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Category:Letter type:JAFP
MONTHYEARJAFP-23-0065, License Amendment Request to Revise Technical Specifications to Adopt TSTF-529, Clarify Use and Application Rules, Revision 4, and Administrative Changes to the Technical Specifications2023-12-14014 December 2023 License Amendment Request to Revise Technical Specifications to Adopt TSTF-529, Clarify Use and Application Rules, Revision 4, and Administrative Changes to the Technical Specifications JAFP-23-0069, Supplemental Response to Part 73 Exemption Request Withdrawal of Request for Exemption from 10 CFR 73, Subpart B, Preemption Authority Requirements2023-12-0707 December 2023 Supplemental Response to Part 73 Exemption Request Withdrawal of Request for Exemption from 10 CFR 73, Subpart B, Preemption Authority Requirements JAFP-23-0057, and Independent Spent Fuel Storage Installation - Request for Exemption from Enhanced Weapons, Firearms Background Checks, and Security Event Notifications Implementation2023-11-22022 November 2023 and Independent Spent Fuel Storage Installation - Request for Exemption from Enhanced Weapons, Firearms Background Checks, and Security Event Notifications Implementation JAFP-23-0064, Emergency Plan Document Revision2023-11-15015 November 2023 Emergency Plan Document Revision JAFP-23-0063, Registration of Spent Fuel Cask Use2023-11-13013 November 2023 Registration of Spent Fuel Cask Use JAFP-23-0059, Registration of Spent Fuel Cask Use2023-10-24024 October 2023 Registration of Spent Fuel Cask Use JAFP-23-0048, Supplemental Information for License Amendment Request to Update the Technical Specification Bases to Change the Fuel Handling Accident Analysis2023-08-31031 August 2023 Supplemental Information for License Amendment Request to Update the Technical Specification Bases to Change the Fuel Handling Accident Analysis JAFP-23-0050, Physical Security Plan, Revision 242023-08-31031 August 2023 Physical Security Plan, Revision 24 JAFP-23-0047, Correction to the 2022 Annual Radioactive Effluent Release Report2023-08-30030 August 2023 Correction to the 2022 Annual Radioactive Effluent Release Report JAFP-23-0040, License Amendment Request to Update the Technical Specification Bases to Change the Fuel Handling Accident Analysis2023-08-0303 August 2023 License Amendment Request to Update the Technical Specification Bases to Change the Fuel Handling Accident Analysis JAFP-23-0043, 10 CFR 50.46 Annual Report2023-07-31031 July 2023 10 CFR 50.46 Annual Report JAFP-23-0038, License Amendment Request to Modify Technical Specification Surveillance Requirement (SR) 3.4.3.1 Safety Relief Valves (S/Rvs) Setpoint Lower Tolerance2023-07-28028 July 2023 License Amendment Request to Modify Technical Specification Surveillance Requirement (SR) 3.4.3.1 Safety Relief Valves (S/Rvs) Setpoint Lower Tolerance JAFP-23-0033, License Amendment Request - Technical Specifications (TS) Section 3.3.1.2, Source Range Monitors (SRM) Instrumentation2023-06-28028 June 2023 License Amendment Request - Technical Specifications (TS) Section 3.3.1.2, Source Range Monitors (SRM) Instrumentation JAFP-23-0025, 2022 Annual Radiological Environmental Operating Report2023-05-10010 May 2023 2022 Annual Radiological Environmental Operating Report JAFP-23-0023, 2022 Annual Radioactive Effluent Release Report2023-04-27027 April 2023 2022 Annual Radioactive Effluent Release Report JAFP-23-0010, 2022 REIRS Transmittal of NRC Form 52023-03-20020 March 2023 2022 REIRS Transmittal of NRC Form 5 JAFP-23-0008, Supplement to Inservice Inspection Summary Report Cycle 252023-02-22022 February 2023 Supplement to Inservice Inspection Summary Report Cycle 25 JAFP-22-0053, Inservice Inspection Summary Report Cycle 252022-12-20020 December 2022 Inservice Inspection Summary Report Cycle 25 JAFP-22-0046, Core Operating Limits Report Cycle 262022-10-17017 October 2022 Core Operating Limits Report Cycle 26 JAFP-22-0040, 10 CFR 50.46 Annual Report2022-07-29029 July 2022 10 CFR 50.46 Annual Report JAFP-22-0033, Core Operating Limits Report Mid-Cycle 252022-06-23023 June 2022 Core Operating Limits Report Mid-Cycle 25 JAFP-22-0032, Response to Request for Additional Information for James A. FitzPatrick Nuclear Power Plant to Adopt Risk Informed Completion Times TSTF-505, Revision 2, Provide Risk-Informed Extended Completion Times - RITSTF Initiative 4b..2022-06-16016 June 2022 Response to Request for Additional Information for James A. FitzPatrick Nuclear Power Plant to Adopt Risk Informed Completion Times TSTF-505, Revision 2, Provide Risk-Informed Extended Completion Times - RITSTF Initiative 4b.. JAFP-22-0030, Oswego County and New York State Participation in the Emergency Plan2022-05-13013 May 2022 Oswego County and New York State Participation in the Emergency Plan JAFP-22-0029, 2021 Annual Radiological Environmental Operating Report2022-05-11011 May 2022 2021 Annual Radiological Environmental Operating Report JAFP-22-0028, 2021 Annual Radioactive Effluent Release Report2022-04-27027 April 2022 2021 Annual Radioactive Effluent Release Report JAFP-22-0026, 2021 REIRS Transmittal of NRC Form 52022-04-0707 April 2022 2021 REIRS Transmittal of NRC Form 5 JAFP-22-2020, Supplemental Information No. 1 for James A. FitzPatrick Nuclear Power Plant to Adopt Risk Informed Completion Times TSTF-505, Revision 2, Provide Risk-Informed Extended Completion Times - RITSTF Initiative 4b. and 10CFR 50.69, Risk-Info2022-03-0404 March 2022 Supplemental Information No. 1 for James A. FitzPatrick Nuclear Power Plant to Adopt Risk Informed Completion Times TSTF-505, Revision 2, Provide Risk-Informed Extended Completion Times - RITSTF Initiative 4b. and 10CFR 50.69, Risk-Informed JAFP-22-0017, Amendments to Indemnity Agreements2022-02-15015 February 2022 Amendments to Indemnity Agreements JAFP-22-0007, Summary of Changes to Exelon Generation Company, LLC, Quality Assurance Topical Report, NO-AA-10, and Decommissioning Quality Assurance Program, NO-DC-102022-01-31031 January 2022 Summary of Changes to Exelon Generation Company, LLC, Quality Assurance Topical Report, NO-AA-10, and Decommissioning Quality Assurance Program, NO-DC-10 JAFP-22-0010, Supplemental Information in Response to Order Consenting to License Transfers and Approval of Draft Conforming License Amendments2022-01-24024 January 2022 Supplemental Information in Response to Order Consenting to License Transfers and Approval of Draft Conforming License Amendments JAFP-22-0008, Response to Request for Supplemental Information by the Office of Nuclear Reactor Regulation to Support Review of a License Amendment Request to Eliminate Selected Response Time Testing for Reactor Protection System and Primary2022-01-14014 January 2022 Response to Request for Supplemental Information by the Office of Nuclear Reactor Regulation to Support Review of a License Amendment Request to Eliminate Selected Response Time Testing for Reactor Protection System and Primary JAFP-21-0093, Propose Change to Eliminate Selected Response Time Testing for Reactor Protection System and Primary Containment Isolation Instrumentation2021-10-18018 October 2021 Propose Change to Eliminate Selected Response Time Testing for Reactor Protection System and Primary Containment Isolation Instrumentation JAFP-21-0089, Application to Revise Technical Specifications to Adopt TSTF-541 Revision 2, Add Exceptions to Surveillance Requirements for Valves and Dampers Locked in the Actuated Position2021-09-27027 September 2021 Application to Revise Technical Specifications to Adopt TSTF-541 Revision 2, Add Exceptions to Surveillance Requirements for Valves and Dampers Locked in the Actuated Position JAFP-21-0087, Response to Request for Additional Information Regarding Request for Approval of Transfer of Licenses and Conforming Amendments2021-09-16016 September 2021 Response to Request for Additional Information Regarding Request for Approval of Transfer of Licenses and Conforming Amendments JAFP-21-0083, Notification of Readiness for NRC 95001 Inspection2021-09-0909 September 2021 Notification of Readiness for NRC 95001 Inspection JAFP-21-0081, Supplement to Application to Revise Technical Specifications to Adopt TSTF-582, Revision 0, Reactor Pressure Vessel Water Inventory Control (RPV WIC) Enhancements2021-09-0303 September 2021 Supplement to Application to Revise Technical Specifications to Adopt TSTF-582, Revision 0, Reactor Pressure Vessel Water Inventory Control (RPV WIC) Enhancements JAFP-21-0075, Proposed Relief Request Associated with Reactor Pressure Vessel Water Level Instrumentation Partial Penetration Nozzle Repairs2021-08-12012 August 2021 Proposed Relief Request Associated with Reactor Pressure Vessel Water Level Instrumentation Partial Penetration Nozzle Repairs JAFP-21-0073, Response to Request for Additional Information Regarding Application to Revise the James A. FitzPatrick Nuclear Power Plant Limiting Condition for Operation (LCO) 3.5.1, ECCS - Operating Surveillance.2021-08-0909 August 2021 Response to Request for Additional Information Regarding Application to Revise the James A. FitzPatrick Nuclear Power Plant Limiting Condition for Operation (LCO) 3.5.1, ECCS - Operating Surveillance. JAFP-21-0069, Application to Adopt 10 CFR 50.69, Risk-informed Categorization and Treatment of Structures, Systems and Components for Nuclear Power Reactors2021-07-30030 July 2021 Application to Adopt 10 CFR 50.69, Risk-informed Categorization and Treatment of Structures, Systems and Components for Nuclear Power Reactors JAFP-21-0070, 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 4b2021-07-30030 July 2021 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 JAFP-21-0071, 10 CFR 50.46 Annual Report2021-07-29029 July 2021 10 CFR 50.46 Annual Report JAFP-21-0064, Supplemental Information - Proposed Alternative Concerning ASME Section XI Repair/Replacement Documentation for Replacement of Pressure Retaining Bolting2021-07-0707 July 2021 Supplemental Information - Proposed Alternative Concerning ASME Section XI Repair/Replacement Documentation for Replacement of Pressure Retaining Bolting JAFP-21-0053, Emergency License Amendment Request - One Time Extension to TS 3.5.1 Condition a, TS 3.6.1.9 Condition a, and TS 3.6.4.1 Condition a Completion Time to Support Residual Heat Removal (RHR) Pump Motor Replacement2021-06-14014 June 2021 Emergency License Amendment Request - One Time Extension to TS 3.5.1 Condition a, TS 3.6.1.9 Condition a, and TS 3.6.4.1 Condition a Completion Time to Support Residual Heat Removal (RHR) Pump Motor Replacement JAFP-21-0052, Emergency License Amendment Request, One Time Extension of Condition a to TS 3.5.1, TS 3.6.1.9, and TS 3.6.4.1 Completion Time to Support Residual Heat Removal (RHR) Pump Motor Replacement2021-06-13013 June 2021 Emergency License Amendment Request, One Time Extension of Condition a to TS 3.5.1, TS 3.6.1.9, and TS 3.6.4.1 Completion Time to Support Residual Heat Removal (RHR) Pump Motor Replacement JAFP-21-0051, Emergency License Amendment Request - One Time Extension to TS 3.5.1 Condition a, TS 3.6.1.9 Condition a, and TS 3.6.4.1 Condition a Completion Time to Support Residual Heat Removal (RHR) Pump Motor Replacement2021-06-13013 June 2021 Emergency License Amendment Request - One Time Extension to TS 3.5.1 Condition a, TS 3.6.1.9 Condition a, and TS 3.6.4.1 Condition a Completion Time to Support Residual Heat Removal (RHR) Pump Motor Replacement JAFP-21-0050, Emergency License Amendment Request, One Time Extension of Condition a to TS 3.5.1, TS 3.6.1.9, and TS 3.6.4.1 Completion Time to Support Residual Heat Removal (RHR) Pump Motor Replacement2021-06-12012 June 2021 Emergency License Amendment Request, One Time Extension of Condition a to TS 3.5.1, TS 3.6.1.9, and TS 3.6.4.1 Completion Time to Support Residual Heat Removal (RHR) Pump Motor Replacement JAFP-21-0044, Response to Request for Additional Information Regarding Request for Approval of Transfer of Licenses and Conforming Amendments2021-06-11011 June 2021 Response to Request for Additional Information Regarding Request for Approval of Transfer of Licenses and Conforming Amendments JAFP-21-0042, Reply to a Notice of Violation; EA-20-1382021-06-0303 June 2021 Reply to a Notice of Violation; EA-20-138 JAFP-21-0041, Supplement to Application to Revise Technical Specifications to Adopt TSTF-545, Revision 3, TS Inservice Testing Program Removal & Clarify SR Usage Rule Application to Section 5.5 Testing2021-05-17017 May 2021 Supplement to Application to Revise Technical Specifications to Adopt TSTF-545, Revision 3, TS Inservice Testing Program Removal & Clarify SR Usage Rule Application to Section 5.5 Testing JAFP-21-0040, Application to Revise Technical Specifications to Adopt TSTF-264-A, Revision 0, 3.3.9 and 3.3.10 - Delete Flux Monitors Specific Overlap Requirement SRs2021-05-14014 May 2021 Application to Revise Technical Specifications to Adopt TSTF-264-A, Revision 0, 3.3.9 and 3.3.10 - Delete Flux Monitors Specific Overlap Requirement SRs 2023-08-31
[Table view] Category:Report
MONTHYEARRS-21-001, Revised Proposed Alternative to Utilize Code Cases N-878 and N-880 for Carbon Steel Piping2021-01-0404 January 2021 Revised Proposed Alternative to Utilize Code Cases N-878 and N-880 for Carbon Steel Piping ML20303A1752020-10-23023 October 2020 Proposed Relief Request from Section XI Repair/Replacement Documentation for Bolting Replacement of Pressure Retaining Bolting ML19301A5052019-08-13013 August 2019 JAF-CALC-NBI-00205 Setpoint Calculation for Vessel Lo-Lo Lvl Primary Containment Isolation Function (02-3LT-57A,B and 02-3STU-258A,B) ML18019A2692018-02-12012 February 2018 Flood Hazard Mitigation Strategies Assessment JAFP-17-0085, High Frequency Confirmation Report for March 12, 2012, Information Request Pursuant to 10 CFR 50.54(f) Regarding Recommendation 2.1, Seismic2017-08-30030 August 2017 High Frequency Confirmation Report for March 12, 2012, Information Request Pursuant to 10 CFR 50.54(f) Regarding Recommendation 2.1, Seismic JAFP-17-0084, Report of Full Compliance with March 12, 2012, Commission Order to Modify Licenses with Regard to Reliable Spent Fuel Pool Instrumentation (Order Number EA-12-051)2017-08-29029 August 2017 Report of Full Compliance with March 12, 2012, Commission Order to Modify Licenses with Regard to Reliable Spent Fuel Pool Instrumentation (Order Number EA-12-051) JAFP-17-0078, Focused Evaluation Summary Pursuant to 10 CFR 50.54(f) Request for Information Regarding Recommendation 2.1: Flooding of the Near-Term Task Force Review of Insights from the Fukushima Dai-ichi Accident2017-07-27027 July 2017 Focused Evaluation Summary Pursuant to 10 CFR 50.54(f) Request for Information Regarding Recommendation 2.1: Flooding of the Near-Term Task Force Review of Insights from the Fukushima Dai-ichi Accident JAFP-17-0054, James a FitzPatrick - Submittal of Stress Analysis Summary for the 24-10-130 Weld Overlay2017-05-25025 May 2017 James a FitzPatrick - Submittal of Stress Analysis Summary for the 24-10-130 Weld Overlay ML17110A2742017-04-20020 April 2017 Proposed Alternative to Utilize Code Case N-513-4, Evaluation Criteria for Temporary Acceptance of Flaws in Moderate Energy Class 2 or 3 Piping Section XI Division 1 ML16043A4112016-02-18018 February 2016 Staff Assessment of Information Provided Pursuant to Title 10 of the Code of Federal Regulations Part 50 Section 50.54(f) Seismic Hazard Reevaluations for Recommendation 2.1 of the Near-Term Task Force Review JAFP-15-0129, Submittal of 10 CFR 71.95 Report Involving 8-120B Cask2015-11-0606 November 2015 Submittal of 10 CFR 71.95 Report Involving 8-120B Cask JAFP-15-0036, J.A. Fitzpatrick Flood Hazard Reevaluation Report - Response to NRC Request for Information Per 10 CFR 50.54(f) Regarding the Flooding Aspects of Recommendation 2.1 of the Near-Term Task Force Review of Insights from the Fuskushima Dai-2015-03-12012 March 2015 J.A. Fitzpatrick Flood Hazard Reevaluation Report - Response to NRC Request for Information Per 10 CFR 50.54(f) Regarding the Flooding Aspects of Recommendation 2.1 of the Near-Term Task Force Review of Insights from the Fuskushima Dai-ichi ML15007A0902015-02-12012 February 2015 Interim Staff Evaluation Relating to Overall Integrated Plan in Response to Order EA-13-109 (Severe Accident Capable Hardened Vents) ML15007A1502015-01-0707 January 2015 Seismic Walkdown Report Update - Entergy'S Response to NRC Request for Information Pursuant to 10 CFR 50.54(f) Re the Seismic Aspects of Recommendation 2.3 of the Near-Term Task Force Review of Insights from the Fukushima Dai-ichi Accident. JAFP-15-0007, Seismic Walkdown Report Update - Entergy'S Response to NRC Request for Information Pursuant to 10 CFR 50.54(f) Re the Seismic Aspects of Recommendation 2.3 of the Near-Term Task Force Review of Insights from the Fukushima Dai-ichi Accid2015-01-0707 January 2015 Seismic Walkdown Report Update - Entergy'S Response to NRC Request for Information Pursuant to 10 CFR 50.54(f) Re the Seismic Aspects of Recommendation 2.3 of the Near-Term Task Force Review of Insights from the Fukushima Dai-ichi Accident. JAFP-14-0143, Expedited Seismic Evaluation Process Report (CEUS Sites), Response NRC Request for Information Pursuant to 10 CFR 50.54(f) Regarding Recommendation 2.1 of the Near-Term Task Force Review of Insights from The.2014-12-30030 December 2014 Expedited Seismic Evaluation Process Report (CEUS Sites), Response NRC Request for Information Pursuant to 10 CFR 50.54(f) Regarding Recommendation 2.1 of the Near-Term Task Force Review of Insights from The. JAFP-14-0094, 10 CFR 50.46 Annual Report - Changes and Errors in Emergency Core Cooling System Evaluation Models2014-08-0101 August 2014 10 CFR 50.46 Annual Report - Changes and Errors in Emergency Core Cooling System Evaluation Models 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 ML14073A1552014-04-0101 April 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 ML14007A6812014-02-21021 February 2014 Interim Staff Evaluation Relating to Overall Integrated Plan in Response to Order EA-12-049 (Mitigation Strategies) ML14043A4282014-02-15015 February 2014 Mega-Tech Services, LLC, Technical Evaluation Report Regarding the Overall Integrated Plan for James A. FitzPatrick Nuclear Power Plant, TAC Nos.: MF1077 ML13247A0362013-08-20020 August 2013 Letter from R. Dale Roberts 10 CFR 71.95 Report on the 8-120B Cask ML13273A6912013-08-20020 August 2013 Redacted - Letter from R. Dale Roberts 10 CFR 71.95 Report on the 8-120B Cask JAFP-13-0096, James a FitzPatrick Nuclear Power Plant, 10 CFR 50.46 Annual Report - Changes and Errors in Emergency Core Cooling System Evaluation Models2013-08-0101 August 2013 James a FitzPatrick Nuclear Power Plant, 10 CFR 50.46 Annual Report - Changes and Errors in Emergency Core Cooling System Evaluation Models JAFP-13-0035, Response to Request for Additional Information on Core for Plate Rim Hold Down Bolting, Plant Specific Analysis and Inspection Plan, License Renewal Commitment 232013-04-17017 April 2013 Response to Request for Additional Information on Core for Plate Rim Hold Down Bolting, Plant Specific Analysis and Inspection Plan, License Renewal Commitment #23 JAFP-12-0134, FitzPatrick - Seismic Walkdown Report, Response to NRC Request for Information Per 10 CFR 50.54 (F) Re the Seismic Aspect of Recommendation 2.3 of the Near-Term Task Force Review of Insights from the Fukushima Dai-ichi Accident2012-11-27027 November 2012 FitzPatrick - Seismic Walkdown Report, Response to NRC Request for Information Per 10 CFR 50.54 (F) Re the Seismic Aspect of Recommendation 2.3 of the Near-Term Task Force Review of Insights from the Fukushima Dai-ichi Accident ML12342A1402012-11-27027 November 2012 Seismic Walkdown Report, Response to NRC Request for Information Per 10 CFR 50.54 (F) Re the Seismic Aspect of Recommendation 2.3 of the Near-Term Task Force Review of Insights from the Fukushima Dai-ichi Accident JAFP-12-0135, Flooding Walkdown Report - Entergys Response to NRC Request for Information Pursuant to 10CFR50.54(f) Flooding Aspects of Recommendation 2.3 of Near-Term Task Force Review of Insights from the Fukushima Dai-ichi Accident2012-11-27027 November 2012 Flooding Walkdown Report - Entergys Response to NRC Request for Information Pursuant to 10CFR50.54(f) Flooding Aspects of Recommendation 2.3 of Near-Term Task Force Review of Insights from the Fukushima Dai-ichi Accident JAFP-12-0134, Engineering Report, JAF-RPT-12-00015, Rev. 0, Jaf Seismic Walkdown Report for Resolution of Fukushima Near-Term Task Force Recommendation 2.3 Seismic, Book 1 of 32012-11-21021 November 2012 Engineering Report, JAF-RPT-12-00015, Rev. 0, Jaf Seismic Walkdown Report for Resolution of Fukushima Near-Term Task Force Recommendation 2.3 Seismic, Book 1 of 3 JAFP-12-0134, Engineering Report JAF-RPT-12-00015, Rev. 0, Jaf Seismic Walkdown Report for Resolution of Fukushima Near-Term Task Force Recommendation 2.3 Seismic, Book 1 of 3, Attachment C, Pages 74 Through 5732012-11-21021 November 2012 Engineering Report JAF-RPT-12-00015, Rev. 0, Jaf Seismic Walkdown Report for Resolution of Fukushima Near-Term Task Force Recommendation 2.3 Seismic, Book 1 of 3, Attachment C, Pages 74 Through 573 ML12342A1392012-11-21021 November 2012 Engineering Report JAF-RPT-12-00015, Rev. 0, Jaf Seismic Walkdown Report for Resolution of Fukushima Near-Term Task Force Recommendation 2.3 Seismic, Book 1 of 3, Attachment C, Pages 74 Through 573 ML12342A1382012-11-21021 November 2012 Engineering Report, JAF-RPT-12-00015, Rev. 0, Jaf Seismic Walkdown Report for Resolution of Fukushima Near-Term Task Force Recommendation 2.3 Seismic, Book 1 of 3 JAFP-12-0134, Engineering Report JAF-RPT-12-00015, Rev. 0, Jaf Seismic Walkdown Report for Resolution of Fukushima Near-Term Task Force Recommendation 2.3 Seismic, Book 2 of 3, Attachment C, Pages 283 Through 5732012-11-21021 November 2012 Engineering Report JAF-RPT-12-00015, Rev. 0, Jaf Seismic Walkdown Report for Resolution of Fukushima Near-Term Task Force Recommendation 2.3 Seismic, Book 2 of 3, Attachment C, Pages 283 Through 573 ML12342A1422012-11-21021 November 2012 Engineering Report JAF-RPT-12-00015, Rev. 0, Jaf Seismic Walkdown Report for Resolution of Fukushima Near-Term Task Force Recommendation 2.3 Seismic, Book 2 of 3, Attachment C, Pages 551 Through 573 ML13163A2662012-05-18018 May 2012 from Bhalchandra Vaidya to Samson Lee: G20120172, 2.206 Petition ML13162A6272012-02-20020 February 2012 Summary of R1 Containment Vents ML1127000692011-09-26026 September 2011 Enclosure 2, Mfn 10-245 R4, Description of the Evaluation and Surveillance Recommendations for BWR/2-5 Plants ML1125101872011-09-0808 September 2011 NRC Investigation Report Nos. 1-2009-041, 1-2010-019, and 1-2010-031 ML1024601002010-09-0202 September 2010 Enclosure 2, RA-ENO-EP1-10-135, Response to NRC Supplemental RAIs 1, 4, 5, and 6 for James a FitzPatrick Cycle 20 SLMCPR Submittal ML1012704392010-05-0505 May 2010 Y020100187 - List of Historical Leaks and Spills at U.S. Commercial Nuclear Power Plants JAFP-09-0091, 10 CFR 50.46 Annual Report - Changes and Errors in Emergency Core Cooling System (ECCS) Evaluation Models2009-07-31031 July 2009 10 CFR 50.46 Annual Report - Changes and Errors in Emergency Core Cooling System (ECCS) Evaluation Models JAFP-09-0056, Summary of Plant and Independent Spent Fuel Storage Installation Changes, Tests, and Experiments for 2007 and 2008 as Required by 10 CFR 50.59 and 10 CFR 72.482009-05-11011 May 2009 Summary of Plant and Independent Spent Fuel Storage Installation Changes, Tests, and Experiments for 2007 and 2008 as Required by 10 CFR 50.59 and 10 CFR 72.48 JAFP-08-0131, Enclosure 2, Structural Integrity Associates, Inc. Calculation 0800769.311 Crack Growth Analysis for Recirculation Inlet Nozzle N-2C2008-12-0404 December 2008 Enclosure 2, Structural Integrity Associates, Inc. Calculation 0800769.311 Crack Growth Analysis for Recirculation Inlet Nozzle N-2C ML0835903392008-12-0404 December 2008 Enclosure 2, Structural Integrity Associates, Inc. Calculation 0800769.311 Crack Growth Analysis for Recirculation Inlet Nozzle N-2C JAFP-08-0131, Enclosure 3, Structural Integrity Associates, Inc. Calculation 0800769.316 Reconciliation of Recirculation Inlet N2 Nozzle-to-Safe End Weld Overlay Repair with Nozzle Stress Report2008-12-0303 December 2008 Enclosure 3, Structural Integrity Associates, Inc. Calculation 0800769.316 Reconciliation of Recirculation Inlet N2 Nozzle-to-Safe End Weld Overlay Repair with Nozzle Stress Report ML0831006632008-11-19019 November 2008 Request for Additional Information Third 10-Year Interval Inservice Inspection Program Plan Relief Request RR-CRV-01 JAFP-08-0103, Certificate of Compliance Not Met During Receipt of Low-Level Radioactive Waste Shipment at Barnwell, Sc Disposal Facility2008-10-0202 October 2008 Certificate of Compliance Not Met During Receipt of Low-Level Radioactive Waste Shipment at Barnwell, Sc Disposal Facility ML0810102772008-09-0808 September 2008 Renewed Facility Operating License No. DPR-59 / Appendix C - Additional Conditions ML0821900132008-08-0707 August 2008 Monthly Operating Reports Second Quarter 2008 JAFP-08-0067, BWR Vessels and Internals Project, Lntegrated Surveillance Program (ISP) Data for FitzPatrick (Excerpted from BWRVIP-135, Revision 1)2008-07-31031 July 2008 BWR Vessels and Internals Project, Lntegrated Surveillance Program (ISP) Data for FitzPatrick (Excerpted from BWRVIP-135, Revision 1) 2021-01-04
[Table view] Category:Technical
MONTHYEARRS-21-001, Revised Proposed Alternative to Utilize Code Cases N-878 and N-880 for Carbon Steel Piping2021-01-0404 January 2021 Revised Proposed Alternative to Utilize Code Cases N-878 and N-880 for Carbon Steel Piping ML19301A5052019-08-13013 August 2019 JAF-CALC-NBI-00205 Setpoint Calculation for Vessel Lo-Lo Lvl Primary Containment Isolation Function (02-3LT-57A,B and 02-3STU-258A,B) JAFP-17-0085, High Frequency Confirmation Report for March 12, 2012, Information Request Pursuant to 10 CFR 50.54(f) Regarding Recommendation 2.1, Seismic2017-08-30030 August 2017 High Frequency Confirmation Report for March 12, 2012, Information Request Pursuant to 10 CFR 50.54(f) Regarding Recommendation 2.1, Seismic JAFP-17-0078, Focused Evaluation Summary Pursuant to 10 CFR 50.54(f) Request for Information Regarding Recommendation 2.1: Flooding of the Near-Term Task Force Review of Insights from the Fukushima Dai-ichi Accident2017-07-27027 July 2017 Focused Evaluation Summary Pursuant to 10 CFR 50.54(f) Request for Information Regarding Recommendation 2.1: Flooding of the Near-Term Task Force Review of Insights from the Fukushima Dai-ichi Accident JAFP-17-0054, James a FitzPatrick - Submittal of Stress Analysis Summary for the 24-10-130 Weld Overlay2017-05-25025 May 2017 James a FitzPatrick - Submittal of Stress Analysis Summary for the 24-10-130 Weld Overlay JAFP-15-0036, J.A. Fitzpatrick Flood Hazard Reevaluation Report - Response to NRC Request for Information Per 10 CFR 50.54(f) Regarding the Flooding Aspects of Recommendation 2.1 of the Near-Term Task Force Review of Insights from the Fuskushima Dai-2015-03-12012 March 2015 J.A. Fitzpatrick Flood Hazard Reevaluation Report - Response to NRC Request for Information Per 10 CFR 50.54(f) Regarding the Flooding Aspects of Recommendation 2.1 of the Near-Term Task Force Review of Insights from the Fuskushima Dai-ichi ML15007A1502015-01-0707 January 2015 Seismic Walkdown Report Update - Entergy'S Response to NRC Request for Information Pursuant to 10 CFR 50.54(f) Re the Seismic Aspects of Recommendation 2.3 of the Near-Term Task Force Review of Insights from the Fukushima Dai-ichi Accident. JAFP-15-0007, Seismic Walkdown Report Update - Entergy'S Response to NRC Request for Information Pursuant to 10 CFR 50.54(f) Re the Seismic Aspects of Recommendation 2.3 of the Near-Term Task Force Review of Insights from the Fukushima Dai-ichi Accid2015-01-0707 January 2015 Seismic Walkdown Report Update - Entergy'S Response to NRC Request for Information Pursuant to 10 CFR 50.54(f) Re the Seismic Aspects of Recommendation 2.3 of the Near-Term Task Force Review of Insights from the Fukushima Dai-ichi Accident. 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 ML14007A6812014-02-21021 February 2014 Interim Staff Evaluation Relating to Overall Integrated Plan in Response to Order EA-12-049 (Mitigation Strategies) ML14043A4282014-02-15015 February 2014 Mega-Tech Services, LLC, Technical Evaluation Report Regarding the Overall Integrated Plan for James A. FitzPatrick Nuclear Power Plant, TAC Nos.: MF1077 JAFP-13-0035, Response to Request for Additional Information on Core for Plate Rim Hold Down Bolting, Plant Specific Analysis and Inspection Plan, License Renewal Commitment 232013-04-17017 April 2013 Response to Request for Additional Information on Core for Plate Rim Hold Down Bolting, Plant Specific Analysis and Inspection Plan, License Renewal Commitment #23 ML12342A1402012-11-27027 November 2012 Seismic Walkdown Report, Response to NRC Request for Information Per 10 CFR 50.54 (F) Re the Seismic Aspect of Recommendation 2.3 of the Near-Term Task Force Review of Insights from the Fukushima Dai-ichi Accident JAFP-12-0134, FitzPatrick - Seismic Walkdown Report, Response to NRC Request for Information Per 10 CFR 50.54 (F) Re the Seismic Aspect of Recommendation 2.3 of the Near-Term Task Force Review of Insights from the Fukushima Dai-ichi Accident2012-11-27027 November 2012 FitzPatrick - Seismic Walkdown Report, Response to NRC Request for Information Per 10 CFR 50.54 (F) Re the Seismic Aspect of Recommendation 2.3 of the Near-Term Task Force Review of Insights from the Fukushima Dai-ichi Accident ML12342A1392012-11-21021 November 2012 Engineering Report JAF-RPT-12-00015, Rev. 0, Jaf Seismic Walkdown Report for Resolution of Fukushima Near-Term Task Force Recommendation 2.3 Seismic, Book 1 of 3, Attachment C, Pages 74 Through 573 ML12342A1382012-11-21021 November 2012 Engineering Report, JAF-RPT-12-00015, Rev. 0, Jaf Seismic Walkdown Report for Resolution of Fukushima Near-Term Task Force Recommendation 2.3 Seismic, Book 1 of 3 JAFP-12-0134, Engineering Report JAF-RPT-12-00015, Rev. 0, Jaf Seismic Walkdown Report for Resolution of Fukushima Near-Term Task Force Recommendation 2.3 Seismic, Book 2 of 3, Attachment C, Pages 283 Through 5732012-11-21021 November 2012 Engineering Report JAF-RPT-12-00015, Rev. 0, Jaf Seismic Walkdown Report for Resolution of Fukushima Near-Term Task Force Recommendation 2.3 Seismic, Book 2 of 3, Attachment C, Pages 283 Through 573 JAFP-12-0134, Engineering Report JAF-RPT-12-00015, Rev. 0, Jaf Seismic Walkdown Report for Resolution of Fukushima Near-Term Task Force Recommendation 2.3 Seismic, Book 1 of 3, Attachment C, Pages 74 Through 5732012-11-21021 November 2012 Engineering Report JAF-RPT-12-00015, Rev. 0, Jaf Seismic Walkdown Report for Resolution of Fukushima Near-Term Task Force Recommendation 2.3 Seismic, Book 1 of 3, Attachment C, Pages 74 Through 573 JAFP-12-0134, Engineering Report, JAF-RPT-12-00015, Rev. 0, Jaf Seismic Walkdown Report for Resolution of Fukushima Near-Term Task Force Recommendation 2.3 Seismic, Book 1 of 32012-11-21021 November 2012 Engineering Report, JAF-RPT-12-00015, Rev. 0, Jaf Seismic Walkdown Report for Resolution of Fukushima Near-Term Task Force Recommendation 2.3 Seismic, Book 1 of 3 ML12342A1422012-11-21021 November 2012 Engineering Report JAF-RPT-12-00015, Rev. 0, Jaf Seismic Walkdown Report for Resolution of Fukushima Near-Term Task Force Recommendation 2.3 Seismic, Book 2 of 3, Attachment C, Pages 551 Through 573 ML13163A2662012-05-18018 May 2012 from Bhalchandra Vaidya to Samson Lee: G20120172, 2.206 Petition ML1024601002010-09-0202 September 2010 Enclosure 2, RA-ENO-EP1-10-135, Response to NRC Supplemental RAIs 1, 4, 5, and 6 for James a FitzPatrick Cycle 20 SLMCPR Submittal JAFP-08-0131, Enclosure 2, Structural Integrity Associates, Inc. Calculation 0800769.311 Crack Growth Analysis for Recirculation Inlet Nozzle N-2C2008-12-0404 December 2008 Enclosure 2, Structural Integrity Associates, Inc. Calculation 0800769.311 Crack Growth Analysis for Recirculation Inlet Nozzle N-2C ML0835903392008-12-0404 December 2008 Enclosure 2, Structural Integrity Associates, Inc. Calculation 0800769.311 Crack Growth Analysis for Recirculation Inlet Nozzle N-2C JAFP-08-0131, Enclosure 3, Structural Integrity Associates, Inc. Calculation 0800769.316 Reconciliation of Recirculation Inlet N2 Nozzle-to-Safe End Weld Overlay Repair with Nozzle Stress Report2008-12-0303 December 2008 Enclosure 3, Structural Integrity Associates, Inc. Calculation 0800769.316 Reconciliation of Recirculation Inlet N2 Nozzle-to-Safe End Weld Overlay Repair with Nozzle Stress Report ML0821900132008-08-0707 August 2008 Monthly Operating Reports Second Quarter 2008 ML0821004582008-07-31031 July 2008 BWR Vessels and Internals Project, Lntegrated Surveillance Program (ISP) Data for FitzPatrick (Excerpted from BWRVIP-135, Revision 1) JAFP-08-0067, BWR Vessels and Internals Project, Lntegrated Surveillance Program (ISP) Data for FitzPatrick (Excerpted from BWRVIP-135, Revision 1)2008-07-31031 July 2008 BWR Vessels and Internals Project, Lntegrated Surveillance Program (ISP) Data for FitzPatrick (Excerpted from BWRVIP-135, Revision 1) JAFP-07-0125, ENT-FLU-002-R-005, Rev 0, James A. FitzPatrick, License Renewal Application - Amendment 14, Fluence Calculation, Non-Proprietary Version2007-10-31031 October 2007 ENT-FLU-002-R-005, Rev 0, James A. FitzPatrick, License Renewal Application - Amendment 14, Fluence Calculation, Non-Proprietary Version ML0732400252007-08-10010 August 2007 Federal Consistency Overview ML0713504332007-04-0505 April 2007 JAF-RPT-05-AMM30, Rev 4, Draft, Aging Management Review of Nonsafety-Related Systems and Components Affecting Safety-Related Systems, (FitzPatrick) ML0705206612007-02-20020 February 2007 E-mail from Entergy to T. Le & Document - GE-NE-B1100732-01, Rev. 1, Final Report, Class II, February 1998 Plant FitzPatrick RPV Surveillance Materials Testing and Analysis of 120 Degree Capsule at 13.4 EFPY ML0701605692007-01-16016 January 2007 JAFNPP - SEIS Web Reference - Mills 2003: Status of the Lake Ontario Food Web in a Changing Ecosystem ML0701203042006-12-27027 December 2006 JAFNPP - SEIS Web References - Chppm Noise Levels of Army Equipments ML0603903722005-09-30030 September 2005 NEDO-33087, Rev. 1, J.A. Fitzpatrick Nuclear Power Plant Aprm/Rbm/Technical Specifications/Maximum Extended Operating Domain (Arts/Meod). ML0701602142005-07-31031 July 2005 JAFNPP - SEIS Web Reference - Mills Et Al 2005 JAFP-05-0118, Request for Exemption from 10 CFR 50, Appendix R, III.G.2.c Requirement for a One-Hour Rated Fire Barrier Wrap2005-07-27027 July 2005 Request for Exemption from 10 CFR 50, Appendix R, III.G.2.c Requirement for a One-Hour Rated Fire Barrier Wrap ML0701103402005-06-30030 June 2005 EPA Shooting Range Bmps - JAFNPP SEIS Reference BVY 04-088, Fitness-for-Duty Program Performance Report for the Period January - June 20042004-08-30030 August 2004 Fitness-for-Duty Program Performance Report for the Period January - June 2004 JAFP-03-0065, Summary of Plant Changes, Tests, & Experiments for 2001/2002 as Required by 10 CFR 50.59 & 10 CFR 72.48, & Summary of Commitment Changes for 2001/20022003-05-12012 May 2003 Summary of Plant Changes, Tests, & Experiments for 2001/2002 as Required by 10 CFR 50.59 & 10 CFR 72.48, & Summary of Commitment Changes for 2001/2002 JAFP-03-0052, 2002 Personnel Radiation Exposure Summary Report2003-04-15015 April 2003 2002 Personnel Radiation Exposure Summary Report ML0305202692003-02-15015 February 2003 Follow-up Report, Reference Event #39516 ML0636202142002-03-31031 March 2002 JAFNPP Er Ref 8-15 DOE Eia International Energy Outlook 2002 ML0732012262001-01-0101 January 2001 Onondago Reservation Census Population NRC Generic Letter 1979-451979-09-25025 September 1979 NRC Generic Letter 1979-045: Transmittal of Reports Regarding Foreign Reactor Operation Experiences ML17037C2661976-01-23023 January 1976 Letter Regarding a Report Entitled 1974 Nine Mile Point Aquatic Ecology Studies. 2021-01-04
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Exelon Generation Company, LLC Exelon Generation(o James A FitzPatnck NPP POBox110 Lycoming , NY 13093 Tel 315-342-3840 William C. Drews Regulatory Assurance Manager - JAF JAFP-17-0054 May 25, 2017 U.S. Nuclear Regulatory Commission Document Control Desk Washington, D. C. 20555
Subject:
Submittal of Stress Analysis Summary for the 24-10-130 Weld Overlay James A FitzPatrick Nuclear Power Plarit Docket No. 50-333 License No. DPR-59
Reference:
- 1. ENOI letter, Proposed Alternative to ASME Code Requirements for Weld Overlay Repair, RR-21 at James A FitzPatrick Nuclear Power Plant, JAFP-17-0010, dated January 27, 2017
- 2. NRC letter, Summary of January 30, 2017, Teleconference with Entergy Nuclear Operations, Inc. Regarding Verbal Authorization for Relief Request RR-21 for the James A. FitzPatrick Nuclear Power Plant (CAC No. MF9128}, ML17031A280, dated February 8, 2017
- 3. ENOl letter, Submittal of the Final Ultrasonic Test Results on the 24 130 Weld Overlay, JAFP-17-0023, dated March 9, 2017
Dear Sir or Madam:
During Refueling Outage R22, an indication was found by ultrasonic testing (UT) examination in "A" Residual Heat Removal Low Pressure Coolant Injection Loop weld number 24-10-130.
Pursuant to 10 CFR 50.55a(z)(1 ), James A FitzPatrick Nuclear Power Plant (JAF) ,proposed a repair in accordance with Relief Request RR-21 [Reference 1] and included a commitment to submit a stress analysis summary demonstrating that the repaired dissimilar metal weld (DMW) will perform its intended design function after the weld overlay installation. The Enclosure of this letter contains the summary of the stress analysis.
This letter contains no new regulatory commitments. Should you have any questions, please contact me at (315) 349-6562.
Very truly yours, William C. Drews Regulatory Assurance Manager WCD/mh
Enclosure:
Stress Analysis Summary for the 24-10-130 Weld Overlay cc: USNRC, Region I Administrator NYSERDA USNRC, Project Manager NYPSC USNRC, Resident Inspector
JAFP-17-0054 Enclosure Stress Analysis Summary for the 24-10-130 Weld Overlay (14 pages)
11515 Vanstory Dr.
Suite 125 Huntersville, NC 28078 Phone: 704-597-5554 Fax: 704-597-0335 www.structint.com therrmann@structint.com May 24, 2017 Report No. 1700151.402.R0 Quality Program: Nuclear Commercial Keith Schoales James A. FitzPatrick Nuclear Power Plant Exelon Corporation P.O. Box 110 Lycoming, NY 13093-0110
Subject:
Summary of Design and Analysis for Full Structural Weld Overlay Repair of James A. FitzPatrick Nuclear Power Plant (JAF), Loop A RHR LPCI Dissimilar Metal Weld - Tee-to-Gate Valve
Reference:
Entergy Letter No. JAFP-17-0010 dated January 27, 2017, Proposed Alternative to ASME Code Requirements for Weld Overlay Repair, RR-21 at James A.
FitzPatrick Nuclear Power Plant
Dear Mr. Schoales:
This letter report is being provided in support of JAFs response to a commitment in the above-referenced proposed alternative:
Commitment:
JAF will submit to the NRC a stress analysis summary demonstrating that the repaired dissimilar metal weld (DMW) will perform its intended design function after weld overlay installation.
This information will be submitted to the NRC within 90 days of completing JAFs Refueling Outage 22.
The body of the referenced Entergy letter states the stress analysis report will include results showing that the requirements of NB-3200 and/or NB-3600 of the ASME Code,Section III are satisfied. The stress analysis will also include results showing that the requirements of IWB-3000 of the ASME Code,Section XI, are satisfied. The results will show that the as-found defect including its growth will not adversely affect the integrity of the overlaid welds.
Keith Schoales May 24, 2017 Report No. 1700151.402.R0 Page 2 of 2 Attachment A to this letter report provides the stress analysis summary to satisfy this commitment.
If you have any questions or comments regarding this summary, please contact one of the undersigned.
Prepared by: Verified by:
05/24/17 05/24/17 Moses Taylor, Jr. Date Terry J. Herrmann Date Senior Associate Senior Associate Approved by:
05/24/17 Terry J. Herrmann Date Senior Associate Attachment cc: Project File No. 1700151.402.R0 M. Lashley
ATTACHMENT A Summary of Design and Analysis for Full Structural Weld Overlay Repair of James A. FitzPatrick Nuclear Plant, Residual Heat Removal Low Pressure Coolant Injection Loop A Weld Number 24-10-130.
Attachment to 1700151.402.R0 1 of 12
1.0 INTRODUCTION
James A. FitzPatrick Nuclear Plant (JAF) applied a full structural weld overlay (FSWOL) on dissimilar metal weld (DMW) 24-10-130 between the reactor water recirculation (RWR)/ residual heat removal (RHR) low pressure coolant injection (LPCI) Loop A tee-to-RHR gate valve.
The purpose of this overlay is to repair the as-found flaw and to mitigate any potential future intergranular stress corrosion cracking (IGSCC) in this weld. The overlay was installed using an IGSCC resistant weld filler material; Alloy 52M [4].
This report, which satisfies one of the JAF commitments of the Relief Request [1], summarizes the results of the component specific residual stress analyses and the fracture mechanics evaluations, and documents that all ASME Code,Section III stress and fatigue criteria are met. This information is to be submitted within 90 days of completing the refueling outage.
The requirements for design of the weld overlay repair are defined in the Relief Request [1], which is based upon ASME Code Cases N-504-4 [2] and N-638-4 [3], and ASME Code,Section XI, Nonmandatory Appendix Q [5]. The analytical basis for the design of the repair is in accordance with the requirements of ASME Code,Section XI [6], IWB-3640. Weld overlay repairs are acceptable long-term repairs for IGSCC-flawed weldments if they meet a conservative set of design assumptions, which qualify them as full structural weld overlays. The three principal design requirements that qualify a weld overlay as full structural are as follows:
- 1. The design basis for the repair is a circumferentially oriented flaw that extends 360º around the component, and is 100% through the original component wall. This conservative assumption eliminates concerns about IGSCC susceptibility of the original Alloy 82/182 DMW. In addition, potential concerns about the integrity of the original butt weld material are not applicable, since no credit is taken for the load carrying capability of this weld.
- 2. As required by ASME Code,Section XI [6], Appendix C, a combination of internal pressure, deadweight, seismic, and other dynamic stresses is used in the design of a weld overlay repair.
Thermal and other secondary stresses are not required to be included for structural sizing calculations (since the repairs are applied using a GTAW process that produces a high toughness weld deposit), but they are addressed later in subsequent stress, fatigue, and stress corrosion cracking evaluations.
- 3. Following the repair, the surface finish of the overlay must be sufficiently smooth to allow preservice and future inservice ultrasonic examinations through the overlay material and into a portion of the original base metal. The purpose of these examinations is to demonstrate that the overlay design basis does not degrade with time due to flaw propagation.
ASME Code,Section III stress and fatigue usage evaluations are also performed to demonstrate that the overlaid components continue to meet ASME Code,Section III requirements. The original construction Code for the RWR / RHR LPCI Loop A tee-to-gate valve DMW was the B31.1 Power Piping Code
[8]. However, the original construction Code has been reconciled to the ASME Code,Section III editions applicable to the stress and fatigue usage analyses [14], and as allowed by ASME Code, Section Attachment to 1700151.402.R0 2 of 12
XI [6], Subsection IWA-4222, Code Editions and Addenda later than the original construction Code may be used. ASME Code,Section III, 2001 Edition with Addenda through 2003 [7] was used for these analyses. In addition to providing structural reinforcement to the IGSCC susceptible locations with a resistant material, weld overlays have also been shown to produce beneficial residual stresses that mitigate IGSCC in the underlying DMWs. The weld overlay approach has been used to repair stress corrosion cracking in U.S. nuclear plants on hundreds of welds, and there have been no reports of subsequent crack extension after application of weld overlays. Thus, the compressive stresses caused by the weld overlay have been effective in mitigating new crack initiation and/or growth of existing cracks.
Finally, evaluations were performed, based on as-built measurements taken after the overlay was applied, to demonstrate that the overlay meet their design basis requirements, and that they will not have an adverse effect on the balance of the piping systems. These include comparison of overlay dimensions to design dimensions, evaluations of shrinkage and added weight effects on the piping systems.
2.0 ANALYSIS
SUMMARY
AND RESULTS 2.1 Weld Overlay Structural Sizing Calculations ASME Code Case N-504-3 [2], which incorporates ASME Code,Section XI [6], IWB-3640 evaluation methodology, was used to determine the thickness of the overlay. The FSWOL thickness design requirements in ASME Code Case N-504-3 are the same as Code Case N-504-4 which is referenced in the Relief Request [1] as the basis for the weld overlay installation. Equations from ASME Code,Section XI [6], Appendix C, and the maximum stresses at the DMW for any Service Level, were used to determine the design FSWOL thickness.
As stated in Section 1.0, a FSWOL was installed using Alloy 52M filler metal. However, Alloy 52M weld metal has demonstrated sensitivity to certain impurities, such as sulfur, when deposited onto austenitic stainless steel base materials. Therefore, buffer (transitional) layers of austenitic stainless steel filler metal were applied across the austenitic stainless steel base material. The austenitic stainless steel buffer layers are not part of the structural weld overlay thickness defined above.
The weld overlay length must consider: (1) sufficient distance on either side of the defect location to provide for effective load transfer across the defect location, (2) length required for access for preservice and inservice examinations of the overlaid weld, and (3) limitation on the area of the component that can be overlaid.
In accordance with the Relief Request [1], which is based on ASME Code Cases N-504-4 [2], the minimum weld overlay length required for structural reinforcement is the length which will provide adequate load transfer from one side of the flaw to the other. Per Reference [2], this criterion is generally satisfied if the overlay full thickness length extends axially at least 0.75Rt on each side of susceptible material where R and t are the outer radius and nominal wall thickness of the overlaid components, prior to depositing the weld overlay.
For this location, with valve and tee transitions increasing sharply on either side of the butt weld the length of the weld overlay is dictated by its thickness and taper on the adjacent valve and tee. This Attachment to 1700151.402.R0 3 of 12
configuration was demonstrated to achieve adequate load transfer and all Code primary stress intensity allowable values are met with overlays that directly tie into the adjacent component at the design minimum thickness of 0.75 inches. Figure 2-1 provides a general illustration of the analyzed configuration of the FSWOL.
The overlay length and profile must also be such that the required post-FSWOL examination volume can be inspected using Performance Demonstration Initiative (PDI) qualified nondestructive examination (NDE) techniques. The final configuration was reviewed by qualified NDE personnel to ensure that it meets inspectability requirements and the Relief Request [1].
ASME Code Case N-638-4 [3] limits the area of the ferritic valve covered by the FSWOL to below 500 square inches.
The as-built weld overlay thickness and length are provided in Table 2-1. These measurements exceed the minimum required structural design dimensions described above, thereby demonstrating the adequacy of the as-installed repair.
Table 2-1: Post-Weld Overlay As-Built Dimensions Thickness or Location Length Minimum Valve Side 1.127 Thickness (in.) Tee Side 1.055 Minimum Valve Side NA*
Length (in.) Tee Side NA*
- Minimum length dictated by weld overlay thickness and taper on the adjacent valve and tee.
Attachment to 1700151.402.R0 4 of 12
2.2 ASME Code,Section III Stress Analyses Stress intensities for the weld overlaid JAF DMW at the RHR LPCI Loop A tee-to-gate valve were determined from finite element analyses for the various specified load combinations and transients using the ANSYS software package [9]. Linearized stresses were evaluated at various stress locations using 3-dimensional solid models. A typical finite element model showing stress path locations is provided in Figure 2-2. The stress intensities at these locations were evaluated in accordance with ASME Code,Section III [7], Sub-articles NB-3200 and NB-3600, and compared to applicable Code limits. A summary of the stress and fatigue usage comparisons for the most limiting locations is provided in Table 2-2. The stresses and fatigue usage in the weld overlaid DMW are within the applicable Code limits.
Table 2-2: Limiting Stress Results for Weld Weld Load Type Calculated Allowable Combination Level A/B Primary + Secondary (P +Q) (ksi)* 45.0 50.5 ASME Code RHR LPCI Cumulative Usage Factor 0.20 1.00 Fatigue A DMW between Gate Valve and Tee Environmentally Assisted Fatigue Cumulative Usage Factor 0.24 1.00 (EAF)
- - Primary stress acceptance criteria are met via the sizing calculations discussed in Section 2.1.
2.3 Residual Stress Analyses Weld residual stresses for the JAF DMW at the RWR/RHR LPCI Loop A tee-to-gate valve weld overlay were determined by detailed elastic-plastic finite element analyses. The approach used in the analysis has been benchmarked and validated through multiple 2-dimensional and 3-dimensional ANSYS finite element analyses following the guidelines discussed in MRP-316 [10] and MRP-317 [11],
and using the information provided in the problem statements for mock-up specimens [10, 12, 13] and experimental measurements [10]. A three-dimensional, non-axisymmetric finite element model was developed for the overlay configuration. The modeling of weld nuggets used in the analysis to lump the combined effects of several weld beads is illustrated in Figure 2-3. The model simulated an inside surface (ID) repair at the DMW location with a depth of approximately 50% of the original wall thickness. This ID weld repair is intended to meet the guidelines of the MRP-169 Safety Evaluation Report, Section 3.2.2, paragraph three [15], which states, The residual stress analysis assumes a highly unfavorable, pre-overlay residual stress condition which would result from an inside diameter surface weld repair during construction.
Attachment to 1700151.402.R0 5 of 12
An analysis is performed to simulate the welding process of the ID weld repair, the FSWOL welding process (including buffer layers and bridge bead), and finally, a slow heatup to operating temperature and pressure. The analysis consists of a thermal pass to determine the temperature response of the model to each individual lumped weld nugget as it is added in sequence, followed by a non-linear elastic-plastic stress pass to calculate the residual stress due to the temperature cycling from the application of each lumped weld pass. Since residual stress is a function of the welding history, the stress pass for each nugget is applied to the residual stress field induced from all previously applied weld nuggets.
After completion of the weld overlay simulation, the model was allowed to cool to a uniform steady state temperature of 70F, and then five cycles of shake down with normal operating temperature and pressure were applied to stabilize the residual stress fluctuations due to stress distribution caused by normal operating loads.
The resulting residual stresses were evaluated on six paths through the DMW. These path definitions are shown in Figure 2-2. The resulting through wall residual stresses along these paths are shown in Figure 2-4.
2.4 Section XI Crack Growth Analyses The residual stress calculations were then utilized, along with stresses due to applied loadings and thermal transients, to demonstrate that assumed cracks that could be missed by inspections or the as-found cracks in the weld will not exceed the overlay design basis during the ASME Section XI inservice inspection interval due to fatigue or IGSCC. In the fatigue crack growth analysis, the 60-year projected cycles for each applied transient was applied.
Given the short length of the overlay and the overall thickness on either side, the FSWOL is not expected to generate sufficient compression to halt IGSCC growth. Therefore, an initial flaw depth of 100% of the original base metal thickness is conservatively used for computing crack growth for an axial or circumferential flaw in the RWR/RHR LPCI Loop A tee-to-RHR gate valve DMW with a FSWOL.
Since the initial flaw depth (100% of the original base metal thickness) starts at the DMW to FSWOL material interface, crack growth rates for Alloy 52M, applicable to a BWR environment, are used to calculate the propagation in the thickness direction of the postulated cracks. The stress intensity factors are determined at 100% of the original base metal thickness for the circumferential and axial flaw. The analysis results showing the time it takes for an initial axial or circumferential flaw of 100% of the original base metal thickness to grow to 75% of the total thickness, including the FSWOL, are provided in Table 2-3 for the DMW.
Attachment to 1700151.402.R0 6 of 12
Table 2-3: Crack Growth Results Time for Initial Flaw Depth to Flaw Type Reach 75% of Total Thickness (Base Metal + FSWOL)
Circumferential Flaw > 20 years Axial Flaw > 20 years 2.5 Evaluation of As-Built Conditions The Relief Request [1] and Code Case N-504-4 [2] require evaluation of the as-built weld overlays to determine the effects of any changes in applied loads, as a result of weld shrinkage from the entire overlay, on other items in the piping system. These evaluations were performed and are documented separately from this report. Calculations were performed based on design dimensions to confirm that the overlays would not adversely affect critical piping components. Specifically, the effect of the added weight of the overlays on the adjacent piping systems, based on maximum design dimensions, was evaluated and found to be insignificant.
Figure 2-1: General Illustration of Weld Overlay Design Attachment to 1700151.402.R0 7 of 12
0º and 180º Cross Sections 90º and 270º Cross Sections Figure 2-2: Finite Element Model for Section III Stress Evaluation showing Stress Paths (RWR/RHR tee on the left - RHR valve on the right)
Attachment to 1700151.402.R0 8 of 12
Figure 2-3: Finite Element Model for Residual Stress Analysis showing Nuggets used for Welding Simulations (RWR/RHR tee on the left - RHR valve on the right)
Attachment to 1700151.402.R0 9 of 12
Figure 2-4: Residual Stress Results along Stress Paths through the DMW Attachment to 1700151.402.R0 10 of 12
3.0 CONCLUSION
S The design of the JAF weld overlay was performed in accordance with the requirements of the Relief Request [1], which is based on ASME Code Case N-504-4 [2] and N-638-4 [3], and ASME Code,Section XI, Nonmandatory Appendix Q [5]. The weld overlay design is demonstrated to be a long-term repair and provide mitigation of IGSCC in this weld based on the following:
- In accordance with ASME Code Case N-504-4, structural design of the overlay was performed to meet the requirements of ASME Code,Section XI, IWB-3640 based on an assumed circumferential flaw 100% through-wall, and 360° around the original weld. The resulting full structural weld overlay thus restores the original safety margins of the original weld, with no credit taken for the underlying IGSCC-susceptible material.
- The weld metal used for the overlay is Alloy 52M, which has been shown to be resistant to IGSCC [4], thus providing an IGSCC resistant barrier. Therefore, little if any IGSCC crack growth is expected to occur in the overlay.
- Application of the weld overlay does not impact the conclusions of the existing system Stress Report. Following application of the overlay, all ASME Code,Section III stress and fatigue criteria are met.
- A weld specific residual stress analysis was performed by first simulating a severe ID weld repair in the DMW, prior to applying the weld overlay. The post weld overlay residual stresses were shown to result in reduced tensile stresses into the thickness of the original DMW.
- A fracture mechanics analysis was performed to determine the amount of future crack growth which would be predicted in the FSWOL, assuming that the initial depth was 100% through wall of the original base material. Both fatigue and IGSCC crack growth were considered, and found to be acceptable.
- A walk down focusing on support H10-347 and review of associated pipe support locations and configurations (EC 619794) indicated that all hangers and other supports were within design dimensional tolerances.
Based on the above observations and the fact that weld overlays have been applied to other plants since 1986 with no subsequent problems identified, it is concluded that the JAF Loop A RWR/RHR low pressure coolant injection tee-to RHR gate valve DMW has received long term resistance to IGSCC.
Attachment to 1700151.402.R0 11 of 12
4.0 REFERENCES
- 1. Relief Request No. JAFP-17-0010 (Relief Request RR 21), Proposed Alternative to ASME Code Requirements for Weld Overlay Repair, RR-21 at James A. FitzPatrick Nuclear Power Plant Docket No. 50-333, dated January 27, 2017. ADAMS Accession No. ML17028A011.
- 2. ASME Boiler and Pressure Vessel Code, Code Case N-504-3/4, Alternate Rules for Repair of Classes 1, 2 and 3 Austenitic Stainless Steel Piping,Section XI, Division 1.
- 3. ASME Boiler and Pressure Vessel Code, Code Case N-638-4, Similar and Dissimilar Metal Welding Using Ambient Temperature Machine GTAW Temper Bead Technique,Section XI, Division 1.
- 4. Peter L. Andresen, et al., GE Global Research Center, SCC of High Cr Alloys in BWR Environments, 15th International Conference on Environmental Degradation, TMS (The Minerals, Metals & Materials Society), 2011.
- 5. ASME Code,Section XI, 2004 Edition with Addenda through 2005, Nonmandatory Appendix Q, Weld Overlay Repairs of Classes 1, 2, and 3 Austenitic Stainless Steel Piping Weldments.
- 6. ASME Boiler and Pressure Vessel Code,Section XI, Rules for Inservice Inspection of Nuclear Power Plant Components, 2001 Edition with Addenda through 2003.
- 7. ASME Boiler and Pressure Vessel Code,Section III, 2001 Edition with Addenda through 2003.
- 8. USAS B31.1.0 Power Piping, 1967 Edition/1969 Addenda.
- 9. ANSYS Mechanical APDL, Release 14.5 (w/ Service Pack 1 UP20120918), ANSYS, Inc.,
September 2012.
- 10. Materials Reliability Program: Finite-Element Model Validation for Dissimilar Metal Butt-Welds (MRP-316, Revision 1): Volumes 1 and 2. EPRI, Palo Alto, CA: 2015. 3002005498.
- 11. Materials Reliability Program: Welding Residual Stress Dissimilar Metal Butt-Weld Finite Element Modeling Handbook (MRP-317, Revision 1). EPRI, Palo Alto, CA: 2015. 3002005499.
- 12. U.S. NRC, International Weld Residual Stress Round Robin Problem Statement, Version 1.0, U.S.
Nuclear Regulatory Commission, Office of Nuclear Regulatory Research, Division of Engineering, Component Integrity Branch, December 2009.
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