17.9045 2.0337 Notes: 1. Relevant uncertainties are explicitly included in the criticality analysis.
For instance, no additional allowance for burnup uncertainty or enrichment uncertainty is required.
For a fuel assembly to meet the requirements of a Fuel Region, the assembly burnup must exceed the "minimum burnup" (GWd/MTU) given by the curve fit for the assembly "decay 'time" and "initial enrichment." The specific minimum burnup (BU) required for each fuel assembly is calculated from the following equation: BU = Ai
- En 3 + A2
- En 2 + A3
- En + A4 2. Initial enrichment, En, is the maximum radial average 2 3enrichment.
Any En value between 2.55 wt% 2 3 5 U and 4.65 wt% 2 3 5 U may be used. Burnup credit is not required for an En below 2.55 wt% 2 3 5 U.3. It is acceptable to linearly interpolate between calculated BU limits based on decay time.4. The 20-year coefficients must be used to calculate the minimum BU for an assembly with a decay time of greater than 20 years.PALO VERDE UNITS 1,2,3 371- MNMN O 2 3.7.17-3 AMENDMENT NO.
After SFP transition Spent Fuel Assembly Storage 3.7.17 Table 3.7.17 -3 Fue] Region 4: Burnup Requirement Coefficients Decay Coefficients Time (yr.) Ai #0 0.4260 -6.2766 40.9264 -54.6813 5 0.2333 -4.1545 32.9080 -46.1161 10 0.4257 -6.2064 39.0371 -51.5889 15 0.5315 -7.3777 42.5706 -54.7524 20 0.5222 -7.3897 42.6587 -54.8201 Notes: 1. Relevant uncertainties are explicitly included in the criticality analysis.
For instance, no additional allowance for burnup uncertainty or enrichment uncertainty is required.
For a fuel assembly to meet the requirements of a Fuel Region, the assembly burnup must exceed the "minimum burnup" (GWd/MTU) given by the curve fit for the assembly "decay time" and "initial enrichment." The specific minimum burnup (BU) required for each fuel assembly is calculated from the following equation: BU =Ai
- En 3 + A2
- En 2 + A3
- En + A4 2. Initial enrichment, En, is the maximum radial average 2 3 5U enrichment.
Any En value between 1.15 wt% 2 3 5U and 4.65 wt% 2 3may be used. Burnup credit is not required for an En below 1.75 wt% 2 3 5 U.3. It is acceptable to linearly interpolate between calculated BU limits based on decay time.4. The 20-year coefficients must be used to calculate the minimum BU for an assembly with a decay time of greater than 20 years.PALO VERDE UNITS 1,2,3 371- MNMN O 2 3.7.17-4 AMENDMENT NO.
After SFP transition I Spent Fuel Assembly Storage 3.7.17 Table 3.7.17 -4 Fue] Region 5: Burnup Requirement Coefficients Decay Coeffi ci ents Time (yr.) Ai km# A4 0 -0.1114 -0.4230 20.9136 -32.8551 5 -0. 1232 -0. 4463 20. 8337 -32. 6068 10 -0. 2357 0.4892 18. 0192 -30. 0042 15 -0.1402 -0. 4523 20. 3745 -31. 7565 20 -0. 0999 -0. 8152 21. 0059 -31. 9911 Notes : 1. Relevant uncertainties are explicitly included in the criticality analysis.
For instance, no additional allowance for burnup uncertainty or enrichment uncertainty is required.
For a fuel assembly to meet the requirements of a Fuel Region, the assembly burnup must exceed the "minimum burnup" (GWd/MTU) given by the curve fit for the assembly "decay time" and "initial enrichment." The speci fi c minimum burnup (BU) requi red for each fuel assembly is calculated from the fol lowing equati on: BU= Ai
- En 3 + A2
- En 2 + A3
- En + A4 2. Initial enrichment, En. is the maximum radial average 2 3 5 U enrichment.
Any En value between 1.65 wt% 2 3 5 U and 4.65 wt% 2 3 5 U may be used. Burnup credit is not required for an En below 1.65 wt% 2 3 5 U.3. It is acceptable to linearly interpolate between calculated BU limits based on decay time.4. The 20-year coefficients must be used to calculate the minimum BU for an assembly with a decay time of greater than 20 years.PALO VERDE UNITS 1,2,3 371- MNMN D 3.7.17-5 AMENDMENT NO.
After SFP transition Spent Fuel Assembly Storage 3.7.17 Table 3.7.17 -5 Fuel Region 6: Burnup Requirement Coefficients Decay Coefficients Time (yr.) A A2 A4 A 0 0.7732 -9.3583 49.6577 -54.6847 5 0.7117 -8.4920 45.1124 -49.7282 10 0.6002 -7.2638 40.2603 -44.9348 15 0.5027 -6.2842 36.6715 -41.4934 20 0.2483 -3.7639 28.8269 -34.6419 Notes: 1. Relevant uncertainties are explicitly included in the criticality analysis.
For instance, no additional allowance for burnup uncertainty or enrichment uncertainty is required.
For a fuel assembly to meet the requirements of a Fuel Region, the assembly burnup must exceed the "minimum burnup" (GWd/MTU) given by the curve fit for the assembly "decay time" and "initial enrichment." The specific minimum burnup (BU) required for each fuel assembly is calculated from the fol lowi ng equati on: BU =A1
- En 3 + Am
- En 2 + A3
- En + A4 2. Initial enrichment, En. is the maximum radial average 235U enrichment.
Any En value between 1.45 wt% 2 3 5 U and 4.65 wt% 2 3 5 U may be used. Burnup credit is not required for an En below 1.45 wt% 2350.3. It is acceptable to linearly interpolate between calculated BU limits based on decay time.4. The 20-year coefficients must be used to calculate the minimum BU for an assembly with a decay time of greater than 20 years.PALO VERDE UNITS 1,2,3 371- MNMN O 3.7.17-6 AMENDMENT NO.
After SFP transition Spent Fuel Assembly Storage 3.7.17 Figure 3.7.17-1 Allowable Storage Arrays Array A 1 X Two Region 1 assemblies (1) checkerboarded with two blocked cells (X).The Region 1 assemblies are each in a cell with a stainless steel L-insert.
No NETCO-SNAP-IN inserts are credited.X 1 Array B 1 TC Two Region 1 assemblies (1) checkerboarded with two cells containing trash cans (TC). The Region 1 assemblies are each in a cell with a stainless steel L-insert.
Every cell without a stainless steel L- TC I insert must contain a NETCO-SNAP-IN insert.Array C Two Region 2 assemblies (2) checkerboarded with one Region 3 assembly 2 X (3) and one blocked cell (X). The Region 2 assemblies are each in a ....... ..............
cell with a stainless steel L-insert.
The Region 3 assembly is in a 3 2 cell containing a NETCO-SNAP-IN insert.Array 0 One Region 2 assembly (2) checkerboarded with three Region 4 2 4 assemblies (4). The Region 2 assembly and the diagonally located ..Region 4 assembly are each in a storage cell with a stainless steel L-insert. The two storage cells without a stainless steel L-insert 4 4 contain a NETCO-SNAP-IN insert.Array E 5 5 Four Region 5 assemblies (5). Two storage cells contain a stainless steel L-insert.
One cell contains a NETCO-SNAP-IN insert. One5 storage cell contains no insert.5 5 Array F 6 6 Four Region 6 assemblies (6). Two storage cells contain a stainless steel L-insert.
The other two cells contain no inserts. 6 6 Notes-1. The shaded locations indicate cells which contain a stainless steel [-insert.2. A blocked cell (X) contains a blocking device and only water in the active fuel region.3. NETCO-SNAP-IN inserts must be oriented in the same direction as the stainless steel [-inserts.
- 4. NETCO-SNAP-IN inserts are only located in cells without a stainless steel [-insert.5. Any cell containing a fuel assembly or a TC may instead be an empty (water-filled) cell in all storage arrays.6.Any storage array location designated for a fuel assembly may be replaced with non-fissile material.PALO VERDE UNITS 1,2,3 371- MNMN O 3.7.17-7 AMENDMENT NO.
Before SFP transition Design Features 4.0 4.0 DESIGN FEATURES (continued) 4.3 Fuel Storage 4.3.1 Criticality 4.3.1.1 The spent fuel storage racks are designed and shall be maintained with: a. Fuel assemblies having a maximum radially averaged U-235 enrichment of 4.80 weight percent;b. keff < 1.0 if fully flooded with unborated water, which includes an allowance for biases and uncertainties as descri bed i n Secti on 9.1I of the UFSAR;c. keff -< 0.-95 if fully flooded with water borated to 900 ppm, which includes an allowance for biases and uncertainties as described in Section 9.1 of the UFSAR.d. A nominal 9.5 inch center-to-center distance between adjacent storage cell locations.
- e. Region 1: Fuel shall be stored in a checkerboard (two-out-of-four) storage pattern. Fuel that qualifies to be stored in Regions 1, 2, 3, or 4 in accordance with Figures 3.7.17-1, 3.7.17-2, or 3.7.17 -3, may be stored in Region 1.f. Region 2: Fuel shall be stored in a repeating 3-by-4 storage pattern in which Region 2 (two-out-of-twelve) assemblies and Region 4 (ten-out-of-twelve) assemblies are mixed as shown in Section 9.1 of the UFSAR. Only fuel that qualifies to be stored in Regions 2, 3, or 4, in accordance with Figures 3.7.17-1, 3.7.17-2, or 3.7.17-3, may be stored in Region 2.g. Region 3: Fuel shall be stored in a four-out-of-four storage pattern. Only fuel that qualifies to be stored in Regions 3 or 4, in accordance with Figures 3.7.17-2 or 3.7.17-3, may be stored in Regi on 3.(conti nued)PALO VERDE UNITS 1,2,3 402AEDETN.1k 4.0-2 AMENDMENT NO.
Before SEP transition Design Features 4.0 4.0 DESIGN FEATURES (continued)
- h. Region 4: Fuel shall be stored in a repeating 3-by-4 storage pattern in which Region 2 (two-out-of-twelve) assemblies and Region 4 (ten-out-of-twelve) assemblies are mixed as shown in Section 9.1 of the UFSAR. Only fuel that qualifies to be stored in Region 4 in accordance with Figure 3.7.17-3 shall be stored in Region 4.4.3.1.2 The new fuel storage racks are designed and shall be maintained with: a. Fuel assemblies having a maximum radially averaged U-235 enrichment of 4.80 weight percent;b. keff 0.95 if fully flooded with unborated water, which includes an allowance for biases and uncertainties as described in Section 9.1 of the UFSAR: c. keff 0.98 if moderated by aqueous foam, which includes an allowance for biases and uncertainties as described in Section 9.1 of the UFSAR; and d. A nominal 17 inch center to center distance between fuel assemblies placed in the storage racks.,.4.3.2 Drainage The spent fuel storage pool is designed and shall be maintained to prevent inadvertent draining of the pool below elevation 137 feet -6 inches.4.3.3 Capacity The spent fuel storage pool is designed and shall be maintained with a storage capacity limited to no more than 1329 fuel assemblies.
PALO VERDE UNITS 1,2,3 403AEDETN.I~
4.0-3 AMENDMENT NO.
After SFP transition Design Features 4.0 4.0 k[DESIGN FEATURES (conti nued)4.3 Fuel Storage 4.3.1 Criticality 4.3.1.1 The spent fuel storage racks are designed and shall be mai ntai ned with : a. Fuel assemblies having a maximum radially averaged U-235 enrichment of 4.65 weight percent;b. keff < 1.0 if fully flooded with unborated water, which includes an allowance for biases and uncertai nti es as descri bed in Secti on 9.1 of the UFSAR;c. keff -< 0.95 if fully flooded with water borated to 1460 ppm, which includes an allowance for biases and uncertainties as described in Section 9.1 of the UFSAR.d. A nominal 9.5 inch center-to-center distance between adjacent storage cell locations.
- e. Fuel assemblies, are classified in Fuel Regions 1-6 as shown in Tables 3.7.17-1 through 3.7.17-5.(conti nued)PALO VERDE UNITS 1,2,340-AMNETNO 4.0-2 AMENDMENT NO.
After SFP transition Design Features 4.0 4.0 DESIGN FEATURES (continued) 4.3.1.2 The new fuel storage racks are designed and shall be maintained with: a. Fuel assemblies having a maximum radially averaged U-235 enrichment of 4.65 weight percent;b. keff 0.95 if fully flooded with unborated water, which includes an allowance for biases and uncertai nti es as descri bed i n Secti on 9.1I of the U FSAR;c. keff 0.98 if moderated by aqueous foam, which i ncl udes an all owance for bi ases and uncertai nti es as described in Section 9.1 of the UFSAR; and d. A nominal 18 inch (east-west) and 31 inch (north-south) center-to-center distance between fuel assemblies placed in the storage racks.4.3.2 Drainage The spent fuel storage pool is designed and shall be maintained to prevent inadvertent draining of the pool below elevation 137 feet -6 inches.4.3.3 Capacity The spent fuel storage pool is designed and shall be maintained with a storage capacity limited to no more than 1329 fuel assemblies.
PALO VERDE UNITS 1,2,340-AMNETNO 2g 4.0-3 AMENDMENT NO.
Before SFP transition Programs and Manuals 5.5 5.5 Programs and Manuals (continued) 5.5.19 Battery Monitoring and Maintenance Program (continued)
- 4. In Regulatory Guide 1.129, Regulatory Position 3, Subsection 5.4.1, "State of Charge Indicator," the following statements in paragraph (d) may be omitted: "When it has been recorded that the charging current has stabilized at the charging voltage for three consecutive hourly measurements, the battery is near full charge.These measurements shall be made after the initially high charging current decreases sharply and the battery voltage rises to approach the charger output voltage." 5. In lieu of RG 1.129, Regulatory Position 7, Subsection 7.6, "Restoration," the following may be used: "Following the test, record the float voltage of each cell of the string." b. The program shall include the following provisions:
- 1. Actions to restore battery cells with float voltage<2.13 V;2. Actions to determine whether the float voltage of the remaining battery cells is 2.13 V when the float voltage of a battery cell has been found to be<2.13 V;3. Actions to equalize and test battery cells that had been discovered with electrolyte level below the top of the plates;4. Limits on average electrolyte temperature, battery connection resistance, and battery terminal voltage;and 5. A requirement to obtain specific gravity readings of all cells at each discharge test, consistent with manufacturer recommendations.
PALO VERDE UNITS 1,2,3 551 MNMN O 9 5.5-19 AMENDMENT NO.
After SFP transition Programs and Manuals 5.5 5.5 Programs and Manuals (continued) 5.5.19 Battery Monitoring and Maintenance Program (continued)
- 4. In Regulatory Guide 1.129, Regulatory Position 3, Subsection 5.4.1, "State of Charge Indicator," the following statements in paragraph (d) may be omitted: "When it has been recorded that the charging current has stabilized at the charging voltage for three consecutive hourly measurements, the battery is near full charge.These measurements shall be made after the initially high charging current decreases sharply and the battery voltage rises to approach the charger output voltage." 5. In lieu of RG 1.129, Regulatory Position 7, Subsection 7.6, "Restoration," the following may be used: "Following the test, record the float voltage of each cell of the string." b. The program shall include the following provisions:
- 1. Actions to restore battery cells with float voltage<2.13 V;2. Actions to determine whether the float voltage of the remaining battery cells is 2.13 V when the float voltage of a battery cell has been found to be<2.13 V;3. Actions to equalize and test battery cells that had been discovered with electrolyte level below the top of the plates;4. Limits on average electrolyte temperature, battery connection resistance, and battery terminal voltage;and 5. A requirement to obtain specific gravity readings of all cells at each discharge test, consistent with manufacturer recommendations.
PALO VERDE UNITS 1,2,35519AEDNT O.49 5.5-19 AMENDMENT NO.
After SFP transition Programs and Manuals 5.5 5.5 Programs and Manuals (continued) 5.5.21 Spent Fuel Storage Rack Neutron Absorber Monitoring Program Certain storage cells in the spent fuel storage racks utilize neutron absorbing material that is credited in the spent fuel storage rack criticality safety analysis to ensure the limitations of Technical Specifications 3.7.17 and 4.3.1.1 are maintained.
In order to ensure the reliability of the neutron absorber material, a monitoring program is provided to confirm the assumptions in the spent fuel pool criticality safety analysis.The Spent Fuel Storage Rack Neutron Absorber Monitoring Program shall require periodic inspection and monitoring of spent fuel pool test coupons and neutron absorber inserts on a performance-based frequency, not to exceed 10 years.Test coupons shall be inspected as part of the monitoring program. These inspections shall include visual, B-IO areal density and corrosion rate.Visual in-situ inspections of inserts shall also be part of the program to monitor for signs of degradation.
In addition, an insert shall be removed periodically for visual inspection, thickness measurements, and determination of retention force.PALO VERDE UNITS 1,2,35520AEDNT O.19 5.5-20 AMENDMENT NO.
Enclosure Description and Assessment of Proposed License Amendment ATTACHM ENT 3 Marked-up Technical Specifications Bases Pages (Pages Provided for Before and After SEP Transition)
B 3.7.15-1 B 3.7.15-2 B 3.7.17-1 B 3.7.17-2 B 3.7.17-3 B 3.7.17-4 B 3.7.17-5 B 3.7.17-6 C IBefore SFP transitionJ Fuel Storage Pool Boron Concentration B 3.7.15 B 3.7 PLANT SYSTEMS B 3.7.15 Fuel Storage Pool Boron Concentration BASES BACKGROUND As described in LCO 3.7.17, 'Spent Fuel Assembly Storage," fuel assemblies are stored in the spent fuel racks in accordance with criteria based on initial enrichment and discharge burnup. Although the water in the spent fuel pool is normally borated to >_ 2150 ppm, the criteria that limit the storage of a fuel assembly to specific rack locations is conservatively developed without taking credit for boron.In order to maintain the spent fuel pool keff < 1.0, a soluble boron concentration of 900 ppm is required to maintain the spent fuel pool keff _< 0.95 assuming the most limiting single fuel mishandling accident.APPLICABLE SAFETY ANALYSES A fuel assembly could be inadvertently loaded into a spent fuel rack location not allowed by LCO 3.7.17 (e.g., an unirradiated fuel assembly or an insufficiently depleted fuel assembly).
Another type of postulated accident is associated with a fuel assembly that is dropped onto the fully loaded fuel pool storage rack or between a rack and the pool walls. These incidents could have a positive reactivity effect, decreasing the margin to criticality.
However, the negative reactivity effect of the soluble boron compensates for the increased reactivity caused by these postulated accident scenarios.
The concentration of dissolved boron in the fuel pool satisfies Criterion 2 of 10 CFR 50.36 (c)(2)(ii).
LCO The specified concentration of dissolved boron in the fuel pool preserves the assumptions used in the analyses of the potential accident scenarios described above. This concentration of dissolved boron is the minimum required concentration for fuel assembly storage and movement within the fuel pool.APPLICABILITY This LCO applies whenever any fuel assembly is stored in the spent fuel pool in order to comply with the TS 4.3.1.1.c design requirement that keff 0.95.(conti nued)PALO VERDE UNITS 1,2,3B37.51RVSO B 3.7.15-1 REVISION 3 Before SFP transition Fuel Storage Pool Boron Concentration B 3.7.15 BASES (continued)
ACTIONS A.1 and A.2 The Required Actions are modified by a Note indicating that LCO 3.0.3 does not apply.When the concentration of boron in the spent fuel pooi is less than required, immediate action must be taken to preclude an accident from happening or to mitigate the consequences of an accident in progress.
This is most efficiently achieved by immediately suspending the movement of fuel assemblies.
This does not preclude the movement of fuel assemblies to a safe position.
In addition, action must be immediately initiated to restore boron concentration to within limit.If moving fuel assemblies while in MODE 5 or 6, LCO 3.0.3 would not specify any action. If moving fuel assemblies while in MODE 1, 2, 3, or 4, the fuel movement is independent of reactor operation.
Therefore, inability to suspend movement of fuel assemblies is not sufficient reason to require a reactor shutdown.SURVEILLANCE SR 3.7.15.1 REQUI REMENTS This SR verifies that the concentration of boron in the spent fuel pool is within the required limit. As long as this SR is met, the analyzed incidents are fully addressed.
The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.REFERENCES
- 1. UFSAR, Section 9.1.2.2. PVNGS Operating License Amendments 82, 69 and 54 for Units 1, 2 and 3, respectively, and associated NRC Safety Evaluation dated September 30, 1994.3. 13-N-001-1900-1221-1, "Palo Verde Spent Fuel Pool Criticality Analysis," ABB calculation A-PV-FE-0106, revision 3, dated January 15, 1999.PALO VERDE UNITS 1,2,3B37152RVSO
.B 3.7.15-2 REVISION Ifer FP ransition Fuel Storage Pool Boron Concentration B 3.7.15 B 3.7 PLANT SYSTEMS B 3.7.15 Fuel Storage Pool Boron Concentration ,an(time.BASES BACKGROUND As dscrbedin.CO 3.7.17. "Spent Fuel Assembly Afuel decieassembliesi tre stored in the spent fuel racks accordance with ,4iteria based on initial enrichment
~discharge burnup' Although the .ater .pent fue pool is .. nomly borte to 2l50 ppm the critei that limt the strg of.. a fue to sp.cifi rac location-s, i soluble boron concentration of m is required to maintain the spent fuel pool kef _< 0.ming the most limiting fuel mishandling accident. 1460 APPLICABLE SAFETY ANALYSES There could also be a misload of multiple fuel assemblies into fuel rack locations not allowed by LCO 3.7.17.A fuel assembly could be inadvertently loaded into a spent fuel rack location not allowed by LCO 3.7.17 (e.g., an unirradiated fuel assembly or an insufficiently depleted fuel Another type of postulated accident is a fuel assembly that is dropped onto the fully 1 ~ted fuel pool storage rack or between a rack and the 4mol walls. These incidents could have a positive activity effect, decreasing the margin to criticality.
However, the negative reactivity effect of the soluble boron compensates for the increased reactivity caused by these postulated accident scenarios.
The concentration of dissolved boron in the fuel pool satisfies Criterion 2 of 10 CFR 50.36 (c)(2)(ii).
LCO The specified concentration of dissolved boron in the fuel pool preserves the assumptions used in the analyses of the potential accident scenarios described above. This concentration of dissolved boron is the minimum required concentration for fuel assembly storage and movement within the fuel pool.APPLICABILITY This LCO applies whenever any fuel assembly is stored in the spent fuel pool in order to comply with the TS 4.3.1.1.c design requirement that keff 0.95.(conti nued)PALO VERDE UNITS 1,2,3B37151RVSO B 3.7.15-1 REVISION IAfter SFP transition Fuel Storage Pool Boron Concentration B 3.7.15 BASES (continued)
ACTIONS A.1 and A.2 The Required Actions are modified by a Note indicating that LCO 3.0.3 does not apply.When the concentration of boron in the spent fuel pool is less than required, immediate action must be taken to preclude an accident from happening or to mitigate the consequences of an accident in progress.
This is most efficiently achieved by immediately suspending the movement of fuel assemblies.
This does not preclude the movement of fuel assemblies to a safe position.
In addition, action must be immediately initiated to restore boron concentration to within limit.If moving fuel assemblies while in MODE 5 or 6, LCO 3.0.3 would not specify any action. If moving fuel assemblies while in MODE 1, 2, 3. or 4, the fuel movement is independent of reactor operation.
Therefore, inability to suspend movement of fuel assemblies is not sufficient reason to require a reactor shutdown.SURVEILLANCE SR 3.7.15.1 REQUIREMENTS This SR verifies that the concentration of boron in the spent fuel pool is within the required limit. As long as this SR is met, the analyzed incidents are fully addressed.
The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.REFERENCES
- 82. 69 3nd{ 51 fo-'r""Criticality Safety Analysis for Palo Verde Nuclear Generating Station Units 1, 2, and 3" (Proprietary), WCAP-1 8030-P, Revision 0, September 2015.PALO VERDE UNITS 1,2,3 B 3.7.15-2 REVISION I Before SFP transition I Spent Fuel Assembly Storage B 3.7.17 B 3.7 PLANT SYSTEMS B 3.7.17 Spent Fuel Assembly Storage BAS ES BAGCKGROUND The spent fuel storage is designed to store either new (nonirradiated) nuclear fuel assemblies, or burned (irradiated) fuel assemblies in a vertical configuration underwater.
The storage pool was originally designed to store up to 1329 fuel assemblies in a borated fuel storage mode.The current storage configuration, which allows credit to be taken for boron concentrati.on, burnup, and decay time, and does not require neutron absorbing (boraflex) storage cans, provides for a maximum storage of 1209 fuel assemblies in a four-region configuration.
The design basis of the spent fuel cooling system, however, is to provide adequate cooling to the spent fuel during all operating conditions (including full core offload) for only 1205 fuel assemblies (UFSAR section 9.1.3). Therefore, an additional four spaces are mechanically blocked to limit the maximum number of fuel assemblies that may be stored in the spent fuel storage pool to 1205.Region 1 is comprised of two 9x8 storage racks and one 12x8 storage rack. Cell blocking devices are placed in every other storage cell location in Region 1 to maintain a two-out-of-four checkerboard configuration.
These cell blocking devices prevent inadvertent insertion of a fuel assembly into a cell that is not allowed to contain a fuel assembly.Region 3 is comprised of three 9x8 storage racks and one 9x9 storage rack in Units 2 and 3. Region 3 is comprised of four 9x8 storage racks and one 9x9 storage rack in Unit 1. Since fuel assemblies may be stored in every Region 3 cell location, no cell blocking devices are installed in Region 3.Regions 2 and 4 are mixed and are comprised of seven 9x8 storage racks and three 12x8 storage racks in Units 2 and 3, Regions 2 and 4 are mixed and are comprised of six 9x8 storage racks and three 12x8 storage racks in Unit 1. Regions 2 and 4 are mixed in a repeating 3x4 storage pattern in which two-out-of-twelve cell locations are designated Region 2 and ten-out-of-twelve cell locations are designated Region 4 (see UFSAR Figures 9.1-7 and 9.1-7A). Since fuel assemblies may be stored in every Region 2 and Region 4 cell location, no cell blocking devices are installed in Region 2 and Region 4.(conti nued)PALO VERDE UNITS 1,2,3B371-1RVSO B 3.7.17-i REVISION
[Before SFP transition Spent Fuel Assembly Storage B 3.7.17 BASES BACKGROUND (conti nued)The spent fuel storage cells are installed in parallel rows with a nominal center-to-center spacing of 9.5 inches. This spacing, a minimum soluble boron concentration of 900 ppm, and the storage of fuel in the appropriate region based on assembly burnup in accordance with TS Figures 3.7.17-1, 3.7.17-2, and 3.7.17-3 is sufficient to maintain a keff of0.95 for fuel of original maximum radially averaged enrichment of up to 4.80%.APPLICABLE SAFETY ANALYSES The spent fuel storage pool is designed for non-criticality by use of adequate spacing, credit for boron concentration, and the storage of fuel in the appropriate region based on assembly burnup in accordance with TS Figures 3.7.17-1, 3.7.17-2, and 3.7.17-3.
The design requirements related to criticality (TS 4.3.1.1) are keff < 1.0 assuming no credit for boron and keff 0.95 taking credit for soluble boron. The burnup versus enrichment requirements (TS Figures 3.7.17-1, 3.7.17-2, and 3.7.17-3) are developed assuming keff < 1.0 with no credit taken for soluble boron, and that keff 0.95 assuming a soluble boron concentration of 900 ppm and the most limiting single fuel mishandling accident.The analysis of the reactivity effects of fuel storage in the spent fuel storage racks was performed by ABB-Combustion Engineering (CE) using the three-dimensional Monte Carlo code KENO-VA with the updated 44 group ENDF/B-5 neutron cross section library. The KENO code has been previously used by CE for the analysis of fuel rack reactivity and have been benchmarked against results from numerous critical experiments.
These experiments simulate the PVNGS fuel storage racks as realistically as possible with respect to parameters important to reactivity such as enrichment and assembly spacing.The modeling of Regions 2, 3, and 4 included several conservative assumptions.
These assumptions neglected the reactivity effects of poison shims in the assemblies and structural grids. These assumptions tend to increase the calculated effective multiplication factor (keff) of the racks. The stored fuel assemblies were modeled as CE 16x16 assemblies with a nominal pitch of 0.5065 inches between fuel rods, a fuel pellet diameter of 0.3255 inches, and a UO(2)density of 10.31 g/cc.(conti nued)PALO VERDE UNITS 1,2,3B37172RVSO 3 B 3.7.17-2 REVISION 3 IBefore SFP transition]
Spent Fuel Assembly Storage B 3.7.17 BASES APPLICABLE SAFETY ANALYSES (conti nued)KENO-Va calculations were used to construct curves of burnup versus initial enrichment for decay times in 5 year increments from 0 to 20 years for both Regions 3 and 4 (TS Figures 3.7.17-2 and 3.7.11-3) such that all points on the curves produce a keff value (including all biases and uncertainties) of < 1.0 for unborated water. Biases associated with methodology and water temperature were included, and uncertainties associated with methodology, KENO-Va calculation, fuel enrichment, fuel rack pitch, fuel rack and L-insert thickness, pellet stack density, and asymmetric fuel assembly loading were included.
KENO-Va calculations were also performed to determine the soluble boron concentration required to maintain the spent fuel pool keff (including all biases and uncertainties) 0.95 at a 95% probability/95%
confidence level. A soluble boron concentration of 900 ppm is required to assure that the spent fuel pool keff remains 0.95 at all times. This soluble boron concentration accounts for the positive reactivity effects of the most limiting single fuel mishandling event and uncertainties associated with fuel assembly reactivity and burnup. This method of reactivity equivalencing has been accepted by the NRC (Reference
- 3) and used for numerous other spent fuel storage pools that take credit for burnup, decay time, and soluble boron.Most abnormal storage conditions will not result in an increase in the keff of the racks. However, it is possible to postulate events, with a burnup and enrichment combination outside of the acceptable area in TS Figure 3.1.17-1, or with a burnup, decay time, and enrichment combination outside of the acceptable area in IS Figures 3.7.17-2 or 3.7.17-3, which could lead to an increase in reactivity.
These events would include an assembly drop on top of a rack or between a rack and the pool walls, or the misloading of an assembly.
For such events, partial credit may be taken for the soluble boron in the spent fuel pool water to ensure protection against a criticality accident since the staff does not require the assumption of two unlikely, independent, concurrent events (double contingency principle).
Although a soluble boron concentration of only 900 ppm is required to assure that keff remains 0.95 assuming the single most limiting fuel mishandling event, TS 3.7.15 conservatively requires the presence of 2150 ppm of soluble boron in the spent fuel pool water. As such, the reduction in keff caused by the required soluble boron concentration more than offsets the reactivity addition caused by credible accidents, and the staff criterion of keff 0.95 is met at all times.(conti nued)PALO VERDE UNITS 1,2,3B37.-3RVSO 3 B 3.7.17-3 REVISION 3 IBefore SEP transitionI Spent Fuel Assembly Storage B 3.7.17 BASES APPLICABLE The criticality aspects of the spent fuel pool meet the SAFETY ANALYSES requirements of General Design Criterion 62 for the (continued) prevention of criticality in fuel storage and handling.The spent fuel pool heat load calculations were based on a full pool with 1205 fuel assemblies.
From the spent fuel pool criticality analysis, the number of fuel assemblies that can be stored in the four-region configuration is 1209 fuel assemblies.
The design basis of the spent fuel cooling system, however, is to provide adequate cooling to the spent fuel during all operating conditions (including full core offload) for only 1205 fuel assemblies (UFSAR section 9.1.3). Therefore, an additional four spaces are mechanically blocked to limit the maximum number of fuel assemblies that may be stored in the spent fuel storage pool to 1205.The original licensing basis for the spent fuel pool allowed for spent fuel to be loaded in either a 4x4 array or a checkerboard array, depending on the use of borated poison.A fuel handling accident was assumed to occur with maximum loading of the pool. The fuel pool rack construction precludes more than one assembly from being impacted in a fuel handling accident.
The UFSAR analysis conclusion regarding the worst scenario for a dropped assembly (in which the horizontal impact of a fuel assembly on top of the Spent fuel assembly damages fuel rods in the dropped assembly but does not impact fuel in the stored assemblies) continues to be limiting.The spent fuel assembly storage satisfies Criterion 2 of 10 CFR 50.36 (c)(2)(ii).
LCO The restrictions on the placement of fuel assemblies within the spent fuel pool, according to Figures 3.7.17-1, 3.7.17-2, and 3.7.17-3 in the accompanying LCO, ensures that the keff of the spent fuel pool will always remain < 1.0 assuming the pool to be flooded with unborated water. The restrictions are consistent with the criticality safety analysis performed for the spent fuel pool according to Figures 3.1.17-1, 3.1.17-2, and 3.7.17-3 in the accompanying LCO.Specification 4.3.1.1 provides additional details for fuel storage in each of the four Regions.(conti nued)PALO VERDE UNITS 1,2,3B371-4RVSO B 3.7.17-4 REVISION IBefore SFP transition Spent Fuel Assembly Storage B 3.7.17 BASES APPLICABILITY This LCO applies whenever any fuel assembly is stored in the spent'fuel pool.ACTIONS A. 1 Requi red Acti on A.1 i is modi fi ed by a Note i ndi cating that LCO 3.0.3 does not apply.When the configuration of fuel assemblies stored in the spent fuel pool is not in accordance with Figures 3.7.17-1, 3.7.17-2, and 3.7.17-3, immediate action must be taken to make the necessary fuel assembly movement(s) to bring the configuration into compliance with Figures 3.7.17-1, 3.7.17-2, and 3.7.17-3.If moving irradiated fuel assemblies while in MODE 5 or 6, LCO 3.0.3 would not specify any action. If moving irradiated fuel assemblies while in MODE 1, 2, 3, or 4, the fuel movement is independent of reactor operation.
Therefore, in either case, inability to move fuel assemblies is not sufficient reason to require a reactor shutdown.SURVEILLANCE SR 3.7.17.1 REQU IREMENTS This SR verifies by administrative means that the initial enrichment and burnup of the fuel assembly is in accordance with Figures 3.7.17-1, 3.7.17-2, and 3.7.17-3 in the accompanying LCO and Speci fi cati on 4.3.1.1.To manually determine the allowed SFP region for a fuel assembly, the actual burnup is compared to the burnup requirement for the given initial enrichment and appropriate decay time from Figure 3.7.17-1, 3.7.17-2, or 3.7.17-3.
If the actual burnup is greater than or equal to the burnup requirement, then the fuel assembly is eligible to be stored in the corresponding region. If the actual burnup is less than the burnup requirement, then the comparison needs to be repeated using another curve for a lower numbered region. Note the fol lowing : (continued)
PALO VERDE UNITS 1,2,3B371-5RVSO 3 B 3.7.17-5 REVISION 3 SBefore SFP transition I Spent Fuel Assembly Storage B 3.7.17 BASES SURVEILLANCE
- that a fuel assembly that does not meet the burnup REQUIREMENTS requirement for Region, 2 must be stored in Region 1, (conti nued)* that any fuel assembly may be stored in Region 1,* that any fuel assembly may be stored in a lower numbered region than the region for which it qualifies because burnup requirements decrease as region numbers decrease (refer also to Tech Spec 4.3.1.1),*and that comparing actual burnup to the burnup requirement for zero decay time will always be correct or conservative.
REFERENCES
- 1. UFSAR, Sections 9.1.2 and 9.1.3.2. PVNGS Operating License Amendments 82, 69, and 54 for Units 1, 2, and 3 respectively, and associated NRC Safety Evaluation, dated September 30, 1994.3. Letter to T. E. Collins, U.S. NRC to T. Greene, WOG,"Acceptance for Referencing of Licensing Topical Report WCAP-14416-P, Westinghouse Spent Fuel Rack Methodology (TAC NO. M93254)", October 25, 1996.4. 13-N-001-1900-1221-1, "Palo Verde Spent Fuel Pool Criticality Analysis," ABB calculation A-PV-FE-0106, revision 03, dated January 15, 1999.5. Westinghouse letter NF-APS-10-19, "Criticality Safety Evaluation of the Spent Fuel Pool Map with a Proposed Region 3 Increase," dated February 25, 2010.PALO VERDE UNITS 1,2,3 B371- EIIN~B 3.7.17-6 REVISION IAfter SFP transition Spent Fuel Assembly Storage B 3.7.17* The design basis of the spent fuel pool cooling system is to B 3.7 PLANT SYSTEMS IProvide adequate cooling to the spent fuel pool during all Ioperating conditions (including full core offload) for up to 1205 B 3.7.17 Spent Fuel Assembl lfuel assemblies (UFSAR Section 9.1.3).BASES /I l t BACKGROUND The spent fuel is designed ostor2/'2 h4 new (nonirradiated) fuel asse ~lies,--g¢ burned (irradiated) fuel assemblies in Jvertical configuration underwater.
The storage pool sorgnally designed to store up to 1329 fuel assemblies i eae ue trg oe The current ,storg config.. ration.. w4. ..hich allow credit. be taken for boron.conentration burnup,..
an dea time, and..dones mnot requirelm mneutron absorbing (boraflex)v ca-ns provides-ltc a maximu~m stor=ag of 1209Q "fuel assemblies¢ in a four reio configuration=.
.The desig basis of the spent fue,,l cooling syste,-, however, is to provide adequatea cooling1 to the core offload frl on nly 1205 asseamblie (UFSTAR sct,,'io~n a 1-- -'- Therefore, an =additional four.spaces.are.mechanically..
bmlockedn to limitf t-he mai~m~um nulmberA t-hatCell de,,'ie are..pnlaceda in ever... other storagea cell location in Region 1 to mainti-mn a out- of four checkerbar configuration.
blockin devices preventf inadverte4-nt-insertion ofc a fueal into a cell that i not ..alloe to contain a fuel assembly.sto-rr~age r~ack- in UI I-4t- 2 and'- 3. Regio n 3 is co .v-mprise of,- fou"r,,, 9x3 sto"rage,- and oneh 9v9 stor,-age,,-
r,.ack. in- Unit- 1. 4Re fue~l asse. mblies- may be kstoredw in evry,,n, Regio"-n 3 ,.-c ll loation4,'
no cell blocking devices are installed in Region 3.Regions 2 andi A are mived- rand4 are seve\-n 9v8ra-,ck and three 12x r~acks,, in Un-it -2c and 3, Regio~ns 2 and I nar m'ixed and are of sixv 9v8 storag are mixed, in a repatnga-n 3vA storwage attern in w.hic-h twor out of cell locations are designated Region-,, 2 and ten out of twelve! cell. locations
..... designated Region I (see-, UFS^R Figres 1 "7 and 9.1 7A). S1nc ..-fuel asse-,mblies may, be stored in e,.ery Region 2 and Region '1 cell location no.. cel blocking de,,ice ..... installed in Region 2 =an Region I secontinued PALO VERDE UNITS 1,2,3B3.171RVSO 4 B 3.7.17-1 REVISION Insert I for TS Bases 3.7.17 page B 3.7.17-1 The spent fuel storage cells are installed in parallel rows with a nominal center-to-center spacing of 9.5 inches. This spacing, a minimum soluble boron concentration of 1460 ppm, the use of neutron-absorbing panels, and the storage of fuel in the appropriate region based on fuel assembly initial enrichment, discharge burnup, and decay time in accordance with TS Tables 3.7.17-1 through 3.7.17-5 is sufficient to maintain keff< 0.95 for fuel of initial maximum radially averaged enrichment of up to 4.65 wt%.Disused CEAs, in-core instruments, and other material is stored in trash cans. A trash can may be stored in any location that is approved to store a fuel assembly.
No special nuclear material (SNM) may be stored in a trash can.
IAfter SFP transitionI Spent Fuel Assembly Storage B 3.7.17 BASES BAAC"KG RUDf/ KF (conti4nued)4 The, fucl storage cclls installed4 in parallel rows, wit4h a nominal cntne-r to nnl cnter-spacing of 9. innHch. ThL spacingn a slubklc, boron co'-nceant-ration of ppm assembkl,, b 'urnp i/n accordance4
... th. 4 -.T--
3.17 1,_3.7.17 2, andl 3.T.1T 3 is sufficie'-4nt-to maintain keTT-of0.9 foK r, fueJl ofz orgna,, nl maxi{mum radia,4:lly.
ofi upn to- /1.0/.APPLICABLE Thc ...nt fu-el strge....
pol is dest"igned,4 for' nonrl SAFETY ANALYSES critcalty' by ' "se of spacing, credi.t- for. boron and3'1 tlhe "fueali4n theappn,'nropriate-TS: FZ{TiguresC 3.7.17 1, 3.7.17 2 3,4 an 37117 3. Th& -.de-l-gm ke, i 1.0 assu-mi ng no credi t for boron and _4 Q--_c_taki"ng credit- fo sonclulehl boron. The burnup The analysis of-effe'tII'ct of fuel_ storage9 in Insrt 1th fue,,l soag k-vr acks whas pefome by BB :Combust/ion Engineering.- (CE)usin the threen,44n dime.'-,.,-nsinlMot-arocd hEO A with th-e updated #an-1f~ ENDF/ neutron:-l cros'/Insert section. library.....Th E.....code ha been.... previously..used.by L I~~~C for the, analysis of~n fuel 4-rack' rea~ctivity~ haver ben.,,-,.,, bencmark\I d agaih l-ns resu:3lts, fAn-,rom numerousR c: nritical storagem, ra:lck as as possblel w.,ith respectn4 to The of DRegio~ns
- 2. :3,nd4 '1 inc'ludeda
- seeal, neglecPted4 theTheeassumptoions',n
'tend4 to increnase tlhe calculated3-n, e,-ffective,,, multlication-:34,-,, facto',r-, (keT of t',-he SThe stored fuel assembli!es w.ere modeled as CE 1..x16v assembl'ies writh a3 nomilnal p'itc-h of 0.5065 "inches betweean fueal rods,,4 a fuerl pellet-of 0.325 inhe,,",ac nd a3 llO2)density of 10.31 g/cc.(conInIued) lbU PALO VERDE UNITS 1,2,3B371-2RVSO B 3.7.17-2 REVISION %
Insert 2 for TS Bases 3.7.17 page B 3.7.17-2 The nuclear criticality safety analysis in References 1 and 2 considered the following reactivity-increasing accidents:
- Misload of a single assembly into an unacceptable storage location* Multiple assemblies misloaded in series due to a common cause* Spent fuel pooi temperature outside the allowable operating range* Dropped and misplaced fresh fuel assembly* Seismic event* Inadvertent removal of a NETCO-SNAP-IN~
rack insert In each case, the spent fuel assembly storage met the requirements of 10 CFR 50.68(b)(4).
Thus, the spent fuel storage facility is designed for noncriticality by use of adequate spacing, and neutron absorbing panels considering initial enrichment, fuel burnup, and decay time.
IAfter SEP transitionI Spent Fuel Assembly Storage B 3.7.17 BASE[S APPLICTfABLEI ,AI'E'TV ,ANAILVSES versus. initial1 enrichmen fo dccay tims in 5 ..e.r increments from 0 to 20 years for both Regions 3 and 4 (TS 3.7.17 2 and 3.7.17 3) uciih that- all poimnt-s orn t'he curves a kr valu(inc/mludingr all bi.ases .and u-ncertainties) of 1.0 for.. unboratcd wae...assccriated,-
wit4h met-hodo"log nd wm- .ater, tenmpr~n3-atre were inluedIcl...
and unerante assuciaedtainthiethodologyat QKEO nr',a calculation fuel, enrichme, nt,l fue rckl pitch, fuel.-,
requret maintainthespenful pool1 ....he1,, (Incui ngm'-4m' allC biasesP un'ertaF'inties) 4h~
95 r, obabiCll ty/95,%PIP confidene-h.-
leve.3P ,--A,,4! soluble, boromn H co4ncentratmiokn of,900ppm i eurdt sueta h pn Most' abnr,,,,mal storage, will not-H resuclt in an increasea in the ke, of te racks.~ Howevr., , it- is possibletlhe aceptab-3le area in TS 3.7.17 1.',, orwith t-he accepntable area in TS Figuresc 3.7.17 2 or 3.7.17 3, w.hich could latonincrease inreactivity.
These evnt wul includen an asse.,,mbly.
dropr- topof a rac orI a rack,the pool walls, or the misloading of a~n assembly.
Fo boron.in the spent fuel poo water to ensure..protection requ~l'ire tlhe of'- tl-wo' unlikenly, indepndnt,~mn solubl boron concentratio of only 900 ppm is require tom assure. that h, remain 4o 0.9 assuming.
t£. he single. mostfuel poom'l water.P As schl," t-he reductiofnm in h,, Pcause by%1 required4 sof' lub~le boroPn concentration'{' t'han staff criterion of ke, _ 0.9 is met at all times.PALO VERDE UNITS 1,2,3B37.73RVSO B 3.7.17-3 REVISION %
IAfter SFP transitionI Spent Fuel Assembly Storage B 3.7.17 BASES ADDI The criticlity a.pect. of t he spent fuel pool meet the The spent poo,, heat load, calculations w;ere based on a fu, ll pool 1205F fuenl assembl-ies.
Frm t-rhe spent fueil pool0+', criticalt nalsiThem Mnumber~-
of fueln assemblies, thatl can beC in~ thel~ four- reginl =configuatin i 1209Efue asse--imblies.
basi oMf thel spntr fuel are mehnhmica=lly blockedH to -imait the mav-imum number oxf fuenl assemblies t-hat- mayl be3 sltoreda -in the spntl fueil pool lomading of the pool. The fuel pool rac~k coinsrctmlion than. oine assmbl fromA being, impacted uiona ful andrling ah .c¥cieant, Th UFSAR analysnis 3cncnml u -ion .h regardn g worst-f scnai for, a dropped assembly (il-n whch-,ni fuenl assembly, damages fuel rodsc in the dropped~,-
assembly, not -imnrImr"'
fuel "in tlhe st-ored assemhliesncontinues,n to The spent fuel assembly storage satisfies Criterion 2 of 10 CFR 50.36 (c)(2)(ii).
I abes3..1-Itheog spent7- fulpoa awyeand Fi1u0e3assuming the.LCOare rsictonsitn woh the crtcli~~ety safetye anlslies performe frthe spent f el pool, according to Figure 3.7..... 1.....and 3..17 3in in theaccmpanying LCO,.nue ht h efo Specification 4.3.1.1 provides additional details for fuel storage in each of the Regios (conti nued)PALO VERDE UNITS 1,2,3 B 3.7.17-4 REVISION IAfter SFP transition Spent Fuel Assembly Storage B 3.7.17 BASES APPLICABILITY This LCO applies whenever any fuel assembly is stored in the spent fuel pooi.ACTIONS A.1 I Tables 3.7.17-1 through 3.7.17-5 and Figure 3.7.17-1'-7 I v w Required Action A.1 is modified by a Note indicating t LCO 3.0.3 does not apply. /When the configuration of fuel assemblies s ed the spent fuel pool is not in accordance with Fgrs 3.7.17 2, and 3.7.17 3, immediate action must; taken to make the necessary fuel assembly movement(s) bring the configuration into compliance with N7!3.7.17-2, and 3.7.17 3.If moving irradiated fuel assemblies while in MODE 5 or 6, LCO 3.0.3 would not specify any action. If moving irradiated fuel assemblies while in MODE 1. 2. 3. or 4, the fuel movement is independent of reactor operation.
Therefore, in either case, inability to move fuel assemblies is not sufficient reason to require a reactor shutdown.SURVEILLANCE REQU IREMENTS This SR administrative means that the initial enrichment of the fuel assembly is in accordance with 3 71!7 1 3_717 2. and 3_7.7 3 in the ofTbe371-thurough 3..1-5 performanc ofh thi SR.....
-ensure compliance with Specification 4.3.1.1. l(ot ud PALO VERDE UNITS 1,2,3B371-5RVSO B 3.7.17-5 REVISION %
IAfter SFP transitionI Spent Fuel Assembly Storage B 3.7.17 BASES requ~lirement-fo~r 2 must e'-h sltored a-in Regnion 1, (continRued)
- that anyJ fuel assembly/
may/ be stored a-in aRegion 1, 4-tht- any\ "fuel assembly, may be stored -in a loer, nuimbereda regionn t-han the rego w: .h-ic-h it- qalifies bc au-se~~~~REFERENE
.USR aSo oTc peci0n 9 1.3 n .1.3),* n thatI_ act al bur ,nup '- to the burnup -,A+," reurmn o eodcytiewllasb orc 4.- 13g_ N 001190 1221 1, "Palo,," \/erdeH Spent- Fuel Pool/Cr'4-it,-1icality Analysi,c " ,ABBclulto PV FE/ 0106, Evaluatio-4n ofc the Spntn4 Fuel Pool Map w.,it-h a Proposedca Reg-,ionr 3 Inrease, ac" datedH Febrary, 25-, 2010.-"Criticality Safety Analysis for Palo Verde Nuclear Generating Station Units 1, 2, and 3" (Proprietary), WCAP-1 8030-P, Revision 0, September 2015.PALO VERDE UNITS 1,2,3 B 3.7.17-6 REVISION Enclosure Description and Assessment of Proposed License Amendment ATTACHMENT 4 List of Regulatory Commitments Enclosure Description and Assessment of Proposed License Amendment Attachment 4 List of Regulatory Commitments Reuaor omimn Due Date/Milestone 1. APS will implement procedural controls to require verification that fresh fuel assemblies are not placed face-adjacent to one another before completing a fuel move.2. The transition to the new SEP configuration will be completed in all three units in accordance with the Spent Fuel Pool Transition Plan within two years of the NRC approval date of the amendment or by December 31, 2019, whichever is later.Within the requested 90-day implementation period following NRC approval.Within two years of the NRC approval date of the amendment or by December 31, 2019, whichever is later.
Enclosure Attachment 8 contains PROPRIETARY information to be withheld under 10 CFR 2.390 10 CFR 50.90 Maria L. Lacal Vice President, Nuclear s~ale Regulatory
&OversightVed Nuclear Generating Station P.O. Box 52034 Phoenix, AZ 85072/° Mail Station 7605 Tel 623.393.6491 102-07149-M LL/TNW November 25, 2015 U. S. Nuclear Regulatory Commission ATI-N: Document Control Desk Washington, DC 20555-0001
Dear Sirs:
Subject:
Palo Verde Nuclear Generating Station (PVNGS)Units 1, 2, and 3 Docket Nos. STN 50-528, 50-529, and 50-530 License Amendment Request to Revise Technical Specifications to Incorporate Updated Criticality Safety Analysis In accordance with the provisions of Section 50.90 of Title 10 of the Code of Federal Regulations (10 CFR), Arizona Public Service Company (APS) is submitting a request for a license amendment to revise the Technical Specifications (TS) for Palo Verde Nuclear Generating Station Units 1, 2, and 3. The proposed amendment would modify TS requirements to incorporate the results of an updated criticality safety analysis for both new and spent fuel storage.The enclosure to this letter provides a description and assessment of the proposed changes including a technical evaluation, a regulatory evaluation, a significant hazards consideration, and an environmental consideration.
The enclosure also contains eight attachments.
Attachment 1 provides the marked-up existing TS pages. Attachment 2 provides the revised (clean) TS pages. Attachment 3 provides the marked-up TS Bases pages to show the proposed changes.This submittal contains new regulatory commitments (as defined by NEI 99-04, Guidelines for Managing NRC Commitment Changes, Revision 0) to be implemented, which are identified in Attachment
- 4. Attachment 5 provides a non-proprietary version of the criticality safety analysis.
Attachment 6 provides a material qualification report for NETCO-SNAP-IN neutron absorbing spent fuel pool rack inserts.Attachment 7 is an affidavit signed by Westinghouse Electric Company LLC that sets forth the basis on which the proprietary information in Attachment 8 may be withheld from public disclosure by the Commission and addresses with specificity the considerations listed in 10 CFR 2.390(b)(4).
Correspondence with respect to the proprietary aspects of Attachment 8 A member of the STAR!S (Strategic Teaming and Resource Sharing) Alliance Callaway *Diablo Canyon " Palo Verde
- Wolf Creek Attachment 8 transmitted herewith contains PROPRIETARY information.
When separated from Attachment 8, this transmittal document is decontrolled.
102-07149-M LL/TNW A-TEN: Document Control Desk U. S. Nuclear Regulatory Commission LAR to Incorporate Updated Criticality Safety Analysis in TS Page 2 or the supporting Westinghouse affidavit should reference Westinghouse letter number CAW-15-4271 and be addressed to James A. Gresham, Manager, Regulatory Compliance, Westinghouse Electric Company, 1000 Westinghouse Drive, Building 3 Suite 310, Cranberry Township, Pennsylvania 16066.Attachment 8 is the Criticality Safety Analysis for Palo Verde Nuclear Generating Station Units 1, 2, and 3, WCAP-18030-P (Proprietary), which contains information proprietary to Westinghouse Electric Company LLC.A public pre-submittal meeting was held with the NRC on May 11, 2015 (Agency Document Access and Management System [ADAMS] accession number ML15140A314) to discuss the criticality safety analysis performed in support of this license amendment request. A follow-up public conference call to address action items from the May 11, 2015, pre-submittal meeting was held on September 1, 2015 (ADAMS accession number ML15286A028).
In accordance with the PVNGS Quality Assurance Program, the Plant Review Board and the Offsite Safety Review Committee have reviewed and approved the proposed amendment.
By copy of this letter, this license amendment request is being forwarded to the Arizona Radiation Regulatory Agency in accordance with 10 CFR 50.91(b)(1).
APS requests approval of the proposed license amendment by October 1, 2017, and will implement the TS amendment within 90 days following NRC approval.
This request is necessary to complete the Spent Fuel Pool Transition Plan by the end of 2019.Should you have any questions concerning the content of this letter, please contact Thomas Weber, Department Leader, Nuclear Regulatory Affairs, at (623) 393-5764.I declare under penalty of perjury that the foregoing is true and correct.Executed on z 2-< .(Date)Sincerely, M LL/TN W/J R/af
Enclosure:
Description and Assessment of Proposed License Amendment cc: M. L. Dapas NRC Region IV Regional Administrator M. M. Watford NRC NRR Project Manager for PVNGS L. J. KIoss NRC NRR Project Manager C. A. Peabody NRC Senior Resident Inspector for PVNGS A. V. Godwin Arizona Radiation Regulatory'Agency (ARRA)T. Morales Arizona Radiation Regulatory Agency (ARRA)
Enclosure Description and Assessment of Proposed License Amendment TABLE OF CONTENTS 1.0
SUMMARY
DESCRIPTION 2.0 DETAILED DESCRIPTION 2.1 Proposed Changes to the Technical Specifications 2.2 Need for Proposed Changes 3.0 TECHNICAL EVALUATION 3.1 Spent Fuel Pool Analysis 3.2 New Fuel Storage and Fuel Transfer Equipment Analysis 3.3 Spent Fuel Pool Transition Plan
4.0 REGULATORY EVALUATION
4.1 Applicable Regulatory Requirements 4.2 Precedent 4.3 Significant Hazards Consideration 4.4 Conclusion 5.0 ENVIRONMENTAL CONSIDERATION
6.0 REFERENCES
ATTACHMENTS
- 1. Marked-up Technical Specifications Pages 2. Revised Technical Specifications Pages (Clean Copy)3. Marked-up Technical Specifications Bases Pages 4. List of Regulatory Commitments
- 5. Criticality Safety Analysis for Palo Verde Nuclear Generating Station Units 1, 2, and 3 (Non-proprietary), WCAP-1 8030-NP, Revision 0, September 2015 6. Material Qualification Report of MAXUS for Spent Fuel Storage, NET-300047-07 Rev 1, November 2015 7. Westinghouse Application for Withholding Proprietary Information from Public Disclosure, CAW-15-4271, September 3, 2015 8. Criticality Safety Analysis for Palo Verde Nuclear Generating Station Units 1, 2, and 3 (Proprietary), WCAP-18030-P, Revision 0, September 2015 LIST OF ACRONYMS ANP AREVA PVNGS Lead Test Assembly Combustion Engineering 16x16 Fuel APS Arizona Public Service Company ENDF Evaluated Nuclear Data File FHE Fuel Handling Equipment IFSR Intermediate Fuel Storage Rack LAR License Amendment Request LER Licensee Event Report NFS New Fuel Storage NGF Combustion Engineering 16x16 Next Generation Fuel PVNGS Palo Verde Nuclear Generating Station SFP Spent Fuel Pool STD Standard Combustion Engineering 16x16 Fuel TS Technical Specification(s)
VAP Value Added Pellet Combustion Engineering 16x16 Fuel ii Enclosure Description and Assessment of Proposed License Amendment 1.0
SUMMARY
DESCRIPTION The proposed amendment would revise Palo Verde Nuclear Generating Station (PVNGS)Renewed Operating License Nos. NPF-41, NPF-51, and NPF-74 to amend the Technical Specifications (TS) to incorporate the results of the updated criticality safety analysis, WCAP-1 8030-P (Reference 6.1). The proposed amendment will correct a non-conservative TS regarding NRC approved License Amendment Number 125, which describes the current licensing bases for the criticality safety analysis for PVNGS. This is further discussed in Section 2.2 of this proposed amendment.
This enclosure provides a description and assessment of the proposed changes including a technical evaluation, a regulatory evaluation, a significant hazards consideration, and an environmental consideration.
The enclosure also contains eight attachments.
Attachment 1 provides the marked-up existing TS pages. Attachment 2 provides the revised (clean) TS pages. Attachment 3 provides the marked-up TS Bases pages to show the proposed changes.This submittal contains new regulatory commitments (as defined by NEI 99-04, Guidelines for Managing NRC Commitment Changes, Revision 0) to be implemented, which are identified in Attachment
- 4. Attachment 5 provides a non-proprietary version of the criticality safety analysis.Attachment 6 provides a material qualification report for NETCO-SNAP-IN neutron absorbing spent fuel pool rack inserts.Attachment 7 is an affidavit signed by Westinghouse Electric Company LLC that sets forth the basis on which the proprietary information in Attachment 8 may be withheld from public disclosure by the Commission and addresses with specificity the considerations listed in 10 CFR 2.390(b)(4).
Attachment 8 is the Criticality Safety Analysis for Palo Verde Nuclear Generating Station Units 1, 2, and 3, WCAP-1 8030-P (Proprietary), which contains information proprietary to Westinghouse Electric Company LLC.2.0 DETAILED DESCRIPTION 2.1 Proposed Changes to the Technical Specifications The following specific TS changes are proposed as part of the updated criticality safety analysis.Marked-up TS pages are provided in Attachment 1 and revised (clean) TS pages are provided in Attachment 2.TS 3.7.17, Spent Fuel Assembly Storage-Revise the LCO statement and surveillance requirement statement to reflect the updated spent fuel assembly storage requirements resulting from WCAP-18030-P
-Add Tables 3.7.17-1 through 3.7.17-5 to define the Fuel Regions-Replace Figure 3.7.17-1 to define the Allowable Storage Arrays-Delete Figures 3.7.17-2 and 3.7.17-3 (information is replaced by Tables 3.7.17-1 through 3.7.17-5)I Enclosure Description and Assessment of Proposed License Amendment*TS 4.3.1, Criticality
-Change TS 4.3.1.1 .a to read "4.65 weight percent" instead of "4.80 weight percent"-Change TS 4.3.1.1.c to read "1460 ppm" [parts per million] instead of "900 ppm"-Change TS 4.3.1.1.e to refer to Fuel Regions 1 -6as shown in Tables 3.7.17-1 through 3.7.17-5-Delete TS 4.3.1.1.f through 4.3.1.1.h (Fuel Regions are defined in Tables 3.7.17-1 through 3.7.17-5)-Change TS 4.3.1 .2.a to read "4.65 weight percent" instead of "4.80 weight percent"-Change TS 4.3.1 .2.d to replace "A nominal 17 inch center to center..." with"A nominal 18 inch (east-west) and 31 inch (north-south) center-to-center..."*Add new program TS 5.5.21, Spent Fuel Storage Rack Neutron Absorber Monitoring Program (Proposed TS 5.5.20, Risk In formed Completion Time Program, was submitted on July 31, 2015 [Agency Document Access and Management System (ADAMS)accession number ML15218A300])
The TS Bases will also be revised for consistency with the proposed TS changes and with WCAP-1 8030-P. A markup of the TS Bases pages reflecting these changes is provided in Attachment 3 for information.
The proposed TS Bases changes will be implemented in accordance with TS 5.5.14, Technical Specifications (TS) Bases Control Program, at the same time that the TS changes in the approved license amendment request (LAR) are implemented.
2.2 Need for Proposed Changes In March of 2000, the NRC approved License Amendment Number 125, which describes the current licensing bases for the criticality safety analysis for PVNGS. That amendment increased the storage capacity of the spent fuel pools (SFPs) by allowing credit for soluble boron and decay time in the criticality safety analysis.
The amendment also increased the maximum radially averaged fuel enrichment from 4.3 weight percent U-235 to 4.8 weight percent U-235.The methodology that was the basis for that amendment was analogous to that developed in WCAP-14416-P-A, Westinghouse Spent Fuel Rack Criticality Analysis Methodology, which was reviewed and approved by the NRC for use [NRC Letter, T. E. Collins (NRC) to T. Greene (WOG), "Acceptance for Referencing of Licensing Topical Report WCAP-14416-P, Westinghouse Spent Fuel Rack Methodology (TAC No. M93254)," dated October 25, 1996]. In 2001 and 2004, Arizona Public Service Company (APS) submitted license amendment requests (LARs) to the NRC that would support replacement of the steam generators and authorize subsequent operation at an increased maximum power level of 3990 Megawatts thermal (a 2.94 percent increase).
The NRC approved the amendments for Unit 2 in 2003 (ADAMS accession number ML032720538) and for Units I and 3 in 2005 (ADAMS accession number ML0531 30275).In May of 2013, APS submitted Licensee Event Report (LER) 2013-001-00 (ADAMS accession number ML13133A002), which reported that certain impacts to the SFP criticality safety analysis approved in License Amendment 125 had not been considered by APS during the increase in the maximum power level to 3990 MWt. One of the corrective actions in the LER was to revise the SFP criticality safety analysis using updated methodology and input parameters and to 2 Enciosure Description and Assessment of Proposed License Amendment submit a LAR to correct the non-conservative TS. The criticality safety analysis methodology included in this LAR is based upon the most recent NRC approved guidance of Staff Guidance Regarding the Nuclear Criticality Safety Analysis for Spent Fuel Pools, DSS-ISG-201 0-01 (Reference 6.2) and satisfies the actions stipulated in the PVNGS Corrective Action Program and PVNGS LER 201 3-001-00.3.0 TECHNICAL EVALUATION This LAR documents an updated criticality safety analysis for the PVNGS SFPs, new fuel storage (NFS) racks, interim fuel storage rack (IFSR) and fuel handling equipment (FHE).Attachment 8 is the plant-specific Westinghouse WCAP-18030-P, Revision 0, Criticality Safety Analysis for Palo Verde Nuclear Generating Station Units 1, 2 and 3. A plant-specific NETCO-SNAP-IN 0 material qualification report is also included (Attachment
- 6) since the LAR credits the presence of neutron poisons in the NETCO-SNAP-IN neutron absorber inserts. The main body of the LAR includes descriptions and summary evaluations, while the attached WCAP and NETCO report provide additional details on topics that include computer codes, fuel design history, depletion analysis, criticality analysis, as well as the interface, normal, and accident conditions for PVNGS.The change to TS 4.3.1 .2.d regarding NFS rack spacing is proposed to more accurately reflect the NFS as-built drawings and the existing NFS criticality safety analysis of record. The text of TS 4.3.1.2 states "The new fuel storage racks are designed and shall be maintained with...", which refers to physical dimensions.
Therefore, TS 4.3.1 .2'.d must accurately reflect the as-built dimensions that must be maintained throughout the life of the plant. The racks have a nominal 18-inch center-to-center pitch on the short axis (east-west) and a 31-inch center-to-center pitch on the long axis (north-south).
According to both the existing NFS and updated criticality safety analyses, this rack design maintains keff <0.95 during all normal and accident conditions.
3.1 Spent Fuel Pool Analysis Design Approach The existing SEP storage racks are evaluated for the placement of fuel within the storage arrays described in the proposed TS changes. Credit is taken for the negative reactivity associated with burnup and post-irradiation cooling time (decay time). Additionally, some SEP storage arrays credit the presence of the neutron poison in the NETCO-SNAP-IN inserts. Finally, credit is taken for the presence of soluble boron in the SFP for specific conditions.
Compliance for the SFP is demonstrated by establishing limits on the minimum allowable burnup as a function of initial enrichment and decay time for each fuel storage array. A conservative combination of best estimate and bounding values has been selected to model the fuel in the analysis to ensure that fuel represented by the proposed TS is less reactive than the fuel modeled in the analysis.
Therefore, burnup limits will conservatively bound fuel to be stored in the SFP.3 Enclosure Description and Assessment of Proposed License Amendment Acceptance Criteria The objective of the SEP criticality safety analysis is to ensure that the SEP operates within the bounds of 10 CFR 50.68(b)(4):
- If no credit for soluble boron is taken, the keff of the spent fuel storage racks loaded with fuel of the maximum fuel assembly reactivity must not exceed 0.95, at a 95 percent probability, 95 percent confidence level, if flooded with unborated water.* If credit is taken for soluble boron, the keff of the spent fuel storage racks loaded with fuel of the maximum fuel assembly reactivity must not exceed 0.95, at a 95 percent probability, 95 percent confidence level, if flooded with borated water, and the keff must remain below 1 .0 (subcritical), at a 95 percent probability, 95 percent confidence level, if flooded with unborated water.Computer Codes The SEP criticality safety analysis methodology employs the following computer codes and cross-section libraries:
- SCALE 6.1.2 (Reference 6.3) with the 238-group cross-section library based on Evaluated Nuclear Data File (ENDF)/B-VII.
- The two-dimensional transport lattice code PARAGON (Reference 6.4) and its 70-group cross-section library based on ENDF/B-VI.3 PARAGON is used for simulation of in-reactor fuel assembly depletion to generate isotopics for burnup credit applications in the SEP. PARAGON is generically approved for depletion calculations (Reference 6.4) and has been chosen for this spent fuel criticality safety analysis because it has all the attributes needed for burnup credit applications.
There are no Safety Evaluation Report limitations for the use of PARAGON in UO 2 criticality analyses.Additional discussion of the computer codes is provided in Section 2.3 of Attachment 8.Code Validation Process The validation of the ENDF/B-VII library with the SCALE 6.1.2 CSAS5 module is documented in Appendix A of Attachment
- 8. The code validation shows that SCALE 6.1.2 is an accurate tool for calculation of kefffor the applications in this LAR. The benchmark calculations utilize the same computer platform and cross-section libraries that are used for the design basis calculations.
3.1.1 Fuel and Fuel Storage Descriptions Four fuel designs were considered for the criticality safety analysis:* Standard Combustion Engineering 16x16 (STD)* Value Added Pellet Combustion Engineering 16x16 (VAP)* Next Generation Fuel Combustion Engineering 16x16 (NGF)* AREVA PVNGS Lead Test Assembly Combustion Engineering 16x16 (ANP)4 Enclosure Description and Assessment of Proposed License Amendment Section 3.1.2 of Attachment 8 discusses the non-mechanical fuel features which are important to criticality safety and how they impact the number of distinct fuel designs to be considered in the analysis.PVNGS has three SFPs, one in each unit, that are identical in layout. Each SFP contains a single rack design and each SEP is surrounded by a concrete wall with a stainless steel liner.The presence of neither the SEP concrete wall nor liner is credited in the criticality safety analysis.
All storage arrays are conservatively assumed to be radially infinite.The SEP storage racks are made up of individual modules, each of which is an array of fuel storage cells. The storage racks are comprised of 17 modules: twelve 8x9 arrays, four 8x12 arrays, and one 9x9 array. The storage racks are stainless steel honeycomb structures with rectangular fuel storage cells compatible with fuel assembly materials and the spent fuel borated water environment.
The fuel assembly spacing of a nominal 9.5 inches center-to-center distance between adjacent storage cell locations is a minimum value after allowances are made for rack fabrication tolerances and predicted deflections resulting from a safe shutdown earthquake.
The storage racks are designed to maximize the number of storage cells available (minimize storage cell pitch) to be used in a checkerboard pattern of fresh (i.e., new) fuel and empty storage locations.
Trashcans in the SEP store various non-fissile materials, such as discarded control element assemblies and in-core instrumentation tubes, filters, and reconstitution materials.
The SEP racks contain a stainless steel L-insert in every other cell location as shown in PVNGS Updated Final Safety Analysis Report (UFSAR) Figure 9.1-5 to help center the new fuel assemblies within this space. The stainless steel L-inserts are offset 11/16 inch from the cell wall. This reduces the positional uncertainty or the eccentric loading positions of fuel in the racks. There is not currently a surveillance program for the stainless steel L-inserts and one will not be included as part of this submittal for the following reasons:* The industry has a large body of operating experience with stainless steel in SEP environments in which nothing suggests the stainless steel L-inserts would become incapable of performing their function.* The stainless steel L-inserts are analogous to the SEP racks in their function and there is no precedent for requiring a surveillance program of the racks themselves.
A NETCO-SNAP-IN rack insert will not be installed in a cell that contains a stainless steel L-insert.
The NETCO-SNAP-IN rack inserts are described in Attachment 6.3.1.2 Depletion Analysis The depletion analysis is a vital part of any SEP criticality safety analysis which uses burnup credit. The isotopic inventory of the fuel as a function of burnup is generated through the depletion analysis, thus the inputs used need to be carefully considered.
Section 4 of Attachment 8 describes the methods used to determine the appropriate inputs for the generation of isotopic number densities to conservatively bound fuel depletion and storage.Some of the salient points regarding the approach to the depletion analysis include: 5 Enclosure Description and Assessment of Proposed License Amendment* The isotopic number densities generated by the fuel depletion calculations are differentiated by fuel enrichment and decay time after discharge.
The fuel has isotopic number densities which are calculated at enrichments of 3.0, 4.0, and 5.0 weight percent U-235 and decay times of 0, 5, 10, 15, and 20 years.* The soluble boron concentration in the reactor during operation impacts the reactivity of fuel being discharged to the SEP. Boron is a strong thermal neutron absorber and its presence hardens the neutron energy spectrum in the core, creating more plutonium.
It is important to account for the presence of soluble boron during reactor operation to ensure this impact is adequately accounted for in the isotopic generation.
- The fuel temperature during operation impacts the reactivity of fuel being discharged to the SFP. Increasing fuel temperature increases resonance absorption in U-238 due to Doppler broadening which leads to increased plutonium production, increasing the reactivity of the fuel.* The limiting distributed axial burnup profiles are used with the uniform axial burnup profile to calculate the burnup limits.* The limiting axial moderator temperature profiles are used with axially distributed and uniform burnup profiles to calculate the isotopics used in generating the burnup limits.Selecting an appropriate moderator temperature profile is important as it impacts the moderator density and the neutron spectrum during depletion, as discussed in NUREG/CR-6665 (Reference 6.5). An appropriate moderator temperature ensures the impact of moderator density on the neutron spectral effects is bounded, conservatively biasing the isotopic inventory of the fuel.* Burnable absorber usage has been considered for the analysis and conservative assumptions have been utilized to bound the effects of burnable absorbers on fuel isotopics.
The burnable absorbers that have been used include both discrete and integral burnable absorbers.
- The PVNGS fuel management strategy uses radial enrichment zoning to control fuel rod power peaking. Individual assemblies may contain two or three different fuel rod enrichments which are used to control peaking factors. A study was performed to determine the reactivity impact of operating with radial enrichment zoning instead of uniform radial zoning.* All four of the different fuel designs listed in Section 3.1.1 and the conditions in which those designs were operated, or are planned to be operated, were considered in the depletion analysis.
It became clear that the NGF design would be limiting throughout life.Therefore, the NGF design was used to develop the isotopics used in the spent fuel reactivity calculations.
- The parameters used in the final depletion calculations include core operation parameters, fuel assembly dimensions, axial burnup profiles, and moderator temperature profiles.3.1.3 Spent Fuel Pool Criticality Analysis KENO is the criticality code used to determine the absolute reactivity of burned and fresh fuel assemblies loaded in storage arrays. The dimensions and tolerances of the design basis fuel assembly and the fuel storage racks are the basis for the KENO models used to determine the burnup requirements for each fuel storage array, and to confirm the safe operation of the SEP 6 Enclosure Description and Assessment of Proposed License Amendment under normal and accident conditions.
The trashcan characteristics are also modeled in the criticality analysis.Differences between fuel types include changes in fuel rod dimensions, such as pellet and cladding dimensions, and structural components, such as grid material and volumes. Each of the fuel types which have been, or are planned to be, operated at the plant were considered.
The bounding fuel assembly design for the analysis has been determined as described in Section 4.3 of Attachment 8.Burnup Limit Generation To ensure safe operation of the PVNGS SFPs, the analysis defines fuel storage arrays which dictate where assemblies can be placed in the SFPs based on enrichment (weight percent U-235), average burnup (GWd/MTU), and decay time (years) since discharge.
Each assembly in the reactor core depletes under slightly different conditions and can have a different reactivity at the same burnup. This is accounted for in the analysis by using a combination of depletion parameters that together produce a bounding isotopic inventory throughout life. Additionally, while fuel manufacturing is a very tightly controlled process, assemblies are not identical.
Reactivity margin is added to the KENO reactivity calculations for the generation of burnup limits to account for manufacturing deviations.
Assembly storage is controlled by defining allowable storage arrays. An array can only be populated by assemblies of the fuel region defined in the array definition or a lower reactivity fuel region. Fuel regions are defined by assembly burnup, initial enrichment, and decay time.Reactivity biases are known variations between the real and analyzed system, and their reactivity impact is added directly to the calculated keff. Uncertainties are random dispersions around a nominal, measured quantity.
Their impact is added to the calculated keff as the square root of the sum of the squares of the uncertainties.
The following biases and uncertainties are accounted for in the analysis.
A detailed discussion of biases and uncertainties is provided in Section 5.2.3 of Attachment 8.* Reactivity effect of manufacturing tolerances
- Burnup measurement uncertainty
- Depletion uncertainty
- Fission product and minor actinide worth bias* An operational uncertainty of 0.002 Ak is an additional conservativism which is added to the conservatism inherent in the specific power histories from reactor operation* Eccentric fuel assembly positioning
- Uncertainty in the predictive capability of SCALE 6.1.2 and the associated cross-section library* SFP temperature bias within the allowable operating range* Borated and unborated biases and uncertainties 7
Enclosure Description and Assessment of Proposed License Amendment Interface Modelingq Interfaces are the locations where there is a change in either the storage racks or the storage requirements of the fuel in question.
At PVNGS, each SFP has a single storage rack design.Therefore, the only interfaces that exist are those between arrays within the single storage rack design and the only interface conditions that need to be addressed in the analysis are those between different fuel storage arrays. Additional details are provided in Section 5.3 of Attachment 8.Normal Conditions Considered in the Criticality Safety Analysis There are five major types of normal conditions beyond the storage of fuel assemblies that are addressed in the criticality safety analysis.Type 1 conditions involve placement of components in or near the intact fuel assemblies while normally stored in the storage racks. This also includes removal and reinsertion of these components into the fuel when stored in the rack positions using specifically designed tooling.Examples include control element assemblies and guide tube inserts, such as in-core instrumentation tubes. The calculation results show that any components designed to be inserted into an assembly may be stored in a fuel assembly guide tube in the SFP.Type 2 conditions involve evolutions where the fuel assembly is removed from the normal storage rack location for a specific procedure and returned to an allowable cell after completion of the procedure, such as fuel assembly cleaning, inspection, reconstitution, or sipping. These are bounded by the criticality analysis.Fuel assembly reconstitution is a normal condition defined as either pulling damaged fuel pins out of an assembly and reinserting intact pins with less reactivity than the damaged pin, or as removing undamaged pins from a damaged assembly for insertion in a new assembly.Damaged pins will be replaced with stainless steel pins or natural uranium pins. Additional information is provided in Section 5.4.2 of Attachment 8.Type 3 conditions involve inserting components that are not intact fuel assemblies into the fuel storage rack cells. Examples include failed fuel rod baskets and miscellaneous maintenance equipment.
Any components that do not contain fissile materials can replace a fuel assembly of any fuel region in one of the approved storage configurations.
Type 4 conditions include temporary installation of non-fissile components on the rack periphery.
Analyses of the storage arrays contained in the criticality analysis assume an infinite array of storage cells. This assumption bounds the installation of any non-fissile components on the periphery of racks.Type 5 conditions involve miscellaneous conditions that do not fit into the first four normal condition types. Examples include usage of fuel handling tools for their intended purpose, miscellaneous debris under the storage racks, and damaged storage cells.Section 5.4 of Attachment 8 provides further details about normal conditions within the SEP.8 Enclosure Description and Assessment of Proposed License Amendment Soluble Boron Credit In accordance with 10 CFR 50.68, the criticality safety analysis ensures that the maximum calculated ke., including all biases and uncertainties, meet the kef limit of less than 1.0 (subcritical) if flooded with unborated water at a 95 percent probability, 95 percent confidence level. Additionally, the criticality safety analysis demonstrates that if the SFP is flooded with borated water, keff does not exceed 0.95, at a 95 percent probability, 95 percent confidence level.The minimum soluble boron concentration in the SEP to maintain keff < 0.95 for the limiting normal condition including biases, uncertainties, and administrative margin is 450 ppm. During normal operation, TS 3.7.15 requires a soluble boron concentration of> 2150 ppm, but the SFP boron concentration is maintained between 4000 and 4400 ppm in accordance with Technical Requirements Manual T3. 1.104, Borated Sources -Shutdown, and T3. 1.105, Borated Sources-Operating.
Consideration of Criticality Accidents in the SEP The following reactivity-increasing accidents are considered and the analysis results are provided in Section 5.6 of Attachment 8.SAssembly misload into the storage racks -this is the limiting accident which addresses both multiple assemblies being misloaded in series into unacceptable storage locations and the misload of a single assembly into an unacceptable storage location.
A multiple assembly misload is a hypothetical accident where assemblies are misloaded in series due to a common cause. A single assembly misload requires 1100 ppm of boron to maintain keff -< 0.95. A multiple assembly misload requires 1460 ppm of boron to maintain kef<0.9 5.* Spent fuel temperature outside operating range -the SEP is to be operated between 60°F and 180°F, but under accident conditions this temperature could be higher.* Dropped and misplaced fresh assembly -the analysis considers the dropping of the fuel assembly from the fuel handling machine during placement of the fuel assemblies in the racks. The dropped assembly could land horizontally on top of the other fuel assemblies in the rack. Additionally, the analysis considers the possibility to misplace a fuel assembly in a location not intended for fuel.* Seismic event -the SEP racks are seismic category I, designed and built to withstand the maximum potential earthquake stresses in this geographic area. Section 5.6.4 of Attachment 8 provides additional details.o The spent fuel pool racks were originally designed to contain a 188 lb. neutron poison insert in every cell. These neutron poison inserts were never installed.
As the mass of the original design is greater than the mass of the NETCO-SNAP-IN, the original analysis bounds the proposed change.* Inadvertent removal of a NETCO-SNAP-IN rack insert -this is a potential reactivity-increasing accident added by the incorporation of NETCO-SNAP-IN rack inserts. The absence of an insert will cause a reactivity increase due to the loss of neutron absorbing material from the storage array.9 Enclosure Description and Assessment of Proposed License Amendment Fuel used to date at PVNGS has an initial radially averaged enrichment of < 4.55 weight percent. Limiting the maximum radially averaged enrichment to 4.65 weight percent mitigates the consequences of a multiple fuel assembly misload event without impacting operational flexibility.
There is no source of water within the fuel building that could reduce the boron concentration of the spent fuel pool from the value of 2150 ppm (Technical Specification LCO 3.7.15) to 1460 ppm. A fire in the fuel building at elevation 140-ft. is the limiting event for boron dilution and it bounds all normal, seismic, and pipe break scenarios.
The current boron dilution analysis demonstrates that the limiting boron dilution event, which is fighting a hypothetical fire on the 140-ft level of the fuel building, will reduce the boron concentration from the TS limit of 2150 ppm to 1900 ppm. This leaves adequate margin to the 1460 ppm credited by the SEP criticality safety analysis.3.1.4 NETCO-SNAP-IN Rack Inserts This proposed change would credit NETCO-SNAP-IN rack inserts for criticality control in individual SEP storage rack cells to ensure that the requirements of TS 3.7.17 and the associated WCAP-1 8030-P are maintained.
The NETCO-SNAP-IN rack inserts are credited in both the borated and unborated conditions.
The installation of the NETCO-SNAP-IN rack inserts will be controlled as a design change implemented under the provisions of 10 CFR 50.59, Changes, Tests and Experiments, from a structural, seismic, and thermal-hydraulic perspective.
Attachment 6 describes the NETCO-SNAP-IN rack inserts, including their manufacture, an engineering evaluation, and corrosion testing information.
Neutron Absorber Monitoring Program (TS 5.5.21)Arizona Public Service will institute a performance-based long-term surveillance program for the NETCO-SNAP-IN inserts based on manufacturer recommendations, current industry operating experience, NEI guidance, and NRC safety evaluations for other plants that are using neutron absorbing inserts. The long-term surveillance program will evolve as information from these sources changes and as the data from the PVNGS-specific inspections accumulate.
The surveillance program consists of periodic inspections of MAXUS material coupons from surveillance assemblies located in the SEPs and periodic inspection of full length inserts.Coupon Inspections Coupons will be selected from MAXUS production material, identical to the material used to manufacture the inserts, for periodic inspection.
Individual coupons will be subjected to pre-test and post-test characterizations.
As appropriate for each coupon type, coupon characterizations may include visual inspection, high resolution photography, neutron attenuation, stress relaxation, blister and pit characterizations, as well as measurement of thickness, length, width, dry weight, and density.A surveillance assembly to which surveillance coupons are attached, also referred to as a coupon tree, will be placed in the related SEP prior to the first installation campaign of NETCO-SNAP-IN inserts and will reside there to support the monitoring program. Periodically, coupons 10 Enclosure Description and Assessment of Proposed License Amendment will be removed and sent to a qualified laboratory for testing. The coupon trees in the related SFP will contain 48 general coupons, 24 galvanic couple coupons, and 24 bend coupons as described below. They will be situated near the center of the active fuel region to maximize exposure from the surrounding fuel. These coupons will be monitored for changes to their physical properties and for changes to their effective areal density or signs of corrosion, which could indicate neutron absorber material degradation.
The frequency for coupon removal and inspection is shown in Table 1.Table I -Frequency for Coupon Removal from Racks After 10 Years with Coupon Type First Ten Years Acpal efrac General 2 coupons every 2 years 2 coupons every 4 years Bend I coupon every 2 years 1 coupon every 4 years Galvanic couples -304L stainless I couple every 6 years Zircaloy 1 couple every 6 years Inconel 718 1 couple every 6 years General and Galvanic Coupons The general coupons in each SFP are designed to carry the largest number of performance indicators for the insert material.
They will be subject to pre-examination, post-examination, and acceptance testing in accordance with Table 2.The galvanic couple coupons are composed of a MAXUS material coupon placed in contact with Zircaloy, Inconel 718, or 304 stainless steel. Eight of each type will be used to produce a total of 24 galvanic couples per coupon tree assembly in each SEP that will be subject to the inspections listed in Table 2.Bend Coupons Once installed, the NETCO-SNAP-IN rack inserts assume a constant strain condition within the SEP storage rack cell. This compression leads to internal stresses, especially at the bend, that might make the rack inserts susceptible to stress corrosion cracking.
An examination of the literature on the subject indicates in general, that high-purity aluminum and low-strength aluminum alloys are not susceptible to stress corrosion cracking.
However, the surveillance bend coupons placed in the related SEP will be maintained under the same strain conditions as the inserts to provide an indication of unexpected crack phenomena.
These coupons will be held in capsules that compress them from their initial manufactured bend angle to an angle of approximately 90 degrees. Table 3 provides the inspection details.11 Enclosure Description and Assessment of Proposed License Amendment Over time, the MAXUS material is expected to release some of the strain built up during the installation process. The material has a metal matrix core made from 1000 series aluminum and boron carbide powder with an outer clad made from 5052 aluminum.
Existing literature for 1100 series aluminum shows a stress relaxation rate of 58 percent over a period of 20 years. Given that 1100 series aluminum is a softer metal than 5052 aluminum, this rate is considered conservative for the MAXUS material due to the 5052 cladding.
The acceptance criterion for stress relaxation is 60 percent over a 20-year period. This rate will be used when determining minimum retention force requirements for the inserts during installation that will still hold the inserts in place during a seismic event after relaxation has occurred.The bend coupon capsules will be removed from the coupon tree and sent to a qualified laboratory for testing where the coupons will be removed, thus relieving the strain on the coupons and allowing them to return to an angle greater than 90 degrees. The change in internal stress can be correlated to the change in bend angle the coupon forms once it is removed from the strained condition.
Deviation from the pre-characterized value will determine the amount of stress relaxation over the life of the coupon.The stress relaxation rate is not linear, rather it tends to follow a logarithmic pattern. Therefore, a more significant loss of stress is expected in the first few years of exposure, but the relaxation rate becomes asymptotic over a longer period of time.*Table 2 -General and Galvanic Coupon Characterizations TetPre- Post- Acceptance I Rejection Characterization Characterization Criteria Visual (high / Evidence of visual resolution digital indications of photo) performance inhibitors.
Dimension
/Min. thickness:
0.005 inch less than nominal thickness (excluding pit locations).
Thickness change: any change of +0.010 inch I-0.004 inch (excluding pit locations).
- Length change: any change of +/- 0.02 inch* Width change: any change of +/- 0.02 inch Density /Any change of +/- 5%Areal density g/cm 2 Boron-10 minimum loading 12 Enclosure Description and Assessment of Proposed License Amendment Weight loss as determined by dry change of +/- 5%weight Corrosion rate /< 0.05 mil/yr Anomaly characterization
/** To be determined at the time of analysis.* Length and width changes are not applicable for galvanic coupons.** At the presence of anomalies Table 3 -Bend Coupon Characterizations TetPre- Post- Acceptance
/Characterization Characterization Rejection Criteria Visual (high Evidence of visual resolution indications of digital photo) performance inhibitors.
Thickness
' V Min. thickness:
0.005 inch less than nominal thickness (excluding pit locations).
Thickness change: any change of +0.010 inch /-0.004 inch (excluding pit locations).
Bending V /Change in stress greater stress than a rate of 60% over 20 years *Weight loss as determined by dry change of +/- 5%weight Anomaly characterization
- To be determined at the time of analysis.Stress relaxation rate is not linear. Stress relaxation will be re-evaluated if 60% is exceeded.**At the presence of anomalies Full-Length Insert Inspections The combined effects of adequate clearance and infrequent fuel assembly movement will preclude significant wear of the rack insert. However, to verify NETCO-SNAP-IN material performance, a portion of the installed inserts will be subject to in-situ visual inspection and removal for detailed inspection of wear performance.
For in-situ inspections at the frequency described in Table 4, rack inserts will be visually inspected by camera (while remaining in the storage racks) to monitor for physical deformities such as bubbling, blistering, corrosion pitting, 13 Enclosure Description and Assessment of Proposed License Amendment cracking, or flaking. Special attention shall be paid to the development of edge or corner defects.A region of high duty spent fuel storage rack cell locations shall be identified for full insert removal and inspection.
These locations will be monitored for fuel insertion and removal events to ensure that their service bounds that of the general population of storage locations.
Once every 10 years, an insert will be fully removed from this region and will be inspected in accordance with Table 5. The thickness measurements at several locations along the full insert length will be compared with the as-built thickness measurements of the removed insert to verify it has sustained uniform wear over its service life. A visual inspection of the removed insert will also be performed.
Table 4 -Frequency for Full Insert Inspections Inspection Type First Ten Years Acpal efrac In-situ 2 inserts every 2 years 2 inserts every 4 years Removal 1 insert every 10 years Table 5 -Full Insert Removal Inspection Characterizations TetPre- Post- Acceptance I Characterization Characterization Rejection Criteria Visual (high of visual resolution indications of digital photo) performance inhibitors.
Thickness thickness:
0.005 inch less than nominal thickness (excluding pit locations).
Thickness change: any change of +0.010 inch I-0.004 inch (excluding pit locations).
Retention force Retention force less than 50 lbs 14 Enclosure Description and Assessment of Proposed License Amendment 3.1.5 Spent Fuel Pool Configuration Control (Human Performance Enhancements)
APS has a multi-tier defense-in-depth program to prevent and mitigate the severity of a scenario in which multiple fuel assemblies are located in the wrong storage locations.
Specific aspects of this program are described below.Control of Move Sheet Generation
- Detailed administrative procedures for the generation of move sheets and revision of move sheets* Training and qualification of individuals responsible for generation and revision of move sheets for use and implementation of the new TS proposed in this LAR.* Graphical representation of approved arrays in TS 3.7.17 to minimize the probability of misinterpretations
- In accordance with the plant special nuclear materials procedures, APS maintains a Region Specification Document for each SEP to aid the move sheet preparer and verifier in selecting and verifying proper placement of fuel assemblies.
This document tracks the following for every fuel assembly at PVNGS:-Fuel assembly initial enrichment
-Fuel assembly burnup-Limiting Fuel Region in which the fuel assembly can be stored* Every fuel move is checked against the availability of the space and the eligibility of fuel to be stored there* A member of management confirms that each move sheet was generated in accordance with PVNGS procedures.
There are at least three signatures on each move sheet sent to the field, including the move sheet preparer, the verifier, and management.
- A move sheet package is a change document that is used to specify and record changes to plant configuration as it relates to special nuclear material.
A move sheet contains as a minimum, the item to be moved, the "from" location, and the "to" location.
Any number of fuel assemblies can be moved using a single move sheet package.Control of Fuel Movement* The spent fuel handling machine is only operated using approved procedures
- All individuals operating the spent fuel handling machine and acting as independent verifiers are trained in their position, including training on industry operating experience pertaining to fuel misload events* Fuel is moved only as directed by approved move sheets* The correct location of the spent fuel handling machine is independently verified before a fuel move begins* The correct location of the spent fuel handing machine is independently verified before the fuel is placed in the SEP racks* Continuous communication is maintained between the fuel mover and verifier 15 Enclosure Description and Assessment of Proposed License Amendment* The fuel handlers visually confirm that fresh fuel is not placed in locations that are "face-adjacent" to other fresh fuel assemblies Use of Blockingq Devices The assumed limiting misload event at PVNGS involves placing a fuel assembly in a location that is required to be empty per TS 3.7.17. APS uses blocking devices to minimize the probability of a fuel assembly being placed in one of these limiting locations.
Each blockintg device meets the following criteria:* Physically configured to prevent insertion of a fuel assembly in a fuel storage location* Requires special tools to install or remove the blocking device from a storage location* The tool used to grapple a fuel assembly is physically incapable of grappling a blocking device* Designed to preclude falling into a storage location or becoming dislodged during normal operation* Will support the full load of a fuel assembly and the fuel assembly grappling tool* Allows continuous water flow through the storage cell 0 S Is easy to distinguish visually from a fuel assembly Blocking devices are administratively controlled with the same level of rigor as fuel assemblies
- A blocking device move sheet package is a change document used to specify and record changes to plant configuration as it relates to blocking devices. A blocking device move sheet contains, as a minimum, the "from" location and the "to" location of the blocking device. Any number of blocking devices may be moved using a single blocking device move sheet package. Blocking devices will not be moved using the same move sheet package as fuel assemblies.
This restriction prevents a single error from removing a blocking device and placing a fuel assembly in a location that is required to be empty.Confirmation of Configquration Control In accordance with existing procedures, a 100 percent serial number check is performed once per calendar year to ensure that every fuel assembly stored in the SFP matches the SFP maps.This provision limits the amount of time that a misload condition could potentially exist.Mitigqation of a Misload Event If the controls discussed above are insufficient to prevent a fuel assembly from being misloaded, the following will mitigate the consequences of such an event:* Misload events, including misload events involving multiple fuel assemblies, have been analyzed-An adequate soluble boron margin mitigates the misload event. TS 3.7.15 requires 2150 ppm of soluble boron* The limiting misload of a single fuel assembly requires 1100 ppm of soluble boron to maintain keff <0.95 16 Enclosure Description and Assessment of Proposed License Amendment* The limiting, credible, multiple misload of placing a fresh fuel assembly into every blocked location in the most limiting array (Array C) requires 1460 ppm of soluble boron to maintain keff <0.95 Placing fresh fuel assemblies face-adjacent to one another is not credible.
A defense-in-depth approach provides multiple, independent barriers to this event. These barriers include:-Move sheets are generated, independently verified, and approved by qualified individuals
-Blocking devices or trash cans are placed in locations that are face-adjacent to locations approved for the storage of fresh fuel-The fuel movers will verify that fresh fuel is not placed in face-adjacent locations prior to completing each fuel move Sufficient rigor is placed into the generation of move sheets, execution of fuel movement, and maintenance of SEP maps that the likelihood of an assembly being misplaced in the SEP is small. The misplacement of multiple fuel assemblies is less probable.
Therefore, the approach used in the analysis is appropriately conservative.
The impact of placing multiple fresh fuel assemblies in face-adjacent locations is not evaluated in the analysis because this event is not considered credible.A regulatory commitment regarding the implementation of procedural controls to require verification that fresh fuel assemblies are not placed face-adjacent to one another before completing a fuel move is provided in Attachment 4.3.2 New Fuel Storage and Fuel Transfer Equipment Analysis A criticality safety analysis was performed to support operation of the NFS racks, the IFSR, the new fuel elevator, and the fuel upender and transfer machine. When discussing the new fuel elevator, and the fuel upender and transfer machine together, they are referred to as the fuel handling equipment (FHE). The existing NFS racks, IFSR, and FHE were evaluated to confirm that each system maintains subcriticality while performing its designed purposes.3.2.1 Storage and Equipment Description N.ew Fuel Storagqe Desigqn The NFS rack assemblies are made up of individual racks similar to those shown in UFSAR Figure 9.1-1. A minimum edge-to-edge spacing between fuel assemblies is maintained in adjacent rows. This spacing is the minimum value after allowances are made for rack fabrication tolerances and the predicted deflections resulting from postulated accident conditions.
The stainless steel construction of the storage racks is compatible with the water and the zirconium-clad fuel. The top structure of the racks is designed such that there is no opening between adjacent fuel cavities that is as large as the cross-section of the fuel bundle. In addition, the outer structure of the racks precludes the inadvertent placement of a bundle against the rack closer than the prescribed edge-to-edge spacing.17 Enclosure Description and Assessment of Proposed License Amendment Two concrete storage cavities are utilized for NFS. Each cavity is approximately 8 feet by 23 feet and contains 45 fuel assemblies in stainless steel racks. Three racks are installed in each cavity, forming a 3x15 array of fuel assemblies.
The rack structure provides at least 10 inches between the top of the active fuel and the top of the rack to preclude criticality in case a fuel assembly is dropped into a horizontal position on the top of the rack. The NFS racks and facilities are qualified as Seismic Category I and will survive a safe shutdown earthquake without loss of safety function.The following postulated accidents were considered in the design of the NFS racks:* Flooding -complete immersion of the entire storage array in pure, unborated, room temperature water* Envelopment of the entire array in a uniform density aqueous foam or mist of optimum density that maximizes the reactivity of the finite array (a condition that could result from firefighting)
- A fuel assembly dropped from a height of 4.5 ft onto the rack that falls horizontally across the top of the rack* Tensile load of 5000 lbs on the rack Although the above accident conditions have been postulated, the FHE, NFS racks, and the building arrangement are designed to minimize the possibility of these accidents and the effects resulting from these accidents.
Intermediate Fuel Storagqe Rack The IFSR is a four-cavity fuel storage rack in a lx4 array designed as an intermediate storage location for fuel bundles during refueling.
The rack is located in the containment adjacent to the core support barrel laydown area, which provides access to the refueling machine for insertion and removal of fuel bundles.Each cavity in the IFSR is a stainless steel can 8.69 inches on a side. The cavities are separated by a fuel center-to-center pitch of 18.56 inches. Each of the cavities is open at the bottom to provide thermal cooling for the worst case fuel bundle. The rack structure is designed to maintain keff < 0.95 by assuring under all normal and accident conditions, which includes SSE, that the minimum edge distance is not violated and that a fuel bundle cannot violate the 12-inch minimum stand-off distance around the cavities.New fuel may be stored in the IFSR before being moved into the core. Partially spent fuel may be moved out of the core and stored temporarily in the IFSR to provide spaces for fuel shuffling.
Spent fuel may be stored in the IFSR before being sent to the SFP.Fuel Upender and Transfer Machine The transfer machine, or carriage, conveys the fuel assemblies through the transfer tube. Two fuel assembly cavities are provided in the fuel carriage to reduce overall fuel handling time. After the refueling machine deposits a spent fuel bundle in the open cavity, it only has to move approximately one foot to pick up the new fuel assembly, which was brought from the fuel building in the other cavity. The handling operation in the fuel building is similar. The dual cavity 18 Enclosure Description and Assessment of Proposed License Amendment arrangement permits both fuel handling machines to travel fully loaded at all times. Fuel assemblies are placed on the transfer carriage in a vertical position, lowered to the horizontal position, moved through the fuel transfer tube on the transfer carriage, and then restored to the vertical position.
Wheels support the carriage and allow it to roll on tracks within the transfer*tube. The track sections at both ends of the transfer tube are mounted on the upending machines to permit the carriage to be properly positioned at the limits of its travel.An upending machine is provided at each end of the transfer tube. Each machine consists of a Structural support base from which is pivoted an upending straddle frame that engages the two-cavity fuel carrier. Hydraulic cylinders attached to the upending frame that engages the support base rotate the fuel carrier between the vertical and horizontal position.
A third fuel assembly was modeled five inches from the transfer carriage to allow for the presence of an additional fuel assembly no closer than five inches from the carriage. (Appendix B of Attachment 8)New Fuel Elevator The new fuel elevator is utilized to lower new fuel from the operating floor to the bottom of the pool where it is grappled by the spent fuel handling tool. The elevator is powered by a cable winch and fuel is contained in a simple support structure whose wheels are captured in two rails.3.2.2 New Fuel Criticality Safety Analysis Acceptance Criteria The objective of the criticality safety analysis is to ensure that the fuel storage operations are within the bounds 10 CFR 50.68(b)(2) and 50.68(b)(3):
- The estimated ratio of neutron production to neutron absorption and leakage (keff) of the fresh fuel in the storage racks shall be calculated assuming the racks are loaded with fuel of the maximum fuel assembly reactivity and flooded with unborated water and must not exceed 0.95, at a 95 percent probability, 95 percent confidence level. This evaluation need not be performed if administrative controls and/or design features prevent such flooding or if fresh fuel storage racks are not used.* If optimum moderation of fresh fuel in the fresh fuel storage racks occurs when the racks are assumed to be loaded with fuel of the maximum fuel assembly reactivity and filled with low-density hydrogenous fluid, the keff corresponding to this optimum moderation must not exceed 0.98, at a 95 percent probability, 95 percent confidence level. This evaluation need not be performed if administrative controls and/or design features prevent such moderation or if fresh fuel storage racks are not used.Design Approach Compliance is shown for the NFS racks, IFSR, and EHE by demonstrating that the system does not exceed 0.95 at a 95 percent probability with a 95 percent confidence level. A conservative combination of best estimate and bounding values has been selected to model the fuel in the analysis to ensure that fuel represented by the proposed TS is less reactive than the fuel modeled in the analysis.19 Enclosure Description and Assessment of Proposed License Amendment Computer Codes The analysis methodology employs SCALE 6.1.2 (Reference 6.3) with the 238-group cross-section library based on ENDF/B-VII.
All analyses performed used the "Fresh fuel without Absorber' validation suite. KENO-Va is used to determine the absolute reactivity of fresh fuel assemblies in the NFS.The validation of the ENDF/B-VII library with the SCALE 6.1.2 CSAS5 module is documented in Appendix A of Attachment
- 8. The code validation shows that SCALE 6.1.2 is an accurate tool for calculation of keff for the applications in this LAR. The benchmark calculations utilize the same computer plafform and cross-section libraries that are used for the design basis calculations.
Limitingq Fuel Desigqn Selection There are four potentially limiting fuel designs that have been used at PVNGS. The VAP design is currently in use on site and NGF is planned for use in the future. Therefore, both VAP and NGF designs are considered the two potentially limiting fuel designs. The STD and ANP fuel designs do not need to be addressed for these calculations because they are bounded by the VAP fuel design. Additional information is provided in Section B.4.2 of Attachment 8.Treatment of Concrete Concrete is a material which has a large variety of different potential compositions, all of which can be labeled as "concrete." Attachment 8 references the most limiting design for each situation.
The analysis has used a bounding treatment for concrete and the methodology remains conservative throughout the life of the concrete.
Additional information is provided in Section B.4.3 of Attachment 8.Biases and Uncertainties Reactivity biases are known variations between the real and analyzed system and their reactivity impact is added directly to the calculated keff. Uncertainties are random dispersions around a nominal, measured quantity.
Their impact is added to the calculated keff as the square root of the sum of the squares of the uncertainties.
The following biases and uncertainties are accounted for in the analysis.
A detailed discussion of biases and uncertainties is provided in Section B.4.5 of Attachment 8.* Reactivity effect of manufacturing tolerances
- Structural material presence* Eccentric fuel assembly positioning
- Uncertainty in the predictive capability of SCALE 6.1.2 and the associated cross-section library* Temperature bias for operating temperature range* Planar enrichment bias 20 Enclosure Description and Assessment of Proposed License Amendment New Fuel Storagqe Rack Criticality Safety Analysis The criticality safety analysis for the NFS rack consists of determining the limiting fuel design under both the fully flooded and optimum moderation condition.
Biases and uncertainties for both fully flooded and optimum moderation conditions are calculated using the limiting fuel design. The best estimate keff of the NES rack under both full density water and optimum moderation conditions is less than the target keff. This demonstrates that the NFS rack complies with the requirements of 10 CFR 50.68. Additional information is provided in Section B.4.6 of Attachment 8.The analysis of the NES rack has demonstrated that it can be operated in its design capacity without risk of exceeding the maximum reactivity imposed by regulation.
The analysis supports use of these components up to a maximum radially averaged enrichment of 4.65 weight percent U-235. All fuel used to date at PVNGS has an initial enrichment of < 4.55 weight percent U-235.Intermediate Fuel Storaqe Rack Criticality Safety Analysis The criticality safety analysis for the IFSR consists of determining the target keff for the IFSR, then confirming that the best estimate system keff (plus 2 o) is below the target keff with the limiting fuel design. The analysis uses the NGF design because the NGF is more reactive with full density water than VAP fuel. The biases and uncertainties are also calculated using the NGF design. The best estimate keff of the IFSR is less than the target keff, which demonstrates that the IFSR complies with the requirements of 10 CFR 50.68. Additional information is provided in Section B.4.7 of Attachment 8.Fuel Upender. Transfer Machine, and New Fuel Elevator Criticality Safety Analysis The criticality safety analysis for the fuel upender and transfer machine, and the analysis for the new fuel elevator, demonstrate that they can be used with fresh 4.65 weight percent U-235 fuel without exceeding a keff of 0.95 at a 95 percent probability, 95 percent confidence level. The design basis fuel is the NGF design. The best estimate keff of the fuel upender and transfer machine, and the new fuel elevator, is less than the target which demonstrates compliance with the requirements of 10 CFR 50.68. Additional information is provided in Section B.4.8 of Attachment 8.3.3 Spent Fuel Pool Transition Plan The SEP transition will be conducted over a total lapsed time of approximately 24 months with a schedule based on the unit refueling outages. Therefore, APS will insert two sets of TS and TS Bases pages during implementation.
One set will be labeled "Before SEP transition" and the other set will be labeled "After SFP transition." The Spent Fuel Pool Transition Plan is based on a SEP module-by-module transition scheme.Each SEP module that has not been transitioned is governed by the 'Before SEP transition" pages. As each module in a SEP is transitioned to the new configuration, the Shift Manager will make an entry in the control room log and declare that module as "transitioned." That particular module is then governed by the "After SEP transition" pages. When all three units have been transitioned, APS will submit an administrative TS change to remove the "before" and "after" pages, and insert the final pages.21 Enclosure Description and Assessment of Proposed License Amendment APS will transition to the proposed TS 3.7.17 in each of the three units in the following manner: 1. Move fuel assemblies as needed in order to neutronically decouple one module from the balance of the SEP. Analysis demonstrates that one row of empty cells is enough to decouple modules in the SEP.2. Perform a shuffle of the decoupled module, including installation of NETCO-SNAP-IN rack inserts.a. Some modules may need to be completely emptied of fuel assemblies.
The assemblies that must be moved may be stored in the appropriate regions of the rest of the SEP.b. Fuel assemblies that already meet the new TS 3.17.17 requirements may not need to be shuffled.3. Upon completion of the fuel shuffle and NETCO-SNAP-IN rack insert installation, the Shift Manager will declare the module has transitioned to the new TS 3.7.17 and enter this information into the control room log.4. Perform additional fuel shuffles, as needed, in order to move fuel from other parts of the SEP to the recently transitioned module.5. Repeat steps 1 through 4 for all 17 modules.6. Perform a 100 percent pool verification to confirm the following a. Fuel has been properly moved as confirmed by a 100 percent serial number check b. Stainless steel L-Jnserts are in the locations assumed in the analysis of record c. NETCO-SNAP-IN rack inserts are in their assumed locations d. Blocking devices are in their assumed locations A generic Westinghouse study (Reference 6.6) investigated the adequacy of assuming a distance of one cell pitch for neutronic isolation.
The study included NGF, which was determined to be the limiting fuel type used in the PVNGS SEP criticality safety analysis.
The results of the study concluded that a distance of approximately 10 cm (3.94 inches) of water was adequate for neutronic decoupling.
Given the required separation distance for neutronic decoupling, fuel assemblies separated by a single cell pitch or more at PVNGS are neutronically decoupled.
Once the Spent Fuel Pool Transition Plan has been started in a particular unit, the insert installation and SFP transition shall be executed in a deliberate, safe, and controlled manner until complete in that unit. The transition to the new SEP configuration will be completed in all three units in accordance with the Spent Fuel Pool Transition Plan within two years of the NRC approval date of the amendment or by December 31, 2019, whichever is later. A regulatory commitment regarding transition plan implementation is provided in Attachment 4.
4.0 REGULATORY EVALUATION
4.1 Applicable Regulatory Requirements The regulations in 10 CFR 50.36(c)(2)(ii)(B), Limiting conditions for operation, state: 22 Enclosure Description and Assessment of Proposed License Amendment Criterion
- 2. A process variable, design feature, or operating restriction that is an initial condition of a design basis accident or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.Technical Specification (TS) 3. 7.17 currently meets this requirement and will continue to meet this requirement after the proposed changes are approved and implemented.
The regulations in 10 CFR 50.36(c)(4), Design features, state: Design features to be included are those features of the facility such as materials of construction and geometric arrangements, which, if altered or modified, would have a significant effect on safety and are not covered in categories described in paragraphs (c)(1), (2), and (3) of this section.TS 4.3.1 currently meets this requirement and will continue to meet this requirement after the proposed changes are approved and implemented.
The regulations in 10 CFR 50.68, Criticality accident requirements, specifically 10 CFR 50.68(b)(1) state: Plant procedures shall prohibit the handling and storage at any one time of more fuel assemblies than have been determined to be safely subcritical under the most adverse moderation conditions feasible by unborated water.This requirement is currently met by existing PVNGS fuel handling procedures and will continue to be met by the same procedures after the proposed changes are approved and implemented.
The regulations in 10 CER 50.68(b)(2) state: The estimated ratio of neutron production to neutron absorption and leakage (k-effective) of the fresh fuel in the fresh fuel storage racks shall be calculated assuming the racks are loaded with fuel of the maximum fuel assembly reactivity and flooded with unborated water and must not exceed 0.95, at a 95 percent probability, 95 percent confidence level. This evaluation need not be performed if administrative controls and/or design features prevent such flooding or if fresh fuel storage racks are not used.The regulations in 10 CFR 50.68(b)(3) state: If optimum moderation of fresh fuel in the fresh fuel storage racks occurs when the racks are assumed to be loaded with fuel of the maximum fuel assembly reactivity and filled with low-density hydrogenous fluid, the k-effective corresponding to this optimum moderation must not exceed 0.98, at a 95 percent probability, 95 percent confidence level. This evaluation need not be performed if administrative controls and/or design features prevent such moderation or if fresh fuel storage racks are not used.The regulations in 10 CFR 50.68(b)(4) state: If no credit for soluble boron is taken, the k-effective of the spent fuel storage racks loaded with fuel of the maximum fuel assembly reactivity must not exceed 0.95, at a 95 percent probability, 95 percent confidence level, if flooded with unborated water. If credit is taken for soluble boron, the k-effective of the spent 23 Enclosure Description and Assessment of Proposed License Amendment fuel storage racks loaded with fuel of the maximum fuel assembly reactivity must not exceed 0.95, at a 95 percent probability, 95 percent confidence level, if flooded with borated water, and the k-effective must remain below 1.0 (subcritical), at a 95 percent probability, 95 percent confidence level, if flooded with unborated water.The requirements in 10 CFR 50.68(b) cited above are met by the nuclear criticality safety analyses provided in WCAP- 18030-P. The results of the criticality analysis form the basis of the proposed TS 3.7.17 changes. TS 3.7.17 currently meets these requirements and will continue to meet these requirements after the proposed changes are approved and implemented.
The regulations in 10 CFR 50.68(b)(7) state: The maximum nominal U-235 enrichment of the fresh fuel assemblies is limited to five (5.0) percent by weight.TS 4.3.1.2 currently meets this requirement and will continue to meet this requirement after the proposed changes are approved and implemented.
The regulations in 10 CFR Part 50, Appendix A, General Design Criteria for Nuclear Power Plants, Criterion 62, Prevention of criticality in fuel storage and handling, state: Criticality in the fuel storage and handling system shall be prevented by physical systems or processes, preferably by use of geometrically safe configurations.
TS 3. 7.17 currently meets this requirement and will continue to meet this requirement after the proposed changes are approved and implemented.
The guidance in DSS-ISG-2010-01 is to be used by NRC staff to review nuclear criticality safety analyses for the storage of new and spent nuclear fuel as they apply to applications for license amendments submitted after September 29, 2011.This license amendment request and WCAP-18030-P were developed using the guidance of DSS-ISG-2010-01.
4.2 Precedent The analysis methodology for the site-specific criticality analysis employs the PARAGON code, which is approved for use by the NRC (Reference 6.4).4.:3 Significant Hazards Consideration As required by 10 CFR 50.91(a), Notice for Public Comment, an analysis of the issue of no significant hazards consideration using the standards in 10 CFR 50.92, Issuance of Amendment, is presented below: 1. Does the proposed amendment involve a significant increase in the probability or consequences of an accident previously evaluated?
Response:
No.24 Enclosure Description and Assessment of Proposed License Amendment The proposed amendment would modify the Palo Verde Nuclear Generating Station (PVNGS) Technical Specifications (TS) to incorporate the results of an updated criticality safety analysis for both new fuel and spent fuel storage. The revised criticality safety analysis provides an updated methodology that allows credit for neutron absorbing NETCO-SNAP-IN rack inserts and corrects non-conservative input assumptions in the previous criticality safety analysis.The proposed amendment does not change or modify the fuel, fuel handling processes, number of fuel assemblies that may be stored in the spent fuel pool (SFP), decay heat generation rate, or the SFP cooling and cleanup system. The proposed amendment was evaluated for impact on the following previously evaluated events and accidents:
- fuel handling accident (FHA)* fuel misload event* SEP boron dilution event* seismic event* loss of SEP cooling event Implementation of the proposed amendment will be accomplished in accordance with the Spent Fuel Pool Transition Plan and does not involve new fuel handling equipment or processes.
The radiological source term of the fuel assemblies is not affected by the proposed amendment request. The EHA radiological dose consequences associated with fuel enrichment at this level are addressed in the PVNGS Updated Final Safety Analysis Report (UFSAR) Section 15.7.4 and remain unchanged.
Therefore, the proposed amendments do not significantly increase the probability or consequences of a FHA.Operation in accordance with the proposed amendment will not change the probability of a fuel misload event because fuel movement will continue to be controlled by approved fuel handling procedures.
Although there will be additional allowable storage arrays defined by the amendment, the fuel handling procedures will continue to require identification of the initial and target locations for each fuel assembly that is moved. The consequences of a fuel misload event are not changed because the reactivity analysis demonstrates that the same subcriticality criteria and requirements continue to be met for the limiting fuel misload event.Operation in accordance with the proposed amendment will not change the probability or consequences of a boron dilution event because the systems and events that could affect SFP soluble boron concentration are unchanged.
The current boron dilution analysis demonstrates that the limiting boron dilution event will reduce the boron concentration from the TS limit of 2150 ppm to 1900 ppm. This leaves sufficient margin to the 1460 ppm credited by the SFP criticality safety analysis.
The analysis confirms that the time needed for dilution to reduce the soluble boron concentration is greater than the time needed for actions to be taken to prevent further dilution.Operation in accordance with the proposed amendment will not change the probability of a seismic event since there are no elements of the updated criticality analysis that influence the occurrence of a seismic event. The consequences of a seismic event are not significantly increased because the forcing functions for seismic excitation are not increased and because the mass of storage racks with NETCO-SNAP-IN inserts is not appreciably 25 Enclosure Description and Assessment of Proposed License Amendment increased.
Seismic analyses demonstrate adequate stress levels in the storage racks when inserts are installed.
Operation in accordance with the proposed amendment will not change the probability of a loss of SEP cooling event because the systems and events that could affect SEP cooling are unchanged.
The consequences are not significantly increased because there are no changes in the SFP heat load or SEP cooling systems, structures, or components.
Furthermore, conservative analyses indicate that the current design requirements and criteria continue to be met with the NETCO-S NAP-IN inserts installed.
Therefore, the proposed amendment does not involve a significant increase in the probability or consequences of an accident previously evaluated.
- 2. Does the proposed amendment create the possibility of a new or different kind of accident from any accident previously evaluated?
Response:
No.The proposed amendment would modify the PVNGS TS to incorporate the results of an updated criticality safety analysis for both new fuel and spent fuel storage. The revised criticality safety analysis provides an updated methodology that allows credit for neutron absorbing NETCO-SNAP-IN rack inserts and corrects non-conservative input assumptions in the previous criticality safety analysis.The proposed amendment does not change or modify the fuel, fuel handling processes, number of fuel assemblies that may be stored in the pool, decay heat generation rate, or the SEP cooling and cleanup system. The effects of operating with the proposed amendment are listed below. The proposed amendment was evaluated for the potential of each effect to create the possibility of a new or different kind of accident:* addition of inserts to the SEP storage racks* additional weight from the inserts* new storage patterns* displacement of SEP water by the inserts, Each NETCO-SNAP-IN insert will be placed between a fuel assembly and the storage cell wall, taking up some of the space available on two sides of the fuel assembly.
Analyses demonstrate that the presence of the inserts does not adversely affect spent fuel cooling, seismic capability, or subcriticality.
The aluminum and boron carbide materials of construction have been shown to be compatible with nuclear fuel, storage racks, and SEP environments, and generate no adverse material interactions.
Therefore, placing the inserts into the SEP storage racks cannot cause a new or different kind of accident.Operation with the added weight of the NETCO-SNAP-IN inserts will not create a new or different accident.
The analyses of the racks with NETCO-SNAP-IN inserts installed demonstrate that the stress levels in the rack modules continue to be considerably less than allowable stress limits. Therefore, the added weight from the inserts cannot cause a new or different kind of accident.26 Enclosure Description and Assessment of Proposed License Amendment Operation with the proposed fuel storage patterns will not create a new or different kind of accident because fuel movement will continue to be controlled by approved fuel handling procedures.
These procedures continue to require identification of the initial and target locations for each fuel assembly that is moved. There are no changes in the criteria or design requirements pertaining to fuel storage safety, including subcriticality requirements.
Analyses demonstrate that the proposed storage patterns meet these requirements and criteria with adequate margins. Therefore, the proposed storage patterns cannot cause a new or different kind of accident.Operation with insert movement above stored fuel will not create a new or different kind of accident.
The insert with its handling tool weighs less than the weight of a single fuel assembly.
Single fuel assemblies are routinely moved safely over fuel assemblies and the same level of safety in design and operation will be maintained when moving the inserts.The installed rack inserts will displace a negligible quantity of the SEP water volume and therefore will not reduce operator response time to previously-evaluated SFP accidents.
The accidents and events previously analyzed remain bounding.
Therefore, the proposed amendment does not create the possibility of a new or different kind of accident from any accident previously evaluated.
- 3. Does the proposed amendment involve a significant reduction in a margin of safety?Response:
No.The proposed amendment would modify the TS to incorporate the results of an updated criticality safety analysis for both new fuel and spent fuel storage. The revised criticality safety analysis provides an updated methodology that allows credit for neutron absorbing NETCO-SNAP-IN rack inserts and corrects non-conservative input assumptions in the previous criticality safety analysis.
It was evaluated for its effect on current margins of safety as they relate to criticality, structural integrity, and spent fuel heat removal capability.
The margin of safety for subcriticality required by 10 CFR 50.68(b)(4) is unchanged.
New criticality analyses confirm that operation in accordance with the proposed amendment continues to meet the required subcriticality margins.The structural evaluations for the racks and spent fuel pool with NETCO-SNAP-IN inserts installed show that the rack and SEP are unimpaired by loading combinations during seismic motion, and there is no adverse seismic-induced interaction between the rack and NETCO-SNAP-IN inserts.The proposed amendment does not affect spent fuel heat generation, heat removal from the fuel assembly, or the SEP cooling systems. The effects of the NETCO-SNAP-IN inserts are negligible with regards to volume of water in the pool, flow in the SEP rack cells, and heat removal system performance.
The addition of a Spent Fuel Pool Rack Neutron Absorber Monitoring program (proposed TS 5.5.21) provides a method to identify potential degradation in the neutron absorber material prior to challenging the assumptions of the criticality safety analysis related to the material.Therefore, the addition of this monitoring program does not reduce the margin of safety;27 Enclosure Description and Assessment of Proposed License Amendment rather it ensures th'e margin of safety is maintained for the planned life of the spent fuel storage racks.Therefore, the proposed amendment does not involve a significant reduction in the margin of safety.4.4 Conclusion APS concludes that operation of the facility in accordance with the proposed amendment does not involve a significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and, accordingly, a finding of "no significant hazards consideration" is justified.
Based on the considerations discussed above, (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or the health and safety of the public.5.0 ENVIRONMENTAL CONSIDERATION A review has determined that the proposed amendment would change a requirement with respect to installation or use of a facility component located within the restricted area, as defined in 10 CFR 20, Standards for Protection Against Radiation.
However, the proposed amendment does not involve (i) a significant hazards consideration, (ii) a significant change in the types or a significant increase in the amounts of any effluents that may be released offsite, or (iii) a significant increase in individual or cumulative occupational radiation exposure.
Accordingly, the proposed amendment meets the eligibility criterion for categorical exclusion set forth in 10 CFR 51.22(c)(9).
Therefore, pursuant to 10 CFR 51 .22(b), no environmental impact statement or environmental assessment need be prepared in connection with the proposed amendment.
6.0 REFERENCES
6.1 Criticality Safety Analysis for Palo Verde Nuclear Generating Station Units 1, 2, and 3 (Proprietary), WCAP-1 8030-P, Revision 0, September 2015.6.2 Staff Guidance Regarding the Nuclear Criticality Safety Analysis for Spent Fuel Pools, DSS-ISG-201 0-01, Revision 0, Nuclear Regulatory Commission Division of Safety Systems, Rockville, MD, September 29, 2011. (ML1 10620086)6.3 Scale: A Comprehensive Modeling and Simulation Suite for Nuclear Safety Analysis and Design, ORNL/TM-2005/39, Version 6.1, Oak Ridge National Laboratory, Oak Ridge, TN, June 2011.6.4 M. Ouisloumen, H. Huria, et al, Qualification of the Two-Dimensional Transport Code PARAGON, WCAP-16045-P-A, Revision 0, Westinghouse Electric Company LLC, Monroeville, PA, August 2004.28 Enclosure Description and Assessment of Proposed License Amendment 6.5 C. V. Parks, et al, Review and Prioritization of Technical Issues Related to Bumup Credit for LWR Fuel, NUREG/CR-6665, Oak Ridge National Laboratory, Oak Ridge, TN, February 2000.6,6 Letter, J. Gresham (WEC) to NRC, Responses to Requests for Additional Information from the Review of WCAP- 1 7483-PA/WCAP-1 7483-NP, Revision 0, 'Westinghouse Methodology for Spent Fuel Pool and New Fuel Rack Criticality Safety Analysis,'
LTR-NRC-15-60, dated July 20, 2015.29 Enclosure ,Description and Assessment of Proposed License Amendment ATTACHMENT 1 Marked-up Technical Specifications Pages (Pages Provided for Before and After SEP Transition) 3.7.17-1-3.7.17-2 3.7.17-3 3.7.17-4 4.0-2 4.0-3 5.5-19 SBefore SFP transitionI Spent Fuel Assembly Storage 3.7.17 3.7 PLANT SYSTEMS 3.7.17 Spent Fuel Assembly Storage LCO 3.7.17 APPLICABILITY:
The combination of initial enrichment, burnup, and decay time of each fuel assembly stored in each of the four regions of the fuel storage pool shall be within the acceptable burnup domain for each region as shown in Figures 3.7.17-1, 3.7.17-2, or 3.7.17-3, and described in Specification 4.3.1.1.Whenever any fuel assembly is stored in the fuel storage pool.ACTIONS__________________________
CONDITION REQUIRED ACTION COMPLETION TIME A. Requirements of the A.1------NOTE----
LCO not met. LCD 3.0.3 is not applicable.
Initiate action to Immediately move the noncomplying fuel assembly into an appropri ate region.SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.17.1 Verify by administrative means the initial Prior to enrichment, burnup, and decay time of the storing the fuel assembly is in accordance with Figures fuel assembly 3.7.17-1, 3.7.17-2, or 3.7.17-3, and in the fuel Specification 4.3.1.1. storage pool.PALO VERDE UNITS 1,2,3 3.7.17-I PAL VEDE NIT 1,,3 .7.7-1AMENDMENT NO. 117, 1£ IBefore SFP transition]I Spent Fuel Assembly Storage 3.7.17 Figure 3.7.17-1 ASSEMBLY BURNUP VERSUS INITIAL ENRICHMENT for Region 2 0 6 0 S S S S 0 ACC EPTABL[" for Reg on 2 15000 500 I-NOT A CEPTA LE for R gion 2 S S S S S S S S S S 0 S S S S S S S_________________
_________________
9 S S S S S S S S S Note: This curve assumes ero decay time.____________
~1~S (1 1.5 2.0 2.5 3.0 3.5 Initial Enrichment, weight %4.0 4.5
- 5.0 4.80%limiting enrichment PALO VERDE UNITS 1,2,3 3.7.17-2 AMENDMENT NO. 117, !2
[Before SFP transition Spent Fuel Assembly Storage 3.7.17 Figure 3.7.17-2 ASSEMBLY BURNUP VERSUS INITIAL ENRICHMENT for Region 3 (at decay times from 0 to 20 years)45000 40000 35000 30000 25000>20000 t'0 E ci 0 ACCEP' ABLE for Region 3 "o ,.S//NOT ACq;EPTABI.
E for Region S 0 ThOUU+J +
I S 0 S S(S-a---S S 5000~~.2 Noe Asn nent and current di S 0 S cay lime.: ly e/igible for Regi jm 3 if actualIBU
> 3U requirement for given initial endch F i, , , , .....m 1.5 2.0 2.5 3.0 3.5 4.0 4.5
- 5.0 Initial Enrichment, weight % 4.80%limiting Decaylime[ -U-1-5years
--4--2Oyears enrichment PALO VERDE UNITS 1,2,33..73AMNETNO 3.7.17-3 AMENDMENT NO.
I Before SFP transition Spent Fuel Assembly Storage 3.7.17 Figure 3.7.17-3 ASSEMBLY BURNUP VERSUS INITIAL ENRICHMENT
~for Region 4 (at decay times from C to 20 years)I-0 E~0 E a)0)3.0 3.5 Initial Enrichment, weight %limiting DecaylimeI
--0years 11-years
--&.-lOyears
-~-'-15 years --o-20years enrdchment PALO VERDE UNITS 1,2,3 371- MNMN O 3.7.17-4 AMENDMENT NO.
SAfter SFP transition Spent Fuel Assembly Storage 3.7.17 3.7 PLANT SYSTEMS 13.7.17-1 through 3.7.17-5.I 3.7.17 Spent Fuel Assembly Storage LCO 3.7.11 The combination of initialnichment, burnup, and decay time of each fuel assemblytreine=hftef'-
acceptable burnup domain for each region as shown in Figures') 7 17 1 ') 7 17 ') ,-~-i Q 7 17 '2 -~A A-,-.~k,-A -vs Specification 1.3.1.1.APPLICABILITY:
Whenever any fuel pool.assembly is stored in the fuel storage ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Requirements of the A.1------NOTE--
--LCO not met. LCO 3.0.3 is not applicable.
Initiate action to Immediately move the noncomplying fuel assembly into an appropriate region.SURVEILLANCE REQUIREMENTS_________
SURVEILLANCE FREQUENCY SR 3.7.17.1 Verify by administrative means the initial Prior to enrichment, burnup, and decay time of the storing the fuel assembly is in accordance with fuel assembly...and in the fuel Specification 4mi.3.1.1
... storage pool.ITables 3.7.17-1 through 3.7.17-5, Figure 3.7.17-PALO VERDE UNITS 1,2,3 3.7.17-1 PAL VEDE NIT 1,,3 .7.7-1AMENDMENT NO. I147, 2 Ilnsert new Tables 3.7.17-1 through 3.7.17-5 and jFigure 3.7.17-1 (total 6 pages) here.I IAfter SFP transitionI Spent Fuel Assembly Storage 3.7.17 FiguJren 3.7.17 1 Acc(rkArnl
\J l \lII-1CI IC TkITTT Al FI-dITriUMCIdT PALO VERDE UNITS 1,2,3 97179 PAL VRD UIT 12, 37.72AMENDMENT NO. 11.7, 2 SAfter SFP transition Spent Fuel Assembly Storage 3.7.17 limiting--me '-O years --U-5 years -* 10 years -U-I--l5 years --4-20 years enrichment PALO VERDE UNITS 1,2,33. 13AMNETNO 15 97]79 AMENDMENT NO.
SAfter SFP transitionI Spent Fuel Assembly Storage 3.7.17 Figure,3.7.17 3-ACSE'MILY VU NUP VERSUS TIT-I-AI ENRT/ICHMEITT
'4,~uuuU 45000 S S 40000 __ __ _ r S ACCEI Region 4 35000 250000 0..~2O000 15000 10000 I 0 S S_________________
_________
I S S S S__________________
___________________
S 0 S S NOT AC E ABLE or Region ____ __S S S S S_________________
_________________
_________________
_________
S______/S S S S 0/ ______ + 4-~---f 5000//N______ 4- + 4-S S 0 S 0 0 0 caYthne. 0 mq~kemwl for tyB~efor Reg M 4&fachiBU
>-hiaI enddi rdw caidwmt 1.2.0 2.5 3.0 3.5 4.0 4.5
- 5 Initial Enrichment, weight % 4.80%-ieI-4-0 years --11-5 years 15 years --4-20Oyears lenicmetnt PALO VERDE UNITS 1,2,33..7AMN ETN. §Q717A AMENDMENT NO.
[After SFP transition]
Spent Fuel Assembly Storage 3.7.17 Table 3.7.17-1 Fuel Regions Ranked by Reactivity Fuel Region 1 Highest Reactivity (See Note 2)Fuel Region 2 Fuel Region 3 Fuel Region 4 Fuel Region 5 Fuel Region 6 Lowest Reactivity Notes: 1. Fuel Regions are defined by assembly average burnup, initial enrichment' and decay time as provided by Table 3.7.17-2 through Table 3.7.17-5.2. Fuel Regions are ranked in order of decreasing reactivity, e.g., Fuel Region 2 is less reactive than Fuel Region 1, etc.3. Fuel Region 1 contains fuel with an initial maximum radially averaged enrichment up to 4.65 wt% 2 3 5 U. No burnup is required.4. Fuel Region 2 contains fuel with an initial maximum radially averaged enrichment up to 4.65 wt% 2 3 5 U with at least 16.0 GWd/MTU of bumup.5. Fuel Regions 3 through 6 are determined from the minimum burnup (BU) equation and coefficients provided in Tables 3.7.17-2 through 3.7.17-5.6. Assembly storage is controlled through the storage arrays defined in Figure 3.7.17-1.7. Each storage cell in an array can only be populated with assemblies of the Fuel Region defined in the array definition or a lower reactivity Fuel Region.SInitial Enrichment is the nominal 2 3 5 U enrichment of the central zone region of fuel, excluding axial blankets, prior to reduction in 2 3 5 U content due to fuel depletion.
If the fuel assembly contains axial regions of different 2 3 5 Uj enrichment values, such as axial blankets, the maximum initial enrichment value is to be utilized.
[After SFP transition[
Spent Fuel Assembly Storage 3.7.17 Table 3.7.17-2 Fuel Region 3: Burnup Requirement Coefficients Coefficients Decay Time (yr.) A 1 A 2 A 3 A 4 0 -1.5473 15.5395 -39.0197 24.1121 5 -1.4149 13.9760 -33.6287 18.3369 10 -1.3012 12.6854 -29.2539 13.6879 15 -1.0850 10.4694 -22.1380 6.3673 20 -0.9568 9.1487 -17.9045 2.0337 Notes: 1. Relevant uncertainties are explicitly included in the criticality analysis.
For instance, no additional allowance for burnup uncertainty or enrichment uncertainty is required.
For a fuel assembly to meet the requirements of a Fuel Region, the assembly burnup must exceed the "minimum burnup" (GWd/MTU) given by the curve fit for the assembly "decay time" and "initial enrichment." The specific minimum burnup (BU) required for each fuel assembly is calculated from the following equation: BU =Al
- En 3 + A 2
- En 2 + A 3
- En + A 4 2. Initial enrichment, En, is the maximum radial average 2 3 5 U enrichment.
Any En value between 2.55 wt% 2 3 5 U and 4.65 wt% 2 3 5 U may be used. Burnup credit is not required for an En below 2.55 wt% 2 3 5 U.3. It is acceptable to linearly interpolate between calculated BU limits based on decay time.4. The 20-year coefficients must be used to calculate the minimum BU for an assembly with a decay time of greater than 20 years.
[After SFP transition Spent Fuel Assembly Storage 3.7.17 Table 3.7.17-3 Fuel Region 4: Burnup Requirement Coefficients Coefficients Decay Time (yr.) A 1 A 2 A 3 A 4 0 0.4260 -6.2766 40.9264 -54.6813 5 0.2333 -4.1!545 32.9080 -46.1161 10 0.4257 -6.2064 39.0371 -51.5889 15 0.53 15 -7.3777 42.5706 -54.7524 20 0.5222 -7.3897 42.6587 -54.8201 Notes: 1. Relevant uncertainties are explicitly included in the criticality analysis.
For instance, no additional allowance for burnup uncertainty or enrichment uncertainty is required.
For a fuel assembly to meet the requirements of a Fuel Region, the assembly bumup must exceed the "minimum burnup" (GWd/MTU) given by the curve fit for the assembly "decay time" and "initial enrichment." The specific minimum burnup (BU) required for each fuel assembly is calculated from the following equation: BU=AI
- En 3 +A 2 *En 2+-bA 3 *En +A 4 2. Initial enrichment, En, is the maximum radial average 2 3 5 U enrichment.
Any En value between 1.75 wt% 2 3 5 U and 4.65 wt% 2 3 5 U may be used. Bumnup credit is not required for an En below 1.75 wt% 235U.3. It is acceptable to linearly interpolate between calculated BU limits based on decay time.4. The 20-year coefficients must be used to calculate the minimum BU for an assembly with a decay time of greater than 20 years.
[After SEP transition Spent Fuel Assembly Storage 3.7.17 Table 3.7.17-4 Fuel Region 5: Burnup Requirement Coefficients Decay Coefficients Time (yr.) A 1 A 2 A 3 A 4 0 -0.1114 -0.4230 20.9136 -32.8551 5 -0.1232 -0.4463 20.8337 -32.6068 10 -0.2357 0.4892 18.0192 -30.0042 15 -0.1402 -0.4523 20.3745 -31.7565 20 -0.0999 -0.8152 21.0059 -31.9911 Notes: 1. Relevant uncertainties are explicitly included in the criticality analysis.
For instance, no additional allowance for burnup uncertainty or enrichment uncertainty is required.
For a fuel assembly to meet the requirements of a Fuel Region, the assembly burnup must exceed the "minimum bumup" (GWd/MTU) given by the curve fit for the assembly "decay time" and "initial enrichment." The specific minimum burnup (BU) required for each fuel assembly is calculated from the following equation: BU =A 1
- En 3 + A 2
- En 2 + A 3
- En + A 4 2. Initial enrichment, En, is the maximum radial average 2 3 5 U enrichment.
Any En value between 1.65 wt% 2 3 5 U and 4.65 wt% 2 3 5 U may be used. Burnup credit is not required for an En below 1.65 wt% 2 3 5 U.3. It is acceptable to linearly interpolate between calculated BU limits based on decay time.4. The 20-year coefficients must be used to calculate the minimum BU for an assembly with a decay time of greater than 20 years.
[After SFP transition Spent Fuel Assembly Storage 3.7.17 Table 3.7.17-5 Fuel Region 6: Burnup Requirement Coefficients Decay Coefficients Time (yr.) A 1 A 2 A 3 A 4 0 0.7732 -9.3583 49.6577 -54.6847 5 0.7117 -8.4920 45.1124 -49.7282 10 0.6002 -7.2638 40.2603 -44.9348 15 0.5027 -6.2842 36.6715 -41.4934 20 0.2483 -3.7639 28.8269 -34.6419 Notes: 1. Relevant uncertainties are explicitly included in the criticality analysis.
For instance, no additional allowance for bumnup uncertainty or enrichment uncertainty is required.
For a fuel assembly to meet the requirements of a Fuel Region, the assembly bumup must exceed the "minimum burnup" (GWd/MTU) given by the curve fit for the assembly "decay time" and "initial enrichment." The specific minimum burnup (BU) required for each fuel assembly is calculated from the following equation: BU =A 1
- En 3 + A 2
- En 2 + A 3
- En + A 4 2. Initial enrichment, En, is the maximum radial average 2 3 5 U enrichment.
Any En value between 1.45 wt% Z3U and 4.65 wt% 23U may be used. Burnup credit is not required for an En below 1.45 wt% 2 3 5 U.3. It is acceptable to linearly interpolate between calculated BU limits based on decay time.4. The 20-year coefficients must be used to calculate the minimum BU for an assembly with a decay time of greater than 20 years.
SAfter SFP trransition Spent Fuel Assembly Storage 3.7.17 Figure 3.7.17-1 Allowable Storage Arrays Fwou Region 6 assemblies (6) Tw steckrbagded cells cotain abtlocess stells L(ise). TheRgo 1. Thsembishaded loathionsa indicatehelhcoti a stainless steel L-insert.NoETOSA-N 2.o Ae block1asedbis()cekrbaddwt w cells(X contains lcing dvcanolyw terainshe actiefulrein 3TC. NTheRgo-n1assemlNinets must bc oientedl winthe saedrcinaah stainless steel L-inserts.Eer 4 NTC -N PN isrsaeolloaeincells without a stainless steel L-inetms oti EC -N P1insert.
5-nsr. Anyhel egontann3 afe assemblyyr iCsa iseaiennemt (aerflld cell intinn aNTCall-Niset stoageayrys 6.e AnR traearagoaion deintdfrafe assembly may cbeckreplaced with noren-gon4fssssiele).Th mein2atserial.
n h ignlylctdRein4asml r ahi I Before SFP transitionI Design Features 4.0 4.0 DESIGN FEATURES (continued) 4.3 Fuel Storage 4.3.1 Criticality 4.3.1.1 The spent fuel storage racks are designed and shall be maintained with: a. Fuel assemblies having a maximum radially averaged U-235 enrichment of 4.80 weight percent;b. keff < 1.0 if fully flooded with unborated water, which includes an allowance for biases and uncertainties as described in Section 9.1 of the UFSAR;c. keff 0.95 if fully flooded with water borated to 900 ppm, which includes an allowance for biases and uncertainties as described in Section 9.1 of the UFSAR.d. A nominal 9.5 inch center-to-center distance between adjacent storage cell locations.
- e. Region 1: Fuel shall be stored in a checkerboard (two-out-of-four) storage pattern. Fuel that qualifies to be stored in Regions 1, 2, 3, or 4 in accordance with Figures 3.7.17-1, 3.7.17-2, or 3.7.17-3, may be stored in Region 1.f. Region 2: Fuel shall be stored in a repeating 3-by-4 storage pattern in which Region 2 (two-out-of-twelve) assemblies and Region 4 (ten-out-of-twelve) assemblies are mixed as shown in Section 9.1 of the UFSAR. Only fuel that qualifies to be stored in Regions 2, 3, or 4, in accordance with Figures 3.7.17-1, 3.1.17-2, or 3.7.17-3, may be stored in Region 2.g. Region 3: Fuel shall be stored in a four-out-of-four storage pattern. Only fuel that qualifies to be stored in Regions 3 or 4, in accordance with Figures 3.7.17-2 or 3.7.17-3, may be stored in Region 3.(conti nued)PALO VERDE UNITS 1,2,3 4.0-2 PALOVERE UITS1.23 40-2AMENDMENT NO. 47~
I Before SEP transition I Design Features 4.0 4.0 DESIGN FEATURES (continued)
- h. Region 4: Fuel shall be stored in a repeating 3-by-4 storage pattern in which Region 2 (two-out-of-twelve) assemblies and Region 4 (ten-out-of-twelve) assemblies are mixed as shown in Section 9.1 of the UFSAR. Only fuel that qualifies to be stored in Region 4 in accordance with Figure 3.7.17-3 shall be stored in Region 4.4.3.1.2 The new fuel storage racks are designed and shall be maintained with: a. Fuel assemblies having a maximum radially averaged U-235 enrichment of 4.80 weight percent;b. keff 0.95 if fully flooded with unborated water, which includes an allowance for biases and uncertainties as described in Section 9.1 of the UFSAR: c. keff 0.98 if moderated by aqueous foam, which includes an allowance for biases and uncertainties as described in Section 9.1 of the UFSAR; and d. A nominal 17 inch center to center distance between fuel assemblies placed in the storage racks.4.3.2 Drainage The spent fuel storage pool is designed and shall be maintained to prevent inadvertent draining of the pool below elevation 137 feet -6 inches.4.3.3 Capacity The spent fuel storage pool is designed and shall be maintained with a storage capacity limited to no more than 1329 fuel assemblies.
PALO VERDE UNITS 1,2,3 4.0-3 PAO EDEUNT 12, .03AMENDMENT NO. 11 o ...
After SFP transition Design Features 4.0 4.0 DESIGN FEATURES (continued) 4.3 Fuel Storage 4.3.1 Criticality 4.3.1.1 The spent fuel storage racks are designed and shall be maintained with:46 a. Fuel assemblies having/ maximum radially averaged U-235 enrichment .weight percent: b. keff < 1.0 if fully flooded with unborated water.which includes an allowance for biases and uncertainties as described in Section 9.1 of the UFSAR-c. keff -< 0.95 if fully flooded with water borated to~ppm, which includes an allowance for biases and unertainties as described in Section 9.1 of the 1460 UFSAR.d. A nominal 9.5 inch center-to-center distance between adjacent storage cell locations.
- e. R~on !:Fuel shall bc stored in a checkerboa=rd Fuel assemblies are .. ..... --.classified in Fuel Regions qualifie to be tore in Regon 1 2 3, or in 1-6 as shown in Tables a ccordance w.i th Figu=res 3.7.17 1, 3.7.17 2, or 3.7.17-1 through 3. 7.17 3 .. mayb in Rcgn 1.3.7.17-5. P Fe,,l shall be a; .......tin-3 byIstrg pattern in ..hich Regon 2 ( .. = atenrutontele seble arfel mixed as. shown accordance wit FiguQres 3.7.17 1, 3.7/.17% 2,-n or Regionnu3.
PALO VERDE UNITS 1,2,340-AMNETNO
.&4.0-2 AMENDMENT NO. 117, 125 After SFP transition Design Featuresi 4.0 4.0 DESIGN FEATURES (continued) 4.3.1.2 Th enewfuel Fuorag rhalls bre dstored ind sa llcat be 3ai bta ine witorgh ateni-hch go (t.oF uteflwlc assemblies hvn m a ndmu ra gionl avrae (t35enrouioctele)t afswembieht aerenmxdt shw incerectiones9.1 dof cthe edUF n S. tion 9.1& of th qualifies to blowne store binaRegio and uccordaincies ait Fiuesrib3 hllb soed in Regtion 9.Ifte FA n m.Ainta ined with inhc, r 465t ene itnc ewe a.fuel assemblies hlavin maximumoradill raverage U-235raiagientrichmenteor eih eret Thesp bn kefue strg 0.95 if fullyflooed withalunboaed waitaiert preveninawhichn drincldeing allowanepfor biaswesevatind u3 et nchertitessdsrbd nScin91o h Thespent includstoane allowanceforsbiase and uncl ermitaintes withstoasgescrpaibyliied iSetiono 9.1ofe then UF3AR andl fuesaseblesplce hisorgeraks PALO VERDE UNITS 1,2,3 4.0-3 PAL VRD UITS1,.34.-3AMENDMENT NO. 1-17,12 ..
IBefore SFP transitionI Programs and Manuals 5.5 5.5 Programs and Manuals (continued) 5.5.19 Battery Monitoring and Maintenance Program (continued)
- 4. In Regulatory Guide 1.129, Regulatory Position 3, Subsection 5.4.1, "State of Charge Indicator," the following statements in paragraph (d) may be omitted: "When it has been recorded that the charging current has stabilized at the charging voltage for three consecutive hourly measurements, the battery is near full charge.These measurements shall be made after the initially high charging current decreases sharply and the battery voltage rises to approach the charger output voltage." 5. In lieu of RG 1.129. Regulatory Position 7, Subsection 7.6, "Restoration," the following may be used: "Following the test, record the float voltage of each cell of the string." b. The program shall include the following provisions:
- 1. Actions to restore battery cells with float voltage<2.13 V;2. Actions to determine whether the float voltage of the remaining battery cells is 2.13 V when the float voltage of a battery cell has been found to be<2.13 V: 3. Actions to equalize and test battery cells that had been discovered with electrolyte level below the top of the plates: 4. Limits on average electrolyte temperature, battery connection resistance, and battery terminal voltage;and 5. A requirement to obtain specific gravity readings of all cells at each discharge test, consistent with manufacturer recommendations.
PALO VERDE UNITS 1,2,3 551 MNMN O 9 5.5-19 AMENDMENT NO.
IAfter SFP TransitionI Programs and Manuals 5.5 5.5 Programs and Manuals (continued) 5.5.19 Battery Monitoring and Maintenance Program (continued)
- 4. In Regulatory Guide 1.129. Regulatory Position 3, Subsection 5.4.1, "State of Charge Indicator." the following statements in paragraph (d) may be omitted: "When it has been recorded that the charging current has stabilized at the charging voltage for three consecutive hourly measurements, the battery is near full charge.These measurements shall be made after the initially high charging current decreases sharply and the battery voltage rises to approach the charger output voltage." 5. In lieu of RG 1.129, Regulatory Position 7, Subsection 1.6. "Restoration." the following may be used: "Following the test, record the float voltage of each cell of the string." b. The program shall include the following provisions:
- 1. Actions to restore battery cells with float voltage<2.13 V: 2. Actions to determine whether the float voltage of the remaining battery cells is 2.13 V when the float voltage of a battery cell has been found to be<2.13 V: 3. Actions to equalize and test battery cells that had been discovered with electrolyte level below the top of the plates: 4. Limits on average electrolyte temperature, battery connection resistance, and battery terminal voltage: and 5. A requirement to obtain specific gravity readings of all cells at each discharge test, consistent with manufacturer recommendations.
page 5.5-19 PALO VERDE UNITS 1,2.3 551 MNMN O 9 5.5-19 AMENDMENT NO.
[After SFP Transition]
Insert for page 5.5-19 5.5.21 Spent Fuel Storagqe Rack Neutron Absorber Monitoring Progqram Certain storage cells in the spent fuel storage racks utilize neutron absorbing material that is credited in the spent fuel storage rack criticality safety analysis to ensure the limitations of Technical Specifications 3.7.17 and 4.3.1.1 are maintained.
In order to ensure the reliability of the neutron absorber material, a monitoring program is provided to confirm the assumptions in the spent fuel pool criticality safety analysis.The Spent Fuel Storage Rack Neutron Absorber Monitoring Program shall require periodic inspection and monitoring of spent fuel pool test coupons and neutron absorber inserts on a performance-based frequency, not to exceed 10 years.Test coupons shall be inspected as part of the monitoring program. These inspections shall include visual, B-10 areal density and corrosion rate.Visual in-situ inspections of inserts shall also be part of the program to monitor for signs of degradation.
In addition, an insert shall be removed periodically for visual inspection, thickness measurements, and determination of retention force.
Enclosure Description and Assessment of Proposed License Amendment ATTACHMENT 2 Revised Technical Specifications Pages (Clean Copy)(Pages Provided for Before and After SEP Transition) 3.7.17-1 3.7.17-2 3.7.17-3 3.7.17-4 3.7.17-5 3.7.17-6 3.7.17-7 4.0-2 4.0-3 5.5-19 5.5-20 Before SFP transition Spent Fuel Assembly Storage 3.7.17 3.7 PLANT SYSTEMS 3.7.17 Spent Fuel Assembly Storage LCO 3.7.17 The combination of initial enrichment, burnup, and decay time of each fuel assembly stored in each of the four regions of the fuel storage pool shall be within the acceptable burnup domain for each region as shown in Figures 3.7.17-1, 3.7.17-2, or 3.7.17-3, and described in Specification 4.3.1.1.APPLICABILITY:
Whenever any fuel pool.assembly is stored in the fuel storage ACTIONS ________________
CONDITION REQUIRED ACTION COMPLETION TIME A. Requirements of the A.1------NOTE----
LCO not met. LCO 3.0.3 is not applicable.
Initiate action to Immediately move the noncomplying fuel assembly into an appropriate region.SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.17.1 Verify by administrative means the initial Prior to enrichment, burnup, and decay time of the storing the fuel assembly is in accordance with Figures fuel assembly 3.7.17-1, 3.7.17-2, or 3.7.17-3, and in the fuel Specification 4.3.1.1. storage pool.PALO VERDE UNITS 1,2,3 371- MNMN O -~3.7.17-1 AMENDMENT NO.
Before SFP transition Spent Fuel Assembly Storage 3.7.17 Figure 3.7.17-1 ASSEMBLY BURNUP VERSUS INITIAL ENRICHMENT for Regi on 2 20000 E E 1.5 2.0 2.5 3.0 3.5 4.0 Initial Enrichment, weight %4.5
- 5.0 4.80%limiting enrichment PALO VERDE UNITS 1,2,3 371- MNMN O 3.7.17-2 AMENDMENT NO.
Before SFP transition Spent Fuel Assembly Storage 3.7.17 Figure 3.7.17-2 ASSEMBLY BURNUP VERSUS INITIAL ENRICHMENT for Regi on 3 (at decay times from 0 to 20 years)E ci 1.5 2.0 2.5 3.0 3.5 4.0 4.5
- 5.0 Initial Enrichment, weight % 4.80%--q Oyers --m-5yars l~ears --X-15ear -- --2yeas ....limiting flmj --Oyars -U-yers ~-1yeas -E-l~yars oyers enrichment Decayl PALO VERDE UNITS 1,2,3 371- MNMN O 2 3.7.17- 3 AMENDMENT NO.
Before SFP transition Spent Fuel Assembly Storage 3.7.17 Figure 3.7.17-3 ASSEMBLY BURNUP VERSUS INITIAL ENRICHMENT for Regi on 4 (at decay times from 0 to 20 years)50000 45000 B S A 1 4 + -~ -S'IvuuU ______________
ACCE TABLE fo Region 4 35000 ____________
30000 B B S~.nnn ___________
I-E 20000 (I)(0J ,I)15000 10000 5000 0-1.5 J S S S S S NOT ACC EPTABLE ~or Region j4/ -U--B S ii~-i I SBB___ I ___ ___ ___ ___ __ B-#A~-- 4 + + +/.J S S S B S---S 0 B S B S S B B caytime.:+ t A------N+ote: Assern4Iy eflgbJ for Regi in 4 facua IBU>* U requiefent forjiI ghen irai ech ent and currnt d 2-.0 2.5 3.0 3.5 4.0 4.5 a 5.0 Initial Enrichment, weight % 4.80%limiting a -U-!--5 years -~-k--l years --UP-15 years --O-20 years .Jenrichment Decaylimej ye.PALO VERDE UNITS 1,2,3 371- MNMN O 2 3.7.17-4 AMENDMENT NO.
After SFP transition I Spent Fuel Assembly Storage 3.7.17 3.7 PLANT SYSTEMS 3.7.17 Spent Fuel Assembly Storage LCO 3.7.17 The combination of i niti al enrichment, burnup, and decay time of each fuel assembly shall be in compliance with the requirements specified in Tables 3.7.17-1 through 3.7.17-5.APPLICABILITY:
Whenever any fuel pool.assembly is stored in the fuel storage ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Requirements of the A.1------NOTE----
LCO not met. LCO 3.0.3 is not applicable.
Initiate action to Immediately move the noncomplying fuel assembly into an appropriate region.SURVEILLANCEREQUIREMENTS__________
SURVEILLANCE FREQUENCY SR 3.7.17.1 Verify by administrative means the initial Prior to enrichment, burnup, and decay time of the storing the fuel assembly is in accordance with Tables fuel assembly 3.7.17-1 through 3.7.17-5, Figure 3.7.17-1, in the fuel and Specification 4.3.1.1. storage pool.PALO VERDE UNITS 1,2,337171AEDNT O.I, 3.7.17-1 AMENDMENT NO.
After SFP transition Spent Fuel Assembly Storage 3.7.17 Table 3.7.17-1 Fuel Regions Ranked by Reactivity Fuel Region 1 Highest Reactivity (See Note 2)Fuel Region 2 Fuel Region 3 Fuel Region 4 Fuel Region 5 Fuel Region 6 Lowest Reactivity Notes: 1. Fuel Regions are defined by assembly average burnup. initial enrichment' and decay time as provided by Table 3.7.11-2 through Table 3.7.17-5.2. Fuel Regions are ranked in order of decreasing reactivity, e.g.. Fuel Region 2 is less reactive than Fuel Region 1. etc.3. Fuel Region 1 contains fuel with an initial maximum radially averaged enrichment up to 4.65 wt% 235U. No burnup is required.4. Fuel Region 2 contains fuel with an initial maximum radially averaged enrichment up to 4.65 wt% 235U with at least 16.0 GWd/MTU of burnup.5. Fuel Regions 3 through 6 are determined from the minimum burnup (BU) equation and coefficients provided in Tables 3.7.17-2 through 3.7.17-5.6. Assembly storage is controlled through the storage arrays defined in Figure 3.7.17-1.7. Each storage cell in an array can only be populated with assemblies of the Fuel Region defined in the array definition or a lower reactivity Fuel Region.'Initial Enrichment is the nominal 2 3 5 U enrichment of the central zone region of fuel, excluding axial blankets, prior to reduction in 2350 content due to fuel depletion.
If the fuel assembly contains axial regions of different 235U enrichment values, such as axial blankets, the maximum initial enrichment value is to be utilized.PALO VERDE UNITS 1,2,337.72AEDNTO.1, 3.7.17-2 AMENDMENT NO.
After SEP transition I Spent Fuel Assembly Storage 3.7.17 Table 3.7.17-2 Fue] Region 3: Burnup Requi rement Coefficients Decay Coefficients Time (yr.) Ai NmA A4 0 -1.5473 15.5395 -39.0197 24.1121 5 -1.4149 13.9760 -33.6287 18.3369 10 -1.3012 12.6854 -29.2539 13.6879 15 -1.0850 10.4694 -22.1380 6.3673 20 -0.9568 9.1487 -17.9045 2.0337 Notes: 1. Relevant uncertainties are explicitly included in the criticality analysis.
For instance, no additional allowance for burnup uncertainty or enrichment uncertainty is required.
For a fuel assembly to meet the requirements of a Fuel Region, the assembly burnup must exceed the "minimum burnup" (GWd/MTU) given by the curve fit for the assembly "decay 'time" and "initial enrichment." The specific minimum burnup (BU) required for each fuel assembly is calculated from the following equation: BU = Ai
- En 3 + A2
- En 2 + A3
- En + A4 2. Initial enrichment, En, is the maximum radial average 2 3enrichment.
Any En value between 2.55 wt% 2 3 5 U and 4.65 wt% 2 3 5 U may be used. Burnup credit is not required for an En below 2.55 wt% 2 3 5 U.3. It is acceptable to linearly interpolate between calculated BU limits based on decay time.4. The 20-year coefficients must be used to calculate the minimum BU for an assembly with a decay time of greater than 20 years.PALO VERDE UNITS 1,2,3 371- MNMN O 2 3.7.17-3 AMENDMENT NO.
After SFP transition Spent Fuel Assembly Storage 3.7.17 Table 3.7.17 -3 Fue] Region 4: Burnup Requirement Coefficients Decay Coefficients Time (yr.) Ai #0 0.4260 -6.2766 40.9264 -54.6813 5 0.2333 -4.1545 32.9080 -46.1161 10 0.4257 -6.2064 39.0371 -51.5889 15 0.5315 -7.3777 42.5706 -54.7524 20 0.5222 -7.3897 42.6587 -54.8201 Notes: 1. Relevant uncertainties are explicitly included in the criticality analysis.
For instance, no additional allowance for burnup uncertainty or enrichment uncertainty is required.
For a fuel assembly to meet the requirements of a Fuel Region, the assembly burnup must exceed the "minimum burnup" (GWd/MTU) given by the curve fit for the assembly "decay time" and "initial enrichment." The specific minimum burnup (BU) required for each fuel assembly is calculated from the following equation: BU =Ai
- En 3 + A2
- En 2 + A3
- En + A4 2. Initial enrichment, En, is the maximum radial average 2 3 5U enrichment.
Any En value between 1.15 wt% 2 3 5U and 4.65 wt% 2 3may be used. Burnup credit is not required for an En below 1.75 wt% 2 3 5 U.3. It is acceptable to linearly interpolate between calculated BU limits based on decay time.4. The 20-year coefficients must be used to calculate the minimum BU for an assembly with a decay time of greater than 20 years.PALO VERDE UNITS 1,2,3 371- MNMN O 2 3.7.17-4 AMENDMENT NO.
After SFP transition I Spent Fuel Assembly Storage 3.7.17 Table 3.7.17 -4 Fue] Region 5: Burnup Requirement Coefficients Decay Coeffi ci ents Time (yr.) Ai km# A4 0 -0.1114 -0.4230 20.9136 -32.8551 5 -0. 1232 -0. 4463 20. 8337 -32. 6068 10 -0. 2357 0.4892 18. 0192 -30. 0042 15 -0.1402 -0. 4523 20. 3745 -31. 7565 20 -0. 0999 -0. 8152 21. 0059 -31. 9911 Notes : 1. Relevant uncertainties are explicitly included in the criticality analysis.
For instance, no additional allowance for burnup uncertainty or enrichment uncertainty is required.
For a fuel assembly to meet the requirements of a Fuel Region, the assembly burnup must exceed the "minimum burnup" (GWd/MTU) given by the curve fit for the assembly "decay time" and "initial enrichment." The speci fi c minimum burnup (BU) requi red for each fuel assembly is calculated from the fol lowing equati on: BU= Ai
- En 3 + A2
- En 2 + A3
- En + A4 2. Initial enrichment, En. is the maximum radial average 2 3 5 U enrichment.
Any En value between 1.65 wt% 2 3 5 U and 4.65 wt% 2 3 5 U may be used. Burnup credit is not required for an En below 1.65 wt% 2 3 5 U.3. It is acceptable to linearly interpolate between calculated BU limits based on decay time.4. The 20-year coefficients must be used to calculate the minimum BU for an assembly with a decay time of greater than 20 years.PALO VERDE UNITS 1,2,3 371- MNMN D 3.7.17-5 AMENDMENT NO.
After SFP transition Spent Fuel Assembly Storage 3.7.17 Table 3.7.17 -5 Fuel Region 6: Burnup Requirement Coefficients Decay Coefficients Time (yr.) A A2 A4 A 0 0.7732 -9.3583 49.6577 -54.6847 5 0.7117 -8.4920 45.1124 -49.7282 10 0.6002 -7.2638 40.2603 -44.9348 15 0.5027 -6.2842 36.6715 -41.4934 20 0.2483 -3.7639 28.8269 -34.6419 Notes: 1. Relevant uncertainties are explicitly included in the criticality analysis.
For instance, no additional allowance for burnup uncertainty or enrichment uncertainty is required.
For a fuel assembly to meet the requirements of a Fuel Region, the assembly burnup must exceed the "minimum burnup" (GWd/MTU) given by the curve fit for the assembly "decay time" and "initial enrichment." The specific minimum burnup (BU) required for each fuel assembly is calculated from the fol lowi ng equati on: BU =A1
- En 3 + Am
- En 2 + A3
- En + A4 2. Initial enrichment, En. is the maximum radial average 235U enrichment.
Any En value between 1.45 wt% 2 3 5 U and 4.65 wt% 2 3 5 U may be used. Burnup credit is not required for an En below 1.45 wt% 2350.3. It is acceptable to linearly interpolate between calculated BU limits based on decay time.4. The 20-year coefficients must be used to calculate the minimum BU for an assembly with a decay time of greater than 20 years.PALO VERDE UNITS 1,2,3 371- MNMN O 3.7.17-6 AMENDMENT NO.
After SFP transition Spent Fuel Assembly Storage 3.7.17 Figure 3.7.17-1 Allowable Storage Arrays Array A 1 X Two Region 1 assemblies (1) checkerboarded with two blocked cells (X).The Region 1 assemblies are each in a cell with a stainless steel L-insert.
No NETCO-SNAP-IN inserts are credited.X 1 Array B 1 TC Two Region 1 assemblies (1) checkerboarded with two cells containing trash cans (TC). The Region 1 assemblies are each in a cell with a stainless steel L-insert.
Every cell without a stainless steel L- TC I insert must contain a NETCO-SNAP-IN insert.Array C Two Region 2 assemblies (2) checkerboarded with one Region 3 assembly 2 X (3) and one blocked cell (X). The Region 2 assemblies are each in a ....... ..............
cell with a stainless steel L-insert.
The Region 3 assembly is in a 3 2 cell containing a NETCO-SNAP-IN insert.Array 0 One Region 2 assembly (2) checkerboarded with three Region 4 2 4 assemblies (4). The Region 2 assembly and the diagonally located ..Region 4 assembly are each in a storage cell with a stainless steel L-insert. The two storage cells without a stainless steel L-insert 4 4 contain a NETCO-SNAP-IN insert.Array E 5 5 Four Region 5 assemblies (5). Two storage cells contain a stainless steel L-insert.
One cell contains a NETCO-SNAP-IN insert. One5 storage cell contains no insert.5 5 Array F 6 6 Four Region 6 assemblies (6). Two storage cells contain a stainless steel L-insert.
The other two cells contain no inserts. 6 6 Notes-1. The shaded locations indicate cells which contain a stainless steel [-insert.2. A blocked cell (X) contains a blocking device and only water in the active fuel region.3. NETCO-SNAP-IN inserts must be oriented in the same direction as the stainless steel [-inserts.
- 4. NETCO-SNAP-IN inserts are only located in cells without a stainless steel [-insert.5. Any cell containing a fuel assembly or a TC may instead be an empty (water-filled) cell in all storage arrays.6.Any storage array location designated for a fuel assembly may be replaced with non-fissile material.PALO VERDE UNITS 1,2,3 371- MNMN O 3.7.17-7 AMENDMENT NO.
Before SFP transition Design Features 4.0 4.0 DESIGN FEATURES (continued) 4.3 Fuel Storage 4.3.1 Criticality 4.3.1.1 The spent fuel storage racks are designed and shall be maintained with: a. Fuel assemblies having a maximum radially averaged U-235 enrichment of 4.80 weight percent;b. keff < 1.0 if fully flooded with unborated water, which includes an allowance for biases and uncertainties as descri bed i n Secti on 9.1I of the UFSAR;c. keff -< 0.-95 if fully flooded with water borated to 900 ppm, which includes an allowance for biases and uncertainties as described in Section 9.1 of the UFSAR.d. A nominal 9.5 inch center-to-center distance between adjacent storage cell locations.
- e. Region 1: Fuel shall be stored in a checkerboard (two-out-of-four) storage pattern. Fuel that qualifies to be stored in Regions 1, 2, 3, or 4 in accordance with Figures 3.7.17-1, 3.7.17-2, or 3.7.17 -3, may be stored in Region 1.f. Region 2: Fuel shall be stored in a repeating 3-by-4 storage pattern in which Region 2 (two-out-of-twelve) assemblies and Region 4 (ten-out-of-twelve) assemblies are mixed as shown in Section 9.1 of the UFSAR. Only fuel that qualifies to be stored in Regions 2, 3, or 4, in accordance with Figures 3.7.17-1, 3.7.17-2, or 3.7.17-3, may be stored in Region 2.g. Region 3: Fuel shall be stored in a four-out-of-four storage pattern. Only fuel that qualifies to be stored in Regions 3 or 4, in accordance with Figures 3.7.17-2 or 3.7.17-3, may be stored in Regi on 3.(conti nued)PALO VERDE UNITS 1,2,3 402AEDETN.1k 4.0-2 AMENDMENT NO.
Before SEP transition Design Features 4.0 4.0 DESIGN FEATURES (continued)
- h. Region 4: Fuel shall be stored in a repeating 3-by-4 storage pattern in which Region 2 (two-out-of-twelve) assemblies and Region 4 (ten-out-of-twelve) assemblies are mixed as shown in Section 9.1 of the UFSAR. Only fuel that qualifies to be stored in Region 4 in accordance with Figure 3.7.17-3 shall be stored in Region 4.4.3.1.2 The new fuel storage racks are designed and shall be maintained with: a. Fuel assemblies having a maximum radially averaged U-235 enrichment of 4.80 weight percent;b. keff 0.95 if fully flooded with unborated water, which includes an allowance for biases and uncertainties as described in Section 9.1 of the UFSAR: c. keff 0.98 if moderated by aqueous foam, which includes an allowance for biases and uncertainties as described in Section 9.1 of the UFSAR; and d. A nominal 17 inch center to center distance between fuel assemblies placed in the storage racks.,.4.3.2 Drainage The spent fuel storage pool is designed and shall be maintained to prevent inadvertent draining of the pool below elevation 137 feet -6 inches.4.3.3 Capacity The spent fuel storage pool is designed and shall be maintained with a storage capacity limited to no more than 1329 fuel assemblies.
PALO VERDE UNITS 1,2,3 403AEDETN.I~
4.0-3 AMENDMENT NO.
After SFP transition Design Features 4.0 4.0 k[DESIGN FEATURES (conti nued)4.3 Fuel Storage 4.3.1 Criticality 4.3.1.1 The spent fuel storage racks are designed and shall be mai ntai ned with : a. Fuel assemblies having a maximum radially averaged U-235 enrichment of 4.65 weight percent;b. keff < 1.0 if fully flooded with unborated water, which includes an allowance for biases and uncertai nti es as descri bed in Secti on 9.1 of the UFSAR;c. keff -< 0.95 if fully flooded with water borated to 1460 ppm, which includes an allowance for biases and uncertainties as described in Section 9.1 of the UFSAR.d. A nominal 9.5 inch center-to-center distance between adjacent storage cell locations.
- e. Fuel assemblies, are classified in Fuel Regions 1-6 as shown in Tables 3.7.17-1 through 3.7.17-5.(conti nued)PALO VERDE UNITS 1,2,340-AMNETNO 4.0-2 AMENDMENT NO.
After SFP transition Design Features 4.0 4.0 DESIGN FEATURES (continued) 4.3.1.2 The new fuel storage racks are designed and shall be maintained with: a. Fuel assemblies having a maximum radially averaged U-235 enrichment of 4.65 weight percent;b. keff 0.95 if fully flooded with unborated water, which includes an allowance for biases and uncertai nti es as descri bed i n Secti on 9.1I of the U FSAR;c. keff 0.98 if moderated by aqueous foam, which i ncl udes an all owance for bi ases and uncertai nti es as described in Section 9.1 of the UFSAR; and d. A nominal 18 inch (east-west) and 31 inch (north-south) center-to-center distance between fuel assemblies placed in the storage racks.4.3.2 Drainage The spent fuel storage pool is designed and shall be maintained to prevent inadvertent draining of the pool below elevation 137 feet -6 inches.4.3.3 Capacity The spent fuel storage pool is designed and shall be maintained with a storage capacity limited to no more than 1329 fuel assemblies.
PALO VERDE UNITS 1,2,340-AMNETNO 2g 4.0-3 AMENDMENT NO.
Before SFP transition Programs and Manuals 5.5 5.5 Programs and Manuals (continued) 5.5.19 Battery Monitoring and Maintenance Program (continued)
- 4. In Regulatory Guide 1.129, Regulatory Position 3, Subsection 5.4.1, "State of Charge Indicator," the following statements in paragraph (d) may be omitted: "When it has been recorded that the charging current has stabilized at the charging voltage for three consecutive hourly measurements, the battery is near full charge.These measurements shall be made after the initially high charging current decreases sharply and the battery voltage rises to approach the charger output voltage." 5. In lieu of RG 1.129, Regulatory Position 7, Subsection 7.6, "Restoration," the following may be used: "Following the test, record the float voltage of each cell of the string." b. The program shall include the following provisions:
- 1. Actions to restore battery cells with float voltage<2.13 V;2. Actions to determine whether the float voltage of the remaining battery cells is 2.13 V when the float voltage of a battery cell has been found to be<2.13 V;3. Actions to equalize and test battery cells that had been discovered with electrolyte level below the top of the plates;4. Limits on average electrolyte temperature, battery connection resistance, and battery terminal voltage;and 5. A requirement to obtain specific gravity readings of all cells at each discharge test, consistent with manufacturer recommendations.
PALO VERDE UNITS 1,2,3 551 MNMN O 9 5.5-19 AMENDMENT NO.
After SFP transition Programs and Manuals 5.5 5.5 Programs and Manuals (continued) 5.5.19 Battery Monitoring and Maintenance Program (continued)
- 4. In Regulatory Guide 1.129, Regulatory Position 3, Subsection 5.4.1, "State of Charge Indicator," the following statements in paragraph (d) may be omitted: "When it has been recorded that the charging current has stabilized at the charging voltage for three consecutive hourly measurements, the battery is near full charge.These measurements shall be made after the initially high charging current decreases sharply and the battery voltage rises to approach the charger output voltage." 5. In lieu of RG 1.129, Regulatory Position 7, Subsection 7.6, "Restoration," the following may be used: "Following the test, record the float voltage of each cell of the string." b. The program shall include the following provisions:
- 1. Actions to restore battery cells with float voltage<2.13 V;2. Actions to determine whether the float voltage of the remaining battery cells is 2.13 V when the float voltage of a battery cell has been found to be<2.13 V;3. Actions to equalize and test battery cells that had been discovered with electrolyte level below the top of the plates;4. Limits on average electrolyte temperature, battery connection resistance, and battery terminal voltage;and 5. A requirement to obtain specific gravity readings of all cells at each discharge test, consistent with manufacturer recommendations.
PALO VERDE UNITS 1,2,35519AEDNT O.49 5.5-19 AMENDMENT NO.
After SFP transition Programs and Manuals 5.5 5.5 Programs and Manuals (continued) 5.5.21 Spent Fuel Storage Rack Neutron Absorber Monitoring Program Certain storage cells in the spent fuel storage racks utilize neutron absorbing material that is credited in the spent fuel storage rack criticality safety analysis to ensure the limitations of Technical Specifications 3.7.17 and 4.3.1.1 are maintained.
In order to ensure the reliability of the neutron absorber material, a monitoring program is provided to confirm the assumptions in the spent fuel pool criticality safety analysis.The Spent Fuel Storage Rack Neutron Absorber Monitoring Program shall require periodic inspection and monitoring of spent fuel pool test coupons and neutron absorber inserts on a performance-based frequency, not to exceed 10 years.Test coupons shall be inspected as part of the monitoring program. These inspections shall include visual, B-IO areal density and corrosion rate.Visual in-situ inspections of inserts shall also be part of the program to monitor for signs of degradation.
In addition, an insert shall be removed periodically for visual inspection, thickness measurements, and determination of retention force.PALO VERDE UNITS 1,2,35520AEDNT O.19 5.5-20 AMENDMENT NO.
Enclosure Description and Assessment of Proposed License Amendment ATTACHM ENT 3 Marked-up Technical Specifications Bases Pages (Pages Provided for Before and After SEP Transition)
B 3.7.15-1 B 3.7.15-2 B 3.7.17-1 B 3.7.17-2 B 3.7.17-3 B 3.7.17-4 B 3.7.17-5 B 3.7.17-6 C IBefore SFP transitionJ Fuel Storage Pool Boron Concentration B 3.7.15 B 3.7 PLANT SYSTEMS B 3.7.15 Fuel Storage Pool Boron Concentration BASES BACKGROUND As described in LCO 3.7.17, 'Spent Fuel Assembly Storage," fuel assemblies are stored in the spent fuel racks in accordance with criteria based on initial enrichment and discharge burnup. Although the water in the spent fuel pool is normally borated to >_ 2150 ppm, the criteria that limit the storage of a fuel assembly to specific rack locations is conservatively developed without taking credit for boron.In order to maintain the spent fuel pool keff < 1.0, a soluble boron concentration of 900 ppm is required to maintain the spent fuel pool keff _< 0.95 assuming the most limiting single fuel mishandling accident.APPLICABLE SAFETY ANALYSES A fuel assembly could be inadvertently loaded into a spent fuel rack location not allowed by LCO 3.7.17 (e.g., an unirradiated fuel assembly or an insufficiently depleted fuel assembly).
Another type of postulated accident is associated with a fuel assembly that is dropped onto the fully loaded fuel pool storage rack or between a rack and the pool walls. These incidents could have a positive reactivity effect, decreasing the margin to criticality.
However, the negative reactivity effect of the soluble boron compensates for the increased reactivity caused by these postulated accident scenarios.
The concentration of dissolved boron in the fuel pool satisfies Criterion 2 of 10 CFR 50.36 (c)(2)(ii).
LCO The specified concentration of dissolved boron in the fuel pool preserves the assumptions used in the analyses of the potential accident scenarios described above. This concentration of dissolved boron is the minimum required concentration for fuel assembly storage and movement within the fuel pool.APPLICABILITY This LCO applies whenever any fuel assembly is stored in the spent fuel pool in order to comply with the TS 4.3.1.1.c design requirement that keff 0.95.(conti nued)PALO VERDE UNITS 1,2,3B37.51RVSO B 3.7.15-1 REVISION 3 Before SFP transition Fuel Storage Pool Boron Concentration B 3.7.15 BASES (continued)
ACTIONS A.1 and A.2 The Required Actions are modified by a Note indicating that LCO 3.0.3 does not apply.When the concentration of boron in the spent fuel pooi is less than required, immediate action must be taken to preclude an accident from happening or to mitigate the consequences of an accident in progress.
This is most efficiently achieved by immediately suspending the movement of fuel assemblies.
This does not preclude the movement of fuel assemblies to a safe position.
In addition, action must be immediately initiated to restore boron concentration to within limit.If moving fuel assemblies while in MODE 5 or 6, LCO 3.0.3 would not specify any action. If moving fuel assemblies while in MODE 1, 2, 3, or 4, the fuel movement is independent of reactor operation.
Therefore, inability to suspend movement of fuel assemblies is not sufficient reason to require a reactor shutdown.SURVEILLANCE SR 3.7.15.1 REQUI REMENTS This SR verifies that the concentration of boron in the spent fuel pool is within the required limit. As long as this SR is met, the analyzed incidents are fully addressed.
The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.REFERENCES
- 1. UFSAR, Section 9.1.2.2. PVNGS Operating License Amendments 82, 69 and 54 for Units 1, 2 and 3, respectively, and associated NRC Safety Evaluation dated September 30, 1994.3. 13-N-001-1900-1221-1, "Palo Verde Spent Fuel Pool Criticality Analysis," ABB calculation A-PV-FE-0106, revision 3, dated January 15, 1999.PALO VERDE UNITS 1,2,3B37152RVSO
.B 3.7.15-2 REVISION Ifer FP ransition Fuel Storage Pool Boron Concentration B 3.7.15 B 3.7 PLANT SYSTEMS B 3.7.15 Fuel Storage Pool Boron Concentration ,an(time.BASES BACKGROUND As dscrbedin.CO 3.7.17. "Spent Fuel Assembly Afuel decieassembliesi tre stored in the spent fuel racks accordance with ,4iteria based on initial enrichment
~discharge burnup' Although the .ater .pent fue pool is .. nomly borte to 2l50 ppm the critei that limt the strg of.. a fue to sp.cifi rac location-s, i soluble boron concentration of m is required to maintain the spent fuel pool kef _< 0.ming the most limiting fuel mishandling accident. 1460 APPLICABLE SAFETY ANALYSES There could also be a misload of multiple fuel assemblies into fuel rack locations not allowed by LCO 3.7.17.A fuel assembly could be inadvertently loaded into a spent fuel rack location not allowed by LCO 3.7.17 (e.g., an unirradiated fuel assembly or an insufficiently depleted fuel Another type of postulated accident is a fuel assembly that is dropped onto the fully 1 ~ted fuel pool storage rack or between a rack and the 4mol walls. These incidents could have a positive activity effect, decreasing the margin to criticality.
However, the negative reactivity effect of the soluble boron compensates for the increased reactivity caused by these postulated accident scenarios.
The concentration of dissolved boron in the fuel pool satisfies Criterion 2 of 10 CFR 50.36 (c)(2)(ii).
LCO The specified concentration of dissolved boron in the fuel pool preserves the assumptions used in the analyses of the potential accident scenarios described above. This concentration of dissolved boron is the minimum required concentration for fuel assembly storage and movement within the fuel pool.APPLICABILITY This LCO applies whenever any fuel assembly is stored in the spent fuel pool in order to comply with the TS 4.3.1.1.c design requirement that keff 0.95.(conti nued)PALO VERDE UNITS 1,2,3B37151RVSO B 3.7.15-1 REVISION IAfter SFP transition Fuel Storage Pool Boron Concentration B 3.7.15 BASES (continued)
ACTIONS A.1 and A.2 The Required Actions are modified by a Note indicating that LCO 3.0.3 does not apply.When the concentration of boron in the spent fuel pool is less than required, immediate action must be taken to preclude an accident from happening or to mitigate the consequences of an accident in progress.
This is most efficiently achieved by immediately suspending the movement of fuel assemblies.
This does not preclude the movement of fuel assemblies to a safe position.
In addition, action must be immediately initiated to restore boron concentration to within limit.If moving fuel assemblies while in MODE 5 or 6, LCO 3.0.3 would not specify any action. If moving fuel assemblies while in MODE 1, 2, 3. or 4, the fuel movement is independent of reactor operation.
Therefore, inability to suspend movement of fuel assemblies is not sufficient reason to require a reactor shutdown.SURVEILLANCE SR 3.7.15.1 REQUIREMENTS This SR verifies that the concentration of boron in the spent fuel pool is within the required limit. As long as this SR is met, the analyzed incidents are fully addressed.
The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.REFERENCES
- 82. 69 3nd{ 51 fo-'r""Criticality Safety Analysis for Palo Verde Nuclear Generating Station Units 1, 2, and 3" (Proprietary), WCAP-1 8030-P, Revision 0, September 2015.PALO VERDE UNITS 1,2,3 B 3.7.15-2 REVISION I Before SFP transition I Spent Fuel Assembly Storage B 3.7.17 B 3.7 PLANT SYSTEMS B 3.7.17 Spent Fuel Assembly Storage BAS ES BAGCKGROUND The spent fuel storage is designed to store either new (nonirradiated) nuclear fuel assemblies, or burned (irradiated) fuel assemblies in a vertical configuration underwater.
The storage pool was originally designed to store up to 1329 fuel assemblies in a borated fuel storage mode.The current storage configuration, which allows credit to be taken for boron concentrati.on, burnup, and decay time, and does not require neutron absorbing (boraflex) storage cans, provides for a maximum storage of 1209 fuel assemblies in a four-region configuration.
The design basis of the spent fuel cooling system, however, is to provide adequate cooling to the spent fuel during all operating conditions (including full core offload) for only 1205 fuel assemblies (UFSAR section 9.1.3). Therefore, an additional four spaces are mechanically blocked to limit the maximum number of fuel assemblies that may be stored in the spent fuel storage pool to 1205.Region 1 is comprised of two 9x8 storage racks and one 12x8 storage rack. Cell blocking devices are placed in every other storage cell location in Region 1 to maintain a two-out-of-four checkerboard configuration.
These cell blocking devices prevent inadvertent insertion of a fuel assembly into a cell that is not allowed to contain a fuel assembly.Region 3 is comprised of three 9x8 storage racks and one 9x9 storage rack in Units 2 and 3. Region 3 is comprised of four 9x8 storage racks and one 9x9 storage rack in Unit 1. Since fuel assemblies may be stored in every Region 3 cell location, no cell blocking devices are installed in Region 3.Regions 2 and 4 are mixed and are comprised of seven 9x8 storage racks and three 12x8 storage racks in Units 2 and 3, Regions 2 and 4 are mixed and are comprised of six 9x8 storage racks and three 12x8 storage racks in Unit 1. Regions 2 and 4 are mixed in a repeating 3x4 storage pattern in which two-out-of-twelve cell locations are designated Region 2 and ten-out-of-twelve cell locations are designated Region 4 (see UFSAR Figures 9.1-7 and 9.1-7A). Since fuel assemblies may be stored in every Region 2 and Region 4 cell location, no cell blocking devices are installed in Region 2 and Region 4.(conti nued)PALO VERDE UNITS 1,2,3B371-1RVSO B 3.7.17-i REVISION
[Before SFP transition Spent Fuel Assembly Storage B 3.7.17 BASES BACKGROUND (conti nued)The spent fuel storage cells are installed in parallel rows with a nominal center-to-center spacing of 9.5 inches. This spacing, a minimum soluble boron concentration of 900 ppm, and the storage of fuel in the appropriate region based on assembly burnup in accordance with TS Figures 3.7.17-1, 3.7.17-2, and 3.7.17-3 is sufficient to maintain a keff of0.95 for fuel of original maximum radially averaged enrichment of up to 4.80%.APPLICABLE SAFETY ANALYSES The spent fuel storage pool is designed for non-criticality by use of adequate spacing, credit for boron concentration, and the storage of fuel in the appropriate region based on assembly burnup in accordance with TS Figures 3.7.17-1, 3.7.17-2, and 3.7.17-3.
The design requirements related to criticality (TS 4.3.1.1) are keff < 1.0 assuming no credit for boron and keff 0.95 taking credit for soluble boron. The burnup versus enrichment requirements (TS Figures 3.7.17-1, 3.7.17-2, and 3.7.17-3) are developed assuming keff < 1.0 with no credit taken for soluble boron, and that keff 0.95 assuming a soluble boron concentration of 900 ppm and the most limiting single fuel mishandling accident.The analysis of the reactivity effects of fuel storage in the spent fuel storage racks was performed by ABB-Combustion Engineering (CE) using the three-dimensional Monte Carlo code KENO-VA with the updated 44 group ENDF/B-5 neutron cross section library. The KENO code has been previously used by CE for the analysis of fuel rack reactivity and have been benchmarked against results from numerous critical experiments.
These experiments simulate the PVNGS fuel storage racks as realistically as possible with respect to parameters important to reactivity such as enrichment and assembly spacing.The modeling of Regions 2, 3, and 4 included several conservative assumptions.
These assumptions neglected the reactivity effects of poison shims in the assemblies and structural grids. These assumptions tend to increase the calculated effective multiplication factor (keff) of the racks. The stored fuel assemblies were modeled as CE 16x16 assemblies with a nominal pitch of 0.5065 inches between fuel rods, a fuel pellet diameter of 0.3255 inches, and a UO(2)density of 10.31 g/cc.(conti nued)PALO VERDE UNITS 1,2,3B37172RVSO 3 B 3.7.17-2 REVISION 3 IBefore SFP transition]
Spent Fuel Assembly Storage B 3.7.17 BASES APPLICABLE SAFETY ANALYSES (conti nued)KENO-Va calculations were used to construct curves of burnup versus initial enrichment for decay times in 5 year increments from 0 to 20 years for both Regions 3 and 4 (TS Figures 3.7.17-2 and 3.7.11-3) such that all points on the curves produce a keff value (including all biases and uncertainties) of < 1.0 for unborated water. Biases associated with methodology and water temperature were included, and uncertainties associated with methodology, KENO-Va calculation, fuel enrichment, fuel rack pitch, fuel rack and L-insert thickness, pellet stack density, and asymmetric fuel assembly loading were included.
KENO-Va calculations were also performed to determine the soluble boron concentration required to maintain the spent fuel pool keff (including all biases and uncertainties) 0.95 at a 95% probability/95%
confidence level. A soluble boron concentration of 900 ppm is required to assure that the spent fuel pool keff remains 0.95 at all times. This soluble boron concentration accounts for the positive reactivity effects of the most limiting single fuel mishandling event and uncertainties associated with fuel assembly reactivity and burnup. This method of reactivity equivalencing has been accepted by the NRC (Reference
- 3) and used for numerous other spent fuel storage pools that take credit for burnup, decay time, and soluble boron.Most abnormal storage conditions will not result in an increase in the keff of the racks. However, it is possible to postulate events, with a burnup and enrichment combination outside of the acceptable area in TS Figure 3.1.17-1, or with a burnup, decay time, and enrichment combination outside of the acceptable area in IS Figures 3.7.17-2 or 3.7.17-3, which could lead to an increase in reactivity.
These events would include an assembly drop on top of a rack or between a rack and the pool walls, or the misloading of an assembly.
For such events, partial credit may be taken for the soluble boron in the spent fuel pool water to ensure protection against a criticality accident since the staff does not require the assumption of two unlikely, independent, concurrent events (double contingency principle).
Although a soluble boron concentration of only 900 ppm is required to assure that keff remains 0.95 assuming the single most limiting fuel mishandling event, TS 3.7.15 conservatively requires the presence of 2150 ppm of soluble boron in the spent fuel pool water. As such, the reduction in keff caused by the required soluble boron concentration more than offsets the reactivity addition caused by credible accidents, and the staff criterion of keff 0.95 is met at all times.(conti nued)PALO VERDE UNITS 1,2,3B37.-3RVSO 3 B 3.7.17-3 REVISION 3 IBefore SEP transitionI Spent Fuel Assembly Storage B 3.7.17 BASES APPLICABLE The criticality aspects of the spent fuel pool meet the SAFETY ANALYSES requirements of General Design Criterion 62 for the (continued) prevention of criticality in fuel storage and handling.The spent fuel pool heat load calculations were based on a full pool with 1205 fuel assemblies.
From the spent fuel pool criticality analysis, the number of fuel assemblies that can be stored in the four-region configuration is 1209 fuel assemblies.
The design basis of the spent fuel cooling system, however, is to provide adequate cooling to the spent fuel during all operating conditions (including full core offload) for only 1205 fuel assemblies (UFSAR section 9.1.3). Therefore, an additional four spaces are mechanically blocked to limit the maximum number of fuel assemblies that may be stored in the spent fuel storage pool to 1205.The original licensing basis for the spent fuel pool allowed for spent fuel to be loaded in either a 4x4 array or a checkerboard array, depending on the use of borated poison.A fuel handling accident was assumed to occur with maximum loading of the pool. The fuel pool rack construction precludes more than one assembly from being impacted in a fuel handling accident.
The UFSAR analysis conclusion regarding the worst scenario for a dropped assembly (in which the horizontal impact of a fuel assembly on top of the Spent fuel assembly damages fuel rods in the dropped assembly but does not impact fuel in the stored assemblies) continues to be limiting.The spent fuel assembly storage satisfies Criterion 2 of 10 CFR 50.36 (c)(2)(ii).
LCO The restrictions on the placement of fuel assemblies within the spent fuel pool, according to Figures 3.7.17-1, 3.7.17-2, and 3.7.17-3 in the accompanying LCO, ensures that the keff of the spent fuel pool will always remain < 1.0 assuming the pool to be flooded with unborated water. The restrictions are consistent with the criticality safety analysis performed for the spent fuel pool according to Figures 3.1.17-1, 3.1.17-2, and 3.7.17-3 in the accompanying LCO.Specification 4.3.1.1 provides additional details for fuel storage in each of the four Regions.(conti nued)PALO VERDE UNITS 1,2,3B371-4RVSO B 3.7.17-4 REVISION IBefore SFP transition Spent Fuel Assembly Storage B 3.7.17 BASES APPLICABILITY This LCO applies whenever any fuel assembly is stored in the spent'fuel pool.ACTIONS A. 1 Requi red Acti on A.1 i is modi fi ed by a Note i ndi cating that LCO 3.0.3 does not apply.When the configuration of fuel assemblies stored in the spent fuel pool is not in accordance with Figures 3.7.17-1, 3.7.17-2, and 3.7.17-3, immediate action must be taken to make the necessary fuel assembly movement(s) to bring the configuration into compliance with Figures 3.7.17-1, 3.7.17-2, and 3.7.17-3.If moving irradiated fuel assemblies while in MODE 5 or 6, LCO 3.0.3 would not specify any action. If moving irradiated fuel assemblies while in MODE 1, 2, 3, or 4, the fuel movement is independent of reactor operation.
Therefore, in either case, inability to move fuel assemblies is not sufficient reason to require a reactor shutdown.SURVEILLANCE SR 3.7.17.1 REQU IREMENTS This SR verifies by administrative means that the initial enrichment and burnup of the fuel assembly is in accordance with Figures 3.7.17-1, 3.7.17-2, and 3.7.17-3 in the accompanying LCO and Speci fi cati on 4.3.1.1.To manually determine the allowed SFP region for a fuel assembly, the actual burnup is compared to the burnup requirement for the given initial enrichment and appropriate decay time from Figure 3.7.17-1, 3.7.17-2, or 3.7.17-3.
If the actual burnup is greater than or equal to the burnup requirement, then the fuel assembly is eligible to be stored in the corresponding region. If the actual burnup is less than the burnup requirement, then the comparison needs to be repeated using another curve for a lower numbered region. Note the fol lowing : (continued)
PALO VERDE UNITS 1,2,3B371-5RVSO 3 B 3.7.17-5 REVISION 3 SBefore SFP transition I Spent Fuel Assembly Storage B 3.7.17 BASES SURVEILLANCE
- that a fuel assembly that does not meet the burnup REQUIREMENTS requirement for Region, 2 must be stored in Region 1, (conti nued)* that any fuel assembly may be stored in Region 1,* that any fuel assembly may be stored in a lower numbered region than the region for which it qualifies because burnup requirements decrease as region numbers decrease (refer also to Tech Spec 4.3.1.1),*and that comparing actual burnup to the burnup requirement for zero decay time will always be correct or conservative.
REFERENCES
- 1. UFSAR, Sections 9.1.2 and 9.1.3.2. PVNGS Operating License Amendments 82, 69, and 54 for Units 1, 2, and 3 respectively, and associated NRC Safety Evaluation, dated September 30, 1994.3. Letter to T. E. Collins, U.S. NRC to T. Greene, WOG,"Acceptance for Referencing of Licensing Topical Report WCAP-14416-P, Westinghouse Spent Fuel Rack Methodology (TAC NO. M93254)", October 25, 1996.4. 13-N-001-1900-1221-1, "Palo Verde Spent Fuel Pool Criticality Analysis," ABB calculation A-PV-FE-0106, revision 03, dated January 15, 1999.5. Westinghouse letter NF-APS-10-19, "Criticality Safety Evaluation of the Spent Fuel Pool Map with a Proposed Region 3 Increase," dated February 25, 2010.PALO VERDE UNITS 1,2,3 B371- EIIN~B 3.7.17-6 REVISION IAfter SFP transition Spent Fuel Assembly Storage B 3.7.17* The design basis of the spent fuel pool cooling system is to B 3.7 PLANT SYSTEMS IProvide adequate cooling to the spent fuel pool during all Ioperating conditions (including full core offload) for up to 1205 B 3.7.17 Spent Fuel Assembl lfuel assemblies (UFSAR Section 9.1.3).BASES /I l t BACKGROUND The spent fuel is designed ostor2/'2 h4 new (nonirradiated) fuel asse ~lies,--g¢ burned (irradiated) fuel assemblies in Jvertical configuration underwater.
The storage pool sorgnally designed to store up to 1329 fuel assemblies i eae ue trg oe The current ,storg config.. ration.. w4. ..hich allow credit. be taken for boron.conentration burnup,..
an dea time, and..dones mnot requirelm mneutron absorbing (boraflex)v ca-ns provides-ltc a maximu~m stor=ag of 1209Q "fuel assemblies¢ in a four reio configuration=.
.The desig basis of the spent fue,,l cooling syste,-, however, is to provide adequatea cooling1 to the core offload frl on nly 1205 asseamblie (UFSTAR sct,,'io~n a 1-- -'- Therefore, an =additional four.spaces.are.mechanically..
bmlockedn to limitf t-he mai~m~um nulmberA t-hatCell de,,'ie are..pnlaceda in ever... other storagea cell location in Region 1 to mainti-mn a out- of four checkerbar configuration.
blockin devices preventf inadverte4-nt-insertion ofc a fueal into a cell that i not ..alloe to contain a fuel assembly.sto-rr~age r~ack- in UI I-4t- 2 and'- 3. Regio n 3 is co .v-mprise of,- fou"r,,, 9x3 sto"rage,- and oneh 9v9 stor,-age,,-
r,.ack. in- Unit- 1. 4Re fue~l asse. mblies- may be kstoredw in evry,,n, Regio"-n 3 ,.-c ll loation4,'
no cell blocking devices are installed in Region 3.Regions 2 andi A are mived- rand4 are seve\-n 9v8ra-,ck and three 12x r~acks,, in Un-it -2c and 3, Regio~ns 2 and I nar m'ixed and are of sixv 9v8 storag are mixed, in a repatnga-n 3vA storwage attern in w.hic-h twor out of cell locations are designated Region-,, 2 and ten out of twelve! cell. locations
..... designated Region I (see-, UFS^R Figres 1 "7 and 9.1 7A). S1nc ..-fuel asse-,mblies may, be stored in e,.ery Region 2 and Region '1 cell location no.. cel blocking de,,ice ..... installed in Region 2 =an Region I secontinued PALO VERDE UNITS 1,2,3B3.171RVSO 4 B 3.7.17-1 REVISION Insert I for TS Bases 3.7.17 page B 3.7.17-1 The spent fuel storage cells are installed in parallel rows with a nominal center-to-center spacing of 9.5 inches. This spacing, a minimum soluble boron concentration of 1460 ppm, the use of neutron-absorbing panels, and the storage of fuel in the appropriate region based on fuel assembly initial enrichment, discharge burnup, and decay time in accordance with TS Tables 3.7.17-1 through 3.7.17-5 is sufficient to maintain keff< 0.95 for fuel of initial maximum radially averaged enrichment of up to 4.65 wt%.Disused CEAs, in-core instruments, and other material is stored in trash cans. A trash can may be stored in any location that is approved to store a fuel assembly.
No special nuclear material (SNM) may be stored in a trash can.
IAfter SFP transitionI Spent Fuel Assembly Storage B 3.7.17 BASES BAAC"KG RUDf/ KF (conti4nued)4 The, fucl storage cclls installed4 in parallel rows, wit4h a nominal cntne-r to nnl cnter-spacing of 9. innHch. ThL spacingn a slubklc, boron co'-nceant-ration of ppm assembkl,, b 'urnp i/n accordance4
... th. 4 -.T--
3.17 1,_3.7.17 2, andl 3.T.1T 3 is sufficie'-4nt-to maintain keTT-of0.9 foK r, fueJl ofz orgna,, nl maxi{mum radia,4:lly.
ofi upn to- /1.0/.APPLICABLE Thc ...nt fu-el strge....
pol is dest"igned,4 for' nonrl SAFETY ANALYSES critcalty' by ' "se of spacing, credi.t- for. boron and3'1 tlhe "fueali4n theappn,'nropriate-TS: FZ{TiguresC 3.7.17 1, 3.7.17 2 3,4 an 37117 3. Th& -.de-l-gm ke, i 1.0 assu-mi ng no credi t for boron and _4 Q--_c_taki"ng credit- fo sonclulehl boron. The burnup The analysis of-effe'tII'ct of fuel_ storage9 in Insrt 1th fue,,l soag k-vr acks whas pefome by BB :Combust/ion Engineering.- (CE)usin the threen,44n dime.'-,.,-nsinlMot-arocd hEO A with th-e updated #an-1f~ ENDF/ neutron:-l cros'/Insert section. library.....Th E.....code ha been.... previously..used.by L I~~~C for the, analysis of~n fuel 4-rack' rea~ctivity~ haver ben.,,-,.,, bencmark\I d agaih l-ns resu:3lts, fAn-,rom numerousR c: nritical storagem, ra:lck as as possblel w.,ith respectn4 to The of DRegio~ns
- 2. :3,nd4 '1 inc'ludeda
- seeal, neglecPted4 theTheeassumptoions',n
'tend4 to increnase tlhe calculated3-n, e,-ffective,,, multlication-:34,-,, facto',r-, (keT of t',-he SThe stored fuel assembli!es w.ere modeled as CE 1..x16v assembl'ies writh a3 nomilnal p'itc-h of 0.5065 "inches betweean fueal rods,,4 a fuerl pellet-of 0.325 inhe,,",ac nd a3 llO2)density of 10.31 g/cc.(conInIued) lbU PALO VERDE UNITS 1,2,3B371-2RVSO B 3.7.17-2 REVISION %
Insert 2 for TS Bases 3.7.17 page B 3.7.17-2 The nuclear criticality safety analysis in References 1 and 2 considered the following reactivity-increasing accidents:
- Misload of a single assembly into an unacceptable storage location* Multiple assemblies misloaded in series due to a common cause* Spent fuel pooi temperature outside the allowable operating range* Dropped and misplaced fresh fuel assembly* Seismic event* Inadvertent removal of a NETCO-SNAP-IN~
rack insert In each case, the spent fuel assembly storage met the requirements of 10 CFR 50.68(b)(4).
Thus, the spent fuel storage facility is designed for noncriticality by use of adequate spacing, and neutron absorbing panels considering initial enrichment, fuel burnup, and decay time.
IAfter SEP transitionI Spent Fuel Assembly Storage B 3.7.17 BASE[S APPLICTfABLEI ,AI'E'TV ,ANAILVSES versus. initial1 enrichmen fo dccay tims in 5 ..e.r increments from 0 to 20 years for both Regions 3 and 4 (TS 3.7.17 2 and 3.7.17 3) uciih that- all poimnt-s orn t'he curves a kr valu(inc/mludingr all bi.ases .and u-ncertainties) of 1.0 for.. unboratcd wae...assccriated,-
wit4h met-hodo"log nd wm- .ater, tenmpr~n3-atre were inluedIcl...
and unerante assuciaedtainthiethodologyat QKEO nr',a calculation fuel, enrichme, nt,l fue rckl pitch, fuel.-,
requret maintainthespenful pool1 ....he1,, (Incui ngm'-4m' allC biasesP un'ertaF'inties) 4h~
95 r, obabiCll ty/95,%PIP confidene-h.-
leve.3P ,--A,,4! soluble, boromn H co4ncentratmiokn of,900ppm i eurdt sueta h pn Most' abnr,,,,mal storage, will not-H resuclt in an increasea in the ke, of te racks.~ Howevr., , it- is possibletlhe aceptab-3le area in TS 3.7.17 1.',, orwith t-he accepntable area in TS Figuresc 3.7.17 2 or 3.7.17 3, w.hich could latonincrease inreactivity.
These evnt wul includen an asse.,,mbly.
dropr- topof a rac orI a rack,the pool walls, or the misloading of a~n assembly.
Fo boron.in the spent fuel poo water to ensure..protection requ~l'ire tlhe of'- tl-wo' unlikenly, indepndnt,~mn solubl boron concentratio of only 900 ppm is require tom assure. that h, remain 4o 0.9 assuming.
t£. he single. mostfuel poom'l water.P As schl," t-he reductiofnm in h,, Pcause by%1 required4 sof' lub~le boroPn concentration'{' t'han staff criterion of ke, _ 0.9 is met at all times.PALO VERDE UNITS 1,2,3B37.73RVSO B 3.7.17-3 REVISION %
IAfter SFP transitionI Spent Fuel Assembly Storage B 3.7.17 BASES ADDI The criticlity a.pect. of t he spent fuel pool meet the The spent poo,, heat load, calculations w;ere based on a fu, ll pool 1205F fuenl assembl-ies.
Frm t-rhe spent fueil pool0+', criticalt nalsiThem Mnumber~-
of fueln assemblies, thatl can beC in~ thel~ four- reginl =configuatin i 1209Efue asse--imblies.
basi oMf thel spntr fuel are mehnhmica=lly blockedH to -imait the mav-imum number oxf fuenl assemblies t-hat- mayl be3 sltoreda -in the spntl fueil pool lomading of the pool. The fuel pool rac~k coinsrctmlion than. oine assmbl fromA being, impacted uiona ful andrling ah .c¥cieant, Th UFSAR analysnis 3cncnml u -ion .h regardn g worst-f scnai for, a dropped assembly (il-n whch-,ni fuenl assembly, damages fuel rodsc in the dropped~,-
assembly, not -imnrImr"'
fuel "in tlhe st-ored assemhliesncontinues,n to The spent fuel assembly storage satisfies Criterion 2 of 10 CFR 50.36 (c)(2)(ii).
I abes3..1-Itheog spent7- fulpoa awyeand Fi1u0e3assuming the.LCOare rsictonsitn woh the crtcli~~ety safetye anlslies performe frthe spent f el pool, according to Figure 3.7..... 1.....and 3..17 3in in theaccmpanying LCO,.nue ht h efo Specification 4.3.1.1 provides additional details for fuel storage in each of the Regios (conti nued)PALO VERDE UNITS 1,2,3 B 3.7.17-4 REVISION IAfter SFP transition Spent Fuel Assembly Storage B 3.7.17 BASES APPLICABILITY This LCO applies whenever any fuel assembly is stored in the spent fuel pooi.ACTIONS A.1 I Tables 3.7.17-1 through 3.7.17-5 and Figure 3.7.17-1'-7 I v w Required Action A.1 is modified by a Note indicating t LCO 3.0.3 does not apply. /When the configuration of fuel assemblies s ed the spent fuel pool is not in accordance with Fgrs 3.7.17 2, and 3.7.17 3, immediate action must; taken to make the necessary fuel assembly movement(s) bring the configuration into compliance with N7!3.7.17-2, and 3.7.17 3.If moving irradiated fuel assemblies while in MODE 5 or 6, LCO 3.0.3 would not specify any action. If moving irradiated fuel assemblies while in MODE 1. 2. 3. or 4, the fuel movement is independent of reactor operation.
Therefore, in either case, inability to move fuel assemblies is not sufficient reason to require a reactor shutdown.SURVEILLANCE REQU IREMENTS This SR administrative means that the initial enrichment of the fuel assembly is in accordance with 3 71!7 1 3_717 2. and 3_7.7 3 in the ofTbe371-thurough 3..1-5 performanc ofh thi SR.....
-ensure compliance with Specification 4.3.1.1. l(ot ud PALO VERDE UNITS 1,2,3B371-5RVSO B 3.7.17-5 REVISION %
IAfter SFP transitionI Spent Fuel Assembly Storage B 3.7.17 BASES requ~lirement-fo~r 2 must e'-h sltored a-in Regnion 1, (continRued)
- that anyJ fuel assembly/
may/ be stored a-in aRegion 1, 4-tht- any\ "fuel assembly, may be stored -in a loer, nuimbereda regionn t-han the rego w: .h-ic-h it- qalifies bc au-se~~~~REFERENE
.USR aSo oTc peci0n 9 1.3 n .1.3),* n thatI_ act al bur ,nup '- to the burnup -,A+," reurmn o eodcytiewllasb orc 4.- 13g_ N 001190 1221 1, "Palo,," \/erdeH Spent- Fuel Pool/Cr'4-it,-1icality Analysi,c " ,ABBclulto PV FE/ 0106, Evaluatio-4n ofc the Spntn4 Fuel Pool Map w.,it-h a Proposedca Reg-,ionr 3 Inrease, ac" datedH Febrary, 25-, 2010.-"Criticality Safety Analysis for Palo Verde Nuclear Generating Station Units 1, 2, and 3" (Proprietary), WCAP-1 8030-P, Revision 0, September 2015.PALO VERDE UNITS 1,2,3 B 3.7.17-6 REVISION Enclosure Description and Assessment of Proposed License Amendment ATTACHMENT 4 List of Regulatory Commitments Enclosure Description and Assessment of Proposed License Amendment Attachment 4 List of Regulatory Commitments Reuaor omimn Due Date/Milestone 1. APS will implement procedural controls to require verification that fresh fuel assemblies are not placed face-adjacent to one another before completing a fuel move.2. The transition to the new SEP configuration will be completed in all three units in accordance with the Spent Fuel Pool Transition Plan within two years of the NRC approval date of the amendment or by December 31, 2019, whichever is later.Within the requested 90-day implementation period following NRC approval.Within two years of the NRC approval date of the amendment or by December 31, 2019, whichever is later.