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Category:Letter type:GO
MONTHYEARGO2-24-004, License Amendment Request to Revise Emergency Plan2024-01-30030 January 2024 License Amendment Request to Revise Emergency Plan GO2-24-003, Relief Requests for the Columbia Generating Station Fifth Ten-Year Interval Inservice Testing2024-01-29029 January 2024 Relief Requests for the Columbia Generating Station Fifth Ten-Year Interval Inservice Testing GO2-24-005, Docket No. 50-397 Supplement to Reply to a Notice of Violation; EA-21-1702024-01-0808 January 2024 Docket No. 50-397 Supplement to Reply to a Notice of Violation; EA-21-170 GO2-23-136, Guarantee of Payment of Deferred Premium2023-12-20020 December 2023 Guarantee of Payment of Deferred Premium GO2-23-135, Notice of Readiness for Supplemental Inspection2023-12-14014 December 2023 Notice of Readiness for Supplemental Inspection GO2-23-130, Reply to a Notice of Violation; EA-23-0542023-12-14014 December 2023 Reply to a Notice of Violation; EA-23-054 GO2-23-105, Licensing Basis Document Update and Biennial Commitment Change Report2023-12-12012 December 2023 Licensing Basis Document Update and Biennial Commitment Change Report GO2-23-107, Application to Revise Technical Specifications to Adopt TSTS-584, Eliminate Automatic RWCU System Isolation on SLC Initiation2023-12-0505 December 2023 Application to Revise Technical Specifications to Adopt TSTS-584, Eliminate Automatic RWCU System Isolation on SLC Initiation GO2-23-126, Docket No. 50-397 Voluntary Response to Regulatory Issue Summary 2023-012023-11-28028 November 2023 Docket No. 50-397 Voluntary Response to Regulatory Issue Summary 2023-01 GO2-23-121, Request for Exemption from Enhanced Weapons, Firearms Background Checks, and Security Event Notifications Implementation2023-11-27027 November 2023 Request for Exemption from Enhanced Weapons, Firearms Background Checks, and Security Event Notifications Implementation GO2-23-097, In-Service Inspection Summary Report for the Twenty-Sixth Refueling Outage (R26)2023-09-0606 September 2023 In-Service Inspection Summary Report for the Twenty-Sixth Refueling Outage (R26) GO2-23-100, Technical Specification Section 5.6.4 Post Accident Monitoring Instrumentation 14-Day Report for Inoperable Suppression Pool Level Indication2023-08-31031 August 2023 Technical Specification Section 5.6.4 Post Accident Monitoring Instrumentation 14-Day Report for Inoperable Suppression Pool Level Indication GO2-23-056, Application to Revise Technical Specifications to Adopt TSTF-230, Add New Condition B to LCO 3.6.2.3, RHR Suppression Pool Cooling2023-08-29029 August 2023 Application to Revise Technical Specifications to Adopt TSTF-230, Add New Condition B to LCO 3.6.2.3, RHR Suppression Pool Cooling GO2-23-093, Supplement to Reply to a Notice of Violation: EA-21-1702023-07-27027 July 2023 Supplement to Reply to a Notice of Violation: EA-21-170 GO2-23-068, Notification of Completion of Commitments Required Prior to Entry Into the Period of Extended Operation2023-07-19019 July 2023 Notification of Completion of Commitments Required Prior to Entry Into the Period of Extended Operation GO2-23-090, Reply to a Notice of Violation2023-07-12012 July 2023 Reply to a Notice of Violation GO2-23-073, Notification of NPDES Permit Issuance2023-06-26026 June 2023 Notification of NPDES Permit Issuance GO2-23-072, Cycle 27 Core Operating Limits Report2023-06-0505 June 2023 Cycle 27 Core Operating Limits Report GO2-23-055, Independent Spent Fuel Storage Installation, 2022 Annual Radiological Environmental Operating Report2023-05-15015 May 2023 Independent Spent Fuel Storage Installation, 2022 Annual Radiological Environmental Operating Report GO2-23-006, License Amendment Request to Clean-Up Operating License and Appendix a Technical Specifications2023-05-0101 May 2023 License Amendment Request to Clean-Up Operating License and Appendix a Technical Specifications GO2-23-053, 2022 Annual Environmental Operating Report2023-04-26026 April 2023 2022 Annual Environmental Operating Report GO2-23-057, 2022 Annual Radioactive Effluent Release Report2023-04-25025 April 2023 2022 Annual Radioactive Effluent Release Report GO2-23-012, Application to Revise Technical Specifications to Adopt Tstf-541, Add Exceptions to Surveillance Requirements for Valves and Dampers Locked in the Actuated Position2023-03-27027 March 2023 Application to Revise Technical Specifications to Adopt Tstf-541, Add Exceptions to Surveillance Requirements for Valves and Dampers Locked in the Actuated Position GO2-23-041, Generation Station - Level of Financial Protection - Annual Reporting Requirement2023-03-15015 March 2023 Generation Station - Level of Financial Protection - Annual Reporting Requirement GO2-23-038, Plant Decommissioning Fund Status Report2023-03-15015 March 2023 Plant Decommissioning Fund Status Report GO2-23-034, Report of Changes or Errors in Emergency Core Cooling System Loss of Coolant Accident Analysis Models Pursuant to 10 CFR 50.462023-03-13013 March 2023 Report of Changes or Errors in Emergency Core Cooling System Loss of Coolant Accident Analysis Models Pursuant to 10 CFR 50.46 GO2-23-035, Supplement to License Amendment Request to Adopt TSTF-580, Provide Exception from Entering Mode 4 with No Operable RHR Shutdown Cooling2023-03-0909 March 2023 Supplement to License Amendment Request to Adopt TSTF-580, Provide Exception from Entering Mode 4 with No Operable RHR Shutdown Cooling GO2-23-019, Emergency Plan - Summary of Changes and Analysis for Revision 68 of the EP-012023-02-0808 February 2023 Emergency Plan - Summary of Changes and Analysis for Revision 68 of the EP-01 GO2-22-131, Guarantee of Payment of Deferred Premium2022-12-19019 December 2022 Guarantee of Payment of Deferred Premium GO2-22-122, Response to Request for Additional Information Related to License Amendment Request to Adopt TSTF-505, Revision 2, Provide Risk-Informed Extended Completion Times - RITSTF Initiative 4b2022-11-28028 November 2022 Response to Request for Additional Information Related to License Amendment Request to Adopt TSTF-505, Revision 2, Provide Risk-Informed Extended Completion Times - RITSTF Initiative 4b GO2-22-076, Status of License Renewal Commitments2022-10-31031 October 2022 Status of License Renewal Commitments GO2-22-110, Supplement to License Amendment Request to Adopt 10 CFR 50.69, Risk-Informed Categorization and Treatment of Structures, Systems and Components for Nuclear Power Reactor2022-10-17017 October 2022 Supplement to License Amendment Request to Adopt 10 CFR 50.69, Risk-Informed Categorization and Treatment of Structures, Systems and Components for Nuclear Power Reactor GO2-22-096, Supplement to License Amendment Request to Adopt TSTF-505, Revision 2, Provide Risk-Informed Extended Completion Times - RITSTF Initiative 4b2022-10-0404 October 2022 Supplement to License Amendment Request to Adopt TSTF-505, Revision 2, Provide Risk-Informed Extended Completion Times - RITSTF Initiative 4b GO2-22-085, Docket No. 50-397, Evacuation Time Estimate Analysis2022-08-30030 August 2022 Docket No. 50-397, Evacuation Time Estimate Analysis GO2-22-075, Response to Request for Additional Information Related to Revised Pressure-Temperature Limit Curves2022-07-11011 July 2022 Response to Request for Additional Information Related to Revised Pressure-Temperature Limit Curves GO2-22-072, Summary of Changes and Analysis for Revision 57 of PPM 13.14.4 - Emergency Equipment Maintenance and Testing2022-07-11011 July 2022 Summary of Changes and Analysis for Revision 57 of PPM 13.14.4 - Emergency Equipment Maintenance and Testing GO2-22-042, Application to Revise Technical Specifications to Adopt TSTF-580, Provide Exception from Entering Mode 4 with No Operable RHR Shutdown Colling Using the Consolidated Line Item Improvement Process2022-05-25025 May 2022 Application to Revise Technical Specifications to Adopt TSTF-580, Provide Exception from Entering Mode 4 with No Operable RHR Shutdown Colling Using the Consolidated Line Item Improvement Process GO2-22-062, Registration of Spent Fuel Cask Use2022-05-19019 May 2022 Registration of Spent Fuel Cask Use GO2-22-049, Supplement to Licensing Amendment Request to Adopt 10CFR50.69, Risk-Informed Categorization and Treatment of Structures, Systems and Components for Nuclear Power Rectors2022-05-10010 May 2022 Supplement to Licensing Amendment Request to Adopt 10CFR50.69, Risk-Informed Categorization and Treatment of Structures, Systems and Components for Nuclear Power Rectors GO2-22-054, Independent Spent Fuel Storage Installation, 2021 Annual Radiological Environmental Operating Report2022-05-0909 May 2022 Independent Spent Fuel Storage Installation, 2021 Annual Radiological Environmental Operating Report GO2-22-056, Registration of Spent Fuel Cask Use2022-04-21021 April 2022 Registration of Spent Fuel Cask Use GO2-22-055, 2021 Annual Environmental Operating Report2022-04-20020 April 2022 2021 Annual Environmental Operating Report GO2-22-048, 2021 Annual Radioactive Effluent Release Report2022-04-15015 April 2022 2021 Annual Radioactive Effluent Release Report GO2-22-045, Registration of Spent Fuel Cask Use2022-04-0606 April 2022 Registration of Spent Fuel Cask Use GO2-22-044, Level of Financial Protection - Annual Reporting Requirement2022-03-21021 March 2022 Level of Financial Protection - Annual Reporting Requirement GO2-22-038, Docket No. 50-397, Report of Changes or Errors in Emergency Core Cooling System Loss of Coolant Accident Analysis Models Pursuant to 10 CFR 50.462022-03-0909 March 2022 Docket No. 50-397, Report of Changes or Errors in Emergency Core Cooling System Loss of Coolant Accident Analysis Models Pursuant to 10 CFR 50.46 GO2-22-031, Response to Audit Plan and Online Reference Portal Request2022-02-15015 February 2022 Response to Audit Plan and Online Reference Portal Request GO2-22-011, Docket No. 72-35; Independent Spent Fuel Storage Installation, 2021 Annual Effluent Release Report2022-02-0808 February 2022 Docket No. 72-35; Independent Spent Fuel Storage Installation, 2021 Annual Effluent Release Report GO2-22-001, License Amendment Request to Adopt TSTF-505, Revision 2, Provide Risk-Informed Extended Completion Times - RITSTF Initiative 4b2022-02-0303 February 2022 License Amendment Request to Adopt TSTF-505, Revision 2, Provide Risk-Informed Extended Completion Times - RITSTF Initiative 4b GO2-22-012, Independent Spent Fuel Storage Installation - Biennial 50.59/72.48 Report2022-01-24024 January 2022 Independent Spent Fuel Storage Installation - Biennial 50.59/72.48 Report 2024-01-08
[Table view] Category:License-Application for Facility Operating License (Amend/Renewal) DKT 50
MONTHYEARGO2-24-004, License Amendment Request to Revise Emergency Plan2024-01-30030 January 2024 License Amendment Request to Revise Emergency Plan GO2-23-107, Application to Revise Technical Specifications to Adopt TSTS-584, Eliminate Automatic RWCU System Isolation on SLC Initiation2023-12-0505 December 2023 Application to Revise Technical Specifications to Adopt TSTS-584, Eliminate Automatic RWCU System Isolation on SLC Initiation GO2-23-056, Application to Revise Technical Specifications to Adopt TSTF-230, Add New Condition B to LCO 3.6.2.3, RHR Suppression Pool Cooling2023-08-29029 August 2023 Application to Revise Technical Specifications to Adopt TSTF-230, Add New Condition B to LCO 3.6.2.3, RHR Suppression Pool Cooling GO2-23-012, Application to Revise Technical Specifications to Adopt Tstf-541, Add Exceptions to Surveillance Requirements for Valves and Dampers Locked in the Actuated Position2023-03-27027 March 2023 Application to Revise Technical Specifications to Adopt Tstf-541, Add Exceptions to Surveillance Requirements for Valves and Dampers Locked in the Actuated Position GO2-23-035, Supplement to License Amendment Request to Adopt TSTF-580, Provide Exception from Entering Mode 4 with No Operable RHR Shutdown Cooling2023-03-0909 March 2023 Supplement to License Amendment Request to Adopt TSTF-580, Provide Exception from Entering Mode 4 with No Operable RHR Shutdown Cooling GO2-22-042, Application to Revise Technical Specifications to Adopt TSTF-580, Provide Exception from Entering Mode 4 with No Operable RHR Shutdown Colling Using the Consolidated Line Item Improvement Process2022-05-25025 May 2022 Application to Revise Technical Specifications to Adopt TSTF-580, Provide Exception from Entering Mode 4 with No Operable RHR Shutdown Colling Using the Consolidated Line Item Improvement Process GO2-22-001, License Amendment Request to Adopt TSTF-505, Revision 2, Provide Risk-Informed Extended Completion Times - RITSTF Initiative 4b2022-02-0303 February 2022 License Amendment Request to Adopt TSTF-505, Revision 2, Provide Risk-Informed Extended Completion Times - RITSTF Initiative 4b GO2-21-133, License Amendment Request to Adopt 10 CFR 50.69, Risk Informed Categorization and Treatment of Structures, Systems and Components for Nuclear Power Reactors2021-11-0909 November 2021 License Amendment Request to Adopt 10 CFR 50.69, Risk Informed Categorization and Treatment of Structures, Systems and Components for Nuclear Power Reactors GO2-21-016, License Amendment Request to Change Technical Specification 3.4.11 Reactor Coolant System Pressure and Temperature Limits2021-10-13013 October 2021 License Amendment Request to Change Technical Specification 3.4.11 Reactor Coolant System Pressure and Temperature Limits GO2-21-042, License Amendment Request to Adopt TSTF-546, Revise APRM Channel Adjustment Surveillance Requirement2021-05-0808 May 2021 License Amendment Request to Adopt TSTF-546, Revise APRM Channel Adjustment Surveillance Requirement GO2-20-104, On-Site Cooling System Sediment Disposal2020-12-21021 December 2020 On-Site Cooling System Sediment Disposal GO2-20-108, License Amendment Request to Adopt TSTF-439, Elinate Second Completion Times Limitng Time from Discovery of Failure to Meet an LCO2020-12-0202 December 2020 License Amendment Request to Adopt TSTF-439, Elinate Second Completion Times Limitng Time from Discovery of Failure to Meet an LCO GO2-20-095, Application to Revise Technical Specifications to Adopt TSTF-582, 'Reactor Pressure Vessel Water Inventory Control (RPV WIC) Enhancements2020-09-24024 September 2020 Application to Revise Technical Specifications to Adopt TSTF-582, 'Reactor Pressure Vessel Water Inventory Control (RPV WIC) Enhancements GO2-20-063, Exigent License Amendment Request for Extension of Implementation Period for Amendment 255 Change to Control Room Air Conditioning System2020-04-15015 April 2020 Exigent License Amendment Request for Extension of Implementation Period for Amendment 255 Change to Control Room Air Conditioning System GO2-20-052, Exigent License Amendment Request for One-Time Change to Completion Times for Inoperable Electrical Distribution Systems2020-04-15015 April 2020 Exigent License Amendment Request for One-Time Change to Completion Times for Inoperable Electrical Distribution Systems GO2-20-004, Application to Revise Technical Specifications to Adopt TSTF-564, Safety Limit MCPR2020-01-27027 January 2020 Application to Revise Technical Specifications to Adopt TSTF-564, Safety Limit MCPR GO2-20-008, License Amendment Request to Revise Technical Specifications to Adopt TSTF-566, Revise Actions for Inoperable RHR Shutdown Cooling Subsystems Using the Consolidated Line Item Improvement Process2020-01-27027 January 2020 License Amendment Request to Revise Technical Specifications to Adopt TSTF-566, Revise Actions for Inoperable RHR Shutdown Cooling Subsystems Using the Consolidated Line Item Improvement Process GO2-19-090, License Amendment Request to Revise Technical Specifications to Adopt TSTF-563, Revise Instrument Testing Definitions to Incorporate the Surveillance Frequency Control Program2019-08-15015 August 2019 License Amendment Request to Revise Technical Specifications to Adopt TSTF-563, Revise Instrument Testing Definitions to Incorporate the Surveillance Frequency Control Program GO2-19-053, License Amendment Request to Remove Operating License Condition 2.C.(11) and Attachment 32019-03-27027 March 2019 License Amendment Request to Remove Operating License Condition 2.C.(11) and Attachment 3 GO2-19-027, License Amendment Request for Licensing Basis Change to Control Room Air Conditioning System2019-02-25025 February 2019 License Amendment Request for Licensing Basis Change to Control Room Air Conditioning System GO2-17-137, License Amendment Request to Update Appendix B to the Renewed Facility Operating License to Incorporate 2017 Biological Opinion2017-12-18018 December 2017 License Amendment Request to Update Appendix B to the Renewed Facility Operating License to Incorporate 2017 Biological Opinion GO2-17-181, License Amendment Request to Revise Technical Specifications to Adopt TSTF-551, Revise Secondary Containment Surveillance Requirements2017-12-12012 December 2017 License Amendment Request to Revise Technical Specifications to Adopt TSTF-551, Revise Secondary Containment Surveillance Requirements GO2-17-038, License Amendment Request to Revise Technical Specifications to Adopt TSTF-542, Reactor Pressure Vessel Water Inventory Control.2017-10-23023 October 2017 License Amendment Request to Revise Technical Specifications to Adopt TSTF-542, Reactor Pressure Vessel Water Inventory Control. GO2-17-028, License Amendment Request for Change to Technical Specification 3.5.1, ECCS-Operating.2017-07-25025 July 2017 License Amendment Request for Change to Technical Specification 3.5.1, ECCS-Operating. GO2-17-009, License Amendment Request - Revise Technical Specification 5.5.12 for Permanent Extension of Type a Test and Type C Leak Rate Test Frequencies2017-03-27027 March 2017 License Amendment Request - Revise Technical Specification 5.5.12 for Permanent Extension of Type a Test and Type C Leak Rate Test Frequencies GO2-17-062, License Amendment Request to Revise Technical Specifications to Adopt TSTF-523, Revision 22017-03-27027 March 2017 License Amendment Request to Revise Technical Specifications to Adopt TSTF-523, Revision 2 GO2-16-119, License Amendment Request for One-Time 7 Day Extension of Completion Time for TS Condition 3.5.1.A, 3.6.1.5.A, and 3.6.2.3.A2016-11-0808 November 2016 License Amendment Request for One-Time 7 Day Extension of Completion Time for TS Condition 3.5.1.A, 3.6.1.5.A, and 3.6.2.3.A ML16243A5152016-08-30030 August 2016 License Amendment Request for Reclassifying Quality Group of Low Temperature Portions of Reactor Water Cleanup System ML16244A8332016-08-30030 August 2016 License Amendment Request for Reclassifying Quality Group of Low Temperature Portions of Reactor Water Cleanup System - Drawing M523-1-3, Revision 10, Flow Diagram Reactor Water Clean-up System Radwaste Bldg. ML16210A5282016-07-28028 July 2016 Request for Amendment to Emergency Plan GO2-16-087, License Amendment Request for Changes to Technical Specification 2.1.1, Reactor Core SLs2016-07-12012 July 2016 License Amendment Request for Changes to Technical Specification 2.1.1, Reactor Core SLs GO2-16-096, License Amendment Request to Revise Operating License and Technical Specifications for Measurement Uncertainty Recapture Power Uprate2016-06-28028 June 2016 License Amendment Request to Revise Operating License and Technical Specifications for Measurement Uncertainty Recapture Power Uprate GO2-16-079, Erratum for License Amendment Request to Modify Technical Specification Surveillance Requirement (SR) 3.4.3.1 and SR 3.4.4.1 Safety/Relief Valves (Srvs) Setpoint Lower Tolerance2016-05-18018 May 2016 Erratum for License Amendment Request to Modify Technical Specification Surveillance Requirement (SR) 3.4.3.1 and SR 3.4.4.1 Safety/Relief Valves (Srvs) Setpoint Lower Tolerance GO2-16-046, License Amendment Request to Modify Technical Specification Surveillance Requirement (SR) 3.4.3.1 and SR 3.4.4.1 Safety/Relief Valves (Srvs) Setpoint Lower Tolerance2016-05-10010 May 2016 License Amendment Request to Modify Technical Specification Surveillance Requirement (SR) 3.4.3.1 and SR 3.4.4.1 Safety/Relief Valves (Srvs) Setpoint Lower Tolerance GO2-16-047, License Amendment Request to Revise Technical Specifications to Adopt TSTF-522, Revision 02016-03-0303 March 2016 License Amendment Request to Revise Technical Specifications to Adopt TSTF-522, Revision 0 GO2-15-070, Application for Technical Specification Change (TSTF-427) to Add LCO 3.0.9 on the Unavailability of Barriers Using the Consolidated Line Item Improvement Process2015-09-0202 September 2015 Application for Technical Specification Change (TSTF-427) to Add LCO 3.0.9 on the Unavailability of Barriers Using the Consolidated Line Item Improvement Process GO2-16-112, License Amendment Request for Reclassifying Quality Group of Low Temperature Portions of Reactor Water Cleanup System - Drawing M523-3, Revision 10, Flow Diagram Reactor Water Clean-up System Radwaste Bldg.2015-06-17017 June 2015 License Amendment Request for Reclassifying Quality Group of Low Temperature Portions of Reactor Water Cleanup System - Drawing M523-3, Revision 10, Flow Diagram Reactor Water Clean-up System Radwaste Bldg. GO2-15-075, Exigent License Amendment Request - Extension of Implementation Period for Amendment 232 Changing Technical Specification Table 3.3.1.1-1 Function 7 Scram Discharge Volume Water Level-High.2015-05-15015 May 2015 Exigent License Amendment Request - Extension of Implementation Period for Amendment 232 Changing Technical Specification Table 3.3.1.1-1 Function 7 Scram Discharge Volume Water Level-High. GO2-15-007, License Amendment Request for Adoption of Technical Specification Task Force Traveler-425, Revision 32015-03-17017 March 2015 License Amendment Request for Adoption of Technical Specification Task Force Traveler-425, Revision 3 ML14336A0062014-11-17017 November 2014 License Amendment Request for Technical Specification Change to Safety Limit Minimum Critical Power Ratio ML14268A2332014-09-0404 September 2014 Erratum for License Amendment Request for Adoption of Technical Specifications Task Force (TSTF) Traveler TSTF-423, Revision 1, Using the Consolidated Line Item Improvement Process GO2-14-126, License Amendment Request to Revise Technical Specification Surveillance Requirement for the Ultimate Heat Sink2014-08-22022 August 2014 License Amendment Request to Revise Technical Specification Surveillance Requirement for the Ultimate Heat Sink ML14234A4572014-08-12012 August 2014 License Amendment Request for Adoption of Technical Specification Task Force (TSTF) Traveler TSTF-423, Revision 1, Using the Consolidated Line Item Improvement Process GO2-14-103, Proposed Revision to License Amendment Request to Adopt TSTF-493, Revision 4, Option a, and Response to Request for Additional Information2014-06-19019 June 2014 Proposed Revision to License Amendment Request to Adopt TSTF-493, Revision 4, Option a, and Response to Request for Additional Information ML14086A3892014-03-18018 March 2014 License Amendment Request to Revise Technical Specifications to Adopt TSTF-535, Revision 0 ML13316A0092013-10-31031 October 2013 License Amendment Request for Change to Emergency Core Cooling Systems Surveillance Requirements GO2-13-138, License Amendment Request for Adoption of TSTF-493, Revision 4, Option a2013-10-0202 October 2013 License Amendment Request for Adoption of TSTF-493, Revision 4, Option a ML13247A6592013-07-25025 July 2013 License Amendment Request for Adoption of Technical Specification Task Force Traveler TSTF 477, Revision 3 GO2-13-002, License Amendment Request to Implement Prnm/Arts/Mellla - Revised Reports2013-01-0707 January 2013 License Amendment Request to Implement Prnm/Arts/Mellla - Revised Reports GO2-12-003, License Amendment Request to Make Administrative and Editorial Changes to Technical Specifications and the Operating License2012-01-0909 January 2012 License Amendment Request to Make Administrative and Editorial Changes to Technical Specifications and the Operating License 2024-01-30
[Table view] Category:Updated Final Safety Analysis Report (UFSAR)
MONTHYEARGO2-21-131, Attachment 5: Columbia Generating Station FSAR, Amendment 66, Part 9 of 102021-12-15015 December 2021 Attachment 5: Columbia Generating Station FSAR, Amendment 66, Part 9 of 10 ML21349B3882021-12-15015 December 2021 Attachment 5: Columbia Generating Station FSAR, Amendment 66, Part 8 of 10 ML21349B3872021-12-15015 December 2021 Attachment 5: Columbia Generating Station FSAR, Amendment 66, Part 7 of 10 ML21349B3852021-12-15015 December 2021 Attachment 5: Columbia Generating Station FSAR, Amendment 66, Part 6 of 10 ML21349B3682021-12-15015 December 2021 Attachment 5: Columbia Generating Station FSAR, Amendment 66, Part 2 of 10 ML21349B3672021-12-15015 December 2021 Attachment 5: Columbia Generating Station FSAR, Amendment 66, Part 1 of 10 ML21349B3842021-12-15015 December 2021 Attachment 5: Columbia Generating Station FSAR, Amendment 66, Part 5 of 10 ML21349B3802021-12-15015 December 2021 Attachment 5: Columbia Generating Station FSAR, Amendment 66, Part 4 of 10 GO2-19-162, Amendment 65 to Final Safety Analysis Report, Chapter 12, Radiation Protection2019-12-0505 December 2019 Amendment 65 to Final Safety Analysis Report, Chapter 12, Radiation Protection GO2-17-190, Amendment 64 to Final Safety Analysis Report, Chapter 12, Radiation Protection2017-12-31031 December 2017 Amendment 64 to Final Safety Analysis Report, Chapter 12, Radiation Protection ML17355A6702017-12-31031 December 2017 Columbia Generating Station - Amendment 64 to Final Safety Analysis Report, Chapter 13, Conduct of Operations ML17355A6682017-12-31031 December 2017 Amendment 64 to Final Safety Analysis Report, Chapter 10, Steam and Power Conversion System ML17355A6672017-12-31031 December 2017 Amendment 64 to Final Safety Analysis Report, Chapter 9, Auxiliary Systems ML17355A6662017-12-31031 December 2017 Amendment 64 to Final Safety Analysis Report, Chapter 8, Electric Power ML17355A6642017-12-31031 December 2017 Amendment 64 to Final Safety Analysis Report, Chapter 7, Instrumentation and Control Systems ML17355A6632017-12-31031 December 2017 Amendment 64 to Final Safety Analysis Report, Chapter 4, Reactor ML17355A6622017-12-31031 December 2017 Columbia Generating Station - Amendment 64 to Final Safety Analysis Report, Chapter 3, Design Criteria - Structures, Components, Equipment, and Systems ML17355A6612017-12-31031 December 2017 Amendment 64 to Final Safety Analysis Report, Chapter 1, Introduction and General Description of Plant ML17355A6602017-12-31031 December 2017 Amendment 64 to Final Safety Analysis Report, General Table of Contents and List of Effective Pages ML15355A5122015-12-21021 December 2015 Final Safety Analysis Report, Amendment 63, Chapter 12, Figure 12.3-19 Through Page 12.5-21 ML15355A4882015-12-21021 December 2015 Final Safety Analysis Report, Amendment 63, General Table of Contents ML15355A4892015-12-21021 December 2015 Final Safety Analysis Report, Amendment 63, Chapter 1 - Introduction and General Description of Plant ML15355A4962015-12-21021 December 2015 Final Safety Analysis Report, Amendment 63, Chapter 3 - Design Criteria - Structures, Components, Equipment, and Systems, Table of Contents GO2-15-159, Final Safety Analysis Report, Amendment 63, Chapter 13 -Conduct of Operations2015-12-21021 December 2015 Final Safety Analysis Report, Amendment 63, Chapter 13 -Conduct of Operations ML15355A5112015-12-21021 December 2015 Final Safety Analysis Report, Amendment 63, Chapter 12 - Radiation Protection ML15355A5102015-12-21021 December 2015 Final Safety Analysis Report, Amendment 63, Chapter 10 - Steam and Power Conversion System ML15355A5092015-12-21021 December 2015 Final Safety Analysis Report, Amendment 63, Chapter 9 - Auxiliary Systems ML15355A5082015-12-21021 December 2015 Final Safety Analysis Report, Amendment 63, Chapter 8 - Electric Power ML15355A5072015-12-21021 December 2015 Final Safety Analysis Report, Amendment 63, Chapter 7 - Instrumentation and Control Systems ML15355A5002015-12-21021 December 2015 Final Safety Analysis Report, Amendment 63, Chapter 4 - Reactor ML15355A4992015-12-21021 December 2015 Final Safety Analysis Report, Amendment 63, Appendix 3A -Plant Design Assessment Report for Safety/Relief Valves and Loss-of-Coolant Accident Loads ML15355A4902015-12-21021 December 2015 Final Safety Analysis Report, Amendment 63, Chapter 2 - Site Characteristics ML15355A4972015-12-21021 December 2015 Final Safety Analysis Report, Amendment 63, Chapter 3, Figure 3.5-48 - Cantilever Barrier Structure to Page 3.6-101 ML15355A4982015-12-21021 December 2015 Final Safety Analysis Report, Amendment 63, Chapter 3, Figures 3.6-21 to 3.12-4 ML14010A3012013-12-30030 December 2013 Final Safety Analysis Report, Amendment 62, Chapter 6 - Engineered Safety Features ML14010A2972013-12-30030 December 2013 Final Safety Analysis Report, Amendment 62, Chapter 3 - Design Criteria - Structures, Components, Equipment, and Systems, - Figure 3.6-21 Through Figure 3.12-4 ML14010A2912013-12-30030 December 2013 Final Safety Analysis Report, Amendment 62, General Table of Contents ML14010A2922013-12-30030 December 2013 Final Safety Analysis Report, Amendment 62, List of Effective Pages ML14010A2932013-12-30030 December 2013 Final Safety Analysis Report, Amendment 62, Chapter 1 - Introduction and General Description of Plant ML14010A2942013-12-30030 December 2013 Final Safety Analysis Report, Amendment 62, Chapter 2 - Site Characteristics ML14010A2952013-12-30030 December 2013 Final Safety Analysis Report, Amendment 62, Chapter 3 - Design Criteria - Structures, Components, Equipment, and Systems, Table of Contents Through 3.5-46 ML14010A2962013-12-30030 December 2013 Final Safety Analysis Report, Amendment 62, Chapter 3 - Design Criteria - Structures, Components, Equipment, and Systems, Page 3.6-1 Through Page 3.6-101 ML14010A2982013-12-30030 December 2013 Final Safety Analysis Report, Amendment 62, Appendix 3A - Plant Design Assessment Report for Safety/Relief Valves and Loss-of-Coolant Accident Loads GO2-13-174, Final Safety Analysis Report, Amendment 62, Appendix J - Shielding Evaluation Report2013-12-30030 December 2013 Final Safety Analysis Report, Amendment 62, Appendix J - Shielding Evaluation Report ML14010A3002013-12-30030 December 2013 Final Safety Analysis Report, Amendment 62, Chapter 5 - Reactor Coolant System and Connected Systems ML14010A3022013-12-30030 December 2013 Final Safety Analysis Report, Amendment 62, Chapter 7 - Instrumentation and Control Systems ML14010A3032013-12-30030 December 2013 Final Safety Analysis Report, Amendment 62, Chapter 8 - Electric Power ML14010A3042013-12-30030 December 2013 Final Safety Analysis Report, Amendment 62, Chapter 9 - Auxiliary Systems ML14010A3052013-12-30030 December 2013 Final Safety Analysis Report, Amendment 62, Chapter 10 - Steam and Power Conversion System ML14010A3062013-12-30030 December 2013 Final Safety Analysis Report, Amendment 62, Chapter 11 - Radioactive Waste Management 2021-12-15
[Table view] |
Text
X ENERGY N NORTHWEST PeoFple Vision* Solutions P.O. Box 968
- 99352-0968 April 17, 2006 G02-06-059 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D.C. 20555-0001
Subject:
COLUMBIA GENERATING STATION, DOCKET NO. 50-397 LICENSE AMENDMENT REQUEST TO ADOPT ORIGEN-ARP FOR CALCULATING FUEL POOL DECAY HEAT LOAD
Reference:
Letter dated November 29, 2005 from DD Chamberlain (NRC) to JV Parrish (Energy Northwest), "Columbia Generating Station Baseline Inspection Report 05000397/2005009"
Dear sir or Madam:
Pursuant to 10 CFR 50.90, Energy Northwest requests approval for a change incorporated into the current amendment 58 of the Columbia Generating Station (Columbia) Final Safety Analysis Report (FSAR). Energy Northwest initially made this change pursuant to a 10 CFR 50.59 evaluation. The referenced inspection report documents a differing opinion on the 10 CFR 50.59 evaluation conclusion and determined the change represented a departure from a method of evaluation described in the FSAR. Therefore, Energy Northwest is seeking NRC review and approval to change the method for calculating fuel pool decay heat load from the original licensing basis methodology of ORIGEN and the Auxiliary Systems Branch Technical Position (ASBTP) 9-2, "Residual Decay Heat Energy for Light Water Reactors for Long-Term Cooling," to ORIGEN-ARP.
Since the change made to the FSAR pursuant to 10 CFR 50.59 was characterized as a violation, Energy Northwest entered this condition into the corrective action program. In accordance with the process described in Part 9900 of the NRC Inspection Manual, "Operability Determinations & Functionality Assessments for Resolution of Degraded and Nonconforming Conditions Adverse to Quality or Safety," dated September 26, 2005, Snergy Northwest has characterized this FSAR change as a nonconformance.
. I LICENSE AMENDMENT REQUEST TO ADOPT ORIGEN-ARP FOR CALCULATING FUEL POOL DECAY HEAT LOAD Page 2 of 2 Energy Northwest is seeking to resolve this nonconformance by obtaining NRC approval of the proposed change in methodology. In the interim, this condition has no impact on plant safety. This nonconformance will have no impact on plant operations until the next refueling outage scheduled for May 2007, barring an unplanned condition that would require a core offload. Energy Northwest requests approval of this amendment by January 2007 to support the timely completion of the fuel pool heat load management plan for the Spring 2007 refueling outage.
Attachment 1 provides an evaluation with a detailed description of the background, proposed change, and technical analysis. The No Significant Hazards Consideration and Environmental Consideration are also contained in Attachment 1. A copy of the revised FSAR page (amendment 58) that replaced the reference to ASBTP 9-2 with a reference to ORIGEN-ARP is provided in Attachment 2 for information.
In accordance with 10 CFR 50.91, a copy of this application, with attachments, is being provided to the designated Washington State Official.
No commitments are made in the submittal. Should you have any questions or require additicnal information regarding this matter, please contact Mr. Michael K Brandon, at (509) :377-4758.
I declare under penalty of perjury that the foregoing is true and correct. Executed orn the date of this letter.
Respezctfully, WS Oxen0 r Vice Presidetnt, Technical Services Mail Drop PE04 Attachment 1: Evaluation of the Proposed Change Attachment 2: Copy of the Affected FSAR Page cc: BS Malleft - NRC RIV BJ Benney- NRC NRR NFRC Senior Resident Inspector/988C RN Sherman - BPA/1 399 WA Horin - Winston & Strawn JO Luce - EFSEC RFZ Cowley - WDOH
LICEN SE AMENDMENT REQUEST TO ADOPT ORIGEN-ARP FOR CALCULATING FUEL POOL DECAY HEAT LOAD Attachment 1 Page 1 of 7 Evaluation of the Proposed Change
1.0 DESCRIPTION
Energy Northwest is requesting NRC approval to replace the original license basis methodologies of ORIGEN and Auxiliary Systems Branch Technical Position (ASBTP) 9-2 with ORIGEN-ARP. This request will resolve the nonconformance associated with this methodology that currently exists in FSAR section 9.1.3, "Spent Fuel Pool Cooling and Cleanup System." ORIGEN-ARP is the current and only actively supported stand-alone version of the ORIGEN code. The ORIGEN-ARP methodology is based on the original version of ORIGEN and ORIGEN-ARP has been approved by the NRC for use in the evaluation of dry cask storage systems. The use of ORIGEN-ARP provides a more rigorous solution to the problem of determining spent fuel decay heat.
2.0 PROPOSED CHANGE
Prior to FSAR amendment 58 (current amendment number), the fuel pool heat load values presented in FSAR Table 9.1-4 were calculated using ORIGEN and based on estimated refueling data. As stated in (pre-amendment 58) FSAR Section 9.1.3.2.1, the fission product decay heat loads shown in FSAR Table 9.1-4 are "in agreement with ASBTP 9-2." The FSAR recognizes that refueling specific analyses may be required to validate the maximum fuel pool temperatures are within the licensing basis acceptance criteria. Since the original ORIGEN code used to calculate fuel pool heat loads is no longer supported and has been replaced by an improved ORIGEN-ARP version, Energy Northwest is requesting to replace the methodologies of ASBTP 9-2 and ORIGEN fcr calculating spent fuel pool decay heat loads with the improved ORIGEN-ARP code.
The use of ORIGEN-ARP will facilitate future compliance with the current FSAR spent fuel pool temperature limits by improving the accuracy and ease of performing future outage specific calculations.
3.0 BACKGROUND
Since both ASBTP 9-2 and ORIGEN were discussed in the Columbia FSAR with respect to calculating decay heat, a brief discussion of each methodology is provided below.
The methodology in ASBTP 9-2 is based on experimental data published from 1958 to 1973 relating to energy release from the decay of fission products. The methodology draws heavily on the proposed ANS standard, "Decay Energy Release Rates Following Shutdown of Uranium-Fueled Thermal Reactors," approved by Subcommittee ANS-5, October 1971 that presents a standard fission product decay heat curve versus cooling time. The actual equation resulting from the curve fit in ASBTP 9-2 is more complex than that in the proposed ANS standard. However, in general, the results of the curve-fit equations agree with each other reasonably well (within +/-5%). In addition, the
LICEN SE AMENDMENT REQUEST TO ADOPT ORIGEN-ARP FOR CALCULATING FUEL POOL DECAY HEAT LOAD Attachment 1 Page 2 of 7 terms for decay heat generation due to the heavy elements and the uncertainty factors in ASBTP 9-2 were taken directly from the above referenced ANS standard. The Foreword of the proposed standard, prepared by Subcommittee ANS-5, provides insight into the development of the methodology and states:
"It is the opinion of the generating group that the state of the knowledge of the relatively complex phenomena that contribute to the total energy release rate does not justify a detailed description of the process. From a very detailed description one might be able to determine the inventory of each radioisotope as a function of time and, knowing the decay constants and energy manifestation, calzulate the total energy deposition. Although such calculations have been made, it was the opinion of the generating group that there is not adequate knowledge of the many physical constants involved for use of this approach in a standard." (published in 1971)
Furthermore, text in ASBTP 9-2 acknowledges the lack of consistent experimental data and the differing results of various calculations but reasons that "the effect of all uncertainties can be treated ... by a suitably conservative multiplying factor." This factor is either 10% or 20% depending on the shutdown time of interest.
By contrast, the ORIGEN computer code models fissile material behavior during periods of irradiation and decay by computing time-dependent concentrations and source terms of a large number of isotopes that are simultaneously generated or depleted through neutronic transmutation, fission, radioactive decay, and physical or chemical removal rates. The primary advantage of ORIGEN over earlier burnup codes was its capability to treat the full isotopic transition matrix rather than a limited number of transmutation chains. The original ORIGEN program utilized a standard cross-section library designed for the analysis of standard light water (LWR) reactor fuel. The required data input for ORIGEN consists of data relating to the specific problem to be analyzed including input for fissile isotope concentrations (i.e., bundle enrichments and uranium weight:), bundle power during irradiation, irradiation length, and length of decay period.
Hence, ORIGEN provides a fairly rigorous treatment of the problem. This assessment is supported by the NRC in Information Notice 96-39 which states:
"ORIGEN does not use empirical methods to calculate decay heat but tracks the buildup and decay of the individual fission products within the reactor core during operation and shutdown. ORIGEN also includes the effect of element transmutation from neutron capture, both in fissile isotopes and fission products.
Because ORIGEN is a rigorous calculation of all decay heat inputs, it was used in the calculations for decay heat ..."
Essentially, all features of the original ORIGEN were retained, expanded or supplemented within new computations in the development of ORIGEN-ARP. The original ORIGEN code released in 1973 was intended principally for the generic
LICENSE AMENDMENT REQUEST TO ADOPT ORIGEN-ARP FOR CALCULATING FUEL POOL DECAY HEAT LOAD Attachment 1 Page a of 7 investigation of fuel cycle operations over a wide range of fuel design parameters.
Soon Users began to apply ORIGEN to problems that required greater precision and specificity than had been intended with the original data libraries (only one cross-section library for all LWR fuel with no burnup dependence). In 1980, ORIGEN2 was released providing a significant update to ORIGEN and the associated database. Like the original ORIGEN code, ORIGEN2 was designed to operate as a stand-alone calculational tool with fixed cross-section data libraries provided for several reactor models. ORIGEN2 is no longer a supported code. Rather ORIGEN-ARP has been developed as a stand-alone tool or as an interface with the SCALE code system to generate problem-specific neutronic data though interaction with other modules of SCALE. For Columbia, this neutronic data is in the form of a cross-section library specific to the bundle being evaluated.
The NRC has previously approved the use of ORIGEN2 for spent fuel pool applications.
In a Safety Evaluation (SE) dated September 21, 2001, the NRC issued an amendment to Duane Arnold for a revised thermal-hydraulic analysis of the spent fuel pool. In discussing the methodology for determining bundle decay heat, the Technical Evaluation Report referenced by the SE states:
"This program can perform decay heat calculations using either Branch Technical Position ASB 9-2 or the ORIGIN2 [sic] computer code.... The ORIGIN2 [sic]
option was used. All fuel assemblies were assumed to have been irradiated to the appropriate maximum burnup level. Based on this review, BNL [Brookhaven National Laboratory] concurs that the methodology and assumptions the licensee used to calculate the decay heat loads meet the intent of the applicable NRC guidelines."
Similarly, in an SE dated August 30, 2002, the NRC issued an amendment to V. C.
Summer for a spent fuel pool re-racking modification. The methodology used and accepted by the NRC for calculating spent fuel pool heat load in the V.C. Summer amendment was the ORIGEN2 code.
Based on the above examples, the ORIGEN2 methodology has been previously approved by the NRC for the calculation of fuel pool decay heat loads.
4.0 TECHNICAL ANALYSIS
The ORIGEN methodology, including the updated ORIGEN-ARP, provides a rigorous treatment of the physical phenomenon involved by computing time-dependent concentrations and source terms of a large number of isotopes, which are simultaneously generated or depleted through neutronic transmutation, fission, radioactive decay and physical or chemical removal rates.
LICENSE AMENDMENT REQUEST TO ADOPT ORIGEN-ARP FOR CALCULATING FUEL POOL DECAY HEAT LOAD Attachment 1 Page 4 of 7 The ORIGEN series of codes is widely used throughout the world for predicting the characteristics of spent reactor fuel. The design and licensing activities associated with storage of spent fuel is dependent on the analytic characteristics (e.g., decay heat, radiation sources). Spent fuel characterization for out-of-reactor applications is routinely obtained with point (i.e., no spatial dependence) depletion/decay methods that are able to capture an almost unlimited number of nuclide inventories and associated characteristic quantities in a flexible and useful format. Perhaps the most widely used point-depletion codes are the ORIGEN series of codes developed at Oak Ridge National Laboratory (ORNL). These codes were developed by ORNL for the NRC and are currently maintained/enhanced under co-sponsorship of NRC and Department of Energy (DOE). These codes are used to predict the characteristics of spent reactor fuel and high level waste for fuel cycle evaluation and out-of-reactor analysis applications.
Like the original ORIGEN computer code, ORIGEN2 and ORIGEN-ARP were designed to operate as a stand-alone calculational tool with fixed cross section data libraries provided for several reactor models.
ORIGEN-ARP uses two modules for calculating decay heat: (1) ARP (Automatic Rapid Process) for providing a problem-specific cross-section library file for input to ORIGEN-S, and (2) ORIGEN-S for performing the depletion and decay calculations described above.
The primary difference between ORIGEN2 and ORIGEN-ARP is the different methods used to develop and supply the burnup-dependent cross-section libraries for use by the ORIGEN code module.
The ORIGEN2 code uses cross-section and fission-product yield libraries that have been developed for a limited number of specific reactor models. For BWRs, only a single library is available for use, corresponding to a 2.75 wt% enriched bundle irradiated for 4 cycles to a burnup of 27.5 MWd/kgU.
ORIGEWN-S, on the other hand, is designed to interact with the other modules of the SCALE code system to produce neutronics data based on the actual assembly design and pertinent reactor history information. The ARP module provides ORIGEN-S with a cross-section library that is based on the actual bundle design and operating parameters such as assembly type (e.g., GE 8x8, Siemens 8x8), enrichment, burnup, and operating history.
At Columbia, the SCALE code system was used to generate cross-section libraries for the GE: 8x8, Siemens (SPC) 8x8, SPC 9x9, SVEA-96, and ATRIUM-1 0 BWR fuel assembly designs that have been irradiated in the Columbia reactor core. These cross-section libraries are employed in ORIGEN-ARP calculations to predict decay heat for the various BWR fuel assemblies as a function of decay time following irradiation. In a similar fashion, SCALE was used to generate cross-section libraries for the GE 7x7 lfuel assembly design that was irradiated in the Cooper reactor core. Measurements of
LICENSE AMENDMENT REQUEST TO ADOPT ORIGEN-ARP FOR CALCULATING FUEL POOL DECAY HEAT LOAD Attachment 1 Page r; of 7 decay heat were performed on several Cooper bundles as published in EPRI report NP-4619 and NUREG/CR-5625. ORIGEN-ARP was employed to calculate the decay heat values of these same assemblies and the results were compared to the measured values in order to establish the methodology bias and uncertainties to be applied to the calculated results for predicting decay heat.
Benchmarking of the ORIGEN2 code to ORIGEN-ARP for Columbia is documented in Westinghouse Report A-GEN-BWR-90. Table 6.8 of A-GEN-BWR-90 documents the!
results of the comparison of the three codes (ORIGEN2, SAS2H/ORIGEN-S, and ORIGEN-ARP) to the measured data. The bias for decay heat from each methodology (when compared to the measured data) is within 1.2% of each other, and the standard deviation is within 0.55%. The total error at the 95% confidence level is 15.83% for ORIGEN2 and 15.76% for ORIGEN-ARP. Based on the results of the benchmarking analysis, the rigorous procedure used to determine the methodology bias and uncertainties as compared to measured data, and the fact that ORIGEN-ARP was developed to provide greater flexibility and accuracy of results, the results of ORIGEN2 and ORIGEN-ARP are essentially the same.
One difference between the assumptions and inputs evaluated in the two referenced SEs for ORIGEN2 and the intended application at Columbia is the treatment of power measurement uncertainty. In the referenced ORIGEN2 applications, a 2% factor for power measurement uncertainty was included. This would normally be applied by multiplying the specific power by 1.02. The ORIGEN-ARP bias and uncertainty terms were derived by comparing code predictions to actual measured data. Thus, the uncertainty in power measurement is inherently included in the error terms.
To ensure that the error term adequately encompasses the required power uncertainty factor, comparisons were made between the results assuming only the 2% power uncertainty factor and the results assuming no power uncertainty but applying the error assoc ated with the best estimate analysis. The decay heat with purely the bias term applied (represents best estimate) was greater than the decay heat calculated with the 1.02 multiplier on specific power for all cases. Since the best estimate approach represents the minimum error that would be applied to the decay heat results, the requirement to include a 2% factor for power measurement uncertainty is indeed satisfied by using the methodology bias procedure defined in A-GEN-BWR-90 for any confidence level.
The SCALE code system, including ORIGEN-ARP, has been classified as Category B software and installed and tested under the Energy Northwest Software QA program.
Based on this discussion, ORIGEN-ARP provides an appropriate methodology for calculating fuel pool heat loads.
LICENSE AMENDMENT REQUEST TO ADOPT ORIGEN-ARP FOR CALCULATING FUEL POOL DECAY HEAT LOAD Page (i of 7 5.0 REGULATORY SAFETY ANALYSIS 5.1 No Significant Hazards Consideration Energy Northwest has evaluated whether or not a significant hazards consideration is involved with the proposed amendment by focusing on the three standards set forth in 10 CFR 50.92, "Issuance of Amendment" as discussed below.
- 1. Does the proposed change involve a significant increase in the probability or consequences of an accident previously evaluated?
Response: No The adoption of ORIGEN-ARP does not affect the probability or consequences of an accident previously evaluated. The calculation of the fuel pool heat load is used to evaluate and demonstrate the ability of the fuel pool cooling system to maintain the fuel pool temperatures within the acceptance limits specified in the Columbia FSAR. The proposed change to the methodology used to calculate the fuel pool heat load has no bearing on the probability or consequences of any previously evaluated accident. Therefore, this change does not involve a significant increase in the probability or consequences of an accident previously evaluated.
- 2. Does the proposed change create the possibility of a new or different kind of accident from any accident previously evaluated?
Response: No The change involves the use of a different methodology for calculating fuel pool decay heat load. This change does not involve any new equipment, it does not change any previously approved acceptance limits, and it does not affect or alter the operation of any equipment. Therefore this change does not create the possibility of a new or different kind of accident from any accident previously evaluated.
- 3. Does the proposed change involve a significant reduction in a margin of safety?
Response: No The margin of safety provided by the fuel pool cooling system is primarily defined by the difference between the maximum allowed fuel pool temperature and the boiling point of water. The margin of safety is supplemented by the ability to make up water to the spent fuel pool if boiling were to occur. The proposed change in methodology for calculating the fuel pool heat load does not alter the current temperature limits or acceptance criteria specified in the FSAR and has
LICENSE AMENDMENT REQUEST TO ADOPT ORIGEN-ARP FOR CALCULATING FUEL POOL DECAY HEAT LOAD Attachment I Page 7 of 7 no effect on the ability to provide make-up water if boiling were to occur. This change will allow Energy Northwest to more accurately calculate the fuel pool heat load to provide added confidence in the ability of the fuel pool cooling system to accommodate the heat load added to the spent fuel pool during refueling activities. Therefore, this change does not involve a significant reduction in a margin of safety.
Based on the above, Energy Northwest concludes that the proposed amendment does not warrant a significant hazard consideration under the standards set forth in 10 CFR 50.92(c) and accordingly, a finding of "no significant hazards consideration" is justified.
5.2 Applicable Regulatory Requirements/Criteria The proposed change does not impact the compliance with any regulatory requirements. The spent fuel pool cooling system design and acceptance limits specified in the FSAR are not changed. The change allows Energy Northwest to adopt a state of the art methodology for calculating the decay heat load in the fuel pool. The proposed methodology is a change from the methodology recommended in Revision 1 of NUREG 0800, "Standard Review Plan," Section 9.1.3, dated July 1, 1981; however, the basis for the acceptability of the proposed methodology is provided in this request. This change will improve the ease and accuracy of heat load calculations and will facilitate future compliance with the existing acceptance criteria with a greater degree of certainty.
In conclusion, 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 after the proposed change, (2) operation after the change 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.
6.0 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, or would change an inspection or surveillance requirement. However, the proposed amendment does not involve; (i) a significant hazards consideration; (ii) a significant change in the types or significant increase in the amounts of any effluent that may be released offsite; or, (iii) a significant increase in individual or cumulative occupational radiation exposure. Accordingly, the proposed amendment meets the eligibility criteria 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 needs be prepared in connection with the proposed amendment.
LICENSE AMENDMENT REQUEST TO ADOPT ORIGEN-ARP FOR CALCULATING FUEL POOL DECAY HEAT LOAD Copy of the Affected FSAR Page
COLUMBIA GENERATING STATION Amendment 58 FINAL SAFETY ANALYSIS REPORT December 2005 9.1.3.2.1 System Operation 9.1.3.2.1.1 FPC System Cooling Function The system normally cools the fuiel pool by transferring the spent fuel decay heat from the tuber side of the two fuel pool cooling heat exchangers to the RCC. Standby SW can be manually aligned as the alternate heat sink for the heat exchangers (see Section 9.1.3.2.3 for additional details). The fuel pool is maintained at or below 1250 F during normal plant operations. The fuel :ool temperature may rise above this value during refueling activities or during an anticipated operational transient of the loss of one train of the FPC system. The RHR system can also be manually aligned into several configurations to provide su lemental cooling of the fuel pool. One of these RHR configurations is the RHR/F ase mode o 5e-Section 9.1.3.2.2 for additional details).
The maximum heat load is present in the spent fue l refueling activities wn recently irradiated fuel bundles are discharged fr m the react or e to t ue e magnitude of this heat load is contingent upon t e cycle-specific refueling activities, i.e., the number of the bundles discharged, the burnup f the discharged bundles, and the decay time of each bundle when it is placed in the eiated with a planned disclarge can be calculated with ORIGEN-ARP computer cod ssuming a 2% thermal power uncertainty or other accept1o s and uncertainties.
During refueling activities, the fuel pool temperature is managed by controlling the number and schedule of fuel assemblies discharged, controlling the number of heat removal systems in service, and controlling the temperatures of the systems (RCC or SW) used to remove the heat from the FPC heat exchangers.
The fuel pool cooling system was originally designed to maintain the pool at a temperature of less than or equal to 125 0 F during refueling activities with both trains of fuel pool cooling in operation and RCC cooling water at 950 F. The decay heat load assumed for the original design basis (normal offload) was based on the original licensed power of 3323 MW-thermal, a one-year fuel cycle, and a quarter core offload with a 20-day decay period.
Since the original design, the licensed power was increased to 3486 MW-thermal and the operating cycle was revised from a one-year cycle to a two-year cycle. These changes resulted in an increased heat load in the fuel pool, particularly during refueling outages. For the current design (3486 MW-thermal and a two-year cycle), a 150'F fuel pool temperature limit (see Table 9.1-6) applies to normal refueling activities for the scenario of both trains of fuel pool sooling in operation.
The FPC system is also designed to provide sufficient cooling for an anticipated operational transient of the loss of one train of the FPC system. For this transient, the maximum bulk fuel pool water temperature is limited to 1550 F. This limit applies to both normal operation and normal refueling activities (i.e., excluding a full core offload).
__j LDCN4)5-005 9.1-23