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Category:Code Relief or Alternative
MONTHYEARML24226A3652024-05-13013 May 2024 Approval of Request for Alternative from Certain Requirements of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code ML23249A1842023-09-18018 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 2024-05-13
[Table view] Category:Letter
MONTHYEARML24134A1522024-09-17017 September 2024 Exemption from the Requirements of 10 CFR 50.71(e)(4) Final Safety Analysis Report Update Schedule (EPID L-2024-LLE-0005) - Letter L-24-038, License Amendment Request to Remove the Expired One-Time Action for Valve Leak Repair in Technical Specification (TS) 3.5.2 and to Correct Core Operating Limits Rep Ort (COLR) References2024-09-17017 September 2024 License Amendment Request to Remove the Expired One-Time Action for Valve Leak Repair in Technical Specification (TS) 3.5.2 and to Correct Core Operating Limits Rep Ort (COLR) References L-24-015, Twenty-Ninth Refueling Outage Inservice Inspection Summary Report2024-09-17017 September 2024 Twenty-Ninth Refueling Outage Inservice Inspection Summary Report ML24260A1912024-09-16016 September 2024 Operator Licensing Examination Approval IR 05000334/20240052024-08-29029 August 2024 Updated Inspection Plan for Beaver Valley Power Station, Units 1 and 2 (Reports 05000334/2024005 and 05000412/2024005) L-24-188, Submittal of Quality Assurance Program Manual, Revision 302024-08-27027 August 2024 Submittal of Quality Assurance Program Manual, Revision 30 IR 05000334/20244022024-08-22022 August 2024 Security Baseline Inspection Report 05000334/2024402 and 05000412/2024402 (Cover Letter Only) IR 05000334/20240102024-08-20020 August 2024 Comprehensive Engineering Team Inspection - Inspection Report 05000334/2024010 and 05000412/2024010 L-24-186, Response to RAI for Exemption Request from 10 CFR 50.71(e)(4) Final Safety Analysis Update Schedule2024-08-15015 August 2024 Response to RAI for Exemption Request from 10 CFR 50.71(e)(4) Final Safety Analysis Update Schedule IR 05000334/20240022024-08-0505 August 2024 Integrated Inspection Report 05000334/2024002 and 05000412/2024002 ML24208A0462024-07-26026 July 2024 NRC Office of Investigations Case No. 1-2023-005 L-24-182, Discharge Monitoring Report (NPDES) Permit No. PA00256152024-07-23023 July 2024 Discharge Monitoring Report (NPDES) Permit No. PA0025615 L-24-161, Discharge Monitoring Report (NPDES) Permit No. PA00256152024-07-19019 July 2024 Discharge Monitoring Report (NPDES) Permit No. PA0025615 L-24-158, Reactor Trip Due to Bypass Feedwater Regulating Valve Failure to Control2024-07-17017 July 2024 Reactor Trip Due to Bypass Feedwater Regulating Valve Failure to Control L-24-014, Annual 10 CFR 50.46 Report of Changes to or Errors in Emergency Core Cooling Svstem Evaluation Models2024-07-16016 July 2024 Annual 10 CFR 50.46 Report of Changes to or Errors in Emergency Core Cooling Svstem Evaluation Models ML24193A2912024-07-16016 July 2024 Revision to Reactor Vessel Surveillance Capsule Withdrawal Schedule IR 05000334/20245012024-07-0808 July 2024 Emergency Preparedness Biennial Exercise Inspection Report 05000334/2024501 and 05000412/2024501 L-24-157, Unanalyzed Containment Bypass Condition Due to Degraded River Water Piping2024-07-0202 July 2024 Unanalyzed Containment Bypass Condition Due to Degraded River Water Piping L-24-164, BV-2, Post Accident Monitor Report2024-06-27027 June 2024 BV-2, Post Accident Monitor Report IR 05000334/20244012024-06-26026 June 2024 Material Control and Accounting Program Inspection Report 05000334/2024401 and 05000412/2024401 (Cover Letter Only) L-24-094, Reactor Vessel Surveillance Capsule Withdrawal Schedule2024-06-24024 June 2024 Reactor Vessel Surveillance Capsule Withdrawal Schedule L-24-152, Response to NRC Regulatory Issue Summary 2024-01 Preparation and Scheduling of Operator Licensing Examinations2024-06-17017 June 2024 Response to NRC Regulatory Issue Summary 2024-01 Preparation and Scheduling of Operator Licensing Examinations L-24-114, Automatic Reactor Trip Due to Low Reactor Coolant Flow Caused by Transformer Overexcitation Protection Relay Actuation2024-06-11011 June 2024 Automatic Reactor Trip Due to Low Reactor Coolant Flow Caused by Transformer Overexcitation Protection Relay Actuation L-24-115, Offsite AC Power Source Inoperable Due to Breaker Failure Resulting in Condition Prohibited by Technical Specification2024-06-0606 June 2024 Offsite AC Power Source Inoperable Due to Breaker Failure Resulting in Condition Prohibited by Technical Specification ML24135A2282024-05-29029 May 2024 Review of the Spring 2023 Outage Generic Letter 95-05 Voltage-Based Alternate Repair Criteria and Steam Generator F Star Reports ML24135A1702024-05-29029 May 2024 – Steam Generator Tube Inspection - Review of the Spring 2023 Tube Inspection Reports L-24-021, Cycle 30 Core Operating Limits Report2024-05-23023 May 2024 Cycle 30 Core Operating Limits Report L-24-121, Discharge Monitoring Report (NPDES) Permit No. PA00256152024-05-23023 May 2024 Discharge Monitoring Report (NPDES) Permit No. PA0025615 ML24141A1052024-05-20020 May 2024 Senior Reactor and Reactor Operator Initial License Examinations L-24-107, CFR 50.55a Request 1-TYP-5-RRP-l, for Alternative Method of Nondestructive Examination for River Water Outlet Piping Weld Repair2024-05-13013 May 2024 CFR 50.55a Request 1-TYP-5-RRP-l, for Alternative Method of Nondestructive Examination for River Water Outlet Piping Weld Repair IR 05000334/20240012024-05-0808 May 2024 Integrated Inspection Report 05000334/2024001 and 05000412/2024001 L-23-269, Application to Revise Technical Specifications to Adopt TSTF-569, Revision of Response Time Testing Definitions2024-05-0707 May 2024 Application to Revise Technical Specifications to Adopt TSTF-569, Revision of Response Time Testing Definitions L-24-054, Submittal of 2023 Annual Radioactive Effluent Release Report 2023 Annual Radiological Environmental Operating Report and 2023 Annual Environmental Operating Report (Non-Radiological)2024-04-29029 April 2024 Submittal of 2023 Annual Radioactive Effluent Release Report 2023 Annual Radiological Environmental Operating Report and 2023 Annual Environmental Operating Report (Non-Radiological) L-24-089, Emergency Preparedness Plan2024-04-23023 April 2024 Emergency Preparedness Plan L-24-088, Discharge Monitoring Report (NPDES) Permit No. PA0025615 for March 20242024-04-22022 April 2024 Discharge Monitoring Report (NPDES) Permit No. PA0025615 for March 2024 ML24101A2752024-04-10010 April 2024 Response to Request for Additional Information Regarding Spring 2023 180-Day Steam Generator Tube Inspection Report L-24-082, Withdrawal of Exemption Request2024-04-0303 April 2024 Withdrawal of Exemption Request L-24-013, Annual Notification of Property Insurance Coverage2024-03-26026 March 2024 Annual Notification of Property Insurance Coverage L-24-064, Submittal of Discharge Monitoring Report (NPDES) Permit No. PA00256152024-03-13013 March 2024 Submittal of Discharge Monitoring Report (NPDES) Permit No. PA0025615 ML24044A0662024-03-0404 March 2024 Exemption from Select Requirements of 10 CFR Part 73 (EPID L-2023-LLE-0083 (Security Notifications, Reports, and Recordkeeping and Suspicious Activity Reporting ML24057A0752024-03-0101 March 2024 the Associated Independent Spent Fuel Storage Installations IR 05000334/20230062024-02-28028 February 2024 Annual Assessment Letter for Beaver Valley Power Station, Units 1 and 2 (Reports 05000334/2023006 and 05000412/2023006) CP-202300502, Notice of Planned Closing of Transaction and Provision of Documents to Satisfy Order Conditions2024-02-23023 February 2024 Notice of Planned Closing of Transaction and Provision of Documents to Satisfy Order Conditions L-23-264, Request for Exemption from 10 CFR 50.71(e)(4) Final Safety Analysis Report Update Schedule2024-02-23023 February 2024 Request for Exemption from 10 CFR 50.71(e)(4) Final Safety Analysis Report Update Schedule L-24-050, Retrospective Premium Guarantee2024-02-22022 February 2024 Retrospective Premium Guarantee IR 05000334/20230042024-02-12012 February 2024 Integrated Inspection Report 05000334/2023004 and 05000412/2023004 ML24025A0922024-01-25025 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 L-23-198, Independent Spent Fuel Storage Installation - Request for Exemption from Enhanced Weapons, Firearms Background Checks, and Security Event Notifications Implementation2023-12-0707 December 2023 Independent Spent Fuel Storage Installation - Request for Exemption from Enhanced Weapons, Firearms Background Checks, and Security Event Notifications Implementation IR 05000334/20230112023-12-0404 December 2023 Age-Related Degradation Inspection Report 05000334/2023011 and 05000412/2023011 2024-09-17
[Table view] Category:Safety Evaluation
MONTHYEARML24193A2912024-07-16016 July 2024 Revision to Reactor Vessel Surveillance Capsule Withdrawal Schedule 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 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 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 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 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 2024-07-16
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Mr. Marty L. Richey Site Vice President UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 November 21, 2016 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 POWER STATION, UNIT 2-RELIEF REQUEST BV2-PZR-01 REGARDING ALTERNATIVE TO REQUIREMENTS FOR COMPONENTS CONNECTED TO THE STEAM SIDE OF THE PRESSURIZER (CAC NO. MF7790) Dear Mr. Richey: By letter dated June 3, 2016 (Agencywide Documents Access and Management System Accession No. ML 16158A306), FirstEnergy Nuclear Operating Company (the licensee) submitted a request to the U.S. Nuclear Regulatory Commission (NRC) for relief from certain requirements of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (the Code) at the Beaver Valley Power Station, Unit 2. The proposed alternative would allow certain piping, tubing, fittings, valves, and supports to remain as designed and constructed in lieu of upgrading the design and replacing these components with those constructed to ASME Code Class 1 and 2 requirements. The NRC staff has determined that the proposed alternative to the requirements of 1 O CFR 50.55a(c) is authorized for BVPS-2 on the basis that compliance with the ASME Code, Section Ill design requirements for Class 1 components would result in hardship without a compensating increase in the level of quality and safety pursuant to 10 CFR 50.55a(a)(z)(2). The licensee's proposed alternative provides reasonable assurance that the pressurizer upper level instrument and other lines and associated components, as designed and constructed, will perform their intended safety function. The alternative is authorized for the remaining life of the plant. All other ASME Code, Section Ill requirements for which relief was not specifically requested and approved remain applicable, including third-party review by the Authorized Nuclear lnservice Inspector.
M. Richey -2 -If you have any questions, please contact the Project Manager, Michael L. Marshall, Jr., at (301) 415-2871 or michael.marshall@nrc.gov. Docket No. 50-412 Enclosure: Safety Evaluation cc w/enclosure: Distribution via Listserv Sincerely, Stephen S. Koenick, Acting Chief Plant Licensing Branch 1-2 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 REQUEST BV2-PZR-01 FIRSTENERGY NUCLEAR OPERATING COMPANY 1.0 INTRODUCTION BEAVER VALLEY POWER STATION. UNIT 2 DOCKET NO. 50-412 By letter dated June 3, 2016 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML 16158A306), FirstEnergy Nuclear Operating Company (the licensee) submitted a request to the U.S. Nuclear Regulatory Commission (NRC or the Commission) for relief from the requirements of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (the Code), Section 111, Class 1 and Class 2, requirements for components connected to the steam side of the pressurizer for the Beaver Valley Power Station, Unit 2 (BVPS-2). Specifically, pursuant to Title 10 of the Code of Federal Regulations (10 CFR) 50.55a(z)(2), the licensee submitted Request BV2-PZR-01, Revision 0, for the use of the proposed alternative on the basis that compliance would result in hardship or unusual difficulty, without a compensating increase in the level of quality and safety. The alternative would allow certain piping, tubing, and valves to remain as designed and constructed in lieu of upgrading the design and replacing these components with those constructed to ASME Code Class 1 and Class 2 requirements. 2.0 REGULATORY EVALUATION The regulations in 1 O CFR 50.55a require that components that are part of the reactor coolant pressure boundary meet the requirements for Class 1 components in Section Ill of the ASME Code, except where alternatives have been authorized by the Commission pursuant to paragraphs (a)(z)(1) or (a)(z)(2) of 10 CFR 50.55a. In proposing alternatives, the licensee must demonstrate that ( 1) the proposed alternatives provide an acceptable level of quality and safety or (2) compliance would result in hardship or unusual difficulty, without a compensating increase in the level of quality and safety. Section 50.55a of 1 O CFR allows the Commission to authorize alternatives upon making the necessary findings. Enclosure
-2 -In addition, 10 CFR 50.55a(c) states, in part: (1) [ ... ] Components that are part of the reactor coolant pressure boundary must meet the requirements for Class 1 components in Section Ill of the ASME BPV [Boiler and Pressure Vessel] Code, except as provided in paragraphs (c)(2), through (4) of this section. (2) [ ... ] Components that are connected to the reactor coolant system and are part of the reactor coolant pressure boundary as defined in§ 50.2 need not meet the requirements of paragraph (c)(1) of this section, provided that: (i) [ ... ]In the event of postulated failure of the component during normal reactor operation, the reactor can be shut down and cooled down in an orderly manner, assuming makeup is provided by the reactor coolant makeup system; [ ... ]. In a letter dated April 3, 2000 (ADAMS Accession No. ML091190189), Westinghouse Electric Company issued its Nuclear Safety Advisory Letter (NSAL), "NSAL-00-006: Pressurizer Upper Level Instrument Safety Classification." This letter identified an issue where a break in the instrument lines for the upper (steam side) pressurizer level instruments may result in a rapid depressurization of the reactor coolant system sufficient to cause an emergency core cooling system actuation. Westinghouse NSAL 07-09, Revision 01, "Safety Classification of Small Lines Connected to the Pressurizer Steam Space," expanded the scope of the aforementioned letter to include all instrument and other small lines connected to the pressurizer steam space. In these letters, Westinghouse indicated that the aforementioned instrument lines should be classified as ASME Code Class 1. Given that a break in these lines would not result in a shutdown and cooldown "in an orderly manner," the licensee determined that the existing affected ASME Code Class 2 instrument and other lines and associated components connected to the pressurizer steam space should be classified as ASME Code Class 1, in accordance with 10 CFR 50.55a(c). The licensee has determined that these existing affected Class 2 lines are not in compliance with 10 CFR 50.55a(c). The licensee stated that the piping, fittings, tubing, valves, and supporting elements identified in the request were constructed using the ASME Code, Section Ill, Subsection NC (Class 2), requirements. Construction as used in Section Ill, Division 1, included requirements for materials, design, fabrication, examination, testing, inspection, and certification required in the manufacture and installation of items. Pursuant to 10 CFR 50.55a(z)(2), the licensee proposes to allow these lines and valves to remain as designed and constructed as Class 2. The NRC staff has approved similar requests for relief regarding classification of pressurizer upper level piping and components for the Wolf Creek Nuclear Operating Corporation on May 31, 2005 (ADAMS Accession No. ML051520526), and for the Callaway Plant, Unit 1 on April 29, 2009 (ADAMS Accession No. ML090840102).
-3 -3.0 TECHNICAL EVALUATION 3.1 Licensee's Relief Request The licensee stated that upgrading the piping, tubing, and valves to ASME Code, Section Ill, Subsection NB (Class 1 ), would result in a hardship or unusual difficulty, without a compensating increase in the level of quality and safety, because the scope of the change would require significant personnel radiation exposure due to component proximity to the pressurizer. The affected piping, tubing, and valves are part of the upper pressurizer instrument lines or pressurizer safety valve loop seal drain lines. The affected piping, tubing, and valves are shown in a table on pages 1 and 2 of the licensee's enclosure in the June 3, 2016, submittal. The licensee proposed alternative is to allow the piping, tubing, and valves identified in the request to remain as designed and constructed to ASME Class 2 and ANSI B31.1 non-nuclear safety (NNS) in lieu of upgrading the current design configuration and replacing these components with components constructed to ASME Section Ill, Class 1 and Class 2 requirements. To justify the proposed alternative, the licensee compared the ASME Code, Section Ill requirements in Subsection NB for Class 1 components to the design rules in Subsection NC for Class 2 components using the applicable editions and addenda of the ASME Code. The licensee also compared ANSI B31.1 Code and ASME, Section Ill, Subsection NC, for the NNS pipe. The comparison considered each article of Subsections NB, NC, and ANSI B31.1 (covering the areas of materials, design, fabrication and installation, examination, testing, protecting against overpressure, nameplates, stamping, and reports) and determined whether the differences were technical, quality, or administrative requirements. Differences in Section Ill administrative requirements such as certification and stamping, furnishing of a stress report, and marking of items, were determined to not reduce the quality or safety of the items and would only affect literal compliance with the ASME Code. Minimal differences were identified in quality requirements between Class 1 and Class 2, because most quality requirements are contained in the General Requirements Subsection NCA and are equally applicable to both Class 1 and Class 2. There are no differences in quality requirements that would reduce the quality or safety of the Class 2 and NNS components. 3.2 NRC Staff Evaluation Regarding the valve design aspects related to the relief request, the licensee indicated that while the requirements for Class 1 small valves are considerably different than the requirements for Class 2 small valves, the affected valves were evaluated to the Class 1 requirements and found to meet all of the technical requirements found in NB-3500. Regarding the material examination facets of the affected piping and components, the later provisions of NB-2501 (a) in the Summer 1983 Addenda exempted :::; 1-inch seamless pipe, tubes, and fittings from the examination requirements of NB-2500. Thus, there are no technical differences between Class 1 and Class 2 rules. Given this information, had the design and construction of these systems been completed at a later date, the current Class 2 configuration would meet the Class 1 material examination requirements of NB-2500. Based on these considerations, the staff finds that for the affected Class 2 valves and material examination requirements, the
-4 -Class 2 requirements provide an equivalent level of safety to Class 1 requirements provided in the ASME Code. The licensee states that NNS piping meets the requirements of ASME Section 111, Subsection NC-3600, for pipe design and has been analyzed to Class 2 requirements. The licensee also states that NNS socket welds were examined by liquid penetration method per NC-5250 during the fall 2015 refueling outage, and a visual examination of the NNS piping was also performed, confirming heat numbers associated with Class 1 piping. The staff reviewed the piping and instrument drawing for the NNS piping part and found that the NNS piping is the drain line located downstream from the isolation valve. The isolation valve is normally closed and the NNS piping does not contain any flow. On this basis, the staff determined that no increase in quality and safety would be realized by considering the NNS piping as Class 2 piping. The Class 2 piping and instrument piping (including tubing) identified in the relief request were designed and analyzed in accordance with the 1971 Edition with the Winter 1972 Addenda of Section Ill of the ASME Code. A provision added in the Summer 1975 Addendum to the 1974 Edition in subparagraph NB-3630(d) allowed Class 2 rules to be used for Class 1 design for piping less than or equal to 1 inch in size. Based on this provision, the affected piping and instrument piping technically meet the design requirements of the Code Class 1 rules. Therefore, the staff finds that the design rules used for the affected Class 2 piping provide an equivalent level of safety to Class 1 design requirements for the affected piping and tubing. Furthermore, the NRC staff concludes that the licensee has demonstrated that the compliance with ASME Code, Section Ill, Class 1 requirements for the lines and associated components described in the alternative would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety because the scope of the change would require substantial time and radiation exposure to upgrade the current design configuration and replace the affected piping, tubing, and valves with components constructed to ASME Section Ill, Class 1 and Class 2 requirements. 4.0 CONCLUSION The NRC concludes that the proposed alternative to the requirements of 10 CFR 50.55a(c) is authorized for BVPS-2 on the basis that compliance with the ASME Code, Section Ill design requirements for Class 1 components would result in hardship without a compensating increase in the level of quality and safety pursuant to 10 CFR 50.55a(a)(z)(2). The licensee's proposed alternative provides reasonable assurance that the pressurizer upper level instrument and other lines and associated components, as designed and constructed, will perform their intended safety function. The alternative is authorized for the remaining life of the plant. Principal Contributor: Kaihwa Hsu Taylor Lamb Date: November 21, 2016 M. Richey -2 -If you have any questions, please contact the Project Manager, Michael L. Marshall, Jr., at (301) 415-2871 or michael.marshall@nrc.gov. Docket No. 50-412 Enclosure: Safety Evaluation cc w/enclosure: Distribution via Listserv DISTRIBUTION: PUBLIC LPLl-2 R/F RidsRgn1 MailCenter Resource RidsACRS_MailCTR Resource JBowen, OEDO RidsNrrDorllpl1-2 Resource ADAMS A ccess1on N ML 16257 A621 o.: OFFICE DORL/LPLl-2/PM DORL/LPLl-2/LA NAME Tlamb LRonewicz DATE 09/27/2016 09/23/2016 Sincerely, IRA/ Stephen S. Koenick, Acting Chief Plant Licensing Branch 1-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation RidsNrrPMBeaverValley Resource RidsNrrLALRonewicz Resource RidsNrrDeEmcb Resource KHsu, NRR *b ., 1y e-ma1 DE/EMCB/BC* DORL/LPLl-2/BC DORL/LPLl-2/PM JQuichocho SKoenick MMarshall 10/17/2016 11/16/2016 11/21/2016 OFFICIAL RECORD COPY