Letter Sequence Supplement |
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Initiation
- Request, Request, Request, Request, Request, Request, Request, Request, Request, Request, Request, Request, Request, Request
- Acceptance...
- Supplement, Supplement
Results
Other: ML21161A299, ML21203A314, ML21211A082, ML21214A178, ML21260A161, ML22013A339, ML22019A279, ML22061A056, ML22109A175, ML22164A861, ML22167A170, ML23188A020, ML23193A938, ML23200A183, ML23243A910
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MONTHYEARML17310B2322017-11-0707 November 2017 Presentation Slides - NextEra Energy/Fpl - GSI-191 Issue Resolution, Pre-submittal Meeting, September 20, 2017 Project stage: Meeting ML17310B2062017-11-20020 November 2017 Summary of September 20, 2017, Meeting with Florida Power & Light Company and NextEra Energy Regarding Closure of NRC Generic Safety Issue 191/NRC Generic Letter 2004-02 Project stage: Meeting ML18136A9052018-05-31031 May 2018 Summary of April 25, 2017, Meeting with Florida Power & Light Company/Nextera Energy Regarding Planned Submittal of Exemption Requests to Support Closure of NRC Generic Safety Issue 191/NRC Generic Letter 2004-02 Project stage: Meeting L-20-162, Supplemental Response to NRC Generic Letter 2004-022020-11-30030 November 2020 Supplemental Response to NRC Generic Letter 2004-02 Project stage: Request 0CAN122001, Final Response to NRC Generic Letter 2004-022020-12-10010 December 2020 Final Response to NRC Generic Letter 2004-02 Project stage: Request ML21062A0642021-03-0202 March 2021 NRR E-mail Capture - Request for Additional Information for Diablo Canyon Generic Letter 2004-02 Submittal (L-2017-LRC-0000) Project stage: RAI NL-21-0020, Final Supplemental Response to NRC Generic Letter 2004-022021-03-23023 March 2021 Final Supplemental Response to NRC Generic Letter 2004-02 Project stage: Request DCL-21-034, Response to Request for Additional Information on Final Supplemental Response to Generic Letter 20042021-04-15015 April 2021 Response to Request for Additional Information on Final Supplemental Response to Generic Letter 2004 Project stage: Supplement ML21118A0072021-04-28028 April 2021 Final Response and Close-out to Generic Letter 2004-02 Project stage: Request PMNS20210610, Public Meeting Regarding Path Forward for Generic Letter 2004-02 Closure for Point Beach Nuclear Plant Units 1 and 22021-05-17017 May 2021 Public Meeting Regarding Path Forward for Generic Letter 2004-02 Closure for Point Beach Nuclear Plant Units 1 and 2 Project stage: Request ML21134A0232021-05-18018 May 2021 GL 2004-02 Resolution Update - NextEra Energy Point Beach, LLC (NextEra) Project stage: Request ML21147A1462021-05-27027 May 2021 Response to Request for Additional Information Regarding Generic Letter 2004-02 Project stage: Response to RAI ML21168A2612021-06-17017 June 2021 (Vcsns), Unit 1 - NRC Generic Letter 2004-02, Potential Impact of Debris Blockage on Emergency Recirculation During Design Basis Accidents at Pressurized-water Reactors - Final Supplemental Response Project stage: Request ML21161A2992021-06-17017 June 2021 Summary of May 18, 2021 Public Webinar with Nextera Energy Point Beach, LLC Regarding Path Forward for Generic Letter 2004-02 Closure Project stage: Other ML21197A0372021-07-16016 July 2021 NRR E-mail Capture - ANO-1 and 2 - Final RAI Final Response to GL 2004-02 Project stage: RAI ML21203A3142021-07-29029 July 2021 Closeout of Generic Letter 2004-02, Potential Impact of Debris Blockage on Emergency Recirculation During Design Basis Accidents at Pressurized-Water Reactors Project stage: Other ML21232A0402021-08-20020 August 2021 Notice of Teleconference with Entergy Operations, Inc. Concerning Final Response to Generic Letter 2004-02 at Arkansas Nuclear One, Units 1 and 2 Project stage: Meeting ML21252A3212021-08-20020 August 2021, 30 August 2021, 14 September 2021 ANO Meeting Summary for September 1, 2020 Public Meeting/Teleconference Project stage: Request ML21242A2792021-08-30030 August 2021 ANO Slides Presentation for 9-1-21 Public Meeting Project stage: Meeting ML21252A2662021-09-14014 September 2021 Summary of September 1, 2021, Teleconference Meeting with Entergy Operations, Inc. Concerning the Final Response to Generic Letter 2004-02 for Arkansas Nuclear One, Units 1 and 2 Project stage: Meeting ML21211A0822021-09-24024 September 2021 Closeout of Generic Letter 2004 02, Potential Impact of Debris Blockage on Emergency Recirculation During Design Basis Accidents at Pressurized-Water Reactors Project stage: Other ML21260A1612021-09-24024 September 2021 Closeout of Generic Letter 2004 02, Potential Impact of Debris Blockage on Emergency Recirculation During Design Basis Accidents at Pressurized-Water Reactors Project stage: Other RA-21-0230, Duke Energy - Final Supplemental Response to NRC Generic Letter 2004-02, Potential Impact of Debris Blockage on Emergency Recirculation During Design Basis Accidents at Pressurized-Water Reactors2021-09-30030 September 2021 Duke Energy - Final Supplemental Response to NRC Generic Letter 2004-02, Potential Impact of Debris Blockage on Emergency Recirculation During Design Basis Accidents at Pressurized-Water Reactors Project stage: Request 0CAN102101, Final Request for Additional Information Concerning Generic Letter 2004-022021-10-0404 October 2021 Final Request for Additional Information Concerning Generic Letter 2004-02 Project stage: Request ML21214A1782021-10-0808 October 2021 Closeout of Generic Letter 2004-02, Potential Impact of Debris Blockage on Emergency Recirculation During Design Basis Accidents at Pressurized Water Reactors Project stage: Other PMNS20211455, Pre-Submittal Public Meeting Regarding License Amendment and Exemption Request for Generic Letter 2004-02 Closure for Point Beach Nuclear Plant Units 1 and 22021-12-0303 December 2021 Pre-Submittal Public Meeting Regarding License Amendment and Exemption Request for Generic Letter 2004-02 Closure for Point Beach Nuclear Plant Units 1 and 2 Project stage: Meeting ML21336A7972021-12-0909 December 2021 GL 2004-02 Resolution Update - NextEra Energy Point Beach, LLC - December 9, 2021 (Slides) Project stage: Request ML22019A2792022-01-20020 January 2022 Summary of Public Webinar with NextEra Energy Point Beach, LLC Regarding Future License Amendment and Exemption Request for Generic Letter 2004-02 Closure for Point Beach Nuclear Plant Project stage: Other ML22019A2652022-01-27027 January 2022 Us NRC Staff Review of Documentation Provided by Firstenergy Nuclear Operating Co. for Beaver Valley, Units 1&2 Concerning Resolution of Generic Letter 2004-02 - Potential Impact of Debris Blockage on Emergency Recirculation During Design B Project stage: Approval ML22013A3722022-01-27027 January 2022, 31 January 2022 Closeout of Generic Letter 2004-02, Potential Impact of Debris Blockage on Emergency Recirculation During Design Basis Accidents at Pressurized-Water Reactors Project stage: Request ML22013A3392022-01-31031 January 2022 Closeout of Generic Letter 2004-02, Potential Impact of Debris Blockage on Emergency Recirculation During Design Basis Accidents at Pressurized-Water Reactors Project stage: Other ML22053A2402022-02-22022 February 2022 Final Supplemental Response to NRC Generic Letter 2004-02 Project stage: Request ML22061A0562022-03-29029 March 2022 Closeout of Generic Letter 2004-02, Potential Impact of Debris Blockage on Emergency Recirculation During Design Basis Accidents at Pressurized-Water Reactors Project stage: Other ML22112A1482022-04-22022 April 2022 Correction_H. B. Robinson Steam Electric Plant, Unit 2 - Request for Additional Information Regarding Supplemental Response to Generic Letter 2004-02 Project stage: RAI ML22109A1752022-04-27027 April 2022 Closeout of Generic Letter 2004-02, Potential Impact of Debris Blockage on Emergency Recirculation During Design Basis Accidents at Pressurized-Water Reactors Project stage: Other RA-22-0144, Response to NRC Request for Additional Information Regarding Supplemental Response to Generic Letter 2004-022022-05-19019 May 2022 Response to NRC Request for Additional Information Regarding Supplemental Response to Generic Letter 2004-02 Project stage: Supplement ML22164A8612022-06-24024 June 2022 Closeout of Generic Letter 2004-02, Potential Impact of Debris Blockage on Emergency Recirculation During Design Basis Accidents at Pressurized-Water Reactors Project stage: Other ML22167A1702022-07-14014 July 2022 Closeout of Generic Letter 2004 02, Potential Impact of Debris Blockage on Emergency Recirculation During Design Basis Accidents at Pressurized-Water Reactors Project stage: Other ML22242A0452022-08-23023 August 2022 NRR E-mail Capture - Dominion GL 04-02 Response Draft RAIs (L-2017-LRC-0000) Project stage: Draft RAI ML22251A1292022-09-0909 September 2022 Request for Additional Information Related to Response to Generic Letter 2004-04 Project stage: RAI ML22312A4432022-11-0707 November 2022 NRC Generic Letter 2004-02, Potential Impact of Debris Blockage on Emergency Recirculation During Design Basis Accidents at Pressurized-Water Reactors Fleet Response to RAI Project stage: Request ML22335A4142022-12-21021 December 2022 Request for Withholding Information from Public Disclosure for Dominion Fleet Response to Request for Additional Information Regarding NRC Generic Letter 2004-02 Project stage: RAI ML23193A9382023-07-18018 July 2023 Closeout of Generic Letter 2004-02, Potential Impact of Debris Blockage on Emergency Recirculation During Design Basis Accidents at Pressurized-Water Reactors Project stage: Other ML23188A0202023-07-26026 July 2023 Closeout of Generic Letter 2004-02, Potential Impact of Debris Blockage on Emergency Recirculation During Design Basis Accidents at Pressurized-Water Reactors Project stage: Other ML23200A1832023-08-0303 August 2023 Closeout of Generic Letter 2004-02, Potential Impact of Debris Blockage on Emergency Recirculation During Design Basis Accidents at Pressurized-Water Reactors Project stage: Other ML23243A9102023-09-0606 September 2023 Closeout of Generic Letter 2004-02, Potential Impact of Debris Blockage on Emergency Recirculation During Design Basis Accidents at Pressurized-Water Reactors Project stage: Other ML24012A0492024-01-11011 January 2024 Request for Additional Information Regarding Final Response to Generic Letter 2004-02 Project stage: RAI 2021-09-14
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Category:Letter type:DCL
MONTHYEARDCL-24-010, Nuclear Material Transaction Report for New Fuel2024-01-29029 January 2024 Nuclear Material Transaction Report for New Fuel DCL-24-009, Nuclear Material Transaction Report for New Fuel2024-01-17017 January 2024 Nuclear Material Transaction Report for New Fuel DCL-24-008, Schedule Considerations for Review of the DCPP License Renewal Application2024-01-17017 January 2024 Schedule Considerations for Review of the DCPP License Renewal Application DCL-24-004, Supplement to License Amendment Request 23-01 Revision to Technical Specifications to Adopt Risk-Informed Completion Times TSTF-505, Revision 2, Provide Risk-Informed Extended Completion Times - RITSTF Initiative 4b2024-01-15015 January 2024 Supplement to License Amendment Request 23-01 Revision to Technical Specifications to Adopt Risk-Informed Completion Times TSTF-505, Revision 2, Provide Risk-Informed Extended Completion Times - RITSTF Initiative 4b DCL-23-129, Nuclear Material Transaction Report for New Fuel2023-12-27027 December 2023 Nuclear Material Transaction Report for New Fuel DCL-23-122, Request for Exemption from Enhanced Weapons, Firearms Background Checks, and Security Event Notifications Implementation2023-12-14014 December 2023 Request for Exemption from Enhanced Weapons, Firearms Background Checks, and Security Event Notifications Implementation DCL-23-128, Emergency Plan Update2023-12-13013 December 2023 Emergency Plan Update DCL-23-125, Core Operating Limits Report for Unit 1 Cycle 252023-12-0606 December 2023 Core Operating Limits Report for Unit 1 Cycle 25 DCL-23-121, Supplement to License Amendment Request 23-03, Revision to Technical Specification3.7.8, Auxiliary Saltwater System2023-11-16016 November 2023 Supplement to License Amendment Request 23-03, Revision to Technical Specification3.7.8, Auxiliary Saltwater System DCL-23-120, License Amendment Request 23-03 Revision to Technical Specification 3.7.8, Auxiliary Saltwater (Asw) System2023-11-14014 November 2023 License Amendment Request 23-03 Revision to Technical Specification 3.7.8, Auxiliary Saltwater (Asw) System DCL-23-118, License Renewal Application2023-11-0707 November 2023 License Renewal Application DCL-2023-520, Discharge Self-Monitoring at Diablo Canyon Power Plant (DCPP)2023-10-19019 October 2023 Discharge Self-Monitoring at Diablo Canyon Power Plant (DCPP) DCL-23-103, Independent Spent Fuel Storage Installation - Withdrawal of License Amendment Request 22-01, Request for Approval of Alternative Security Measures for Early Warning System2023-10-13013 October 2023 Independent Spent Fuel Storage Installation - Withdrawal of License Amendment Request 22-01, Request for Approval of Alternative Security Measures for Early Warning System DCL-23-092, Material Status Report for the Period Ending August 31, 20232023-09-28028 September 2023 Material Status Report for the Period Ending August 31, 2023 DCL-23-077, License Amendment Request 23-02 Application to Adopt 10 CFR 50.69, Risk-Informed Categorization and Treatment of Structures, Systems and Components for Nuclear Power Reactors2023-09-27027 September 2023 License Amendment Request 23-02 Application to Adopt 10 CFR 50.69, Risk-Informed Categorization and Treatment of Structures, Systems and Components for Nuclear Power Reactors DCL-23-083, Upcoming Meeting with Nuclear Security and Incident Response Staff2023-09-13013 September 2023 Upcoming Meeting with Nuclear Security and Incident Response Staff DCL-23-078, Nuclear Material Transaction Report for New Fuel2023-09-0606 September 2023 Nuclear Material Transaction Report for New Fuel DCL-23-070, Withdrawal of Request Regarding Senior Reactor Operator License Application2023-08-16016 August 2023 Withdrawal of Request Regarding Senior Reactor Operator License Application DCL-23-068, Nuclear Material Transaction Report for New Fuel2023-08-0909 August 2023 Nuclear Material Transaction Report for New Fuel DCL-23-066, Nuclear Material Transaction Report for New Fuel2023-08-0303 August 2023 Nuclear Material Transaction Report for New Fuel DCL-23-065, Letter DCL-23-065 Request That Exam Material Be Withheld from Public Disclosure2023-07-27027 July 2023 Letter DCL-23-065 Request That Exam Material Be Withheld from Public Disclosure DCL-23-064, Nuclear Material Transaction Report for New Fuel2023-07-26026 July 2023 Nuclear Material Transaction Report for New Fuel DCL-23-054, License Amendment Request 23-01 Revision to Technical Specifications to Adopt Risk-Informed Completion Times TSTF-505, Revision 2, Provide Risk-Informed Extended Completion Times - RITSTF Initiative 4b2023-07-13013 July 2023 License Amendment Request 23-01 Revision to Technical Specifications to Adopt Risk-Informed Completion Times TSTF-505, Revision 2, Provide Risk-Informed Extended Completion Times - RITSTF Initiative 4b DCL-23-055, Nuclear Material Transaction Report for New Fuel2023-07-0606 July 2023 Nuclear Material Transaction Report for New Fuel DCL-23-042, Independent Spent Fuel Storage Installation, Submittal of Quality Assurance Program Description, Revision 02023-05-24024 May 2023 Independent Spent Fuel Storage Installation, Submittal of Quality Assurance Program Description, Revision 0 DCL-23-032, Independent Spent Fuel Storage Installation Docket No. 72-27, Materials License Number SNM-2514 Humboldt Bay Independent Spent Fuel Storage Installation - NRC Requested Notification of Withdrawal from Voluntary Program2023-05-22022 May 2023 Independent Spent Fuel Storage Installation Docket No. 72-27, Materials License Number SNM-2514 Humboldt Bay Independent Spent Fuel Storage Installation - NRC Requested Notification of Withdrawal from Voluntary Program DCL-23-041, 10 CFR 50.46 Annual Report of Emergency Core Cooling System Evaluation Model Changes for Peak Cladding Temperature for 20222023-05-22022 May 2023 10 CFR 50.46 Annual Report of Emergency Core Cooling System Evaluation Model Changes for Peak Cladding Temperature for 2022 DCL-23-038, Revision to the Unit 1 Reactor Vessel Material Surveillance Program Withdrawal Schedule2023-05-15015 May 2023 Revision to the Unit 1 Reactor Vessel Material Surveillance Program Withdrawal Schedule DCL-23-034, 2022 Annual Nonradiological Environmental Operating Report2023-05-0101 May 2023 2022 Annual Nonradiological Environmental Operating Report DCL-23-036, 2022 Annual Radiological Environmental Operating Report2023-05-0101 May 2023 2022 Annual Radiological Environmental Operating Report DCL-23-025, 2022 Annual Radioactive Effluent Release Report2023-05-0101 May 2023 2022 Annual Radioactive Effluent Release Report DCL-2023-512, Submittal of Receiving Water Monitoring Program 2022 Annual Report2023-04-27027 April 2023 Submittal of Receiving Water Monitoring Program 2022 Annual Report DCL-23-035, Response to Request for Additional Information on Request Regarding Senior Reactor Operator License Application2023-04-24024 April 2023 Response to Request for Additional Information on Request Regarding Senior Reactor Operator License Application DCL-23-028, Technical Specification Bases, Revision 142023-04-24024 April 2023 Technical Specification Bases, Revision 14 DCL-2023-511, Report on Discharge Monitoring for the First Quarter of 20232023-04-20020 April 2023 Report on Discharge Monitoring for the First Quarter of 2023 DCL-23-031, Submittal of Annual Report of Occupational Radiation Exposure for 20222023-04-19019 April 2023 Submittal of Annual Report of Occupational Radiation Exposure for 2022 DCL-23-024, Independent Spent Fuel Storage Installation, Response to Nrg Request for Additional Information for Revision to License Amendment Request 22-01 Request for Approval of Alternative Security Measures for Early Warning System2023-04-0404 April 2023 Independent Spent Fuel Storage Installation, Response to Nrg Request for Additional Information for Revision to License Amendment Request 22-01 Request for Approval of Alternative Security Measures for Early Warning System DCL-23-022, 2023 Annual Statement of Insurance for Pacific Gas and Electric Companys Diablo Canyon Power Plant2023-03-29029 March 2023 2023 Annual Statement of Insurance for Pacific Gas and Electric Companys Diablo Canyon Power Plant DCL-23-023, Decommissioning Funding Report for Diablo Canyon Power Plant, Units 1 and 22023-03-28028 March 2023 Decommissioning Funding Report for Diablo Canyon Power Plant, Units 1 and 2 DCL-23-021, Independent Spent Fuel Storage Installation, Emergency Plan Update2023-03-23023 March 2023 Independent Spent Fuel Storage Installation, Emergency Plan Update DCL-23-020, Responses to NRC Questions Regarding License Renewal Efforts2023-03-17017 March 2023 Responses to NRC Questions Regarding License Renewal Efforts DCL-23-009, Owner'S Activity Report for Unit 2 Twenty-Third Refueling Outage2023-02-22022 February 2023 Owner'S Activity Report for Unit 2 Twenty-Third Refueling Outage DCL-23-008, Request Regarding Senior Reactor Operator License Application2023-02-22022 February 2023 Request Regarding Senior Reactor Operator License Application DCL-2023-503, Annual Sea Turtle Report2023-01-26026 January 2023 Annual Sea Turtle Report DCL-2023-502, (DCPP) 4th Quarter 2022 Report on Discharge Self-Monitoring2023-01-18018 January 2023 (DCPP) 4th Quarter 2022 Report on Discharge Self-Monitoring DCL-22-091, Revised Steam Generator Tube Inspection Report for Twenty-First Refueling Outage2022-12-20020 December 2022 Revised Steam Generator Tube Inspection Report for Twenty-First Refueling Outage DCL-22-089, Core Operating Limits Report for Unit 2 Cycle 242022-12-20020 December 2022 Core Operating Limits Report for Unit 2 Cycle 24 DCL-22-085, Request to Resume Review of the License Renewal Application or, Alternatively, for an Exemption from 10 CFR 2.109(b), Concerning a Timely Renewal Application2022-10-31031 October 2022 Request to Resume Review of the License Renewal Application or, Alternatively, for an Exemption from 10 CFR 2.109(b), Concerning a Timely Renewal Application DCL-22-041, Site-Specific Decommissioning Cost Estimate, Revision 12022-10-12012 October 2022 Site-Specific Decommissioning Cost Estimate, Revision 1 DCL-22-042, Post-Shutdown Decommissioning Activities Report, Revision 12022-10-12012 October 2022 Post-Shutdown Decommissioning Activities Report, Revision 1 2024-01-29
[Table view] Category:Response to Request for Additional Information (RAI)
MONTHYEARDCL-23-035, Response to Request for Additional Information on Request Regarding Senior Reactor Operator License Application2023-04-24024 April 2023 Response to Request for Additional Information on Request Regarding Senior Reactor Operator License Application DCL-23-024, Independent Spent Fuel Storage Installation, Response to Nrg Request for Additional Information for Revision to License Amendment Request 22-01 Request for Approval of Alternative Security Measures for Early Warning System2023-04-0404 April 2023 Independent Spent Fuel Storage Installation, Response to Nrg Request for Additional Information for Revision to License Amendment Request 22-01 Request for Approval of Alternative Security Measures for Early Warning System DCL-23-020, Responses to NRC Questions Regarding License Renewal Efforts2023-03-17017 March 2023 Responses to NRC Questions Regarding License Renewal Efforts DCL-22-058, Responses to NRC Requests for Additional Information on LAR 20-03, Proposed Technical Specifications and Revised License Conditions for the Permanently Defueled Condition2022-07-20020 July 2022 Responses to NRC Requests for Additional Information on LAR 20-03, Proposed Technical Specifications and Revised License Conditions for the Permanently Defueled Condition ML22172A1642022-06-21021 June 2022 Response to Public Follow-up Question Regarding Unit 1 Reactor Vessel Neutron Embrittlement Coupon Testing - June 21, 2022 DCL-22-018, Independent Spent Fuel Storage Installation Responses to NRC Requests for Additional Information on License Amendment Request 21-04, Proposed Changes to the Emergency Plan for Post-Shutdown and Permanently.2022-03-23023 March 2022 Independent Spent Fuel Storage Installation Responses to NRC Requests for Additional Information on License Amendment Request 21-04, Proposed Changes to the Emergency Plan for Post-Shutdown and Permanently. DCL-21-060, Response to Request for Additional Information on License Amendment Request 21-03, Request for Revision to Technical Specification 3.8.1, 'Ac Sources - Operating,' to Support Diesel Fuel Oil Transfer System.2021-09-14014 September 2021 Response to Request for Additional Information on License Amendment Request 21-03, Request for Revision to Technical Specification 3.8.1, 'Ac Sources - Operating,' to Support Diesel Fuel Oil Transfer System. DCL-21-048, Response to Request for Additional Information on Emergency License Amendment Request 21-05, Revision to Technical Specification 3.7.8, 'Auxiliary Saltwater (Asw) System'2021-07-0707 July 2021 Response to Request for Additional Information on Emergency License Amendment Request 21-05, Revision to Technical Specification 3.7.8, 'Auxiliary Saltwater (Asw) System' DCL-21-040, Responses to NRC Requests for Additional Information on LAR 20-03, Proposed Technical Specifications and Revised License Conditions for the Permanently Defueled Condition2021-05-13013 May 2021 Responses to NRC Requests for Additional Information on LAR 20-03, Proposed Technical Specifications and Revised License Conditions for the Permanently Defueled Condition DCL-21-034, Response to Request for Additional Information on Final Supplemental Response to Generic Letter 20042021-04-15015 April 2021 Response to Request for Additional Information on Final Supplemental Response to Generic Letter 2004 DCL-20-088, Response to NRC Request for Additional Information Regarding Diablo Canyon Unit 2 Fall 2019 Steam Generator Tube Inspection Report2020-10-16016 October 2020 Response to NRC Request for Additional Information Regarding Diablo Canyon Unit 2 Fall 2019 Steam Generator Tube Inspection Report DCL-20-072, Response to Additional NRC Request for Additional Information Regarding License Amendment Request 20-01, Exigent Request for Revision to Technical Specification 3.7.5, 'Auxiliary Feedwater System'2020-08-20020 August 2020 Response to Additional NRC Request for Additional Information Regarding License Amendment Request 20-01, Exigent Request for Revision to Technical Specification 3.7.5, 'Auxiliary Feedwater System' DCL-20-069, Response to NRC Request for Additional Information Regarding License Amendment Request 20-01, Exigent Request for Revision to Technical Specification 3.7.5, 'Auxiliary Feedwater System'2020-08-18018 August 2020 Response to NRC Request for Additional Information Regarding License Amendment Request 20-01, Exigent Request for Revision to Technical Specification 3.7.5, 'Auxiliary Feedwater System' DCL-20-068, Response to NRC Request for Additional Information Regarding License Amendment Request 20-01, Exigent Request for Revision to Technical Specification 3.7.5, 'Auxiliary Feedwater System2020-08-16016 August 2020 Response to NRC Request for Additional Information Regarding License Amendment Request 20-01, Exigent Request for Revision to Technical Specification 3.7.5, 'Auxiliary Feedwater System DCL-20-012, Response to NRC Request for Additional Information Regarding Diablo Canyon Power Plant, Units 1 and 2 - Post-Shutdown Decommissioning Activities Report2020-03-11011 March 2020 Response to NRC Request for Additional Information Regarding Diablo Canyon Power Plant, Units 1 and 2 - Post-Shutdown Decommissioning Activities Report DCL-19-088, Supplemental Information Response to NRC Request for Additional Information Regarding 'License Amendment Request 19-01, Proposed Changes to the Intake Structure Physical Security Classification'2019-10-24024 October 2019 Supplemental Information Response to NRC Request for Additional Information Regarding 'License Amendment Request 19-01, Proposed Changes to the Intake Structure Physical Security Classification' DCL-19-056, Response to NRC Request for Additional Information (Supplemental) Regarding License Amendment Request 18-01, Request to Revise Emergency Plan Response Organization Staffing and Augmentation.2019-07-0303 July 2019 Response to NRC Request for Additional Information (Supplemental) Regarding License Amendment Request 18-01, Request to Revise Emergency Plan Response Organization Staffing and Augmentation. DCL-19-039, Response to NRC Request for Additional Information Regarding License Amendment Request 18-01, Request to Revise Emergency Plan Response Organization Staffing and Augmentation2019-05-0202 May 2019 Response to NRC Request for Additional Information Regarding License Amendment Request 18-01, Request to Revise Emergency Plan Response Organization Staffing and Augmentation ML19074A1092019-03-14014 March 2019 Response to Letter Dated February 26, 2019, Requesting NRC Opinion on Potential Impacts Due to Pacific Gas and Electric Company'S Bankruptcy Filing DCL-17-108, Spent Fuel Pool Evaluation Supplemental Report, Response to NRC Request for Information Pursuant to 10 CFR 50.54(f)Regarding Recommendation 2.1 of the Near-Term Task Force Review of Insights from the Fukushima Dai-ichi..2017-12-18018 December 2017 Spent Fuel Pool Evaluation Supplemental Report, Response to NRC Request for Information Pursuant to 10 CFR 50.54(f)Regarding Recommendation 2.1 of the Near-Term Task Force Review of Insights from the Fukushima Dai-ichi.. ML17324B3442017-11-20020 November 2017 Response to NRC Request for Additional Information Regarding Request for Approval of Alternative for Application of Full Structural Weld Overlay, REP-RHRSWOL, Units 1 and 2 DCL-17-073, Supplement to License Amendment Request 16-04 Revision 6. Development of Emergency Action Levels for Non-Passive Reactors.2017-08-17017 August 2017 Supplement to License Amendment Request 16-04 Revision 6. Development of Emergency Action Levels for Non-Passive Reactors. DCL-17-060, Response to NRC Request for Additional Information Regarding Relief Requests NDE-SLH U2, NDE-LSL U2, NDE-LHC U2, NDE-LHM U2, and NDE-ONV U22017-06-21021 June 2017 Response to NRC Request for Additional Information Regarding Relief Requests NDE-SLH U2, NDE-LSL U2, NDE-LHC U2, NDE-LHM U2, and NDE-ONV U2 ML17179A0192017-06-21021 June 2017 PG&E Response to NRC Request for Additional Information Regarding License Amendment Request 16-04, Request to Adopt Emergency Action Level Schemes Pursuant to NEI 99-01, Revision 6, 'Development of Emergency Action Levels for Non-Passive .. DCL-16-114, Response to Generic Letter 2016-012016-11-0303 November 2016 Response to Generic Letter 2016-01 DCL-16-101, Diablo Canyon, Units 1 and 2, Response to NRC Request for Additional Information Regarding License Amendment Request 15-03, Application of Alternative Source Term2016-10-0606 October 2016 Diablo Canyon, Units 1 and 2, Response to NRC Request for Additional Information Regarding License Amendment Request 15-03, Application of Alternative Source Term DCL-16-101, Response to NRC Request for Additional Information Regarding License Amendment Request 15-03, Application of Alternative Source Term2016-10-0606 October 2016 Response to NRC Request for Additional Information Regarding License Amendment Request 15-03, Application of Alternative Source Term ML16287A7552016-10-0606 October 2016 Technical Assessment Implementation of Alternative Source Terms Summary of Dose Analyses and Results Revision 4 DCL-16-094, Response to NRC Request for Additional Information Regarding License Amendment Request 16-02, License Amendment Request to Revise Technical Specification 3.4.12, 'Low Temperature Overpressure Protection (LTOP) System'.2016-09-28028 September 2016 Response to NRC Request for Additional Information Regarding License Amendment Request 16-02, License Amendment Request to Revise Technical Specification 3.4.12, 'Low Temperature Overpressure Protection (LTOP) System'. DCL-16-062, Response to RAI Regarding License Amendment Request 13-02, Revision to Technical Specifications to Adopt Risk Informed Completion Times TSTF-505, Revision 1, 'Provide Risk-Informed Extended Completion Times - Ritstf..2016-07-0707 July 2016 Response to RAI Regarding License Amendment Request 13-02, Revision to Technical Specifications to Adopt Risk Informed Completion Times TSTF-505, Revision 1, 'Provide Risk-Informed Extended Completion Times - Ritstf.. DCL-16-044, Response to NRC Request for Additional Information Regarding License Amendment Request 15-03, Application of Alternative Source Term.2016-04-21021 April 2016 Response to NRC Request for Additional Information Regarding License Amendment Request 15-03, Application of Alternative Source Term. DCL-16-023, Response to NRC Letter Dated February 2, 2016, Requests for Additional Information for the Review of the License Renewal Application - Set 39.2016-02-25025 February 2016 Response to NRC Letter Dated February 2, 2016, Requests for Additional Information for the Review of the License Renewal Application - Set 39. DCL-16-019, Response to NRC Request for Additional Information Regarding License Amendment Request 15-03, Application of Alternative Source Term.2016-02-10010 February 2016 Response to NRC Request for Additional Information Regarding License Amendment Request 15-03, Application of Alternative Source Term. ML16039A3662016-02-0808 February 2016 Units and 2 - Updated Response to Request for Information Pursuant to 10 CFR 50.54(f) Regarding Recommendation 2.1 Flooding. Part 1 of 2 DCL-16-016, Units and 2 - Updated Response to Request for Information Pursuant to 10 CFR 50.54(f) Regarding Recommendation 2.1 Flooding. Part 2 of 22016-02-0808 February 2016 Units and 2 - Updated Response to Request for Information Pursuant to 10 CFR 50.54(f) Regarding Recommendation 2.1 Flooding. Part 2 of 2 DCL-16-015, Response to NRC Request for Additional Information Regarding License Amendment Request 15-03, Application of Alternative Source Term.2016-02-0101 February 2016 Response to NRC Request for Additional Information Regarding License Amendment Request 15-03, Application of Alternative Source Term. DCL-16-013, Response to NRC Request for Additional Information Regarding Relief Request NDE-FWNS-U1/U22016-01-27027 January 2016 Response to NRC Request for Additional Information Regarding Relief Request NDE-FWNS-U1/U2 DCL-16-012, Clarifications to Environmental Request for Additional Information Responses for the Review of the License Renewal Application2016-01-26026 January 2016 Clarifications to Environmental Request for Additional Information Responses for the Review of the License Renewal Application ML16049A0112016-01-25025 January 2016 Attachment 2 - Pacific Gas and Electric Company Diablo Canyon Power Plant Units 1 and 2, Process Protection System (PPS) Replacement Interface Requirements Specification Nuclear Safety-Related, Rev 9 DCL-16-011, Attachment 2 - Pacific Gas and Electric Company Diablo Canyon Power Plant Units 1 and 2, Process Protection System (PPS) Replacement Interface Requirements Specification Nuclear Safety-Related, Rev 92016-01-25025 January 2016 Attachment 2 - Pacific Gas and Electric Company Diablo Canyon Power Plant Units 1 and 2, Process Protection System (PPS) Replacement Interface Requirements Specification Nuclear Safety-Related, Rev 9 ML16049A0092016-01-25025 January 2016 Pacific Gas & Electric Company Diablo Canyon Units 1 and 2 - Response to Request for Additional Information and Submittal of Advanced Logic System Phase 2 Documents for the License Amendment Request for Process Protection System Replacement DCL-16-011, Pacific Gas & Electric Company Diablo Canyon Units 1 and 2 - Response to Request for Additional Information and Submittal of Advanced Logic System Phase 2 Documents for the License Amendment Request for Process Protection System Replaceme2016-01-25025 January 2016 Pacific Gas & Electric Company Diablo Canyon Units 1 and 2 - Response to Request for Additional Information and Submittal of Advanced Logic System Phase 2 Documents for the License Amendment Request for Process Protection System Replacement DCL-15-156, Response to NRC Request for Additional Information - National Fire Protection Association Standard 805 and Supplement2015-12-31031 December 2015 Response to NRC Request for Additional Information - National Fire Protection Association Standard 805 and Supplement ML15355A5502015-12-21021 December 2015 Transmittal of Response to NRC Request for Additional Information Dated October 1, 2015, and November 13, 2015, Regarding Recommendation 2.1 of the Near-Term Task Force Seismic Hazard and Screening Report DCL-15-154, Response to NRC Request for Additional Information Dated October 1, 2015, and November 13, 2015, Regarding Recommendation 2.1 of the Near-Term Task Force Seismic Hazard and Screening Report2015-12-21021 December 2015 Response to NRC Request for Additional Information Dated October 1, 2015, and November 13, 2015, Regarding Recommendation 2.1 of the Near-Term Task Force Seismic Hazard and Screening Report ML16004A3612015-12-17017 December 2015 Response to NRC Request for Additional Information Regarding License Amendment Request 15-03 - Updated Final Safety Analysis Report Markup, Revision 1. Part 1 of 3 ML16004A3552015-12-17017 December 2015 Response to NRC Request for Additional Information Regarding License Amendment Request 15-03, Regulatory Guide 1.183 Conformance Tables, Revision 1 ML16004A3562015-12-17017 December 2015 Response to NRC Request for Additional Information Regarding License Amendment Request 15-03, Application of Alternative Source Term. ML16004A3572015-12-17017 December 2015 Response to NRC Request for Additional Information Regarding License Amendment Request 15-03 - Updated Final Safety Analysis Report Markup, Revision 1. Part 2 of 3 DCL-15-152, Response to NRC Request for Additional Information Regarding License Amendment Request 15-03 - Updated Final Safety Analysis Report Markup, Revision 1. Part 3 of 32015-12-17017 December 2015 Response to NRC Request for Additional Information Regarding License Amendment Request 15-03 - Updated Final Safety Analysis Report Markup, Revision 1. Part 3 of 3 2023-04-04
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mPacHic Gas and Electric Company*
Paula Gerfen Site Vice President Diablo Canyon Power Plant Mail code 104/6/605 P.O. Box 56 Avila Beach, CA 93424 805.545.4596 Internal: 691.4596 Fax: 805.545.4234 April 15, 2021 PG&E Letter DCL-21-034 U.S. Nuclear Regulatory Commission 10 CFR 50.54(f)
ATTN: Document Control Desk Washington, DC 20555-0001 Docket No. 50-275, OL-DPR-80 Docket No. 50-323, OL-DPR-82 Diablo Canyon Units 1 and 2 Response to Request for Additional Information on Final Supplemental Response to Generic Letter 2004-02
Reference:
- 1. PG&E Letter DCL-20-031, Final Supplemental Response to Generic Letter 2004-02, dated April 30, 2020 [ML20121A095]
- 2. NRC email, Request for additional information for Diablo Canyon Generic Letter 2004-02 Submittal (L-2017-LRC-0000), dated March 2, 2021
Dear Commissioners and Staff:
In Reference 1, Pacific Gas and Electric Company (PG&E) submitted the final supplemental response for Diablo Canyon Units 1 and 2 to Generic Letter 2004-02, dated September 13, 2004, Potential Impact of Debris Blockage on Emergency Recirculation Design Basis Accidents at Pressurized-Water Reactors. In Reference 2, the NRC Staff provided a request for additional information (RAI) via an e-mail, dated March 2, 2021. The Enclosure to this letter provides PG&E responses to the RAI.
This letter does not include any new or revised regulatory commitment (as defined by NEI 99-04).
If you have any questions or require additional information, please contact Mr. James Morris, Regulatory Services Manager, at (805) 545-4609.
I state under penalty of perjury that the foregoing is true and correct.
Executed on April 15, 2021.
A member of the STARS Alliance Callaway
Document Control Desk PG&E Letter DCL-21-034 April 15, 2021 Page 2 Sincerely,
~~7)
Paula Gerfen Site Vice President kjse/51105667 Enclosure cc: Diablo Distribution cc/enc: Donald R. Krause, NRC Senior Resident Inspector Samson S. Lee, NRR Senior Project Manager Scott A. Morris, NRC Region IV Administrator Gonzalo L. Perez, Branch Chief, California Department of Public Health A member of the STARS Alliance Callaway
PG&E Letter DCL-21-034 Enclosure Response to Request for Additional Information on Final Supplemental Response to Generic Letter 2004-02 NRC Question 1 By letter dated April 30, 2020 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML20121A095), Pacific Gas and Electric Company (PG&E or the licensee) submitted a final response to close Generic Letter (GL) 2004-02, dated September 13, 2004 (ADAMS Accession No. ML042360586),
Potential Impact of Debris Blockage on Emergency Recirculation During Design Basis Accidents at Pressurized-Water Reactors, for the Diablo Canyon Power Plant, Units 1 and 2. 10 CFR 50.46 requires that plants are able to maintain adequate long-term core cooling (LTCC) to ensure that the fuel in the core can be cooled and maintained in a safe and stable configuration following a postulated accident. GL 2004-02 requested that licensees provide information confirming that their plants are in compliance with the regulation. During its review of the licensees submittal, the NRC staff identified that it required additional information to confirm the licensees evaluation.
Please provide the following information:
Table 3.b.1-1 of the April 30, 2020 submittal states that WCAP-17561 was used to determine the zone of influence (ZOI) for Temp-Mat. The ZOI credited is 3.7D. The NRC staff reviewed WCAP-17561 and found that the methods and results in the WCAP were generally acceptable. However, the NRC staff determined that some of the test geometries for the Temp-Mat tests may not have been representative or conservative with respect to those installed in plants. The text in Section 3.b.1 states that the amount of debris included in the testing exceeded the debris quantity that would result if a ZOI of 11.7D (NRC generically approved ZOI for Temp-Mat) were applied. As used in the current debris generation and head loss analysis, this is a conservative assumption and is acceptable. This issue regarding the test geometry for the WCAP-17561 testing is not relevant to Diablo Canyons current submittal but could be relevant to future modifications or operability determinations.
If future actions assume the reduced ZOI based on WCAP-17561, and the use of this ZOI is not appropriate for the plant specific geometry, the amount of fibrous debris generated could exceed that included in the plant specific testing. Please justify that the WCAP-17561 ZOI is representative of the plant geometry or explain how the potential discrepancy will be managed.
PG&E Response As noted in the question, applicability of the test geometry for the WCAP-17561 testing to the current Diablo Canyon Power Plant (DCPP) configuration is not in question. To ensure that WCAP-17561 will continue to be representative to potential 1
PG&E Letter DCL-21-034 Enclosure future design configurations at DCPP, two scenarios are considered: insulation changes due to planned modifications and unexpected discovery of new debris sources that have not been included in previous analyses. The potential for each of these scenarios to impact the applicability of WCAP-17561 to DCPP is discussed below.
Insulation changes due to planned modifications are evaluated according to the DCPP design change procedure. This procedure discourages the use of fibrous thermal insulation inside containment and requires the review of the debris generation and transport calculations by design engineering personnel if the use of fibrous insulation is desired. Additionally, DCPP controls the fabrication and installation requirements for encapsulated Temp-Mat insulation that complies with the WCAP-17561-P tested configuration. This includes the encapsulation foil material, thickness, stitching composition, seam orientation, and seam coverage.
Discovery of new Temp-Mat debris sources in quantities significant enough to challenge existing margins is considered unlikely given the insulation reduction and controls implemented over the past years, as described in Section 2 in PG&E's Letter DCL-20-031 providing the final response to close GL 2004-02
[ML20121A095]. Any future discovery of new Temp-Mat debris would constitute a non-conforming condition which is within the scope of the DCPP operability determination procedure. It is expected that, for an operability determination, any newly-discovered Temp-Mat condition would follow the procedural guidance and be initially evaluated by assuming that its full quantity would count against the existing containment debris margins. Should a more detailed analysis be required, the existing containment debris margin calculation provides a roadmap back to the debris generation calculation, where the 3.7D ZOI is referred to the WCAP-17561-P testing configuration (Temp-Mat encapsulated in stainless steel) and results. The calculation also states that non-encapsulated Temp-Mat has a ZOI of 11.7D per the NRC safety evaluation of NEI 04-07, Volume 2, Revision 0, Pressurized Water Reactor Sump Performance Evaluation Methodology - Safety Evaluation by the Office of Nuclear Reactor Regulation Related to NRC Generic Letter 2004-02
[ML050550156].
NRC Question 2 Provide an overview of the analysis method for the Temp-Mat ZOI length for the hot and cold-leg nozzle breaks. Also, describe how the break geometries and sizes were determined and how these geometries relate to the assigned ZOI volumes.
PG&E Response Due to the presence of pipe whip restraints, the maximum allowable axial and radial separations are less than the pipe wall thicknesses for both the hot and cold legs.
The analysis of Temp-Mat ZOI for the reactor nozzle breaks used the American National Standards Institute/American Nuclear Society (ANSI/ANS) standard 2
PG&E Letter DCL-21-034 Enclosure ANSI/ANS-58.2-1988 laterally restrained double-ended guillotine break (DEGB) jet model. This approach is reasonable because the shape of the jet resulting from this configuration will not be significantly affected by the slight radial offset. The flow from one end of the ruptured pipe will not escape in the axial direction past the outside surface of the opposite pipe, and the jet will continue to form radially. The jet configuration is illustrated in Figure 2-1.
Figure 2-1: Illustration of a laterally restrained DEGB jet The Temp-Mat ZOI length was determined according to the ANSI 58.2-1988 standard, as well as Appendix I in the NRC safety evaluation of NEI 04-07, Volume 2, Revision 0, where appropriate since the NRC safety evaluation focused on the evaluation of a fully separated break.
The first step of the analysis is to determine the mass flux from the break, which depends on the temperature and pressure conditions upstream of the break. As recommended in the NRC safety evaluation for subcooled conditions, the Henry-Fauske model was used for the hot and cold leg nozzle breaks. The initial condition necessary for the calculation of mass flux requires the determination of thermodynamic parameters for different regions within the break jet. ANSI 58.2-1988 defines three jet regions that characterize an expanding jet. Figure 2-2 illustrates the break jet regions for a partially separated circumferential break.
Region 1 is the jet core which is characterized by the upstream stagnation pressure and temperature, and the length of the core defines the boundary between Region 1 and Region 2. Region 2 occurs between the jet core and the asymptotic plane. It was assumed that the jet can only interact with the environment after crossing the asymptotic plane to maintain the maximum jet force. Region 3 occurs after the asymptotic plane and is the low pressure region of the jet.
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PG&E Letter DCL-21-034 Enclosure Asymptotic Plane Region 2 L
Region 1 Break Flow Plane - --\.- - --r..._ _.,
(top and bottom)
Figure 2-2: Partially separated circumferential break The next step in the analysis is to evaluate the lengths and diameters of the jet at different intervals using Appendix D of standard ANSI 58.2-1988, which determines the jet pressure as a function of radius from the centerline of the pipe and the distance from the break plane. This allows for calculation of the pressure isobars for contour mapping of a jet emission from the pipe break including pressure of the core and localized pressure inside the expanding jet. The conversion of the jet pressure to isobars was performed using the methodology in Appendix I of the NRC safety evaluation of NEI 04-07, Volume 2, Revision 0. The pressure isobars for the hot leg break are illustrated in Figure 2-3. The ZOI for the reactor nozzle breaks with restricted separation results from the revolution of the isobars around the circumference of the pipe, which resembles the ZOI shape shown in Figure 2-1.
The final step of the analysis is to determine the points at which the isobar corresponding to the destruction pressure of Temp-Mat insulation (10.2 pounds per square inch gage) crosses the centerline of the jet pressure volume.
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PG&E Letter DCL-21-034 Enclosure Figure 2-3: Hot leg break pressure isobars This point represents the farthest radial reach of the ZOI from the reactor nozzle break location. The ZOI lengths for hot leg and cold leg breaks were overlayed onto a model of the reactor coolant system (RCS) piping to demonstrate that the ZOI for any RCS nozzle break would not reach the primary shield penetrations of the adjacent legs, as shown in Figure 2-4. Therefore, the debris generation analysis assumed that 100 percent of the Temp-Mat in the primary shield penetration of the broken leg becomes fines while the Temp-Mat in the adjacent legs become intact pieces. Equivalent spherical ZOIs were calculated based on the volume within the pressure isobar. However, the spherical ZOIs were not used in the debris generation analysis since the radius of an equivalent volume sphere is smaller than the calculated jet length.
Figure 2-4: Hot leg and cold leg nozzle break ZOI lengths 5
PG&E Letter DCL-21-034 Enclosure NRC Question 3 Please provide additional details regarding the submergence of the strainer for the Small Break Loss-of-Coolant Accident (SBLOCA) scenario. Section 3.f.11 of the submittal states that the submerged height of the strainer is 1.01 ft. for the front and 1.79 ft. for the rear sections. It also states that the headloss is 0.758 ft. This is less than half of the rear section submergence, but greater than half of the front section submergence. Considering Regulatory Guide (RG) 1.82 guidance, please explain why this condition is acceptable? Also, please provide details of the timing of additional submergence for this case. For example, how long does it take the height of the pool to increase such that the strainer is fully submerged? If this occurs relatively quickly, it may be demonstrated that headloss will not increase quickly enough to cause partially submerged strainer failure.
PG&E Response Section 3.f.11 of PG&E's Letter DCL-20-031 providing the final response to close GL 2004-02 (hereafter referred to as the submittal) states that half of the strainer submerged depth is 1.01 ft for the front and 1.79 ft for the rear sections. These values have already been halved from the full strainer submergence depths at those locations. Therefore, the head loss is required by RG 1.82 to be less than 1.01 ft.
The maximum strainer head loss is 0.758 ft, which meets the requirement of RG 1.82 independent of the timing of the submergence for the evaluated limiting SBLOCA scenario.
NRC Question 4 Related to question 3 above, Section 3.g.5 of the submittal states that the Containment Spray (CS) pumps may not be actuated. Table 3.g.12-1 implies that the volume injected from the Refueling Water Storage Tank (RWST) for the SBLOCA assumes CS flow. Please describe how the status of CS affects strainer submergence for the SBLOCA case. The submittal demonstrates that there is adequate margin to account for the sump level reduction that may occur due to reduced injection from the RWST. Please confirm that future changes to net positive suction head (NPSH) calculations will account for the actual minimum sump level that may occur. The potential for a reduction in sump level of 1.3 ft. may be evaluated against increased NPSH available from other sources. For example, if CS does not actuate, there is additional inventory available because CS piping is not filled. In addition, debris headloss is very low at start of recirculation. Other aspects of the scenario may be considered.
PG&E Response DCPP analyzed the minimum containment water level using the injection volume associated with the RWST low level trip setpoint and recirculation switchover timing.
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PG&E Letter DCL-21-034 Enclosure The additional contribution from CS for those cases when it is actuated is the result of its operation during the time between RWST low level trip and the beginning of recirculation, for which a conservative time of three minutes was assumed. As a result, the injection volume is not significantly affected by CS actuation.
Minimum containment water levels for break sizes between 1.5 inches and 6 inches were analyzed both with and without operation of CS under minimum and maximum safeguard conditions. Break sizes below 3 inches do not automatically actuate CS; however, due to conservative inputs affecting the selected water level case (described below), the minimum water level of 93.17 ft presented in Section 3.g.1 of the submittal results from a 4-inch break with CS and minimum safeguards.
In Table 4-1, the credited volumes from Table 3.g.12-1 of the submittal are shown alongside the volumes from another case evaluated in the water level calculation.
The additional case is for a 2-inch break with minimum safeguards but no CS. For the 4-inch break case, a conservatively small volume from the accumulators was used due to lack of break-specific accumulator injection time curve. This led to a water level lower than that determined for smaller breaks without CS. Therefore, the case resulting in the minimum water level is bounding for all other breaks of 6 inches and smaller, including those which do not actuate CS.
In summary, the minimum sump pool water level shown in Section 3.g.1 of the submittal for the breaks of 6 inches and smaller was 93.17 ft, resulting from a 4-inch break with CS and minimum safeguards flow. Table 3.g.12-1 in the submittal shows the source water volumes used for the analysis of this 4-inch break (see Table 4-1 below). The 4-inch break minimum water level bounds (i.e., is smaller than) the cases for breaks down to 1.5 inches without CS. Therefore, it is conservative to use this minimum water level in Section 3.f.11 to evaluate the strainer head loss and NPSH for SBLOCAs against the RG 1.82 acceptance criterion for a partially submerged strainer. The current NPSH calculation already uses a sump minimum level value of 93.17 feet, the lowest level for breaks less than six inches, and therefore a future revision to the NPSH calculation to account for the actual minimum sump level that may occur is not required.
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PG&E Letter DCL-21-034 Enclosure Table 4-1: Source water volumes and resulting water level 4-inch break 2-inch break without Source with containment spray containment spray RWST (gallons) 283,388 274,850 Accumulators 6,927 24,358 (gallons)
RCS (gallons) 2,164 825 Spray Additive 2,279 0 Tank (gallons)
Water level (feet) 93.17 93.35 NRC Question 5 In Section 3.f.14 of the submittal, it was determined that the maximum amount of entrained gases that can reach the pump suction is 0.17 percent. Was the NPSH required value used in the NPSH margin calculation adjusted per the guidance in Regulatory Guide 1.82, Appendix A-3, to account for the entrained gases? If not, please describe how the effect of entrained gases on pump performance was evaluated.
PG&E Response The NPSH required value used in the NPSH margin calculation was not adjusted per the guidance in RG 1.82, Appendix A-3 As stated in Section 3.f.14 of the submittal, the void fraction is 0.17 percent at the pump suction, which is much lower than the 2 percent limit from NEI 09-10, Guidelines for Effective Prevention and Management of System Gas, Revision 1, December 2010, to prevent mechanical damage and significant impact on the pump head. Various conservativisms were built into the void fraction analysis, as summarized below.
Void fraction immediately downstream of the strainer was maximized by:
- Using the maximum strainer head loss at 60°F.
- Using the smaller minimum strainer submergence between the front and rear sections of the strainer.
The 0.17 percent void fraction was based on the assumption that the voids formed at the strainer will transport intact to the pump suction. When crediting compression of the voids at the pump suction using the ideal gas law, the pressure ratio between the strainer and pump suction was maximized by:
- Neglecting strainer head loss when calculating the strainer pressure but including the maximum strainer head loss when calculating the pump suction pressure.
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PG&E Letter DCL-21-034 Enclosure
- Using the greater strainer submergence between the front and rear strainer disks for the strainer pressure term When the voids formed at the strainer are transported to the pump, the increased elevation head generated in moving the fluid down to the pump suction overcomes the head loss associated with the piping, strainer, and debris bed, resulting in a net pressure increase. For DCPP, the increase in pressure is over 10 pounds per square inch. Although not credited in the DCPP analysis, the increasing pressure tends to compress and collapse the bubbles formed at the strainer as they transport to the pump suction. This is similar to cavitation, where bubbles form when water near its saturation point experiences a rapid pressure drop and collapse as the pressure recovers.
In summary, PG&E calculated the void fraction using a conservative method and showed that the resulting void fraction is well below the NEI 09-10 acceptance criterion. Although not credited in the analysis, the voids formed at the strainer are expected to collapse as they transport to the pump suction and experience higher pressures. As a result, the voids will not degrade pump performance.
NRC Question 6 For the in-vessel evaluation, please provide the WCAP-17788 chemical effects test group number that was applied to the Diablo Canyon in-vessel analysis and confirm it is representative of projected post-LOCA plant conditions.
PG&E Response WCAP-17788 autoclave Test Group 45, including Test 45-01 and Test IBOB 45-01, is applied as representative of the DCPP post-LOCA plant conditions for the in-vessel analysis. This test group demonstrates that the minimum chemical precipitation time is greater than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> as stated in Table 3.n.1-3 of the submittal.
Table 6-1 shows the critical projected post-LOCA conditions and debris loads at plant scale.
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PG&E Letter DCL-21-034 Enclosure Table 6-1: Critical Projected Diablo Canyon Post-LOCA Plant Conditions (plant scale)
Diablo Canyon Unit 1 Diablo Canyon Unit 2 Parameter (Plant Scale) (Plant Scale)
Buffer Sodium Hydroxide Sodium Hydroxide Sump pH (Long-term) 8.0 - 9.5 8.0 - 9.5 Minimum Sump Volume 68,925 ft3 68,925 ft3 Maximum Sump Pool 261°F 261°F Temperature Maximum Calcium Silicate 34,800 g 52,300 g Maximum E-Glass 119,100 g 72,500 g Maximum Silica 0g 0g Maximum Mineral Wool 0g 0g Maximum Aluminum Silicate 30,800 g 37,900 g Maximum Concrete Not Determined Not Determined Maximum Interam' 0g 0g Aluminum 4,039 ft2 4,039 ft2 Galvanized Steel Not Determined Not Determined Table 6-2 shows the above parameters scaled by volume to the WCAP-17788 autoclave test scale for comparison with Test Group 45.
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PG&E Letter DCL-21-034 Enclosure Table 6-2: Critical Projected Diablo Canyon Post-LOCA Plant Conditions (test scale)
Diablo Canyon Unit 1 Diablo Canyon Unit 2 Parameter (Test Scale) (Test Scale)
Buffer Sodium Hydroxide Sodium Hydroxide Sump pH (Long-term) 8.0 - 9.5 8.0 - 9.5 Minimum Sump Volume 1.76 ft3 (50 L*) 1.76 ft3 (50 L)
Maximum Sump Pool 261°F 261°F Temperature Maximum Calcium Silicate 0.889 g 1.335 g Maximum E-Glass 3.04 g 1.85 g Maximum Silica 0g 0g Maximum Mineral Wool 0g 0g Maximum Aluminum Silicate 0.786 g 0.968 g Maximum Concrete Not Determined Not Determined Maximum Interam' 0g 0g Maximum Aluminum 0.1031 ft2 0.1031 ft2 Galvanized Steel Not Determined Not Determined
- Refer to Page 3-2 of WCAP-17788-NP, Volume 5 for this volume of test solution.
Test 45-01 and Test IBOB 45-01 used sodium hydroxide as the buffer. The test pH values fall within the representative range for DCPP. The maximum test temperatures are greater than the maximum DCPP sump temperatures. Finally, the maximum test debris and aluminum amounts exceed the projected DCPP post-LOCA plant amounts.
The maximum post-LOCA exposed concrete surface area was not determined for DCPP. As stated in the response in part 3.o.2.3 of the submittal, exposed concrete does not significantly impact chemical product generation.
The maximum post-LOCA galvanized steel surface area was not determined for DCPP. The presence of zinc from galvanized steel is postulated to impact the release of aluminum from aluminum metal. Despite the variability in the galvanized steel surface area included in Test 45-01 versus Test IBOB 45-01, the aluminum concentrations measured in Test 45-01 and IBOB 45-01 were not significantly different relative to the aluminum precipitation boundary.
Filtration tests did not detect precipitates for Test 45-01 or Test IBOB 45-01 down to a temperature of 120°F for the 24-hour duration. A DCPP containment sump temperature of 120°F after only 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> is indicative of a significantly less severe accident than simulated within the autoclaves. Therefore, the WCAP-17788 autoclave Test Group 45 results demonstrate that the minimum post-LOCA precipitation time for DCPP is greater than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
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