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Results
- Approval, Approval, Approval, Approval, Approval, Approval, Approval, Approval, Approval, Approval, Approval, Approval, Approval, Approval, Approval, Approval
Other: JAFP-17-0083, Report of Full Compliance with March 12, 2012, Commission Order to Modify Licenses with Regard to Requirements for Mitigation Strategies for Beyond-Design-Basis External Events, ML17306A484, ML18156A151, ML18180A314, ML18249A169
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MONTHYEARNRC 2017-0037, High Frequency Seismic Evaluation Confirmation Report2017-08-0202 August 2017 High Frequency Seismic Evaluation Confirmation Report Project stage: Request ML17230A0882017-08-10010 August 2017 Submittal of High Frequency Supplement for Information Per 10CFR50.54(f) Re Recommendation 2.1 of the Near-Term Task Force Review of Insights from the Fukushima Dai-ichi Accident Project stage: Supplement ML17234A4782017-08-22022 August 2017 Mitigating Strategies Assessment (MSA) Report for the Reevaluated Seismic Hazard Information - NEI 12-06, Appendix H, Revision 2, H.4.4 Path 4: GMRS ≪ 2xSSE Project stage: Request JAFP-17-0083, Report of Full Compliance with March 12, 2012, Commission Order to Modify Licenses with Regard to Requirements for Mitigation Strategies for Beyond-Design-Basis External Events2017-08-29029 August 2017 James a Fitzpatrick, Report of Full Compliance with March 12, 2012, Commission Order to Modify Licenses with Regard to Requirements for Mitigation Strategies for Beyond-Design-Basis External Events Project stage: Other ML17244A2692017-08-29029 August 2017 Catawba Nuclear Station High Frequency Supplement to Seismic Hazard Screening Report, Response NRC Request for Information Pursuant to 10 CFR 50.54(f) Re Recommendation 2.1 of the Near-Term Task Force Review of Insights from the Fukushima D Project stage: Supplement ML17256A7752017-09-11011 September 2017 NRR E-mail Capture - (External_Sender) Seabrook Flooding MSA Project stage: Request MNS-17-040, (Mns), Units 1 and 2 - Supplement to the High Frequency Supplement to Seismic Hazard Screening Report, Response NRC Request for Information Pursuant to 10 CFR 50.54(f) Regarding Recommendation 2.1 of the Near-Term Task Force ..2017-09-27027 September 2017 (Mns), Units 1 and 2 - Supplement to the High Frequency Supplement to Seismic Hazard Screening Report, Response NRC Request for Information Pursuant to 10 CFR 50.54(f) Regarding Recommendation 2.1 of the Near-Term Task Force .. Project stage: Supplement ML17291A7082017-10-23023 October 2017 Staff Review of Mitigating Strategies Assessment Report of the Impact of the Re-evaluated Seismic Hazard Developed in Response to the March 12, 2012, 50.54(f) Letter Project stage: Approval ML17277B1092017-10-23023 October 2017 Staff Review of Mitigating Strategies Assessment Report of the Impact of the Reevaluated Seismic Hazard Developed in Response to the March 12, 2012, 50.54(F) Letter Project stage: Approval ML17310B5312017-11-16016 November 2017 Staff Review of Mitigating Strategies Assessment Report of the Impact of the Re-Evaluated Seismic Hazard Developed in Response to the March 12, 2012, 50.54(f) Letter Project stage: Approval ML17306A4842017-11-29029 November 2017 Flood Hazard Mitigation Strategies Assessment Project stage: Other NL-17-1889, NEI 12-06, Appendix H, Revision 4, H.4.5 Path 5: GMRS Greater than 2 X SSE, Mitigating Strategies Assessment (MSA) Report for the New Seismic Hazard Information2017-12-0505 December 2017 NEI 12-06, Appendix H, Revision 4, H.4.5 Path 5: GMRS Greater than 2 X SSE, Mitigating Strategies Assessment (MSA) Report for the New Seismic Hazard Information Project stage: Request ML17349A9912017-12-21021 December 2017 Units 1 and 2 - Staff Review of Mitigating Strategies Assessment Report of the Impact of the Re-Evaluated Seismic Hazard Developed in Response to the March 12, 2012, 50.54(f) Letter (CAC Nos. MF7843 and MF7844, ... Project stage: Approval CNS-17-058, Supplement to the High Frequency Supplement to Seismic Hazard Screening 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 Fukus2017-12-28028 December 2017 Supplement to the High Frequency Supplement to Seismic Hazard Screening 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 Fukushi Project stage: Supplement ML17313A8812018-01-22022 January 2018 Staff Review of Mitigating Strategies Assessment Report of the Impact of the Reevaluated Seismic Hazard Developed in Response to the March 12, 2012, 50.54(f) Letter (CAC Nos. MF7809 and MF7810; EPID L-2016-JLD-0006) Project stage: Approval ML18017A1212018-01-30030 January 2018 Staff Review of Mitigating Strategies Assessment Report of the Impact of the Reevaluated Seismic Hazard Developed in Response to the March 12, 2012, 50.54(f) Letter Project stage: Approval ML18033A2092018-02-0707 February 2018 Staff Review of Mitigating Strategies Assessment Report of the Impact of the Re-Evaluated Seismic Hazard Developed in Response to the March 12, 2012, 50.54(F) Letter Project stage: Approval ML18040A4542018-02-20020 February 2018 Staff Review of Mitigating Strategies Assessment Report of the Impact of the Re-Evaluated Seismic Hazard Developed in Response to the March 12, 2012, 50.54(F) Letter Project stage: Approval ML18068A6542018-03-22022 March 2018 Staff Review of Mitigating Strategies Assessment Report of the Impact of the Re-Evaluated Seismic Hazard Developed in Response to the March 12, 2012, 50.54(F) Letter Project stage: Approval CNL-18-007, Seismic Probabilistic Risk Assessment Supplemental Information2018-04-10010 April 2018 Seismic Probabilistic Risk Assessment Supplemental Information Project stage: Request ML18115A5082018-04-30030 April 2018 Staff Review of Mitigating Strategies Assessment Report of the Impact of the Reevaluated Seismic Hazard Developed in Response to the March 12, 2012, 50.54(F) Letter Project stage: Approval ML18130A7502018-05-0909 May 2018 Report of Compliance with March 12, 2012 Commission Order Modifying Licenses with Regard to Requirements of Mitigation Strategies for Beyond-Design-Basis External Events Project stage: Request ML18159A2892018-06-13013 June 2018 Staff Review of Mitigation Strategies Assessment Report of the Impact of the Reevaluated Seismic Hazard Develop in Response to the March 12, 2012, 50.54(f) Letter Project stage: Approval NL-18-0684, Fukushima Near-Term Task Force Recommendation 2.1: Seismic Supplemental Information Regarding NEI 12-06. Appendix H. Revision 4. H.4.5 Path 5 Mitigating Strategies Assessment (MSA) Report2018-06-25025 June 2018 Fukushima Near-Term Task Force Recommendation 2.1: Seismic Supplemental Information Regarding NEI 12-06. Appendix H. Revision 4. H.4.5 Path 5 Mitigating Strategies Assessment (MSA) Report Project stage: Supplement ML18180A3142018-07-10010 July 2018 Seismic Hazard Mitigation Strategies Assessment (CAC Nos. MF7888 and MF7889; EPID L-2016-JLD-0006) Project stage: Other ML18156A1512018-07-12012 July 2018 Seismic Hazard Mitigation Strategies Assessment (CAC Nos. MF7891 and MF7892; EPID L-21016-JLD-0006) Project stage: Other ML18184A2732018-07-18018 July 2018 Staff Review of Mitigation Strategies Assessment Report of the Impact of the Reevaluated Seismic Hazard Developed in Response to the March 12, 2012, 50.54(f) Letter (CAC No. MF7819; EPID L-2016-JLD-0006) Project stage: Approval ML18173A1652018-07-19019 July 2018 Staff Review of Mitigation Strategies Assessment Report of the Impact of the Reevaluated Seismic Hazard Developed in Response to the March 12, 2012, 50.54(f) Letter (CAC No. MF7828; EPID L-2016-JLD-0006) Project stage: Approval ML18207A8542018-08-14014 August 2018 Staff Review of Mitigating Strategies Assessment Report of the Impact of the Reevaluated Seismic Hazard Developed in Response to the March 12, 2012, 50.54(F) Letter Project stage: Approval ML18236A1912018-08-29029 August 2018 Staff Review of Mitigation Strategies Assessment Report of the Impact of the Reevaluated Seismic Hazard Developed in Response to the March 12, 2012, 50.54(f) Letter (CAC No. MF7875; EPID L-2016-JLD-0006 Project stage: Approval ML18249A1692018-09-0707 September 2018 Seismic Hazard Mitigation Strategies Assessment (CAC Nos. MF7800 and MF7801; EPID L-2016-JLD-0006) Project stage: Other ML18262A4152018-09-27027 September 2018 Staff Review of Mitigating Strategies Assessment Report of the Impact of the Reevaluated Seismic Hazard Developed in Response to the March 12, 2012, 50.54(f) Letter (CAC No. MF7893; EPID No. L-2016-JLD-0006) Project stage: Approval 2018-02-20
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Category:Letter type:CNL
MONTHYEARCNL-24-052, Response to Request for Additional Information Regarding Application to Modify the Technical Specification Surveillance Requirement 3.9.5.1 (WBN-TS-21-14)2024-06-27027 June 2024 Response to Request for Additional Information Regarding Application to Modify the Technical Specification Surveillance Requirement 3.9.5.1 (WBN-TS-21-14) CNL-24-018, License Amendment Request for Adoption of Technical Specification Task Force Traveler TSTF-276-A, Revision 2, Regarding TS 3.8.1 AC Sources – Operating to Clarify Requirements for Diesel Generator Testing (WBN-TS2024-06-25025 June 2024 License Amendment Request for Adoption of Technical Specification Task Force Traveler TSTF-276-A, Revision 2, Regarding TS 3.8.1 AC Sources – Operating to Clarify Requirements for Diesel Generator Testing (WBN-TS CNL-24-040, Tennessee Valley Authority - Central Emergency Control Center Emergency Plan Implementing Procedure Revisions2024-05-0808 May 2024 Tennessee Valley Authority - Central Emergency Control Center Emergency Plan Implementing Procedure Revisions CNL-24-037, Clinch River, Sequoyah, Units 1 and 2, Watts Bar, Unit 1 and 2, Nuclear Quality Assurance Plan, TVA-NQA-PLN89-A, Revision 422024-04-22022 April 2024 Clinch River, Sequoyah, Units 1 and 2, Watts Bar, Unit 1 and 2, Nuclear Quality Assurance Plan, TVA-NQA-PLN89-A, Revision 42 CNL-24-010, License Amendment Request to Recapture Low-Power Testing Time (WBN-TS-23-19)2024-04-17017 April 2024 License Amendment Request to Recapture Low-Power Testing Time (WBN-TS-23-19) CNL-24-024, Hydrologic Engineering Center River Analysis System Project Milestone Status Update2024-04-17017 April 2024 Hydrologic Engineering Center River Analysis System Project Milestone Status Update CNL-24-033, Central Emergency Control Center Emergency Plan Implementing Procedure Revisions2024-04-17017 April 2024 Central Emergency Control Center Emergency Plan Implementing Procedure Revisions CNL-24-017, Tennessee Valley Authority - Central Emergency Control Center Emergency Plan Implementing Procedure and Emergency Preparedness Department Procedure Revisions2024-01-17017 January 2024 Tennessee Valley Authority - Central Emergency Control Center Emergency Plan Implementing Procedure and Emergency Preparedness Department Procedure Revisions CNL-24-016, Supplement to Application to Modify the Technical Specification Surveillance Requirement 3.9.5.1 (WBN-TS-21-14)2024-01-10010 January 2024 Supplement to Application to Modify the Technical Specification Surveillance Requirement 3.9.5.1 (WBN-TS-21-14) CNL-23-052, Application to Adopt TSTF-427-A, Revision 2, Allowance for Non-Technical Specification Barrier Degradation on Supported System Operability2024-01-0909 January 2024 Application to Adopt TSTF-427-A, Revision 2, Allowance for Non-Technical Specification Barrier Degradation on Supported System Operability CNL-23-062, Application to Revise the Technical Specifications Section 3.8.2, AC Sources-Shutdown, to Remove Reference to the C-S Diesel Generator (WBN-TS-23-018)2024-01-0808 January 2024 Application to Revise the Technical Specifications Section 3.8.2, AC Sources-Shutdown, to Remove Reference to the C-S Diesel Generator (WBN-TS-23-018) CNL-23-069, Request for Exemption from Enhanced Weapons, Firearms Background Checks, and Security Event Notifications Implementation2023-12-21021 December 2023 Request for Exemption from Enhanced Weapons, Firearms Background Checks, and Security Event Notifications Implementation CNL-23-036, Application to Revise Function 5 of Technical Specification Table 3.3.2-1, Engineered Safety Feature Actuation System Instrumentation, for the Sequoyah Nuclear Plant and Watts Bar Nuclear Plant (SQN-TS-23-02 and WBN-TS-23-08)2023-12-18018 December 2023 Application to Revise Function 5 of Technical Specification Table 3.3.2-1, Engineered Safety Feature Actuation System Instrumentation, for the Sequoyah Nuclear Plant and Watts Bar Nuclear Plant (SQN-TS-23-02 and WBN-TS-23-08) CNL-23-001, Rebaseline of Sections 3.1 and 3.2 of the Technical Specifications (WBN-TS-23-01)2023-12-13013 December 2023 Rebaseline of Sections 3.1 and 3.2 of the Technical Specifications (WBN-TS-23-01) CNL-23-067, Central Emergency Control Center Emergency Plan Implementing Procedure Revisions2023-11-27027 November 2023 Central Emergency Control Center Emergency Plan Implementing Procedure Revisions CNL-23-059, Supplement to Sequoyah Nuclear Plant, Units 1 and 2, and Watts Bar Nuclear Plant, Units 1 and 2, Application to Revise Technical Specifications to Adopt TSTF-567-A, Revision 1, Add Containment Sump TS to Address GSI-191 Issues (SQN-TS-23-2023-09-20020 September 2023 Supplement to Sequoyah Nuclear Plant, Units 1 and 2, and Watts Bar Nuclear Plant, Units 1 and 2, Application to Revise Technical Specifications to Adopt TSTF-567-A, Revision 1, Add Containment Sump TS to Address GSI-191 Issues (SQN-TS-23-03 CNL-23-061, Tennessee Valley Authority - Central Emergency Control Center Emergency Plan Implementing Procedure Revision2023-09-20020 September 2023 Tennessee Valley Authority - Central Emergency Control Center Emergency Plan Implementing Procedure Revision CNL-23-057, Central Emergency Control Center Emergency Plan Implementing Procedure Revisions. Includes CECC-EPIP-1, Revision 76 and CECC-EPIP-9, Revision 642023-09-0505 September 2023 Central Emergency Control Center Emergency Plan Implementing Procedure Revisions. Includes CECC-EPIP-1, Revision 76 and CECC-EPIP-9, Revision 64 CNL-23-045, License Amendment Request to Revise Technical Specification Table 1.1-1 Regarding the Required Number of Fully Tensioned Reactor Pressure Vessel Head Closure Bolts (WBN-TS-23-010)2023-08-0707 August 2023 License Amendment Request to Revise Technical Specification Table 1.1-1 Regarding the Required Number of Fully Tensioned Reactor Pressure Vessel Head Closure Bolts (WBN-TS-23-010) CNL-23-028, Application to Revise Technical Specifications to Adopt TSTF-567-A, Revision 1, Add Containment Sump TS to Address GSI-191 Issues (SQN-TS-23-03 and WBN-TS-23-06)2023-08-0202 August 2023 Application to Revise Technical Specifications to Adopt TSTF-567-A, Revision 1, Add Containment Sump TS to Address GSI-191 Issues (SQN-TS-23-03 and WBN-TS-23-06) CNL-23-055, Tennessee Valley Authority - Emergency Preparedness Department Procedure Revision. Includes EPDP-3, Revision 3, Emergency Plan Exercises and Preparedness Drills2023-07-25025 July 2023 Tennessee Valley Authority - Emergency Preparedness Department Procedure Revision. Includes EPDP-3, Revision 3, Emergency Plan Exercises and Preparedness Drills CNL-23-053, Tennessee Valley Authority - Radiological Emergency Plan Revisions2023-07-18018 July 2023 Tennessee Valley Authority - Radiological Emergency Plan Revisions CNL-23-020, Application to Revise Technical Specifications to Adopt TSTF-501-A, Revision 1, Relocate Stored Fuel Oil and Lube Oil Volume Values to Licensee Control (WBN-TS-22-06)2023-06-28028 June 2023 Application to Revise Technical Specifications to Adopt TSTF-501-A, Revision 1, Relocate Stored Fuel Oil and Lube Oil Volume Values to Licensee Control (WBN-TS-22-06) CNL-23-049, Tennessee Valley Authority - Radiological Emergency Plan and Central Emergency Control Center Emergency Plan .2023-06-26026 June 2023 Tennessee Valley Authority - Radiological Emergency Plan and Central Emergency Control Center Emergency Plan . CNL-23-044, Transmittal of Revision 3 to WCAP-18774-P and WCAP-18774-NP, Addendum to the Rotterdam Dockyard Company Final Stress Report for 173 P.W.R. Vessels TVA III & IV (Report No. 30749-B-030, Rev. 3) - Evaluation of One Closure Stud Out2023-06-0101 June 2023 Transmittal of Revision 3 to WCAP-18774-P and WCAP-18774-NP, Addendum to the Rotterdam Dockyard Company Final Stress Report for 173 P.W.R. Vessels TVA III & IV (Report No. 30749-B-030, Rev. 3) - Evaluation of One Closure Stud Out CNL-23-042, Radiological Emergency Plan and Central Emergency Control Center Emergency Plan Implementing Procedure Revisions2023-05-16016 May 2023 Radiological Emergency Plan and Central Emergency Control Center Emergency Plan Implementing Procedure Revisions CNL-23-043, Emergency License Amendment Request to Relax the Required Number of Fully Tensioned Reactor Pressure Vessel Head Closure Studs in Technical Specification Table 1.1-1, Modes (WBN-TS-23-09)2023-05-0404 May 2023 Emergency License Amendment Request to Relax the Required Number of Fully Tensioned Reactor Pressure Vessel Head Closure Studs in Technical Specification Table 1.1-1, Modes (WBN-TS-23-09) CNL-23-030, Tennessee Valley Authority - Hydrologic Engineering Center River Analysis System Project Milestone Status Update2023-04-27027 April 2023 Tennessee Valley Authority - Hydrologic Engineering Center River Analysis System Project Milestone Status Update CNL-23-032, Clinch River, Sequoyah, Units 1 and 2, Watts Bar, Unit 1 and 2, Nuclear Quality Assurance Plan, TVA-NQA-PLN89-A, Revision 412023-04-27027 April 2023 Clinch River, Sequoyah, Units 1 and 2, Watts Bar, Unit 1 and 2, Nuclear Quality Assurance Plan, TVA-NQA-PLN89-A, Revision 41 CNL-23-033, Radiological Emergency Plan and Central Emergency Control Center Emergency Plan Implementing Procedure Revision2023-04-24024 April 2023 Radiological Emergency Plan and Central Emergency Control Center Emergency Plan Implementing Procedure Revision CNL-23-029, Radiological Emergency Plan and Central Emergency Control Center Emergency Plan Implementing Procedure Revisions2023-04-11011 April 2023 Radiological Emergency Plan and Central Emergency Control Center Emergency Plan Implementing Procedure Revisions CNL-23-023, Annual Insurance Status Report2023-03-30030 March 2023 Annual Insurance Status Report CNL-23-022, Decommissioning Funding Status Report2023-03-29029 March 2023 Decommissioning Funding Status Report CNL-23-024, TVA Guarantee of Payment of Deferred Premiums - 2022 Annual Report2023-03-29029 March 2023 TVA Guarantee of Payment of Deferred Premiums - 2022 Annual Report CNL-23-002, Application to Revise Watts Bar Nuclear Plant Units 1 and 2 Technical Specifications to Change the Number of Tritium Producing Burnable Absorber Rods (WBN-TS-21-02) and Proposed Revision to Reactor Vessel Surveillance Capsule Removal Sche2023-03-20020 March 2023 Application to Revise Watts Bar Nuclear Plant Units 1 and 2 Technical Specifications to Change the Number of Tritium Producing Burnable Absorber Rods (WBN-TS-21-02) and Proposed Revision to Reactor Vessel Surveillance Capsule Removal Schedu CNL-23-021, Radiological Emergency Plan and Central Emergency Control Center Emergency Plan Implementing Procedure Revisions2023-03-0808 March 2023 Radiological Emergency Plan and Central Emergency Control Center Emergency Plan Implementing Procedure Revisions CNL-23-015, Expedited Application to Modify the Watts Bar Nuclear Plant, Unit 1 and Unit 2 Technical Specifications for Main Control Room Chiller Completion Time Extension (WBN-TS-22-08)2023-02-27027 February 2023 Expedited Application to Modify the Watts Bar Nuclear Plant, Unit 1 and Unit 2 Technical Specifications for Main Control Room Chiller Completion Time Extension (WBN-TS-22-08) CNL-23-003, Browns Nuclear Plant, Units 1, 2 & 3; Sequoyah Nuclear Plant, Units 1 & 2; and Watts Bar Nuclear Plant, Units 1 & 2 - Organizational Topic Report TVA-NPOD89-A2023-01-30030 January 2023 Browns Nuclear Plant, Units 1, 2 & 3; Sequoyah Nuclear Plant, Units 1 & 2; and Watts Bar Nuclear Plant, Units 1 & 2 - Organizational Topic Report TVA-NPOD89-A CNL-23-009, Response to Request for Additional Information Request to Revise Technical Specification 3.4.122023-01-0404 January 2023 Response to Request for Additional Information Request to Revise Technical Specification 3.4.12 CNL-23-008, Tennessee Valley Authority Radiological Emergency Plan and Central Emergency Control Center Emergency Plan Implementing Procedure Revisions2022-12-22022 December 2022 Tennessee Valley Authority Radiological Emergency Plan and Central Emergency Control Center Emergency Plan Implementing Procedure Revisions CNL-22-109, Steam Generator Tube Inspection Report to Reflect TSTF-577 Reporting Requirements2022-12-22022 December 2022 Steam Generator Tube Inspection Report to Reflect TSTF-577 Reporting Requirements CNL-22-101, American Society of Mechanical Engineers Boiler and Pressure Vessel Code Section XI, Request for Alternative WBN-2-ISI-012022-11-28028 November 2022 American Society of Mechanical Engineers Boiler and Pressure Vessel Code Section XI, Request for Alternative WBN-2-ISI-01 CNL-22-106, Tennessee Valley Authority - Radiological Emergency Plan and Central Emergency Control Center Emergency Plan Implementing Procedure Revision. Includes CECC EPIP-14, Revision 38, Nuclear Emergency Public Information Organization and Operat2022-11-28028 November 2022 Tennessee Valley Authority - Radiological Emergency Plan and Central Emergency Control Center Emergency Plan Implementing Procedure Revision. Includes CECC EPIP-14, Revision 38, Nuclear Emergency Public Information Organization and Operatio CNL-22-099, Tennessee Valley Authority - Hydrologic Engineering Center River Analysis System Project Milestone Status Update2022-10-31031 October 2022 Tennessee Valley Authority - Hydrologic Engineering Center River Analysis System Project Milestone Status Update CNL-22-098, Tennessee Valley Authority - Radiological Emergency Plan and Central Emergency Control Center Emergency Plan Implementing Procedure Revisions. Includes CECC EPIP-7, Rev. 45 & CECC EPIP-8, Rev. 552022-10-17017 October 2022 Tennessee Valley Authority - Radiological Emergency Plan and Central Emergency Control Center Emergency Plan Implementing Procedure Revisions. Includes CECC EPIP-7, Rev. 45 & CECC EPIP-8, Rev. 55 CNL-22-085, Response to Request for Additional Information Regarding Sequoyah Nuclear Plant (Sqn), Units 1 and 2 and Watts Bar Nuclear Plant (Wbn), Units 1 and 2, American Society of Mechanical Engineers Operation and Maintenance Code, Request for Al2022-09-0202 September 2022 Response to Request for Additional Information Regarding Sequoyah Nuclear Plant (Sqn), Units 1 and 2 and Watts Bar Nuclear Plant (Wbn), Units 1 and 2, American Society of Mechanical Engineers Operation and Maintenance Code, Request for Alte CNL-22-077, Radiological Emergency Plan and Central Emergency Control Center Emergency Plan Implementing Procedure Revisions2022-08-11011 August 2022 Radiological Emergency Plan and Central Emergency Control Center Emergency Plan Implementing Procedure Revisions CNL-22-030, Application to Revise Technical Specification 3.4.12, Low Temperature Overpressure Protection System for Sequoyah Nuclear Plant (SQN-TSC-22-01) and TS 3.4.12 Cold Overpressure Mitigation System for Watts Bar Nuclear Plant (WBN-TS-22-03)2022-07-27027 July 2022 Application to Revise Technical Specification 3.4.12, Low Temperature Overpressure Protection System for Sequoyah Nuclear Plant (SQN-TSC-22-01) and TS 3.4.12 Cold Overpressure Mitigation System for Watts Bar Nuclear Plant (WBN-TS-22-03) CNL-22-039, Application to Revise Technical Specifications to Adopt TSTF-554-A, Revision 1, Revise Reactor Coolant Leakage Requirements (BFN TS-537) (SQN-21-05) (WBN-TS-21-04)2022-07-13013 July 2022 Application to Revise Technical Specifications to Adopt TSTF-554-A, Revision 1, Revise Reactor Coolant Leakage Requirements (BFN TS-537) (SQN-21-05) (WBN-TS-21-04) CNL-22-071, Supplement to Application to Revise Technical Specifications to Adopt TSTF-577, Revised Frequencies for Steam Generator Tube Inspections (SQN-TS-21-03 and WBN-TS-21-08)2022-07-13013 July 2022 Supplement to Application to Revise Technical Specifications to Adopt TSTF-577, Revised Frequencies for Steam Generator Tube Inspections (SQN-TS-21-03 and WBN-TS-21-08) 2024-06-27
[Table view] Category:Report
MONTHYEARWBL-24-022, Cycle 5 Steam Generator Tube Inspection Report2024-05-16016 May 2024 Cycle 5 Steam Generator Tube Inspection Report CNL-24-016, Supplement to Application to Modify the Technical Specification Surveillance Requirement 3.9.5.1 (WBN-TS-21-14)2024-01-10010 January 2024 Supplement to Application to Modify the Technical Specification Surveillance Requirement 3.9.5.1 (WBN-TS-21-14) ML23346A1382024-01-0303 January 2024 Regulatory Audit Summary Related to Request to Increase the Number of Tritium Producing Burnable Absorber Rods ML23192A4472023-07-31031 July 2023 Staff Assessment of Updated Seismic Hazards at TVA Sites Following the NRC Process for the Ongoing Assessment of Natural Hazards Information WBL-23-018, Revised Pressure and Temperature Limits Report (PTLR)2023-04-10010 April 2023 Revised Pressure and Temperature Limits Report (PTLR) CNL-23-002, Application to Revise Watts Bar Nuclear Plant Units 1 and 2 Technical Specifications to Change the Number of Tritium Producing Burnable Absorber Rods (WBN-TS-21-02) and Proposed Revision to Reactor Vessel Surveillance Capsule Removal Sche2023-03-20020 March 2023 Application to Revise Watts Bar Nuclear Plant Units 1 and 2 Technical Specifications to Change the Number of Tritium Producing Burnable Absorber Rods (WBN-TS-21-02) and Proposed Revision to Reactor Vessel Surveillance Capsule Removal Schedu WBL-22-034, Update to Fire Protection Report Figures Redacted2022-08-0101 August 2022 Update to Fire Protection Report Figures Redacted WBL-21-057, Cycle 4 Mid-Cycle Outage Generic Letter 95-05 Voltage-Based Alternate Repair Criteria Final Report2021-12-16016 December 2021 Cycle 4 Mid-Cycle Outage Generic Letter 95-05 Voltage-Based Alternate Repair Criteria Final Report CNL-21-018, Application to Adopt TSTF-205-A, Revision of Channel Calibration, Channel Functional Test, and Related Definitions, and TSTF-563-A, Revise Instrument Testing Definitions to Incorporate the Surveillance .2021-12-0909 December 2021 Application to Adopt TSTF-205-A, Revision of Channel Calibration, Channel Functional Test, and Related Definitions, and TSTF-563-A, Revise Instrument Testing Definitions to Incorporate the Surveillance . WBL-21-056, Revised Pressure and Temperature Limits Report (PTLR)2021-12-0909 December 2021 Revised Pressure and Temperature Limits Report (PTLR) ML21244A3452021-09-20020 September 2021 Proposed Alternative IST RR 9 to the Requirements of the ASME OM Code for Test Plan Group 6 Relief Valves CNL-21-055, Revision to Refueling Outage 3 Generic Letter 95-05 Voltage-Based Alternate Repair Criteria Final Report2021-07-20020 July 2021 Revision to Refueling Outage 3 Generic Letter 95-05 Voltage-Based Alternate Repair Criteria Final Report CNL-21-040, Supplement to Expedited Application for Approval to Use a Growth Rate Temperature Adjustment When Implementing the Generic Letter 95-05 Analysis for the (WBN TS-391-21-002)2021-03-23023 March 2021 Supplement to Expedited Application for Approval to Use a Growth Rate Temperature Adjustment When Implementing the Generic Letter 95-05 Analysis for the (WBN TS-391-21-002) ML21060A9132021-03-17017 March 2021 Final Environmental Assessment and Finding of No Significant Impact for Initial and Updated Decommissioning Funding Plans for Watts Bar ISFSI CNL-21-011, Expedited Application for Approval to Use a Growth Rate Temperature Adjustment When Implementing the Generic Letter 95-05 Analysis for the (WBN TS-391-21-002)2021-02-25025 February 2021 Expedited Application for Approval to Use a Growth Rate Temperature Adjustment When Implementing the Generic Letter 95-05 Analysis for the (WBN TS-391-21-002) WBL-21-006, Refueling Outage 3 Generic Letter 95-05 Voltage-Based Alternate Repair Criteria Final Report2021-02-11011 February 2021 Refueling Outage 3 Generic Letter 95-05 Voltage-Based Alternate Repair Criteria Final Report WBL-20-066, Revised Pressure and Temperature Limits Report (PTLR)2020-12-16016 December 2020 Revised Pressure and Temperature Limits Report (PTLR) CNL-20-102, 10 CFR 71.95 Report for 3-60B Casks User2020-12-16016 December 2020 10 CFR 71.95 Report for 3-60B Casks User CNL-20-074, Submittal of Additional Supplement to License Amendment Request for Measurement Uncertainty Recapture Power Uprate (WBN-TS-19-06)2020-08-28028 August 2020 Submittal of Additional Supplement to License Amendment Request for Measurement Uncertainty Recapture Power Uprate (WBN-TS-19-06) WBL-20-004, Analysis of Capsule U from Watts Bar Unit 2 Reactor Vessel Radiation Surveillance Program2020-04-16016 April 2020 Analysis of Capsule U from Watts Bar Unit 2 Reactor Vessel Radiation Surveillance Program CNL-19-082, License Amendment Request for Measurement Uncertainty Recapture Power Uprate (WBN-TS-19-06)2019-10-10010 October 2019 License Amendment Request for Measurement Uncertainty Recapture Power Uprate (WBN-TS-19-06) L-19-034, Wafts Bar Nuclear Plant, Unit 1 - Technical Specification (TS) 5.9.8 - Post Accident Monitoring System (Pams) Report2019-06-18018 June 2019 Wafts Bar Nuclear Plant, Unit 1 - Technical Specification (TS) 5.9.8 - Post Accident Monitoring System (Pams) Report L-19-026, Revised Pressure and Temperature Limits Report (PTLR)2019-04-0404 April 2019 Revised Pressure and Temperature Limits Report (PTLR) ML19039A0492019-02-0808 February 2019 Amd 1 to USAR Chapter 9 Auxiliary System NRC Additional Redactions ML19003A5692019-01-16016 January 2019 Review of the Fall 2017 Steam Generator Tube Inspection Report ML18242A0382018-08-30030 August 2018 Technical Specification (TS) 5.9.8 - Post Accident Monitoring System (Pams) Report CNL-18-092, Application to Revise the Technical Specifications to Adopt TSTF-266-A, Revision 3, Eliminate the Remote Shutdown System Table of Instrumentation and Controls (WBN-TS-18-02)2018-08-0101 August 2018 Application to Revise the Technical Specifications to Adopt TSTF-266-A, Revision 3, Eliminate the Remote Shutdown System Table of Instrumentation and Controls (WBN-TS-18-02) CNL-18-007, Seismic Probabilistic Risk Assessment Supplemental Information2018-04-10010 April 2018 Seismic Probabilistic Risk Assessment Supplemental Information ML17356A2692017-12-20020 December 2017 Construction Lessons Learned Report ML17313A1282017-11-0909 November 2017 Revised Pressure and Temperature Limits Report (PTLR) CNL-17-134, Transmittal of WCAP-18191-NP, Watts Bar Unit 2 Heatup and Cooldown Limit Curves for Normal Operation and Supplemental Reactor Vessel Integrity Evaluations2017-10-13013 October 2017 Transmittal of WCAP-18191-NP, Watts Bar Unit 2 Heatup and Cooldown Limit Curves for Normal Operation and Supplemental Reactor Vessel Integrity Evaluations ML17272A0192017-09-29029 September 2017 Technical Specification (TS) 5.9.8 - Post Accident Monitoring System (Pams) Report ML17263A1162017-09-20020 September 2017 Technical Specification (TS) 5.9.8 - Post Accident Monitoring System (Pams) Report ML17209A5542017-07-28028 July 2017 Cycle 14 Steam Generator Tube Inspection Report CNL-17-050, Submission of Technical Reports to Support a Public Meeting Regarding the Transition to the Reactor Oversight Process for the Mitigating Systems Performance Index2017-05-30030 May 2017 Submission of Technical Reports to Support a Public Meeting Regarding the Transition to the Reactor Oversight Process for the Mitigating Systems Performance Index ML16215A1042016-08-0202 August 2016 Technical Specification (TS) 5.9.8 - Post Accident Monitoring System Report ML16113A0202016-04-22022 April 2016 Submittal of Title 10, Code of Federal Regulations 50.59 Summary Report CNL-16-038, Application to Revise Technical Specification 4.2.1, Fuel Assemblies and Radioactive Waste System Design Basis Source Term Supplement to Response to NRC Request for Additional Information2016-03-31031 March 2016 Application to Revise Technical Specification 4.2.1, Fuel Assemblies and Radioactive Waste System Design Basis Source Term Supplement to Response to NRC Request for Additional Information CNL-16-034, TMI Task Action I.D.1 Commitment Closure Regarding Control Room Design Review Special Program2016-02-19019 February 2016 TMI Task Action I.D.1 Commitment Closure Regarding Control Room Design Review Special Program CNL-15-263, Application to Revise Technical Specification 4.2.1, Fuel Assemblies (WBN-TS-15-03) - Supplemental Information Related to the Onsite Regulatory Audit at Pacific Northwest National Laboratory2015-12-29029 December 2015 Application to Revise Technical Specification 4.2.1, Fuel Assemblies (WBN-TS-15-03) - Supplemental Information Related to the Onsite Regulatory Audit at Pacific Northwest National Laboratory CNL-15-204, Response to NRC Question Regarding the Revised Containment Analysis, Including Enclosure 3, WCAP-17834-NP, Revision 2, Wcobra/Trac Long Term LOCA M&E and Containment Integrity Analysis and Enclosure 4, Affidavit2015-09-21021 September 2015 Response to NRC Question Regarding the Revised Containment Analysis, Including Enclosure 3, WCAP-17834-NP, Revision 2, Wcobra/Trac Long Term LOCA M&E and Containment Integrity Analysis and Enclosure 4, Affidavit CNL-15-165, Submittal of Electromagnetic Interference (EMI) Survey Results2015-08-20020 August 2015 Submittal of Electromagnetic Interference (EMI) Survey Results CNL-15-143, the Tennessee Valley Authority (TVA) Nuclear Power Group Commercial Grade Dedication Recovery Project - Closure Report2015-07-31031 July 2015 the Tennessee Valley Authority (TVA) Nuclear Power Group Commercial Grade Dedication Recovery Project - Closure Report CNL-15-131, Individual Plant Examination of External Events (IPEEE) Report, Revision 32015-07-15015 July 2015 Individual Plant Examination of External Events (IPEEE) Report, Revision 3 CNL-15-106, 3, Sequoyah and Watts Bar, Units 1 & 2 - Provides Service List Update for Routine Information2015-07-0808 July 2015 3, Sequoyah and Watts Bar, Units 1 & 2 - Provides Service List Update for Routine Information CNL-15-097, Flood Hazard Reevaluation Report for Watts Bar Plant, Response to NRC Request for Information Per 10CFR50.54(f) Regarding Recommendations 2.1, 2.3 and 9.3 of the Near-Term Task Force Review of Insights from the Fukushima Dai-ichi Accident2015-06-16016 June 2015 Flood Hazard Reevaluation Report for Watts Bar Plant, Response to NRC Request for Information Per 10CFR50.54(f) Regarding Recommendations 2.1, 2.3 and 9.3 of the Near-Term Task Force Review of Insights from the Fukushima Dai-ichi Accident ML15121A6562015-05-0101 May 2015 NRC Region II - CIB1 Watts Bar 2 Ip&S 194 Additional Questions Request List CNL-15-039, Severe Accident Management Alternatives for Reactor Coolant Pump Seals2015-04-10010 April 2015 Severe Accident Management Alternatives for Reactor Coolant Pump Seals CNL-15-043, Flood Hazard Reevaluation Report for Watts Bar, Response to NRC Request for Information Per Title 10 of CFR 50.54(f) Regarding Recommendations 2.1, 2.3 and 9.3 of the Near-Term Task Force Review of Insights from the Fukushima Dai-ichi Acc2015-03-25025 March 2015 Flood Hazard Reevaluation Report for Watts Bar, Response to NRC Request for Information Per Title 10 of CFR 50.54(f) Regarding Recommendations 2.1, 2.3 and 9.3 of the Near-Term Task Force Review of Insights from the Fukushima Dai-ichi Accid ML15030A5082015-01-30030 January 2015 Tritium Production Program, Updated Plans for Cycle 13 Operation and Updated Evaluation of the Radiological Impacts of Tritium Permeation Into the Reactor Coolant System 2024-05-16
[Table view] Category:Miscellaneous
MONTHYEARML23346A1382024-01-0303 January 2024 Regulatory Audit Summary Related to Request to Increase the Number of Tritium Producing Burnable Absorber Rods WBL-22-034, Update to Fire Protection Report Figures Redacted2022-08-0101 August 2022 Update to Fire Protection Report Figures Redacted ML21244A3452021-09-20020 September 2021 Proposed Alternative IST RR 9 to the Requirements of the ASME OM Code for Test Plan Group 6 Relief Valves ML19003A5692019-01-16016 January 2019 Review of the Fall 2017 Steam Generator Tube Inspection Report ML18242A0382018-08-30030 August 2018 Technical Specification (TS) 5.9.8 - Post Accident Monitoring System (Pams) Report CNL-18-092, Application to Revise the Technical Specifications to Adopt TSTF-266-A, Revision 3, Eliminate the Remote Shutdown System Table of Instrumentation and Controls (WBN-TS-18-02)2018-08-0101 August 2018 Application to Revise the Technical Specifications to Adopt TSTF-266-A, Revision 3, Eliminate the Remote Shutdown System Table of Instrumentation and Controls (WBN-TS-18-02) CNL-18-007, Seismic Probabilistic Risk Assessment Supplemental Information2018-04-10010 April 2018 Seismic Probabilistic Risk Assessment Supplemental Information ML17313A1282017-11-0909 November 2017 Revised Pressure and Temperature Limits Report (PTLR) ML17272A0192017-09-29029 September 2017 Technical Specification (TS) 5.9.8 - Post Accident Monitoring System (Pams) Report ML17263A1162017-09-20020 September 2017 Technical Specification (TS) 5.9.8 - Post Accident Monitoring System (Pams) Report ML17209A5542017-07-28028 July 2017 Cycle 14 Steam Generator Tube Inspection Report ML16215A1042016-08-0202 August 2016 Technical Specification (TS) 5.9.8 - Post Accident Monitoring System Report ML16113A0202016-04-22022 April 2016 Submittal of Title 10, Code of Federal Regulations 50.59 Summary Report CNL-16-038, Application to Revise Technical Specification 4.2.1, Fuel Assemblies and Radioactive Waste System Design Basis Source Term Supplement to Response to NRC Request for Additional Information2016-03-31031 March 2016 Application to Revise Technical Specification 4.2.1, Fuel Assemblies and Radioactive Waste System Design Basis Source Term Supplement to Response to NRC Request for Additional Information CNL-16-034, TMI Task Action I.D.1 Commitment Closure Regarding Control Room Design Review Special Program2016-02-19019 February 2016 TMI Task Action I.D.1 Commitment Closure Regarding Control Room Design Review Special Program CNL-15-263, Application to Revise Technical Specification 4.2.1, Fuel Assemblies (WBN-TS-15-03) - Supplemental Information Related to the Onsite Regulatory Audit at Pacific Northwest National Laboratory2015-12-29029 December 2015 Application to Revise Technical Specification 4.2.1, Fuel Assemblies (WBN-TS-15-03) - Supplemental Information Related to the Onsite Regulatory Audit at Pacific Northwest National Laboratory CNL-15-165, Submittal of Electromagnetic Interference (EMI) Survey Results2015-08-20020 August 2015 Submittal of Electromagnetic Interference (EMI) Survey Results CNL-15-131, Individual Plant Examination of External Events (IPEEE) Report, Revision 32015-07-15015 July 2015 Individual Plant Examination of External Events (IPEEE) Report, Revision 3 CNL-15-106, 3, Sequoyah and Watts Bar, Units 1 & 2 - Provides Service List Update for Routine Information2015-07-0808 July 2015 3, Sequoyah and Watts Bar, Units 1 & 2 - Provides Service List Update for Routine Information ML15121A6562015-05-0101 May 2015 NRC Region II - CIB1 Watts Bar 2 Ip&S 194 Additional Questions Request List CNL-15-043, Flood Hazard Reevaluation Report for Watts Bar, Response to NRC Request for Information Per Title 10 of CFR 50.54(f) Regarding Recommendations 2.1, 2.3 and 9.3 of the Near-Term Task Force Review of Insights from the Fukushima Dai-ichi Acc2015-03-25025 March 2015 Flood Hazard Reevaluation Report for Watts Bar, Response to NRC Request for Information Per Title 10 of CFR 50.54(f) Regarding Recommendations 2.1, 2.3 and 9.3 of the Near-Term Task Force Review of Insights from the Fukushima Dai-ichi Accid CNL-14-212, Expedited Seismic Evaluation Process Report (CEUS Sites) Response to NRC Request for Information Pursuant to 10 CFR 50.54(f) Re Recommendation 2.1 of the Near-Term Task Force Review of Insights from Fukushima Dai-ichi Accident2014-12-30030 December 2014 Expedited Seismic Evaluation Process Report (CEUS Sites) Response to NRC Request for Information Pursuant to 10 CFR 50.54(f) Re Recommendation 2.1 of the Near-Term Task Force Review of Insights from Fukushima Dai-ichi Accident ML14163A6582014-09-18018 September 2014 Closeout of Generic Letter, 2004-02, Potential Impact of Debris Blockage on Emergency Recirculation During Design Basis Accidents at Pressurized-Water Reactors ML14212A6032014-07-31031 July 2014 WBRD-50-391/86-60 - Final Report and Revised Completion Schedule IR 05000391/19860602014-07-31031 July 2014 WBRD-50-391/86-60 - Final Report and Revised Completion Schedule ML14149A1502014-06-16016 June 2014 Staff Assessment of the Flooding Walkdown Report Supporting Implementation of Near-Term Task Force Recommendation 2.3 Related to the Fukushima Dai-Ichi Nuclear Power Plant Accident ML14133A5422014-05-23023 May 2014 Staff Assessment of the Seismic Walkdown Report Supporting Implementation of Near-Term Task Force Recommendation 2.3 Related to the Fukushima Dai-Ichi Nuclear Power Plant Accident ML13246A0222013-08-28028 August 2013 Submittal of Pre-op Test Instruction ML13178A2812013-06-26026 June 2013 10 CFR 50.59 Summary Report Supplement ML13144A5762013-05-22022 May 2013 Watt Bar, Units 1 & 2, Report of Drug Testing Error in Accordance with 10 CFR 26.719(c)(1) ML13126A2942013-04-29029 April 2013 10 CFR 50.59 Summary Report ML13121A0602013-04-29029 April 2013 Commitment Summary Report ML13080A3632013-03-18018 March 2013 Enclosure 3, Summer 2011 Compliance Survey for Watts Bar Nuclear Plant Outfall Passive Mixing Zone ML13080A3662013-03-18018 March 2013 Enclosure 1, Summer 2010 Compliance Survey for Watts Bar Nuclear Plant Outfall Passive Mixing Zone ML13080A0732013-03-12012 March 2013 Extension Request Regarding the Flooding Hazard Reevaluation Report Required by NRC Request for Information Pursuant to Title 10 of the Code of Federal Regulations 50.54(f) Regarding Recommendation.. ML13175A1352013-03-0505 March 2013 2-PTI-092-03, Revision 0, Nuclear Instrumentation Source Range Noise Checks During Hot Functional Testing. ML12356A3172012-12-17017 December 2012 Submittal of Pre-op Test Instruction, 2-PTI-063-06, Revision 0, Safety Injection System Check Valve Test. ML12335A3402012-11-27027 November 2012 Tennessee Valley Authority - Fleet Response to NRC Request for Information Pursuant to Title 10 of the Code of Federal Regulations 50.54(f) Regarding the Flooding Walkdown Results of Recommendation 2.3 of the Near-Term Task Force Review of ML13108A2842012-11-12012 November 2012 Unit 1, Fukushima Near-Term Task Force Recommendation 2.3: Seismic Response Report ML13175A1362012-11-0808 November 2012 2-PTI-099-06, Revision 0, Reactor Protection Setpoint Verification. ML13175A1342012-11-0101 November 2012 2-PTI-082-02, Revision 0, Rod Control - Non Hft. ML13175A1332012-10-22022 October 2012 2-PTI-085-01, Revision 0, Rod Control Functional Test. ML12236A1642012-07-19019 July 2012 Enclosure 1 Evaluation of Proposed Changes Tennessee Valley Authority Watts Bar Nuclear Plant, Unit 1 ML12223A1832012-03-29029 March 2012 Environmental Protection Agency 2012 - Facility Detail Report - Environmental Facts Warehouse Fii - Moccasin Bend Wwtp ML11362A0562011-12-20020 December 2011 Status of Regulatory Framework for the Completion of Construction and Licensing for Unit 2 - Revision 7 (TAC No. MD6311), and Status of Generic Communications for Unit 2 - Revision 7 ML11341A1572011-11-30030 November 2011 Attachments 7 Through 9, WNA-CN-00157-WBT-NP, Revision 1, CAW-11-3316, and WBT-D-3566 Np, Incore Instrument System Signal Processing System Isolation Requirement ML11326A2842011-11-18018 November 2011 Commitment Summary Report ML11257A0502011-08-31031 August 2011 Attachment 7, WCAP-17427-NP, Rev. 1, Watts Bar Nuclear Plant Unit 2 Common Q Post Accident Monitoring System Computer Security Assessment, Attachment 8, Application for Withholding Proprietary Information from Public Disclosure and Attachme ML1104003852011-02-0707 February 2011 Enclosure 2, Appendix a, Hydrothermal Effects on the Ichthyhoplankton from the Watts Bar Nuclear Plant Supplemental Condenser Cooling Water Outfall in Upper Chickamauga Reservoir ML1104003842011-02-0707 February 2011 Enclosure 1, Hydrothermal Effects on the Ichthyoplankton from Watts Bar Nuclear Plant Supplemental Condenser Cooling Water Outfall in Upper Chickamauga Reservoir 2024-01-03
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Text
Tennessee Valley Authority, 1101 Market Street, Chattanooga, Tennessee 37402 CNL-18-007 April 10, 2018 10 CFR 50.4 10 CFR 50.54(f)
ATTN: Document Control Desk U.S. Nuclear Regulatory Commission Washington, D.C. 20555-0001 Watts Bar Nuclear Plant, Units 1 and 2 Facility Operating License Nos. NPF-90 and NPF-96 NRC Docket Nos. 50-390 and 50-391
Subject:
TENNESSEE VALLEY AUTHORITY (TVA) - WATTS BAR NUCLEAR PLANT SEISMIC PROBABILISTIC RISK ASSESSMENT SUPPLEMENTAL INFORMATION
Reference:
TVA letter to NRC, Seismic Probabilistic Risk Assessment for Watts Bar Nuclear Plant, Units 1 and 2 - Response to NRC Request for Information Pursuant to Title 10 of the Code of Federal Regulations 50.54(f) Regarding Recommendation 2.1 of the Near-Term Task Force Review of Insights from the Fukushima Dai-ichi Accident, dated June 30, 2017 (ML17181A485)
In the Reference letter dated June 30, 2017, TVA provided the Seismic Probabilistic Risk Assessment Summary Report for Watts Bar Nuclear Plant, Units 1 and 2, as requested by NRCs letter dated October 27, 2015 (ML15194A015). The Enclosure to the Reference letter provided the information requested in Item (8)B of the 50.54(f) letter associated with NTTF Recommendation 2.1: Seismic.
The purpose of this letter is to provide supplemental information for the Seismic Probabilistic Risk Assessment Summary Report for Watts Bar Nuclear Plant, Units 1 and 2.
This letter contains no new regulatory commitments.
U.S. Nuclear Regulatory Commission CNL-18-007 Page 2 April 10, 2018 If you have any questions regarding this submittal, please contact Russell Thompson at (423) 751-2567.
I declare under penalty of perjury that the foregoing is true and correct. Executed on the 10th day of April 2018.
Vice resident, Nuclear Regulatory Affairs and Support Services
Enclosure:
Supplemental Information for the Watts Bar Nuclear Plant, Units 1 and 2, Seismic Probabilistic Risk Assessment in Response to 50.54(f) Letter with Regard to NTTF 2.1 Seismic Summary Report cc (Enclosure):
NRR Director - NRC Headquarters NRC Regional Administrator - Region II NRC Project Manager - Watts Bar Nuclear Plant NRC Senior Resident Inspector - Watts Bar Nuclear Plant
Enclosure Supplemental Information for the Watts Bar Nuclear Plant, Units 1 and 2, Seismic Probabilistic Risk Assessment in Response to 50.54(f) Letter with Regard to NTTF 2.1 Seismic Summary Report In response to the 10 CFR 50.54(f) letter issued by the NRC on March 12, 2012 [Ref 1], a Seismic Probabilistic Risk Assessment (SPRA) was developed for Watts Bar Nuclear Plant (WBN), Units 1 and 2. The WBN SPRA was submitted to NRC by letter dated June 30, 2017.
The SPRA shows that the point estimate seismic Core Damage Frequency (CDF) is 2.6x10-6 per reactor calendar year (rcy) for each unit. The seismic Large Early Release Frequency (LERF) is 1.7x10-6/rcy for each unit. Sensitivity studies were performed to identify critical assumptions, test the sensitivity to quantification parameters and the seismic hazard, and identify potential areas to consider for the reduction of seismic risk. These sensitivity studies demonstrated that the model results were robust to the modeling and assumptions used. No seismic hazard vulnerabilities were identified, and no plant actions have been taken or are planned given the insights from the seismic risk assessment.
The evaluation of the WBN SPRA relative to the 16 EPRI 1025287 (commonly referred to as the SPID) [Ref 2] requirement-related items as identified by the NRC Staff Review Guidance for Seismic PRA Submittal Technical Review Checklist (ML17041A342), demonstrate that the WBN SPRA is of sufficient quality and level of detail for the response to NTTF 2.1 Seismic.
Below are information summaries associated with the NRC Staff Review Checklist. The information is provided as a guide for the NRC Staff to aid in the location of various pieces of information. References noted are detailed at the end of this enclosure, and are the same as those provided in the WBN SPRA. Specific WBN SPRA Sections are referenced that provide information pertinent to the given topic.
Topic 1: Seismic Hazard (SPID Section 2.1, Section 2.2, and Section 2.3)
Prior to the NTTF 2.1 activities, a probabilistic seismic hazard analysis was initiated to support potential licensing efforts for WBN Unit 2. This analysis [Ref 19] was used for the WBN SPRA in lieu of the NTTF 2.1 submittal [Ref 3] since the site analysis develops the additional elements required for the SPRA, such as Foundation Input Response Spectra (FIRS),
hazard-consistent strain-compatible properties, and vertical ground motions. The guidance in the SPID was followed for developing the site analysis. The site analysis is described in WBN SPRA Section 3.1. Figure 3.3-1 of the SPRA compares the mean control point hazard curves at 1 Hz, 10Hz, and PGA for the seismic hazard submitted to the NRC for NTTF 2.1 seismic hazard analysis and the SPRA seismic hazard analysis and shows that, depending on the selected annual frequency of exceedance, the shape of the site analysis is either similar to or slightly above the hazard submitted to the NRC. Peer review findings related to the seismic hazard, except for one, were closed utilizing the process given in Appendix X of NEI 12-13.
The open peer review finding, related to screening out of the seismic hazards other than vibratory ground motion, was judged Technically Resolved - Open Documentation by the peer reviewers and thus does not affect the seismic hazard calculation.
Page 1 of 12
Enclosure Supplemental Information for the Watts Bar Nuclear Plant, Units 1 and 2, Seismic Probabilistic Risk Assessment in Response to 50.54(f) Letter with Regard to NTTF 2.1 Seismic Summary Report Topic 2: Site Seismic Response (SPID Section 2.4)
The site seismic response is provided in Reference 19. The guidance in the SPID was followed for developing the site analysis. The site analysis is described in WBN SPRA Section 3.1. Figure 3.3-2 of the WBN SPRA compares the GMRS submitted to the NRC for NTTF 2.1 with the site response used for the SPRA and shows that the site response is equivalent to or slightly above the response submitted to the NRC for NTTF 2.1. Peer review findings related to the site analysis were closed utilizing the process given in Appendix X of NEI 12-13.
Topic 3: Definition of the Control Point for the SSE-to-GMRS-Comparison Aspect of the Site Analysis (SPID Section 2.4.2)
The GMRS is defined at the Reactor Building foundation control point at a depth of 64 ft. below plant grade of 728 which corresponds to elevation 664 ft mean sea level. Information is provided in Section 3.2 of the WBN SPRA. The site-analysis for the SPRA and the Seismic Hazard Screening Report use the same control point.
Topic 4: Adequacy of the Structural Model (SPID Section 6.3.1)
New three-dimensional (3D) finite element models (FEM) were built for the Auxiliary-Control Building (ACB), Diesel Generator Building (DGB), Intake Pumping Station (IPS), and North Steam Valve Room (NSVR). The new FEM, which capture building torsion, out-of-plane floor response, and in-plane floor diaphragm stiffness, satisfy the criteria 1 through 7 in the SPID
[Ref 2] Section 6.3.1. The Reactor Building (RB) and the Refueling Water Storage Tank (RWST) use lumped-mass-stick-models. For the Reactor Building, a 3D FEM is used to calibrate the complex interior concrete structure portion of the model. For the refueling water storage tank the existing lumped-mass-stick-model (LMSM) is enhanced. For the Reactor Building and Refueling water storage tank the criteria 1 through 6 in the SPRA Section 6.3.1 is judged to be met. Information is provided in Section 4.3.3 of the WBN SPRA. Peer review findings related to the adequacy of structural modeling were closed utilizing the process given in Appendix X of NEI 12-13.
The 3D models developed incorporated the geometry, configuration, and dimensions of the structural components of the building, such as the foundation and floor slabs, walls, and openings with reference to the respective mid-planes. The models automatically incorporate coupling between horizontal directions and also coupling between vertical and horizontal directions. With the exception of the DGB, which was reanalyzed after peer team review using one combined structural model, the other buildings are analyzed using two different models in the horizontal and vertical directions. Horizontal models assume all concrete sections are cracked, whereas vertical models assume only floor and roof concrete elements are cracked.
Since 3D FEM analyses for these buildings reveal that the coupling is not significant (especially between the horizontal and the vertical directions), the combined SSI result from the horizontal and vertical models effectively reduces the out-of-plane bending and in-plane shear stiffness to about 50 percent, and preserves the axial stiffness to uncracked stiffness.
As stated above, the DGB was reanalyzed after peer review using one combined structural model.
Page 2 of 12
Enclosure Supplemental Information for the Watts Bar Nuclear Plant, Units 1 and 2, Seismic Probabilistic Risk Assessment in Response to 50.54(f) Letter with Regard to NTTF 2.1 Seismic Summary Report The RB and the RWST use the LMSMs reported in the FSAR. The RB includes three structural systems, Interior Concrete Structure (ICS), Steel Containment Vessel (SCV), and Shield Building (SB) supported by a common foundation mat. The use of LMSMs for the RWST, SB and the SCV is justified on the basis that these structures are relatively simple and symmetric.
The LMSM for the ICS is verified and validated by means of independent evaluations using FEMs. Both the horizontal and vertical models of the ICS are configured to capture the coupling between vertical and horizontal directions.
Criteria 4 in SPID Section 6.3.1 states that the number of nodal or dynamic degrees of freedom should be sufficient to represent significant structural modes. Cutoff frequencies are 50 Hz for the ACB, RB, and NSVR, 90 Hz for the IPS and RSWT, and 100 Hz for the DGB.
Structural finite element models are sufficiently discretized to capture modes higher than 20 Hz in all directions.
Topic 5: Use of Fixed-Based Dynamic Seismic Analysis of Structures for Sites Previously Defined as Rock (SPID Section 6.3.3)
The NSVR uses a fixed-based dynamic seismic analysis. Because the NSVR is a relatively small structure located on the side of the much larger RB, the seismic analysis of the NSVR assumes fixed-base foundation conditions subjected to the seismic motion of the RB at elevations below grade. The analysis method is described in Section 4.3.1 of the WBN SPRA.
Peer review findings related to the adequacy of structural modeling were closed utilizing the process given in Appendix X of NEI 12-13.
Topic 6: Use of Seismic Response Scaling (SPID Section 6.3.2)
Scaling of In-Structure-Response-Spectra (ISRS) to account for higher ground motion levels was not used for the WBN SPRA.
The fragilities of five nuclear steam supply system (NSSS) components, Reactor Vessel, NSSS Piping, Steam Generator, Reactor Coolant Pump, and Pressurizer, are based on scaling the reported components stresses, displacements, and support reactions. The scaling is based on the previously developed in-structure response spectra (SSE spectra), NSSS natural frequencies, mode shapes, and participation factors. Review level earthquake (RLE) seismic demand is obtained by scaling the reported stress due to the SSE by the ratio of the spectral accelerations at appropriate frequencies represented in the SSE spectra and the in-structure response spectra due to the RLE. Accordingly, seismic safety factors for NSSS components are calculated using the scaled demand and the capacity. Then, appropriate inelastic absorption factors are utilized to incorporate non-linear effects.
Topic 7: Use of New Response Analysis for Building Response, ISRS, and Fragilities The requirements for new analysis are found in the ASME/ANS SPRA [Ref 4] standard under high level requirement SFR-C. The site response is developed with appropriate building specific soil profiles that captures the uncertainty and variability in material dynamic properties as described in Sections 3.0 and 4.3 of the WBN SPRA. Peer review findings related to the adequacy of new response analysis were closed utilizing the process given in Appendix X of NEI 12-13.
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Enclosure Supplemental Information for the Watts Bar Nuclear Plant, Units 1 and 2, Seismic Probabilistic Risk Assessment in Response to 50.54(f) Letter with Regard to NTTF 2.1 Seismic Summary Report Topic 8: Screening by Capacity to Select SSCs for Seismic Fragility Analysis (SPID Section 6.4.3)
The approach to screening out rugged SSCs from the model followed the guidance in Section 6.4.3 of the SPID. Other component capacity screening was not performed for WBN because the WBN GMRS seismic hazard is considerably higher than the available tools for seismic capacity screening. The screening approach is documented in the WBN SPRA Sections 4.2 and 4.3.6. Peer review findings related to the adequacy of selection of SSCs for seismic fragility analysis were closed utilizing the process given in Appendix X of NEI 12-13.
Topic 9: Use of the CDFM/Hybrid Methodology for Fragility Analysis (SPID Section 6.4.1)
The CDFM/Hybrid methodology used for fragility analysis is documented in Section 4.3.7 of the WBN SPRA and meets the recommendations in Section 6.4.1 of the SPID [Ref 2].
Recommended values from Table 6-2 of the SPID [Ref 2] were used to develop full seismic fragility curves. Peer review findings related to the adequacy of CDFM/Hybrid Methodology for fragility analysis response analysis were closed utilizing the process given in Appendix X of NEI 12-13.
Topic 10: Capacities of SSCs Sensitive to High-Frequencies (SPID Section 6.4.2)
Devices of the type identified in EPRI Phase 2 testing (EPRI 3002002997, High Frequency Program, High Frequency Testing Summary, September 2014) as being potentially sensitive to high-frequency seismic motion were included and documented in Sections 4.1.1 and 4.1.2 of the WBN SPRA. Peer review findings related to capacities of SSCs sensitive to high-frequencies were closed utilizing the process given in Appendix X of NEI 12-13.
Topic 11: Capacities of Relays Sensitive to High-Frequencies (SPID Section 6.4.2)
Circuit analysis was relied on for screening relays and other components potentially sensitive to high-frequency vibratory motion. Circuit analysis was performed to identify relays that can potentially impact plant SSCs if chatter were to occur, and screen out the relay devices that do not pose a safety concern. The circuit analysis was performed in accordance with the requirements in the ASME/ANS SPRA Standard [Ref 4] and meets the SPID [Ref 2]
requirements, and is documented in Section 4.1.1 of the WBN SPRA. The circuit analysis resulted in most relay chatter scenarios screened from further evaluation based on no impact to component function. However, some relays did not screen from further evaluation. Further fragility calculations and evaluations were performed for the relays that did not screen. The uncertainty parameters (r and u) for the relays were based on SPID Table 6-2. Uncertainty parameters different from the recommended values in the SPID were not used. The further calculations and evaluations showed that the unscreened relays were acceptable because either the relay median ground acceleration (Am) exceeded the fragility cutoff used in the SPRA model or the relay is in a system was not credited in the SPRA model. No operator actions were used to resolve relay chatter. Peer review findings related to capacities of relays sensitive to high-frequencies were closed utilizing the process given in Appendix X of NEI 12-13.
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Enclosure Supplemental Information for the Watts Bar Nuclear Plant, Units 1 and 2, Seismic Probabilistic Risk Assessment in Response to 50.54(f) Letter with Regard to NTTF 2.1 Seismic Summary Report Topic 12: Selection of Dominant Risk Contributors that Require Fragility Analysis Using the Separation of Variables Methodology (SPID Section 6.4.1)
The CDFM methodology has been used in the seismic PRA for analysis of the bulk of the SSCs requiring seismic fragility analysis. Selected SSCs that are significant contributors to the seismic risk were evaluated using separation of variables approach. This is consistent with the requirement in SPID Section 6.4.1 [Ref 2] and documented in the WBN SPRA Section 4.3.8.
To account for differences between test response and required response spectra bandwidths in determination of component fragilities, both the test response spectra and required response spectra for a tested component are clipped in accordance with the methodology presented in Appendix Q of EPRI-NP-6041. The clipping is done prior to any comparisons of the spectra to ensure that only broad band comparisons are made when determining the limiting TRS/RRS ratio.
To account for appropriate reduction of test levels performed using single- or dual-axis testing in the determination of component fragilities, FMS factors from EPRI-NP-6041 App. Q (page Q-9) are used. Note that most of the tests are multi-axis but when single axis or dual axis testing is completed, the appropriate factor is used.
To account for multi-mode response in determination of component fragilities for components that were tested using narrowband excitation, component TRS/RRS ratios are determined over a frequency range of interest where damage to the component is expected to occur, usually at the dominant mode of the component. In instances where damage may be expected to occur at higher modes, the TRS/RRS ratio is calculated as the minimum ratio over the frequency range of interest.
In general, the fragilities consider seismic motion in all three directions. Each component is reviewed and if it is determined that a particular direction would be non-damaging to a component, then the fragility analysis focused on the directions of motion leading to potential damage. This is done to be as realistic as possible. As an example, Motor Control Centers are commonly installed in long rows within a building. Since all cabinets are bolted together, then the side to side natural frequency is much higher than the front to back (and much higher than the side to side natural frequency in a test report that investigated only one cabinet section). This natural frequency in the side-to-side direction is well above 20 Hz, so the seismic displacements are insignificant and judged to be non-damaging for component functionality (except that sensitive relays are specifically evaluated). In this scenario, TRS/RRS ratios are only investigated in the front to back and vertical directions. The minimum ratio in either of those two directions over the frequency range of interest is used to calculate the fragility.
The method used to select the dominant risk contributors on which SoV fragility calculations were performed as described below.
A component is designated as a significant contributor to the model if it has a Fussel Vesely value greater than 0.005. On an iterative basis, fragility components were refined until a subsequent iteration of the quantification showed no significant change in the number of components or rankings of the components on the top contributor list. The tables in the SPRA submittal are the result of multiple iterations.
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Enclosure Supplemental Information for the Watts Bar Nuclear Plant, Units 1 and 2, Seismic Probabilistic Risk Assessment in Response to 50.54(f) Letter with Regard to NTTF 2.1 Seismic Summary Report For all WBN building structures, the CDFM method was used for fragility calculations. For all components aside from those within the DGB, the CDFM methodology was used. A sensitivity study was performed that demonstrates, based on the soil and site conditions of the plant, that there is little difference in component fragility between the results of the detailed CDFM fragility calculations performed for WBN components, and the SoV Method. The sensitivity study investigated a top contributor relay logic panel, a representative switchgear cabinet, and a top contributor block wall. The sensitivity study concluded that the detailed CDFM fragility calculation method used at WBN resulted in realistic fragility parameters.
After the second-to-last quantification, more detailed analysis was performed for the DGB and all of its components, because (1) many of the components in the DGB were showing up as top contributors, (2) the WBN hazard range of interest was determined to be above the hazard range considered for the strain compatible soil properties used for SSI analysis of the DGB, and (3) the peer review team noted some anomalies in a DGB block wall calculation. The results of the DGB re-analysis using the SoV method showed that the capacities for all dominant contributors from the second-to-last quantification remained the same or slightly increased, with one exception being for N/S shear wall capacities, which decreased by up to 20%. However, these shear walls have such high capacity that the 20% strength decrease was not significant.
Prior to the final quantification, Flexible and Diverse Coping Strategies (FLEX) equipment was only evaluated through a sensitivity study. For F&O resolution, FLEX equipment was included in the model for the final quantification. The FLEX component fragilities were calculated using margin assessment from the Expedited Seismic Evaluation Process (ESEP) and therefore are conservative. To determine the significance of the conservative FLEX component fragilities, a sensitivity study was completed increasing FLEX fragility values by 20%. The results demonstrated that increase of these fragilities did not make a significant change to SCDF or SLERF (2% and 1% reductions, respectively) and support the conclusion that current fragility calculations are sufficient for these items.
In the final quantification, several new items came up to the top contributor list for SCDF and SFERF, and no explicit SoV analyses were performed for the majority of these items. Based on the level of rigor in the CDFM fragility calculations performed for WBN, and the results of the SoV sensitivity studies and DGB re-investigation, it is concluded that the current fragility calculations are sufficiently realistic and representative of as-installed plant conditions.
Topic 13: Evaluation of LERF (SPID Section 6.5.1)
The evaluation of LERF is judged to be consistent with section 6.5.1 and Table 6-3 of the SPID [Ref 2]. Sections 5.1 and 5.5 of the WBN SPRA detail the evaluation of LERF. Peer review findings related to the evaluation of LERF were closed utilizing the process given in Appendix X of NEI 12-13.
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Enclosure Supplemental Information for the Watts Bar Nuclear Plant, Units 1 and 2, Seismic Probabilistic Risk Assessment in Response to 50.54(f) Letter with Regard to NTTF 2.1 Seismic Summary Report Topic 14: Peer Review of the Seismic PRA, Accounting for NEI 12-13 (SPID Section 6.7)
The peer review of the seismic PRA meets the elements described in Section 6.7 of the SPID
[Ref 2]. Peer review findings, except for one, were closed utilizing the process given in Appendix X of NEI 12-13. The open peer review finding related to screening out seismic hazards other than vibratory ground motion was judged Technically Resolved - Open Documentation by the peer reviewers and thus is not significant to the SPRA conclusions for this review application.
Information is provided in Appendix A of the WBN SPRA. ASME / ANS-Sb-2013 was used by the SPRA peer review team and the independent assessment team.
Many of the internal events F&Os provided in Appendix A of the submittal have been subsequently closed through the closure review process, including 1-8, 3-1, 3-8, 5-1, 7-4, and 7-8. The closure review was performed in accordance with the process documented in Appendix X to NEI 05-04, as well as the requirements published in the ASME/ANS PRA Standard (RA-Sa-2009) and Regulatory Guide 1.200, Revision 2. The model that was reviewed in the closure review is the same model that was used as the basis for the WBN SPRA. Therefore, these F&Os have no impact on the WBN SPRA.
Internal events F&O 3-6 remains open because the state of knowledge (SOK) correlation was not applied in the ISLOCA calculation of valve failure probabilities (as required by QU-A3).
However, the SPRA model is not affected since similar components are grouped fragility groups that are completely correlated. In addition, common cause grouping of random failures for similar components is included WBN SPRA, which accounts for a majority of the effects of the SOK Correlation. Therefore, there is no impact on the WBN SPRA results.
No PRA upgrades, as defined per PRA Standard ASME/ANS-Sa-2009, were made between the time of the 2009 internal events PRA peer review and the version of the internal events PRA model which was used to develop the subsequent SPRA model. The table in Appendix A of the WBN SPRA states that for each F&O, there is no upgrade and no new method used to resolve the F&O. Therefore, a follow-on peer review was not required.
NEI 05-04/07-12/12-06 Appendix X describes the process by which an independent assessment team can close out F&Os. The NRC staff accepted this process, as modified by two conditions, by letter dated May 3, 2017 (ML17079A427). The independent assessment review team adhered to the requirements of Appendix X and the conditions in the NRC letter. The independent assessment team reviewed the documented finding closure basis prepared by TVA, including updated PRA models and documentation. Appendix C of the Finding Closure Review report provides a table that summarizes both TVAs input to the review team and the review teams assessment of adequacy of closure. This table includes a column labeled Independent Review Team Assessment in which the review teams basis for determining whether each finding was adequately addressed is stated, with references to the revised PRA documentation as appropriate. The same process was used to assess adequacy of resolution of each finding within the review teams scope, including the following findings associated with SRs SFR-A2, SFR-F1, and SFR-G2. SFR-A2 associated findings 23-5, 23-8, 24-15. SFR-F1 associated findings 22-6, 22-7, 23-14, 23-15, 23-21, 23-22, 24-10, 24-12, 24-14. SFR-G2 associated findings 23-3, 23-6, 23-7, 23-9, 23-11, 24-4, 24-7, 24-11, 24-13. The independent assessment team members with responsibility for fragility review conducted their review and agreed with plant response on the associated findings for SFR-A2, SFR-F1, and SFR-G2.
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Enclosure Supplemental Information for the Watts Bar Nuclear Plant, Units 1 and 2, Seismic Probabilistic Risk Assessment in Response to 50.54(f) Letter with Regard to NTTF 2.1 Seismic Summary Report Section 2.2.5 of the Finding Closure Review report notes that TVA did not identify any of the finding dispositions as representing PRA Upgrade, i.e., all were identified as PRA Maintenance, and further confirms that the review team did not identify any of the finding closure dispositions as upgrades.
Topic 15: Documentation of the Seismic PRA (SPID Section 6.8)
WBN SPRA Table 2.0-1 provides a cross-reference for 50.54(f) Enclosure SPRA Reporting and Table 2.0-2 provides a cross-reference for additional SPID Section 6.8 Reporting. The tables show how the documentation requirements of Section 6.8 of the SPID [Ref 2] are met.
Appendix A of the WBN SPRA provides the summary of the SPRA Peer Review and Assessment of the PRA Technical Adequacy.
Topic 16: Review of Plant Modifications and Licensee Actions, If Any There are no modifications necessary to achieve the appropriate risk profile as documented in Section 6 of the WBN SPRA. No seismic hazard vulnerabilities were identified, and no plant actions have been taken or are planned given the insights from the seismic risk assessment.
TVA reviewed the risk levels and found the risk level to be low and tolerable with respect to total plant risk, therefore no modifications were considered. Prior to SPRA development and in conjunction with the Expedited Seismic Evaluation Program, a program to evaluate the seismic margin was initiated. The purpose of the program was to conduct thorough review of the seismic IPEEE for possible increase in the seismic margin from a 0.30g review level earthquake (RLE) to a HCLPF capacity of at least 0.50g RLE. These efforts included reviewing possible modifications to increase capacities. In addition, the IPEEE was updated from a focused scope to a full-scope IPEEE. Through these efforts, Watts Bar improved the plant HCLPF seismic capacity to 0.50g RLE. Anchorage modifications were made to the 480V Shutdown Board Transformers to improve the HCLPF from 0.38g to 0.85g. No additional modifications were found to be required under this review. No single vulnerability has been shown to significantly impact CDF or LERF results, with the exception of the assumed fragility for Loss of Offsite Power. Therefore, no plant modifications were considered following the issuance of the WBN SPRA model. Any refinements in the fragility values of the top contributors to risk are not expected to significantly affect overall seismic risk. In addition, any improvements in the fragilities of the top contributors are likely to affect both CDF and LERF, so the ratio of CDF to LERF is expected to be the same.
For seismically induced failure of human actions due to instrumentation failures (SEIS_HRAINSTR), the individual instrumentation components were not explicitly modeled in the WBN SPRA. Instead, a bounding event, SEIS_HRAINSTR, was modeled with operator actions that represented failure of all instrumentation, simultaneously. The bounding instrument fragility was conservatively calculated considering the types of racks and panels on which they are mounted. Additionally, the control room ceiling and instrumentation power supplies were also considered in the calculation. Improving the fragility of one or more Auxiliary Instrument Room panels would not improve the overall instrument failure probability unless all panels were modified. Additionally, if the Am value for all panels were improved to greater than or equal to 2.08g, the main control room (MCR) ceiling becomes the bounding fragility. Similarly, improving the fragility of the MCR 120 Power Supply Distribution Panels will improve the overall instrument failure probability, up until it is equal to or greater than 2.09g which is the fragility of the MCR 120V Instrument Power Transformers.
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Enclosure Supplemental Information for the Watts Bar Nuclear Plant, Units 1 and 2, Seismic Probabilistic Risk Assessment in Response to 50.54(f) Letter with Regard to NTTF 2.1 Seismic Summary Report Seismically induced internal flooding (SEIS-IF), the fragility represents a failure of the bracing for the piping, and effectively represents failure of any piping brace that can lead to a modeled internal flooding event throughout the plant. This is a collection of several uncorrelated fragility events conservatively modeled to initiate all internal flooding events. If the bracing was improved at one or more locations, this event would have the same effect unless all pipe bracing was improved. Furthermore, the other failure modes or failures of other equipment would likely increase in importance if all pipe bracing was improved.
Breaker chatter for low voltage switchgear (LVS) and breaker chatter of medium voltage switchgear (MVS) are represented by SEIS_0-24 and SEIS_0-25, respectively. The failure of these events conservatively represents failure of all breakers in each group, respectively. The fragility calculations for these breakers have been highly refined and further evaluation is not expected to have a significant ability to reduce risk. It should be noted that these fragility groups predominantly contribute to CDF and LERF in the higher seismic bins (>1.2g) and further refinement would likely result in higher importance of other fragility groups with similar fragility values. No modifications that would significantly improve breaker fragilities have been identified.
From the review of the plant seismic margin prior to SPRA development and the review of the SPRA risk levels and top contributors, potential modifications that are easily completed to improve equipment capacities have been completed and no additional modifications were found to significantly improve plant seismic risk levels.
Sensitivities were performed on the SPRA results. Conservatisms and non-conservatisms in the model exist and are discussed as follows:
- 1. Complete seismic correlation of most fragility groups is a conservatism. However, a sensitivity analysis performed for this item revealed a maximum of a 16.8% reduction in SLERF if all groups are completely uncorrelated, and complete correlation is expected to be much closer to the actual result. Therefore, the level of conservatism introduced by excessive correlation is not deemed to be a significant factor.
- 2. The modeling of seismic impacts on evacuation may be conservative, given that all large seismic releases were treated as SLERF for events >0.5g. This effectively assumes that for all events greater than 0.5g, no evacuation of the surrounding area is possible.
However, a sensitivity study identified the maximum possible contribution from the treatment to be 5.1%, and since large seismic events are expected to significantly impact evacuation capabilities, the resulting skew in SLERF should again be significantly lower.
- 3. A fragility cutoff value of 3.5g was used, a potentially non-conservative approach. Thus, anything that had an Am>3.5g was screened from inclusion in the model, with the exception of some events directly tied to SIET (seismic initiating event tree) events.
Sensitivity showed a maximum contribution to SLERF from events greater than this of 2%, by introducing a direct core damage event at this fragility.
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Enclosure Supplemental Information for the Watts Bar Nuclear Plant, Units 1 and 2, Seismic Probabilistic Risk Assessment in Response to 50.54(f) Letter with Regard to NTTF 2.1 Seismic Summary Report
- 4. The lowest applicable failure mode fragility for a building was conservatively assumed to fail all components within that building. The only structure included in the SLERF importances was the NSVR, which was split into an A state and a C state to partially address this concern. This change resulted in a FV of the A mode failure of the NSVR of 0.008, which is above the screening limit of 0.005, but still represents less than a 1%
contribution to SLERF. In addition, though the turbine building did not appear on the importances directly, HFEs in the turbine building were considered failed due to inaccessibility, which failed FLTB1C and OPCMPA for all but bin 1. Note that this does not necessarily assume a guaranteed failure of SCCW piping leading to major flooding conditions in the turbine building due to the buildings collapse, a scenario which was analyzed to increase SLERF by 0.7% in another sensitivity study.
- 5. In general, components inside non-Seismic Category 1 structures are assumed to fail.
This is conservative, but a sensitivity study (which included all equipment in non-cat 1 structures, except for the SCCW piping failure described above in point 4) shows no significant impact on the model due to this assumption.
- 6. The PCS is assumed to fail. This approach is conservative and the sensitivity is examined in the SQU Notebook.
- 7. Seismically-rugged components (check valves, manual valves, strainers, and filters) are assumed to not fail for the seismic events modeled. Given that the worst case fragility values for these components were above the fragility cutoff mentioned in point 3 above, this does not introduce further non-realism.
- 8. Assuming complete seismic correlation for ruptures in the Main Steam lines most likely results in those events resulting in direct core damage rather than being transferred to the SSBI and SSBO event trees. Modeling this event in the SSBI and SSBO event trees (which do not automatically lead to core damage) is potentially non-conservative.
Sensitivity study performed showed essentially no impact due to this modeling.
- 9. The baseline SPRA modeling generally does not include > 24-h mission potential impacts such as additional room cooling, coolant makeup, steam supply, and 72-h mission for failure to run. This approach is non-conservative with respect to the SPRA base case results. Sensitivity study indicated a potential increase in SLERF of 0.7% due to this change.
- 10. Seismically induced failure of the control room ceiling (as a bounding condition to control room inoperability) in conjunction with failure of the operators to shut down the plant remotely is assumed to lead to direct SLERF. This is a conservative assumption, but the CR ceiling event does not appear in the SLERF importances.
- 11. Sub-components that were mounted on other SEL equipment were considered boxed by the larger component, e.g. control switches boxed on a board or limit switches boxed on a valve. In cases where all components can contribute to the overall loss of the larger component, only the governing fragility was used. This excludes chatter sensitive components (relays). Any chatter sensitive component is analyzed separately, independent of the larger component and will have a separate event in the SPRA model.
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Enclosure Supplemental Information for the Watts Bar Nuclear Plant, Units 1 and 2, Seismic Probabilistic Risk Assessment in Response to 50.54(f) Letter with Regard to NTTF 2.1 Seismic Summary Report
- 12. EPRI HRA Bin 4 (seismic bins 7 and 8) assume that all operator actions are unsuccessful due mainly to lack of HRA instrumentation. This includes seismic events in excess of 2.0g. This may potentially have the effect of raising HFE FVs and lower HFE RAWs, but the potential impact of this event is very minimal given the other large probabilities throughout these bins.
- 13. HFEs involving main feedwater and fire water alignment were set to 1.0 for all bins, since those systems were set to guaranteed failure in the model (see point 5 above).
- 14. For the relay chatter analysis, chatter was assumed to create the most undesirable combination of individual contact pair chattering. Given that no relay chatter events were ultimately included in the SPRA, this did not impact the SLERF results.
The approach for determining the importance measures in the SPRA in the context of the binning is discussed as follows. Per the standard, seismic events were split into 8 bins, each with a representative fragility. For importances, the resulting cutsets were modified from CCDP/CLERP and into CDF/LERF cutsets, merged, and then analyzed using the ACUBE importances tool. There is no expected non-realism introduced into any of these importance measures with the exception of the HRA RAW results. Due to the HRA combination method, it is extremely difficult to develop a precise number. Therefore, the HRA RAW events were generated using a conservative process. Specifically, all HRA combo and recovery events that related to an HFE were set to TRUE (rather than just converted to the corresponding combo event that did not contain the HFE). Since this did not result in any risk-significant RAW HFEs, this method was not refined.
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Enclosure Supplemental Information for the Watts Bar Nuclear Plant, Units 1 and 2, Seismic Probabilistic Risk Assessment in Response to 50.54(f) Letter with Regard to NTTF 2.1 Seismic Summary Report References (as listed in WBN SPRA Section 7.0)
- 1) NRC (E Leeds and M Johnson) Letter to All Power Reactor Licensees et al., Request for Information Pursuant to Title 10 of the Code of Federal Regulations 50.54(f)
Regarding Recommendations 2.1, 2.3 and 9.3 of the Near-Term Task Force Review of Insights from the Fukushima Dai-ichi Accident, March 12, 2012, ML12056A046.
- 2) EPRI 1025287, Seismic Evaluation Guidance: Screening, Prioritization and Implementation Details (SPID) for the Resolution of Fukushima Near-Term Task Force Recommendation 2.1: Seismic. Electric Power Research Institute, Palo Alto, CA:
February 2013.
- 3) TVA, (Shea) Letter to NRC, Tennessee Valley Authoritys Seismic Hazard and Screening Report (CEUS Sites), Response to NRC Request for Information Pursuant to 10CFR50.54(f) Regarding Recommendation 2.1 of the Near-Term Task Force Review of Insights from the Fukushima Dai-ichi Accident, March 31, 2014, ML14098A478.
- 4) ASME/ANS RA-S-2008, Standard for Level 1/Large Early Release Frequency Probabilistic Risk Assessment for Nuclear Power Plant Applications, including Addenda B, 2013, American Society of Mechanical Engineers, New York, September 30, 2013.
- 19) Seismic Hazard Analysis - Seismic Probabilistic Risk Assessment, TVA Calculation CDN0000002015000739, Rev 2.
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