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Category:Letter type:NL
MONTHYEARNL-24-0014, Revised Response to Request for Additional Information Regarding License Amendment Request to Relax the Required Number of Fully Tensioned Reactor Pressure Vessel Head Closure Studs in Technical.2024-01-30030 January 2024 Revised Response to Request for Additional Information Regarding License Amendment Request to Relax the Required Number of Fully Tensioned Reactor Pressure Vessel Head Closure Studs in Technical. NL-23-0889, Units 1 and 2 - License Amendment Request to Revise Technical Specifications to Adopt Risk Informed Completion Times for Residual Heat Removal Service Water (RHRSW) and Plant Service Water (Psw) Systems2023-12-0606 December 2023 Units 1 and 2 - License Amendment Request to Revise Technical Specifications to Adopt Risk Informed Completion Times for Residual Heat Removal Service Water (RHRSW) and Plant Service Water (Psw) Systems NL-23-0879, Request for Exemption from Security Event Notification Implementation2023-11-29029 November 2023 Request for Exemption from Security Event Notification Implementation NL-23-0841, Update to Notice of Intent to Pursue Subsequent License Renewal2023-11-20020 November 2023 Update to Notice of Intent to Pursue Subsequent License Renewal NL-23-0658, Southern Nuclear Operating Company - Response to Request for Additional Information Regarding Quality Assurance Topical Report Submittal2023-08-11011 August 2023 Southern Nuclear Operating Company - Response to Request for Additional Information Regarding Quality Assurance Topical Report Submittal NL-23-0542, CFR 50.46 ECCS Evaluation Model Annual Report for 20222023-08-0909 August 2023 CFR 50.46 ECCS Evaluation Model Annual Report for 2022 NL-23-0624, Report of Changes to Emergency Plan and Summary of 50.54(q) Analysis2023-08-0404 August 2023 Report of Changes to Emergency Plan and Summary of 50.54(q) Analysis NL-23-0566, ISFSI and Edwin I. Hatch Nuclear Plant ISFSI - Registration of Spent Fuel Cask Use2023-07-13013 July 2023 ISFSI and Edwin I. Hatch Nuclear Plant ISFSI - Registration of Spent Fuel Cask Use NL-23-0555, Request for Exemption from Physical Barrier Requirement2023-07-0707 July 2023 Request for Exemption from Physical Barrier Requirement NL-23-0444, Quality Assurance Topical Report Submittal2023-06-15015 June 2023 Quality Assurance Topical Report Submittal NL-23-0457, ISFSI, and Edwin I. Hatch Nuclear Plant, ISFSI - Registration of Spent Fuel Cask Use2023-06-12012 June 2023 ISFSI, and Edwin I. Hatch Nuclear Plant, ISFSI - Registration of Spent Fuel Cask Use NL-23-0422, Inservice Inspection Program Owner'S Activity Report for Refueling Outage 2R272023-05-30030 May 2023 Inservice Inspection Program Owner'S Activity Report for Refueling Outage 2R27 NL-23-0383, SNC Response to Regulatory Issue Summary 2023-01:Preparation And.2023-05-19019 May 2023 SNC Response to Regulatory Issue Summary 2023-01:Preparation And. NL-23-0372, Units 1 & 2, Joseph M. Farley Nuclear Plant - Units 1 & 2, Vogtle Electric Generating Plant - Units 1 & 2, Annual Radiological Environmental Operating Reports for 20222023-05-10010 May 2023 Units 1 & 2, Joseph M. Farley Nuclear Plant - Units 1 & 2, Vogtle Electric Generating Plant - Units 1 & 2, Annual Radiological Environmental Operating Reports for 2022 NL-23-0310, Annual Non-Radiological Environmental Operating Reports and Annual Radioactive Effluent Release Reports for 20222023-04-25025 April 2023 Annual Non-Radiological Environmental Operating Reports and Annual Radioactive Effluent Release Reports for 2022 NL-23-0019, GEN-ISI-ALT-2023-01, Request to Use Provision of a Later Edition of the ASME Boiler and Pressure Vessel Code, Section XI2023-04-12012 April 2023 GEN-ISI-ALT-2023-01, Request to Use Provision of a Later Edition of the ASME Boiler and Pressure Vessel Code, Section XI NL-23-0263, Southern Nuclear Company Submittal of Drug and Alcohol Testing Errors Identified 10 CFR 26.719(c) 30-Day Report2023-04-0505 April 2023 Southern Nuclear Company Submittal of Drug and Alcohol Testing Errors Identified 10 CFR 26.719(c) 30-Day Report NL-23-0208, Independent Spent Fuel Storage Installation ISFSI, Decommissioning Funding Plan Triennial Update2023-03-29029 March 2023 Independent Spent Fuel Storage Installation ISFSI, Decommissioning Funding Plan Triennial Update NL-23-0014, Southern Nuclear Operating Co Submittal of Report on Status of Decommissioning Funding2023-03-29029 March 2023 Southern Nuclear Operating Co Submittal of Report on Status of Decommissioning Funding NL-23-0228, Nuclear Property Insurance Coverage as of April 1, 2023 and Licensee Guarantees of Payment of Deferred Premiums (10 CFR 140.21)2023-03-20020 March 2023 Nuclear Property Insurance Coverage as of April 1, 2023 and Licensee Guarantees of Payment of Deferred Premiums (10 CFR 140.21) NL-23-0179, National Pollutant Discharge Elimination System (NPDES) Permit Renewal2023-03-13013 March 2023 National Pollutant Discharge Elimination System (NPDES) Permit Renewal NL-23-0101, Cycle 28 Core Operating Limits Report Version 12023-02-23023 February 2023 Cycle 28 Core Operating Limits Report Version 1 NL-23-0003, Response to Request for Additional Information Regarding License Amendment Request to Relax the Required Number of Fully Tensioned Reactor Pressure Vessel Head Closure Studs in Technical Specification Table 12023-01-20020 January 2023 Response to Request for Additional Information Regarding License Amendment Request to Relax the Required Number of Fully Tensioned Reactor Pressure Vessel Head Closure Studs in Technical Specification Table 1 NL-23-0069, Correction of Technical Specification Omission and Typographical Error2023-01-20020 January 2023 Correction of Technical Specification Omission and Typographical Error NL-22-0316, Emergency License Amendment Request for Technical Specification 3.7.2 Regarding One-Time Extension of Completion Time for Plant Service Water (Psw) Pump Inoperable2022-11-30030 November 2022 Emergency License Amendment Request for Technical Specification 3.7.2 Regarding One-Time Extension of Completion Time for Plant Service Water (Psw) Pump Inoperable NL-22-0850, Supplement to License Amendment Request to Revise the NFPA 805 Fire Protection Program2022-11-18018 November 2022 Supplement to License Amendment Request to Revise the NFPA 805 Fire Protection Program NL-22-0832, Supplement to License Amendment Request to Revise Technical Specifications to Adopt Risk Informed Completion Times TSTF-505, Revision 2, Provide Risk-Informed Extended Completion Times - RITSTF .2022-10-27027 October 2022 Supplement to License Amendment Request to Revise Technical Specifications to Adopt Risk Informed Completion Times TSTF-505, Revision 2, Provide Risk-Informed Extended Completion Times - RITSTF . NL-22-0800, Supplement to Authorized Relief Request HNP-ISI-RR-05-022022-10-27027 October 2022 Supplement to Authorized Relief Request HNP-ISI-RR-05-02 NL-22-0801, Response to Request for Additional Information Regarding: Application to Revise Technical Specifications to Adopt Technical Specifications Task Force (TSTF) Traveler TSTF-208, Revision 0, Extension of .2022-10-14014 October 2022 Response to Request for Additional Information Regarding: Application to Revise Technical Specifications to Adopt Technical Specifications Task Force (TSTF) Traveler TSTF-208, Revision 0, Extension of . NL-22-0757, Units 1 & 2, Supplement to License Amendment Request to Revise Technical Specifications to Adopt Risk Informed Completion Times TSTF-505, Revision 2, Provide Risk-Informed Extended Completion Times - RITSTF Initiative 4b2022-09-30030 September 2022 Units 1 & 2, Supplement to License Amendment Request to Revise Technical Specifications to Adopt Risk Informed Completion Times TSTF-505, Revision 2, Provide Risk-Informed Extended Completion Times - RITSTF Initiative 4b NL-22-0687, Notice of Intent to Pursue Subsequent License Renewal for Hatch Units 1 & 22022-08-31031 August 2022 Notice of Intent to Pursue Subsequent License Renewal for Hatch Units 1 & 2 NL-22-0406, Request to Relax the Required Number of Fully Tensioned Reactor Pressure Vessel Head Closure Studs in Technical Specification Table 1.1-1, Modes2022-08-19019 August 2022 Request to Relax the Required Number of Fully Tensioned Reactor Pressure Vessel Head Closure Studs in Technical Specification Table 1.1-1, Modes NL-22-0650, Units 1 and 2 - Response to Request for Confirmation of Information Regarding License Amendment Request to Revise the NFPA 805 Fire Protection Program2022-08-19019 August 2022 Units 1 and 2 - Response to Request for Confirmation of Information Regarding License Amendment Request to Revise the NFPA 805 Fire Protection Program NL-22-0590, 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 Exten2022-08-18018 August 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 Exten NL-22-0672, Registration of Spent Fuel Cask Use2022-08-17017 August 2022 Registration of Spent Fuel Cask Use NL-22-0564, and Vogtle Electric Generating Plant, Registration of Spent Fuel Cask Use2022-07-21021 July 2022 and Vogtle Electric Generating Plant, Registration of Spent Fuel Cask Use NL-22-0510, Plants Units 1 and 2, 10 CFR 50.46 ECCS Evaluation Model Annual Report for 20212022-07-14014 July 2022 Plants Units 1 and 2, 10 CFR 50.46 ECCS Evaluation Model Annual Report for 2021 NL-22-0499, Registration of Spent Fuel Cask Use2022-06-27027 June 2022 Registration of Spent Fuel Cask Use NL-22-0362, Southern Nuclear Operating Company, Inc. - Foreign Ownership, Control, or Influence Five-Year Renewal Filing and NRC Facility Clearance Update2022-06-0101 June 2022 Southern Nuclear Operating Company, Inc. - Foreign Ownership, Control, or Influence Five-Year Renewal Filing and NRC Facility Clearance Update NL-22-0365, Inservice Inspection Program Owner'S Activity Report (OAR-1) for Outage 1R302022-05-31031 May 2022 Inservice Inspection Program Owner'S Activity Report (OAR-1) for Outage 1R30 NL-22-0335, Supplement to Application to Revise Technical Specifications to Adopt Technical Specifications Task Force (TSTF) Traveler TSTF-208, Revision 0, Extension of Time to Reach Mode 2 in LCO 3.0.3 And.2022-05-17017 May 2022 Supplement to Application to Revise Technical Specifications to Adopt Technical Specifications Task Force (TSTF) Traveler TSTF-208, Revision 0, Extension of Time to Reach Mode 2 in LCO 3.0.3 And. NL-22-0344, Annual Radiological Environmental Operating Reports for 20212022-05-10010 May 2022 Annual Radiological Environmental Operating Reports for 2021 NL-22-0315, Technical Specification 5.6.6 Post Accident Monitoring Instrumentation Report for Drywell Temperature Indicator Inoperable for Greater than 30 Days2022-05-0404 May 2022 Technical Specification 5.6.6 Post Accident Monitoring Instrumentation Report for Drywell Temperature Indicator Inoperable for Greater than 30 Days NL-22-0325, Withdrawal of Emergency License Amendment Request for Technical Specification 3.7.2 Regarding One-Time Extension of Completion Time for Plant Service Water (Psw) Pump Inoperable2022-05-0303 May 2022 Withdrawal of Emergency License Amendment Request for Technical Specification 3.7.2 Regarding One-Time Extension of Completion Time for Plant Service Water (Psw) Pump Inoperable NL-22-0283, Supplement to License Amendment Request to Revise Technical Specifications to Adopt Risk Informed Completion Times TSTF-505, Revision 2, Provide Risk-Informed Extended Completion Times - Ritstf.2022-04-29029 April 2022 Supplement to License Amendment Request to Revise Technical Specifications to Adopt Risk Informed Completion Times TSTF-505, Revision 2, Provide Risk-Informed Extended Completion Times - Ritstf. NL-22-0271, Changes to the Physical Security Plan2022-04-19019 April 2022 Changes to the Physical Security Plan NL-22-0240, SNC Response to Regulatory Issue Summary 2022-01: Preparation and Scheduling of Operator Licensing Examinations2022-03-31031 March 2022 SNC Response to Regulatory Issue Summary 2022-01: Preparation and Scheduling of Operator Licensing Examinations NL-21-0918, License Amendment Request to Revise the NFPA 805 Fire Protection Program. with Attachments 3 to 52022-03-31031 March 2022 License Amendment Request to Revise the NFPA 805 Fire Protection Program. with Attachments 3 to 5 NL-22-0220, Units 1 and 2 - Nuclear Property Insurance Coverage as of April 1, 20222022-03-28028 March 2022 Units 1 and 2 - Nuclear Property Insurance Coverage as of April 1, 2022 NL-22-0221, Licensee Guarantees of Payment of Deferred Premiums (10 CFR 140.21)2022-03-28028 March 2022 Licensee Guarantees of Payment of Deferred Premiums (10 CFR 140.21) 2024-01-30
[Table view] Category:Report
MONTHYEARML24005A1142024-01-0505 January 2024 Recommendation for 2023-301 Cr/Sim 3 (Emergency Depress the Reactor Using Main Steam Line Drains) NL-22-0510, Plants Units 1 and 2, 10 CFR 50.46 ECCS Evaluation Model Annual Report for 20212022-07-14014 July 2022 Plants Units 1 and 2, 10 CFR 50.46 ECCS Evaluation Model Annual Report for 2021 A000412, National Pollutant Discharge Elimination System (NPDES) Permit Renewal Application2021-12-0202 December 2021 National Pollutant Discharge Elimination System (NPDES) Permit Renewal Application NL-20-1295, 10 CFR 71.95 Report on Non-Compliance Involving Radwaste Cask 3-60B2020-12-14014 December 2020 10 CFR 71.95 Report on Non-Compliance Involving Radwaste Cask 3-60B ML20303A1782020-09-29029 September 2020 Submittal of Revision 38 to Updated Final Safety Analysis Report NL-19-0674, Proposed Alternative to Use ASME Code Case N-831-1, Ultrasonic Examination in Lieu of Radiography for Welds in Ferritic or Austenitic Pipe Section XI, Division 12019-09-30030 September 2019 Proposed Alternative to Use ASME Code Case N-831-1, Ultrasonic Examination in Lieu of Radiography for Welds in Ferritic or Austenitic Pipe Section XI, Division 1 NL-18-0863, CFR 50.55a Request for Alternative HNP-ISI-ALT-05-04 Implementation of BWRVIP Documents in Lieu of Certain 8-N-1 and 8-N-2 Examinations2018-06-21021 June 2018 CFR 50.55a Request for Alternative HNP-ISI-ALT-05-04 Implementation of BWRVIP Documents in Lieu of Certain 8-N-1 and 8-N-2 Examinations NL-18-0506, Failure to Observe the Certificate of Compliance Condition of the 8-120B Cask Pre-Shipment Leak Test2018-04-16016 April 2018 Failure to Observe the Certificate of Compliance Condition of the 8-120B Cask Pre-Shipment Leak Test NL-17-1713, Proposed Alternative GEN-ISI-ALT-2017-03, and HNP-ISI-ALT-05-07, Version 1.0 Service Water Evaluation for Code Case N-513-4 for Moderate Pressure, and for Higher Pressure2018-04-0606 April 2018 Proposed Alternative GEN-ISI-ALT-2017-03, and HNP-ISI-ALT-05-07, Version 1.0 Service Water Evaluation for Code Case N-513-4 for Moderate Pressure, and for Higher Pressure NL-18-0282, Enclosure 1: NFPA 805 LAR Transition Report for Edwin I. Hatch2018-04-0404 April 2018 Enclosure 1: NFPA 805 LAR Transition Report for Edwin I. Hatch NL-17-1916, Pressure and Temperature Limits Report2017-11-27027 November 2017 Pressure and Temperature Limits Report ML18012A0582017-10-31031 October 2017 NEDO-33884, Revision 0, Gnf Fecrai ATF Lead Test Assembly for Edwin I. Hatch Nuclear Plant, Unit 1. NL-18-0026, NEDO-33884, Revision 0, Gnf Fecrai Atf Lead Test Assembly for Edwin I. Hatch Nuclear Plant, Unit 1.2017-10-31031 October 2017 NEDO-33884, Revision 0, Gnf Fecrai Atf Lead Test Assembly for Edwin I. Hatch Nuclear Plant, Unit 1. NL-18-0026, NEDO-33884, Revision 0, Gnf Fecrai ATF Lead Test Assembly for Edwin I. Hatch Nuclear Plant, Unit 1.2017-10-31031 October 2017 NEDO-33884, Revision 0, Gnf Fecrai ATF Lead Test Assembly for Edwin I. Hatch Nuclear Plant, Unit 1. NL-18-0026, NEDO-33883, Revision 0, Gnf Armor Lead Test Assembly for Edwin I. Hatch Nuclear Plant, Unit 1.2017-09-30030 September 2017 NEDO-33883, Revision 0, Gnf Armor Lead Test Assembly for Edwin I. Hatch Nuclear Plant, Unit 1. NL-17-1255, Fukushima Near-Term Task Force Recommendation 2.1 Seismic Limited-Scope High Frequency Confirmation Evaluation2017-08-22022 August 2017 Fukushima Near-Term Task Force Recommendation 2.1 Seismic Limited-Scope High Frequency Confirmation Evaluation NL-17-0955, Alternatives HNP-ISI-ALT-05-05 and HNP-ISI-ALT-05-062017-06-0505 June 2017 Alternatives HNP-ISI-ALT-05-05 and HNP-ISI-ALT-05-06 ML17069A2402017-04-13013 April 2017 Mitigating Strategies Assessment (CAC Nos. MF7932 and MF7933) - Redacted ML16356A0172016-12-16016 December 2016 Information Report for Lead Use Assemblies NL-16-2466, Fukushima Near-Term Task Force Recommendation 2.1 Expedited Seismic Evaluation Process Report Completion2016-12-15015 December 2016 Fukushima Near-Term Task Force Recommendation 2.1 Expedited Seismic Evaluation Process Report Completion NL-16-1136, Fukushima Near-Term Task Force Recommendation 2.1 Seismic Limited-Scope Low Frequency Evaluation2016-08-10010 August 2016 Fukushima Near-Term Task Force Recommendation 2.1 Seismic Limited-Scope Low Frequency Evaluation NL-16-0463, Submittal of 10 CFR 50.46 ECCS Evaluation Model, Annual Report for 2015 and Significant Change/Error Report2016-04-0101 April 2016 Submittal of 10 CFR 50.46 ECCS Evaluation Model, Annual Report for 2015 and Significant Change/Error Report NL-15-2010, E.I. Hatch - Submits 10 CFR 71.95 Report on Non-Conformance Involving Radwaste Cask 8-120B2015-11-0202 November 2015 E.I. Hatch - Submits 10 CFR 71.95 Report on Non-Conformance Involving Radwaste Cask 8-120B NL-15-1461, Submittal of 10 CFR 71.95 Report on Potential Issues Involving Radwaste Cask 8-120B2015-08-21021 August 2015 Submittal of 10 CFR 71.95 Report on Potential Issues Involving Radwaste Cask 8-120B ML15097A4242015-04-27027 April 2015 Staff Assessment of Information Provided Pursuant to Title 10 of the Code of Federal Regulations Part 50, Section 50.54(f), Seismic Hazard Reevaluations Relating to Recommendation 2.1 of the NTTF Review ML15106A3102015-04-17017 April 2015 April 22, 2015, Meeting with Southern Nuclear Company, Draft Minimum Shift Staffing Analysis ML14335A1372015-03-25025 March 2015 Interim Staff Evaluation Relating to Overall Integrated Plan in Response to Order EA-13-109 (Severe Accident Capable Hardened Vents) NL-14-1989, Expedited Seismic Evaluation Process Report - Fukushima Near-Term Task Force Recommendation 2.12014-12-30030 December 2014 Expedited Seismic Evaluation Process Report - Fukushima Near-Term Task Force Recommendation 2.1 NL-14-1876, Proposed Lnservice Inspection Alternative HNP-ISI-ALT-HDPE-01, Version 2.0, Conceptual Design Information Package2014-11-24024 November 2014 Proposed Lnservice Inspection Alternative HNP-ISI-ALT-HDPE-01, Version 2.0, Conceptual Design Information Package NL-14-1245, 10 CFR 26.719(c) Report: False Negative Results for a Blind Performance Test Sample2014-08-22022 August 2014 10 CFR 26.719(c) Report: False Negative Results for a Blind Performance Test Sample ML14223A7942014-07-31031 July 2014 Enclosure 2 - Non-Proprietary Gnf Report GNF-001N6296-R1-NP, Gnf Additional Information Regarding the Requested Changes to the Technical Specification SLMCPR, Hatch 2 Cycle 24 ML14155A4052014-06-30030 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 (TAC MF0234-35) ML14176A9612014-06-24024 June 2014 Submittal of Non-Proprietary BWROG Technical Product, BWROGTP-11-006 - ECCS Containment Walkdown Procedure, Rev 1 (January 2011), as Formally Requested During the Public Meeting Held on April 30, 2014 ML14155A3612014-06-0606 June 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 NL-14-0343, Seismic Hazard and Screening Report for CEUS Sites2014-03-31031 March 2014 Seismic Hazard and Screening Report for CEUS Sites NL-14-0326, Units 1 and 2, Recommendation 2.1 Flood Hazard Reevaluation Report, Requested by NRC Letter Dated March 12, 20122014-03-0606 March 2014 Units 1 and 2, Recommendation 2.1 Flood Hazard Reevaluation Report, Requested by NRC Letter Dated March 12, 2012 ML13364A2022014-02-27027 February 2014 Interim Staff Evaluation Related to Overall Integrated Plan in Response to Order EA-12-049 (Mitigation Strategies) ML14045A1472014-02-12012 February 2014 Mega-Tech Services, LLC Technical Evaluation Report Regarding the Overall Integrated Plan for Edwin I. Hatch Nuclear Plant, Units 1 and 2, TAC Nos.: MF0712 and MF0713 NL-13-1898, 10 CFR 71.95 Report on Potential Issues Involving Radwaste Cask 8-120B2013-08-30030 August 2013 10 CFR 71.95 Report on Potential Issues Involving Radwaste Cask 8-120B ML13193A3662013-08-0707 August 2013 Request for Concurrence on the Effects of the Edwin I. Hatch Nuclear Plant, Units 1 and 2 on the Federally-Listed Endangered Species Altamaha Spinymussel NL-13-0172, Southern Nuclear Operating Company'S Overall Integrated Plan in Response to 3/12/2012 Commission Order Modifying Licenses with Regard to Requirements for Reliable Spent Fuel Pool Instrumentation (EA-12-051)2013-02-27027 February 2013 Southern Nuclear Operating Company'S Overall Integrated Plan in Response to 3/12/2012 Commission Order Modifying Licenses with Regard to Requirements for Reliable Spent Fuel Pool Instrumentation (EA-12-051) NL-13-0214, Southern Nuclear Operating Company'S Overall Integrated Plan in Response to March 12, 2012 Commission Order Modifying Licenses with Regard to Requirements for Mitigation Strategies for Beyond-Design-Basis External Events..2013-02-27027 February 2013 Southern Nuclear Operating Company'S Overall Integrated Plan in Response to March 12, 2012 Commission Order Modifying Licenses with Regard to Requirements for Mitigation Strategies for Beyond-Design-Basis External Events.. ML13115A4732013-01-31031 January 2013 GNF-0000-0079-7396NP, Rev. 6, Technical Basis Supporting GNF-Ziron Lead Test Assembly Introduction Into the Hatch Nuclear Plant, March 2008 NL-13-0402, GNF-0000-0079-7396NP, Rev. 6, Technical Basis Supporting GNF-Ziron Lead Test Assembly Introduction Into the Hatch Nuclear Plant, March 20082013-01-31031 January 2013 GNF-0000-0079-7396NP, Rev. 6, Technical Basis Supporting GNF-Ziron Lead Test Assembly Introduction Into the Hatch Nuclear Plant, March 2008 ML12355A6332012-11-26026 November 2012 SNCH082-RPT-02, Ver. 1.0, Edwin I. Hatch, Unit 2 Seismic Walkdown Report for Resolution of Fukushima Near-Term Task Force Recommendation 2.3: Seismic, Part 3 of 11 ML12355A6312012-11-26026 November 2012 SNCH082-RPT-02, Ver. 1.0, Edwin I. Hatch, Unit 2 Seismic Walkdown Report for Resolution of Fukushima Near-Term Task Force Recommendation 2.3: Seismic, Part 2 of 11 ML12355A0592012-11-26026 November 2012 SNCH082-RPT-01, Ver. 1.0, Seismic Walkdown Report for Resolution of Fukushima Near-Term Task Force Recommendation 2.3: Seismic, Part 11 of 11 ML12355A0562012-11-26026 November 2012 SNCH082-RPT-01, Ver. 1.0, Seismic Walkdown Report for Resolution of Fukushima Near-Term Task Force Recommendation 2.3: Seismic, Part 8 of 11 ML12355A0552012-11-26026 November 2012 SNCH082-RPT-01, Ver. 1.0, Seismic Walkdown Report for Resolution of Fukushima Near-Term Task Force Recommendation 2.3: Seismic, Part 7 of 11 ML12355A0542012-11-26026 November 2012 SNCH082-RPT-01, Ver. 1.0, Seismic Walkdown Report for Resolution of Fukushima Near-Term Task Force Recommendation 2.3: Seismic, Part 6 of 11 2024-01-05
[Table view] Category:Miscellaneous
MONTHYEARML24005A1142024-01-0505 January 2024 Recommendation for 2023-301 Cr/Sim 3 (Emergency Depress the Reactor Using Main Steam Line Drains) NL-18-0863, CFR 50.55a Request for Alternative HNP-ISI-ALT-05-04 Implementation of BWRVIP Documents in Lieu of Certain 8-N-1 and 8-N-2 Examinations2018-06-21021 June 2018 CFR 50.55a Request for Alternative HNP-ISI-ALT-05-04 Implementation of BWRVIP Documents in Lieu of Certain 8-N-1 and 8-N-2 Examinations NL-18-0506, Failure to Observe the Certificate of Compliance Condition of the 8-120B Cask Pre-Shipment Leak Test2018-04-16016 April 2018 Failure to Observe the Certificate of Compliance Condition of the 8-120B Cask Pre-Shipment Leak Test NL-17-1713, Proposed Alternative GEN-ISI-ALT-2017-03, and HNP-ISI-ALT-05-07, Version 1.0 Service Water Evaluation for Code Case N-513-4 for Moderate Pressure, and for Higher Pressure2018-04-0606 April 2018 Proposed Alternative GEN-ISI-ALT-2017-03, and HNP-ISI-ALT-05-07, Version 1.0 Service Water Evaluation for Code Case N-513-4 for Moderate Pressure, and for Higher Pressure NL-17-0955, Alternatives HNP-ISI-ALT-05-05 and HNP-ISI-ALT-05-062017-06-0505 June 2017 Alternatives HNP-ISI-ALT-05-05 and HNP-ISI-ALT-05-06 ML17069A2402017-04-13013 April 2017 Mitigating Strategies Assessment (CAC Nos. MF7932 and MF7933) - Redacted ML16356A0172016-12-16016 December 2016 Information Report for Lead Use Assemblies NL-16-0463, Submittal of 10 CFR 50.46 ECCS Evaluation Model, Annual Report for 2015 and Significant Change/Error Report2016-04-0101 April 2016 Submittal of 10 CFR 50.46 ECCS Evaluation Model, Annual Report for 2015 and Significant Change/Error Report NL-15-1461, Submittal of 10 CFR 71.95 Report on Potential Issues Involving Radwaste Cask 8-120B2015-08-21021 August 2015 Submittal of 10 CFR 71.95 Report on Potential Issues Involving Radwaste Cask 8-120B ML15097A4242015-04-27027 April 2015 Staff Assessment of Information Provided Pursuant to Title 10 of the Code of Federal Regulations Part 50, Section 50.54(f), Seismic Hazard Reevaluations Relating to Recommendation 2.1 of the NTTF Review ML15106A3102015-04-17017 April 2015 April 22, 2015, Meeting with Southern Nuclear Company, Draft Minimum Shift Staffing Analysis ML14335A1372015-03-25025 March 2015 Interim Staff Evaluation Relating to Overall Integrated Plan in Response to Order EA-13-109 (Severe Accident Capable Hardened Vents) NL-14-1989, Expedited Seismic Evaluation Process Report - Fukushima Near-Term Task Force Recommendation 2.12014-12-30030 December 2014 Expedited Seismic Evaluation Process Report - Fukushima Near-Term Task Force Recommendation 2.1 NL-14-1245, 10 CFR 26.719(c) Report: False Negative Results for a Blind Performance Test Sample2014-08-22022 August 2014 10 CFR 26.719(c) Report: False Negative Results for a Blind Performance Test Sample ML14155A4052014-06-30030 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 (TAC MF0234-35) ML14155A3612014-06-0606 June 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 NL-14-0326, Units 1 and 2, Recommendation 2.1 Flood Hazard Reevaluation Report, Requested by NRC Letter Dated March 12, 20122014-03-0606 March 2014 Units 1 and 2, Recommendation 2.1 Flood Hazard Reevaluation Report, Requested by NRC Letter Dated March 12, 2012 NL-13-1898, 10 CFR 71.95 Report on Potential Issues Involving Radwaste Cask 8-120B2013-08-30030 August 2013 10 CFR 71.95 Report on Potential Issues Involving Radwaste Cask 8-120B ML13193A3662013-08-0707 August 2013 Request for Concurrence on the Effects of the Edwin I. Hatch Nuclear Plant, Units 1 and 2 on the Federally-Listed Endangered Species Altamaha Spinymussel NL-13-0172, Southern Nuclear Operating Company'S Overall Integrated Plan in Response to 3/12/2012 Commission Order Modifying Licenses with Regard to Requirements for Reliable Spent Fuel Pool Instrumentation (EA-12-051)2013-02-27027 February 2013 Southern Nuclear Operating Company'S Overall Integrated Plan in Response to 3/12/2012 Commission Order Modifying Licenses with Regard to Requirements for Reliable Spent Fuel Pool Instrumentation (EA-12-051) NL-13-0214, Southern Nuclear Operating Company'S Overall Integrated Plan in Response to March 12, 2012 Commission Order Modifying Licenses with Regard to Requirements for Mitigation Strategies for Beyond-Design-Basis External Events..2013-02-27027 February 2013 Southern Nuclear Operating Company'S Overall Integrated Plan in Response to March 12, 2012 Commission Order Modifying Licenses with Regard to Requirements for Mitigation Strategies for Beyond-Design-Basis External Events.. ML12355A0502012-11-26026 November 2012 SNCH082-RPT-01, Ver. 1.0, Seismic Walkdown Report for Resolution of Fukushima Near-Term Task Force Recommendation 2.3: Seismic, Part 2 of 11 NL-12-2268, SNCH082-RPT-01, Ver. 1.0, Seismic Walkdown Report for Resolution of Fukushima Near-Term Task Force Recommendation 2.3: Seismic, Part 1 of 112012-11-26026 November 2012 SNCH082-RPT-01, Ver. 1.0, Seismic Walkdown Report for Resolution of Fukushima Near-Term Task Force Recommendation 2.3: Seismic, Part 1 of 11 NL-12-2268, Edwin I. Hatch, Unit 1, SNCH082-RPT-01, Ver. 1.0, Seismic Walkdown Report for Resolution of Fukushima Near-Term Task Force Recommendation 2.3: Seismic, Part 9 of 112012-11-26026 November 2012 Edwin I. Hatch, Unit 1, SNCH082-RPT-01, Ver. 1.0, Seismic Walkdown Report for Resolution of Fukushima Near-Term Task Force Recommendation 2.3: Seismic, Part 9 of 11 NL-12-2268, Edwin I. Hatch, Unit 1, SNCH082-RPT-01, Ver. 1.0, Seismic Walkdown Report for Resolution of Fukushima Near-Term Task Force Recommendation 2.3: Seismic, Part 5 of 112012-11-26026 November 2012 Edwin I. Hatch, Unit 1, SNCH082-RPT-01, Ver. 1.0, Seismic Walkdown Report for Resolution of Fukushima Near-Term Task Force Recommendation 2.3: Seismic, Part 5 of 11 NL-12-2268, Edwin I. Hatch, Unit 1, SNCH082-RPT-01, Ver. 1.0, Seismic Walkdown Report for Resolution of Fukushima Near-Term Task Force Recommendation 2.3: Seismic, Part 6 of 112012-11-26026 November 2012 Edwin I. Hatch, Unit 1, SNCH082-RPT-01, Ver. 1.0, Seismic Walkdown Report for Resolution of Fukushima Near-Term Task Force Recommendation 2.3: Seismic, Part 6 of 11 NL-12-2268, Edwin I. Hatch, Unit 1, SNCH082-RPT-01, Ver. 1.0, Seismic Walkdown Report for Resolution of Fukushima Near-Term Task Force Recommendation 2.3: Seismic, Part 7 of 112012-11-26026 November 2012 Edwin I. Hatch, Unit 1, SNCH082-RPT-01, Ver. 1.0, Seismic Walkdown Report for Resolution of Fukushima Near-Term Task Force Recommendation 2.3: Seismic, Part 7 of 11 ML12355A0512012-11-26026 November 2012 SNCH082-RPT-01, Ver. 1.0, Seismic Walkdown Report for Resolution of Fukushima Near-Term Task Force Recommendation 2.3: Seismic, Part 3 of 11 NL-12-2268, Edwin I. Hatch, Unit 1, SNCH082-RPT-01, Ver. 1.0, Seismic Walkdown Report for Resolution of Fukushima Near-Term Task Force Recommendation 2.3: Seismic, Part 8 of 112012-11-26026 November 2012 Edwin I. Hatch, Unit 1, SNCH082-RPT-01, Ver. 1.0, Seismic Walkdown Report for Resolution of Fukushima Near-Term Task Force Recommendation 2.3: Seismic, Part 8 of 11 NL-12-2269, SNCH082-RPT-02, Ver. 1.0, Edwin I. Hatch, Unit 2 Seismic Walkdown Report for Resolution of Fukushima Near-Term Task Force Recommendation 2.3: Seismic, Part 8 of 112012-11-26026 November 2012 SNCH082-RPT-02, Ver. 1.0, Edwin I. Hatch, Unit 2 Seismic Walkdown Report for Resolution of Fukushima Near-Term Task Force Recommendation 2.3: Seismic, Part 8 of 11 NL-12-2269, SNCH082-RPT-02, Ver. 1.0, Edwin I. Hatch Unit 2 Seismic Walkdown Report for Resolution of Fukushima Near-Term Task Force Recommendation 2.3: Seismic, Part 1 of 112012-11-26026 November 2012 SNCH082-RPT-02, Ver. 1.0, Edwin I. Hatch Unit 2 Seismic Walkdown Report for Resolution of Fukushima Near-Term Task Force Recommendation 2.3: Seismic, Part 1 of 11 NL-12-2269, SNCH082-RPT-02, Ver. 1.0, Edwin I. Hatch, Unit 2 Seismic Walkdown Report for Resolution of Fukushima Near-Term Task Force Recommendation 2.3: Seismic, Part 5 of 112012-11-26026 November 2012 SNCH082-RPT-02, Ver. 1.0, Edwin I. Hatch, Unit 2 Seismic Walkdown Report for Resolution of Fukushima Near-Term Task Force Recommendation 2.3: Seismic, Part 5 of 11 NL-12-2269, SNCH082-RPT-02, Ver. 1.0, Edwin I. Hatch, Unit 2 Seismic Walkdown Report for Resolution of Fukushima Near-Term Task Force Recommendation 2.3: Seismic, Part 4 of 112012-11-26026 November 2012 SNCH082-RPT-02, Ver. 1.0, Edwin I. Hatch, Unit 2 Seismic Walkdown Report for Resolution of Fukushima Near-Term Task Force Recommendation 2.3: Seismic, Part 4 of 11 NL-12-2269, SNCH082-RPT-02, Ver. 1.0, Edwin I. Hatch, Unit 2 Seismic Walkdown Report for Resolution of Fukushima Near-Term Task Force Recommendation 2.3: Seismic, Part 3 of 112012-11-26026 November 2012 SNCH082-RPT-02, Ver. 1.0, Edwin I. Hatch, Unit 2 Seismic Walkdown Report for Resolution of Fukushima Near-Term Task Force Recommendation 2.3: Seismic, Part 3 of 11 NL-12-2269, SNCH082-RPT-02, Ver. 1.0, Edwin I. Hatch, Unit 2 Seismic Walkdown Report for Resolution of Fukushima Near-Term Task Force Recommendation 2.3: Seismic, Part 2 of 112012-11-26026 November 2012 SNCH082-RPT-02, Ver. 1.0, Edwin I. Hatch, Unit 2 Seismic Walkdown Report for Resolution of Fukushima Near-Term Task Force Recommendation 2.3: Seismic, Part 2 of 11 NL-12-2268, Edwin I. Hatch, Unit 1, SNCH082-RPT-01, Ver. 1.0, Seismic Walkdown Report for Resolution of Fukushima Near-Term Task Force Recommendation 2.3: Seismic, Part 10 of 112012-11-26026 November 2012 Edwin I. Hatch, Unit 1, SNCH082-RPT-01, Ver. 1.0, Seismic Walkdown Report for Resolution of Fukushima Near-Term Task Force Recommendation 2.3: Seismic, Part 10 of 11 ML12355A6332012-11-26026 November 2012 SNCH082-RPT-02, Ver. 1.0, Edwin I. Hatch, Unit 2 Seismic Walkdown Report for Resolution of Fukushima Near-Term Task Force Recommendation 2.3: Seismic, Part 3 of 11 ML12356A4102012-11-26026 November 2012 SNCH082-RPT-02, Ver. 1.0, Edwin I. Hatch, Unit 2 Seismic Walkdown Report for Resolution of Fukushima Near-Term Task Force Recommendation 2.3: Seismic, Part 9 of 11 ML12356A4092012-11-26026 November 2012 SNCH082-RPT-02, Ver. 1.0, Edwin I. Hatch, Unit 2 Seismic Walkdown Report for Resolution of Fukushima Near-Term Task Force Recommendation 2.3: Seismic, Part 11 of 11 ML12356A4082012-11-26026 November 2012 SNCH082-RPT-02, Ver. 1.0, Edwin I. Hatch, Unit 2 Seismic Walkdown Report for Resolution of Fukushima Near-Term Task Force Recommendation 2.3: Seismic, Part 10 of 11 ML12355A6662012-11-26026 November 2012 SNCH082-RPT-02, Ver. 1.0, Edwin I. Hatch, Unit 2 Seismic Walkdown Report for Resolution of Fukushima Near-Term Task Force Recommendation 2.3: Seismic, Part 7 of 11 NL-12-2269, SNCH082-RPT-02, Ver. 1.0, Edwin I. Hatch, Unit 2, Seismic Walkdown Report for Resolution of Fukushima Near-Term Task Force Recommendation 2.3: Seismic, Part 6 of 112012-11-26026 November 2012 SNCH082-RPT-02, Ver. 1.0, Edwin I. Hatch, Unit 2, Seismic Walkdown Report for Resolution of Fukushima Near-Term Task Force Recommendation 2.3: Seismic, Part 6 of 11 ML12355A6622012-11-26026 November 2012 SNCH082-RPT-02, Ver. 1.0, Edwin I. Hatch, Unit 2 Seismic Walkdown Report for Resolution of Fukushima Near-Term Task Force Recommendation 2.3: Seismic, Part 8 of 11 ML12355A6372012-11-26026 November 2012 SNCH082-RPT-02, Ver. 1.0, Edwin I. Hatch Unit 2 Seismic Walkdown Report for Resolution of Fukushima Near-Term Task Force Recommendation 2.3: Seismic, Part 1 of 11 ML12355A6352012-11-26026 November 2012 SNCH082-RPT-02, Ver. 1.0, Edwin I. Hatch, Unit 2 Seismic Walkdown Report for Resolution of Fukushima Near-Term Task Force Recommendation 2.3: Seismic, Part 5 of 11 ML12355A6342012-11-26026 November 2012 SNCH082-RPT-02, Ver. 1.0, Edwin I. Hatch, Unit 2 Seismic Walkdown Report for Resolution of Fukushima Near-Term Task Force Recommendation 2.3: Seismic, Part 4 of 11 ML12355A0522012-11-26026 November 2012 SNCH082-RPT-01, Ver. 1.0, Seismic Walkdown Report for Resolution of Fukushima Near-Term Task Force Recommendation 2.3: Seismic, Part 4 of 11 ML12355A6312012-11-26026 November 2012 SNCH082-RPT-02, Ver. 1.0, Edwin I. Hatch, Unit 2 Seismic Walkdown Report for Resolution of Fukushima Near-Term Task Force Recommendation 2.3: Seismic, Part 2 of 11 ML12355A0592012-11-26026 November 2012 SNCH082-RPT-01, Ver. 1.0, Seismic Walkdown Report for Resolution of Fukushima Near-Term Task Force Recommendation 2.3: Seismic, Part 11 of 11 ML12355A0582012-11-26026 November 2012 SNCH082-RPT-01, Ver. 1.0, Seismic Walkdown Report for Resolution of Fukushima Near-Term Task Force Recommendation 2.3: Seismic, Part 10 of 11 2024-01-05
[Table view] |
Text
t* ';'"'
SOUTHERN'\
COMPANY July 31, 2009 Ellngy {O Serlle }Ollr World Docket No.: 50-366 NL-09-1145 U. S. Nuclear Regulatory Commission ATIN: Document Control Desk Washington, D. C. 20555-0001 Edwin I. Hatch Nuclear Plant-Unit 2 Inservice Inspection Program Owner's Activity Report for Outage 2R20 Ladies and Gentlemen:
Enclosed is the ASME Section XI Code Case N-532-4 OAR-1 Owner's Activity Report for the 2R20 Refueling Outage which includes:
- Table 1, Items with Flaws or Relevant Conditions that Required Evaluation for Continued Service, see Structural Integrity Associates, Inc report for the flaw evaluation of the Unit 2 Recirculation Inlet Nozzle (N2)
G5 Nozzle-to-Safe End Weld Axial Indication that exceeded the Code acceptance criteria.
- Table 2, Abstract of Repairs, Replacement, or Corrective Measures Required for Continued Service which includes those repairs and replacements that occurred during operating cycle twenty (20) and Refueling Outage 2R20.
This report is for the second period of the 4th Interval lSI activities (Interval 4, Period 2, Outage 1).
This letter contains no NRC commitments. If you have any questions, please advise.
Sincerely, M. J. Ajluni Manager, Nuclear Licensing MJAlPAH/lac
Enclosure:
OAR-1 2R20 Owner's Report
U. S. Nuclear Regulatory Commission Log: NL-1145 Page 2 cc: Southern Nuclear Operating Company Mr. J. 1. Gasser, Executive Vice President Mr. D. R. Madison, Vice President - Hatch Ms. P. M. Marino, Vice President - Engineering RTYPE: CHA02.004 U. S. Nuclear Regulatory Commission Mr. L. A. Reyes, Regional Administrator Ms. D. N. Wright, NRR Project Manager - Hatch Mr. J. A. Hickey, Senior Resident Inspector - Hatch
Enclosure OAR-1 2R20 Owner's Report
FORM OAR*! OWNER'S ACTIVITY REPORT Report Number 2-4-2-1 (Unit 2, 4TH Interval, 2 ND Period, I ST Report)
Plant Edwin 1. Hatch Nuclear Plant, J 1028 Hatch Parkway North, Baxley, Georgia 31513 Unit No. 2 Commercial service date W5179 Refueling outage no. 2R20 (if applicable)
Current inspection interval Currenl inspection period Edition and Addenda of Section XI applicable to the inspection plans ASME Section XI, 200 I Edition with 2003 Addenda Date and revision of inspection plans 01/15/09, Revision 2 Edition and Addenda of Section XI applicable to repair/replacement activities, if different than the inspection plans Same Code Cases used: N/A (if applicable)
CERTIFICATE OF CONFORMANCE I certify that (a) the statements made in this report are correct; (b) the examinations and tests meet the Inspection Plan as required by the ASME Code,Section XI; and (c) the repair/replacement at:tivities and evaluations supporting the completion of 2R20 conform to the requ' ements of Section XI. (refueling outage number)
Date CERTIFICATE OF INSERVICE INSPECTION I, the undersi~ned, holding a valid commission issued by the National Board o~iler and Pressure Vessel Inzectors an2.
the State or Province of Get-'[~ : £.1 and employed by NS () C I of t.JtU-+ . J, r CI have inspected the items describe in this Owner's Activity Report, and state that to the best of my kno'wledge and belief, the Owner has performed all activities represented by this report in accordance with the requirements of Section XI.
By signing this certificate neither the Inspector nor his employer makes any warranty, expressed or implied, concerning the repair/replacement activities and evaluation described in this report. Furthermore, neither the Inspector nor his employer shall be liable in any manner for any personal injury or property damage or a loss of any kind arising from or connected with this inspection.
ar4-At~
IT Inspector's Signature GA~75 National Board. State, Province, and Endorsements Date () z/361z tJ £J 7
Examination Evaluation Category and Description Item Number Item Description B-F / B5.1O One indication was discovered during scheduled lSI The fatigue crack growth analysis was performed based examination (ultrasonic testing (UT)) of RPV nozzle to upon examination data and crack growth rates utilizing safe end weld for the Recirc inlet nozzle 2N2G weld pertinent BWRVIP documents. The allowable flaw size 2B31-1 RC-12AR-G-5 which exceeded the acceptance evaluation was also based upon the guidelines of Section criteria of IWB-3514. The indication was axially XI IWB*3600 and Appendix C. Following evaluation of oriented with a depth of .26 inches and a length of the stress intensity factor, stress corrosion, and fatigue approximately .50 inches. crack growth analyses in order to compare end of evaluation period flaw size to the allowable flaw size, it was concluded to be acceptable to operate for one cycle (2 years). Reference Plant Hatch Unit 2 2R20 OAR-I Report, Table I, Enclosure I for a copy of the flaw evaluation performed by Structural Integrity Associates.
~!.....,~... Jl',~;"*, ,.L,~ ~d,i:;:~\ '".'",,~ '~ "1of1
Plant Hatch Unit 2 2R20 OAR-1 Report, Table 1, Enclosure 1 l) Structural Integrity Associates, Inc. I File No.: 0900257.301
) CALCULATION PACKAGE Project No.: 0900257 Quality Program: ~ Nuclear 0 Commercial PROJECT ~A;\,lE:
Hatch Unit 2 Recirculation Inlet Nozzle G-5 Nozzle-to-Safe End Weld Flaw Evaluation CONTRACT ~O.:
Later CLIENT: PLANT:
Southern Nuclear Operating Company Hatch Unit 2 CALCULATION TITLE:
Flaw Evaluation of the Hatch Unit 2 Recirculation Inlet Nozzle (N2) G-5 Nozzle-to-Safe End Weld Axial Indication Project Manager Preparer(s) &
Document Affected Revision Desuiption Approval Checker(s)
Revision Pages Signature & Date Signatures & Date A I -9 Initial Draft Issue M. L. Herrera S. S. Tang 2/23/09 2/23/09 M. L. Herrera 2/23/09
) o I - 10 Initial Issue jt~~~
M. L. Herrera 2/24/09 M. L. Herrera 2/24/09
\)
Page I of 10 FOJU6*()I RU
Plant Hatch Unit2 2R20 OAR-1 Report, Table 1, Enclosure 1 1.0 I~TRODL"CTION
)
A tlaw evaluation is performed to disposition an indication observed in the Hatch Unit 2 Recirculation Inlet Nozzle (N.2) G-5 nozzle-to-safe end weld (2B31-1 RC-12AR-G-5). The indication is axially oriented with a depth of 0.26 inches and a length of approximately 0.50 inches [1.2]. The indication based on inspection results is contained in the Alloy 182 weld. This weld is a dissimilar metal weld (Alloy 182) joining the stainless steel safe end and the low alloy steel (LAS) nozzle [2]. This location was stress mitigated lIsing Induction Heating Stress Improvement (IHSI) [3].
This evaluation includes flaw evaluation to determine the acceptability of the observed indication for the next operating cycle.
2.0 TECHNICAL APPROACH The evaluation consists of:
- l. Crack growth analysis based on inspection data and crack growth rates using pertinent BWRVIP documents.
- 2. Allowable tlaw size evaluation based on the guidelines of ASME B&PV Code.Section XI, lWB 3600 and Appendix C [4]. The allowable flaw size was determined using the tabular solutions as allowed by Appendix C of Section XL
- 3. Evaluation of the stress intensity factor, stress corrosion and fatigue crack growth analyses to compare end of evaluation period flaw size to the allowable flaw size. The evaluation period is I
)':
operating cycle for a total of 2 years.
3.0 DESIGN INPUTS The outside radius at the weld location is 7 3116 inches (7.1875) Figure I [2]. The local wall thickness is 1.22 inches [1,2].
The safe end material is SA 182 Gr. F304 and the nozzle material is SA 508 CI. .2 [2]. The weld material is Alloy 182 [2].
Since the flaw is an axial flaw, the only stress of interest is the hoop stress due to pressure. The design pressure stress is 1423 psi for the recirculation line [5].
4.0 ASSli:VIPTIONS The following assumptions are used in the flaw evaluation:
I. Although this weld was subjected to IHSJ, no credit is taken for the beneficial residual Slress produced by the IHSI process.
File No.: 0900257.301 Page 2 of 10
)
Revision: 0 FOJ06-01
Plant Hatch Unit 2 2R20 OAR-1 Report, Table 1. Enclosure 1
, The indication is in the Alloy 182 wc Id material.
~
J. Alloy 182 is assumed to have the same matcrial properties as Alloy 600. Alloy 182 is the
) corresponding SMA W wire for Alloy 600 and the composition closely matches. Use of Alloy 600 propcrties for the Alloy 182 material is consistent with general industry practice.
- 4. The tlaw is postulated to be an active stress corrosion Haw.
- 5. The crack is initially contained in Alloy 182 weld material. but growth into the stainless steel safe end cannot be ruled out. Also, growth to the low alloy steel nozzle interface is also possible.
- 6. The depth of the Haw is also calculated using the observed tlaw depth and detennining growth using stainless steel stress corrosion crack (SCC) growth rate.
Note that since the weld has been subjected to IHSI, using a residual stress for the as-welded condition wIth a maximum stress equivalent to the yield strength producing high stress intensity factors, is conservative (Assumption I). It is likely that the IHSI process improves the residual stress with regards to intergranular stress corrosion cracking.
5.0 CALCl:LATIONS 5.1 Stress Corrosion Crack Growth Analysis A crack growth analysis is perfonned for the observed indication to obtain the crack growth due to stress corrosion and fatigue. Hatch Unit 2 is currently injecting Hydrogen at sufficient levels to fully protect this weld location. Since crack growth into the safe end or nozzle cannot be ruled out due to growth in the length direction. crack growth (depth direction) for Alloy 182. stainless steel and low alloy steel must be considered. The crack growth rate for LAS is bounded by that for Alloy 182. thus the naw depth detennined for Alloy 182 will be conservatively applied to the LAS. Figure 2 illustrates the postulated naw configuration that considers lengthwise growth of the flaw.
From Reference 6. BWRVIP-59-A. stress corrosion crack growth rates are obtained for Alloy 182. SCC growth rates are provided for Hydrogen Water Chemistry injection conditions depending on the stress intensity tactor levels.
da/dt = 5x J 0- 6 in/hr for K > 25 ksi...Jin (I) da/dt = 3.2x 10- 10 K3 in/hr for K.:::: 25 ksi...Jin (2)
For purposes of this evaluation, the crack growth rate for a K greater than 25 ksi...Jin will be conservatively used.
Crack growth in the Alloy 182 over the next operating cycle (2 years) is:
<'la = 365(24)(2)(5x 10-6 ) =:: 0.09 inches.
The final flaw depth in the Alloy 182 after two years is:
af = 3 0 +.1.a = (0.26) + 0.09 = 0.35 inches File No: 0900257.301 Page 3 of 10
)
Revision: 0 F0306-0 I
Plant Hatch Unit 2 2R20 OAR-1 Report, Table 1, Enclosure 1 where ~a = change in tlaw depth
) a,) = initial tlaw depth at = tinal tlaw depth This depth is equivalent to 0.35/1.22 = 0.287 or 28.7% of the pipe wall. This depth will be used to dctermlne the acceptability of the !law in the Alloy 182 and low alloy steel nozzle material since Alloy 182 crack growth rates bound those for low alloy steel.
Although the tlaw is not identified as being in the stainless steel, crack growth into the safe-end is also considered. For stainless steel in the Hydrogen Water Chemistry environment, a constant crack growth rate of 1.1 x I0.5 inihr [7] can be used. This results in a crack growth in the stainless steet material of 0.193 inches in 2 years:
c'1a = 365( 24)( 2)( 1.1 x 10'5) = 0.193 inches.
at' = au + 0.193 inches = (0.26) + 0.193 = 0.453 inches This depth is cquivalent to 0.453/1.22 = 37% of the pipe wall.
The tlaw growth in length (along the longitudinal axis of the nozzle/pipe) over the next operating cycle (2 years) will be conservatively estimated using 5xlO- 5 inlhr [6] at each end of the axial tlaw. Therefore, the tinallength (ft*) is ff = 0.5 inches + (5x I 0'5(365)(24)(2)(2>> inches = 2.25 inches This gives a frl.,J(Rrnt) = 2.25/~(6.5775'" 1.22) = 0.79. This is conservative since the tlaw would grow to the interface of the low alloy steel and essentially arrest. However, tor purposes of this evaluation, this calculated length will be used.
5.2 Screening Criteria Although the tlaw is fully contained in the Alloy 182 material, this calculation also considers the potential for lengthwise growth into the stainless steel sate end and low alloy steel nozzle. Note that growth into the LAS will be small compared to that in the Alloy 182. Thus for purposes of this evaluation. the tlaw depth predicted in the Alloy 182 will be used for the low alloy steel. Figure 2 illustrated the tlaw configuration.
The failure mode is determined according to Article C-4000 per Figure C-4210-1 (tor austenitic piping) and C-4220.1 (for ferritic piping). For austenitic piping, Article 6000, "Flaw Evaluation for Ductile Fracture using EPFM Criteria" can be used for the Alloy 182 flux weld [4] and stainless steel piping. It is assumed that the weld is a nux weld in this evaluation.
For the fcrritic steel (nozzle), the failure mode tor an axial tlaw is determined based on the screening criteria SC dctined as tallow:
SC = K,'iS,' (3)
K,' = [IOOOK,2/(E'J,c>)"5 (4)
File No.: 0900257.301 Page 4 of 10 R.evision: 0 F0306-01
Plant Hatch Unit 2 2R20 OAR-1 Report, Table 1, Enclosure 1 Sr' = (pR,,/t)/crt (5)
K 1 = (pRmit)(rta;Q)Il'F (6)
) Q = I + 4.593(a/f)1 65 (7)
F= I. L2+0.053a+0.0055a c+( 1.0+0.02a+0.0 19 lac)(20-R,nlt)2/l 400 (8) a = (a;t )/(ail) (9) crt= cry i[ 1-(a/t)]1 ll-(a/t)/M c]} ( 10)
M 2 ='[I+(1.61/(4R on t>>)it 5 ( I I)
E' = E/( l_v 2 ) (12)
Where: a = crack depth
- f. = crack length cry = yield strength Ron = mean pipe radius t = pipe wall thickness E = Young's modulus v = Poisson's ratio F = parameter for axial flaw stress intensity factor 2
Using the material properties per Table C-8322-1 for an axial naw in ferritic steel, J,e = 300 in-lb/in is lIsed for temperature ~ upper shelf temperature.
6 l]sing a = 0.35 inch. t = 2.25 inch, cry =42.38 ksi, E=25.45x I 0 psi and v=0.3 for SA-508 Class 2 [8] at 575 of, the followings were calculated:
) alt = 0.2869 atl=0.1556 a = 1.8437 Q = 1.2133 F = 1.4044 M 2 = 1.1198 crt = 40.6314 Sr'= 0.1888 E'= 27967 ksi K, = 10.2574 K r ' = O,ll2 SC =- 0.5931 Pcr Figurc C-4220-1 [4], when O.l ~ SC ~ 1.8, EPFM failure criteria in Article C-6000 is used.
5.3 Stress Ratio To lise the allowable flaw tables in Appendix C. a stress ratio is necessary as well as the flaw length. As defined on Table C -6410 of Appendix C of the ASME Code. the stress ratio is given as:
File No.: 0900257.301 Page 5 of 10
)
Revision: 0 F0306*0 I
Plant Hatch Unit 2 2R20 OAR-1 Report, Table 1, Enclosure 1
( 13)
Where: O"h = hoop stress from pressure Of = !low stress = (Sy +S\I)/2 Sy = yield strength S\I = tcnsi Ie strength Note that for this particular evaluation, the design pressure will be applied to the normal/upset (AlB) allowable tlaw tables. This is conservative as the normal and upset conditions require higher safety factors than the emergency and faulted conditions (C/O).
The membrane stress is a result of the pressure, P and direct load using the nominal dimensions is:
crh= PRrr!(t) = 1423<<7.1875+5.9675)12)/1.22 = 7.67 ksi (14)
The yield and tensile strength of Alloy 182, SA 508 Class 2 and SA 182 Grade F304 at the design temperature of 575°F are obtained from Reference 8 and summarized in Table I along with the flow stresses and stress ratios.
5.4 Allowable Flaw Size Appendix C of the ASME Code, Section Xl [4], provides the procedure for determining the allowable naw sizes in piping requiring a determination of the failure mode as provided in Article C-4000 per Figure C -4210-1. Article C-6400 provides the approach for calculating the allowable axial flaw size
.) using EPFM criteria.
Based on the stress ratios presented in Table I for the different materials, the end of evaluation period allowable naw sizes for service levels A and B conditions are obtained from Table C-6410-1 (since EPFM failure mode governs tor all materials present) and summarized in Table 2.
Based on the final naw depth to thickness ratio of 0.287 in the stainless steel and 0.37 in the low alloy steel and Alloy I ~Q, and the tth'{R mt)=0.79, this is well below the allowable naw size of 75% of wall.
2 In Table C-MI 0-1, it is noted lie 2= 600 in-lb/in in the CL direction. In the screening criteria, the lie is 2
taken as 300 in -lb/in per Table C-8322-1. Using a higher lie in the screening criteria does not change the failure mode for evaluation.
5.5 Fatigue Crack Growth Fatigue crack growth tor the next operating cycle is not signiticant due to a limited number of cycles that result in signiticant stress. For the purpose of this evaluation a representative calculation for the Alloy 182 material is performed. For this evaluation, 5 startup/shutdown cycles per year will be assumed (10 cycles total over 2 years), which conservatively bounds actual plant experience. Even with a ~K of 50 ksi --Jin, the fatigue crack growth rate in the Alloy 1~2 exposed to the water environment is estimated to be 3. 75x 1O*J inches/cycle using data from Reference 9. ~ote that this conservatively uses a high range File No.: 0900257.301 Page 6 of 10
)
Revision: 0 1'0306*01
Plant Hatch Unit 2 2R20 OAR-1 Report, Table 1, Enclosure 1 l)f stress intensity factor and does not include the beneticial etfect of HWC on the fatigue crack growth rate as mentioned in Reference 9. This is also conservative since no credit is taken for the beneficial
) wmpressive residual stress caused by IHSI. Fatigue crack growth in stainless steel and low alloy steel would result in similar conclusions. Thus, fatigue crack growth, even with the conservative assumption on the number of signi ticant cycles, is small and does not change the final flaw depth significantly.
6.0 CO~CLL510~5 A tlaw evaluation has been performed assuming that the tlaw is an active stress corrosion Haw for an operating period of 2 years. Fatigue crack growth was also evaluated and was determined to be very small. Results of this evaluation demonstrate that growth of the observed tlaw, will remain less than the allowable tlaw size of75% of the pipe wall. Therefore, the required safety factors will be maintained during operation with this tlaw over the next operating cycle.
7.0 REFERE~CES I. GE-Hitachi Examination Summary Sheet, Report No. H2R20-APR-014, 51 File 0900257.202.
- 2. E-mail, Robin Dyle to Marcos Herrera, Subject RE:2B31-1 RC-12AR-G-5, February 21,2009, 3:23pm. 51 File 0900257.201.
- 3. E-mail, Robin Dyle to Marcos Herrera, Subject RE:2B31-IRC-12AR-G-5, February 22,2009, 5:30pm. SI File 0900257.205.
- 4. ASME, Boiler and Pressure Vessel Code,Section XI, 2001 Edition with 2003 Addenda.
) 5. Combustion Engineering Inc. Report CENC-L232, ApriL L975. "Analytical Report for Hatch No.2 Reactor Vessel tor Georgia Power Company". SI File 0800742.2 L5.
- 6. BWRVIP-59A, BWR Vessel and Internals Project, Evaluation of Crack Growth in BWR Nickel Base Austenitic Alloy in RPV Internals,' EPRI Report TR-I 08710, Palo Alto, CA 1998.
- 7. BWRVIP-14A, BWR Vessel and Internals Project, Evaluation of Crack Growth in BWR Stainless Steel RPV Internals (BWRVLP-L4-A), EPRI Report TR-105873-A, Final Report, March 2003.
- 8. ASME, Boiler and Pressure Vessel Code, Section LI, Part D, 2001 Edition with 2003 Addenda.
- 9. NUREG/CR-672I, "Effects of Alloy Chemistry, Cold Work, and Water Chemistry on Corrosion Fatigue and Stress Corrosion Cracking of Nickel Alloys and Welds," U.S. Nuclear Regulatory Commission (Argonne National Laboratory), April 200 I.
File No.: 0900257.301 Page 7 of 10
)
Revision: 0 F030h-OJ
Plant Hatch Unit 2 2R20 OAR-1 Report, Table 1, Enclosure 1 Table 1: Stress Ratio Results
)
Alloy 182 SA 508 Class 2 SA 182 Grade F304 Temperature (oF) I 575 575 575 Stress Intensity Sm (ksi) 23.3 26.7 16.83 Yield Strength Sv (ksi) 30.0 42.38 18.65 Tensile Strength SUI (ksi) 80.0 80.0 59.2 Flow Strength Sr(ksi) 55.0 61.19 38.93 (Jh= PRm/t (ksi) 7.67 7.67 7.67 Stress Ratio (Jh ; Sr 0.1395 0.1253 0.1970 Table 2: Allowable End of Evaluation Period Flaw Depth to Thickness Ratio (a/f) for Axial Flaws Allowable alt End of Material Stress Ratio Nondimensional Flaw Length ftj'-J(Rmt)
Cycle alt
! 0.6 0.8 0.79 Alloy 182 0.1395 0.75 0.75 0.75 0.28 SA 508 Class 2 0.1253 i 0.75 0.75 0.75 0.28 SA IlQ Grade 304 0.1970 0.75 0.75 0.75 0.37
) File No.: 0900257.301 Page 8 of 10 Revision: 0 FOJ06*01
Plant Hatch Unit 2 2R20 OAR-1 Report, Table 1, Enclosure 1 JOITRATD IoflJ .. ~
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) File No.: 0900257.301 Page 9 of 10 Revision: 0 FOJ06-0J
Plant Hatch Unit 2 2R20 OAR-1 Report, Table 1, Enclosure 1
)
Safe End
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Observed Flaw Figure 2: Predicted Crack Growth Configuration
) File No.: 0900257.301 PagelDoflO Revision: 0 1'0)06*01
Repair Flaw or Relevant Condition Replacement Found during Scheduled Repair/
Date Code Corrective Item Section XI Examination or Replacement Complete Class Measure Description Description of Work Test (Yes/No) Plan Number 3 Replacement SRV Tee Quencher Su Replaced all 44 SRV tee quencher support bolts within No 3/ I0/2009 2090432101 the Torus due to failure of2 bolts. Failure mechanism was determined to be hydrogen embrittlement and was exacerbated by stress with misalignment contributing (at least at one connection). Reference CR 2009101926 and 2009101935. MPL 2T23.
2 Replacement Pipe Hanger U-Bolt Replaced broken u-bolt on hanger due to adjacent No 3/10/2009 209038090 I broken support 2EII-RHR-H82. Support was broken due to system transient. Reference CR 2009101161.
TlMdaJ. July 28. 2009 '-ge1of7
Repair Flaw or Relevant Condition Replacement Found during Scheduled Repair/
Date Replacement Code Corrective Item Section XI Examination or Complete Class Measure Description Description of Work Test (Yes/No) Plan Number 2 Replacement Pipe Hanger Replaced hanger struts, base plates, end attachments. No 3/11/2009 2090342901 and pins due to discovered failed condition of support 2EII-RHR-H82. Reference CR 2009101161. The failure is due to a system transient, likely a dynamic event due to gas voiding at a piping high point without venting.
3 Repair 14" Check Valve Machined valve body hinge pin holes for RHRSW No 11/20/2008 2081399701 discharge check valve 2£ II-F005D to facilitate the addition of bushings. The bushings are being added to compensate for the larger hole being machined due to excessive wear between the hinge pin and valve body.
The addition of bushings is not a design change and is permissable per vendor manual SX25652, page 194.
The max depth for boring to not encroach upon threaded area of hinge pin plugs or be greater than 1/4" radially.
T..-day. July 28. 2001 P8ge2of7
Repair Flaw or Relevant Condition Replacement Found during Scheduled Repair/
Code Corrective Date Item Section XI Examination or Replacement Class Measure Complete Description Description of Work Test (Yes/No) Plan Number 3 Replacement RHRSW 18" Piping S Modified support 2EII-RSW-R 15 per improved No 5/29/2008 2080663404 design in accordance with DCP 2080663404 due to failure. Failure was caused due to vibrational loads of RHRSW piping due to operational conditions caused by flow control valves and operation of multiple pumps in single loop of RHRSW piping. Acceptable per IWA-4223.
3 Replacement 4" Gate Valve Replaced valve 2P41-F 1158 due to internals No 2/28/2009 2070532501 degradation noted during 42 IT-TET-O 12-2 examination. Reference CR 2007\02073. Acceptable per IWA-4223(a).
T. . . . . July a 2008 Page3of7
Repair Flaw or Relevant Condition Replacement Found during Scheduled Repair/
Date Replacement Code Corrective Item Section XI Examination or Complete Class Measure Description Description of Work Test (Yes/No) Plan Number 3 Replacement 4" Gate Valve Replaced valve 2P4 I-F 1176 due to noted degradation No 317/2009 207052330J noted during internals examination per OE 16653.
Reference CR 20071 02086. Acceptable per IWA 4224.I(a).
3 Repair 18" Check Valve Repaired body gasket seating surface of valve 2P41 No 4/23/2009 2062407201 F311 D due to erosion caused by nonnal inservice conditions. Reference CR 2006110988.
T~. July 28. 200t hge4of7
Repair Flaw or Relevant Condition Replacement Found during Scheduled Repair/
Code Corrective Date Item Section XI Examination or Replacement Class Measure Complete Description Description of Work Test (Yes/No) Plan Number 3 Replacement PSW Strainer Replaced bolting in PSW strainer to backwash piping No 3/912009 2061671601 flange due to noted degradation due to normal inservicc conditions. Acceptable per IWA-4224. I(a).
Reference CR 2009102285. MPL 2P41-DOOIB.
3 Replacement Pipe and AOV, 3" Replaced valve 2P4I-F036A and adjacent piping due No 12/12/2007 2050001804 to erosion in valve body and corrosion products in piping. Reference CR 20071 10914. Acceptable per IWA-4223(a) and IWA-4224.1 (a).
TundIIy, July 2B. 2001 Page50f7
Repair Replacement Flaw or Relevant Condition Code Corrective Found during Scheduled Repair/
Item Section XI Examination or Date Class Measure Replacement Description Description of Work Complete Test (Yes/No) Plan Number 3 Replacement Rigid Strut, \8" Piping Replace rigid strut 2Ell-RSW-R25 which has broken No 6/23/2004 2041499601 paddle. Reference CR 2004106672 for documentation of broken paddle, and CR 2007106393 for lost WO package. Acceptable per IWA-4223(a).
3 Replacement Ball Valve, 3" Replaced valve to end piece bolting (studs/nuts) for No 10/5/2007 2040852001 valve 2P41-F I 191 B due to corrosion. New bolting to be corrosion resistant. Reference ED 1071287701 and CR 2004100584.
Tueeday, July 21, 2001
"'8017
Repair Flaw or Relevant Condition Replacement Found during Scheduled Repairl Date Code Corrective Item Section XI Examination or Replacement Complete Class Measure Description Description of Work Test (Yes/No) Plan Number 3 Replacement 3" Ball Valve Replace body to end piece bolting (studs and nuts) for No 11/9/2007 2040152901 valve 2P41-F 1191 A due to corrosion. New bolting material selected due to being corrosion resistant.
Reference CR 2004100584 and ED 1071287701.
Acceptable per IWA-4224. J (a).
TtMday, July 28, 200t Page 7 on