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MONTHYEARCP-201500402, Relief Request 1B3-3 Inservice Inspection for Application of an Alternative to ASME Boiler & Pressure Vessel Code Section XI Examination Requirements for Reactor Pressure Vessel Cold Leg Weld Inspection Frequency2015-04-20020 April 2015 Relief Request 1B3-3 Inservice Inspection for Application of an Alternative to ASME Boiler & Pressure Vessel Code Section XI Examination Requirements for Reactor Pressure Vessel Cold Leg Weld Inspection Frequency Project stage: Request CP-201500999, Revision to Relief Request 1B3-3 Inservice Inspection for Application of an Alternative to the ASME Boiler and Pressure Vessel Code Section XI Examination Requirements for Reactor Pressure Vessel Cold Leg Weld Inspection..2015-10-15015 October 2015 Revision to Relief Request 1B3-3 Inservice Inspection for Application of an Alternative to the ASME Boiler and Pressure Vessel Code Section XI Examination Requirements for Reactor Pressure Vessel Cold Leg Weld Inspection.. Project stage: Request ML16015A0032016-01-14014 January 2016 NRR E-mail Capture - Request for Additional Information - Relief Request 1B3-3 for Comanche Peak Nuclear Power Plant, Unit 1 Project stage: RAI ML16057A3102016-02-12012 February 2016 Westinghouse LTR-PAFM-16-8 Attachment B - Responses to the NRC RAIs Re Comanche Peak Unit 1 Extension of Required Inspection Frequency for Reactor Vessel Inlet Nozzle Dissimilar Metal Welds from 7 to 9 Calendar Years Project stage: Other ML16057A3092016-02-15015 February 2016 Response to Request for Additional Information Re Relief Request 1B3-3 Project stage: Response to RAI CP-201600097, Westinghouse LTR-PAFM-16-2-NP, Technical Justification to Support Extended Volumetric Examination Interval for Comanche Peak Unit 1 Reactor Vessel Inlet Nozzle to Safe End Dissimilar Metal Welds.2016-02-29029 February 2016 Westinghouse LTR-PAFM-16-2-NP, Technical Justification to Support Extended Volumetric Examination Interval for Comanche Peak Unit 1 Reactor Vessel Inlet Nozzle to Safe End Dissimilar Metal Welds. Project stage: Other ML16053A4462016-03-10010 March 2016 Request for Withholding Information from Public Disclosure, 2/10/16 Affidavit Executed by J. Gresham, Westinghouse Electric Company LLC LTR-PAFM-16-8 and LTR-PAFM-16-2-P, Revision 0, Relief Request 1B3-3 Project stage: Withholding Request Acceptance ML16074A0012016-03-14014 March 2016 Relief Request 1B3-3, Alternative for Reactor Pressure Vessel Cold Leg Weld Inspection Frequency from Not to Exceed 7 Years to 9 Years for the Third 10-Year Inservice Inspection Interval Project stage: Other 2016-02-12
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Category:Code Relief or Alternative
MONTHYEARML23198A3212023-07-27027 July 2023 Authorization and Safety Evaluation for Proposed Alternative RV-1 ML23121A2202023-05-0404 May 2023 Proposed Alternative P-1 Regarding Safeguard Building Sump Pumps ML22077A8412022-03-24024 March 2022 Proposed Alternative SNB-3 for the Snubber Inservice Program Third Interval Extension ML21021A1002021-02-0101 February 2021 Proposed Alternative to the Requirements of the ASME OM Code to Extend the Inservice Testing Program Interval for Certain Snubbers (EPID L-2020-LLR-0095 (Covid 19)) ML21013A1282021-01-28028 January 2021 Approval of Proposed Alternative to the Requirements of the ASME Code to Extend the Third Inservice Inspection Interval (EPID L-2020-LLR-0092 (COVID-19)) ML21021A0022021-01-28028 January 2021 Approval of Proposed Alternatives 1A4-1 and 1A4-2 from Certain Requirements of 10 CFR 50.55a for Inservice Inspection of Nuclear Power Plants ML20253A1172020-10-0505 October 2020 Approval of Request for Alternative from Certain Requirements of 10 CFR 50.55a for Operation and Maintenance of Nuclear Power Plants (EPIDs L-2020-LLR-0063, L-2020-LLR-0064, and L-2020-LLR-0065 (COVID-19)) CP-202000431, Relief Request 1A4-1 Supplement - Reactor Vessel (Rv) Upper Head Examinations2020-08-0505 August 2020 Relief Request 1A4-1 Supplement - Reactor Vessel (Rv) Upper Head Examinations ML20196L8232020-07-14014 July 2020 Relief Request SNB-1, Snubber Testing ML20196L8242020-07-14014 July 2020 Relief Request 1A3-2, Inservice Inspection (Isi), Table 1 ML20196L8252020-07-14014 July 2020 Relief Request SNB-1, Snubber Testing, Table 1 ML20196L8262020-07-14014 July 2020 Relief Request V-3, Inservice Testing (IST) ML20196L8292020-07-14014 July 2020 Relief Request 1A3-2, Inservice Inspection (ISI) ML20196L8272020-07-14014 July 2020 Relief Request 1A4-2, Reactor Vessel (Rv) Bottom Mounted Instrumentation (Bmi) Nozzle Penetration Examination ML20196L8302020-07-14014 July 2020 Relief Request V-3, Inservice Testing (Ist), Table 1 CP-202000262, Snubber Testing and Snubber Visual Examinations Relief Request2020-04-10010 April 2020 Snubber Testing and Snubber Visual Examinations Relief Request CP-202000261, Inservice Inspection (ISI) and Inservice Testing (IST) Program Relief Requests2020-04-0707 April 2020 Inservice Inspection (ISI) and Inservice Testing (IST) Program Relief Requests ML17095A2512017-04-0707 April 2017 Request for Alternative 2B3-1 Re Examination of Reactor Vessel Closure Head Penetration Nozzles ML16179A4052016-07-11011 July 2016 Relief Request Nos. B-9, B-3, B-10, and B-11 for Piping Welds, Second 10-year Inservice Inspection Interval ML16074A0012016-03-14014 March 2016 Relief Request 1B3-3, Alternative for Reactor Pressure Vessel Cold Leg Weld Inspection Frequency from Not to Exceed 7 Years to 9 Years for the Third 10-Year Inservice Inspection Interval ML16063A0012016-03-11011 March 2016 Relief Request No. B-15, C-2, and C-4 for Welds in the RPV and Containment Spray and Residual Heat Removal Heat Exchanger Shells, Second 10-year Inservice Inspection Interval ML16011A0732016-01-19019 January 2016 Relief Request T-1; Alternative to the ASME OM Code Frequency Specifications for Inservice Testing for the Third 10-Year IST Interval ML15259A0042015-10-30030 October 2015 Relief Request 1B3-4, Alternative for Reactor Pressure Vessel Head Penetration Weld Inspection Frequency (from 10 to 15 Years), Third 10-Year Inservice Inspection Interval ML15257A2422015-09-18018 September 2015 Relief Request 2A3-1, from ASME Code Section XI Requirements, Risk-Informed Process for Selection of Class 1 and Class 2 Piping Weld Examinations, Third 10-Year Inservice Inspection Interval ML15257A2402015-09-15015 September 2015 Relief Request 2C3-1, from ASME Code, Section XI Requirements Reactor Pressure Vessel Leak-Off Flange for the Third 10-Year Inservice Inspection Interval CP-201500671, Relief Request T-1 for Inservice Testing Program for Application of an Alternative to the ASME OM Code Frequency Specifications, (2007 Edition of ASME Code, Section XI, 2008 Addenda Third Interval Start Date: August 3.2015-06-30030 June 2015 Relief Request T-1 for Inservice Testing Program for Application of an Alternative to the ASME OM Code Frequency Specifications, (2007 Edition of ASME Code, Section XI, 2008 Addenda Third Interval Start Date: August 3. ML15090A1042015-04-0303 April 2015 Relief Requests B-7, B-12, and B-13 - Steam Generator Head-to-Tube Sheet Welds for the Second 10-Year Inservice Inspection Interval ML14087A0662014-04-10010 April 2014 Relief Request No. 1/2E-1 for Containment Electrical Penetrations for the Third 10-Year Inservice Inspection Interval ML14073A5442014-04-0101 April 2014 Relief Request No. B-14 for Reactor Pressure Vessel Hot Leg Nozzle Weld Examinations for the Second 10-Year Inservice Inspection Interval ML13158A0932013-06-26026 June 2013 Relief Request No. E-1 Containment Electrical Penetrations, for the Second 10-Year Inservice Inspection Interval ML13148A4372013-06-26026 June 2013 Relief Request No. P-1 for Pumps and Valves, Third 10-Year Inservice Testing Plan Interval ML13113A3792013-05-0808 May 2013 Relief Request No. V-1, from ASME Code for Operation and Maintenance of Nuclear Power Plants Requirements for Pumps and Valves, for the Third 10-Year Inservice Testing Interval ML13046A3852013-03-19019 March 2013 Relief Request C-2 for the Reactor Pressure Vessel Flange Leak-Off Piping, Third 10-Year Inservice Inspection Interval ML13056A5032013-03-15015 March 2013 Relief Request B-2, Alternative to ASME Code Requirements for Examination of Reactor Vessel Hot-Leg Nozzle Welds, for the Third 10-Year Inservice Inspection Interval CP-201300003, Submittal of Relief Request No. B-2, Third 10 Year ISI Interval from 10CFR50.55a Requirements for Reactor Vessel Hot Leg Nozzle Weld Examinations (Third ISI Interval Start Date August 13, 2010)2013-01-16016 January 2013 Submittal of Relief Request No. B-2, Third 10 Year ISI Interval from 10CFR50.55a Requirements for Reactor Vessel Hot Leg Nozzle Weld Examinations (Third ISI Interval Start Date August 13, 2010) CP-201201384, Response to Request for Additional Information for Relief Request No. E-12012-11-14014 November 2012 Response to Request for Additional Information for Relief Request No. E-1 ML12194A2502012-08-14014 August 2012 Relief Request A-1 for Approval of Risk-Informed Alternative to ASME Code, Section XI for Class 1 and 2 Piping Welds, Third 10-Year Inservice Inspection Interval ML1131100922011-12-19019 December 2011 Relief Request No. C-9, Reactor Pressure Vessel Flange Leak-off Piping Configuration, Second 10-Year Inservice Inspection Interval ML1126500832011-11-10010 November 2011 Approval of Relief Request Nos. B-10 and B-11 for the Second 10-Year Inservice Inspection Interval CP-201001546, Relief Request No. B-11 for the Second 10 Years ISI Interval from 10 CFR 50.55a Inspection Requirements Due to Physical Interferences (Second Interval Start Date: August 13, 2000)2010-12-15015 December 2010 Relief Request No. B-11 for the Second 10 Years ISI Interval from 10 CFR 50.55a Inspection Requirements Due to Physical Interferences (Second Interval Start Date: August 13, 2000) CP-201001544, Relief Request No. C-9 for the Second 10 Year Interval from 10 CFR 50.55a Inspection Requirements Due to Hardship (Second Interval Start Date: August 13, 2000)2010-12-15015 December 2010 Relief Request No. C-9 for the Second 10 Year Interval from 10 CFR 50.55a Inspection Requirements Due to Hardship (Second Interval Start Date: August 13, 2000) CP-201000890, Relief Request No. C-7 for the Second 10 Year ISI Interval from 10 CFR 50.55a Inspection Requirements Due to Physical Interferences (Second Interval Start Date: August 13, 2000)2010-12-15015 December 2010 Relief Request No. C-7 for the Second 10 Year ISI Interval from 10 CFR 50.55a Inspection Requirements Due to Physical Interferences (Second Interval Start Date: August 13, 2000) ML0928706372009-12-22022 December 2009 Relief Request B-9 for Unit 1 and B-8 for Unit 2 to Extend the Inservice Inspection Interval for the Reactor Vessel Weld Examination CP-200901227, Revision to Request for Relief to Extend the Unit 1 and 2 in Service Inspection Interval for the Reactor Vessel Weld Examination and Withdrawal of License Amendment Request 09-004 to Add License Condition for Submittal of ISI Informatio2009-09-14014 September 2009 Revision to Request for Relief to Extend the Unit 1 and 2 in Service Inspection Interval for the Reactor Vessel Weld Examination and Withdrawal of License Amendment Request 09-004 to Add License Condition for Submittal of ISI Information an ML0916205482009-07-23023 July 2009 Relief Request P-2, Inservice Testing Plan for Pumps and Valves for Second Interval CP-200801132, Relief Request No. P-2 for the Unit 1 and Unit 2 Inservice Testing Plan for Pumps and Valves (ASME OM Code 1998 Edition, Through 2000 Addenda; Interval Start Date: August 3, 2004, Second Interval)2008-09-24024 September 2008 Relief Request No. P-2 for the Unit 1 and Unit 2 Inservice Testing Plan for Pumps and Valves (ASME OM Code 1998 Edition, Through 2000 Addenda; Interval Start Date: August 3, 2004, Second Interval) ML0821301472008-08-22022 August 2008 Request for Relief B-2 for Second 10-Year Inservice Inspection Interval from 10 CFR 50.55a Inspections Requirements Due to Physical Interferences CP-200800927, (Cpnpp), Relief Request B-8 for Unit 1 Second 10 Year ISI Interval & B-6 for Unit 2 Second 10 Year ISI Interval from 10 CFR 50.55a Requirements for Reactor Vessel Hot & Cold Leg Nozzle Weld Examinations (Unit 1 Second Interval Start..2008-07-10010 July 2008 (Cpnpp), Relief Request B-8 for Unit 1 Second 10 Year ISI Interval & B-6 for Unit 2 Second 10 Year ISI Interval from 10 CFR 50.55a Requirements for Reactor Vessel Hot & Cold Leg Nozzle Weld Examinations (Unit 1 Second Interval Start.. ML0805201952008-03-10010 March 2008 Relief Request B-5 for Second 10-Year ISI Interval from 10 CFR 50.55a Requirements for Class 1 Repair/Replacement of Control Rod Drive Mechanism Canopy Seal Welds ML0804306622008-02-29029 February 2008 Request for Relief No. B-4 from Certain Requirements of ASME Code, Section XI for Implementation of the EPRI-PDI Supplement 11 Program and Application of Weld Overlays 2023-07-27
[Table view] Category:Letter
MONTHYEARCP-202400030, License Renewal Application Revision 0 - Supplement 3, Revision 12024-01-31031 January 2024 License Renewal Application Revision 0 - Supplement 3, Revision 1 IR 05000445/20230042024-01-29029 January 2024 Integrated Inspection Report 05000445/2023004 and 05000446/2023004 CP-202400034, (CPNPP) - Core Operating Limits Report (Colr), Unit 2 Cycle 21, (ERX-23-001, Revision 1)2024-01-29029 January 2024 (CPNPP) - Core Operating Limits Report (Colr), Unit 2 Cycle 21, (ERX-23-001, Revision 1) ML24024A2102024-01-29029 January 2024 Summary of Regulatory Audit Regarding a License Amendment Request to Adopt 10 CFR 50.69, Risk-Informed Categorization and Treatment of Structures, Systems, and Components for Nuclear Power Reactors ML24025A0052024-01-25025 January 2024 Review of the Spring 2023 Steam Generator Tube Inspection Report ML24023A0242024-01-24024 January 2024 Correction to Amendment Nos. 185 and 185 Regarding Implementation of Full Spectrum Loss-of-Coolant Accident Methodology ML24018A1072024-01-18018 January 2024 Notification of Commercial Grade Dedication Inspection (05000445/2024012 and 05000446/2024012) and Request for Information ML23159A2082023-12-20020 December 2023 Request for Withholding Information from Public Disclosure ML23319A3872023-12-20020 December 2023 Issuance of Amendment Nos. 185 and 185 Regarding Implementation of Full Spectrum Loss-of-Coolant Accident (Fsloca) Methodology ML23348A2392023-12-19019 December 2023 Nonacceptance of License Amendment Request to Relocate Technical Specification 3.9.3, Nuclear Instrumentation, to the Technical Requirements Manual CP-202300575, (Cpnpp), License Amendment Request to Adopt 10 CFR 50.69, Risk-Informed Categorization and Treatment of Structures, Systems and Components for Nuclear Power Reactors Supplement 22023-12-13013 December 2023 (Cpnpp), License Amendment Request to Adopt 10 CFR 50.69, Risk-Informed Categorization and Treatment of Structures, Systems and Components for Nuclear Power Reactors Supplement 2 ML23333A0872023-12-13013 December 2023 Transmittal of Dam Safety Inspection Report - Public CP-202300566, (Cpnpp), Special Report 1-SR-23-001-00, Inoperable Post Accident Monitoring Instrumentation2023-12-12012 December 2023 (Cpnpp), Special Report 1-SR-23-001-00, Inoperable Post Accident Monitoring Instrumentation CP-202300494, License Renewal Application Revision 0, Supplement 32023-12-0606 December 2023 License Renewal Application Revision 0, Supplement 3 ML23313A0732023-12-0606 December 2023 Issuance of Amendment Nos. 184 and 184 Regarding Revision to Technical Specifications to Implement WCAP-17661-P-A, Rev. 1, Improved Roac and CAOC Fq Surveillance Technical Specifications ML23291A4382023-11-30030 November 2023 Notice of Availability of the Draft Plant-Specific Supplement 60, to the Generic Environmental Impact Statement for License Renewal of Nuclear Plants Regarding Comanche Peak Nuclear Power Plant, Unit Numbers 1 and 2, License Renewal Applica ML23325A0182023-11-30030 November 2023 Schedule Revision for the License Renewal Application Review IR 05000445/20234022023-11-30030 November 2023 NRC Security Inspection Report 05000445/2023402 and 05000446/2023402 CP-202300349, License Amendment Request (Lar) 23-004 Technical Specifications (TS) 3.9.3, Nuclear Instrumentation2023-11-20020 November 2023 License Amendment Request (Lar) 23-004 Technical Specifications (TS) 3.9.3, Nuclear Instrumentation ML23308A0032023-11-17017 November 2023 Letter to R. Nelson, Executive Director; Achp; Re., Comanche Peak Draft Environmental Impact Statement ML23308A0022023-11-17017 November 2023 Letter to M. Wolfe, Executive Director; Shpo; Re., Comanche Peak Draft Environmental Impact Statement ML23317A3002023-11-13013 November 2023 Letter to R. Sylestine, Chairman, Alabama-Coushatta Tribe of Texas Regarding Comanche Peak Draft Environmental Impact Statement ML23317A2972023-11-13013 November 2023 Letter to R. Martin, President, Tonkawa Tribe of Oklahoma Regarding Comanche Peak Draft Environmental Impact Statement ML23317A2872023-11-13013 November 2023 Letter to J. Garza, Chairman, Kickapoo Traditional Tribe of Texas Regarding Comanche Peak Draft Environmental Impact Statement ML23317A2832023-11-13013 November 2023 Letter to D. Dotson, President, Delaware Nation Regarding Comanche Peak Draft Environmental Impact Statement ML23317A2852023-11-13013 November 2023 Letter to E. Martinez, President, Mescalero Apache Tribe Regarding Comanche Peak Draft Environmental Impact Statement ML23306A0302023-11-13013 November 2023 Letter to J. Cernek, Chairman; Coushatta Tribe of Louisiana Regarding Comanche Peak Draft Environmental Impact Statement ML23317A2902023-11-13013 November 2023 Letter to M. Pierite, Chairman, Tunica Biloxi Tribe of Louisiana Regarding Comanche Peak Draft Environmental Impact Statement ML23317A2982023-11-13013 November 2023 Letter to R. Morrow, Town King, Thlopthlocco Tribal Town Regarding Comanche Peak Draft Environmental Impact Statement ML23317A2842023-11-13013 November 2023 Letter to D. Kaskaske, Chairman, Kickapoo Tribe of Oklahoma Regarding Comanche Peak Draft Environmental Impact Statement ML23317A2822023-11-13013 November 2023 Letter to D. Cooper, Chairman, Apache Tribe of Oklahoma Regarding Comanche Peak Draft Environmental Impact Statement ML23317A2962023-11-13013 November 2023 Letter to M. Woommavovah, Chairman, Comanche Nation Regarding Comanche Peak Draft Environmental Impact Statement ML23317A2812023-11-13013 November 2023 Letter to C. Hoskin, Principal Chief, Cherokee Nation; Regarding Comanche Peak Draft Environmental Impact Statement ML23317A2862023-11-13013 November 2023 Letter to J. Bunch, Chief, United Keetoowah Band of Cherokee Indians Regarding Comanche Peak Draft Environmental Impact Statement ML23317A3032023-11-13013 November 2023 Letter to W. Yargee, Chief, Alabama-Quassarte Tribal Town Regarding Comanche Peak Draft Environmental Impact Statement ML23317A2882023-11-13013 November 2023 Letter to L. Johnson, Chief, Seminole Nation of Oklahoma Regarding Comanche Peak Draft Environmental Impact Statement ML23317A3012023-11-13013 November 2023 Letter to S. Yahola, Mekko, Kialegee Tribal Town Regarding Comanche Peak Draft Environmental Impact Statement ML23317A2792023-11-13013 November 2023 Letter to B. Gonzalez, Chairman, Caddo Nation Regarding Comanche Peak Draft Environmental Impact Statement ML23317A3022023-11-13013 November 2023 Letter to T. Parton, President, Wichita and Affiliated Tribes Regarding Comanche Peak Draft Environmental Impact Statement ML23317A2892023-11-13013 November 2023 Letter to L. Spottedbird, Chairman, Kiowa Indian Tribe Regarding Comanche Peak Draft Environmental Impact Statement ML23311A2082023-11-0909 November 2023 Reassignment of U.S. Nuclear Regulatory Commission Branch Chief in the Division of Operating Reactor Licensing for Plant Licensing Branch IV ML23360A6312023-10-26026 October 2023 FEMA, Submittal of Radiological Emergency Preparedness Final Report for the Comanche Peak Nuclear Power Plant Medical Services Drill Evaluated on August 23, 2023 CP-202300416, Supplemental Information to Facilitate Acceptance of Licensee Amendment Request 23-002, Application Regarding GDC-5 Shared System Requirements2023-10-12012 October 2023 Supplemental Information to Facilitate Acceptance of Licensee Amendment Request 23-002, Application Regarding GDC-5 Shared System Requirements CP-202300432, Response to Request for Additional Information Regarding the Safety Review of the License Renewal Application - Set 42023-10-0404 October 2023 Response to Request for Additional Information Regarding the Safety Review of the License Renewal Application - Set 4 ML23237B4222023-09-28028 September 2023 Energy Harbor Nuclear Corp. - Vistra Operations Company LLC - Letter Regarding Order Approving Transfer of Licenses and Draft Conforming License Amendments ML23263A0242023-09-21021 September 2023 Revision of Schedule for the Environmental Review of the Comanche Peak Nuclear Power Plant Units 1 and 2 License Renewal Application 2024-01-31
[Table view] Category:Safety Evaluation
MONTHYEARML23319A3872023-12-20020 December 2023 Issuance of Amendment Nos. 185 and 185 Regarding Implementation of Full Spectrum Loss-of-Coolant Accident (Fsloca) Methodology ML23313A0732023-12-0606 December 2023 Issuance of Amendment Nos. 184 and 184 Regarding Revision to Technical Specifications to Implement WCAP-17661-P-A, Rev. 1, Improved Roac and CAOC Fq Surveillance Technical Specifications ML23237B4282023-09-28028 September 2023 Energy Harbor Nuclear Corp. - Vistra Operations Company LLC - Enclosure 2, Draft Conforming License Amendments ML23237B4302023-09-28028 September 2023 Energy Harbor Nuclear Corp. - Vistra Operations Company LLC - Enclosure 4, Safety Evaluation for Transfer of Licenses and Draft Conforming License Amendments (EPID L-2023-LLM-0000) (Public) ML23198A3212023-07-27027 July 2023 Authorization and Safety Evaluation for Proposed Alternative RV-1 ML22192A0072022-08-22022 August 2022 Issuance of Amendment Nos. 183 and 183 Regarding the Adoption of Technical Specifications Task Force Traveler TSTF-505, Revision 2 ML22194A0592022-07-14014 July 2022 Correction to Amendment Nos. 182 and 182 to Revise Technical Specifications to Adopt TSTF-577, Revision 1, Revised Frequencies for Steam Generator Tube Inspections ML22129A0722022-05-16016 May 2022 Review of Quality Assurance Program Changes ML22077A8412022-03-24024 March 2022 Proposed Alternative SNB-3 for the Snubber Inservice Program Third Interval Extension ML21321A3492022-02-24024 February 2022 Issuance of Amendment Nos. 182 and 182 to Revise Technical Specifications to Adopt TSTF-577, Rev. 1, Revised Frequencies for Steam Generator Tube Inspections ML21322A1032021-12-0707 December 2021 Proposed Alternative for the Continued Use of a Risk-Informed Process for the Selection of Class 1 and 2 Piping Welds for Inservice Inspection ML21132A0892021-06-0909 June 2021 Issuance of Amendment Nos. 181 and 181 Regarding the Adoption of Technical Specifications Task Force Traveler TSTF-567, Rev. 1, Add Containment Sump TS to Address GSI - 191 Issues ML21061A2172021-05-19019 May 2021 Issuance of Amendment Nos. 180 and 180 to Authorize Revision of Certain Emergency Action Levels of the Emergency Plan ML21103A0392021-04-23023 April 2021 Issuance of Amendment Nos. 179 and 179 the Adoption of Technical Specifications Task Force Traveler, TSTF-569, Revision 2, Revise Response Time Testing Definition (EPID L-2020-0147) ML21015A2122021-02-12012 February 2021 Issuance of Amendment Nos. 178 and 178 Regarding One-Time Revision to Technical Specifications 3.7.8 Station Service Water System (Ssws) and 3.8.1 AC Sources - Operating ML21022A1622021-02-0808 February 2021 Proposed Alternative to the Requirements of the ASME OM Code to Extend the Inservice Testing Program Interval for Certain Check and Relief Valves (EPID L-2020-LLR-0096 (COVID-19)) ML20346A0192021-02-0101 February 2021 Issuance of Amendment Nos. 177 and 177 Regarding Revision to Technical Specifications 3.8.1, AC Sources - Operating ML21021A1002021-02-0101 February 2021 Proposed Alternative to the Requirements of the ASME OM Code to Extend the Inservice Testing Program Interval for Certain Snubbers (EPID L-2020-LLR-0095 (Covid 19)) ML21013A1282021-01-28028 January 2021 Approval of Proposed Alternative to the Requirements of the ASME Code to Extend the Third Inservice Inspection Interval (EPID L-2020-LLR-0092 (COVID-19)) ML21021A0022021-01-28028 January 2021 Approval of Proposed Alternatives 1A4-1 and 1A4-2 from Certain Requirements of 10 CFR 50.55a for Inservice Inspection of Nuclear Power Plants ML20282A7092020-11-17017 November 2020 Proposed Alternative to the Requirements of the ASME Omcode to Extend the Inservice Testing Program Interval for Certain Snubbers (EPID L-2020-LLR-0060 (COVID-19)) ML20281A4722020-11-0404 November 2020 Use of Later Code Edition to the Requirements of the ASME OM Code ML20255A1002020-10-0707 October 2020 Proposed Alternative to the Requirements of the ASME OM Code to Extend the Inservice Test Program Interval for Certain Check and Relief Valves (EPID L-2020-LLR-0061 and EPID L-2020-LLR-0062 (COVID-19)) ML20223A3492020-08-31031 August 2020 Issuance of Amendment Nos. 175 and 175 Regarding One-Time Revision to Technical Specification 3.7.19, Safety Chilled Water ML20226A0132020-08-17017 August 2020 Use of Later Code Edition to the Requirements of the ASME Code ML20167A3182020-07-0606 July 2020 Issuance of Amendment Nos. 174 and 174 Regarding Revision to Technical Specifications to Adopt TSTF-563, Revision 0 ML20108E8782020-04-17017 April 2020 Issuance of Amendment Nos. 173 and 173 Revision to TS 5.5.9, Unit 1 Model D76 and Unit 2 Model D5 Steam Generator (SG) Program (Exigent Circumstances) ML20054A2762020-02-27027 February 2020 Approval of Change to the Quality Assurance Program as Described in the Comanche Peak Nuclear Power Plant Final Safety Analysis Report ML19267A0182019-11-0404 November 2019 Issuance of Amendment Nos. 172 and 172 to Revise Augmentation Times and Emergency Response Organization Staffing for the Emergency Plan ML18304A4872018-11-30030 November 2018 Issuance of Amendments 171 and 171 Regarding Revision to Technical Specifications for Engineered Safety Feature Actuation System Instrumentation ML18267A3842018-10-25025 October 2018 Issuance of Amendment Nos. 170 and 170 Revision to Technical Specification 3.8.4, DC Sources - Operating, Condition B (Exigent Circumstances) ML18221A6322018-08-15015 August 2018 Eicb Safety Evaluation - Comanche Peak Nuclear Power Plant, Units 1 and 2, License Amendment Request to Revise Technical Specification 3.3.2 Engineered Safety Feature Actuation System Instrumentation Docket/Epid 000976/05000446/L-2018-LLA-0 ML17129A0242017-06-29029 June 2017 Comanche Peak Nuclear Power Plant, Units 1 and 2 - Issuance of Amendment Nos. 169 and 169 Re: Administrative Change to Licensee Name (CAC Nos. MF8933 and MF8934) ML17074A4942017-04-13013 April 2017 Issuance of Amendment Nos. 168 and 168 Adoption of TSTF-545, Revision 3, TS Inservice Testing Program Removal & Clarify SR Usage Rule Application to Section 5.5 Testing ML17095A2512017-04-0707 April 2017 Request for Alternative 2B3-1 Re Examination of Reactor Vessel Closure Head Penetration Nozzles ML16334A1732016-12-14014 December 2016 Comanche Peak Nuclear Power Plant, Units 1 And 2; Safety Evaluation Regarding Implementation Of Mitigating Strategies And Reliable Spent Fuel Pool Instrumentation Related To Orders EA-12-049 And EA-12-051 ML16179A4052016-07-11011 July 2016 Relief Request Nos. B-9, B-3, B-10, and B-11 for Piping Welds, Second 10-year Inservice Inspection Interval ML16137A0562016-06-14014 June 2016 Issuance of Amendment Nos. 166 and 166, Request to Revise Emergency Action Levels Based on Nuclear Energy Institute (NEI) 99-01, Revision 6 ML16096A2662016-05-0606 May 2016 Redacted Letter, Order, Safety Evaluation, and Draft Conforming Amendments, Application for Order Approving Direct and Indirect Transfer of Licenses and Conforming License Amendments ML16096A2642016-05-0606 May 2016 Draft Conforming Amendments, Application for Order Approving Direct and Indirect Transfer of Licenses and Conforming License Amendments ML16074A0012016-03-14014 March 2016 Relief Request 1B3-3, Alternative for Reactor Pressure Vessel Cold Leg Weld Inspection Frequency from Not to Exceed 7 Years to 9 Years for the Third 10-Year Inservice Inspection Interval ML16063A0012016-03-11011 March 2016 Relief Request No. B-15, C-2, and C-4 for Welds in the RPV and Containment Spray and Residual Heat Removal Heat Exchanger Shells, Second 10-year Inservice Inspection Interval ML16011A0732016-01-19019 January 2016 Relief Request T-1; Alternative to the ASME OM Code Frequency Specifications for Inservice Testing for the Third 10-Year IST Interval ML15309A0732015-12-30030 December 2015 Issuance of Amendment Nos. 165 and 165, Revise Technical Specification 5.5.16, Containment Leakage Rate Testing Program, to Increase ILRT Test Interval from 10 to 15 Years and Type C Tests from 60 to 75 Mos ML15259A0042015-10-30030 October 2015 Relief Request 1B3-4, Alternative for Reactor Pressure Vessel Head Penetration Weld Inspection Frequency (from 10 to 15 Years), Third 10-Year Inservice Inspection Interval ML15257A2422015-09-18018 September 2015 Relief Request 2A3-1, from ASME Code Section XI Requirements, Risk-Informed Process for Selection of Class 1 and Class 2 Piping Weld Examinations, Third 10-Year Inservice Inspection Interval ML15257A2402015-09-15015 September 2015 Relief Request 2C3-1, from ASME Code, Section XI Requirements Reactor Pressure Vessel Leak-Off Flange for the Third 10-Year Inservice Inspection Interval ML15090A1042015-04-0303 April 2015 Relief Requests B-7, B-12, and B-13 - Steam Generator Head-to-Tube Sheet Welds for the Second 10-Year Inservice Inspection Interval ML15008A1332015-02-24024 February 2015 Issuance of Amendment Nos. 164 and 164, Revise Technical Specification 3.8.1 for a 14-Day Completion Time for Offsite Circuits ML14183A3422014-09-0808 September 2014 Issuance of Amendment Nos. 163 and 163 to Revise License Condition Related to Approval of Revised Cyber Security Plan Implementation Schedule 2023-09-28
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UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 March 14, 2016 Mr. Ken J. Peters Senior Vice President and Chief Nuclear Officer (Acting)
Attention: Regulatory Affairs Luminant Generation Company LLC P.O. Box 1002 Glen Rose, TX 76043
SUBJECT:
COMANCHE PEAK NUCLEAR POWER PLANT, UNIT 1- RELIEF REQUEST 1B3-3, ALTERNATIVE TO THE ASME CODE, SECTION XI, EXAMINATION REQUIREMENTS FOR REACTOR PRESSURE VESSEL COLD-LEG WELD INSPECTION FREQUENCY (CAC NO. MF6125)
Dear Mr. Peters:
By letter dated April 20, 2015 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML15119A216), as supplemented by letters dated October 15, 2015, and February 15, 2016 (ADAMS Accession Nos. ML15300A013 and ML16057A309, respectively), Luminant Generation Company, LLC (the licensee) submitted Relief Request (RR) 1B3-3 to the U.S. Nuclear Regulatory Commission (NRC) for Comanche Peak Nuclear Power Plant (CPNPP), Unit 1, for the third 10-year inservice inspection program interval. Relief Request 1B3-3 requests relief from certain requirements of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (Code) associated with the weld inspection frequency for the reactor pressure vessel cold leg at CPNPP, Unit 1.
Specifically, pursuant to Title 10 of the Code of Federal Regulations (10 CFR) 50.55a(z)(2), the licensee requested to use the proposed alternative for the reactor pressure vessel cold-leg weld inspection frequency as specified in Code Case N-770-1 for a period not to exceed 9 years, instead of a period of not to exceed 7 years, on the basis that complying with the specified requirement would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety.
The NRC staff has reviewed the subject request and concludes, as set forth in the enclosed safety evaluation, that for RR 1B3-3, the proposed alternative provides reasonable assurance of structural integrity of the affected components, and that complying with the specified requirement would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety. Accordingly, the NRC staff concludes that the licensee has adequately addressed all of the regulatory requirements set forth in 10 CFR 50.55a(z)(2) for RR 1B3-3. Therefore, the NRC staff authorizes the use of alternative RR 1B3-3 for CPNPP, Unit 1, until startup from the spring 2019 refueling outage (1 RF20).
K. Peters All other ASME Code,Section XI, requirements for which relief was not specifically requested and approved in this relief request remain applicable, including third-party review by the Authorized Nuclear lnservice Inspector.
If you have any questions, please contact Ms. Margaret Watford of my staff at 301-415-1233 or via e-mail at Margaret.Watford@nrc.gov.
Sincerely, Robert J. Pascarelli, Chief Plant Licensing Branch IV-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-445
Enclosure:
Safety Evaluation cc w/encl: Distribution via Listserv
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELIEF REQUEST 1B3-3 LUMINANT GENERATION COMPANY LLC COMANCHE PEAK NUCLEAR POWER PLANT, UNIT 1 DOCKET NO. 50-445
1.0 INTRODUCTION
By letter dated April 20, 2015 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML15119A216), as supplemented by letters dated October 15, 2015, and February 15, 2016 (ADAMS Accession Nos. ML15300A013 and ML16057A309, respectively), Luminant Generation Company, LLC (the licensee) submitted Relief Request (RR) 1B3-3 to the U.S. Nuclear Regulatory Commission (NRC) for Comanche Peak Nuclear Power Plant (CPNPP), Unit 1, for the third 10-year inservice inspection (ISi) program interval.
Relief Request 1B3-3 requests relief from certain requirements of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (Code) associated with the weld inspection frequency for the reactor pressure vessel (RPV) cold leg at CPNPP, Unit 1.
Specifically, pursuant to Title 10 of the Code of Federal Regulations (10 CFR) 50.55a(z)(2), the licensee requested to use the proposed alternative for the RPV cold-leg weld inspection frequency as specified in Code Case N-770-1 for a period not to exceed 9 years, instead of a period of not to exceed 7 years, on the basis that complying with the specified requirement would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety.
2.0 REGULATORY EVALUATION
The ISi of ASME Code Class 1, 2, and 3 components is to be performed in accordance with Section XI, "Rules for lnservice Inspection of Nuclear Power Plant Components," of the ASME Code and applicable editions and addenda as required by 10 CFR 50.55a(g), except where specific written relief has been granted by the Commission.
Paragraph 10 CFR 50.55a(g)(6)(ii) states that the Commission may require the licensee to follow an augmented ISi program for systems and components for which the Commission deems that added assurance of structural reliability is necessary. Paragraph 10 CFR 50.55a(g)(6)(ii)(F) requires, in part, augmented inservice volumetric inspection of Class 1 piping and nozzle dissimilar welds (OM) butt welds of pressurized-water reactors in accordance with ASME Code Case N-770-1, subject to the conditions specified in paragraphs (2) through (10) of 10 CFR 50.55a(g)(6)(ii)(F).
Enclosure
Alternatives to requirements under 10 CFR 50.55a(g) may be authorized by the NRC pursuant to 10 CFR 50.55a(z)(1) or 10 CFR 50.55a(z)(2). In proposing alternatives or requests for relief, the licensee must demonstrate that: (1) the proposed alternatives would provide an acceptable level of quality and safety; or (2) compliance with the specified requirements would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety.
Based on the analysis of the regulatory requirements, and subject to the following technical evaluation, the NRC staff concludes that regulatory authority exists for the licensee to request the use of an alternative and the NRC to authorize the alternative proposed by the licensee on the basis that compliance with the specified requirements would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety. Accordingly, the NRC staff has reviewed and evaluated the licensee's request pursuant to 10 CFR 50.55a(z)(2).
3.0 TECHNICAL EVALUATION
3.1 Licensee Relief Request 3.1.1 Component Identification The affected components are as follows:
Weld TBX-1-4100-14, Loop 1 cold-leg nozzle to safe-end weld Weld TBX-1-4200-14, Loop 2 cold-leg nozzle to safe-end weld Weld TBX-1-4300-14, Loop 3 cold-leg nozzle to safe-end weld Weld TBX-1-4400-14, Loop 4 cold-leg nozzle to safe-end weld 3.1.2 Code Requirements for Which Relief is Requested Volumetric inspection of RPV inlet cold-leg nozzle to safe-end DM welds of pressurized-water reactors are required in accordance with ASME Code Case N-770-1, subject to the conditions specified in paragraphs (2) through (10) of 10 CFR 50.55a(g)(6)(ii)(F). ASME Code Case N-770-1, Table 1, Inspection Item B requires volumetric examination of essentially 100 percent of each weld once every second inspection period not to exceed 7 years.
3.1.3 Licensee's Proposed Alternative and Duration of Relief The licensee proposed a one-time extension to the Code Case N-770-1, Table 1, Inspection Item 8, volumetric examinations from a period of "not to exceed 7 years" to a period of "not to exceed 9 years."
The licensee proposed an alternative to the regulatory requirement, which would reschedule the inspections from spring 2016 refueling outage to the spring 2019 refueling outage. This is a one-time extension inspection frequency request.
3.1.4 Licensee's Basis for Relief The licensee stated that the relief request was due to the need to examine the RPV inlet cold-leg nozzle to safe-end welds from the inside surface of the weld. This requires access to the lower portion of the RPV to insert automated volumetric inspection equipment to perform the examination. As such, it would be necessary to remove the core barrel and other RPV internals. The core barrel is scheduled to be removed for inspection of the vessel shell welds and vessel internal inspection required by Electric Power Research Institute (EPRI) report 1022863, "Materials Reliability Program: Pressurized Water Reactor Internals Inspection and Evaluation Guidelines (MRP-227-A)," December 2011 (ADAMS Accession No. ML120170453),
during the spring 2019 refueling outage. Requiring the additional removal of the core barrel and other internals during the spring 2016 refueling outage would result in additional radiological personnel dose and an additional heavy load lift in containment.
The licensee discussed in its submittal that the technical basis for the relief request included the considerations that: ( 1) there has been no service experience with cracking found in any RPV inlet cold-leg DM welds; (2) crack growth rates in RPV inlet cold-leg DM welds are slow; (3) the likelihood of initial cracking; crack growth and a subsequent through-wall leak is very small in these welds; and (4) the specific axial and circumferential flaw evaluation showing that any indication detected during the previous inspection, as well as any flaw size which could have been reasonably missed during RPV inlet cold-leg DM weld examination, would not grow to the allowable size flaw specified by ASME Code,Section XI, during the requested inspection interval. The licensee provided this technical basis to demonstrate that it is acceptable to extend the re-examination interval.
The licensee also stated that since primary water stress-corrosion cracking (PWSCC) is temperature dependent, it would be expected that hot-leg temperature welds would show evidence of crack initiation before cold-leg temperature welds, and no evidence of cracking has been identified in either hot-leg or cold-leg welds. Further, the cold-leg temperature welds that are the subject of this relief request were inspected in spring 2010 with volumetric techniques.
No indications were identified in the welds.
The licensee provided a plant-specific flaw analysis for the RPV inlet cold-leg DM welds to support its proposal. The analysis was developed and based on the most recent guidance of EPRI report 1021023, "Materials Reliability Program: Primary Water Stress Corrosion Cracking (PWSCC) Flaw Evaluation Guidance (MRP-287)," December 2010 (publicly available at http://www.epri.com/abstracts/Pages/ProductAbstract. aspx?Productld=OOOOOOOOOOO 1021023).
In development of the weld residual stresses, the licensee included the effects of a hypothetical 50 percent through-wall inside diameter surface weld repair. In summary, the licensee stated that the calculations show that in order for a flaw to have grown to a depth of 75 percent through-weld by the next inspection in spring 2019, an axial flaw would have had to have been 57 percent through-weld thickness or a circumferential flaw would have had to have been 33 percent through-weld thickness, during the previous inspection in spring of 2010. The licensee stated that based on the current inspection capabilities, the flaw sizes above are significantly larger than the theoretical flaw detection limit and the minimum size flaw actually detected during the previous inspection.
Finally, the licensee included a discussion on the probability of cracking or leakage from these welds. The licensee stated that there were analyses and sensitivity studies which showed the likelihood of cracking or through-wall leaks was very small and that more frequent inspections had only a small benefit in terms of risk. Further, the licensee noted that there was no operational experience of cracking in RPV inlet cold leg DM welds, and the number of indications in RPV hot-leg DM welds, a leading indicator of cracking due to temperature, was small when compared to the number of those welds in service.
Additionally, the licensee stated that volumetric inspection of the RPV inlet cold-leg nozzle to safe-end welds from the outside surface would be undesirable due to the welds being located inside a sandbox and covered with insulation. The sandbox was installed during original plant construction after all welding was completed.
Therefore, the licensee concluded that extending the required RPV inlet cold-leg DM weld volumetric examination until spring 2019 is justified.
As such, the licensee concluded that the technical basis was sufficient to ensure public health and safety by extending the inspection frequency of the RPV inlet cold-leg nozzle to safe-end DM welds from a maximum of 7 years to a new maximum of 9 years.
3.2 NRC Staff Evaluation The NRC staff notes that the generic rules for the frequency of volumetric examination of DM butt welds were established to provide reasonable assurance of the leak tightness and structural integrity of the reactor coolant pressure boundary. As such, the staff finds that a plant-specific analysis could be used to provide a reasonable basis for inspection relief if the inspection frequency can be shown to maintain reasonable assurance of the leak tightness and structural integrity of the weld. As such, the staff reviewed the licensee's proposed alternative under the requirements of 10 CFR 50.55a(z)(2) that:
Compliance with the specified requirements would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety.
The NRC staff reviewed the licensee's basis for hardship. Since the RPV inlet cold-leg DM welds are located in sandboxes, and inspection of the welds would require the licensee to remove the RPV core barrel only for these examinations in spring 2016 refueling outage (1RF18), the staff concludes that the licensee has a sufficient basis for hardship due to the expected radiological dose for the work.
Given the basis for hardship, the licensee's technical justification for the proposed alternative is that it provides reasonable assurance of structural integrity and leak tightness. The licensee also stated that no flaw of a size that could potentially have been missed during the 2010 refueling outage inspection could reasonably grow to an unacceptable size prior to the proposed inspection in 2019. The NRC staff reviewed the licensee's inspection results and flaw analysis to assess the acceptability of the proposed alternative.
The NRC staff reviewed each of the licensee's four bases: (1) there has been no service experience with cracking found in any RPV inlet cold-leg DM welds; (2) crack growth rates in
RPV inlet cold-leg OM welds are slow; (3) the likelihood of initial cracking, crack growth, and a subsequent through-wall leak is very small in these welds; and (4) the specific axial and circumferential flaw evaluation. The NRC staff notes that cracking has been identified in cold-leg temperature OM welds of smaller pipe size than the RPV inlet nozzle. Furthermore, cracking has been found in other locations at cold-leg temperatures in the reactor coolant pressure boundary. Additionally, reviews of the inspection data from those flaws have shown faster than average growth rates for the cold-leg temperatures. The staff notes that the RPV inlet cold-leg OM welds are made with weld materials susceptible to PWSCC and that this type of cracking can initiate and grow at cold-leg temperatures. As such, the staff concludes that for the licensee's items (1) - (3), there is insufficient basis to assume that cracking could not occur in these welds over time, and those flaws could not grow to a size that could challenge leak tightness or structural integrity.
However, the licensee's fourth basis, regarding the plant-specific axial and circumferential flaw evaluation, assumes that a hypothetical flaw could exist and provides an assessment of the potential growth of that flaw over time. The NRC staff reviewed the analysis and concluded that it could provide a basis to demonstrate leak tightness and structural integrity. Therefore, the staff focused its review on this aspect of the licensee's basis for the proposed alternative.
The licensee's flaw analysis is composed of a stress analysis and a flaw growth calculation.
The NRC staff reviewed the licensee's stress analysis and found that it followed the recommendations of MRP-287 on effective weld residual stress calculations to address PWSCC flaw analysis. To add significant conservatism, a 50 percent inside surface weld repair 360 degrees around the circumference was simulated in the weld residual stress analysis. The fabrication sequence was simulated based on information provided in the plant-specific drawings. The staff also concluded that that the use of two stress paths, calculated for both hoop and axial stresses, was effective and consistent with NRC staff expectations. The staff reviewed the final proprietary stress analysis through the thickness of the weld and found both the hoop and axial residual stress curve contours were consistent with the MRP-287 analyses using similar geometries and fabrication methods. Based on its review, the NRC staff concluded that the licensee's plant-specific stress analysis for the subject welds to have conservative inputs and assumptions and, therefore, was adequate to be used in the flaw evaluation.
The NRC staff reviewed the licensee's previous inspection methods and results to assess the licensee's basis for assuming a maximum hypothetical initial flaw size during the 2010 outage.
The basis included an ASME Code,Section XI, Appendix VIII demonstrated volumetric examination obtaining essentially 100 percent coverage that found no indications of surface connected flaws. The staff concluded that the licensee's qualified inspection techniques provide a reasonable basis that any flaw connected to the wetted surface with a size of 10 percent in depth or greater should have been identified. Also, the staff concluded that the licensee's data and supporting inspection results provided a reasonable basis for the initial flaw size assumptions.
The NRC staff assessed the licensee's proposed alternative by performing a series of flaw evaluations. The staff's flaw evaluations demonstrated that there is sufficient margin between the hypothetical maximum size of the postulated flaw after 9 years of growth and the ASME Code allowable flaw size, which supports the licensee's proposed alternative.
Therefore, based on the hardship of the increased radiological dose required to perform the required examinations due to the location of the RPV inlet cold-leg nozzle to safe-end DM welds in the sandboxes, and the licensee's flaw analysis demonstrating a sufficient safety margin, the NRC staff concludes that the licensee has provided an adequate technical basis to demonstrate that compliance with the requirements of 10 CFR 50.55a(g)(6)(ii)(F) for the volumetric inspection of the RPV inlet cold-leg nozzle to safe-end DM welds during the spring 2016 refueling outage would result in a hardship or unusual difficulty without a compensating increase in the level of quality and safety.
4.0 CONCLUSION
Based on the above, the NRC staff concludes that that the proposed alternative provides reasonable assurance of structural integrity of the subject components and that complying with the requirement would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety. Accordingly, the NRC staff concludes that the licensee has adequately addressed all of the regulatory requirements set forth in 10 CFR 50.55a(z)(2).
Therefore, the NRC staff authorizes the use of the proposed alternative RR 183-3 at CPNPP, Unit 1, until startup from the spring 2019 refueling outage (1 RF20).
All other ASME Code,Section XI, requirements for which relief was not specifically requested and approved in this relief request remain applicable, including third-party review by the Authorized Nuclear lnservice Inspector.
Principal Contributor: M. Audrain, NRR/DE/EPNB Date: March 14, 2016
K. Peters All other ASME Code,Section XI, requirements for which relief was not specifically requested and approved in this relief request remain applicable, including third-party review by the Authorized Nuclear lnservice Inspector.
If you have any questions, please contact Ms. Margaret Watford of my staff at 301-415-1233 or via e-mail at Margaret.Watford@nrc.gov.
Sincerely, IRA/
Robert J. Pascarelli, Chief Plant Licensing Branch IV-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-445
Enclosure:
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