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MONTHYEARML18094A7992018-04-0303 April 2018 Acceptance Review for Beaver Valley - Relief Request from ASME Section XI Code - 5th ISI Interval Project stage: Acceptance Review ML18227A7332018-08-27027 August 2018 Request for Relief from the Requirements of the ASME Code Project stage: Approval 2018-04-03
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Category:Code Relief or Alternative
MONTHYEARML23249A1842023-09-18018 September 2023 Authorization and Safety Evaluation for Alternative Request No. 2-TYP-4-RV-06 ML22082A2532022-03-29029 March 2022 Correction Letter - Issuance of Authorization of Proposed Alternative to Use ASME OM Code Case OMN-27 ML22012A2972022-01-21021 January 2022 Proposed Alternative to Use ASME OM Code Case OMN-27 L-20-256, Request to Use Provision in Later Edition of the ASME Boiler and Pressure Vessel Code, Section XI2020-09-28028 September 2020 Request to Use Provision in Later Edition of the ASME Boiler and Pressure Vessel Code, Section XI ML20223A0142020-08-20020 August 2020 Issuance of Relief Request SSR-1, Revision 0, from the Requirements of the ASME Code (EPID L-2020-LLR-0050 (COVID-19)) ML20206K8652020-08-0707 August 2020 Relief Requests 2-TYP-4-RV-06 and 2-TYP-4-RV-07 from the Requirements of the ASME Code (EPID L-2020-LLR-0053 and EPID L-2020-LLR-0054 (COVID-19)) ML20145A0002020-06-0303 June 2020 Relief from the Requirements of the ASME Code for Requests VRR6 and VRR5 (EPID L-2020-LLR-0049 and EPID L-2020-LLR-0051 (COVID-19)) ML20100N3222020-04-0909 April 2020 Verbal Relief for Penetration Evaluation and Hot Leg Nozzles - Delivered 4/9/2020 at 10:00 Am ML20099B2572020-04-0808 April 2020 Verbal Relief Unit 2 CIV ML20098F3012020-04-0707 April 2020 Verbal Relief for Appendix I Safety Relief Valves ML20095J2192020-04-0404 April 2020 Email Beaver Valley Power Station, Unit 2 - Verbal Relief for MOVs - Delivered 4/4/2020 at 4:00 Pm ML20095J0992020-04-0404 April 2020 Email Beaver Valley Power Station, Unit 2 - Verbal Relief for Snubbers - Delivered 4/4/2020 at 4:00 P.M L-20-117, 10 CFR 50.55a Request Number 2-TYP-4-RV-06, Hardship for Hot Leg Nozzle Inspections2020-04-0303 April 2020 10 CFR 50.55a Request Number 2-TYP-4-RV-06, Hardship for Hot Leg Nozzle Inspections L-20-118, CFR 50.55a Request Number SRR-1, Revision 0, Snubber Testing2020-04-0303 April 2020 CFR 50.55a Request Number SRR-1, Revision 0, Snubber Testing L-20-060, CFR 50.55a Request Number: VRR4, Revision 0, Containment Isolation Valve Test Frequency2020-04-0202 April 2020 CFR 50.55a Request Number: VRR4, Revision 0, Containment Isolation Valve Test Frequency L-20-116, CFR 50.55a Request Number VRR6, Revision 0, Motor-Operated Valve Test Frequency2020-04-0101 April 2020 CFR 50.55a Request Number VRR6, Revision 0, Motor-Operated Valve Test Frequency ML20079F8162020-03-26026 March 2020 Relief Requests 1-TYP-4-C2.21-1 and 1-TYP-4-RA-1 Regarding Weld Examination Coverage for the Fourth Inservice Inspection Interval ML19275E2942019-10-16016 October 2019 Issuance of Relief Request Proposed Alternative to Reactor Vessel Nozzle Weld Examination Frequency Requirements in Lieu of Specific ASME Code Requirements L-19-107, Impractical American Society of Mechanical Engineers Boiler and Pressure Vessel Code Section XI Examination Requirements2019-08-27027 August 2019 Impractical American Society of Mechanical Engineers Boiler and Pressure Vessel Code Section XI Examination Requirements ML19051A1082019-02-20020 February 2019 Request Alternative Examination Frequency for Reactor Vessel Nozzle to Safe-End Welds (Request 2-TYP-4-RV-05, Revision 0) ML18227A7332018-08-27027 August 2018 Request for Relief from the Requirements of the ASME Code ML18004A1222018-01-22022 January 2018 FENOC-Beaver Valley, Davis-Besse, and Perry - Alternative for the Use of ASME Code Case N-513-4 (CAC Nos. MG0120, MG0121, MG0122, and MG0123; EPID L-2017-LLR-0088) L-17-317, Request to Extend Certain Reactor Vessel Inspections from 10 to 20 Years (Request 1-TYP-4-BN-01)2017-11-15015 November 2017 Request to Extend Certain Reactor Vessel Inspections from 10 to 20 Years (Request 1-TYP-4-BN-01) L-17-308, 10 CFR 50.55a Request for Alternate Reactor Vessel Nozzle Flaw Depth Sizing Criteria (Request 2-TYP-4-RVSE-2)2017-10-25025 October 2017 10 CFR 50.55a Request for Alternate Reactor Vessel Nozzle Flaw Depth Sizing Criteria (Request 2-TYP-4-RVSE-2) ML17167A0672017-06-26026 June 2017 Requests for Alternatives and Requests for Relief Fourth 10-Year Inservice Test Program Interval (CAC Nos. MF8333-MF8356). Note: Correction Safety Evaluation See ML17255A526 ML17159A4422017-06-26026 June 2017 Requests for Alternatives and Requests for Relief Fifth 10-Year Inservice Testing Program Interval (CAC Nos. MF8332 Through MF8357). Note: Correction Safety Evaluation See ML17255A508 ML17047A6102017-03-0202 March 2017 Relief from the Requirements of the American Society of Mechanical Engineers Code ML17041A1852017-03-0202 March 2017 Relief from the Requirements of the ASME Code ML17048A0042017-03-0202 March 2017 Relief from the Requirements of the ASME Code ML16328A1252017-01-23023 January 2017 Relief from the Requirements of the American Society of Mechanical Engineers (ASME) Code for Operation and Maintenance of Nuclear Power Plants (OM Code) (TAC No. MF7780 - MF7783) ML16190A1332016-12-27027 December 2016 Relief from the Requirements of the ASME Code ML16319A0572016-12-0101 December 2016 Acceptance of Requested Licensing Action Relief Request for Proposed Alternative to Perform As-Found Set-Pressure Test ML16257A6212016-11-21021 November 2016 Relief Request BV2-PZR-01, Regarding Alternative to Requirements for Components Connected to the Steam Side of the Pressurizer ML16228A4082016-10-21021 October 2016 Correction to Relief Request 2-TYP-3-RV-04, Revision 0, Regarding Repair Activities for Reactor Vessel Head Penetration Nozzles and Associated J-Groove Welds ML16147A3622016-06-17017 June 2016 Relief Request No. 2-TYP-3-RV-04, Revision 0, Regarding Repair Activities for Reactor Vessel Head Penetration Nozzles and Associated J-Groove Welds ML14363A4092015-01-28028 January 2015 Relief Request No. 1-TYP-4-RV-04 Regarding the Examination Requirements of Code Case N-729-1 ML1202702982012-02-0707 February 2012 Relief Request VRR3 Regarding Solenoid Operated Valve Remote Position Verification Frequency ML1131304282011-11-22022 November 2011 Relief Request VRR5 Regarding Turbine Driven Auxiliary Feedwater Valve Test Frequency for the 10-Year Inservice Testing Program Interval ML1126404122011-09-20020 September 2011 Acceptance Review Results for VC Summer Relief Request (ME6879) ML1107705512011-04-26026 April 2011 Relief Request VRR2 Regarding the 10-Year Inservice Testing Program Interval ML1104705572011-02-25025 February 2011 Relief Request Regarding an Alternative Weld Repair Method for Reactor Vessel Head Penetrations J-Groove Welds ML1006807812010-03-12012 March 2010 Third 10-Year ISI Interval Relief Request (ME2608) L-08-362, Beaver, Units 1 and 2, Proposed Alternative to American Society of Mechanical Engineers Boiler and Pressure Vessel Code Section XI, Paragraph IWA-5244 Examination Requirements2008-12-0202 December 2008 Beaver, Units 1 and 2, Proposed Alternative to American Society of Mechanical Engineers Boiler and Pressure Vessel Code Section XI, Paragraph IWA-5244 Examination Requirements L-08-207, Proposed Alternative to American Society of Mechanical Engineers Boiler and Pressure Vessel Code Section XI Inspection Period Extension Requirement2008-09-24024 September 2008 Proposed Alternative to American Society of Mechanical Engineers Boiler and Pressure Vessel Code Section XI Inspection Period Extension Requirement L-08-069, Impractical American Society of Mechanical Engineers Code Section XI Weld Examination Requirements (Request Nos. 1-TYP-3-RA-1, 1-TYP-3-RA-2, 1-TYP-3-RA-3)2008-04-0909 April 2008 Impractical American Society of Mechanical Engineers Code Section XI Weld Examination Requirements (Request Nos. 1-TYP-3-RA-1, 1-TYP-3-RA-2, 1-TYP-3-RA-3) ML0720504882007-09-17017 September 2007 Relief Request No. BV1-PZR-01 Regarding Weld Overlay Repairs on Pressurizer Nozzle Welds ML0705905552007-04-30030 April 2007 Relief, Relief Request No. BV2-PZR-01 Regarding Weld Overlay Repairs on Pressurizer Nozzle Welds L-07-056, Requests Approval of Proposed Alternatives and Relief for Inservice Testing Program Ten-Year Update2007-03-28028 March 2007 Requests Approval of Proposed Alternatives and Relief for Inservice Testing Program Ten-Year Update ML0625801202006-10-0202 October 2006 (BVPS-1 and 2), Inservice Inspection (ISI) Program, Alternative Examination of Reactor Coolant Pipe Welds, Request for Relief No. BV3-RV-2 L-06-042, Proposed Alternative to American Society of Mechanical Engineers Code Section XI Examination Requirements2006-04-0707 April 2006 Proposed Alternative to American Society of Mechanical Engineers Code Section XI Examination Requirements 2023-09-18
[Table view] Category:Letter
MONTHYEARML24025A0922024-01-25025 January 2024 Requalification Program Inspection L-23-267, Submittal of Discharge Monitoring Report (Npdes), Permit No. PA00256152023-12-18018 December 2023 Submittal of Discharge Monitoring Report (Npdes), Permit No. PA0025615 IR 05000334/20230112023-12-0404 December 2023 Age-Related Degradation Inspection Report 05000334/2023011 and 05000412/2023011 L-23-229, Request for Additional Information Regarding the Spring 2023 Generic Letter 95-05 Voltage-Based Alternate Repair Criteria and Steam Generator F-Star Reports2023-11-29029 November 2023 Request for Additional Information Regarding the Spring 2023 Generic Letter 95-05 Voltage-Based Alternate Repair Criteria and Steam Generator F-Star Reports IR 05000334/20233012023-11-27027 November 2023 Initial Operator Licensing Examination Report 05000334/2023301 L-23-247, Discharge Monitoring Report (NPDES) Permit No. PA00256152023-11-17017 November 2023 Discharge Monitoring Report (NPDES) Permit No. PA0025615 IR 05000334/20230032023-11-0606 November 2023 Integrated Inspection Report 05000334/2023003 and 05000412/2023003 L-23-227, Discharge Monitoring Report (NPDES) Permit No. PA0025615 for Third Quarter 20232023-10-20020 October 2023 Discharge Monitoring Report (NPDES) Permit No. PA0025615 for Third Quarter 2023 ML23279A0612023-10-0505 October 2023 Paragon Energy Solutions LLC, Part 21 Final Report Re Potential Defect with Eaton Jd and Hjd Series Molded Case Circuit Breakers (Mccbs) ML23198A3592023-10-0202 October 2023 Issuance of Amendment Nos. 322 and 212 Analytical Methodology to the Core Operating Limits Report for a Full Spectrum Loss of Coolant Accident (EPID L-2022-LLA-0129) - Nonproprietary IR 05000334/20234022023-09-29029 September 2023 Security Baseline Inspection Report 05000334/2023402 and 05000412/2023402 (Cover Letter Only) 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 ML23269A1242023-09-27027 September 2023 Request for Withholding Information from Public Disclosure IR 05000334/20234012023-09-21021 September 2023 Cybersecurity Inspection Report 05000334/2023401 and 05000412/2023401 (Cover Letter Only) ML23263A0192023-09-20020 September 2023 Operator Licensing Examination Approval ML23243A9272023-09-19019 September 2023 Review of the Fall 2022 Steam Generator Tube Inspection Report ML23249A1842023-09-18018 September 2023 Authorization and Safety Evaluation for Alternative Request No. 2-TYP-4-RV-06 L-23-208, Submittal of Discharge Monitoring Report Cnpdes), Permit No. PA00256152023-09-14014 September 2023 Submittal of Discharge Monitoring Report Cnpdes), Permit No. PA0025615 L-23-167, Twenty-Third Refueling Outage Inservice Inspection Summary Report2023-09-13013 September 2023 Twenty-Third Refueling Outage Inservice Inspection Summary Report L-23-205, Supplement to Application for Order Consenting to Transfer of Licenses and Conforming License Amendments2023-09-12012 September 2023 Supplement to Application for Order Consenting to Transfer of Licenses and Conforming License Amendments IR 05000334/20230052023-08-31031 August 2023 Updated Inspection Plan for Beaver Valley Power Station Units 1 and 2 (Report 05000334/2023005 and 05000412/2023005) L-23-172, Quality Assurance Program Manual2023-08-31031 August 2023 Quality Assurance Program Manual ML23129A1722023-08-25025 August 2023 Request for Withholding Information from Public Disclosure for Beaver Valley Power Station, Units 1 and 2; Davis Besse Nuclear Power Station, Unit 1; and Perry Nuclear Power Plant, Unit 1 IR 05000334/20230022023-08-0909 August 2023 Integrated Inspection Report 05000334/2023002 and 05000412/2023002 ML23219A0592023-08-0707 August 2023 Steam Generator Inspection Reports - Spring 2023 Spring Refueling Outrage L-23-188, Energy Harbor Nuclear Corp., Supplement to Application for Order Consenting to Transfer of Licenses and Conforming License Amendments2023-08-0707 August 2023 Energy Harbor Nuclear Corp., Supplement to Application for Order Consenting to Transfer of Licenses and Conforming License Amendments L-23-179, Submittal of Discharge Monitoring Report, (NPDES) Permit No. PA00256152023-07-18018 July 2023 Submittal of Discharge Monitoring Report, (NPDES) Permit No. PA0025615 ML23188A0982023-07-17017 July 2023 Correction to the Safety Evaluation Issued Related to Amendment Nos. 321 and 211 Consolidate Fuel Decay Time Technical Specifications in a New Limiting Condition for Operation Titled Decay Time ML23188A0672023-07-10010 July 2023 Project Manager Assignment L-23-165, Discharge Monitoring Report (NPDES) Permit No. PA00256152023-06-26026 June 2023 Discharge Monitoring Report (NPDES) Permit No. PA0025615 L-23-139, Response to Request for Additional Information Regarding Fall 2022 180-Day Steam Generator Tube Inspection Report2023-06-13013 June 2023 Response to Request for Additional Information Regarding Fall 2022 180-Day Steam Generator Tube Inspection Report ML23157A1072023-06-0707 June 2023 Request for Information and Notification of Conduct of IP 71111.21.N.04, Age-Related Degradation, Reference Inspection Report 05000334/2023011 and 05000412/2023011 ML23143A0272023-05-23023 May 2023 Licensed Operator Positive Fitness-For-Duty Test L-23-055, Submittal of the Updated Final Safety Analysis Report, Revision 342023-05-23023 May 2023 Submittal of the Updated Final Safety Analysis Report, Revision 34 ML23102A1472023-05-22022 May 2023 Issuance of Amendment Nos. 321 and 211 Consolidate Fuel Decay Time Technical Specifications in a New Limiting Condition for Operation Titled Decay Time L-23-065, Annual Financial Report2023-05-22022 May 2023 Annual Financial Report ML23135A7872023-05-19019 May 2023 Summary of Conference Calls Regarding the Spring 2023 Steam Generator Tube Inspections L-23-137, Discharge Monitoring Report (NPDES) Permit No. PA00256152023-05-18018 May 2023 Discharge Monitoring Report (NPDES) Permit No. PA0025615 ML23124A1742023-05-17017 May 2023 Energy Harbor Fleet Vistra License Transfer - Request for Withholding Information from Public Disclosure for Commance Peak Plant, Units 1 & 2, Beaver Valley Station, Units 1 & 2, Davis Besse Station, Unit 1 and Perry Plant, Unit 1 L-23-125, Cycle 24 Core Operating Limits Report2023-05-17017 May 2023 Cycle 24 Core Operating Limits Report ML23124A3882023-05-16016 May 2023 Summary of Regulatory Audit in Support of License Amendment Request to Revise Technical Specification 5.6.3, Core Operating Limits Reports (COLR) and TS 4.2.1 Fuel Assemblies ML23129A0112023-05-16016 May 2023 Notice of Consideration of Approval of Indirect and Direct License Transfer for Comanche Peak Plant, Units 1 & 2, Beaver Valley Station, Units 1 & 2, Davis Besse Station, Unit 1 and Perry Plant, Unit 1 (EPID L-2023-LLM-0000) (Letter) L-23-132, Response to NRC Regulatory Issue Summary 2023-01 Preparation and Scheduling of Operator Licensing Examinations2023-05-10010 May 2023 Response to NRC Regulatory Issue Summary 2023-01 Preparation and Scheduling of Operator Licensing Examinations IR 05000334/20230012023-05-0808 May 2023 Integrated Inspection Report 05000334/2023001 and 05000412/2023001 L-23-129, Response to Request for Additional Information for 10 CFR 50.55a Request 2-TYP-4-RV-06 for Alternative Repair Methods for Reactor Pressure Vessel Head Penetrations2023-05-0505 May 2023 Response to Request for Additional Information for 10 CFR 50.55a Request 2-TYP-4-RV-06 for Alternative Repair Methods for Reactor Pressure Vessel Head Penetrations ML23118A3812023-04-28028 April 2023 10 CFR 50.55a Request 2-TYP-4-RV-06 for Alternative Repair Methods for Reactor Pressure Vessel Head Penetrations L-23-115, Submittal of 2022 Annual Radioactive Effluent Release Report, 2022 Annual Radiological Environmental Operating Report, and 2022 Annual Environmental Operating Report (Non-Radiological2023-04-27027 April 2023 Submittal of 2022 Annual Radioactive Effluent Release Report, 2022 Annual Radiological Environmental Operating Report, and 2022 Annual Environmental Operating Report (Non-Radiological L-23-126, Discharge Monitoring Report (Npdes), Permit No. PA00256152023-04-22022 April 2023 Discharge Monitoring Report (Npdes), Permit No. PA0025615 CP-202300181, ISFSI, Beaver Valley, Units 1 and 2, ISFSI, Davis-Besse, Unit 1, ISFSI, Perry, Unit 1, ISFSI, Corrected Affidavit for Application for Order Consenting to Transfer of Licenses and Conforming License Amendments2023-04-20020 April 2023 ISFSI, Beaver Valley, Units 1 and 2, ISFSI, Davis-Besse, Unit 1, ISFSI, Perry, Unit 1, ISFSI, Corrected Affidavit for Application for Order Consenting to Transfer of Licenses and Conforming License Amendments ML23107A2502023-04-18018 April 2023 Requalification Program Inspection 2024-01-25
[Table view] Category:Safety Evaluation
MONTHYEARML23198A3592023-10-0202 October 2023 Issuance of Amendment Nos. 322 and 212 Analytical Methodology to the Core Operating Limits Report for a Full Spectrum Loss of Coolant Accident (EPID L-2022-LLA-0129) - Nonproprietary 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) ML23249A1842023-09-18018 September 2023 Authorization and Safety Evaluation for Alternative Request No. 2-TYP-4-RV-06 ML23188A0982023-07-17017 July 2023 Correction to the Safety Evaluation Issued Related to Amendment Nos. 321 and 211 Consolidate Fuel Decay Time Technical Specifications in a New Limiting Condition for Operation Titled Decay Time ML23102A1472023-05-22022 May 2023 Issuance of Amendment Nos. 321 and 211 Consolidate Fuel Decay Time Technical Specifications in a New Limiting Condition for Operation Titled Decay Time ML23019A0032023-03-16016 March 2023 Issuance of Amendment Nos. 320 and 210 Adoption of Technical Specifications Task Force (Tstf) Traveler Tstf-501, Revision 1, Relocate Stored Fuel Oil and Lube Oil Volume Values to Licensee Control ML23062A5212023-03-0606 March 2023 Issuance of Amendment No. 319 Revise Technical Specification (TS) 3.5.2, ECCS Operating, Limiting Condition for Operation (LCO) 3.5.2, ML22277A8142022-10-0707 October 2022 Correction of Errors in Safety Evaluation Associated with License Amendment Nos. 315 and 205 Regarding Changes to the Emergency Preparedness Plan ML22210A0102022-09-16016 September 2022 Energy Harbor Fleet- Issuance of Amendments Regarding Adoption of TSTF 554, Revise Reactor Coolant Leakage Requirements ML22222A0862022-09-0101 September 2022 Issuance of Amendment Nos. 317 and 208 Revise Technical Specification 3.3.5, Loss of Power (LOP) Diesel Generator (DG) Start and Bus Separation Instrumentation ML22235A7672022-09-0101 September 2022 Correction of Errors in Safety Evaluation Associated with License Amendment Nos. 315 and 205 Regarding Changes to the Emergency Preparedness Plan ML22202A4642022-06-29029 June 2022 Emergency Plan Safety Evaluation ML22140A2092022-06-28028 June 2022 Issuance of Amendment No. 207 Correct TS 3.1.7 Change Made by TSTF-547 ML22095A2352022-05-10010 May 2022 Issuance of Amendment Nos. 316 and 206 Revise Technical Specification 5.6.3, Core Operating Limits Report (COLR) ML21286A7822022-05-0606 May 2022 Issuance of Amendment Nos. 315 and 205 Regarding Changes to the Emergency Plan ML22077A1342022-05-0202 May 2022 Issuance of Amendment Nos. 314 and 204 Revise Technical Specifications to Adopt TSTF-577, Revision 1, Revised Frequencies for Steam Generator Tube Inspections ML22082A2532022-03-29029 March 2022 Correction Letter - Issuance of Authorization of Proposed Alternative to Use ASME OM Code Case OMN-27 ML22012A2972022-01-21021 January 2022 Proposed Alternative to Use ASME OM Code Case OMN-27 ML21197A0092021-11-0101 November 2021 Issuance of Amendment Nos. 313 and 203 Reactor Coolant System, Pressure and Temperature Limits Report ML21214A2752021-10-15015 October 2021 Issuance of Amendment Nos. 312 and 202 Atmospheric Dump Valves ML21153A1762021-06-30030 June 2021 Issuance of Amendment No. 201 Revision of Technical Specifications Related to Steam Generator Tube Inspection, and Repair Methods ML21075A1132021-04-16016 April 2021 Issuance of Amendment Nos. 311, 200, and 302 to Incorporate the Applicable Standard Technical Specification 5.2.2, Unit Staff, Into the Technical Specifications ML21070A0002021-03-22022 March 2021 Issuance of Amendment 310, Revise Technical Specification 5.5.5.1, Unit 1 SG Program, to Defer Spring 2021 Refueling Outage Steam Generator Inspections to Fall 2022 Refueling Outage ML20346A0222021-03-10010 March 2021 Issuance of Amendment Nos. 309 and 199 to Change Technical Specifications to Implement New Surveillance Methods for the Heat Flux Hot Channel Factor ML20335A0522021-02-18018 February 2021 Issuance of Amendment Nos. 308 and 198 to Modify Certified Fuel Handler Related Technical Specifications for Permanently Defueled Condition ML20345A2362021-01-28028 January 2021 Issuance of Amendment Nos. 307 and 197 to Add Containment Sump Technical Specifications to Address GSI-191 Issues ML20335A0232020-12-28028 December 2020 Issuance of Amendment Nos. 306 and 196 to Remove License Conditions B and C Related to the Irradiated Fuel Management Plans ML20285A2662020-10-21021 October 2020 Correction to Safety Evaluation for Amendment Nos. 305 and 195 Issued September 23, 2020, Modify Primary and Secondary Coolant Activity Technical Specifications ML20279A4402020-10-0808 October 2020 Energy Harbor Fleet-Beaver Valley Power Station; Davis-Besse Nuclear Power Station, and Perry Nuclear Power Plant - Request to Use a Provision of a Later Edition of the ASME Engineers Boiler and Pressure Vessel Code, Section XI ML20213A7312020-09-23023 September 2020 Issuance of Amendment Nos. 305 and 195 to Modify Primary and Secondary Coolant Activity Technical Specifications ML20223A0142020-08-20020 August 2020 Issuance of Relief Request SSR-1, Revision 0, from the Requirements of the ASME Code (EPID L-2020-LLR-0050 (COVID-19)) ML20206K8652020-08-0707 August 2020 Relief Requests 2-TYP-4-RV-06 and 2-TYP-4-RV-07 from the Requirements of the ASME Code (EPID L-2020-LLR-0053 and EPID L-2020-LLR-0054 (COVID-19)) ML20153A0142020-07-16016 July 2020 Relief from the Requirements of the American Society of Mechanical Engineers Code Regarding Request VRR4 (EPID L-2020-LLR-0052 (COVID-19)) ML20145A0002020-06-0303 June 2020 Relief from the Requirements of the ASME Code for Requests VRR6 and VRR5 (EPID L-2020-LLR-0049 and EPID L-2020-LLR-0051 (COVID-19)) ML20080J7892020-04-28028 April 2020 Relief Requests 2 TYP-3-B3.110-1, 2-TYP-3-C2.21-1, 2-TYP-3-C1.30-1, and 2-TYP-3-RA-1 Regarding Weld Examination Coverage for the Third Inservice Inspection Interval ML20079F8162020-03-26026 March 2020 Relief Requests 1-TYP-4-C2.21-1 and 1-TYP-4-RA-1 Regarding Weld Examination Coverage for the Fourth Inservice Inspection Interval ML19336A0282019-12-18018 December 2019 Safety Evaluation Irradiated Fuel Management Plans ML19305B1312019-12-0202 December 2019 Firstenergy Nuclear Operating Company - Enclosure 6, Non-Proprietary Safety Evaluation, Direct and Indirect Transfer of Licenses and Draft Conforming Amendments for Beaver Valley Units 1 and 2, Davis-Besse Unit 1, and Perry Unit 1 ML19275E2942019-10-16016 October 2019 Issuance of Relief Request Proposed Alternative to Reactor Vessel Nozzle Weld Examination Frequency Requirements in Lieu of Specific ASME Code Requirements ML19028A0302019-04-11011 April 2019 FENOC - Beaver Valley Power Station, Unit 1 & 2; Davis-Besse Nuclear Power Station Unit 1; and Perry Nuclear Power Plant Unit 1 - Approval of Certified Fuel Handler Training and Retraining Program ML18348B2062019-02-25025 February 2019 Issuance of Amendment No. 193 Revise Steam Generator Technical Specifications ML18249A1692018-09-0707 September 2018 Seismic Hazard Mitigation Strategies Assessment (CAC Nos. MF7800 and MF7801; EPID L-2016-JLD-0006) ML18227A7332018-08-27027 August 2018 Request for Relief from the Requirements of the ASME Code ML18205A2922018-08-10010 August 2018 Correction of Errors in Safety Evaluation Associated with Relief Request 2-TYP-4-RV-02 ML18179A4672018-07-30030 July 2018 FENOC-Beaver Valley Power Station, Unit No. 1 and 2; Davis-Besse Nuclear Power Station, Unit 1, and Perry Nuclear Power Plant, Unit 1 - Issuance of Amendments Request to Adopt TSTF-529, Clarify Use and Application Rules ML18178A1122018-07-0202 July 2018 Relief Request No. 2-TYP-4-RV-02 Regarding the American Society of Mechanical Engineers Boiler and Pressure Vessel Code Case N-729-4 Examination Requirements ML18065A4032018-04-0505 April 2018 Correction of Errors in Safety Evaluation Associated with License Amendment Nos. 301 and 190 ML18073A1062018-03-28028 March 2018 Safety Evaluation of Proposed Alternatives 1-TYP-4-BA-01 and TYP-4-BN-01 Regarding the Fourth 10- Year Interval of the Inservice Inspection Program ML18075A0962018-03-27027 March 2018 Safety Evaluation of Proposed Alternative to Use ASME Code Cases N-695-1 and N-696-1 in Lieu of Certain Requirements of the ASME Code 2023-09-28
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UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, O.C. 20555-0001 August 27, 2018 Mr. Richard D. Bologna Site Vice President FirstEnergy Nuclear Operating Company Beaver Valley Power Station Mail Stop A-BV-SEB1 P.O. Box 4, Route 168 Shippingport, PA 15077
SUBJECT:
BEAVER VALLEY, UNIT NO. 2 -REQUEST FOR RELIEF FROM THE REQUIREMENTS OF THE ASME CODE (EPID L-2018-LLR-0026)
Dear Mr. Bologna:
By letter dated March 12, 2018 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML 18072A288), FirstEnergy Nuclear Operating Company (FENOC, the licensee) proposed an alternative to the requirements of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code) regarding alternative requirements for repair of reactor vessel head penetrations (VHPs) at Beaver Valley Power Station, Unit No. 2 (Beaver Valley). Specifically, pursuant to Title 10 of the Code of Federal Regulations (10 CFR) 50.55a(z)(1
), the licensee requested to use the proposed alternative on the basis that its requirements for repair of VHPs provides an acceptable level of quality and safety. The U.S. Nuclear Regulatory Commission (NRC) staff has reviewed the subject request and concludes, as set forth in the enclosed safety evaluation, that the licensee has adequately addressed all of the regulatory requirements set forth in 10 CFR 50.55a(z)( 1 ). The NRC staff finds that the proposed alternative provides an acceptable level of quality and safety. Therefore, the NRC staff authorizes the use of the proposed alternative in relief request (RR) 2-TYP-4-RV-04, for the fourth 10-year inservice inspection (ISi) interval of Beaver Valley, Unit 2, which is scheduled to begin on August 29, 2018 and end on August 28, 2028. All other ASME Code,Section XI, requirements for which relief was not specifically requested and authorized herein by the NRC staff remain applicable, including the third-party review by the Authorized Nuclear lnservice Inspector.
R. Bologna If you have any questions, please contact Jennifer Tobin, Project Manager, at 301-415-2328 or via e-mail at Jennifer.Tobin@nrc.gov.
Docket No.: 50-412
Enclosure:
Safety Evaluation cc: Listserv Sincerely, fJ~U~ Jals G. Danna, Chief Plant Licensing Branch I Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELIEF REQUEST N0.2-TYP-4-RV-04, REVISION 0, REGARDING REACTOR VESSEL HEAD PENETRATION REPAIRS FIRSTENERGY NUCLEAR OPERATING COMPANY BEAVER VALLEY POWER STATION, UNIT NO. 2 DOCKET NO. 50-412
1.0 INTRODUCTION
By letter dated March 12, 2018 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML 18072A288), FirstEnergy Nuclear Operating Company (FENOC, the licensee) proposed an alternative to the requirements of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code) regarding alternative requirements for repair of reactor vessel head penetrations (VHPs) at Beaver Valley Power Station, Unit No. 2 (Beaver Valley). Specifically, pursuant to Title 10 of the Code of Federal Regulations (10 CFR) 50.55a(z)(1
), the licensee requested to use the proposed alternative on the basis that its alternative requirements for repair of VHP's provides an acceptable level of quality and safety.
2.0 REGULATORY EVALUATION
The licensee proposes to use alternatives to the requirements of ASME Code,Section XI, IWA-4000, regarding removal of defects from and welded repair of VHPs. Pursuant to 10 CFR 50.55a(g)(4), ASME Code Class 1, 2, and 3, components (including supports) must meet the requirements, except the design and assess provisions and the pre-service examination requirements, set forth in the ASME Code,Section XI, "Rules for In-service Inspection (ISi) of Nuclear Power Plant Components," to the extent practical within the limitations of design, geometry, and materials of construction of the components.
Pursuant to 10 CFR 50.55a(g)(4)(ii), ISi examination of components during successive 120-month inspection intervals must comply with the requirements of the latest edition and addenda of the ASME Code incorporated by reference in paragraph (a) of 10 CFR 50.55a, 12 months before the start of the 120-month inspection interval ( or the optional ASME Code Cases listed in NRC Regulatory Guide (RG) 1.147, Revision 18, when using Section XI) that are incorporated by reference in paragraphs (a)(3)(ii) of 10 CFR 50.55a), subject to the conditions listed in paragraph (b) of 10 CFR 50.55a. Enclosure Pursuant to 10 CFR 50.55a(g)(6)(ii)(D), "Augmented ISi requirements:
Reactor vessel head inspections
-1) Implementation." holders of operating licenses or combined licenses for pressurized-water reactors (PWRs) as of or after August 17, 2017, shall implement the requirements of ASME Code Case N-729-4 "Alternative Examination Requirements for PWR Reactor Vessel Upper Heads With Nozzles Having Pressure-Retaining Partial-Penetration Welds,Section XI, Division 1," instead of ASME Code Case N-729-1, subject to the conditions specified in paragraphs (g)(6)(ii)
(0)(2) through (4) of 10 CFR 50.55a, by the first refueling outage starting after August 17, 2017. Pursuant to 10 CFR 50.55a(z), alternatives to the requirements of paragraph (g) of 10 CFR 50.55a may be used when authorized by the Director, Office of Nuclear Reactor Regulation.
A proposed alternative must be submitted and authorized prior to implementation.
The licensee must demonstrate:
(1) the proposed alternative would provide an acceptable level of quality and safety; or (2) compliance with the specified requirements of this section 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 U.S. Nuclear Regulatory Commission (NRC) staff finds that regulatory authority exists for the licensee to request and the NRC to authorize the alternative requested by the licensee.
3.0 TECHNICAL EVALUATION 3.1 The Licensee's Proposed Alternative ASME Code Components Affected The ASME Code Class 1 VHP nozzles (1 through 65) and their associated partial penetration J-groove attachment welds are affected.
Applicable Code Edition and Addenda The code of record for the fourth 10-year ISi interval for Beaver Valley Unit 2, is the 2013 Edition of the ASME Code,Section XI. Examinations of the VHPs are performed in accordance with 10 CFR 50.55a(g)(6)(ii)(D), which specifies the use of Code Case N-729-4, with conditions.
The licensee stated that the Construction Code for Beaver Valley, Unit 2, is the 1971 Edition through summer 1972 Addenda of the ASME Code, Section Ill. Duration of Relief The licensee submitted relief request (RR) 2-TYP-4-RV-04 for the fourth 10-year ISi interval at Beaver Valley, Unit 2, which is scheduled to begin on August 29, 2018, and end on August 28, 2028. ' Applicable Code Requirements ASME Code,Section XI, 2013 Edition, subparagraph IWA-4000 contains requirements for the removal of defects from and welded repairs performed on ASME Code components.
For the removal or mitigation of defects by welding, ASME Code,Section XI, IWA-4411, requires that repairs and installation of replacement items shall be performed in accordance with the owner's design specification and the original Construction Code of the component or system. The original Construction Code of the reactor vessel is the ASME Code, Section Ill, 1971 Edition through summer 1972 Addenda. The licensee requests relief from the ASME Code, Section Ill, subparagraphs NB-4131, NB-2538, and NB-2539, which pertain to the removal of base material defects prior to repair by welding, and NB-4451, NB-4452, and NB-4453 which pertain to the removal of weld material defects prior to repair by welding. Proposed Alternative and Basis for Use As an alternative to the requirements of ASME Code, Sections Ill and XI, the licensee proposed to repair VHPs using the methodology described in Westinghouse Commercial Atomic Power (WCAP)-15987-P, Revision 2-P-A, "Technical Basis for the Embedded Flaw Process for Repair of Reactor Vessel Head Penetrations," December 2003 (WCAP-15987-2) (ADAMS Accession No. ML040290246), with modifications as described in Section 5 of the licensee's submittal.
Design and implementation of the repairs will be consistent with WCAP-15987-2 and WCAP-16158-P, Revision 1, "Technical Basis for Repair Options for Reactor Vessel Head Penetration Nozzles and Attachment welds: Beaver Valley Unit 2," January 2018. The nonproprietary version of WCAP-16158 is also included in Appendix D of the licensee's March 12, 2018 letter. Basis for Relief The licensee stated that the embedded flaw repair (EFR) technique is considered a permanent repair. The licensee stated that as long as a primary water stress corrosion cracking (PWSCC) flaw remains isolated from the primary coolant environment, it cannot propagate.
Further, the licensee stated that since Alloys 52/52M weldment is considered resistant to PWSCC, a new PWSCC flaw cannot initiate and grow through Alloys 52/52M overlay to reconnect the primary water environment with the embedded flaw. The licensee stated that the residual stresses produced by the embedded flaw technique have been measured and found to be relatively low, indicating that no new flaws will initiate and grow in the area adjacent to the repair weld. Therefore, fatigue-driven crack growth is not a mechanism for further crack growth into Alloys 52/52M overlay after the EFR process is implemented.
According to the licensee, the small residual stresses produced by the embedded flaw will act constantly, and, therefore, will have no impact on the fatigue effects in this region. Since the residual stress would be additive to the maximum and minimum stress, the stress range will not change and the negligible fatigue usage factor for the region will not change. The licensee stated that WCAP-16158, Revision 1, provides the plant-specific analysis performed for Beaver Valley, Unit 2, using the same methodology as WCAP-15987-2.
WCAP-16158, Revision 1, provides the means to evaluate a broad range of postulated repair scenarios to the VPH penetrations and J-groove welds relative to ASME Code requirements for allowable size and service life. 3.2 NRC Staff Evaluation The licensee requested authorization of its alternative RR 2-TYP-4-RV-04 under 10 CFR 50.55a(z)(1
). The licensee requested to use the proposed alternative on the basis that its requirements for repair of VHP's provides an acceptable level of quality and safety. The purpose of the licensee's proposed repair is to address PWSCC, which typically initiates in susceptible materials, such as Alloy 600 material and Alloys 82/182 weld materials, in areas of tensile stress and certain environmental conditions, such as higher temperatures and corrosive environments.
The reactor VHPs and their associated J-groove attachment welds at Beaver Valley meet these conditions to be susceptible to PWSCC. The proposed repair technique isolates the susceptible material using a seal weld of Alloy 52M weld material which is less susceptible to PWSCC. The licensee's basis for the design, implementation, and inspection of the repairs for VHPs is Westinghouse WCAP-15987, Revision 2-A. In a letter dated July 3, 2003, from H. N. Berkow (NRC) to H. A. Sepp (Westinghouse Electric Company), (ADAMS Accession No. ML031840237), the NRC staff provided a safety evaluation (SE), in which the NRC staff found WCAP-15987-2 to be acceptable for referencing in licensing applications as an alternative to Section XI of the ASME Code, with the following conditions:
- 1. Licensees must follow the NRC flaw evaluation guidelines provided in the R. J. Barrett (NRC) letter to A. Marion (Nuclear Energy Institute), "Flaw Evaluation Guidelines," April 11, 2003. (ADAMS Accession No. ML030980322)
- 2. The crack growth rate referenced in WCAP-15987-P, Revision 2, is not applicable to Alloy 600 or Alloy 690 weld material, i.e., Alloys 52, 82, 152, and 182, filler material.
- 3. The nondestructive examination (NOE) requirements listed in the table below must be implemented for examinations of repairs made using the embedded flaw process. Repair Location Flaw Repair Repair NOE ISi NOE of the Repair Orientation Weld Note 2 VHP Nozzle ID Axial Seal UT (ultrasonic testing) UT or Surface (inside diameter) and Surface VHP Nozzle ID Circumferential Note 1 Note 1 Note 1 VHP Nozzle OD Axial or Note 1 Note 1 Note 1 (outside diameter)
Circumferential above J-groove weld VHP Nozzle OD Axial or UT or Surface UT or Surface below J-groove Circumferential Seal weld J-groove weld Axial Seal UT and Surface, Note 3 UT and Surface, Note 3 J-groove weld Circumferential Seal UT and Surface, Note 3 UT and Surface, Note 3 Notes: 1. Repairs must be reviewed and approved separately by the NRC. 2. Inspection consistent with the NRC Order EA-03-009
[Order Establishing Interim Inspection Requirements for Reactor Pressure Vessel Heads at Pressurized Water Reactors]
dated February 11, 2003, and any subsequent changes (ADAMS Accession No. ML030380470).
- 3. Inspect with personnel and procedures qualified with UT performance-based criteria.
Examine the accessible portion of the repaired region. The UT coverage plus surface coverage must equal 100 percent. The licensee stated that their proposed alternative will use the methodology of the NRC approved WCAP-15987-2 as described by Section 5.1 of the submittal, with some modification.
NRC staff reviewed Section 5.1 to ensure the licensee's proposed actions would meet the requirements of WCAP-15987-2, and any modifications would be acceptable under 10 CFR 50.55a(z)( 1 ). As part of this review the NRC staff identified the following technical changes between the requirements of WCAP-15987-2, and Section 5.1, of the licensee's submittal:
A. The Alloy 600 tube material with a flaw will be repaired with two Alloy 52 isolation weld layers rather than the three layers required in WCAP-15987-2-A.
B. A circumferential flaw on the nozzle or tube inside diameter can be repaired using the seal weld technique without additional submission of the repair method for approval by the NRC. C. Prior to the application of the Alloy 52 or 52M seal weld repair on the reactor pressure vessel clad surface, the stainless steel (SS) head cladding will have three beads of 309L SS installed 360 degrees around the interface of the clad and the J-groove weld metal as a buffer layer. The J-groove weld will be covered with three layers of Alloys 52/52M deposited 360 degrees around the nozzle over and extend to the SS buffer layer. D. In lieu of Notes 2 and 3 of the NRC acceptance for WCAP-15987-P, Revision 2-A, the NOE of the repair will be performed in accordance with ASME Code Case N-729-4, as conditioned by 10 CFR 50.55a(g)(6)(ii)(D).
E. Surface examination of the embedded flaw repair (EFR), for repaired J-groove welds, shall be performed to ensure the repair satisfies ASME Section Ill, NB-5350 acceptance standards.
The frequency of examination shall be as follows: a. Perform surface examination during the first and second refueling outage after installation or repair of the EFR. b. When the examination results in a. above verify acceptable results then reinspection of the EFR will be continued at a frequency of every other refueling outage. If these examinations identify unacceptable results that require flaw removal, flaw reduction to acceptable dimensions or welded repair the requirements of a. above shall be applied during the next refueling outage. The inspection frequency described in a. and b. above is in lieu of performing reinspections every refueling outage. The NRC staff verified that the changes in methodology identified in RR 2-TYP-4-RV-04 from the previously approved WCAP-15987-2 would still meet the methodology approved by the NRC for effective EFR and provide an acceptable level of quality and safety. The NRC staff reviewed the licensee's proposal in paragraph A (above) to allow a reduction in the maximum three layers of the seal weld over the Alloy 600 nozzle material to only two layers. The licensee's basis is that the flaw would be isolated from the primary coolant environment necessary for continued PWSCC growth with less stress being introduced in the base metal with the proposed repair. The NRC staff finds that operational experience has shown that two layers of Alloy 52 material have been sufficient to address dilution layer effects of the high chromium content of the Alloy 52 material, which is the principle reason for the material's resistance to PWSCC. In addition, the concern of increased residual stresses in the Alloy 600 material that might still be exposed to primary coolant is a cause for concern for future flaw initiation.
Therefore, since the repair can be effective using two weld layers, a smaller seal weld that generates less weld residual stresses in the base metal would be more preferable than using three weld layers. Therefore, the NRC staff finds this change to be acceptable.
The NRC staff reviewed the licensee's proposal in paragraph B (above) which alleviates the licensee from the need to submit a repair plan for each circumferential flaw identified that initiates from the inside diameter of the tube or nozzle surface. The licensee detailed a generic repair plan in accordance with Section 5.2 of the licensee's submittal.
The licensee's proposed alternative repair would be to partially excavate the flaw to reduce it to an acceptable size, examine it by UT or surface examination, inlay with Alloys 52 or 52M, and examine by UT and surface examination.
This is in compliance with embedded flaw technique.
Further, operational experience has shown this repair technique is effective in arresting growth of PWSCC flaws. Hence, the NRC staff finds it is an effective generic repair plan to address circumferential flaws that initiate on the inside tube or nozzle surface. Therefore, the NRC staff finds this change to be acceptable.
The NRC staff reviewed the licensee's proposal in paragraph C (above) that the head cladding will have three beads of SS 309L buffer layer installed 360 degrees around the interface of the clad and the J-groove weld metal. The NRC staff notes that it is common practice to apply a buffer layer of SS weld metal over existing SS base material or existing SS weld metal before applying Alloy 52. Existing SS material may potentially contain sufficient levels of sulphur, phosphorous and silicon that can contribute to weld solidification cracking when diluted into Alloy 52 welds. Applying a buffer layer of SS weld metal that contains controlled levels of these elements, which can be detrimental to Alloy 52 welds, decreases the potential for weld cracking.
The NRC staff finds the proposed alternative to be appropriate as the buffer layer of SS 309L weld metal will improve weld quality by decreasing the potential for welding defects at the SS cladding/Alloy 52 interface.
The NRC staff reviewed the licensee's proposal in paragraph D (above) for alternatives for NOE examination requirements of the seal weld and future ISi requirements.
During the time period in which WCAP-15987-2 was approved by the NRC staff, the regulatory requirements for upper head inspection were dictated under NRC Order EA-03-009.
In September 2008, by rule, the NRC established 10 CFR 50.55a(g)(6)(ii)(D) which defined the regulatory requirements for upper head inspections as using ASME Code Case N-729-1, with conditions, and rescinded NRC Order EA-03-009.
In July 2017, by rule, the NRC modified 10 CFR 50.55a(g)(6)(ii)(D) to replace ASME Code Case N-729-1 with ASME Code Case N-729-4. The NRC staff finds that the licensee's proposed alternative inspections for the upper head penetration nozzles, under the current regulatory guidelines of ASME Code Case N-729-4, satisfy the previous NRC conditions on the NDE required for implementation of an EFR under WCAP-15987-2.
Therefore, the NRC staff finds these changes to be acceptable.
The NRC staff reviewed the licensee's proposal in paragraph E (above) regarding alternatives for ISi of repaired J-groove welds. The licensee proposes to perform penetrant testing examinations of repaired J-groove welds for the first two outages after the repairs and then every other outage in lieu of every outage following J-groove weld repairs. The alternative in paragraph E and the licensee's supporting information are identical to the the previous NRC approved alternative for Beaver Valley, as described in the NRC staff SE of alternative RR 2-TYP-3-RV-04, Revision 1, dated March 2, 2017 (ADAMS Accession No. ML 17041A185).
The NRC staff is unaware of any plant operating conditions or industry operating experience that would invalidate the staff's previous conclusion.
Therefore, the NRC staff finds the alternative described in paragraph E acceptable.
In order to support the use of WCAP-15987-2 with a plant-specific technical basis for the use of the EFR, the licensee submitted WCAP-16158 Revision 1, "Technical Basis for Repair Options for Reactor Vessel Head Penetration Nozzles and Attachment Welds: Beaver Valley Unit 2," January 2018, to support its current fourth 10-year ISi interval request. WCAP-16158, Revision 0, Technical Basis for Repair Options for Reactor Vessel Head Penetration Nozzles and Attachment Welds: Beaver Valley Unit 2," November 2003, was referenced by the licensee in support of its VHP nozzle repair alternative request, 2-TYP-3-RV-01, for the third ISi interval which ends August 28, 2018. For the most limiting condition, WCAP-16158, Revision O supported any remaining ligaments of the flaws identified by the licensee in VHP nozzle J-groove weld material to be safely encapsulated for 5 years of operation.
WCAP-16158, Revision 1, provides a basis for any remaining ligaments of the flaws identified by the licensee in the VHP nozzle J-groove weld material to be safely encapsulated for 22 years of operation.
The increase in repair life is based on the use of more realistic assumptions such as use of the design pressure of 2500 pounds per square inch (psi) in lieu of the primary hydro test condition of 3125 (psi) and a maximum embedded flaw aspect ratio of 2 instead of 6. The NRC staff notes that the vessel received a hydrostatic test at the completion of fabrication and will not have the primary hydro test condition again. An aspect ratio of 2 is more realistic than 6 because an aspect ratio of 2 bounds the weld dimensions of all of the VHP nozzle attachment welds. The NRC staff finds WCAP-16401, Revision 1, provides a basis for any remaining ligaments of the flaws identified by the licensee in VHP nozzle J-groove weld material to be safely encapsulated for 22 years of operation.
In accordance with the previous NRC conditions imposed on the use of WCAP-15987-2, and the plant-specific technical basis for the EFR, the NRC staff confirms that the licensee has followed the NRC flaw evaluation guidelines and will implement the appropriate NDE for the repairs to VHP nozzles and their associated J-groove welds at Beaver Valley, Unit 2. As previously shown in the July 3, 2003, NRC SE, the EFR process is considered to be an alternative to ASME Code requirements that provides an acceptable level of quality and safety, as required by 10 CFR 50.55a(z)(1
).
4.0 CONCLUSION
Based on the above evaluation, the NRC staff finds that the proposed alternative provides an acceptable 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)(1
). Therefore, the NRC staff authorizes the use of the proposed alternative in RR 2-TYP-4-RV-04, Revision 0, for the fourth 10-year ISi interval of Beaver Valley, Unit 2, which is scheduled to begin on August 29, 2018 and end on August 28, 2028. All other ASME Code,Section XI, requirements for which relief was not specifically requested and authorized herein by the NRC staff remain applicable, including the third-party review by the Authorized Nuclear lnservice Inspector.
Principal Contributor:
Robert Davis Date of issuance:
August 27, 2018 R. Bologna
SUBJECT:
BEAVER VALLEY, UNIT NO. 2 -REQUEST FOR RELIEF FROM THE REQUIREMENTS OF THE ASME CODE (EPID: L-2018-LLR-0026)
DATED AUGUST 27, 2018 DISTRIBUTION:
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