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EPID:L-2022-LLE-0027, Registration of Spent Fuel Storage Cask Pursuant to 10 CFR 72.212(b)(2) (Approved, Closed) |
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MONTHYEARML22059B0612022-02-24024 February 2022 Registration of Spent Fuel Storage Cask Pursuant to 10 CFR 72.212(b)(2) Project stage: Request ML22216A0782022-08-0404 August 2022 Request for Exemption from Various 10 CFR Part 72 Regulations Resulting from Non-Destruction-Examination Compliance Project stage: Request ML22318A1482022-11-18018 November 2022 Letter - Request for Supplemental Information Pertaining to Sequoyah ISFSI August 4, 2022 Exemption Request A148 Project stage: Other ML22318A1472022-11-18018 November 2022 Request for Supplemental Information Pertaining to Sequoyah ISFSI August 4, 2022 Exemption Request Project stage: Request ML22318A1492022-11-18018 November 2022 Enclosure- Request for Supplemental Information A149 Project stage: Other ML22346A2732022-12-12012 December 2022 International - Attachment 2 - Affidavit of Kimberly Manzione Project stage: Other ML22346A2712022-12-12012 December 2022 International - Supporting Document for Sequoyah Nuclear Plant ISFSI Exemption Request Project stage: Other ML22353A0662022-12-19019 December 2022 Response to Request for Supplemental Information (D-RSI) Request for Exemption from Non-Destruction-Examination Compliance Project stage: Supplement ML23047A2952023-02-0808 February 2023 Conversation Record - Clarification Call with TVA Regarding Its 080422 Exemption Request for Sequoyah Nuclear Plant Independent Spent Fuel Storage Installation Project stage: Approval ML23117A1162023-04-27027 April 2023 Response to Follow-up Questions to TVAs Response to NRCs Request for Supplemental Information (Rsi) Related to August 4, 2022, Sequoyah ISFSI Exemption Request Project stage: Supplement ML23307A0822023-11-0808 November 2023 Request for Additional Information August 4, 2022, Exemption Request for Deviating from the Conditions of Certificate of Compliance No. 1032, Amendment No. 3, Related to Sequoyah Nuclear Plant Independent Spent Fuel Storage Installation Project stage: RAI ML24004A0402024-01-0303 January 2024 Response to NRCs November 8, 2023, Request for Additional Information - Related to Independent Spent Fuel Storage Installation Project stage: Request 2022-08-04
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Category:Letter
MONTHYEARML24032A0202024-01-31031 January 2024 NPDES Biocide/Corrosion Treatment Plan Annual Report, Cy 2023 ML23319A2452024-01-29029 January 2024 Issuance of Amendment Nos. 366 and 360; 164 and 71 Regarding the Adoption of TSTF-567, Revision 1, Add Containment Sump TS to Address GSI-191 Issues 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 ML24018A0142024-01-17017 January 2024 Engine Systems, Inc., Report No. 10CFR21-0137, Rev. 1, 56913-EN 56913 ML24011A3182024-01-11011 January 2024 Submittal of Discharge Monitoring Report (Dmr), October 2023 ML24011A3172024-01-11011 January 2024 Submittal of Discharge Monitoring Report (Dmr), September 2023 ML24011A3202024-01-11011 January 2024 Submittal of Discharge Monitoring Report (Dmr), December 2023 ML24011A3162024-01-11011 January 2024 Submittal of Discharge Monitoring Report (Dmr), August 2023 ML24011A3192024-01-11011 January 2024 Submittal of Discharge Monitoring Report (Dmr), November 2023 IR 05000327/20234422024-01-11011 January 2024 95001 Supplemental Inspection Report 05000327/2023442 and 05000328/2023442 and Follow-Up Assessment Letter ML24010A2132024-01-10010 January 2024 CFR 21.21 Final Report Regarding Siemens Medium Voltage Circuit Breakers ML24018A0952024-01-0404 January 2024 Engine Systems, Inc., 10CFR21 Reporting of Defects and Non-Compliance Report No. 10CFR21-0137, Rev. 0 ML24004A0332024-01-0303 January 2024 Interim Report of a Deviation or Failure to Comply Crompton Instruments Type 077 Ammeter ML24004A0402024-01-0303 January 2024 Response to NRCs November 8, 2023, Request for Additional Information - Related to Independent Spent Fuel Storage Installation CNL-23-068, 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) ML23346A1222023-12-12012 December 2023 Annual Non-Radiological Environmental Operating Report - 2023 IR 05000327/20234202023-11-28028 November 2023 Security Baseline Inspection Report 05000327/2023420 and 05000328/2023420 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 ML23324A4362023-11-0909 November 2023 Exam Corporate Notification Letter Aka 210-day Letter ML23307A0822023-11-0808 November 2023 Request for Additional Information August 4, 2022, Exemption Request for Deviating from the Conditions of Certificate of Compliance No. 1032, Amendment No. 3, Related to Sequoyah Nuclear Plant Independent Spent Fuel Storage Installation IR 05000327/20230032023-11-0303 November 2023 Integrated Inspection Report 05000327/2023003 and 05000328/2023003 ML23306A1592023-11-0202 November 2023 Enforcement Action EA-22-129 Inspection Readiness Notification ML23292A0792023-10-19019 October 2023 Tennessee Valley Authority - Emergency Plan Implementing Procedure Revision, Includes EPIP-5, Revision 58, General Emergency IR 05000327/20230112023-10-16016 October 2023 Triennial Fire Protection Inspection Report 05000327/2023011 and 05000328/2023011 ML23285A0882023-10-12012 October 2023 Submittal of Sequoyah Nuclear Plant, Units 1 and 2, Submittal of Updated Final Safety Analysis Report Amendment 31 ML23284A4252023-10-11011 October 2023 10 CFR 50.59 and 10 CFR 72.48 Changes, Tests, and Experiments Summary Report; Commitment Summary Report; and Update to the Fire Protection Report ML23283A2792023-10-10010 October 2023 Revisions to the Sequoyah Nuclear Plant Units 1 and 2 Technical Requirements Manual ML23279A0612023-10-0505 October 2023 Paragon Energy Solutions LLC, Part 21 Final Report Re Potential Defect with Eaton Jd and Hjd Series Molded Case Circuit Breakers (Mccbs) ML23277A0462023-10-0404 October 2023 Revisions to the Sequoyah Nuclear Plant Units 1 and 2 Technical Specification Bases ML23275A0272023-09-29029 September 2023 Submittal of Discharge Monitoring Report (DMR) Quality Assurance Study 43 Final Report 2023 ML23271A1662023-09-28028 September 2023 Registration of Spent Fuel Storage Cask Pursuant to 10 CFR 72.212(b)(2) 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 IR 05000327/20234032023-09-14014 September 2023 Cyber Security Inspection Report 05000327/2023403 and 05000328/2023403 (Cover Letter) ML23257A0062023-09-14014 September 2023 Enforcement Action EA-22-129 Inspection Postponement Request ML23254A2192023-09-11011 September 2023 Emergency Plan Implementing Procedure Revisions ML23254A0652023-09-0707 September 2023 Registration of Spent Fuel Storage Cask Pursuant to 10 CFR 72.212(b)(2) 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 IR 05000327/20230052023-08-29029 August 2023 Updated Inspection Plan for Sequoyah Nuclear Plant, Units 1 and 2 - Report 05000327/2023005 and 05000328/2023005 ML23233A0122023-08-17017 August 2023 Unit 1 Cycle 25 Refueling Outage - 90-Day Inservice Inspection Summary Report - Supplement ML23233A0142023-08-15015 August 2023 Discharge Monitoring Report (Dmr), July 2023 ML23215A1212023-08-0303 August 2023 301 Exam Administrative Items (2B) Normal Release ML23215A1572023-08-0303 August 2023 Enforcement Action EA-22-129 Inspection Readiness Notification 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) 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 2024-01-04
[Table view] Category:Response to Request for Additional Information (RAI)
MONTHYEARML24004A0402024-01-0303 January 2024 Response to NRCs November 8, 2023, Request for Additional Information - Related to Independent Spent Fuel Storage Installation ML23117A1162023-04-27027 April 2023 Response to Follow-up Questions to TVAs Response to NRCs Request for Supplemental Information (Rsi) Related to August 4, 2022, Sequoyah ISFSI Exemption Request 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 ML22353A0662022-12-19019 December 2022 Response to Request for Supplemental Information (D-RSI) Request for Exemption from Non-Destruction-Examination Compliance CNL-22-084, Response to Request for Additional Information Regarding License Amendment Request to License Amendment Request to Modify Approved 10 CFR 50.69 Categorization Process (SQN-TS-21-07)2022-09-16016 September 2022 Response to Request for Additional Information Regarding License Amendment Request to License Amendment Request to Modify Approved 10 CFR 50.69 Categorization Process (SQN-TS-21-07) 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-22-053, Response to Request for Additional Information Regarding Application to Revise Sequoyah Nuclear Plant Units 1 and 2 Updated Final Safety Analysis Report Regarding Changes to Hydrologic Analysis - Second Partial Response to Additional Requ2022-08-22022 August 2022 Response to Request for Additional Information Regarding Application to Revise Sequoyah Nuclear Plant Units 1 and 2 Updated Final Safety Analysis Report Regarding Changes to Hydrologic Analysis - Second Partial Response to Additional Reques CNL-22-069, Response to Request for Additional Information Regarding License Amendment Request to Revise Technical Specifications to Adopt Risk-Informed Completion Times TSTF-505, Revision 2, Provide Risk-Informed Extended Completion Times - Ritstf.2022-07-0101 July 2022 Response to Request for Additional Information Regarding License Amendment Request to Revise Technical Specifications to Adopt Risk-Informed Completion Times TSTF-505, Revision 2, Provide Risk-Informed Extended Completion Times - Ritstf. CNL-22-062, Response to Request for Additional Information Regarding American Society of Mechanical Engineers Operation and Maintenance Code, Request for Alternative RV-022022-06-28028 June 2022 Response to Request for Additional Information Regarding American Society of Mechanical Engineers Operation and Maintenance Code, Request for Alternative RV-02 CNL-22-064, Response to Request for Additional Information Regarding Tennessee Valley Authority License Amendment Request to Revise Emergency Plan Implementing Procedure Regarding Seismic Event Emergency Action Level Change2022-06-0909 June 2022 Response to Request for Additional Information Regarding Tennessee Valley Authority License Amendment Request to Revise Emergency Plan Implementing Procedure Regarding Seismic Event Emergency Action Level Change CNL-21-024, Partial Response to Additional Request for Information Re Application to Revise Updated Final Safety Analysis Report Re Changes to Hydrologic Analysis (TS-19-02)2021-06-15015 June 2021 Partial Response to Additional Request for Information Re Application to Revise Updated Final Safety Analysis Report Re Changes to Hydrologic Analysis (TS-19-02) CNL-21-039, Response to Request for Additional Information Re Application to Modify Technical Specification to Allow for Transition to Westinghouse RFA-2 Fuel (SQN-TS-20-09)2021-05-0505 May 2021 Response to Request for Additional Information Re Application to Modify Technical Specification to Allow for Transition to Westinghouse RFA-2 Fuel (SQN-TS-20-09) CNL-20-082, Partial Response to Request for Additional Information Regarding Application to Revise Updated Final Safety Analysis Report Regarding Changes to Hydrologic Analysis (TS-19-02)2020-11-10010 November 2020 Partial Response to Request for Additional Information Regarding Application to Revise Updated Final Safety Analysis Report Regarding Changes to Hydrologic Analysis (TS-19-02) CNL-20-076, Response to Request for Additional Information Re Application to Revise Sequoyah Nuclear Plant (SQN) Unit 1 Technical Specifications for Steam Generator Tube Inspection Frequency (SQN-TS-20-01)2020-09-23023 September 2020 Response to Request for Additional Information Re Application to Revise Sequoyah Nuclear Plant (SQN) Unit 1 Technical Specifications for Steam Generator Tube Inspection Frequency (SQN-TS-20-01) CNL-20-057, Tennessee Valley Authority - Application to Revise Sequoyah Nuclear Plant Units 1 & 2 Updated FSAR Changes to Hydrologic Analysis - Partial Response to Additional Request for Additional Information (TS-19-02)2020-08-12012 August 2020 Tennessee Valley Authority - Application to Revise Sequoyah Nuclear Plant Units 1 & 2 Updated FSAR Changes to Hydrologic Analysis - Partial Response to Additional Request for Additional Information (TS-19-02) ML20196L6892020-07-16016 July 2020 TVA Non-Proprietary Slides for Open Session of July 16, 2020, Partially Closed Public Meeting to Discuss Sequoyah License Amendment Request CNL-20-032, Response to Request for Additional Information (TS-19-02) Application to Revise Updated Final Safety Analysis Report Regarding Changes to Hydrologic Analysis2020-05-14014 May 2020 Response to Request for Additional Information (TS-19-02) Application to Revise Updated Final Safety Analysis Report Regarding Changes to Hydrologic Analysis CNL-20-026, Supplement to Application to Revise Sequoyah Nuclear Plant Units 1 and 2 Updated Final Safety Analysis Report Regarding Changes to Hydrologic Analysis, (TS-19-02)2020-02-18018 February 2020 Supplement to Application to Revise Sequoyah Nuclear Plant Units 1 and 2 Updated Final Safety Analysis Report Regarding Changes to Hydrologic Analysis, (TS-19-02) CNL-19-119, Response to Request for Additional Information Regarding Sequoyah Nuclear Plant, Unit 1 Exigent License Amendment Request to Revise Technical Specification 4.2.2, Control Rod Assemblies (SQN-TS-19-05)2019-11-19019 November 2019 Response to Request for Additional Information Regarding Sequoyah Nuclear Plant, Unit 1 Exigent License Amendment Request to Revise Technical Specification 4.2.2, Control Rod Assemblies (SQN-TS-19-05) CNL-19-061, Seismic Probabilistic Risk Assessment for Sequoyah 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 Task2019-10-18018 October 2019 Seismic Probabilistic Risk Assessment for Sequoyah 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 Fo ML19231A2082019-08-19019 August 2019 Response to Request for Information Regarding Unit 2 Cycle 22 - 180-Day Steam Generator Tube Inspection Report CNL-19-068, Response to Request for Additional Information Regarding the 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 for2019-07-22022 July 2019 Response to Request for Additional Information Regarding the 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 .. CNL-19-057, Response to Request for Additional Information Regarding American Society of Mechanical Engineers Boiler and Pressure Vessel Code Section XI, Inservice Inspection Program, Request for Alternative, 18-ISl-12019-06-19019 June 2019 Response to Request for Additional Information Regarding American Society of Mechanical Engineers Boiler and Pressure Vessel Code Section XI, Inservice Inspection Program, Request for Alternative, 18-ISl-1 CNL-19-002, Response to Request for Additional Information Regarding Application to Modify Sequoyah Nuclear Plant Units 1 and 2, Application to Adopt 10 CFR 50.69, Risk-Informed Categorization and Treatment of Structures,.2019-03-21021 March 2019 Response to Request for Additional Information Regarding Application to Modify Sequoyah Nuclear Plant Units 1 and 2, Application to Adopt 10 CFR 50.69, Risk-Informed Categorization and Treatment of Structures,. CNL-19-016, Response to Request for Additional Information on License Amendment Request to Modify Essential Raw Cooling Water Motor Control Center Breakers and to Revise the Updated Final Safety Analysis Report2019-01-30030 January 2019 Response to Request for Additional Information on License Amendment Request to Modify Essential Raw Cooling Water Motor Control Center Breakers and to Revise the Updated Final Safety Analysis Report CNL-18-089, Response to Request for Additional Information Regarding Decommissioning Funding Plan Update for Browns Ferry Nuclear Plant and Sequoyah Nuclear Plant Independent Spent Fuel Storage Installations, Docket Nos. 72-052 and 72-0342018-07-23023 July 2018 Response to Request for Additional Information Regarding Decommissioning Funding Plan Update for Browns Ferry Nuclear Plant and Sequoyah Nuclear Plant Independent Spent Fuel Storage Installations, Docket Nos. 72-052 and 72-034 NL-18-044, Browns Ferry Units 1, 2, and 3; Sequoyah Units 1 and 2; Watts Bar Units 1 and 2 - Tennessee Valley Authority Response to NRC Request for Additional Information Related to Topical Report TVA-NPG-AWA16, TVA Overall Basin Probable Maximum Pre2018-04-19019 April 2018 Browns Ferry Units 1, 2, and 3; Sequoyah Units 1 and 2; Watts Bar Units 1 and 2 - Tennessee Valley Authority Response to NRC Request for Additional Information Related to Topical Report TVA-NPG-AWA16, TVA Overall Basin Probable Maximum Prec CNL-18-044, Tennessee Valley Authority Response to NRC Request for Additional Information Related to Topical Report TVA-NPG-AWA16, TVA Overall Basin Probable Maximum Precipitation.2018-04-19019 April 2018 Tennessee Valley Authority Response to NRC Request for Additional Information Related to Topical Report TVA-NPG-AWA16, TVA Overall Basin Probable Maximum Precipitation. CNL-17-063, Response to Request for Additional Information (RAI) Regarding Application to Modify Technical Specifications Regarding Diesel Generator Steady State Frequency (SQN-TS-14-02)2017-10-26026 October 2017 Response to Request for Additional Information (RAI) Regarding Application to Modify Technical Specifications Regarding Diesel Generator Steady State Frequency (SQN-TS-14-02) CNL-17-104, Response to Request for Additional Information Regarding Proposed Technical Specification Change to Revise the Note Modifying SR 3.8.1.17 of TS 3.8.1 AC Sources - Operating (TS-SQN-16-04)2017-08-0707 August 2017 Response to Request for Additional Information Regarding Proposed Technical Specification Change to Revise the Note Modifying SR 3.8.1.17 of TS 3.8.1 AC Sources - Operating (TS-SQN-16-04) CNL-17-098, Revised Response to NRC Request for Additional Information Related to TVA Fleet License Amendment Request to Adopt NEI 99-01 Revision 6 Emergency Action Levels2017-07-27027 July 2017 Revised Response to NRC Request for Additional Information Related to TVA Fleet License Amendment Request to Adopt NEI 99-01 Revision 6 Emergency Action Levels CNL-17-085, Response to NRC Request for Additional Information Related to TVA Fleet License Amendment Request to Adopt NEI 99-01 Revision 6 Emergency Action Levels2017-07-0707 July 2017 Response to NRC Request for Additional Information Related to TVA Fleet License Amendment Request to Adopt NEI 99-01 Revision 6 Emergency Action Levels CNL-17-039, 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 - Withdrawal of Regulatory..2017-03-10010 March 2017 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 - Withdrawal of Regulatory.. CNL-16-194, 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 Accid2016-12-23023 December 2016 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 Accid CNL-16-178, Response to Request for Additional Information (RAI) Regarding Application to Modify Technical Specifications Regarding Diesel Generator Steady State Frequency (SQN-TS-14-02)2016-12-23023 December 2016 Response to Request for Additional Information (RAI) Regarding Application to Modify Technical Specifications Regarding Diesel Generator Steady State Frequency (SQN-TS-14-02) CNL-16-196, Revised Response to NRC Generic Letter 2016-01, Monitoring of Neutron-Absorbing Materials in Spent Fuel Pools2016-12-19019 December 2016 Revised Response to NRC Generic Letter 2016-01, Monitoring of Neutron-Absorbing Materials in Spent Fuel Pools CNL-16-123, Response to NRC Request for Additional Information Request for Approval for Use of Alternate Calibration Block Reflector Requirements 16-PDI-52016-07-28028 July 2016 Response to NRC Request for Additional Information Request for Approval for Use of Alternate Calibration Block Reflector Requirements 16-PDI-5 CNL-16-086, Response to Request for Additional Information Technical Specifications 3.7.8 Change - Essential Raw Cooling Water System2016-05-31031 May 2016 Response to Request for Additional Information Technical Specifications 3.7.8 Change - Essential Raw Cooling Water System CNL-15-218, Application to Revise Technical Specification 6.8.4.h, Containment Leakage Rate Testing Program, (SQN-TS-14-03) (TAC Nos. MF5366 and MF5367), Supplement to Response to Request for Additional Information2015-10-23023 October 2015 Application to Revise Technical Specification 6.8.4.h, Containment Leakage Rate Testing Program, (SQN-TS-14-03) (TAC Nos. MF5366 and MF5367), Supplement to Response to Request for Additional Information CNL-15-171, Application to Revise Technical Specification 6.8.4.h, Containment Leakage Rate Testing Program, (SQN-TS-14-03) - Response to Request for Additional Information2015-09-11011 September 2015 Application to Revise Technical Specification 6.8.4.h, Containment Leakage Rate Testing Program, (SQN-TS-14-03) - Response to Request for Additional Information CNL-15-158, Supplement to TVA Letter, Sequoyah Nuclear Plant - Revision to Commitment No. 28 and Review of Impacts to the SQN Reactor Vessel Internals Aging Management Program Due to Dislodged Reactor Vessel Surveillance Capsules in Unit 1 Reactor.2015-08-0303 August 2015 Supplement to TVA Letter, Sequoyah Nuclear Plant - Revision to Commitment No. 28 and Review of Impacts to the SQN Reactor Vessel Internals Aging Management Program Due to Dislodged Reactor Vessel Surveillance Capsules in Unit 1 Reactor. CNL-15-146, Response to NRC Request for Information Regarding the Review of the License Renewal Application, Set 25 (TAC Nos. MF0481 and MF0482)2015-08-0303 August 2015 Response to NRC Request for Information Regarding the Review of the License Renewal Application, Set 25 (TAC Nos. MF0481 and MF0482) NL-15-158, Supplement to TVA Letter, Sequoyah Nuclear Plant - Revision to Commitment No. 28 and Review of Impacts to the SQN Reactor Vessel Internals Aging Management Program Due to Dislodged Reactor Vessel Surveillance Capsules in Unit 1 Reactor.2015-08-0303 August 2015 Supplement to TVA Letter, Sequoyah Nuclear Plant - Revision to Commitment No. 28 and Review of Impacts to the SQN Reactor Vessel Internals Aging Management Program Due to Dislodged Reactor Vessel Surveillance Capsules in Unit 1 Reactor. CNL-15-136, Response to NRC Request for Additional Information Regarding ASME Request for Relief RP-072015-07-22022 July 2015 Response to NRC Request for Additional Information Regarding ASME Request for Relief RP-07 ML15176A6822015-06-19019 June 2015 Sequoyah Nuclear Plants, Units 1 and 2 Technical Specifications Conversion to NUREG-1431, Rev. 4.0 (SQN-TS-11-10) - Supplement 2. Part 4 of 8 ML15176A6812015-06-19019 June 2015 Sequoyah Nuclear Plants, Units 1 and 2 Technical Specifications Conversion to NUREG-1431, Rev. 4.0 (SQN-TS-11-10) - Supplement 2. Part 3 of 8 ML15176A7482015-06-19019 June 2015 Plants, Units 1 and 2 Technical Specifications Conversion to NUREG-1431, Rev. 4.0 (SQN-TS-11-10) - Supplement 2. Part 8 of 8 ML15176A7402015-06-19019 June 2015 Plants, Units 1 and 2 Technical Specifications Conversion to NUREG-1431, Rev. 4.0 (SQN-TS-11-10) - Supplement 2. Part 7 of 8 ML15176A6792015-06-19019 June 2015 Sequoyah Nuclear Plants, Units 1 and 2 Technical Specifications Conversion to NUREG-1431, Rev. 4.0 (SQN-TS-11-10) - Supplement 2. Part 2 of 8 ML15176A6642015-06-19019 June 2015 Sequoyah Nuclear Plants, Units 1 and 2 Technical Specifications Conversion to NUREG-1431, Rev. 4.0 (SQN-TS-11-10) - Supplement 2. Part 1 of 8 2024-01-03
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Sequoyah Nuclear Plant, P.O. Box 2000, Soddy Daisy, Tennessee 37384 December 19, 2022 10 CFR 72.7 ATTN: Document Control Desk Director, Division of Fuel Management Office of Nuclear Material Safety and Safeguards U.S. Nuclear Regulatory Commission Washington, D.C. 20555-0001 Sequoyah Nuclear Plant, Units 1 and 2 Renewed Facility Operating License Nos. DPR-77 and DPR-79 NRC Docket Nos. 50-327, 50-328, and 72-034
Subject:
Sequoyah Nuclear Plant - Response to Request for Supplemental Information (D-RSI) Request for Exemption from Non-Destruction-Examination Compliance (EPID No. L-2022-LLE-0027)
Reference:
- 1. TVA letter to NRC , Sequoyah Nuclear Plant - Request for Exemption from Various 10 CFR Part 72 Regulations Resulting from Non-Destruction-Examination Compliance, dated August 4, 2022
- 2. NRC letter to SQN, Request For Supplemental Information - Request for Exemption from Various 10 CFR Part 72 Regulations Related to Non-Destructive Examination Compliance, Sequoyah Nuclear Plant Independent Spent Fuel Storage Installation [Enterprise Project Identification Number L-2022-LLE-0027], dated November 18, 2022
- 3. Holtec International letter to NRC, Supporting Document for Sequoyah Nuclear Plant ISFSI Exemption Request, dated December 12, 2022 Pursuant to 10 CFR 72.7, Specific exemption, TVA requested an exemption from the requirements of 10 CFR 72.212(a)(2), 72.212(b)(3), 72.212(b)(5)(i), 72.212(b)(11) and 72.214 by Reference 1.
The NRC review of the exemption request identified information necessary to continue evaluation. This necessary information is sought under Reference 2. Enclosure 1 provides the TVAs response to Reference 2. In response to the NRC Structural Discipline RSI-S1, Holtec International has submitted a copy of Holtec Report HI-2094418, Structural Calculation Package for HI-STORM FW System, in Reference 3. The Holtec letter refers to the Sequoyah Nuclear Plant docket numbers and the Enterprise Project Identification Number associated with this license activity.
U.S. Nuclear Regulatory Commission Page 2 December 19, 2022 This document contains no new regulatory commitments. If you have any questions, please contact Rick Medina, Site Licensing Manager, at (423) 843-8129.
Respectfully, Digitally signed by Reneau, Reneau, William William Christopher Christopher Date: 2022.12.19 07:53:54
-05'00' Thomas Marshall Site Vice President Sequoyah Nuclear Plant
Enclosure:
TVA Response to Request for Supplemental Information cc:
NRC Regional Administrator - Region II NRR Project Manager - Sequoyah Nuclear Plant NRC Senior Resident Inspector - Sequoyah Nuclear Plant
ENCLOSURE TENNESSEE VALLEY AUTHORITY (TVA)
SEQUOYAH NUCLEAR PLANT (SQN)
INDEPENDENT SPENT FUEL STORAGE INSTALLATION TVA RESPONSE TO REQUEST FOR SUPPLEMENTAL INFORMATION NRC Material Discipline RSI-M1 Provide the following supplemental information on the original shell-to-shell longitudinal weld, the radiography testing (RT) results that identified a defect located in the shell-to-shell longitudinal weld between 4 to 14 inches from the bottom of the multipurpose canister (MPC) baseplate, and the method used in the repair weld.
(a) Describe the welding process used for the original weld.
(b) Describe the original RT results that identified the initial defect(s) located between 4 to 14 inches from the bottom of the MPC baseplate including the size, location, orientation, and type of defect(s).
(c) Describe the original weld joint geometry and a sketch or detailed dimensions of the weld excavation.
(d) Describe the welding process was used for the first repair weld.
The applicant provided that the original RT identified a defect which located in the shell-to-shell longitudinal weld between 4 to 14 inches from the bottom of the MPC baseplate. The applicant stated that this defect required an excavation of approximately 6.5 long, 5/8 wide, and 9/32 deep. However, it is unclear where the defect was located with respect to the original joint configuration. In addition, the welding process used for the repair weld is not described in the exemption request.
This information is necessary to determine compliance with Title 10 of the Code of Federal Regulations (10 CFR), 72.236(g), (j), and (l).
TVA Response to RSI-M1 (a) The original weld was welded from both sides using the submerged arc welding (SAW) process utilizing WPS-227HC. SAW is a welding process that involves the formation of an electric arc between a continuously fed consumable electrode and the workpiece. The molten weld pool is shielded from the atmosphere beneath a blanket of powdered flux.
(b) The initial RT identified 3.7 inches of lack of fusion within view 0-1 for Weld No. 21 which is located approximately 4 to 14 inches from the bottom of the MPC baseplate with 0 being at the toe of the circumferential weld as illustrated in Figure 1 on the following page.
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Figure 1: Location Markers on Longitudinal Weld (c) The original MPC shell-to-shell longitudinal weld joint was a double V-groove. See Figure 2 for additional details. Refer to Weld No. 21 on Figure 3 (MTR 12306-657) for dimensions of the weld excavation area.
Figure 2: Original MPC shell-to-shell longitudinal weld joint Figure 3: MTR 12306-657 Repair Excavation Map E2 of E8
(d) The repair of the MPC shell-to-shell longitudinal weld involved the excavation of the weld defect from a single side using mechanical means (grinding). A sketch of the shell-to-shell joint after excavation of the weld defect is displayed in Figure 4. Once the defect was verified to be removed through visual and liquid penetrant examination, the excavated area was blended uniformly into the surrounding surface with not less than a 3:1 taper and re-welded using the gas tungsten arc welding (GTAW) process utilizing WPS 47HC. GTAW is a welding process that involves the formation of an electric arc between a non-consumable tungsten electrode and the workpiece. During welding the weld filler wire is manually fed into the molten weld pool to create the weld. The molten weld pool is shielded from the atmosphere using inert gas (Argon). The weld repair area is examined by the method(s)
Visual (VT), Liquid Penetrant (PT) and RT that originally disclosed the defect.
Figure 4: MPC shell-to-shell joint after excavation of the weld defect NRC Material Discipline RSI-M2 Provide the following supplemental information on the RT results of the repaired shell-to-shell longitudinal weld that identified a defect located in the shell-to-shell longitudinal weld between 8.5 to 25 inches from the bottom of the MPC baseplate.
(a) Describe the RT results for the repaired weld that identified the defect(s) located between 8.5 to 25 inches from the bottom of the MPC baseplate.
(b) Clarify whether the initial RT did not identify indications that were observed in a second RT in the area from 8.5 to 25 inches from the bottom of the MPC baseplate.
(c) Describe the RT results that identified the defect(s) located between 8.5 to 25 inches from the bottom of the MPC baseplate including the size, location, orientation, and type of defect(s).
(d) Describe the original and repair weld joint geometry and a sketch or detailed dimensions of the weld excavation between 8.5 to 25 inches from the bottom of the MPC baseplate.
(e) Describe the welding process that was used for the second repair weld. The applicant stated that a second RT was performed after the first repair which identified another defect located in the shell-to-shell longitudinal weld. This defect required an excavation of approximately 16.5 long, 5/8 wide, and 1/4 deep. It appears that there is an overlap of 5.5 inches between the two repairs. It is unclear whether the second defect was missed by the original RT, or this excavation was necessary to remove defect(s) in the first repair weld.
This information is necessary to determine compliance with 10 CFR 72.236(g), (j), and (l).
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TVA Response to RSI-M2 (a) The RT of the repair area on Weld No. 21 did not identify defect(s) located between 8.5 to 25 inches from the bottom of the MPC baseplate.
(b) The initial RT of view 1-2 which is approximately between 8.5 to 25 inches from the bottom of the MPC baseplate was acceptable and free of any indications. The second iteration of RT identified 0.327 inches of lack of fusion within view 0-1 for Weld No. 21 which is located approximately 4 to 14 inches from the bottom of the MPC baseplate. The removal of the defect via excavation caused the repair area to extend into the adjacent view 1-2 which is approximately 8.5 to 25 inches from the bottom of the baseplate.
(c) No defects were identified within view 1-2 which is approximately 8.5 to 25 inches from the bottom of the baseplate. The second iteration of RT identified 0.327 inches of lack of fusion within view 0-1 for Weld No. 21 which is located approximately 4 to 14 inches from the bottom of the MPC baseplate. The removal of the defect via excavation caused the repair area to extend into the adjacent view 1-2 which is approximately 8.5 to 25 inches from the bottom of the baseplate.
(d) The original MPC shell-to-shell longitudinal weld joint was a double V-groove. See Figure 2 for additional details. A sketch of the shell-to-shell joint after excavation of the weld defect is displayed in Figure 4. Refer to Weld No. 21 on Figure 5 (MTR 12306-667) for dimensions of the weld excavation area.
Figure 5: MTR 12306-667 Repair Excavation Map (e) The second defect was not missed by the initial RT. The second iteration of RT identified 0.327 inches of lack of fusion in the repaired weld. The excavation was necessary to remove the lack of fusion in the repaired weld.
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NRC Structural Discipline RSI-S1 Provide revision 20 of Holtec Report HI-2094418, Structural Calculation Package for HI-STORM FW System, and state why revision 20 is the version used and/or applicable for this exemption request.
As stated in safety evaluation report for the missing radiographic examination (RT), the evaluation mirrors the methodology, acceptance criteria, and finite element model from Supplement 1 of Holtec Report HI-2094418, revision 20, with some exemptions. The evaluation also utilizes the stress results from the finite element analysis provided in the report. However, this analysis report was not included in support of the evaluation.
This information is necessary to evaluate the requested exemption, under 10 CFR 72.7, from the requirements of 10 CFR 72.154(b), 10 CFR 72.212(a)(2), 10 CFR 72.212(b)(5)(i),
10 CFR 72.212(b)(11) and 10 CFR 72.214.
TVA Response to RSI-S1 A copy of Holtec Report HI-2094418, Revision 20, has been provided in Holtec letter from Kimberly Manzione to NRC, Supporting Document for Sequoyah Nuclear Plant ISFSI Exemption Request, dated December 12, 2022. The Holtec letter references SQNs docket numbers and the Enterprise Project Identification Number for this request.
Revision 20 of the report is consistent with HI-STORM FW Final Safety Analysis Report (FSAR),
Revision 6 and Certificate of Compliance, Amendment 3, which are applicable to SQN Independent Spent Fuel Storage Facility licensing basis for this loaded cask.
NRC Structural Discipline RSI-S2 Evaluate the potential of crack propagation from thermal cycling on potential flaws or imperfections on sections of the multipurpose canister (MPC) longitudinal shell-to-shell weld where the RT was missed.
As described in Holtec Report No. HI-2114830, HI-STORM FW FSAR, structural welds rely, in part, in welding operations to be performed in accordance with the requirements of codes and standards to ensure, in part, that no risk of crack propagation under the applicable stress levels will occur. However, the structural evaluation does not address the potential of crack propagation of flaws or imperfections that may have not been identified at the MPC longitudinal shell-to-shell weld sections where the RT was missed.
This information is necessary to evaluate the requested exemption, under 10 CFR 72.7, from the requirements of 10 CFR 72.154(b), 10 CFR 72.212(a)(2), 10 CFR 72.212(b)(5)(i),
10 CFR 72.212(b)(11) and 10 CFR 72.214.
TVA Response to RSI-S2 A detailed analysis for crack propagation of flaws or imperfections is not deemed necessary for the section of the MPC shell-to-shell weld that cannot be confirmed by documentary evidence to be radiographed since there is no potential for stress cycling in the unexamined portion of the E5 of E8
repair. As discussed in Paragraph 3.1.2.5 of HI-STORM FW FSAR, Revision 6, fatigue failure is not a credible concern for the MPC since it is not an active system (i.e., no moving parts) and is not subject to significant stress cycling due to rapid temperature changes or significant pressure changes. Therefore, there is no credible concern of fatigue failure even if the unlikely scenario of a flaw introduction during the weld repair is considered.
NRC Structural Discipline RSI-S3 The safety analysis does not address the effects of local membrane plus primary bending stress at the sections of the MPC longitudinal shell-to-shell weld where the RT was missed.
The safety evaluation of the missing RT on the MPC longitudinal shell-to-shell weld did not consider the impact of local membrane plus primary bending stresses at the affected region.
Therefore, it is not clear whether the proposed stress-reduction factor (SRF) the MPC confinement boundary stress intensity limits are met.
This information will be necessary to evaluate the requested exemption, under 10 CFR 72.7, from the requirements of 10 CFR 72.154(b), 10 CFR 72.212(a)(2), 10 CFR 72.212(b)(5)(i),
10 CFR 72.212(b)(11) and 10 CFR 72.214.
TVA Response to RSI-S3 The safety analysis presented in Holtec RRTI-3087-007, Revision 2, presents the maximum stress intensity in the MPC shell between 10 inches and 30 inches from the bottom of the MPC baseplate for all five (5) cases. The presented maximum stress intensity includes contributions from primary membrane, local membrane plus primary bending and secondary stresses although the contributions from local membrane plus primary bending and secondary stresses are expected to be minimal given the separation distance from the shell-to-baseplate or shell-to-lid discontinuities. The maximum stress intensity values are then compared conservatively with the applicable primary membrane stress limits (which are lower than local membrane plus primary bending stress and secondary stress limits) to compute the safety factors.
Therefore, the presented safety analysis considers the effects of local membrane plus primary bending stress and compares the maximum stress intensity results for all cases against the primary membrane stress limits to arrive at conservative safety factors.
NRC Observations: Structural Discipline The staff has observed the following based on the preliminary review of the information as provided by Sequoyah Nuclear Plant. Note that additional information will be required for resolution. The staff notes that a response to this observational question is not required at this time as part of the RSI response from SQN.
Adequate technical justification is necessary to demonstrate why a qualification factor of 0.75, as required in ASME Section III, Subsection NG, for a Category A full penetration, is not applicable for the safety evaluation of the missing RT on MPC longitudinal shell-to-shell weld.
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Section 3 of exemption request states that the use of a quality factor of 0.75 is an overly conservative lower bound for the SRF associated with the shell-to-shell plate weld since more than 98% of the weld has been examined by RT. However, this argument does not provide adequate technical basis to justify the use of less conservative SRF (e.g., 0.8 vs. 0.75).
This information will be necessary to evaluate the requested exemption, under 10 CFR 72.7, from the requirements of 10 CFR 72.154(b), 10 CFR 72.212(a)(2), 10 CFR 72.212(b)(5)(i),
10 CFR 72.212(b)(11) and 10 CFR 72.214.
TVA Response to Observation The basis for using a stress reduction factor (SRF) of 0.8 is based on the known condition of the MPC shell-to-shell longitudinal weld and consideration of Interim Staff Guidance (ISG)-15, NUREG-2215, American Society of Mechanical Engineers (ASME)Section VIII, Division 1, and ASME Section III, Subsections ND and NG. The detailed justification is presented below.
As discussed in the request for exemption, the shell-to-shell weld that cannot be confirmed by documentary evidence to be radiographed after repair is about 7.5 inches in length, 5/8 inch wide and 1/4 inch deep. Root and final PT examinations were performed on the entire repaired weld, and the results were acceptable with no further indications.
The MPC shell-to-shell weld is a full penetration joint, which is classified as a Category A joint per Subsection NB. Full penetration welds that are 100% examined using volumetric methods are considered fully effective under Subsection NB of the ASME Code, meaning the stress reduction factor (SRF) is equal to 1.0.
Interim Staff Guidance (ISG)-15 and Standard Review Plan for Spent Fuel Dry Storage Systems and Facilities - Final Report (NUREG-2215) both endorse a SRF of 0.8 for austenitic canisters with lid-to-shell (LTS) weld subject to progressive PT examination. The overall level of inspection performed on the shell-to-shell weld for MPC # 234 is more thorough and statistically more significant than that of a progressive PT examination over the full length of the weld.
ASME Section VIII, Div. 1 (Table UW-12) and ASME Section III, Subsection ND (Paragraph ND-3352) both specify a SRF of 0.85 for Category A joints subject to spot radiography. The level of inspection performed on the entirety of shell-to-shell weld for MPC # 234 far exceeds the minimum Code requirements for spot radiography (6-inch section to be examined for every 50-foot increment of weld per Section UW-52 of ASME Section VIII, Div. 1). It is also noted that, for SA-240 304 stainless steel, the design stress values applicable to Section VIII, Div. 1 and Section III, Subsection ND are generally equal to the design stress intensity values applicable to Section III, Subsection NB, except for minor variances at 300 and 400 degrees Fahrenheit.
ASME Section III, Subsection NG, which applies to core support structures and has the same design stress intensity values as Subsection NB, specifies a joint efficiency (also known as SRF) of 0.75 for a Category A full penetration weld subjected to root and final PT only (i.e., no RT) per Table NG-3352-1. Clearly, this is a conservative lower bound for the SRF associated with the shell-to-shell plate weld for MPC # 234 since more than 98% of the MPC welds has been examined by RT.
Based on the above, it is concluded that a SRF of 0.8 is an appropriate and justifiable value for the re-evaluation of the shell-to-shell repaired weld for MPC # 234. Per the re-evaluation in Holtec RRTI 3087-007, Revision 2, all calculated safety factors remain above 1.0 for all design basis load conditions. This indicates that the shell-to-shell weld has sufficient strength even in E7 of E8
the presence of possible, albeit unlikely, flaws in the small volume of weld that was not volumetrically examined after repair. It is also noted that all calculated safety factors for the MPC shell-to-shell weld would continue to remain above 1.0 even if a SRF of 0.75 is used.
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