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MONTHYEARML0521404632005-07-21021 July 2005 Inservice Inspection Summary Report Project stage: Request ML0606200632006-03-0101 March 2006 Proposed Alternative to the American Society of Mechanical Engineers Code, Section XI Repair Requirements Request for Additional Information Project stage: Request ML0617104642006-06-0909 June 2006 Proposed Alternative to the American Society of Mechanical Engineers Code, Section XI Repair Requirements Project stage: Request ML0623600032006-08-23023 August 2006 Draft RAI on Preemptive Weld Overlay Project stage: Draft RAI ML0626301422006-09-15015 September 2006 Supplement to Proposed Alternative to the American Society of Mechanical Engineers Code, Section XI, Repair Requirements Project stage: Supplement ML0626803132006-09-25025 September 2006 Two Questions Re. Cook Unit 1 Preemptive Weld Overlay Relief Project stage: Other ML0627201222006-09-26026 September 2006 Supplement to Proposed Alternative to the American Society of Mechanical Engineers Code, Section XI Repair Requirements Project stage: Supplement ML0627802032006-09-26026 September 2006 Supplement to Proposed Alternative to ASME Code, Section XI, Repair Requirements Project stage: Supplement ML0631902952006-11-0606 November 2006 American Society of Mechanical Engineers Code, Section XI Repair Requirements Preemptive Weld Overlay - Ultrasonic Examination Results Project stage: Request ML0635303352006-12-0707 December 2006 American Society of Mechanical Engineers Code, Section XI Repair Requirements Preemptive Weld Overlay - Stress Summaries Project stage: Request ML0707200212007-04-26026 April 2007 Alternative Regarding Use of Preemptive Weld Overlays on Certain Dissimilar Metal Welds Project stage: Other 2006-06-09
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Category:Letter
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Cook Nuclear Power Plant, July 2024 AEP-NRC-2024-56, Unit 2, Independent Spent Fuel Storage Installation, Registration of Dry Spent Fuel Storage Cask2024-07-0808 July 2024 Unit 2, Independent Spent Fuel Storage Installation, Registration of Dry Spent Fuel Storage Cask AEP-NRC-2024-48, Response to Request for Additional Information (RAI) for License Amendment Request for One-Time Extension of Completion Time for Inoperable AC Source - Operating2024-07-0202 July 2024 Response to Request for Additional Information (RAI) for License Amendment Request for One-Time Extension of Completion Time for Inoperable AC Source - Operating ML24176A1012024-06-21021 June 2024 57143-EN 57143 - Paragon Energy Solutions - Update 1 (Final) - 10CFR Part 21 Final Notification: P21-05242024-FN, Rev. 0 AEP-NRC-2024-45, Report Per Technical Specification 5.6.6 for Inoperability of Post Accident Monitoring Neutron Flux Monitoring2024-06-13013 June 2024 Report Per Technical Specification 5.6.6 for Inoperability of Post Accident Monitoring Neutron Flux Monitoring ML24163A0132024-06-12012 June 2024 Request for Information for the NRC Age-Related Degradation Inspection: Inspection Report 05000315/2024012 and 05000316/2024012 ML24159A2522024-05-30030 May 2024 10 CFR 50.71(e) Update and Related Site Change Reports AEP-NRC-2024-23, Core Operating Limits Report2024-05-23023 May 2024 Core Operating Limits Report ML24141A2162024-05-20020 May 2024 —Information Request to Support the NRC Annual Baseline Emergency Action Level and Emergency Plan Changes Inspection 05000316/LER-2024-001, Reactor Trip Due to Main Turbine Trip from a High-High Thrust Bearing Position Trip2024-05-20020 May 2024 Reactor Trip Due to Main Turbine Trip from a High-High Thrust Bearing Position Trip AEP-NRC-2024-40, Unit 2 - Supplement to License Amendment Request for One-Time Extension of Completion Time for Inoperable AC Source - Operating2024-05-16016 May 2024 Unit 2 - Supplement to License Amendment Request for One-Time Extension of Completion Time for Inoperable AC Source - Operating AEP-NRC-2024-41, Annual Radiological Environmental Operating Report2024-05-15015 May 2024 Annual Radiological Environmental Operating Report AEP-NRC-2024-26, Transmittal of Donald C. Cook Nuclear Plant, Emergency Plan Revision 492024-05-14014 May 2024 Transmittal of Donald C. Cook Nuclear Plant, Emergency Plan Revision 49 IR 05000315/20244012024-05-14014 May 2024 – Security Baseline Inspection Report 05000315/2024401 and 05000316/2024401 AEP-NRC-2024-07, Unit 2 - Transmittal of Report of Changes to the Emergency Plan2024-05-14014 May 2024 Unit 2 - Transmittal of Report of Changes to the Emergency Plan IR 05000315/20240012024-05-14014 May 2024 Integrated Inspection Report 05000315/2024001 and 05000316/2024001 ML24115A2152024-05-0707 May 2024 LTR: CNP Non-Acceptance with Opportunity TS 3-8-1 AEP-NRC-2024-24, Form OAR-1, Owners Activity Report2024-05-0707 May 2024 Form OAR-1, Owners Activity Report ML24256A1472024-05-0606 May 2024 DC Cook 2024 NRC Examination Submittal Letter: Submittal ML24116A0002024-05-0202 May 2024 – Regulatory Audit in Support of Review of the Application to Adopt 10 CFR 50.69, Risk-Informed Categorization and Treatment of Structures, Systems and Components for Nuclear Power Reactors AEP-NRC-2024-35, Response to NRC Regulatory Issue Summary 2024-01 Preparation and Scheduling of Operator Licensing Examinations2024-04-30030 April 2024 Response to NRC Regulatory Issue Summary 2024-01 Preparation and Scheduling of Operator Licensing Examinations AEP-NRC-2024-28, 2023 Annual Radioactive Effluent Release Report2024-04-29029 April 2024 2023 Annual Radioactive Effluent Release Report AEP-NRC-2024-31, Annual Report of Individual Monitoring2024-04-24024 April 2024 Annual Report of Individual Monitoring AEP-NRC-2024-29, (CNP) Unit 2 - Request for Relief Related to American Society of Mechanical Engineers (ASME) Code Case N-729-6 Supplemental Examination Requirements, ISIR-5-072024-04-0303 April 2024 (CNP) Unit 2 - Request for Relief Related to American Society of Mechanical Engineers (ASME) Code Case N-729-6 Supplemental Examination Requirements, ISIR-5-07 AEP-NRC-2024-02, Unit 2 License Amendment Request for One-Time Extension of Completion Time for Inoperable AC Source - Operating2024-04-0303 April 2024 Unit 2 License Amendment Request for One-Time Extension of Completion Time for Inoperable AC Source - Operating 2024-09-09
[Table view] Category:Report
MONTHYEARML24183A0162024-07-25025 July 2024 Review of Reactor Vessel Material Surveillance Program Capsule W Technical Report AEP-NRC-2024-45, Report Per Technical Specification 5.6.6 for Inoperability of Post Accident Monitoring Neutron Flux Monitoring2024-06-13013 June 2024 Report Per Technical Specification 5.6.6 for Inoperability of Post Accident Monitoring Neutron Flux Monitoring ML24159A2702024-05-30030 May 2024 R1900-0024-001, Rev. 16, NFPA 805 Nuclear Safety Capability Assessment AEP-NRC-2022-66, Report Per Technical Specification 5.6.6 for Inoperability of Unit 2 Post Accident Monitoring Neutron Flux Monitoring2022-12-15015 December 2022 Report Per Technical Specification 5.6.6 for Inoperability of Unit 2 Post Accident Monitoring Neutron Flux Monitoring AEP-NRC-2022-46, Notification of Deviation from Electric Power Research Institute (EPRI) Materials Reliability Program 2019-008, Interim Guidance for NEI 03-08 Needed Requirements for Us PWR Plants for Management of Thermal Fatigue in2022-12-12012 December 2022 Notification of Deviation from Electric Power Research Institute (EPRI) Materials Reliability Program 2019-008, Interim Guidance for NEI 03-08 Needed Requirements for Us PWR Plants for Management of Thermal Fatigue in ML22340A1992022-11-30030 November 2022 Report of Changes, Tests, Experiments Pursuant to 10 CFR 50.59(d)(2) ML22340A2132022-11-30030 November 2022 R1900-0024-001, Revision 13, NFPA 805 Nuclear Safety Capability Assessment ML22340A1762022-11-30030 November 2022 Commitment Change Summary October 2020 to May 2022 AEP-NRC-2022-58, U1C31 Steam Generator Tube Inspection Report2022-10-24024 October 2022 U1C31 Steam Generator Tube Inspection Report AEP-NRC-2021-44, Form OAR-1, Owner'S Activity Report2021-08-12012 August 2021 Form OAR-1, Owner'S Activity Report ML21125A5582021-04-19019 April 2021 Report 2019, Commitment Change Summary, May 2019 to October 2020 ML21125A5392021-04-19019 April 2021 Report of Changes, Tests, Experiments Pursuant to 10 CFR 50.59(d)(2), Boration System Functionality Requirement Change in Mode 4 EA-21-024, Notice of Enforcement Discretion2021-03-0404 March 2021 Notice of Enforcement Discretion for Donald C. Cook Nuclear Plant, Units 1 and 2 AEP-NRC-2021-07, Supplement to Report Per Technical Specification 5.6.6, Lnoperability of Unit 1, Post Accident Monitoring, Containment Water Level2021-01-28028 January 2021 Supplement to Report Per Technical Specification 5.6.6, Lnoperability of Unit 1, Post Accident Monitoring, Containment Water Level AEP-NRC-2020-28, CFR 72.48(d)(2) Summary Report of Completed Changes, Tests, and Experiments 1O CFR 72.48 Evaluations2020-05-0606 May 2020 CFR 72.48(d)(2) Summary Report of Completed Changes, Tests, and Experiments 1O CFR 72.48 Evaluations AEP-NRC-2020-23, Request for Use of Encoded Phased Array Ultrasonic Examination Techniques in Lieu of Radiography2020-04-30030 April 2020 Request for Use of Encoded Phased Array Ultrasonic Examination Techniques in Lieu of Radiography ML20108E9992020-03-0505 March 2020 Enclosure 7 - LTR-SCS-19-50, Revision 0, D.C. Cook Unit 1 Low Temperature Overpressure Protection System (Ltops) Analysis for 48 EFPY, Dated March 5, 2020, Attachment 2 Only (Non-Proprietary) ML20108F0002020-02-28028 February 2020 Enclosure 5 - WCAP-18455-NP, Revision 1, D.C. Cook Unit 1 Heatup and Cooldown Limit Curves for Normal Operation, Westinghouse Electric Company, February 2020. (Non-Proprietary) ML18274A0952018-09-30030 September 2018 WCAP-18394-NP, Revision 1, Fatigue Crack Growth Evaluations of D. C. Cook Units 1 and 2 RHR, Accumulator, and Safety Injection Lines Supporting Expanded Scope Leak-Before-Break, September 2018 (Non-Proprietary) ML18334A2712018-09-30030 September 2018 WCAP-18394-NP, Revision 1, Fatigue Crack Growth Evaluations of D.C. Cook, Units 1 and 2 RHR, Accumulator, and Safety Injection Lines Supporting Expanded Scope Leak-Before-Break. AEP-NRC-2018-36, Notification of Initial Renewable Operating Permit2018-05-0909 May 2018 Notification of Initial Renewable Operating Permit AEP-NRC-2018-21, 30-Day Report of Changes to or Errors in an Evaluation Model2018-05-0404 May 2018 30-Day Report of Changes to or Errors in an Evaluation Model ML18026A8822018-02-0505 February 2018 Staff Assessment of Flooding Focused Evaluation ML18334A2702018-01-31031 January 2018 WCAP-18309-NP, Revision 0, Technical Justification for Eliminating Safety Injection Line Rupture as the Structural Design Basis for D.C. Cook, Units 1 and 2, Using Leak-Before-Break Methodology. ML18334A2692018-01-31031 January 2018 WCAP-18302-NP, Revision 0, Technical Justification for Eliminating Residual Heat Removal Line Rupture as the Structural Design Basis for D.C. Cook Units 1 and 2, Using Leak-Before-Break Methodology. ML18334A2682018-01-31031 January 2018 WCAP-18295-NP, Revision 0, Technical Justification for Eliminating Accumulator Line Rupture as the Structural Design Basis for D.C. Cook Units 1 and 2, Using Leak-Before-Break Methodology (Non-Proprietary) ML17151A9672017-06-14014 June 2017 Flood Hazard Mitigation Strategies Assessment ML16313A1172016-10-10010 October 2016 1BTl1V001-RPT-01, Donald C. Cook Focused Scope Peer Review - Pre-Initiator Human Reliability Analysis. ML16127A3352016-05-0606 May 2016 Reactor Oversight Process Task Force FAQ Log-April 13, 2016 ML16113A1982016-04-20020 April 2016 Precursor Screening Analysis- Reject ML16169A1182016-03-31031 March 2016 RWA-1313-015, Rev. 1, AST Radiological Analysis Technical Report. ML15308A0932015-10-15015 October 2015 Pressurized Water Reactor Owners Group (Pwrog), 15066-NP, Revision 1, Responses to Follow-Up NRC RAI 2 on the D.C. Cook, Units 1 and 2, Reactor Internals Aging Management Program. AEP-NRC-2015-83, Revision 1 of Final Integrated Plan Regarding March 12, 2012, NRC Order Regarding Mitigation Strategies for Beyond-Design-Basis External Events (Order Number EA-12-049)2015-10-0101 October 2015 Revision 1 of Final Integrated Plan Regarding March 12, 2012, NRC Order Regarding Mitigation Strategies for Beyond-Design-Basis External Events (Order Number EA-12-049) ML15233A0242015-08-19019 August 2015 Transmittal of Annual Report of Loss-of-Coolant Accident Evaluation Model Changes ML14253A3172014-09-0404 September 2014 Enclosure 2: I&M CAP Document AR 2010-1804-10, Root Cause Evaluation Attachment, Rx Vessel Core Support Lug Bolting Anomalies ML14147A3292014-06-18018 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 ML14181A5382014-06-0505 June 2014 Enclosure 5 to AEP-NRC-2014-42, Attachment #2 (NP-Attachment) of Westinghouse Letter, LTR-PL-14-22, Westinghouse Responses to NRC, Request for Additional Information on the Application for Amendment to Restore.. ML14073A7592014-03-31031 March 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 ML14092A3302014-03-17017 March 2014 Document No. 13Q3208-RPT-003, Revion 1, Seismic Hazard and Screening Report for the Cook Nuclear Plant (Cnp), Enclosure 2 to AEP-NRC-2014-25 AEP-NRC-2014-15, 30 Day Report of Changes to or Errors in an Evaluation Model2014-02-27027 February 2014 30 Day Report of Changes to or Errors in an Evaluation Model ML13337A3252014-01-24024 January 2014 Interim Staff Evaluation and Audit Report Relating to Overall Integrated Plan in Response to Order EA-12-049 (Mitigation Strategies) AEP-NRC-2014-08, SD-121023-001, Rev. 2, in Response to the 50.54(f) Information Request Regarding Fukushima Near-Term Task Force Recommendation 2.3: Seismic for the D. C. Cook Unit 1 and 2, Page D-322 Through Page D-4042014-01-13013 January 2014 SD-121023-001, Rev. 2, in Response to the 50.54(f) Information Request Regarding Fukushima Near-Term Task Force Recommendation 2.3: Seismic for the D. C. Cook Unit 1 and 2, Page D-322 Through Page D-404 ML14035A3632014-01-13013 January 2014 SD-121023-001, Rev. 2, in Response to the 50.54(f) Information Request Regarding Fukushima Near-Term Task Force Recommendation 2.3: Seismic for the D. C. Cook Unit 1 and 2, Page C-845 Through Page C-962 ML14035A3682014-01-13013 January 2014 SD-121023-001, Rev. 2, in Response to the 50.54(f) Information Request Regarding Fukushima Near-Term Task Force Recommendation 2.3: Seismic for the D. C. Cook Unit 1 and 2, Page D-241 Through Page D-321 ML14035A3672014-01-13013 January 2014 SD-121023-001, Rev. 2, in Response to the 50.54(f) Information Request Regarding Fukushima Near-Term Task Force Recommendation 2.3: Seismic for the D. C. Cook Unit 1 and 2, Page D-170 Through Page D-240 ML14035A3662014-01-13013 January 2014 SD-121023-001, Rev. 2, in Response to the 50.54(f) Information Request Regarding Fukushima Near-Term Task Force Recommendation 2.3: Seismic for the D. C. Cook Unit 1 and 2, Page D-92 Through Page D-169 ML14035A3642014-01-13013 January 2014 SD-121023-001, Rev. 2, in Response to the 50.54(f) Information Request Regarding Fukushima Near-Term Task Force Recommendation 2.3: Seismic for the D. C. Cook Unit 1 and 2, Page D-1 Through Page D-91 ML14035A3522014-01-13013 January 2014 SD-121023-001, Rev. 2, in Response to the 50.54(f) Information Request Regarding Fukushima Near-Term Task Force Recommendation 2.3: Seismic for the D. C. Cook Unit 1 and 2, Cover Through Page B-312 ML14035A3532014-01-13013 January 2014 SD-121023-001, Rev. 2, in Response to the 50.54(f) Information Request Regarding Fukushima Near-Term Task Force Recommendation 2.3: Seismic for the D. C. Cook Unit 1 and 2, Page C-1 Through Page C-114 ML14035A3552014-01-13013 January 2014 SD-121023-001, Rev. 2, in Response to the 50.54(f) Information Request Regarding Fukushima Near-Term Task Force Recommendation 2.3: Seismic for the D. C. Cook Unit 1 and 2, Page D-405 Through End 2024-07-25
[Table view] Category:Miscellaneous
MONTHYEARML24183A0162024-07-25025 July 2024 Review of Reactor Vessel Material Surveillance Program Capsule W Technical Report AEP-NRC-2024-45, Report Per Technical Specification 5.6.6 for Inoperability of Post Accident Monitoring Neutron Flux Monitoring2024-06-13013 June 2024 Report Per Technical Specification 5.6.6 for Inoperability of Post Accident Monitoring Neutron Flux Monitoring AEP-NRC-2022-46, Notification of Deviation from Electric Power Research Institute (EPRI) Materials Reliability Program 2019-008, Interim Guidance for NEI 03-08 Needed Requirements for Us PWR Plants for Management of Thermal Fatigue in2022-12-12012 December 2022 Notification of Deviation from Electric Power Research Institute (EPRI) Materials Reliability Program 2019-008, Interim Guidance for NEI 03-08 Needed Requirements for Us PWR Plants for Management of Thermal Fatigue in AEP-NRC-2021-44, Form OAR-1, Owner'S Activity Report2021-08-12012 August 2021 Form OAR-1, Owner'S Activity Report ML21125A5582021-04-19019 April 2021 Report 2019, Commitment Change Summary, May 2019 to October 2020 ML21125A5392021-04-19019 April 2021 Report of Changes, Tests, Experiments Pursuant to 10 CFR 50.59(d)(2), Boration System Functionality Requirement Change in Mode 4 EA-21-024, Notice of Enforcement Discretion2021-03-0404 March 2021 Notice of Enforcement Discretion for Donald C. Cook Nuclear Plant, Units 1 and 2 AEP-NRC-2020-28, CFR 72.48(d)(2) Summary Report of Completed Changes, Tests, and Experiments 1O CFR 72.48 Evaluations2020-05-0606 May 2020 CFR 72.48(d)(2) Summary Report of Completed Changes, Tests, and Experiments 1O CFR 72.48 Evaluations AEP-NRC-2018-36, Notification of Initial Renewable Operating Permit2018-05-0909 May 2018 Notification of Initial Renewable Operating Permit AEP-NRC-2018-21, 30-Day Report of Changes to or Errors in an Evaluation Model2018-05-0404 May 2018 30-Day Report of Changes to or Errors in an Evaluation Model ML18026A8822018-02-0505 February 2018 Staff Assessment of Flooding Focused Evaluation ML17151A9672017-06-14014 June 2017 Flood Hazard Mitigation Strategies Assessment ML16127A3352016-05-0606 May 2016 Reactor Oversight Process Task Force FAQ Log-April 13, 2016 ML16113A1982016-04-20020 April 2016 Precursor Screening Analysis- Reject AEP-NRC-2015-83, Revision 1 of Final Integrated Plan Regarding March 12, 2012, NRC Order Regarding Mitigation Strategies for Beyond-Design-Basis External Events (Order Number EA-12-049)2015-10-0101 October 2015 Revision 1 of Final Integrated Plan Regarding March 12, 2012, NRC Order Regarding Mitigation Strategies for Beyond-Design-Basis External Events (Order Number EA-12-049) ML15233A0242015-08-19019 August 2015 Transmittal of Annual Report of Loss-of-Coolant Accident Evaluation Model Changes ML14181A5382014-06-0505 June 2014 Enclosure 5 to AEP-NRC-2014-42, Attachment #2 (NP-Attachment) of Westinghouse Letter, LTR-PL-14-22, Westinghouse Responses to NRC, Request for Additional Information on the Application for Amendment to Restore.. ML14073A7592014-03-31031 March 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 AEP-NRC-2014-15, 30 Day Report of Changes to or Errors in an Evaluation Model2014-02-27027 February 2014 30 Day Report of Changes to or Errors in an Evaluation Model ML12324A4182012-12-20020 December 2012 Review of the 2011 Refueling Outage Steam Generator Tube Inservice Inspection Results AEP-NRC-2012-86, Flooding Walkdown Report in Response to the 50.54(F) Information Request Regarding Near-Term Task Force Recommendation 2.3: Flooding for the D.C. Cook Nuclear Power Plant2012-11-13013 November 2012 Flooding Walkdown Report in Response to the 50.54(F) Information Request Regarding Near-Term Task Force Recommendation 2.3: Flooding for the D.C. Cook Nuclear Power Plant ML12362A0762012-11-0505 November 2012 Donald C. Cook, Units 1 & 2, Seismic Walkdown Report SD-121023-001, Rev. 0 in Response to the 50.54(f) Information Request Regarding Fukushima Near-Term Task Force Recommendation 2.3: Seismic, Appendix C, Page C-268 Through C-353 AEP-NRC-2012-87, Seismic Walkdown Report SD-121023-001, Rev. 0 in Response to the 50.54(f) Information Request Regarding Fukushima Near-Term Task Force Recommendation 2.3: Seismic, Appendix C, Page C-437 Through C-4862012-11-0505 November 2012 Seismic Walkdown Report SD-121023-001, Rev. 0 in Response to the 50.54(f) Information Request Regarding Fukushima Near-Term Task Force Recommendation 2.3: Seismic, Appendix C, Page C-437 Through C-486 ML12362A0772012-11-0505 November 2012 Donald C. Cook, Units 1 & 2, Seismic Walkdown Report SD-121023-001, Rev. 0 in Response to the 50.54(f) Information Request Regarding Fukushima Near-Term Task Force Recommendation 2.3: Seismic, Appendix C, Page C-354 Through C-436 ML12362A0752012-11-0505 November 2012 Donald C. Cook, Units 1 & 2, Seismic Walkdown Report SD-121023-001, Rev. 0 in Response to the 50.54(f) Information Request Regarding Fukushima Near-Term Task Force Recommendation 2.3: Seismic, Appendix C, Page C-200 Through C-267 ML12362A0582012-11-0505 November 2012 Donald C. Cook, Units 1 & 2, Seismic Walkdown Report SD-121023-001, Rev. 0 in Response to the 50.54(f) Information Request Regarding Fukushima Near-Term Task Force Recommendation 2.3: Seismic, Appendix C, Page C-487 Through C-548 ML12362A0592012-11-0505 November 2012 Donald C. Cook, Units 1 & 2, Seismic Walkdown Report SD-121023-001, Rev. 0 in Response to the 50.54(f) Information Request Regarding Fukushima Near-Term Task Force Recommendation 2.3: Seismic, Appendix C, Page C-549 Through C-620 ML12362A0612012-11-0505 November 2012 Donald C. Cook, Units 1 & 2, Seismic Walkdown Report SD-121023-001, Rev. 0 in Response to the 50.54(f) Information Request Regarding Fukushima Near-Term Task Force Recommendation 2.3: Seismic, Appendix C, Page C-698 Through C-776 ML12362A0742012-11-0505 November 2012 Donald C. Cook, Units 1 & 2, Seismic Walkdown Report SD-121023-001, Rev. 0 in Response to the 50.54(f) Information Request Regarding Fukushima Near-Term Task Force Recommendation 2.3: Seismic, Appendix C, Page C-157 Through C-199 ML12362A0732012-11-0505 November 2012 Donald C. Cook, Units 1 & 2, Seismic Walkdown Report SD-121023-001, Rev. 0 in Response to the 50.54(f) Information Request Regarding Fukushima Near-Term Task Force Recommendation 2.3: Seismic, Appendix C, Page C-113 Through C-156 ML12362A0722012-11-0505 November 2012 Donald C. Cook, Units 1 & 2, Seismic Walkdown Report SD-121023-001, Rev. 0 in Response to the 50.54(f) Information Request Regarding Fukushima Near-Term Task Force Recommendation 2.3: Seismic, Appendix C, Page C-47 Through C-112 ML12362A0712012-11-0505 November 2012 Donald C. Cook, Units 1 & 2, Seismic Walkdown Report SD-121023-001, Rev. 0 in Response to the 50.54(f) Information Request Regarding Fukushima Near-Term Task Force Recommendation 2.3: Seismic, Cover Through Appendix C, Page C-46 ML12362A0692012-11-0505 November 2012 Donald C. Cook, Units 1 & 2, Seismic Walkdown Report SD-121023-001, Rev. 0 in Response to the 50.54(f) Information Request Regarding Fukushima Near-Term Task Force Recommendation 2.3: Seismic, Appendix D, Page D-388 Through End ML12362A0682012-11-0505 November 2012 Donald C. Cook, Units 1 & 2, Seismic Walkdown Report SD-121023-001, Rev. 0 in Response to the 50.54(f) Information Request Regarding Fukushima Near-Term Task Force Recommendation 2.3: Seismic, Appendix D, Page D-337 Through Page D-387 ML12362A0672012-11-0505 November 2012 Donald C. Cook, Units 1 & 2, Seismic Walkdown Report SD-121023-001, Rev. 0 in Response to the 50.54(f) Information Request Regarding Fukushima Near-Term Task Force Recommendation 2.3: Seismic, Appendix D, Page D-276 Through Page D-336 ML12362A0662012-11-0505 November 2012 Donald C. Cook, Units 1 & 2, Seismic Walkdown Report SD-121023-001, Rev. 0 in Response to the 50.54(f) Information Request Regarding Fukushima Near-Term Task Force Recommendation 2.3: Seismic, Appendix D, Page D-220 Through Page D-275 ML12362A0602012-11-0505 November 2012 Donald C. Cook, Units 1 & 2, Seismic Walkdown Report SD-121023-001, Rev. 0 in Response to the 50.54(f) Information Request Regarding Fukushima Near-Term Task Force Recommendation 2.3: Seismic, Appendix C, Page C-621 Through C-697 ML12362A0622012-11-0505 November 2012 Donald C. Cook, Units 1 & 2, Seismic Walkdown Report SD-121023-001, Rev. 0 in Response to the 50.54(f) Information Request Regarding Fukushima Near-Term Task Force Recommendation 2.3: Seismic, Appendix C, Page C-777 Through Appendix D, Page ML12362A0632012-11-0505 November 2012 Donald C. Cook, Units 1 & 2, Seismic Walkdown Report SD-121023-001, Rev. 0 in Response to the 50.54(f) Information Request Regarding Fukushima Near-Term Task Force Recommendation 2.3: Seismic, Appendix D, Page D-54 Through Page D-114 ML12362A0642012-11-0505 November 2012 Donald C. Cook, Units 1 & 2, Seismic Walkdown Report SD-121023-001, Rev. 0 in Response to the 50.54(f) Information Request Regarding Fukushima Near-Term Task Force Recommendation 2.3: Seismic, Appendix D, Page D-115 Through Page D-175 ML12362A0652012-11-0505 November 2012 Donald C. Cook, Units 1 & 2, Seismic Walkdown Report SD-121023-001, Rev. 0 in Response to the 50.54(f) Information Request Regarding Fukushima Near-Term Task Force Recommendation 2.3: Seismic, Appendix D, Page D-176 Through Page D-219 AEP-NRC-2012-83, Communications Assessment Requested by Nuclear Regulatory Commission Letter, Request for Information Pursuant to Title 10 of the Code of Federal Regulation 50.54(f) Regarding Recommendation 2.1, 2.3, and 9.3.2012-10-31031 October 2012 Communications Assessment Requested by Nuclear Regulatory Commission Letter, Request for Information Pursuant to Title 10 of the Code of Federal Regulation 50.54(f) Regarding Recommendation 2.1, 2.3, and 9.3. AEP-NRC-2012-38, Response to Request for Information, 10 CFR 50.46 Report for Emergency Core Cooling System Model Change or Error Associated with Thermal Conductivity Degradation2012-06-11011 June 2012 Response to Request for Information, 10 CFR 50.46 Report for Emergency Core Cooling System Model Change or Error Associated with Thermal Conductivity Degradation AEP-NRC-2009-25, Small Break Loss-of-Coolant Accident Evaluation Model Reanalysis2009-03-30030 March 2009 Small Break Loss-of-Coolant Accident Evaluation Model Reanalysis ML0807400532008-02-29029 February 2008 Response to Request for Additional Information Regarding Reanalysis of Small Break Loss-Of-Coolant Accident ML0807703952008-02-29029 February 2008 AEP:NRC:8054-02, Attachment 1, References, Through Attachment 3, Supplemental Response to GL 2004-02 and Request for Additional Information. ML0807200622008-02-29029 February 2008 License Amendment Request to Revise Ice Condenser Licensing Basis ML0807703962008-02-29029 February 2008 AEP:NRC:8054-02, Attachment 3, I&M Response to Information Item 3.f.4, to NRC Information Item 3 - Conclusions. ML0807704002008-02-29029 February 2008 AEP:NRC:8054-02, Attachment 4, Figure A4-1, General Arrangement of Recirculation Sump to Attachment 5, Figure A5-40, Scotch 77 Fire Retardant Tape Test 1 Pre-Test Picture. ML0807704042008-02-29029 February 2008 AEP:NRC:8054-02, Attachment 5, Figure A5-41, Scotch 77 Fire Retardant Tape Test 1 Post-Test Picture to Attachment 7, Regulatory Commitments. 2024-07-25
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Indiana Michigan Power INDIANA Cook Nuclear Plant MICHIGAN One Cook Place Bridgman, MI 49106 POWERO AEPcom A unit of American Electric Power December 7, 2006 AEP:NRC:6055-19 10 CFR 50.55a Docket No.: 50-315 U. S. Nuclear Regulatory Commission ATTN: Documnent Control Desk Mail Stop O-P1-17
.Washington, DC 20555-0001 Donald C. Cook Nuclear Plant Unit 1 AMERICAN SOCIETY OF MECHANICAL ENGINEERS CODE,.
SECTION XI REPAIR REQUIREMENTS PREEMPTIVE WELD OVERLAY - STRESS SUMMARIES
Reference:
Letter firom Mark A. Peifer, Indiana Michigan Power Company, to Nuclear Regulatory Conmmission Document Control Desk, "Donald C. Cook Nuclear Plant Unit 1, Supplement to Proposed Alternative to the American Society Of Mechanical Engineers Code,Section XI Repair Requirements," AEP:NRC:6055-17, Accession Number ML062780203, dated September 26, 2006.
In the referenced letter, Indiana Michigan Power Company (I&M), the licensee for Donald C. Cook Nuclear Plant Unit 1, proposed an alternative to the repair requirements of the American Society of Mechanical Engineers Code,Section XI. .Approval of the proposed alternative was requested to allow I&M to apply full structural preemptive weld overlays (PWOLs) on pressurizer nozzle safe end to nozzle welds where NiCrFe Alloy 82/182 was originally used to weld the safe ends thereto.
In requesting approval of the proposed alternative, I&M committed to providing the Nuclear Regulatory Commission with the stress summaries for the PWOLs. The attachment to this letter provides the PWOL stress summaries and the associated flaw growth evaluation, which has been conservatively calculated by assuming that a 360 degree circumferential flaw would propagate by primary water stress corrosion cracking through the thickness of the Alloy 82/182 weld, to the interface with the Alloy 52/52M overlay material.
This letter contains no new commitments. Should you have any questions, please contact Ms. Susan D. Simpson, Regulatory Affairs Manager, at (269) 466-2428.
Vice President Site Support Services RGV/rdw Aoq7
U. S. Nuclear Regulatory Commission AEP:NRC:6055-19 Page 2
Attachment:
Donald C. Cook Unit 1, Preemptive Weld Overlay Structural Evaluation Summnary c: R. Aben - Department of Labor and Economic Growth J. L. Caldwell - NRC Region III K. D. Curry - AEP Ft. Wayne, w/o attachment J. T. King - MPSC, w/o attachment MDEQ - WHMD/RPMWS, w/o attachment NRC Resident Inspector P. S. Tarn - NRC Washington DC
ATTACHMENT TO AEPiNRC:6055-19 Donald C. Cook Unit 1 PREEMPTIVE WELD OVERLAY STRUCTURAL EVALUATION
SUMMARY
Abbreviations and Symbols Used in this Attachment ASME American Society of Mechanical Engineers B&PV Boiler and Pressure Vessel Code CNP Donald C. Cook Nuclear Plant DM Dissimilar Metal FCG Flaw Crack Growth HUCD Heatup Cooldown in. inches ksi thousand pounds per square inch OBE Operating Basis Earthquake psi pounds per square inch PWSCC Primary Water Stress Corrosion Cracking SI Stress Intensity SS Stainless Steel
'Equal to
< Less than
> Greater than
Attachment to AEP:NRC:6055-19 Page 2 PRESSURIZER PIPING WELD OVERLAY STRESS ANALYSIS
SUMMARY
ASME SECTION III CRITERIA 1.0 Introduction Due to the susceptibility of Alloy 600 and its associated weldments, Alloy 82/182, to PWSCC, Indiana Michigan Power Company applied full structural weld overlays to the safety, relief, spray, and surge nozzles of the CNP Unit 1 pressurizer. A repair procedure was developed where the DM Alloy 82/182 weld and butter, the SS safe end and weld, and a portion of both the nozzle and attached pipe were overlaid with PWSCC resistant Alloy 52/52M material.
ASME B&PV Code,Section III stress analyses were performed for the CNP Unit 1 pressurizer nozzles repaired with weld overlays in compliance with ASME Code Case N-504-2, Paragraph g (1). Three dimensional ANSYS computer code finite element models for the three nozzles with weld overlays were developed, and detailed finite element analyses (thermal and structural) were performed. The purpose of these calculations was to qualify the weld overlay design to the requirements of the 1998 ASME B&PV Code, including Addenda through 2000, Section 11 criteria. The weld overlay size (thickness and length) was calculated per ASME B&PV Code, Section Xl, Division 1 and ASME Code Case N-504-2.
Thermal stresses were deternmined for the appropriate design transients and a fatigue analysis was performed. The design conditions as well as the thermal transients were evaluated with the finite element models. The results of the thermal analysis were reviewed by examining the magnitude of the temperature difference between critical locations in the models at times when the maximum thermal stresses would develop. The stresses due to the nozzle external loads were calculated and added to the stresses resulting from internal pressure and thermal gradients. The applicable criteria of the 1998 ASME B&PV Code,Section III requirements were met.
2.0 Results 2.1 Primary Stress Intensity Criteria for Design Conditions and All Service Level Loadings The weld overlay applied on the outside surface relieves the nozzle primary stress burden resulting from the applied internal pressure and external loads. Therefore, ASME B&PV Code,Section III primary stress requirements for design conditions and all service level loadings as specified in Paragraphs NB-3221, NB-3222, NB-3223, NB-3224, and NB-3225 have been satisfied for the nozzles, welds with overlays, safe ends, and piping elbows that were evaluated.
Therefore, the primary stress intensity criteria for design conditions and all service level loadings are bounded by the original design.
Attachment to AEP:NRC:6055-19 Page 3 2.2 Minimum Required Pressure Thickness and Reinforcement Area Criteria Adding the weld overlay will increase the nozzle wall thickness. As a result the ASME B&PV Code Section III requirements contained in Paragraphs NB-3324 and NB-3330 are satisfied.
2.3 Primary Plus Secondary Stress Intensity The final SI range is obtained by adding the. maximum membrane plus bending SI range during transients to that due to the applied external loads (thermal plus OBE). Per ASME B&PV Code,
- Section III, Subparagraph NB-3222.2, the SI range limit is 3Sin, Although the final SI range at most locations evaluated is below the 3S.. limit, there are several locations where the limit is exceeded. When~the 3 S..limit is exceeded, the shear bending SI range was subtracted flom the total menmbrane plus bending SI range. The highest SI range in each nozzle is listed as follows:
Safety/Relief Nozzle 69.66 ksi <.3S,,= 69.90 ksi Spray Nozzle 97.57 ksi > 'S,..= 49.38 ksi Surge Nozzle 121.30 ksi > 3S= 56.10 ksi As can be seen, the spray nozzle and the surge nozzle have locations that exceed the 3S,, limiit.
Per the ASME B&PV Code, Section II1, the 3S,, limit on the prilnary plus secondary SI range may be exceeded provided that the requirements of Subparagraph NB-3228.5 (a) through (f) are met: The evaluations of the spray and surge nozzles are provided in the following paragraphs..
2.3.1 Spray Nozzle The spray nozzle meets all criteria at all locations where the 3Sin limit is exceeded.
2.3.2 Surge Nozzle The surge nozzle did not meet the requirement (a) criterion that the primary plus secondary membrane plus bending SI range, excluding thermal bending stresses, shall be less than 3 SM.
The 3S.. limit is still exceeded at the thermal sleeve, 87.6 ksi > 3S.. = 56.1 ksi. Therefore, the ASME B&PV Code,Section III requirement is not met at this location and a detailed evaluation based on the elastic-plastic approach was performed for the HUCD transients with an insurge-outsurge fluid temperature difference of 320 degrees Fahrenheit.
Elastic - Plastic Analysis of the Surge Nozzle Weld Overlay for HUCD Transients: The elastic-plastic analysis was performed in accordance with ASME B&PV Code, Section lII, Subparagraph NB-3228.4, "Shakedown Analysis." The Subparagraph NB-3228.4 criteria are met.
Attachment to AEP:NRC:6055-19 Page 4 2.4 Fatigue Analysis The fatigue usage factor of the three nozzles is conservatively calculated for 60 years of operation (40 design life plus 20 years life extension). Below is a summary:
Safety/Relief Nozzle: the highest cumulative fatigue usage factor = 0.025 < 1.0 (ASME Criteria)
Spray Nozzle: the highest cumulative fatigue:usage factor = 0.890 < 1.0 (ASME Criteria)
Surge:Nozzle: the highest cumulative fatigue usage. factor = 0.214 < 1.0 (ASME Criteria) 3.0 Conclusion Based on the above results, the requirements of Paragraph (g)(1) of ASME Code Case N-504-2 are met, and the repair has been shown to be acceptable for the remaining service life of CNP Unit 1.
Attachment to AEP:NRC:6055-19 Page 5 PRESSURIZER PIPING WELD OVERLAY FATIGUE CRACK GROWTH ANALYSIS
SUMMARY
1.0 Introduction The overlays applied to the pressurizer piping were analyzed for potential growth of a worst case flaw in the nozzle/pipe welds. It was postulated that a 360 degree circumferential flaw would propagate by PWSCC through the thickness of the Alloy 82/182 weld, to the interface with the Alloy 52/52M overlay material. Although PWSCC would not continue to Occur in the Alloy 52/52M overlay, it was postulated that a small fatigue initiated flaw forms in the Alloy 52/52M overlay and combines with the PWSCC crack in the Alloy 82/182 weld to form a large part-through-wall full circumferential flaw that would propagate into the Alloy 52/52M overlay by fatigue crack growth under cyclic loading conditions.
Fracture mechanics analyses were performed to evaluate this worst case flaw in the repair configuration in compliance with ASME Code Case N-504-2, Paragraph (g)(2).:- These evaluations considered the residual welding, steady state, and normal/upset condition transient stresses with the associated number of transient cycles to predict the final flaw size at the end of license extension at CNP Unit 1, which equates to a 29 year service life. These evaluations demonstrated that the postulated circumferential flaw met the 1989 ASME B&PV Code,Section XI, Appendix C acceptance criteria. An additional check was made on the applied membrane stresses in the remaining ligament under normal operating conditions. These analyses were performed for both the Alloy 82/182 weld as well as the SS weld joining the safe end to the piping.
2.0 Results 2.1 Safety/Relief Nozzles 2.1.1 Flaw Growth Results DM Weld SS Weld Minimum Weld Overlay Thickness, in. 0.5370 0.4850 Additional Weld Overlay Thickness for FCG, in. 0.0300 0.0000 Initial Flaw Size, in. 1.4800 0.7150 Final Flaw Size after 29 Years, in. 1.4858 0.7150 Flaw Growth, in. 0.0058 0.0000 Final Crack Depth to Thickness Ratio 0.7497 0.5958
Attachment to AEP:NRC:6055-19 Page 6 2.1.2 Limit Load Analysis Results At the final crack depth, the plastic collapse stress calculated in accordance with ASMIE B&PV Code,Section XI, Appendix C is compared to the failure bending stress in the pipe, accounting for safety factors for normal/upset and emergency/faulted conditions. At both overlaid locations (the DM and SS welds), the requirement that the plastic collapse stress exceed the failure bending stress is met.
NormallUpset Emergency/Faulted Plastic collapse stress at DM weld, psi 30,473 30,349 Failure bending stress at DM weld, psi 9,347 8,980 Plastic collapse stress at SS weld, psi 45,788 45,602 Failure bending stress at SS weld, psi 15,761 15,828 2.1.3 Applied Membrane Stress Considerations The applied niemnbrane stress in the remaining ligament is less than the operating temperature yield stress:.
- Yield stress at DM weld, psi 27,500 Memibt~ane stress at DM weld, psi 10,588 Yield stress at SS weld, psi 27,500 Menibrane stress at SS weld, psi 9,057 2.2 Spray Nozzle 2.2.1 Flaw Growth Results DM Weld SS Weld Minimum Weld Overlay Thickness, in. 0.397000 0.35140 Additional Weld Overlay Thickness for FCG, in. 0.023000 0.00000 Initial Flaw Size, in. 1.060000 0.44000 Final Flaw Size after 29 Years, in. 1.060004 0.44003 Flaw Growth, in. 0.000004 0.00003 Final Crack Depth to Thickness Ratio 0.7491 0.5560 2.2.2 Limit Load Analysis Results At the final crack depth, the plastic collapse stress calculated according to ASME B&PV Code,Section XI, Appendix C is compared to the failure bending stress in the pipe, accounting for safety factors for normal/upset and emergency/faulted conditions. At both overlaid locations
Attachment to AEP:NRC:6055-19 Page 7 (the DM and SS welds), the requirement that the plastic collapse stress exceed the failure bending stress is met.
Normal/Upset Emergency/Faulted Plastic collapse stress at DM weld, psi 30,300 30,091 Failure bending stress at DM weld, psi 12,959 10,064 Plastic collapse stress at SS weld, psi 49,474 49,123 Failure bending stress at SS weld, psi 23,369 18,557 2.2.3 Applied Membrane Stress Consideration The applied membrane stress in the remaining ligament is less than the operating temnperature yield stress.
Yield stress at DM weld, psi 27,500 Membrane stress at DM weld, psi 11,169 Yield stress at SS weld, psi 27,500 Membrane stress at SS weld, psi 8,732 2.3 Surge Nozzle 2.3.1 Flaw Growth Results DM Weld SS Weld Minimum Weld Overlay Thickness, in. 0.5200 0.7600 Additional Weld Overlay Thickness for FCG, in. 0.0120 0.0000 Initial Flaw Size, in. 1.5600 1.3900 Final Flaw Size after 29 Years, in. 1.5678 1.3900 Flaw Growth, in. 0.0078 0.0000 Final Crack Depth to Thickness Ratio 0.7494 0.6465 2.3.2 Limit Load Analysis Results At the final crack depth, the plastic collapse stress calculated according to ASME B&PV Code,Section XI, Appendix C is compared to the failure bending stress in the pipe, accounting for safety factors for normal/upset and emergency/faulted conditions. At both overlaid locations
Attachment to AEP:NRC:6055-19 Page 8 (the DM and SS welds), the requirement that the plastic collapse stress exceed the failure bending stress is met.
Normal/Upset Emergency/Faulted Plastic collapse stress at DM Weld, psi 27,490 27,130 Failure bending stress at DM weld, psi 11,165 20,808 Plastic collapse stress at SS weld, psi 39,800 39,493 Failure bending stress at SS weld, psi 10,724 20,157 2.3.3 Applied Membrane Stress Consideration The applied membrane stress in the remaining ligament is less than the operating temperature yield stress.
Yield stress at DM weld, psi 27,500 Membrane stress at DM weld, psi 17,844 Yield stress at SS weld, psi 27,500 Membrane stress at SS weld, psi 11,633 3.0 Conclusion Based on the above results, the requirements of Paragraph (g)(2) of ASME Code Case N-504-2 are met, and the repair has been shown to be acceptable for the remaining service life of CNP Unit 1.