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EPID:L-2020-LLR-0050, 10 CFR 50.55a Request Number 2-TYP-4-RV-06, Hardship for Hot Leg Nozzle Inspections (Approved, Closed) |
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MONTHYEARML20093G6512020-04-0202 April 2020 Email Beaver Valley Unit 2 - Request for Additional Information - Snubber RR Project stage: RAI 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 Project stage: Request L-20-125, Response to Request for Additional Information Regarding 10 CFR 50.55a Request Number SRR-1, Rev. 0, Snubber Testing2020-04-0303 April 2020 Response to Request for Additional Information Regarding 10 CFR 50.55a Request Number SRR-1, Rev. 0, Snubber Testing Project stage: Response to RAI 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 Project stage: Request 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 Project stage: Other 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)) Project stage: Approval 2020-04-03
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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 type:L
MONTHYEARL-23-267, Submittal of Discharge Monitoring Report (Npdes), Permit No. PA00256152023-12-18018 December 2023 Submittal of Discharge Monitoring Report (Npdes), Permit No. PA0025615 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 L-23-247, Discharge Monitoring Report (NPDES) Permit No. PA00256152023-11-17017 November 2023 Discharge Monitoring Report (NPDES) Permit No. PA0025615 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 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 L-23-172, Quality Assurance Program Manual2023-08-31031 August 2023 Quality Assurance Program Manual 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 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 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 L-23-065, Annual Financial Report2023-05-22022 May 2023 Annual Financial Report L-23-137, Discharge Monitoring Report (NPDES) Permit No. PA00256152023-05-18018 May 2023 Discharge Monitoring Report (NPDES) Permit No. PA0025615 L-23-125, Cycle 24 Core Operating Limits Report2023-05-17017 May 2023 Cycle 24 Core Operating Limits Report 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 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 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 L-23-053, 2022 Annual 10 CFR 50.46 Report of Changes to or Errors in Emergency Core Cooling System Evaluation Models2023-04-14014 April 2023 2022 Annual 10 CFR 50.46 Report of Changes to or Errors in Emergency Core Cooling System Evaluation Models L-23-061, Submittal of the Decommissioning Funding Status Reports2023-03-31031 March 2023 Submittal of the Decommissioning Funding Status Reports L-23-058, 180-Day Steam Generator Tube Inspection Report2023-03-27027 March 2023 180-Day Steam Generator Tube Inspection Report L-23-066, Annual Notification of Property Insurance Coverage2023-03-21021 March 2023 Annual Notification of Property Insurance Coverage L-23-036, Report of Facility Changes, Tests and Experiments2023-03-13013 March 2023 Report of Facility Changes, Tests and Experiments L-23-086, Response to Request for Additional Information for Emergency License Amendment Request for Technical Specification 3.5.2 Regarding One-Time Action for Valve Leak Repair2023-03-0404 March 2023 Response to Request for Additional Information for Emergency License Amendment Request for Technical Specification 3.5.2 Regarding One-Time Action for Valve Leak Repair L-23-087, Supplement to Emergency License Amendment Request for Technical Specification 3.5.2 Regarding One-Time Action for Valve Leak Repair (L-2023-LLA-0027)2023-03-0404 March 2023 Supplement to Emergency License Amendment Request for Technical Specification 3.5.2 Regarding One-Time Action for Valve Leak Repair (L-2023-LLA-0027) L-23-073, Emergency License Amendment Request for Technical Specification 3.5.2 Regarding One-Time Action for Valve Leak Repair2023-03-0101 March 2023 Emergency License Amendment Request for Technical Specification 3.5.2 Regarding One-Time Action for Valve Leak Repair L-23-016, Twenty-Eighth Refueling Outage Inservice Inspection Summary Report2023-02-21021 February 2023 Twenty-Eighth Refueling Outage Inservice Inspection Summary Report L-23-064, Discharge Monitoring Report, (NPDES) Permit No. PA00256152023-02-21021 February 2023 Discharge Monitoring Report, (NPDES) Permit No. PA0025615 L-23-057, Energy Harbor Nuclear Corp Retrospective Premium Guarantee2023-02-20020 February 2023 Energy Harbor Nuclear Corp Retrospective Premium Guarantee L-22-193, Request for Exemption from Specific Provisions in 10 CFR 50 Appendix H2023-02-14014 February 2023 Request for Exemption from Specific Provisions in 10 CFR 50 Appendix H L-22-286, Supplement to Request for an Amendment to Consolidate Fuel Decay Time Technical Specifications in a New Limiting Condition for Operation Titled Decay Time2023-02-14014 February 2023 Supplement to Request for an Amendment to Consolidate Fuel Decay Time Technical Specifications in a New Limiting Condition for Operation Titled Decay Time L-23-032, Discharge Monitoring Report (NPDES) Permit No. PA0025615 for Second Half of 20222023-01-23023 January 2023 Discharge Monitoring Report (NPDES) Permit No. PA0025615 for Second Half of 2022 L-22-281, Discharge Monitoring Report (NPDES) Permit No. PA00256152022-12-16016 December 2022 Discharge Monitoring Report (NPDES) Permit No. PA0025615 L-22-246, Response to Request for Additional Information Regarding a Request to Consolidate Fuel Decay Time Technical Specifications to New LCO2022-12-0707 December 2022 Response to Request for Additional Information Regarding a Request to Consolidate Fuel Decay Time Technical Specifications to New LCO L-22-217, Submittal of Discharge Monitoring Report (NPDES) Permit No. PA00256152022-11-21021 November 2022 Submittal of Discharge Monitoring Report (NPDES) Permit No. PA0025615 L-22-226, Emergency Preparedness Plan2022-11-0404 November 2022 Emergency Preparedness Plan L-22-222, Cycle 29-1 Core Operating Limits Report2022-10-31031 October 2022 Cycle 29-1 Core Operating Limits Report L-22-228, Independent Spent Fuel Storage Installation Changes, Tests, and Experiments2022-10-26026 October 2022 Independent Spent Fuel Storage Installation Changes, Tests, and Experiments L-22-200, Response to Request for Additional Information Regarding a 180-Day Steam Generator Tube Inspection Report Fall 2021 Refueling Outage2022-10-21021 October 2022 Response to Request for Additional Information Regarding a 180-Day Steam Generator Tube Inspection Report Fall 2021 Refueling Outage L-22-232, Withdrawal of a License Amendment Request Associated with Fire Protection Program Changes2022-10-21021 October 2022 Withdrawal of a License Amendment Request Associated with Fire Protection Program Changes L-22-238, Submittal of Discharge Monitoring Report (NPDES) Permit No. PA00256152022-10-20020 October 2022 Submittal of Discharge Monitoring Report (NPDES) Permit No. PA0025615 L-22-227, Revised August 2022 Outfall 003 & 004 NPDES Discharge Monitoring Report (Dmr), Revised Daily Effluent Monitoring Report Forms for Outfalls 003 & 004 and Corrective Action Letter from Eurofins2022-10-0303 October 2022 Revised August 2022 Outfall 003 & 004 NPDES Discharge Monitoring Report (Dmr), Revised Daily Effluent Monitoring Report Forms for Outfalls 003 & 004 and Corrective Action Letter from Eurofins L-22-219, Submittal of Discharge Monitoring Report (NPDES) Permit No. PA00256152022-09-26026 September 2022 Submittal of Discharge Monitoring Report (NPDES) Permit No. PA0025615 L-22-204, Submittal of Evacuation Time Estimates2022-09-0707 September 2022 Submittal of Evacuation Time Estimates L-22-137, Request for Fire Protection Program Changes2022-09-0606 September 2022 Request for Fire Protection Program Changes L-21-238, License Amendment Request for Addition of Analytical Methodology to the Core Operating Limits Report for a Full Spectrum Loss of Coolant Accident2022-08-31031 August 2022 License Amendment Request for Addition of Analytical Methodology to the Core Operating Limits Report for a Full Spectrum Loss of Coolant Accident L-22-188, Response to Request for Additional Information Regarding Steam Generator Inspection Report- Fall 2021 Refueling Outage2022-08-22022 August 2022 Response to Request for Additional Information Regarding Steam Generator Inspection Report- Fall 2021 Refueling Outage L-22-191, Spent Fuel Storage Cask Registration2022-08-17017 August 2022 Spent Fuel Storage Cask Registration 2023-09-14
[Table view] |
Text
Energy Harbor Nuclear Corp.
Beaver Valley Power Station P. O. Box 4 Shippingport, PA 15077 Rod L. Penfield 724-682-5234 Site Vice President, Beaver Valley Nuclear April 3, 2020 L-20-117 10 CFR 50.55a ATTN: Document Control Desk U.S. Nuclear Regulatory Commission Washington, DC 20555-0001
SUBJECT:
Beaver Valley Power Station, Unit No. 2 Docket No. 50-412, License No. NPF-73 10 CFR 50.55a Request Number: 2-TYP-4-RV-06, Hardship for Hot Leg Nozzle Inspections In accordance with 10 CFR 50.55a(z)(2), Energy Harbor Nuclear Corp. hereby requests Nuclear Regulatory Commission (NRC) approval of request 2-TYP-4-RV-06 that proposes to eliminate the examination of the Beaver Valley Power Station, Unit No. 2 (BVPS-2) reactor vessel hot leg nozzle-to-safe end dissimilar metal welds 2RCS-REV21-N-24, 2RCS-REV21-N-26, and 2RCS-REV21-N-28.
As a result of the hardship produced by the recent pandemic and the resulting national state of emergency, Energy Harbor Nuclear Corp. is requesting expedited approval of 2-TYP-4-RV-06. The proposed alternative will eliminate performance of the examination during 2R21 and resume the normal outage examination frequency at the next opportunity, currently expected to be the next refueling outage (2R22) set to begin on October 10, 2021.
To support the startup and critical generation of BVPS-2 from its scheduled refuel outage, Energy Harbor Nuclear Corp. requests approval of the proposed alternative by April 12, 2020.
The enclosed request identifies the affected components, applicable code requirements, and a description and basis for the proposed alternative.
Beaver Valley Power Station, Unit No. 2 L-20-117 Page 2 There are no regulatory commitments contained in this submittal. If there are any questions or additional information is required, please contact Mr. Phil H. Lashley, Acting Manager - Nuclear Licensing and Regulatory Affairs, at (330) 315-6808.
Sincerely, Rod L. Penfield
Enclosure:
Beaver Valley Power Station, Unit No. 2, 10 CFR 50.55a Request Number: 2-TYP-4-RV-06 cc: NRC Region I Administrator NRC Resident Inspector NRC Project Manager Director BRP/DEP Site BRP/DEP Representative
Beaver Valley Power Station, Unit No. 2 10 CFR 50.55a Request Number: 2-TYP-4-RV-06 Proposed Alternative in Accordance with 10 CFR 50.55a(z)(2)
Page 1 of 6
--Hardship or Unusual Difficulty without Compensating Increase in Level of Quality or Safety--
- 1. ASME Code Component(s) Affected Component Numbers: Beaver Valley Power Station, Unit No. 2 (BVPS-2) reactor vessel (RV) hot leg nozzle-to-safe end dissimilar metal (DM) welds (2RCS-REV21-N-24, 2RCS-REV21-N-26, and 2RCS-REV21-N-28)
Code Class: Class 1 Examination Category: Class 1 PWR Pressure Retaining Dissimilar Metal Piping and Vessel Nozzle Butt Welds Containing Alloy 82/182 (ASME Code Case N-770-2, Table 1) and Class 1 PWR Components Containing Alloy 600/82/182 (ASME Code Case N-722-1, Table 1)
Item Number: A-2 in Code Case N-770-2 and B15.90 in Code Case N-722-1
==
Description:==
Alternative Examination Requirements and Acceptance Standards for Class 1 PWR Piping and Vessel Nozzle Butt Welds (Code Case N-770-2) and Additional Examinations for PWR Pressure Retaining Welds in Class 1 Components Fabricated with Alloy 600/82/182 Materials (Code Case N-722-1)
2. Applicable Code Edition and Addenda
American Society of Mechanical Engineers (ASME), Boiler Pressure Vessel (BPV)
Code, Division 1,Section XI, 2013 Edition.
ASME BPV Code Case N-770, Revision 2, Alternative Examination Requirements and Acceptance Standards for Class 1 PWR Piping and Vessel Nozzle Butt Welds Fabricated With UNS N06082 or UNS W86182 Weld Filler Metal With or Without Application of Listed Mitigation Activities.
Beaver Valley Power Station, Unit No. 2 10 CFR 50.55a Request Number: 2-TYP-4-RV-06, Revision 0 Page 2 of 6 ASME BPV Code Case N-722, Revision 1, Additional Examinations of PWR Pressure Retaining Welds in Class 1 Components Fabricated with Alloy 600/82/182 Materials.
3. Applicable Code Requirement
10 CFR 50.55a(g)(6)(ii)(F) states the following: Holders of operating licenses or combined licenses for pressurized-water reactors as of or after August 17, 2017, shall implement the requirements of ASME BPV Code Case N-770-2 instead of ASME BPV Code Case N-770-1, subject to the conditions specified in paragraphs (g)(6)(ii)(F)(2) through (13) of this section ASME Code Case N-770-2 states in Table 1, Examination Categories, for item A-2 Unmitigated butt weld at Hot Leg operating Temperature 625°F, that the weld surface must be visually examined each refueling outage.
10 CFR 50.55a(g)(6)(ii)(E) states the following: All licensees of pressurized water reactors must augment their inservice inspection program by implementing ASME Code Case N-722-1, subject to the conditions specified in paragraphs (g)(6)(ii)(E)(2) through (4) of this section.
ASME Code Case N-722-1 states in Table 1, Examination Categories, for item B15.90 Hot leg nozzle-to-pipe connections, that the weld surface must be visually examined each refueling outage.
4. Reason for Request
Beaver Valley Power Station Unit 2 (BVPS-2) is scheduled to start its 21st refueling outage (2R21) on April 12, 2020. The hot leg nozzle visual examination is required to be performed in 2R21, as specified by ASME Code Case N-770-2 and ASME Code Case N-722-1. For the reasons specified below, performance of the hot leg nozzle visual examinations creates a hardship due to expected challenges with obtaining and maintaining staffing levels sufficient to perform the examinations during 2R21.
Elimination of these examinations could reduce the risk of exposure for critical contract and direct hire personnel to the COVID-19 virus.
On March 13, 2020, the President of the United States declared a national emergency due to the spread and infectious nature of the Coronavirus-2019 (COVID-19) virus and resulting pandemic. The most recent guidance from the Centers for Disease Control and Prevention (CDC) includes recommendations for social distancing by maintaining approximately six feet from other personnel to limit the spread of the virus. On March 28, 2020, the Governor of Pennsylvania issued a Stay at Home order for Beaver County and the surrounding counties of Allegheny and Butler. Furthermore, on March 28,
Beaver Valley Power Station, Unit No. 2 10 CFR 50.55a Request Number: 2-TYP-4-RV-06, Revision 0 Page 3 of 6 2020, the Department of Homeland Security identified workers in the nuclear energy sector as essential critical infrastructure workers.
To prevent the spread of COVID-19 at Beaver Valley Power Station (BVPS), and to protect the health and safety of plant personnel while maintaining responsibilities to support critical infrastructure, Energy Harbor Nuclear Corp. intends to reduce the amount of personnel on-site, which will pose a hardship for completing the currently planned 2R21 refueling outage work scope. Energy Harbor Nuclear Corp. is also contingency planning in case some of its workforce becomes unavailable due to the COVID-19 outbreak. With the current work scope and potential loss of personnel, the company may not be able to complete the refueling outage in a timely manner, which could negatively impact critical infrastructure that is needed during this time.
This request is submitted due to the expected hardship of obtaining and maintaining onsite staff sufficient to prepare, perform, and recover from this examination. At BVPS-2, this exam requires construction trades to open hatches in the floor of the refueling cavity, install temporary lighting, remove neutron shield material, and remove insulation.
Additional contract and onsite staff are required to perform radiological surveys and the weld examinations. Because of the rapid spread and infection rates of the virus, BVPS anticipates challenges to maintain staff levels throughout the outage and is requesting relief where appropriate to reduce necessary staff and ensure being able to return the unit to power production in a reasonable time to support the power needs of the surrounding area.
5. Proposed Alternative and Basis for Use
BVPS-2 is requesting relief to not perform the reactor vessel hot leg nozzle visual examinations during the 2R21 refueling outage. The visual examinations of the reactor vessel hot leg nozzles will be performed during the next refueling outage (2R22) and subsequent refueling outages in accordance with the requirements of ASME Code Cases N-770-2 and N-722-1.
The proposed alternative is based on past BVPS-2 inspection results, a review of industry operating experience related to hot leg nozzle primary water stress corrosion cracking (PWSCC) indications, compensatory actions that would detect leakage if it were to occur, and the chemical mitigation benefits that result from zinc addition to the RCS.
The primary degradation mechanism addressed by the examinations of ASME Code Cases N-770-2 and N-722-1 is PWSCC. This degradation mechanism occurs when a susceptible material is exposed to a primary water environment, elevated stress levels, and high operating temperatures.
Beaver Valley Power Station, Unit No. 2 10 CFR 50.55a Request Number: 2-TYP-4-RV-06, Revision 0 Page 4 of 6 Hot leg nozzle examinations were performed at BVPS-2 during the 2R14 refueling outage (fall 2009), 2R15 refueling outage (spring 2011), 2R16 refueling outage (fall 2012), 2R18 refueling outage (fall 2015), and 2R19 refueling outage (spring 2017). No evidence of pressure boundary leakage or corrosion of ferritic steel components was identified during any of the visual and ultrasonic examinations.
An assessment of recent industry operating experience (OE) revealed that with respect to reactor vessel hot leg nozzle weld examinations, PWSCC flaws are generally discovered using ultrasonic examination techniques. During the last BVPS-2 refueling outage (2R20 in fall 2018), full coverage was obtained with both volumetric (ultrasonic) and eddy current examinations of the reactor vessel hot leg nozzle welds. The examinations were performed in accordance with the requirements of ASME Code Case N-770-2, with no flaws identified. The industry hot leg nozzle PWSCC indications that have been discovered to date were found prior to the issuance of ASME Code Case N-770-2. Since the ultrasonic examinations at BVPS-2 are performed at the frequency required by ASME code case N-770-2, the relevant ultrasonic examination OE has been adequately addressed to ensure that any PWSCC flaws in the BVPS-2 reactor vessel hot leg nozzles would be discovered and mitigated before they would become a safety concern.
A review of recent industry OE also revealed that in one case a visual examination technique did identify leakage in a reactor vessel nozzle weld. However, this was the first flaw identified in a reactor vessel hot leg nozzle weld and it was identified prior to the time that more routine examinations were required. Additionally, in the one case where a visual examination of a reactor vessel hot leg nozzle identified a through-wall flaw, there was significant weld repairs performed in the nozzle weld during fabrication that would have likely increased the PWSCC susceptibility of the hot leg nozzle weld. A review of immediately available records for BVPS-2 did not identify any known weld repairs within the BVPS-2 reactor vessel hot leg nozzle welds.
Zinc addition to the primary reactor coolant was implemented at BVPS-2 in October of 2010, and the zinc deposits have been building since that time. Zinc addition decreases the PWSCC susceptibility of alloy 600/82/182 components in the reactor coolant system, including the reactor vessel hot leg nozzles. The progress of the zinc buildup on the primary system surfaces is measured in terms of ppb-months, which is the product of the concentration of zinc in the chemical mitigation, and the time over which it has been applied. In October 2017, the plant reached 300 ppb-months, at which time significant chemical mitigation against PWSCC was achieved. This decreases the PWSCC susceptibility of the BVPS-2 reactor vessel hot leg nozzles.
Without inspecting the hot leg nozzle welds directly, there are other inspections and equipment in the area that would identify a leak should one occur. The general area around and below the reactor is examined as part of the pressure test program walkdown during mode 3 start up. Any leakage noted would be investigated to
Beaver Valley Power Station, Unit No. 2 10 CFR 50.55a Request Number'. 2-TYP-4-RV-06, Revision 0 Page 5 of 6 determine the source. During operation, an increase in radiation levels within containment would be noted if there were significant leakage.
Also, BVPS-2 has an integrated leakage monitoring program that monitors RCS leakage. The Unit 2 reactor coolant system (RCS) inventory is calculated by procedures 205T-6 .2 or 205T-6 .2A reactor coolant system water inventory balance, with a surveillance test requirement to be performed every 72hours. The RCS lntegrated Leakage Program (112-ADM-0710) provides guidance where the RCS leakage is quantified and compared to the recent history of RCS leakage to better understand if current changes are outside normally expected values.
The RCS lntegrated Leakage Program includes guidance to determine if any program action level criteria are exceeded. The action level criteria are provided in the following table.
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lftl E irrrrs:r-lrirr LrrL, ii:r,rir'":r l.rrs faJ{^ iitr.t .t " . " r Ithan O.15 gprnl l,qru &rnl of tfte larf #rrss mnmedft,ry {#rdrft#fird ftC$ hn* rfiffi Latc{ f NOru*,rr theo Irnmn {JmdarrtflIed nCS tsrkilste * .? rdtrd laav*aaboir
- rfu Uniden#fied ftfS twk rarfe greafer flron 0.i glpml lls Lerrsl J l{r **u 1.m{enfr#ed ftCS lapk raf+ grearsr lhan fr.nenn + J Standarid The 0.1 gpm is consistent with WCAP-16465-NP, Pressurized Water Reactor Owners Group Standard RCS Leakage Action Levels and Response Guidelines for Pressurized Water Reactors and is one-tenth of the technical specification (TS) limit for unidentified leakage. The RCS lntegrated Leakage Program includes the requirement to identify the leakage source and could include entering containment to identify the source of the leakage. lf the source of the leakage is found and isolated, the program directs operation personnel to re-perform an RCS leak rate calculation to confirm that the source of leakage has been addressed.
The RCS leakage iuantity is reviewed against the TS associated with RCS leakage criteria. Depending on the source identified, a shutdown could be required in
Beaver Valley Power Station, Unit No. 2 10 CFR 50.55a Request Number: 2-TYP-4-RV-06, Revision 0 Page 6 of 6 accordance with TS Limiting Condition for Operation (LCO) 3.4.13 that has the following specific limits:
- a. No pressure boundary LEAKAGE;
- b. 1 gpm unidentified LEAKAGE;
- c. 10 gpm identified LEAKAGE; and
- d. 150 gallons per day primary to secondary LEAKAGE through any one steam generator (SG).
A through-wall leak from the reactor vessel hot leg nozzle weld would constitute pressure boundary leakage.
Should any of these limitations be exceeded, the appropriate LCO Condition would be entered and the required actions performed within the specified completion time, including plant shutdown if required.
Based on previous BVPS-2 examination results, a review of relevant industry operating experience, compensatory actions that would detect leakage if it were to occur, the benefits of zinc addition, and the ability to detect leakage while the plant is operating, it is determined that the structural integrity of the BVPS-2 reactor vessel hot leg nozzle welds will be maintained even if the visual exams are not performed during the 2R21 refueling outage.
6. Duration of Proposed Alternative
BVPS-2 is requesting relief to eliminate performance of this examination during 2R21 and resume the normal outage examination frequency at the next opportunity, currently expected to be the next refueling outage (2R22) set to begin fall 2021.
- 7. References
- 1. Code Case N-770-2, Alternative Examination Requirements and Acceptance Standards for Class 1 PWR Piping and Vessel Nozzle Butt Welds Fabricated with UNS N06082 or UNS W86182 Weld Filler Material with or Without Application of listed Mitigation Activities,Section XI, Division 1.
- 2. Code Case N-722-1, Additional Examinations of PWR Pressure Retaining Welds in Class 1 Components Fabricated with Alloy 600/82/182 Materials,Section XI, Division 1.