|
---|
Category:Code Relief or Alternative
MONTHYEARML23256A0882023-09-25025 September 2023 Issuance of Alternative to Steam Generator Welds ML23151A3482023-05-30030 May 2023 Duke Fleet - Request for Additional Information Proposed Alternative for Pressurizer Welds in Accordance with 10 CFR 50.55a(z)(1) ML23118A0762023-05-0101 May 2023 Approval for Use of Specific Provision of a Later Edition of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code, Section XI RA-22-0257, Proposed Alternative for Pressurizer Welds in Accordance with 10 CFR 50.55a(z)(1)2023-02-17017 February 2023 Proposed Alternative for Pressurizer Welds in Accordance with 10 CFR 50.55a(z)(1) ML22096A0032022-11-18018 November 2022 McGuire Nuclear Station and Shearon Harris Nuclear Power Plant Authorization of RA-19-0352 Regarding Use of Alternative for RPV Head Closure Stud Examinations ML21029A3352021-02-16016 February 2021 Request to Use a Provision of a Later Edition of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code, Section XI RA-20-0312, Relief Request, Service Water Pinhole Leak in a Socket Weld, Inservice Inspection Program, Fourth Ten-Year Interval2020-11-0404 November 2020 Relief Request, Service Water Pinhole Leak in a Socket Weld, Inservice Inspection Program, Fourth Ten-Year Interval ML20080G9502020-04-0505 April 2020 Relief from the Requirements of the ASME Code ML19018A0252019-02-26026 February 2019 Relief Request I3R-18, Regarding Alternative Repair and Replacement Testing Requirements for the Containment Building Equipment Hatch Sleeve Weld, Inservice Inspection Program for Containment, Third Ten-Year Interval RA-19-0026, Response to Request for Additional Information (RAI) Regarding Proposed Alternative to the Depth Sizing Qualification Requirement of Appendix Viii, Supplements 2 and 10 (18-GO-001)2019-02-11011 February 2019 Response to Request for Additional Information (RAI) Regarding Proposed Alternative to the Depth Sizing Qualification Requirement of Appendix Viii, Supplements 2 and 10 (18-GO-001) ML18283B5442018-11-20020 November 2018 Relief from the Requirements of the ASME Code, Section XI, Reactor Vessel Closure Head Penetration Nozzle Repair Technique RA-18-0180, Supplement to Relief Request for an Alternative to the Depth Sizing Qualification Requirement of Appendix Viii, Supplements 22018-11-12012 November 2018 Supplement to Relief Request for an Alternative to the Depth Sizing Qualification Requirement of Appendix Viii, Supplements 2 ML17331A0862017-12-26026 December 2017 Alternative to Inservice Inspection Regarding Reactor Pressure Vessel Threads ML15342A0432016-01-0606 January 2016 Supplemental Safety Evaluation to Relief Request I3R-15 for Reactor Vessel Closure Head Penetration Nozzle Repair Technique, Inservice Inspection Program - Third 10-Year Interval ML15203A7022015-09-18018 September 2015 Relief Request 13R-15 for Reactor Vessel Closure Head Penetration Nozzle Repair Technique, Inservice Inspection Program - Third 10 - Year Interval ML14093A0752014-04-11011 April 2014 Relief Request I3R-13 for Reactor Vessel Closure Head Penetration Nozzle 37 Repair Inservice Inspection Program - Third 10-Year Interval ML13238A1542013-09-13013 September 2013 Relief Request I3R-11 for Reactor Vessel Closure Head Penetration Nozzles Repair Inservice Inspection Program - Third 10-Year Interval ML13017A3492013-01-28028 January 2013 Relief Request I3R-10 for Spent Fuel Pool Cooling Piping Inservice Inspection Program - Third 10-Year Interval HNP-12-133, Relief Request I3R-10 Spent Fuel Pool Cooling Piping, Request for Additional Information Response2012-12-17017 December 2012 Relief Request I3R-10 Spent Fuel Pool Cooling Piping, Request for Additional Information Response ML12270A2582012-10-0202 October 2012 Relief Request 13R-09 for Reactor Vessel Closure Head Penetration Nozzles Repair Inservice Inspection Program Third 10-Year Interval HNP-12-060, Relief Request I3R-09 Reactor Vessel Closure Head Nozzle Repairs Inservice Inspection Program - Third Interval, Request for Additional Information Response2012-05-18018 May 2012 Relief Request I3R-09 Reactor Vessel Closure Head Nozzle Repairs Inservice Inspection Program - Third Interval, Request for Additional Information Response HNP-12-054, Relief Request I3R-09 Reactor Vessel Closure Head Nozzles Inservice Inspection Program - Third Interval2012-05-0303 May 2012 Relief Request I3R-09 Reactor Vessel Closure Head Nozzles Inservice Inspection Program - Third Interval ML1014800292010-06-17017 June 2010 Acceptance Review Regarding a Relief Request to Defer an American Society of Mechanical Engineers Code-Required Repair of a Service Water System Weld Unit the Refueling Outage ML1008204632010-04-0808 April 2010 Request for Relief 2R1-018 for the Second 10-Year Interval Inservice Inspection Program Plant ML0912701882009-05-19019 May 2009 Acceptance Review Regarding Relief Requests on Final Documentation for the Second 10-Year Inservice Inspection Program for Limited Coverage of Welds in Multiple Exam Categories HNP-08-045, Second Ten-Year Interval Inservice Inspection Program-Final Documentation Including Requests for Relief in Accordance with 10 CFR 50.55a2009-02-0505 February 2009 Second Ten-Year Interval Inservice Inspection Program-Final Documentation Including Requests for Relief in Accordance with 10 CFR 50.55a ML0832601362009-01-21021 January 2009 Request for Relief No. 13R-04 for the Third 10-Year Inservice Inspection Interval Regarding System Pressure Testing of Class 1 Systems (Tac MD8744) ML0832600992009-01-15015 January 2009 Request for Relief No. 13R-03 for the Third 10-Year Inservice Inspection Interval Regarding the System Pressure Test ML0727607372007-10-10010 October 2007 Relief Request No. 1 Regarding Proposed Alternative to the American Society of Mechanical Engineers Boiler and Pressure Vessel Code Requirements for Pressurizer Nozzle Weld Overlay Repairs HNP-07-073, Withdrawal of the Relief Request from ASME Code Inservice Testing Requirements for the Second 10-Year Interval2007-05-15015 May 2007 Withdrawal of the Relief Request from ASME Code Inservice Testing Requirements for the Second 10-Year Interval ML0608703872006-04-17017 April 2006 Relief Request No. 2R1-015 for the Second 10-Year Inservice Inspection Interval ML0608700432006-04-0707 April 2006 Relief, Relief Request 2R1-016 and 2R1-017 for the Second 10 Yr Inservice Inspection Interval HNP-05-128, Relief Request from Inservice Inspection Program No. 2R1-015 Alternative to ASME Code Section XI, IWA-4000 Requirements2005-11-18018 November 2005 Relief Request from Inservice Inspection Program No. 2R1-015 Alternative to ASME Code Section XI, IWA-4000 Requirements ML0412900022004-05-0303 May 2004 Relief, ASME Code, Section XI, Appendix Viii, Supplement 10 2023-09-25
[Table view] Category:Letter
MONTHYEARIR 05000400/20230042024-01-30030 January 2024 Integrated Inspection Report 05000400/2023004 ML23317A3462023-11-14014 November 2023 Duke Fleet - Correction Letter to License Amendment Nos. 312 & 340 Issuance of Amendments Regarding the Adoption of Technical Specifications Task Force Traveler TSTF-554, Revision 1 IR 05000400/20230032023-11-0909 November 2023 Integrated Inspection Report 05000400/2023003 ML23346A1322023-10-0606 October 2023 Communication from C-10 Research & Education Foundation Regarding NextEra Common Emergency Fleet Plan License Amendment Request and Related Documents Subsequently Published ML23234A1702023-10-0303 October 2023 Issuance of Amendment No. 199 Regarding Administrative Changes to the Renewed Facility Operating License and Technical Specifications ML23256A0882023-09-25025 September 2023 Issuance of Alternative to Steam Generator Welds ML23195A0782023-08-29029 August 2023 Issuance of Amendments Regarding the Adoption of Technical Specifications Task Force Traveler TSTF-554, Revision 1 IR 05000400/20230052023-08-23023 August 2023 Updated Inspection Plan for Shearon Harris Nuclear Power Plant, Unit 1 (Report 05000400/2023005) ML23234A2542023-08-22022 August 2023 RQ Inspection Notification Letter IR 05000400/20234022023-07-26026 July 2023 Security Baseline Inspection Report 05000400/2023402 IR 05000400/20230022023-07-24024 July 2023 Integrated Inspection Report 05000400/2023002 IR 05000400/20234402023-07-17017 July 2023 Special Inspection Report 05000400/2023440 and Preliminary Greater than Green Finding and Apparent Violation Cover Letter IR 05000400/20243012023-05-15015 May 2023 Notification of Licensed Operator Initial Examination 05000400/2024301 IR 05000400/20230012023-05-10010 May 2023 Integrated Inspection Report 05000400 2023001 IR 05000400/20234042023-05-0404 May 2023 Cyber Security Inspection Report 05000400/2023404 ML23118A0762023-05-0101 May 2023 Approval for Use of Specific Provision of a Later Edition of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code, Section XI ML23118A1392023-04-28028 April 2023 Submittal of Updated Final Safety Analysis Report (Amendment 65), Technical Specification Bases Revision, Report of Changes Pursuant to 10 CFR 50.59 and Summary of Commitment Changes IR 05000400/20234032023-04-0505 April 2023 Security Baseline Inspection Report 05000400/2023403 IR 05000400/20230102023-03-15015 March 2023 Comprehensive Engineering Team Inspection (CETI) Inspection Report 05000400/2023010 ML22332A4932023-03-10010 March 2023 William States Lee III 1 and 2 - Issuance of Amendments Regarding the Relocation of the Emergency Operations Facility IR 05000400/20220062023-03-0101 March 2023 Annual Assessment Letter for Shearon Harris Nuclear Power Plant - NRC Inspection Report 05000400/2022006 ML23033A5272023-02-0808 February 2023 Correction of Typographical Errors Incurred During Issuance of License Amendment No. 196 IR 05000400/20220042023-02-0707 February 2023 Integrated Inspection Report 05000400/2022004 ML23020A1252023-01-23023 January 2023 Notification of Target Set Inspection and Request for Information (NRC Inspection Report 05000400/2023403) ML22096A0032022-11-18018 November 2022 McGuire Nuclear Station and Shearon Harris Nuclear Power Plant Authorization of RA-19-0352 Regarding Use of Alternative for RPV Head Closure Stud Examinations ML22256A2532022-11-14014 November 2022 Issuance of Amendments Regarding the Adoption of Technical Specifications Task Force Traveler TSTF-541, Rev. 2 IR 05000400/20220032022-11-10010 November 2022 Integrated Inspection Report 05000400/2022003 IR 05000400/20223012022-11-0202 November 2022 NRC Operator License Examination Report 05000400/2022301 ML22271A6202022-09-29029 September 2022 Notification of Shearon Harris Nuclear Power Plant Comprehensive Engineering Team Inspection - U.S. Nuclear Regulatory Commission Inspection Report 05000400/2023010 ML22258A1262022-09-14014 September 2022 NRC Operator Licensing Examination Approval 05000400/2022301 ML22242A0022022-09-12012 September 2022 Issuance of Amendments to Adopt TSTF 569, Revision 2, Revise Response Time Testing Definition ML22227A0682022-09-0202 September 2022 Alternative to Certain Inservice Testing Requirements in the American Society of Mechanical Engineers Code for Operation and Maintenance for Certain Target Rock Solenoid Valves IR 05000400/20220052022-08-24024 August 2022 Updated Inspection Plan for Shearon Harris Nuclear Power Plant (Report 05000400/2022005) IR 05000400/20224032022-08-23023 August 2022 Security Baseline Inspection Report 05000400/2022403 ML22126A0082022-08-0909 August 2022 Issuance of Amendment No. 194 Revise Technical Specifications Related to Reactor Protection System Instrumentation P7 IR 05000400/20220022022-08-0505 August 2022 Integrated Inspection Report 05000400/2022002 ML22161B0332022-07-28028 July 2022 Issuance of Amendment No. 193 Regarding Revision of Surveillance Requirements to Remove Shutdown Limitation IR 05000400/20220112022-07-20020 July 2022 Biennial Problem Identification and Resolution Inspection Report 05000400 2022011 IR 05000400/20224022022-06-30030 June 2022 Material Control and Accounting Program Inspection Report 05000400/2022402 (OUO Removed) IR 05000400/20220102022-06-13013 June 2022 Design Basis Assurance Inspection (Programs) Inspection Report 05000400/2022010 ML22138A4012022-05-26026 May 2022 Project Manager Reassignment ML22101A2822022-05-0606 May 2022 Regulatory Audit Summary Related to the Review of License Amendment Regarding Revising the Flood Hazard Protection Scheme IR 05000400/20220012022-05-0404 May 2022 Integrated Inspection Report 05000400/2022001 IR 05000400/20224012022-03-24024 March 2022 Security Baseline Inspection Report 05000400/2022401 ML22020A0072022-03-10010 March 2022 Issuance of Amendment No. 192 Regarding Removal of Extraneous Content and Requirements from the Renewed Facility Operating License and Technical Specifications ML22010A2812022-03-0404 March 2022 Issuance of Amendments to Adopt TSTF-577, Rev. 1 Revised Frequencies for Steam Generator Tube Inspections (EPID L-2021-LLA-0161 IR 05000400/20210062022-03-0202 March 2022 Annual Assessment Letter for Shearon Harris Nuclear Power Plant (Report No. 05000400/2021006) ML21351A4722022-02-10010 February 2022 Issuance of Amendment No. 190 Regarding Revision to Containment Spray Nozzle Test Frequency IR 05000400/20210042022-02-0404 February 2022 Integrated Inspection Report 05000400/2021004 ML22034A5352022-02-0202 February 2022 Notification of Licensed Operator Initial Examination 05000400/2022301 2024-01-30
[Table view] Category:Safety Evaluation
MONTHYEARML23234A1702023-10-0303 October 2023 Issuance of Amendment No. 199 Regarding Administrative Changes to the Renewed Facility Operating License and Technical Specifications ML23256A0882023-09-25025 September 2023 Issuance of Alternative to Steam Generator Welds ML23195A0782023-08-29029 August 2023 Issuance of Amendments Regarding the Adoption of Technical Specifications Task Force Traveler TSTF-554, Revision 1 ML23118A0762023-05-0101 May 2023 Approval for Use of Specific Provision of a Later Edition of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code, Section XI ML22332A4932023-03-10010 March 2023 William States Lee III 1 and 2 - Issuance of Amendments Regarding the Relocation of the Emergency Operations Facility ML22096A0032022-11-18018 November 2022 McGuire Nuclear Station and Shearon Harris Nuclear Power Plant Authorization of RA-19-0352 Regarding Use of Alternative for RPV Head Closure Stud Examinations ML22256A2532022-11-14014 November 2022 Issuance of Amendments Regarding the Adoption of Technical Specifications Task Force Traveler TSTF-541, Rev. 2 ML22242A0022022-09-12012 September 2022 Issuance of Amendments to Adopt TSTF 569, Revision 2, Revise Response Time Testing Definition ML22227A0682022-09-0202 September 2022 Alternative to Certain Inservice Testing Requirements in the American Society of Mechanical Engineers Code for Operation and Maintenance for Certain Target Rock Solenoid Valves ML22126A0082022-08-0909 August 2022 Issuance of Amendment No. 194 Revise Technical Specifications Related to Reactor Protection System Instrumentation P7 ML22161B0332022-07-28028 July 2022 Issuance of Amendment No. 193 Regarding Revision of Surveillance Requirements to Remove Shutdown Limitation ML22083A0272022-03-23023 March 2022 Safety Evaluation -Harris Safety Evaluation -EPID L-2021-LLA-0149 - Revise TS Related to Reactor Protection System ML22020A0072022-03-10010 March 2022 Issuance of Amendment No. 192 Regarding Removal of Extraneous Content and Requirements from the Renewed Facility Operating License and Technical Specifications ML22010A2812022-03-0404 March 2022 Issuance of Amendments to Adopt TSTF-577, Rev. 1 Revised Frequencies for Steam Generator Tube Inspections (EPID L-2021-LLA-0161 ML21351A4722022-02-10010 February 2022 Issuance of Amendment No. 190 Regarding Revision to Containment Spray Nozzle Test Frequency ML21320A0012022-01-20020 January 2022 Issuance of Amendment No. 189 to Regarding Administrative Change to Reflect Development of a Technical Requirements Manual ML21316A2482022-01-19019 January 2022 Issuance of Amendment No. 188, Regarding Revision of 10 CFR 50.69, Risk-Informed Categorization and Treatment of Structures, Systems and Components for Nuclear Power Reactors ML21155A2132021-08-26026 August 2021 Plant ML21113A0132021-05-0606 May 2021 Request for Use of a Later Edition of ASME Boiler and Pressure Vessel Code, Section XI for Repair and Replacement Activities ML21047A4702021-04-0808 April 2021 Issuance of Amendment No. 185 Regarding Reduction of Reactor Coolant System Minimum Flow Rate and Update to the Core Operating Limits Report References ML21047A3142021-04-0202 April 2021 Issuance of Amendment No. 184 Regarding Technical Specifications Task Force (TSTF) Traveler TSTF-505, Revision 2, Provide Risk Informed Extended Completion Times RITSTF Initiative 4B ML21033B0072021-03-22022 March 2021 Issuance of Amendment No, 183 to Correct Non-Conservative Technical Specification Related to Reactor Coolant System Pressure/Temperature Limits ML21069A0642021-03-18018 March 2021 Alternative to ASME Section XI Code Requirements to Defer Repair of Emergency Service Water System Socket Weld ML21035A1322021-03-16016 March 2021 Issuance of Amendment No. 182 Regarding Technical Specifications Related to Accident Monitoring Instrumentation, and Refueling Operations Instrumentation ML21029A3352021-02-16016 February 2021 Request to Use a Provision of a Later Edition of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code, Section XI ML20259A5122020-12-0808 December 2020 Issuance of Amendment No. 181 Regarding Adoption of 10 CFR Part 50, Appendix J, Option B, for Type B and C Testing and for Permanent Extension of Type a, B, and C Leak Rate Testing Frequencies(L-2019-LLA-0256) ML20300A2062020-11-0909 November 2020 Duke Fleet - Request to Use a Provision of a Later Edition of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code, Section XI ML20281A2792020-10-13013 October 2020 Correction Letter to Amendment No. 179 Regarding Departure from Nucleate Boiling Ratio Safety Limit to Address Transition to New Fuel Type ML20212L5942020-09-29029 September 2020 Issuance of Amendment No. 179 Regarding Departure from Nucleate Boiling Ratio Safety Limit to Address Transition to New Fuel Type ML20258A2712020-09-21021 September 2020 Correction Letter to Amendment Nos. 301, 329, 307, 303, 178, 317, 296, 416, 418, 417, and 268 ML20237F4352020-09-0404 September 2020 Issuance of Amendments Regarding the Technical Specification Requirements Pertaining to Mode Change Limitations ML20083F9272020-08-12012 August 2020 Issuance of Amendments to Relocate the Unit/Facility/Plant Staff Qualification Requirements to the Duke Energy Quality Assurance Program Description ML20099F5052020-06-29029 June 2020 Issuance of Amendment No. 177 Regarding Elimination of Certain Technical Specification Requirements in Alignment with Improved Standard Technical Specifications ML19225C0692019-09-19019 September 2019 Issuance of Amendment No. 175 Modify Reactor Trip System and Engineered Safety Features Actuation System Instrumentation Trip Setpoints ML19192A0122019-09-17017 September 2019 Issuance of Amendment No. 174 Adopt Title 10 of the Code of Federal Regulations 50.69, Risk-Informed Categorization and Treatment of Structures, Systems and Components ML19254A6202019-09-17017 September 2019 Duke Energy Fleet (Brunswick, Catawba, Robinson, McGuire, Oconee, and Shearon Harris) - Alternative to Use Encoded Phased Array Ultrasonic Examination in Lieu of Radiography ML19108A1732019-07-18018 July 2019 Issuance of Amendment No. 173 Regarding Emergency Plan Emergency Action Level Scheme Change ML19058A6322019-07-0101 July 2019 Issuance of Amendments to Revise Emergency Action Level Schemes to Incorporate Clarifications Provided by Emergency Preparedness FAQs ML19128A3262019-06-20020 June 2019 Alternative to Depth Sizing Qualification Examination of Welds ML18288A1392019-04-29029 April 2019 Issuance of Amendments 171, 263 Revising Technical Specifications to Support Self Performance of Core Reload Design and Safety Analyses ML19050A2972019-04-18018 April 2019 McGuire Nuclear Station, Units 1 and 2, Shearon Harris Nuclear Power Plant, Unit 1, H B. Robinson Steam Electric Plant, Unit 2 - Issuance of Amendments Regarding Removal ML18347A3852019-03-29029 March 2019 Issuance of Amendment No. 169, Revise Updated Final Safety Analysis Report to Incorporate Tornado Missile Risk Evaluator Methodology Into Licensing Basis ML19018A0252019-02-26026 February 2019 Relief Request I3R-18, Regarding Alternative Repair and Replacement Testing Requirements for the Containment Building Equipment Hatch Sleeve Weld, Inservice Inspection Program for Containment, Third Ten-Year Interval ML18303A0482018-12-0404 December 2018 Extension of Reactor Pressure Vessel Surveillance Report for Capsule Z ML18283B5442018-11-20020 November 2018 Relief from the Requirements of the ASME Code, Section XI, Reactor Vessel Closure Head Penetration Nozzle Repair Technique ML18262A3032018-11-19019 November 2018 Issuance of Amendment No. 168 Regarding License Amendment Request for Rod Control Movable Assemblies Technical Specifications Changes ML18204A2862018-10-22022 October 2018 Issuance of Amendment No. 167 Regarding (CAC MF9996; EPID 2017-LLA-0303) ML18172A1722018-08-15015 August 2018 Issuance of Amendments to Adopt TSTF-545, Revision 3, TS Inservice Testing Program Removal & Clarify SR Usage Rule Application ML18060A4012018-04-10010 April 2018 Redacted Issuance of Amendments Revising TSs for Methodology Reports DPC-NE-3008-P, Revision 0, and DPC-NE-3009-P, Revision 0) (CAC Nos. MF8439 and MF8440; EPID L-2016-LLA-0012) ML17331A0862017-12-26026 December 2017 Alternative to Inservice Inspection Regarding Reactor Pressure Vessel Threads 2023-09-25
[Table view] |
Text
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 January 6, 2016 Mr. Benjamin C. Waldrep Site Vice President Shearon Harris Nuclear Power Plant 5413 Shearon Harris Rd.
M/C HNP01 New Hill, NC 27562-0165
SUBJECT:
SHEARON HARRIS NUCLEAR POWER PLANT, UNIT 1 - SUPPLEMENTAL SAFETY EVALUATION TO RELIEF REQUEST 13R-15 REGARDING RELIEF REQUESTS 13R-09, 13R-11, AND 13R-13 FOR REACTOR VESSEL CLOSURE HEAD PENETRATION NOZZLE REPAIR TECHNIQUE, INSERVICE INSPECTION PROGRAM - THIRD 10-YEAR INTERVAL (CAC NO. MF6053)
Dear Mr. Waldrep:
On May 4, 2015, the U.S. Nuclear Regulatory Commission (NRC) staff verbally authorized the use of Relief Request (RR) 13R-15 for the repair of degraded reactor vessel closure head (RVCH) nozzles 14, 18, and 23 at the Shearon Harris Nuclear Power Plant, Unit 1. The script for the verbal authorization was issued on May 7, 2015 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML15126A542). By letter dated September 18, 2015, the NRC staff issued a safety evaluation (SE) (ADAMS Accession No. ML15203A702) documenting the staffs detailed technical basis for the verbal authorization, concluding that the licensee adequately addressed all of the regulatory requirements set forth in Title 10 of the Code of Federal Regulations (1 O CFR) 50.55a(z)(1 ).
The purpose of the enclosed supplemental SE is to address the calculation error identified during the review of RR 13R-15. The calculation was previously submitted to support RRs 13R-09, 13R-11, and 13R-13. The NRC authorized RRs 13R-09, 13R-11, and 13R-13 by letters dated October 2, 2012; September 13, 2013; and April 4, 2014 (ADAMS Accession Nos. ML12270A258, ML13238A154, and ML14093A075, respectively). In the enclosed supplemental SE, the NRC staff finds that the error identified in calculation 32-9176350-001, for the heat flow calculations, will not exceed the required interpass temperature limits. The NRC staff finds that the error in calculation 32-9176350-001 did not affect the conclusion in the SEs for RRs 13R-09, 13R-11, and 13R-13, and the subject repairs for RVCH nozzles stated above continue to provide an acceptable level of quality and safety in accordance with 10 CFR 50.55a(z)(1).
The RRs 13R-09, 13R-11, 13R-13, and 13R-15 apply to the third 10-year inservice inspection interval, which is scheduled to end on May 1, 2017.
B. Waldrep If you have any questions, please contact the Project Manager, Martha Barillas, at 301-415-2760 or Martha.Barillas@nrc.gov.
Sincerely, Benjamin G. Beasley, Chief Plant Licensing Branch 11-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-400
Enclosure:
Supplemental Safety Evaluation cc w/enclosure: Distribution via Listserv
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SUPPLEMENTAL SAFETY EVALUATION REGARDING RELIEF REQUESTS 13R-09, 13R-11. AND 13R-13 ALTERNATIVE REPAIR OF REACTOR VESSEL CLOSURE HEAD PENETRATION NOZZLES DUKE ENERGY PROGRESS. INC.
SHEARON HARRIS NUCLEAR POWER PLANT, UNIT 1 DOCKET NUMBER 50-400
1.0 INTRODUCTION
By letters dated October 2, 2012; September 13, 2013; and April 11, 2014 (Agencywide Documents Access and Management System (ADAMS) Accession Nos. ML12270A258, ML13238A154, and ML14093A075, respectively), the U.S. Regulatory Commission (NRC) staff authorized Relief Requests (RRs) 13R-09, 13R-11, and 13R-13 for repairs to various degraded reactor vessel closure head (RVCH) penetration nozzles at the Shearon Harris Nuclear Power Plant, Unit 1 (Harris). In RR 13R-09, the NRC authorized repair of nozzles 5, 17, 38, and 63. In RR 13R-11, the NRC authorized repair of nozzle 49; in RR 13R-13, the NRC authorized repair of nozzle 37.
In a letter dated April 29, 2015 (ADAMS Accession No. ML15120A406), the licensee notified the NRC that its evaluation of the impact of a residual flaw in the J-groove weld contained an error as documented in the calculation 32-9176350-001, Shearon Harris Unit 1 CRDM/CET Nozzle As-Left J-groove Weld Analysis. The calculation error affects all nozzles repaired as part of RRs 13R-09, 13R-11, and 13R-13. The licensee previously submitted calculation 32-9176350-001 to support RR 13R-13 (ADAMS Accession No. ML13330A996). The licensee referenced calculation 32-9176350-001 in the original RR 13R-15 submittal dated April 2, 2015 (ADAMS Accession No. ML15092A236). Subsequently, the licensee corrected the original calculation and submitted the revised calculation 32-9176350-002 as documented in Enclosure 3 of the submittal dated April 29, 2015 (ADAMS Accession No. ML15120A406 (non-proprietary version)).
By letter dated September 18, 2015, the NRG issued the safety evaluation (SE) (ADAMS Accession No. ML15203A702) following the verbal authorization granted on May 4, 2015, and documented via script dated May 7, 2015 (ADAMS Accession No. ML15126A542), documenting the staff's technical basis for the verbal authorization of RR 13R-15 for the repair of degraded RVCH nozzles 14, 18, and 23 at Harris.
Enclosure
The purpose of this supplemental SE is to address the calculation error in previously authorized RRs 13R-09, 13R-11, and 13R-13. The NRC staff finds that the error identified in calculation 32-9176350-001, for proposed heat flow calculations, will not exceed the required interpass temperature limits.
The NRC staff finds that the error in calculation 32-9176350-001 did not have a significant impact, because the subject repairs for RVCH nozzles stated above did not significantly reduce the level of quality and safety in accordance with Title 10 of the Code of Federal Regulations (10 CFR) 50.55a(z)(1 ).
2.0 REGULATORY EVALUATION
Pursuant to 10 CFR 50.55a(z)(1 ), the licensee requested authorization of alternatives to various requirements in IWA-3000, IWA-4000, and IWB-3000 of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code),Section XI, and ASME Code Case N-638-1, "Similar and Dissimilar Metal Welding Using Ambient Temperature Machine GTAW [Gas Tungsten Arc Welding] Temper Bead Technique,Section XI, Division 1."
Adherence to Section XI of the ASME Code is mandated by 10 CFR 50.55a(g)(4), which states, in part, that ASME Code Class 1, 2, and 3 components will meet the requirements, except the design and access provisions and the pre-service examination requirements, set forth in the ASME Code,Section XI.
Pursuant to 10 CFR 50.55a(g)(6)(ii), the Commission may require the licensee to follow an augmented inservice inspection (ISi) program for systems and components for which the Commission deems that added assurance of structural reliability is necessary.
As stated in 10 CFR 50.55a(g)(6)(ii)(D), "Reactor vessel head inspections," licensees of pressurized-water reactors (PWRs) are required to augment their ISi of the RVCH with ASME Code Case N-729-1, "Alternative Examination Requirements for PWR Reactor Vessel Upper Heads With Nozzles Having Pressure-Retaining Partial-Penetration Welds,Section XI, Division 1," with conditions.
ASME Code Case N-638-1 provides requirements for automatic or machine GTAW of Class 1 components without the use of preheat or postweld heat treatment.
As stated, in part, in 10 CFR 50.55a(z), alternatives to the requirements of paragraph (g) of 10 CFR 50.55a may be used, when authorized by the NRC, if the licensee demonstrates that (1) the proposed alternative provides 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 above, and subject to the following technical evaluation, the NRC staff finds that regulatory authority exists for the licensee to request, and the Commission to authorize, the alternative requested by the licensee.
3.0 TECHNICAL EVALUATION
The licensee's original calculation 32-9176350-001 contains an error, because the calculation did not consider the impact of the base metal of the reactor vessel head surrounding the repaired RVCH nozzle on the flaw growth calculation. In addition, the revised calculation 32-9176350-002 calculated the RVCH service life considering all the repaired RVCH nozzle configurations that have been performed to date from 2012 to the spring 2015 refueling outage.
The licensee performed the fracture mechanics analysis to determine the acceptability of leaving degraded J-groove weld and butter material in the reactor vessel head following the repair of either an RVCH nozzle or Core Exit Thermocouple nozzle by the inside diameter temper bead weld procedure. The licensee postulated that a flaw in the RVCH would combine with a large stress corrosion crack in the remnant J-groove weld and butter to form a radial corner flaw that would propagate into the low alloy steel RVCH by fatigue crack growth under cyclic loading conditions.
The licensee used the screening criteria of the ASME Code,Section XI, Appendix C to determine the failure mode and appropriate method of analysis (linear elastic fracture mechanics, elastic-plastic fracture mechanics, or limit load) for flaws in ferritic materials of the RVCH, considering the applied stress, temperature, and material toughness. The licensee used the linear elastic fracture mechanics method to derive the stress intensity factor for the final flaw size to satisfy the available fracture toughness of the RVCH. For the more ductile material, the licensee used the elastic-plastic fracture mechanics method to evaluate flaw stability and crack driving force in the RVCH. The licensee also performed a limit load analysis to ensure that the postulated flaw in the RVCH satisfies the primary stress limits of the ASME Code, Section Ill.
The flaw evaluation using the linear elastic fracture mechanics method and elastic plastic fracture mechanics method did not change significantly between the original calculation 32-9176350-001 and revised calculation 32-9176350-002. The major change involved the limit load analysis.
In the revised calculation 32-9176350-002, the licensee's limit load analysis was based on primary stress limits of the ASME Code, Section 111, NB-3332. As shown in Figures C-1 to C-11 of calculation 32-9176350-002, the licensee modeled the details of the region that surrounds the repaired CROM nozzle bore. The approach of calculating available years of service is based on determining the available area of reinforcement as required per the ASME Code, Section Ill, NB-3332.
The nozzle repair results in removal of the structural material in the RVCH. In addition, the remnant J-groove weld is not considered as structural material as it contains flaws. The licensee also discounted additional area caused by postulated crack growth into the carbon steel of the RVCH from the available structural area. As such, the licensee's nozzle model considered the available reinforcement area surrounding the nozzle, the removed bore area, the area partially removed caused by strength ratio, the removed J-groove weld area caused by crack growth, the removed J-groove weld and buttering area, and the reduced new weld area of reinforcement caused by strength ratio.
The licensee's limit load analysis demonstrated that the as-repaired RVCH continues to satisfy the primary stress limits of NB-3000, considering postulated flaws emanating from the original J-groove weld. This is accomplished by comparing the available reinforcement areas in the vicinity of the repaired nozzles with the areas removed from consideration of carrying primary load, in accordance with NB-3332.
The results of the licensee's revised limit load analysis showed that among the nine nozzle repairs (numbers 5, 14, 17, 18, 23, 37, 38, 49, and 63), RVCH nozzle number 14 yields the most limiting service life of 15 years. The licensee noted that as the earliest repairs were performed in May 2012, the RVCH nozzles are acceptable for 12 years of additional operation, starting from April 2015.
The licensee also noted that based on a combination of linear elastic and elastic-plastic fracture mechanics analysis of a postulated remaining flaw in the original Alloy 182 J-groove weld and butter material, a Harris CROM or CET nozzle is considered to be acceptable for 30 years of operation following the nozzle repair.
For RRs 13R-09, 13R-11, 13R-13, and 13R-15, the NRC staff approved the use of the RR for the third 10-year ISi interval, which commenced on May 2, 2007, and will end on May 1, 2017.
In the NRC's evaluation of RR 13R-9, dated October 2012, the NRC staff noted that the RVCH repaired nozzles are acceptable for the design life of the repair, which is 14.8 effective full power years.
The NRC staff recognizes that the linear elastic fracture mechanics and elastic-plastic fracture mechanics method predicted a service life of 30 years for the RVCH following the nozzle repairs. However, based on the licensee's latest limit load analysis in calculation 32-9176350-002, the service life for the RVCH following the nozzle repairs should be limited to 15 years, based on nozzle 14 being the most limiting case.
The NRC staff notes that since the discovery of indications in the RVCH nozzles in 2012, the licensee has performed, and is required to perform, bare metal visual examinations of every RVCH nozzle every subsequent refueling outage at Harris in accordance with 10 CFR 50.55a(g)(6)(ii)(D). This inspection frequency will adequately monitor the RVCH nozzles to ensure that the repaired nozzles are structurally sound. The inspection frequency also provides a means of verifying the flaw growth calculations to ensure that the projected flaw growth in the remnant J-groove weld will not affect the structural integrity of the RVCH.
The NRC staff finds that as indicated by the results of the licensee's revised calculation for the remnant J-groove weld, the error in the licensee's original calculation did not affect the structural integrity of the RVCH. Therefore, the NRC staff determines that the error in the original calculation did not affect the safety function of the RVCH.
4.0 CONCLUSION
Based on the information submitted, the NRC staff concludes that the error in the licensee's calculation 32-9176350-001 did not affect the structural integrity of the repaired RVCH nozzles nor the RVCH itself. Accordingly, the NRC staff concludes that the repaired nozzles remain in compliance with the requirements of the ASME Code, Sections Ill and XI, Code Case N-638-1 and Code Case N-729-1, as conditioned by 10 CFR 50.55a(g)(6)(ii)(D), for which relief was not requested. Therefore, the NRC staff concludes that the licensee's calculation error did not affect the conclusion in the SE for RRs 13R-09, 13R-11, and 13R-13, and the repairs continue to provide an acceptable level of quality and safety in accordance with 10 CFR 50.55a(z)(1 ).
Principal Contributor: John Tsao Date:January 6, 2016
B. Waldrep If you have any questions, please contact the Project Manager, Martha Barillas, at 301-415-2760 or Martha.Barillas@nrc.gov.
Sincerely,
/RAJ Benjamin G. Beasley, Chief Plant Licensing Branch 11-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-400
Enclosure:
Supplemental Safety Evaluation cc w/enclosure: Distribution via Listserv DISTRIBUTION:
PUBLIC LPL2-2 R/F RidsNrrDorlLpl2-2 RidsNrrLABClayton RidsACRS_MailCenter RidsRgn2MailCenter RidsNrrDeEpnb RidsNrrPMShearonHarris JTsao, NRR ADAMS Access1on No.: ML15342A043 *b1v e-ma1*1 OFFICE NRR/DORL/LPLll-2/PM NRR/DORL/LPLll-2/LA NRR/DE/EPNB/BC*
NAME MBarillas BClayton (LRonewicz for) DAiiey DATE 1/05/2016 1/04/2016 10/24/2015 OFFICE NRR/DORL/LPLll-2/BC N RR/DORL/LP LI 1-2/PM NAME BBeasley MBarillas DATE 1/06/2016 1/06/2016 OFFICIAL RECORD COPY