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| Enclosure to NL-18-0611 SNC Response to NRC Request for Additional Information (RAis) | | Enclosure to NL-18-0611 SNC Response to NRC Request for Additional Information (RAis) |
| NRC RAI37 Paragraph (b) of Title 10 of the Code of Federal Regulations, Section 50.46, .. Acceptance criteria for emergency core cooling systems for light-water nuclear power reactors, .. requires, in part, that an emergency core cooling system (ECCS) be provided for long-term cooling after successful initial operation of the ECCS. Water for long-term cooling is recirculated through the plant's sump strainer. GL 2004-02 contained a request that licensees provide verification that the sump screens (i.e., sump strainers) are capable of withstanding the loads imposed by the accumulation of debris and pressure differentials caused by blockage under flow conditions. Item 3.k of the NRC statt*s revised content guide for GL 2004-02 supplement responses (ADAMS Accession No. | | NRC RAI37 Paragraph (b) of Title 10 of the Code of Federal Regulations, Section 50.46, .. Acceptance criteria for emergency core cooling systems for light-water nuclear power reactors, .. requires, in part, that an emergency core cooling system (ECCS) be provided for long-term cooling after successful initial operation of the ECCS. Water for long-term cooling is recirculated through the plant's sump strainer. GL 2004-02 contained a request that licensees provide verification that the sump screens (i.e., sump strainers) are capable of withstanding the loads imposed by the accumulation of debris and pressure differentials caused by blockage under flow conditions. Item 3.k of the NRC statt*s revised content guide for GL 2004-02 supplement responses (ADAMS Accession No. ML073110389) requests licensees summarize the structural qualification results and design margins for various components of the sump strainer structural assembly. |
| ML073110389) requests licensees summarize the structural qualification results and design margins for various components of the sump strainer structural assembly. | |
| Table 3.k.1-3 of Enclosure 2, states that the crush pressure on the strainer due to suction strainer operation is equivalent to 10.1 ft. of head loss. This pressure is used in the load combinations for the structural analysis of the strainer. However, several locations in Enclosure 2 (e.g., Tables 3.f.14-1 and 3.g.16-1) identify strainer head loss values greater than 10.1 ft. and the supplemental response to GL 2004-02 item 3.f.7 notes that the strainer structural margin is 24ft. It is not clear what the head loss limit is for the strainer. | | Table 3.k.1-3 of Enclosure 2, states that the crush pressure on the strainer due to suction strainer operation is equivalent to 10.1 ft. of head loss. This pressure is used in the load combinations for the structural analysis of the strainer. However, several locations in Enclosure 2 (e.g., Tables 3.f.14-1 and 3.g.16-1) identify strainer head loss values greater than 10.1 ft. and the supplemental response to GL 2004-02 item 3.f.7 notes that the strainer structural margin is 24ft. It is not clear what the head loss limit is for the strainer. |
| : a. Please identify the head loss limit for structural qualification of the strainer and explain how this value was determined. | | : a. Please identify the head loss limit for structural qualification of the strainer and explain how this value was determined. |
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Category:Letter type:NL
MONTHYEARNL-24-0337, Interim 10 CFR 21.21(a)(2) Report Regarding Operation Technology, Inc., ETAP Software Error in Transient Stability Program2024-09-0909 September 2024 Interim 10 CFR 21.21(a)(2) Report Regarding Operation Technology, Inc., ETAP Software Error in Transient Stability Program NL-24-0299, Exemption Request: Final Safety Analysis Report Update Schedule, Response to Request for Additional Information2024-08-14014 August 2024 Exemption Request: Final Safety Analysis Report Update Schedule, Response to Request for Additional Information NL-24-0282, License Amendment Request and Exemption Request: Remove Tier 1 and Tier 2* Requirements2024-07-25025 July 2024 License Amendment Request and Exemption Request: Remove Tier 1 and Tier 2* Requirements NL-24-0126, – Units 3 and 4, License Amendment Request: Changes to Technical Specification 3.7.6, Main Control Room Emergency Habitability System (Ves) Action a and SR 3.7.6.62024-07-25025 July 2024 – Units 3 and 4, License Amendment Request: Changes to Technical Specification 3.7.6, Main Control Room Emergency Habitability System (Ves) Action a and SR 3.7.6.6 NL-24-0286, Emergency Request to Revise Technical Specification 3.7.9 for a One-Time Change to Support a Unit 1 Nuclear Service Cooling Water Transfer Pump Repair2024-07-20020 July 2024 Emergency Request to Revise Technical Specification 3.7.9 for a One-Time Change to Support a Unit 1 Nuclear Service Cooling Water Transfer Pump Repair NL-24-0261, 10 CFR 50.46 ECCS Evaluation Model Annual Report for 20232024-07-19019 July 2024 10 CFR 50.46 ECCS Evaluation Model Annual Report for 2023 NL-24-0227, Proposed Inservice Inspection Alternative GEN-ISI-AL T-2024-03 for Pressurizer Welds in Accordance with 10 CFR 50.55a(z)(1)2024-07-0303 July 2024 Proposed Inservice Inspection Alternative GEN-ISI-AL T-2024-03 for Pressurizer Welds in Accordance with 10 CFR 50.55a(z)(1) NL-24-0234, Application to Revise Technical Specifications to Adopt TSTF-589, Eliminate Automatic Diesel Generator Start During Shutdown2024-06-28028 June 2024 Application to Revise Technical Specifications to Adopt TSTF-589, Eliminate Automatic Diesel Generator Start During Shutdown NL-24-0143, Proposed Alternative to Use ASME Code Case N-752, Risk-Informed Categorization and Treatment for Repair/ Replacement Activities in2024-06-27027 June 2024 Proposed Alternative to Use ASME Code Case N-752, Risk-Informed Categorization and Treatment for Repair/ Replacement Activities in NL-24-0087, License Amendment Request: Changes to Technical Specification 3.7.6, Main Control Room Emergency Habitability System (Ves) Air Storage Tanks2024-06-21021 June 2024 License Amendment Request: Changes to Technical Specification 3.7.6, Main Control Room Emergency Habitability System (Ves) Air Storage Tanks NL-24-0201, Proposed Inservice Inspection Alternative GEN-ISI-ALT-2024-002 for Steam Generator Welds in Accordance with 10 CFR 50.55a(z)(1)2024-06-18018 June 2024 Proposed Inservice Inspection Alternative GEN-ISI-ALT-2024-002 for Steam Generator Welds in Accordance with 10 CFR 50.55a(z)(1) NL-24-0243, Registration of Spent Fuel Cask Use2024-06-18018 June 2024 Registration of Spent Fuel Cask Use NL-24-0202, SNC Response to Regulatory Issue Summary 2024-01: Preparation and Scheduling of Operator Licensing Examinations2024-05-24024 May 2024 SNC Response to Regulatory Issue Summary 2024-01: Preparation and Scheduling of Operator Licensing Examinations NL-24-0191, Annual Radiological Environmental Operating Reports for 20232024-05-10010 May 2024 Annual Radiological Environmental Operating Reports for 2023 NL-24-0194, Revised Request for Relief and Alternative Requirements for Squib Valves First Test Interval (V34-IST-ALT-03-R1)2024-05-0707 May 2024 Revised Request for Relief and Alternative Requirements for Squib Valves First Test Interval (V34-IST-ALT-03-R1) NL-24-0170, Responses to Second Round NRC Request for Additional Information for Refueling Outage 1 R24 Steam Generator Tube Inspection Report2024-04-25025 April 2024 Responses to Second Round NRC Request for Additional Information for Refueling Outage 1 R24 Steam Generator Tube Inspection Report NL-24-0165, Annual Non-Radiological Environmental Operating Reports and Annual Radioactive Effluent Release Reports for 20232024-04-25025 April 2024 Annual Non-Radiological Environmental Operating Reports and Annual Radioactive Effluent Release Reports for 2023 NL-24-0154, Preservice Inspection Progam Owners Activity Report2024-04-16016 April 2024 Preservice Inspection Progam Owners Activity Report NL-24-0128, License Amendment Request: Technical Specification 5.5.13, Ventilation Filter Testing Program (VFTP) Testing Frequency2024-04-0404 April 2024 License Amendment Request: Technical Specification 5.5.13, Ventilation Filter Testing Program (VFTP) Testing Frequency NL-24-0115, Response to Request for Additional Information Exemption Requests for Physical.2024-04-0404 April 2024 Response to Request for Additional Information Exemption Requests for Physical. NL-24-0116, Nuclear Property Insurance Coverage as of April 1, 2024 and Licensee Guarantees of Payment of Deferred.2024-03-29029 March 2024 Nuclear Property Insurance Coverage as of April 1, 2024 and Licensee Guarantees of Payment of Deferred. NL-24-0098, Report on Status of Decommissioning Funding2024-03-28028 March 2024 Report on Status of Decommissioning Funding NL-24-0112, Request for Revision to NRC Staff Assessment of Vogtle Site Updated Seismic Hazards2024-03-28028 March 2024 Request for Revision to NRC Staff Assessment of Vogtle Site Updated Seismic Hazards NL-24-0097, Exemption Request: Final Safety Analysis Report Update Schedule2024-03-22022 March 2024 Exemption Request: Final Safety Analysis Report Update Schedule NL-22-0267, License Amendment Request to Revise Diesel Generator Frequency and Voltage Ranges for Technical Specification 3.8.1, AC Sources – Operating, Surveillance Requirements2024-03-20020 March 2024 License Amendment Request to Revise Diesel Generator Frequency and Voltage Ranges for Technical Specification 3.8.1, AC Sources – Operating, Surveillance Requirements NL-24-0060, Request for Relief and Alternative Requirements for Squib Valves First Test Interval (V34-IST-ALT-03)2024-02-15015 February 2024 Request for Relief and Alternative Requirements for Squib Valves First Test Interval (V34-IST-ALT-03) NL-24-0042, Response to Request for Additional Information Exemption Requests for Physical Barriers2024-02-13013 February 2024 Response to Request for Additional Information Exemption Requests for Physical Barriers NL-24-0045, Supplement to Response to Requests for Additional Information for Proposed Alternative ALT-VR-022024-02-0808 February 2024 Supplement to Response to Requests for Additional Information for Proposed Alternative ALT-VR-02 NL-24-0033, Response to Question for Withholding Information from Public Disclosure Exemption Requests for Physical Barriers2024-02-0505 February 2024 Response to Question for Withholding Information from Public Disclosure Exemption Requests for Physical Barriers NL-24-0038, National Pollutant Discharge Elimination System (NPDES) Permit Renewal Application2024-02-0202 February 2024 National Pollutant Discharge Elimination System (NPDES) Permit Renewal Application NL-24-0031, Inservice Inspection Program Owners Activity Report (OAR-1) for Outage 2R232024-01-29029 January 2024 Inservice Inspection Program Owners Activity Report (OAR-1) for Outage 2R23 NL-24-0020, Plan, Unit 1 Responses to NRC Request for Additional Information for Refueling Outage 1R24 Steam Generator Tube Inspection Report2024-01-22022 January 2024 Plan, Unit 1 Responses to NRC Request for Additional Information for Refueling Outage 1R24 Steam Generator Tube Inspection Report NL-23-0926, Correction of Technical Specification Typographical Error2024-01-12012 January 2024 Correction of Technical Specification Typographical Error NL-23-0878, Request for Exemption from Security Event Notification Implementation2023-11-29029 November 2023 Request for Exemption from Security Event Notification Implementation NL-23-0827, Response to Requests for Additional Information for a License Amendment Request and a Proposed Alternative Related to TS 3.4.142023-11-17017 November 2023 Response to Requests for Additional Information for a License Amendment Request and a Proposed Alternative Related to TS 3.4.14 NL-23-0750, Response to Second Request for Additional Information Regarding License Amendment Request: Technical Specification Revision to Adopt TSTF-339-A, Relocate Technical Specification Parameters to the Colr2023-10-0404 October 2023 Response to Second Request for Additional Information Regarding License Amendment Request: Technical Specification Revision to Adopt TSTF-339-A, Relocate Technical Specification Parameters to the Colr NL-23-0745, Refueling Outage 1R24 Steam Generator Tube Inspection Report2023-09-22022 September 2023 Refueling Outage 1R24 Steam Generator Tube Inspection Report NL-23-0742, to Non-Voluntary License Amendment Request: Technical Specification Revision to Adopt WCAP-17661-P-A, Improved RAOC and CAOC Fo Surveillance Technical Specifications2023-09-20020 September 2023 to Non-Voluntary License Amendment Request: Technical Specification Revision to Adopt WCAP-17661-P-A, Improved RAOC and CAOC Fo Surveillance Technical Specifications NL-23-0695, Response to Round 2 Request for Additional Information Regarding License Amendment Request for Technical Specification 3.8.3, Inverters Operating, Completion Time Extension (LAR-22-002S2)2023-08-31031 August 2023 Response to Round 2 Request for Additional Information Regarding License Amendment Request for Technical Specification 3.8.3, Inverters Operating, Completion Time Extension (LAR-22-002S2) NL-23-0666, License Amendment Request: Remove Combined License Appendix C (LAR 23-008)2023-08-28028 August 2023 License Amendment Request: Remove Combined License Appendix C (LAR 23-008) NL-23-0670, Response to Request for Additional Information Regarding License Amendment Request: Technical Specification Revision to Adopt TSTF-339-A, Relocate Technical Specification Parameters to the COLR Consistent with WCAP-144832023-08-11011 August 2023 Response to Request for Additional Information Regarding License Amendment Request: Technical Specification Revision to Adopt TSTF-339-A, Relocate Technical Specification Parameters to the COLR Consistent with WCAP-14483 NL-23-0658, Southern Nuclear Operating Company - Response to Request for Additional Information Regarding Quality Assurance Topical Report Submittal2023-08-11011 August 2023 Southern Nuclear Operating Company - Response to Request for Additional Information Regarding Quality Assurance Topical Report Submittal NL-23-0672, Licensee Guarantees of Payment of Deferred Premiums2023-08-10010 August 2023 Licensee Guarantees of Payment of Deferred Premiums NL-23-0542, CFR 50.46 ECCS Evaluation Model Annual Report for 20222023-08-0909 August 2023 CFR 50.46 ECCS Evaluation Model Annual Report for 2022 NL-23-0624, Report of Changes to Emergency Plan and Summary of 50.54(q) Analysis2023-08-0404 August 2023 Report of Changes to Emergency Plan and Summary of 50.54(q) Analysis NL-23-0611, Core Operating Limits Report, Cycle 11 Version 12023-07-19019 July 2023 Core Operating Limits Report, Cycle 11 Version 1 NL-23-0584, Preservice Inspection Program Owner'S Activity Report2023-07-10010 July 2023 Preservice Inspection Program Owner'S Activity Report NL-23-0555, Request for Exemption from Physical Barrier Requirement2023-07-0707 July 2023 Request for Exemption from Physical Barrier Requirement NL-23-0506, to Non-Voluntary License Amendment Request: Technical Specification Revision to Adopt WCAP-17661-P-A, Improved RAOC and CAOC Fq Surveillance Technical Specifications2023-07-0505 July 2023 to Non-Voluntary License Amendment Request: Technical Specification Revision to Adopt WCAP-17661-P-A, Improved RAOC and CAOC Fq Surveillance Technical Specifications NL-23-0541, Inservice Inspection Program Owners Activity Report for Outage 1R242023-06-28028 June 2023 Inservice Inspection Program Owners Activity Report for Outage 1R24 2024-09-09
[Table view] Category:Response to Request for Additional Information (RAI)
MONTHYEARNL-24-0299, Exemption Request: Final Safety Analysis Report Update Schedule, Response to Request for Additional Information2024-08-14014 August 2024 Exemption Request: Final Safety Analysis Report Update Schedule, Response to Request for Additional Information NL-24-0194, Revised Request for Relief and Alternative Requirements for Squib Valves First Test Interval (V34-IST-ALT-03-R1)2024-05-0707 May 2024 Revised Request for Relief and Alternative Requirements for Squib Valves First Test Interval (V34-IST-ALT-03-R1) NL-24-0170, Responses to Second Round NRC Request for Additional Information for Refueling Outage 1 R24 Steam Generator Tube Inspection Report2024-04-25025 April 2024 Responses to Second Round NRC Request for Additional Information for Refueling Outage 1 R24 Steam Generator Tube Inspection Report NL-24-0042, Response to Request for Additional Information Exemption Requests for Physical Barriers2024-02-13013 February 2024 Response to Request for Additional Information Exemption Requests for Physical Barriers NL-24-0045, Supplement to Response to Requests for Additional Information for Proposed Alternative ALT-VR-022024-02-0808 February 2024 Supplement to Response to Requests for Additional Information for Proposed Alternative ALT-VR-02 NL-24-0033, Response to Question for Withholding Information from Public Disclosure Exemption Requests for Physical Barriers2024-02-0505 February 2024 Response to Question for Withholding Information from Public Disclosure Exemption Requests for Physical Barriers NL-24-0020, Plan, Unit 1 Responses to NRC Request for Additional Information for Refueling Outage 1R24 Steam Generator Tube Inspection Report2024-01-22022 January 2024 Plan, Unit 1 Responses to NRC Request for Additional Information for Refueling Outage 1R24 Steam Generator Tube Inspection Report NL-23-0827, Response to Requests for Additional Information for a License Amendment Request and a Proposed Alternative Related to TS 3.4.142023-11-17017 November 2023 Response to Requests for Additional Information for a License Amendment Request and a Proposed Alternative Related to TS 3.4.14 NL-23-0750, Response to Second Request for Additional Information Regarding License Amendment Request: Technical Specification Revision to Adopt TSTF-339-A, Relocate Technical Specification Parameters to the Colr2023-10-0404 October 2023 Response to Second Request for Additional Information Regarding License Amendment Request: Technical Specification Revision to Adopt TSTF-339-A, Relocate Technical Specification Parameters to the Colr NL-23-0695, Response to Round 2 Request for Additional Information Regarding License Amendment Request for Technical Specification 3.8.3, Inverters Operating, Completion Time Extension (LAR-22-002S2)2023-08-31031 August 2023 Response to Round 2 Request for Additional Information Regarding License Amendment Request for Technical Specification 3.8.3, Inverters Operating, Completion Time Extension (LAR-22-002S2) NL-23-0658, Southern Nuclear Operating Company - Response to Request for Additional Information Regarding Quality Assurance Topical Report Submittal2023-08-11011 August 2023 Southern Nuclear Operating Company - Response to Request for Additional Information Regarding Quality Assurance Topical Report Submittal NL-23-0670, Response to Request for Additional Information Regarding License Amendment Request: Technical Specification Revision to Adopt TSTF-339-A, Relocate Technical Specification Parameters to the COLR Consistent with WCAP-144832023-08-11011 August 2023 Response to Request for Additional Information Regarding License Amendment Request: Technical Specification Revision to Adopt TSTF-339-A, Relocate Technical Specification Parameters to the COLR Consistent with WCAP-14483 NL-23-0339, Response to Request for Additional Information Regarding License Amendment Request for Technical Specification 3.8.3, Inverters - Operating, Completion Time Extension (LAR-22-002S1)2023-06-13013 June 2023 Response to Request for Additional Information Regarding License Amendment Request for Technical Specification 3.8.3, Inverters - Operating, Completion Time Extension (LAR-22-002S1) NL-23-0467, Response to Request for Additional Information Regarding License Amendment Request: Timing of Unit 4 Technical Specifications Effectiveness Prior to Initial Criticality (LAR-23-005S1)2023-06-0909 June 2023 Response to Request for Additional Information Regarding License Amendment Request: Timing of Unit 4 Technical Specifications Effectiveness Prior to Initial Criticality (LAR-23-005S1) NL-23-0196, Response to Request for Additional Information Regarding License Amendment Request for Alternative Source Term, TSTF-51, TSTF-471, and TSTF-4902023-03-24024 March 2023 Response to Request for Additional Information Regarding License Amendment Request for Alternative Source Term, TSTF-51, TSTF-471, and TSTF-490 NL-23-0010, Response to Request for Additional Information Regarding License Amendment Request for Alternative Source Term, TSTF-51, TSTF-471, and TSTF-4902023-02-0606 February 2023 Response to Request for Additional Information Regarding License Amendment Request for Alternative Source Term, TSTF-51, TSTF-471, and TSTF-490 NL-23-0013, Response to Request for Additional Information Related to 10CFR 50.75 Decommissioning Funding Assurance Requirements2023-01-30030 January 2023 Response to Request for Additional Information Related to 10CFR 50.75 Decommissioning Funding Assurance Requirements NL-23-0068, Responses to NRC Request for Additional Information Related to the Refueling Outage 2R22 Steam Generator Tube Inspection Report2023-01-27027 January 2023 Responses to NRC Request for Additional Information Related to the Refueling Outage 2R22 Steam Generator Tube Inspection Report NL-23-0012, Post-Audit Supplement to License Amendment Request and Exemptions to Allow Use of Lead Test Assemblies for Accident-Tolerant Fuel2023-01-20020 January 2023 Post-Audit Supplement to License Amendment Request and Exemptions to Allow Use of Lead Test Assemblies for Accident-Tolerant Fuel NL-22-0609, Response to NRC Requests for Information License Amendment Request and Exemptions to Allow Use of Lead Test Assemblies for Accident-Tolerant Fuel2022-09-13013 September 2022 Response to NRC Requests for Information License Amendment Request and Exemptions to Allow Use of Lead Test Assemblies for Accident-Tolerant Fuel NL-22-0549, Responses to NRC Request for Additional Information for Refueling Outage 1R23 Steam Generator Tube Inspection Report2022-07-19019 July 2022 Responses to NRC Request for Additional Information for Refueling Outage 1R23 Steam Generator Tube Inspection Report ND-22-0328, Enclosuvogtle Electric Generating Plant, Units 3 & 4 - Fukushima Response NEI 12-01 On-Shift Staffing Analysis Phase 2 Report, Revision 3.0, Standard Emergency Plan Annex, Revision 6.02022-06-29029 June 2022 Enclosuvogtle Electric Generating Plant, Units 3 & 4 - Fukushima Response NEI 12-01 On-Shift Staffing Analysis Phase 2 Report, Revision 3.0, Standard Emergency Plan Annex, Revision 6.0 NL-22-0055, Response to Request for Additional Information Regarding License Amendment Request to Revise Technical Specification 3.7.2 Limiting Condition for Operation to Remove One Main Steam Isolation Valve System2022-02-28028 February 2022 Response to Request for Additional Information Regarding License Amendment Request to Revise Technical Specification 3.7.2 Limiting Condition for Operation to Remove One Main Steam Isolation Valve System NL-21-1015, Response to Request for Additional Information Regarding License Amendment Request to Revise Technical Specification 3.7.2 Limiting Condition for Operation to Remove One Main Steam Isolation Valve System2022-01-13013 January 2022 Response to Request for Additional Information Regarding License Amendment Request to Revise Technical Specification 3.7.2 Limiting Condition for Operation to Remove One Main Steam Isolation Valve System ND-21-0618, Response to Request for Additional Information Regarding Request for Alternative: Alternative Requirements for ASME Section Ill Remediation of Containment Vessel Unistrut Welding (VEGP 3-ALT-16)2021-07-0202 July 2021 Response to Request for Additional Information Regarding Request for Alternative: Alternative Requirements for ASME Section Ill Remediation of Containment Vessel Unistrut Welding (VEGP 3-ALT-16) NL-21-0208, Response to Request for Additional Information Related to Refueling Outage 1R22 Steam Generator Tube Inspection Report2021-03-0202 March 2021 Response to Request for Additional Information Related to Refueling Outage 1R22 Steam Generator Tube Inspection Report NL-20-1417, Vogltle, Units 1 and 2, Response to Request for Additional Information Regarding Risk-Informed Resolution to GSI-1912020-12-17017 December 2020 Vogltle, Units 1 and 2, Response to Request for Additional Information Regarding Risk-Informed Resolution to GSI-191 NL-20-1312, Response to Request for Additional Information Related to Proposed Inservice Inspection Alternative VEGP-ISI-ALT-04-04 Version 2.02020-11-23023 November 2020 Response to Request for Additional Information Related to Proposed Inservice Inspection Alternative VEGP-ISI-ALT-04-04 Version 2.0 NL-20-1273, Response to NRC Requests for Information License Amendment Request to Revise the Emergency Plan to Change Staffing and Extend Staff Augmentation Times for2020-11-20020 November 2020 Response to NRC Requests for Information License Amendment Request to Revise the Emergency Plan to Change Staffing and Extend Staff Augmentation Times for ND-20-1005, Supplement to Request for License Amendment: Vacuum Relief Valve Technical Specification Changes (LAR-20-005S1)2020-09-18018 September 2020 Supplement to Request for License Amendment: Vacuum Relief Valve Technical Specification Changes (LAR-20-005S1) NL-20-0640, Response to Request for Additional Information for Exemption Request2020-05-27027 May 2020 Response to Request for Additional Information for Exemption Request ND-19-1097, Response to Request for Additional Information - LAR-19-013R12019-11-0101 November 2019 Response to Request for Additional Information - LAR-19-013R1 NL-19-1183, Response to Request for Additional Information Regarding End State Revision from Hot Shutdown to Cold Shutdown2019-10-17017 October 2019 Response to Request for Additional Information Regarding End State Revision from Hot Shutdown to Cold Shutdown ND-19-1023, Enclosure 7 - Response to Draft Request for Additional Information (LAR-19-005R1)2019-10-10010 October 2019 Enclosure 7 - Response to Draft Request for Additional Information (LAR-19-005R1) ND-19-1129, Supplement to the Request for License Amendment Regarding Protection and Safety Monitoring System Surveillance Requirement Reduction Technical Specification ...2019-10-0707 October 2019 Supplement to the Request for License Amendment Regarding Protection and Safety Monitoring System Surveillance Requirement Reduction Technical Specification ... ND-19-1177, Supplement to Request for License Amendment and Exemption: Addition of In-Containment Refueling Water Storage Tank to Radiation Analyses (LAR-19-003S1)2019-10-0303 October 2019 Supplement to Request for License Amendment and Exemption: Addition of In-Containment Refueling Water Storage Tank to Radiation Analyses (LAR-19-003S1) NL-19-0709, Core Operating Limits Report, Cycle 21, Version 1, SNC Response to NRC Request for Additional Information2019-06-17017 June 2019 Core Operating Limits Report, Cycle 21, Version 1, SNC Response to NRC Request for Additional Information NL-19-0498, License Amendment Request for Technical Specification Improvement to Revise Actions for One Steam Supply to Turbine Driven Auxiliary Feedwater Pump Inoperable SNC Response to NRC Request for ...2019-04-30030 April 2019 License Amendment Request for Technical Specification Improvement to Revise Actions for One Steam Supply to Turbine Driven Auxiliary Feedwater Pump Inoperable SNC Response to NRC Request for ... ND-19-0289, Enclosure 1 - Revised Response to Request for Additional Information Question 1b Related to Request for Exemption from Operator Written Examination and Operating Test2019-04-16016 April 2019 Enclosure 1 - Revised Response to Request for Additional Information Question 1b Related to Request for Exemption from Operator Written Examination and Operating Test ND-19-0178, Supplement to Request for Exemption: 10 CFR Part 26 Visitor Access Requirements (Supplement 1)2019-03-0808 March 2019 Supplement to Request for Exemption: 10 CFR Part 26 Visitor Access Requirements (Supplement 1) ND-19-0216, Response to Request for Additional Information Related to Request for Exemption from Operator Written Examination and Operating Test2019-03-0404 March 2019 Response to Request for Additional Information Related to Request for Exemption from Operator Written Examination and Operating Test NL-18-1514, Response to Request for Information Regarding License Amendment Request to Revise Technical Specification 5.2.2.g and Update Emergency Plan Minimum On-Shift.2019-01-31031 January 2019 Response to Request for Information Regarding License Amendment Request to Revise Technical Specification 5.2.2.g and Update Emergency Plan Minimum On-Shift. ND-19-0007, Supplement to Request for Alternative: Preservice and Inservice Inspection Requirements for Specific Valve-to-Pipe Welds (VEGP 3&4-PSI/ISI-ALT-06S1)2019-01-28028 January 2019 Supplement to Request for Alternative: Preservice and Inservice Inspection Requirements for Specific Valve-to-Pipe Welds (VEGP 3&4-PSI/ISI-ALT-06S1) ND-18-1439, Revised Request for License Amendment: Technical Specification Changes for Spent Fuel Pool Level - Low 2 and In-containment Refueling Water Storage Tank (IRWST) Wide Range Level - Low Operability (LAR-18-017R1)2018-12-0303 December 2018 Revised Request for License Amendment: Technical Specification Changes for Spent Fuel Pool Level - Low 2 and In-containment Refueling Water Storage Tank (IRWST) Wide Range Level - Low Operability (LAR-18-017R1) NL-18-1181, Relief Requests VEGP-ISI-RR-05 and VEGP-ISI-RR-06, SNC Response to NRC Request for Additional Information2018-10-17017 October 2018 Relief Requests VEGP-ISI-RR-05 and VEGP-ISI-RR-06, SNC Response to NRC Request for Additional Information ND-18-1058, Supplement to Request for License Amendment and Exemption: Containment Pressure Analysis (LAR-17-043S1)2018-09-28028 September 2018 Supplement to Request for License Amendment and Exemption: Containment Pressure Analysis (LAR-17-043S1) NL-18-1130, Units 1 and 2 - Relief Request VEGP-ISI-RR-03: SNC Response to NRC Request for Additional Information2018-09-26026 September 2018 Units 1 and 2 - Relief Request VEGP-ISI-RR-03: SNC Response to NRC Request for Additional Information NL-18-1179, Response to NRC Request for Additional Information Tornado Missile Risk Evaluator SNC Supplemental2018-09-14014 September 2018 Response to NRC Request for Additional Information Tornado Missile Risk Evaluator SNC Supplemental ND-18-0635, Response to Request for Additional Information Regarding Application of VT-1 Visual Examination Methodology for Preservice Inspection of the Reactor Vessel Nozzle Inner Radius Sections (VEGP 3&4-PSI-ALT-07S3)2018-08-31031 August 2018 Response to Request for Additional Information Regarding Application of VT-1 Visual Examination Methodology for Preservice Inspection of the Reactor Vessel Nozzle Inner Radius Sections (VEGP 3&4-PSI-ALT-07S3) ND-18-1040, Supplement to the Request for License Amendment: Technical Specification Changes to Support Operability During Mode 5 Vacuum Fill Operations (LAR-18-009S1)2018-08-10010 August 2018 Supplement to the Request for License Amendment: Technical Specification Changes to Support Operability During Mode 5 Vacuum Fill Operations (LAR-18-009S1) 2024-08-14
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~ Souther Nuclea Cheryl A. Gayheart Regulatory Affairs Director 40 ln vcmcss Ccnlcr Parkway Posl Otlicc Box 12lJ5 Birm ingham, AI. ::\5242 205 992 53 1fi lei
. 05 992 7fi0 I fa x MAY 2 3 2018 *ugayhca<P)sc>ulhcmcn.ccJm Docket Nos.: 50-424 NL-18-0611 50-425 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D. C. 20555-0001 Voqtle Electric Generating Plant Units1 &2 Systematic Risk-Informed Assessment of Debris Technical Report SNC Response to NRC Request for Additional Information (RAis #37-39)
Ladies and Gentlemen:
By letter dated April21, 2017 (Agencywide Documents Access and Management System Accession No. ML17116A098) as supplemented by letters dated July 11, 2017; November 9, 2017; January 2, 2018; January 9, 2018; February 6, 2018, February 12, 2018 and February 21, 2018; Southern Nuclear Operating Company, Inc. (SNC) submitted a plant-specific technical report for Vogtle Electric Generating Plant (VEGP), Units 1 and 2 and requested U.S. Nuclear Regulatory Commission (NRC) approval of the methods and inputs described in the technical report. The plant-specific technical report describes a risk-informed methodology to evaluate debris effects with the exception of in-vessel fiber limits. By letter dated May 1, 2018, the NRC staff notified SNC that additional information is needed for the staff to complete their review. The Enclosure provides the SNC response to the NRC requests for additional information (RAis) for RAis 37-39.
This letter contains no NRC commitments. If you have any questions, please contact Jamie Coleman at 205.992.6611.
I declare under penalty of perjury that the foregoing is true and correct. Executed on the 2..?, day of May 2018.
Respectfully submitted, Cheryl A. G. e Director, Regulatory Affairs Southern Nuclear Operating Company CAG/PDB/SCM
U.S. Nuclear Regulatory Commission NL-18-0611 Page2
Enclosure:
SNC Response to NRC Request for Additional Information cc: Regional Administrator, Region II NRR Project Manager- Vogtle 1 & 2 Senior Resident Inspector- Vogtle 1 & 2 State of Georgia Environmental Protection Division RType: CVC700
Vogtle Electric Generating Plant- Units 1 & 2 Systematic Risk-Informed Assessment of Debris Technical Report SNC Response to NRC Request for Additional Information (RAis #37-39)
Enclosure SNC Response to NRC Request for Additional Information (RAis)
Enclosure to NL-18-0611 SNC Response to NRC Request for Additional Information (RAis)
NRC RAI37 Paragraph (b) of Title 10 of the Code of Federal Regulations, Section 50.46, .. Acceptance criteria for emergency core cooling systems for light-water nuclear power reactors, .. requires, in part, that an emergency core cooling system (ECCS) be provided for long-term cooling after successful initial operation of the ECCS. Water for long-term cooling is recirculated through the plant's sump strainer. GL 2004-02 contained a request that licensees provide verification that the sump screens (i.e., sump strainers) are capable of withstanding the loads imposed by the accumulation of debris and pressure differentials caused by blockage under flow conditions. Item 3.k of the NRC statt*s revised content guide for GL 2004-02 supplement responses (ADAMS Accession No. ML073110389) requests licensees summarize the structural qualification results and design margins for various components of the sump strainer structural assembly.
Table 3.k.1-3 of Enclosure 2, states that the crush pressure on the strainer due to suction strainer operation is equivalent to 10.1 ft. of head loss. This pressure is used in the load combinations for the structural analysis of the strainer. However, several locations in Enclosure 2 (e.g., Tables 3.f.14-1 and 3.g.16-1) identify strainer head loss values greater than 10.1 ft. and the supplemental response to GL 2004-02 item 3.f.7 notes that the strainer structural margin is 24ft. It is not clear what the head loss limit is for the strainer.
- a. Please identify the head loss limit for structural qualification of the strainer and explain how this value was determined.
- b. If the value does not bound all postulated head loss values, please provide a justification for any exceedances.
SNC Response to RAI 37
- a. The results and methodology presented in Sections 3.k.1 and 3.k.2 of Enclosure 5 (Reference 1) are based on the structural qualification performed by the hardware vendor, GEH, prior to the strainer installation. The head loss limit used for the structural qualification (referred to as crush pressure hereafter) presented in Table 3.k.1-3 (Reference 1 pp. E5-116) was an assumed value based on what was believed to be adequate at that time. As part of the change to a risk-informed methodology, GEH revisited the stress model and results to determine if a higher crush pressure could be justified. As shown in Table 3.k.2-2 of Enclosure 5 (Reference 1 pp. E5-120), the most limiting stress is in the welds between the perforated plate to finger and the perforated plate to frame for the Service Level D Load combination. Linear scaling was used conservatively assuming that total stress is scalable with crush pressure. The resulting max allowable crush pressure is 10.7 psi for the plate to finger weld condition, and 5.03 psi for the plate to frame weld. However, this approach is too conservative and not appropriate for the plate to frame weld. A single disk FEA model was run using ANSYS with the same mesh as the original analysis and determined that the impact of the crush pressure has E-1
Enclosure to NL-18-0611 SNC Response to NRC Request for Additional Information (RAis) very limited impacts to the stress of the plate to frame weld. Using this new model, the allowable crush pressure for the plate to frame weld was calculated to be 30.47 psi. This concluded that the limiting factor for the crush pressure was the plate to finger weld which has a max allowable crush pressure of 10.7 psi.
Using ASME Code Section Ill Level 0 allowables, and the methodology above, the crush pressure was determined to be 10.39 psi for a 15-disk strainer, compared to 10. 7 psi for the 18-disk strainer. The lower crush pressure for the 15-disk strainer is due to the larger debris weight for each disk (same total debris weight with fewer disks). Therefore, a crush pressure of 10.39 psi (24.0 ft) was conservatively used for the final configuration of 16 disks, as shown in Section 3.f.7 of the submittal (Reference 1 pp. E5-73).
- b. The crush pressure of 10.39 psi (24.0 ft) bounded all postulated head loss values, as shown in Table 3.g.16-1 (Reference 1 pp. E5-103).
NRC RAI38 Title 10 of the Code of Federal Regulations, Section 50.46, Subsection (a)(1) requires in part that cooling performance be calculated with an acceptable evaluation model and must be calculated for a number of postulated loss-of-coolant accidents (LOCA) of different sizes, locations, and other properties sufficient to provide assurance that the most severe postulated loss-of-coolant accidents are calculated. Additionally, the subsection requires that the evaluation includes sufficient supporting justification to show that the analytical technique realistically describes the behavior of the reactor system during a LOCA. In its risk-informed methodology, the licensee used guidance and acceptance guidelines described in Regulatory Guide (RG) 1.174, Rev. 2, "An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis," (ADAMS Accession No. ML100910006).
Section 2.3.3 of RG 1.174, Rev. 2, states that:
... technical adequacy will be understood as being determined by the adequacy of the actual modeling and the reasonableness of the assumptions and approximations. , Section 2.0, of the submittal states that a screening process can be used to eliminate scenarios that are not relevant, not affected by debris, or have an insignificant contribution. The results of this screening process are provided, but not a description (i.e., justification) of the systematic approach implemented to identify relevant initiating events and how scenarios are eliminated.
- a. Please describe in detail the process leading to the identification of relevant internal initiating events (e.g., LOCA, open safety relief valve, water hammer-induced LOCAs, non-piping LOCAs). Please include any criteria (quantitative or qualitative) used in the process for screening (i.e., eliminating) any initiating events or scenarios.
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Enclosure to NL-18-0611 SNC Response to NRC Request for Additional Information (RAis)
- b. Please describe in more detail how the high-likelihood scenarios were determined and how the change in risk associated with low likelihood scenarios were determined for this application. Please include a summary of the process used to make these determinations.
- c. Please describe in detail the systematic process applied to evaluate the impact of secondary side breaks. Please include a summary of how and why these breaks were screened in or out.
- d. Please explain which secondary side breaks were screened from detailed analyses, and the basis for their screening.
SNC Response to RAI 38
- a. A systematic process was used to determine the hazards, initiating events, and operating modes to be addressed in the Vogtle GSI-191 analysis. The process was based on the identification of hazards and initiating events with the potential to (1) generate debris inside containment, (2) require sump recirculation for mitigation of the event, and (3) result in debris transport to the containment sump. Hazards or initiating events that do not meet these three criteria were excluded from the analysis.
Among internal plant hazards, the following initiating events do not have the potential to generate debris inside containment and were screened from the analysis:
o transients, o steam generator tube rupture, o inadvertent safety injection, o inadvertent or stuck-open power operated relief valves (PORVs) that discharge to the pressurizer relief tank (PAT),
o secondary side breaks outside containment, and o interfacing systems loss of coolant accidents (LOCAs) that discharge outside containment The internal initiating events that do have the potential to generate debris inside containment are LOCAs (small, medium, and large) due to breaks inside containment and secondary side breaks inside containment.
Internal flood hazards do not have the potential to generate debris inside containment.
Pipe breaks that flood inside containment are evaluated as LOCA or secondary side break internal events.
The internal hazards and initiating events identified above that have the potential to generate debris inside containment may also require sump recirculation for mitigation of the event and result in debris transport to the containment sump. Therefore, the following E-3
Enclosure to NL-18-0611 SNC Response to NRC Request for Additional Information (RAis) events were included in the scope of the Vogtle GSI-191 analysis and considered with a detailed or conservative quantitative assessment or a qualitative evaluation:
- 1. Large, medium, and small LOCAs due to:
- i. Pipe breaks ii. Failure of non-piping components iii. Water hammer
- 2. Secondary side breaks inside containment that result in a consequential LOCA upon failure to terminate safety injection or a stuck open PORV, requiring sump recirculation.
- b. The Vogtle PRA model was used to identify the high-likelihood equipment configurations that can occur in response to a LOCA, to focus the analysis of debris generation, transport, and resulting GSI-191 phenomena. The identification of high-likelihood equipment configurations followed a systematic process:
- 1. All possible combinations of system/train failures that affect the likelihood of debris-induced failure of ECCS following a LOCA were identified. The configurations considered were specific to each LOCA size (small, medium, or large). Examples of such configurations include successful operation of all ECCS equipment (e.g.,
RHR, containment spray, and containment coolers following a large LOCA), failure of one train of RHR, failure of both trains of containment spray, etc.
- 2. For each possible configuration, the functional failure probability (FFP) was quantified for each system/train failed in that configuration, using the associated logic in the PRA model. Those FFP values were then used to calculate total probability for each equipment failure combination.
- 3. For each possible configuration identified for each LOCA size, the annual scenario frequency was calculated based on the associated initiating event frequency and total probability of the equipment failure configuration. The scenario frequencies for each LOCA size were summed and the significant scenarios were identified as those that comprised 95o/o, or individually contributed 1°/o, of the total frequency.
The equipment failure combinations from those significant scenarios represent the high-likelihood configurations.
The high-likelihood equipment configurations were included in the detailed NARWHAL analysis for GSI-191 effects and were explicitly modeled in the PRA for quantification of the risk impacts due to GSI-191 phenomena.
The remaining low-likelihood configurations also have a risk impact from GSI-191 phenomena, but were not evaluated in detail. Instead, a bounding risk impact was E-4
Enclosure to NL-18-0611 SNC Response to NRC Request for Additional Information (RAis) calculated for the low-likelihood configurations. Calculation of the bounding risk impact for the low-likelihood configurations also followed a systematic process, as described below.
- 1. Because NARWHAL analysis was not performed to determine sump strainer or core cooling conditional failure probability (CFP) values for all low-likelihood configurations, a representative or bounding CFP was selected for each low-likelihood configuration using the following logic:
- i. Each low-likelihood configuration was binned based upon the impact on sump strainer debris accumulation, relative to the high-likelihood configurations evaluated in NARWHAL. For example, the failure of a containment cooler was captured by bounding temperature profiles used in the NARWHAL analysis, implicitly assuming the containment cooler failed.
Therefore, low-likelihood configurations with containment cooler failure (e.g., RHR train A and containment cooler A failed) were considered equivalent to the high-likelihood configuration with only RHR train A failed.
ii. Based on the binning performed, the appropriate CFP was selected to use as a representative or bounding value for the low-likelihood configuration.
For example, the low-likelihood configuration with one RHR train failed and one CS train failed, has the same impact on sump strainer accumulation as the high-likelihood configuration with one ECCS train failed due to loss of nuclear service cooling water (NSCW). Therefore, the CFP calculated for one ECCS train failure was applied to the configuration with one RHR and one CS train failed.
iii. If there was no similar high-likelihood configuration for a low-likelihood configuration, such as failure of one RHR train and both CS trains, a CFP of 1.0 was assumed.
- 2. For each low-likelihood configuration identified for each LOCA size, the bounding annual core damage frequency (CDF) was calculated based on the associated initiating event frequency, probability of the equipment failure configuration, and the selected representative or bounding CFP. The bounding CDF values from the low-likelihood configurations for each LOCA size were then summed. The base case CDF (assuming no GSI-191 failures) was neglected, so the sum of the bounding CDF from the low-likelihood configurations was assumed to be the risk increase (b.CDF).
- 3. The bounding LEAF due to low-likelihood configurations was calculated by multiplying the bounding CDF by the conditional large early release probability (CLERP) given core damage. The CLERP was obtained by dividing the base case (i.e., the case with no GSI-191 failures) LEAF by the base case CDF. (Note that the CLERP for the high-likelihood configurations is identical to that for the base E-5
Enclosure to NL-18-0611 SNC Response to NRC Request for Additional Information (RAis) case with no GSI-191 failures, 2.9E-03, so this approach to calculate bounding LEAF for the low-likelihood configurations is reasonable.) The bounding LEAF from the low-likelihood configurations was assumed to be the risk increase
(~LEAF).
- c. Secondary side breaks inside containment (SSBI) that result in a consequential LOCA upon failure to terminate safety injection or a stuck open PORV may generate debris inside containment and require sump recirculation resulting in the transport of debris to the containment sump. A bounding evaluation was performed for the SSBI risk contribution to GSI-191 failures by quantifying the SSBI accident sequences in the Vogtle internal events PRA with a failure probability of 1.0 assigned to the sump strainers. The resulting risk increase was 1.18E-7 for CDF and 6.89E-9 for LEAF. This was considered overly conservative to include in the determination of the overall GSI-191 risk impact.
Therefore, a conservative evaluation of the SSBI risk contribution was performed based on CFPs calculated in NARWHAL for secondary side breaks. The NARWHAL evaluation assumed all secondary side breaks (feedwater or main steam line breaks inside containment) were double-ended guillotine breaks, and that one or both trains of containment spray (CS) had failed. Sump strainer failure due to GSI-191 phenomena occurred only for main steam line breaks when both trains of CS fail. Using the CFP from NARWHAL to quantify the SSBI risk contribution to GSI-191 failures produced a risk increase of 1.39E-9 for CDF and 8.25E-11 for LEAF, nearly two orders of magnitude below the bounding evaluation results and well within the Regulatory Guide 1.174 Region Ill risk acceptance guidelines (Reference 2).
- d. As described in the response to RAI 38.a, secondary side breaks outside containment do not have the potential to generate debris inside containment, and therefore were screened from further analysis.
NRC RAI39 Section 2.3.1 of RG 1.174, Rev. 2, states that:
... the scope of a PRA is defined in terms of the causes of initiating events and the plant operating modes it addresses. Typical hazard groups considered in a nuclear power plant PRA include internal events, internal floods, seismic events, internal fires, high winds, external flooding, etc.
It is not apparent that the impacts of internal fire and external hazards, other than seismic, have been addressed in the submittal. These other external hazards may affect the total change in risk for this application of the PRA.
- a. Please provide a justification (e.g., qualitative arguments or bounding analyses) that demonstrates that the risk contributions from internal fire would not affect this application of the PRA.
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Enclosure to NL-18-0611 SNC Response to NRC Request for Additional Information (RAis)
- b. Please provide a justification (e.g., qualitative arguments or bounding analyses) that demonstrates that the risk contributions from external events other than seismic events would not affect this application of the PRA.
SNC Response to RAI 39
- a. Consistent with the guidance in NUREG/CR-6850 (Reference 3), internal fire hazards were not assumed to result in pipe breaks. However, fire-induced LOCAs can occur, including spurious opening of a pressurizer PORV or safety valve, spurious reactor head vent, continuous letdown, spurious interfacing system LOCA, or reactor coolant pump (RCP) seal LOCA due to loss of seal cooling. Of these, only an RCP seal LOCA has the potential to generate debris inside containment. A spurious opening of a pressurizer PORV or safety valve, or spurious reactor head vent is discharged to the PAT, which has negligible sources of debris near the rupture disk. Spurious interfacing system LOCAs or continuous letdown all discharge outside containment. Therefore, these scenarios were all screened from the analysis. The quantity of debris generated by an RCP seal LOCA is equivalent to the quantity generated by a small or medium LOCA, which was found to not challenge the sump strainers; therefore, fire-induced RCP seal LOCAs were also screened from the analysis.
- b. An evaluation of external hazards conducted for Vogtle concluded that in addition to internal flood, internal fire and seismic events, the only external hazards applicable to Vogtle are:
o aircraft impact, o extreme winds and tornadoes, o external flooding including intense local precipitation, o industrial and military facility accidents, o pipeline accidents, o transportation accidents, and o turbine-generated missiles None of these external hazards listed have the potential to generate debris inside containment and were screened from the GSI-191 analysis. Therefore, seismic events are the only external hazard that affect this application of the Vogtle PRA.
References
- 1. ML17116A098. Vogtle Electric Generating Plant- Units 1 & 2 Supplemental Response to NRC Generic Letter 2004-02. April21, 2017.
- 2. Regulatory Guide 1.174. An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis. Revision 2.
- 3. NUREG/CR-6850. Fire PRA Methodology for Nuclear Power Facilities. September 2005.
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