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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:Report
MONTHYEARL-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-22-154, Pressure and Temperature Limits Report Revision2022-07-0606 July 2022 Pressure and Temperature Limits Report Revision L-22-054, Steam Generator Inspection Report- Fall 2021 Refueling Outage2022-02-10010 February 2022 Steam Generator Inspection Report- Fall 2021 Refueling Outage IR 05000334/20210052021-09-0101 September 2021 Updated Inspection Plan for the Beaver Valley Power Station, Units 1 and 2 (Report 05000334/2021005 and 05000412/2021005) ML21131A0242021-05-17017 May 2021 Review of Reactor Pressure Vessel Capsule Y Analysis Report ML21055A0582021-03-10010 March 2021 Review of the Refueling Outage 21 Generic Letter 95-05 Voltage-Based Alternate Repair Criteria and F* Reports L-21-063, Facility Id 04-13361, Unit 2 Emergency Diesel Generator Day Tanks2021-02-0101 February 2021 Facility Id 04-13361, Unit 2 Emergency Diesel Generator Day Tanks ML21026A3372021-01-26026 January 2021 Steam Generator Inspection Report Revision - Spring 2020 Refueling Outage L-20-257, Request for One-Time Exemption from 10 CFR 50, Appendix E, Biennial Emergency Preparedness Exercise Requirements Due to COVID-19 Pandemic2020-09-22022 September 2020 Request for One-Time Exemption from 10 CFR 50, Appendix E, Biennial Emergency Preparedness Exercise Requirements Due to COVID-19 Pandemic ML21127A1652020-07-17017 July 2020 ODCM: Index, Matrix and History of ODCM Changes L-21-135, ODCM: Liquid Effluents2020-07-17017 July 2020 ODCM: Liquid Effluents ML21127A1482020-07-17017 July 2020 ODCM: Controls for RETS and REMP Programs 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 ML19280A0382019-10-22022 October 2019 Ti 2515-191 Summary of Findings ML21127A1562019-07-17017 July 2019 ODCM: Gaseous Effluents ML19051A1092019-02-20020 February 2019 LTR-SDA-18-054-NP, Revision 0 - Technical Justification to Support the Extended Volumetric Examination Interval for Beaver Valley Unit 2 Reactor Vessel Inlet Nozzle to Safe End Dissimilar Metal Welds L-19-053, Emergency Response Data System Data Point Library Update2019-02-20020 February 2019 Emergency Response Data System Data Point Library Update L-18-037, Annual 10 CFR 50.46 Report of Changes to or Errors in Emergency Core Cooling System Evaluation Models2018-06-18018 June 2018 Annual 10 CFR 50.46 Report of Changes to or Errors in Emergency Core Cooling System Evaluation Models L-18-017, Request to Extend Enforcement Discretion Provided in Enforcement Guidance Memorandum 15-002 for Tornado-Generated Missile Protection Non-Conformances Identified in Response to Regulatory Issue Summary 2015-06, Tornado.2018-03-15015 March 2018 Request to Extend Enforcement Discretion Provided in Enforcement Guidance Memorandum 15-002 for Tornado-Generated Missile Protection Non-Conformances Identified in Response to Regulatory Issue Summary 2015-06, Tornado. ML21127A1592018-01-22022 January 2018 ODCM: Radiological Environmental Monitoring Program L-17-277, Pressure and Temperature Limits Reports, Revisions 8 and 92017-10-0404 October 2017 Pressure and Temperature Limits Reports, Revisions 8 and 9 L-17-259, Mitigating Strategies Assessment (MSA) for Flooding2017-09-20020 September 2017 Mitigating Strategies Assessment (MSA) for Flooding ML17213A0162017-05-11011 May 2017 Enclosure a: Beaver Valley, Units 1, Seismic Probabilistic Risk Assessment in Response to 50.54(f) Letter with Regard to NTTF 2.1 ML17213A0172017-05-11011 May 2017 Enclosure B: Beaver Valley, Unit 2, Seismic Probabilistic Risk Assessment in Response to 50.54(0 Letter with Regard to NTTF 2.1 ML17024A0362016-12-31031 December 2016 Operating Data Report for 2016 ML16312A3112016-11-0707 November 2016 Spent Fuel Pool Evaluation Supplemental Report. Response to NRC Request for Information Pursuant to 10 CFR 50.54(f) Regarding Recommendation 2.1 of the Near-Term Task Force (Ntif) Review... L-16-327, Technical Basis for Optimization or Elimination of Liquid Penetrant Exams for the Embedded Flaw Repair2016-10-31031 October 2016 Technical Basis for Optimization or Elimination of Liquid Penetrant Exams for the Embedded Flaw Repair L-16-093, Report of Facility Changes, Tests and Experiments2016-06-0303 June 2016 Report of Facility Changes, Tests and Experiments L-16-065, Transmittal of 180-Day Steam Generator Tube Inspection Report - Technical Specification 5.6.6.22016-04-0606 April 2016 Transmittal of 180-Day Steam Generator Tube Inspection Report - Technical Specification 5.6.6.2 L-16-006, Submittal of Annual Report Pursuant to 10 CFR 75.11, Location Information2016-01-20020 January 2016 Submittal of Annual Report Pursuant to 10 CFR 75.11, Location Information L-15-386, Post Accident Monitoring Report Regarding Inoperability of One of the Source Range Nuclear Excore Instrumentation Channel for a Period in Excess of the 30-Day Restoration Time2016-01-0505 January 2016 Post Accident Monitoring Report Regarding Inoperability of One of the Source Range Nuclear Excore Instrumentation Channel for a Period in Excess of the 30-Day Restoration Time L-15-337, Completion of Required Action by NRC Order EA-12-049, Order Modifying Licenses with Regard to Requirements for Mitigation Strategies for Beyond-Design-Basis External Events and NRC Order EA-12-051, Reliable Spent Fuel Pool..2015-12-21021 December 2015 Completion of Required Action by NRC Order EA-12-049, Order Modifying Licenses with Regard to Requirements for Mitigation Strategies for Beyond-Design-Basis External Events and NRC Order EA-12-051, Reliable Spent Fuel Pool.. ML16277A1142015-10-28028 October 2015 ERS-MPD-93-007, Rev. 11, BVPS-U1 Gaseous Radioactivity Monitor Emergency Action Levels. ML15274A3072015-10-0505 October 2015 Staff Assessment of Information Provided Pursuant to Title 10 of the Code of Federal Regulations Part 50 Section 50.54 Seismic Hazard Reevaluations for Recommendation 2.1of the Near-Term Task Force Review L-15-187, Submittal of 10 CFR 50.46 Report of Changes to or Errors in Emergency Core Cooling System Evaluation Models2015-09-16016 September 2015 Submittal of 10 CFR 50.46 Report of Changes to or Errors in Emergency Core Cooling System Evaluation Models ML15132A4962015-05-22022 May 2015 Review of the 2014 Steam Generator Tube Inspection Report (TAC No. M4620) ML16277A1222015-04-20020 April 2015 ERS-LMR-14-001, Rev. 1, Liquid Monitor Alert Emergency Action Level (EAL) Set Points. ML15119A1042015-03-31031 March 2015 SG-CCOE-14-4-NP, Revision 1, Examination of Steam Generator Tubes Removed, Part 1 of 4 L-15-091, SG-CCOE-14-4-NP, Revision 1, Examination of Steam Generator Tubes Removed, Part 2 of 42015-03-31031 March 2015 SG-CCOE-14-4-NP, Revision 1, Examination of Steam Generator Tubes Removed, Part 2 of 4 ML21127A1582015-02-11011 February 2015 ODCM: Bases for ODCM Controls ML16277A1152015-01-30030 January 2015 ERS-HHM-87-014, Rev. 8, Unit 1 / Unit 2 ODCM Gaseous Effluent Monitor Setpoints. ML15027A2352015-01-28028 January 2015 Regulatory Audit in Support of the License Amendment Request to Implement a Risk-Informed, Performance-Based, Fire Protection Program (Tac Nos. MF3301 & MF3302) L-14-401, First Energy Nuclear Operating Company (FENOC) Expedited Seismic Evaluation Process (ESEP) Reports Response to NRC Request for Information Pursuant to 10 CFR50.54(f) Regarding Recommendation.1 of the Near-Term Task Force (NTTF) Review of In2014-12-19019 December 2014 First Energy Nuclear Operating Company (FENOC) Expedited Seismic Evaluation Process (ESEP) Reports Response to NRC Request for Information Pursuant to 10 CFR50.54(f) Regarding Recommendation.1 of the Near-Term Task Force (NTTF) Review of In L-14-352, Steam Generator Tube Inspection Report - Technical Specification 5.6.6.22014-11-0505 November 2014 Steam Generator Tube Inspection Report - Technical Specification 5.6.6.2 ML21127A1442014-10-0707 October 2014 ODCM: Information Related to 40 CFR 190 L-14-259, Firstenergy Nuclear Operating Company'S (Fenoc'S) Third Six-Month Status Report in Response to March 12, 2012 Commission Order Modifying Licenses with Regard to Reliable Spent Fuel Pool Instrumentation (Order Number EA-12-051)2014-08-28028 August 2014 Firstenergy Nuclear Operating Company'S (Fenoc'S) Third Six-Month Status Report in Response to March 12, 2012 Commission Order Modifying Licenses with Regard to Reliable Spent Fuel Pool Instrumentation (Order Number EA-12-051) ML14224A5772014-07-31031 July 2014 SG-SGMP-14-17, Revision 1, Beaver Valley, Unit 2, End-of-Cycle 17 Analysis and Prediction for End-of-Cycle 18 Voltage-Based Repair Criteria 90-Day Report. ML14224A5762014-07-31031 July 2014 SG-CCOE-14-1, Revision 0, Generic Letter 95-05 Tube Intersection Burst Test, Leakage Test, and Morphology Conclusions. ML14224A5752014-06-25025 June 2014 R17 Steam Generator F* (F Star) Report ML14009A0892014-02-10010 February 2014 Staff Assessment of the Seismic Walkdown Report Supporting Implementation of Near-Term Task Force Recommendation 2.3 Related to the Fukushima Daiichi Nuclear Power Plant Accident 2023-08-07
[Table view] Category:Miscellaneous
MONTHYEARL-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 IR 05000334/20210052021-09-0101 September 2021 Updated Inspection Plan for the Beaver Valley Power Station, Units 1 and 2 (Report 05000334/2021005 and 05000412/2021005) ML21131A0242021-05-17017 May 2021 Review of Reactor Pressure Vessel Capsule Y Analysis Report ML21055A0582021-03-10010 March 2021 Review of the Refueling Outage 21 Generic Letter 95-05 Voltage-Based Alternate Repair Criteria and F* Reports L-21-063, Facility Id 04-13361, Unit 2 Emergency Diesel Generator Day Tanks2021-02-0101 February 2021 Facility Id 04-13361, Unit 2 Emergency Diesel Generator Day Tanks L-20-257, Request for One-Time Exemption from 10 CFR 50, Appendix E, Biennial Emergency Preparedness Exercise Requirements Due to COVID-19 Pandemic2020-09-22022 September 2020 Request for One-Time Exemption from 10 CFR 50, Appendix E, Biennial Emergency Preparedness Exercise Requirements Due to COVID-19 Pandemic L-19-053, Emergency Response Data System Data Point Library Update2019-02-20020 February 2019 Emergency Response Data System Data Point Library Update L-18-037, Annual 10 CFR 50.46 Report of Changes to or Errors in Emergency Core Cooling System Evaluation Models2018-06-18018 June 2018 Annual 10 CFR 50.46 Report of Changes to or Errors in Emergency Core Cooling System Evaluation Models L-18-017, Request to Extend Enforcement Discretion Provided in Enforcement Guidance Memorandum 15-002 for Tornado-Generated Missile Protection Non-Conformances Identified in Response to Regulatory Issue Summary 2015-06, Tornado.2018-03-15015 March 2018 Request to Extend Enforcement Discretion Provided in Enforcement Guidance Memorandum 15-002 for Tornado-Generated Missile Protection Non-Conformances Identified in Response to Regulatory Issue Summary 2015-06, Tornado. ML17024A0362016-12-31031 December 2016 Operating Data Report for 2016 ML16312A3112016-11-0707 November 2016 Spent Fuel Pool Evaluation Supplemental Report. Response to NRC Request for Information Pursuant to 10 CFR 50.54(f) Regarding Recommendation 2.1 of the Near-Term Task Force (Ntif) Review... L-16-093, Report of Facility Changes, Tests and Experiments2016-06-0303 June 2016 Report of Facility Changes, Tests and Experiments L-16-065, Transmittal of 180-Day Steam Generator Tube Inspection Report - Technical Specification 5.6.6.22016-04-0606 April 2016 Transmittal of 180-Day Steam Generator Tube Inspection Report - Technical Specification 5.6.6.2 L-16-006, Submittal of Annual Report Pursuant to 10 CFR 75.11, Location Information2016-01-20020 January 2016 Submittal of Annual Report Pursuant to 10 CFR 75.11, Location Information L-15-386, Post Accident Monitoring Report Regarding Inoperability of One of the Source Range Nuclear Excore Instrumentation Channel for a Period in Excess of the 30-Day Restoration Time2016-01-0505 January 2016 Post Accident Monitoring Report Regarding Inoperability of One of the Source Range Nuclear Excore Instrumentation Channel for a Period in Excess of the 30-Day Restoration Time L-15-337, Completion of Required Action by NRC Order EA-12-049, Order Modifying Licenses with Regard to Requirements for Mitigation Strategies for Beyond-Design-Basis External Events and NRC Order EA-12-051, Reliable Spent Fuel Pool..2015-12-21021 December 2015 Completion of Required Action by NRC Order EA-12-049, Order Modifying Licenses with Regard to Requirements for Mitigation Strategies for Beyond-Design-Basis External Events and NRC Order EA-12-051, Reliable Spent Fuel Pool.. ML15274A3072015-10-0505 October 2015 Staff Assessment of Information Provided Pursuant to Title 10 of the Code of Federal Regulations Part 50 Section 50.54 Seismic Hazard Reevaluations for Recommendation 2.1of the Near-Term Task Force Review L-15-187, Submittal of 10 CFR 50.46 Report of Changes to or Errors in Emergency Core Cooling System Evaluation Models2015-09-16016 September 2015 Submittal of 10 CFR 50.46 Report of Changes to or Errors in Emergency Core Cooling System Evaluation Models ML15132A4962015-05-22022 May 2015 Review of the 2014 Steam Generator Tube Inspection Report (TAC No. M4620) ML15027A2352015-01-28028 January 2015 Regulatory Audit in Support of the License Amendment Request to Implement a Risk-Informed, Performance-Based, Fire Protection Program (Tac Nos. MF3301 & MF3302) L-14-401, First Energy Nuclear Operating Company (FENOC) Expedited Seismic Evaluation Process (ESEP) Reports Response to NRC Request for Information Pursuant to 10 CFR50.54(f) Regarding Recommendation.1 of the Near-Term Task Force (NTTF) Review of In2014-12-19019 December 2014 First Energy Nuclear Operating Company (FENOC) Expedited Seismic Evaluation Process (ESEP) Reports Response to NRC Request for Information Pursuant to 10 CFR50.54(f) Regarding Recommendation.1 of the Near-Term Task Force (NTTF) Review of In L-14-352, Steam Generator Tube Inspection Report - Technical Specification 5.6.6.22014-11-0505 November 2014 Steam Generator Tube Inspection Report - Technical Specification 5.6.6.2 L-14-259, Firstenergy Nuclear Operating Company'S (Fenoc'S) Third Six-Month Status Report in Response to March 12, 2012 Commission Order Modifying Licenses with Regard to Reliable Spent Fuel Pool Instrumentation (Order Number EA-12-051)2014-08-28028 August 2014 Firstenergy Nuclear Operating Company'S (Fenoc'S) Third Six-Month Status Report in Response to March 12, 2012 Commission Order Modifying Licenses with Regard to Reliable Spent Fuel Pool Instrumentation (Order Number EA-12-051) ML14009A0892014-02-10010 February 2014 Staff Assessment of the Seismic Walkdown Report Supporting Implementation of Near-Term Task Force Recommendation 2.3 Related to the Fukushima Daiichi Nuclear Power Plant Accident ML13284A0242013-09-0404 September 2013 Near-Term Task Force Recommendation 2.3, Seismic Walkdown Report Revision 1, Appendix B, Seismic Walkdown Checklists (Swcs), Part 3 of 5 ML13284A0272013-09-0404 September 2013 Near-Term Task Force Recommendation 2.3, Seismic Walkdown Report Revision 1, Appendix B, Seismic Walkdown Checklists (Swcs), Continued, Part 4 of 5 ML13284A0262013-09-0404 September 2013 Near-Term Task Force Recommendation 2.3, Seismic Walkdown Report Revision 1, Appendix a, Resumes and Qualifications, Part 2 of 5 ML13284A0252013-09-0404 September 2013 Near-Term Task Force Recommendation 2.3, Seismic Walkdown Report Revision 1, Part 1 of 5 ML13284A0232013-09-0404 September 2013 Near-Term Task Force Recommendation 2.3 Seismic Walkdown Report, Revision 1, Appendix a, Resumes and Qualifications, Part 2 of 5 ML13284A0192013-09-0404 September 2013 Near-Term Task Force Recommendation 2.3 Seismic Walkdown Report, Revision 1, Appendix B Seismic Walkdown Checklists (Swcs), Part 3 of 5 ML13284A0222013-09-0404 September 2013 Near-Term Task Force Recommendation 2.3 Seismic Walkdown Report, Revision 1, Part 1 of 5 ML13284A0202013-09-0404 September 2013 Near-Term Task Force Recommendation 2.3 Seismic Walkdown Report, Revision 1, Appendix B Seismic Walkdown Checklists (Swcs) Continued. Part 4 of 5 ML13284A0212013-09-0404 September 2013 Near-Term Task Force Recommendation 2.3 Seismic Walkdown Report, Revision 1, Appendix C Area Walk-by Checklists (Awcs), Part 5 of 5 L-13-036, Proposed Revision to Reactor Vessel Surveillance Capsule Withdrawal Schedules2013-05-28028 May 2013 Proposed Revision to Reactor Vessel Surveillance Capsule Withdrawal Schedules L-13-095, Discharge Monitoring Report (NPDES) Permit No. PA00256152013-02-22022 February 2013 Discharge Monitoring Report (NPDES) Permit No. PA0025615 L-12-437, Containment Liner Random and Non-Random Examinations Report2012-12-18018 December 2012 Containment Liner Random and Non-Random Examinations Report L-12-347, FENOC Response to NRC Request for Information Pursuant to 10 CFR 50.54(f) Regarding Flooding Aspects of Recommendation 2.3 of Near-Term Task Force Review of Insights from Fukushima Dai-ichi Accident2012-11-27027 November 2012 FENOC Response to NRC Request for Information Pursuant to 10 CFR 50.54(f) Regarding Flooding Aspects of Recommendation 2.3 of Near-Term Task Force Review of Insights from Fukushima Dai-ichi Accident ML13008A0422012-10-31031 October 2012 Near-Term Task Force Recommendation 2.3 Seismic Walkdown Report, Appendix B Seismic Walkdown Checklits (Swcs), Battery 2N. Pages 1 - 82 ML13008A0412012-10-31031 October 2012 Near-Term Task Force Recommendation 2.3 Seismic Walkdown Report. Cover Through A-40 ML13135A2382012-10-31031 October 2012 Near-Term Task Force Recommendation 2.3 Seismic Walkdown Report, Appendix B - Seismic Walkdown Checklists (Swcs), Sheet 1 of 513 Through Sheet 301 of 513 ML13008A0432012-10-31031 October 2012 Near-Term Task Force Recommendation 2.3 Seismic Walkdown Report, Appendix B Seismic Walkdown Checklists (Swcs), Purge Valve Isolation. Pages 83 - 161 ML13008A0442012-10-31031 October 2012 Near-Term Task Force Recommendation 2.3 Seismic Walkdown Report, Appendix B Seismic Walkdown Checklists (Swcs), Bus F1, Low Voltage Switchgear. Pages 162 - 240 ML13008A0452012-10-31031 October 2012 Near-Term Task Force Recommendation 2.3 Seismic Walkdown Report, Appendix B Seismic Walkdown Checklists (Swcs), SGI Level Switch SP9B6. Pages 241 - 318 ML13008A0462012-10-31031 October 2012 Near-Term Task Force Recommendation 2.3 Seismic Walkdown Report, Appendix B Seismic Walkdown Checklists (Swcs), Relief Valves SW 3963. Pages 319 - End ML13135A2412012-10-31031 October 2012 Near-Term Task Force Recommendation 2.3 Seismic Walkdown Report, Appendix C - Area Walk-by Checklists (Awcs) and Appendix D - Component List for Anchorage Configuration Check ML13135A2402012-10-31031 October 2012 Near-Term Task Force Recommendation 2.3 Seismic Walkdown Report, Appendix B - Seismic Walkdown Checklists (Swcs), Sheet 302 of 513 Through Sheet 513 of 513 ML13135A2362012-10-31031 October 2012 Near-Term Task Force Recommendation 2.3 Seismic Walkdown Report ML13135A2372012-10-31031 October 2012 Near-Term Task Force Recommendation 2.3 Seismic Walkdown Report, Appendix a - Resumes and Qualifications ML13008A0402012-10-23023 October 2012 Near-Term Task Force Recommendation 2.3 Seismic Walkdown Report, Appendix C Area Walk-by Checklists (Awcs), Prim. Aux. Bldg 735'6. Pages 135 - 191 ML13008A0392012-10-23023 October 2012 Near-Term Task Force Recommendation 2.3 Seismic Walkdown Report, Appendix C Area Walk-by Checklists (Awcs), Emer. Swgr. Room Train A. Pages 66 - 134 2023-08-07
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RrstEnergy Nuclear Operating Company Eric A. Larson Site Vice President Beaver Valley Power Station P.O. Box 4 Shippingport, PA 15077 724-682-5234 Fax: 724-643-8069 September 16, 2015 L-15-187 10 CFR 50.46(a)(3)(ii)
ATTN: Document Control Desk U.S. Nuclear Regulatory Commission Washington, DC 20555-0001
SUBJECT:
Beaver Valley Power Station, Unit Nos. 1 and 2 Docket No. 50-334, License No. DPR-66 Docket No. 50-412, License No. NPF-73 10 CFR 50.46 Report of Changes to or Errors in Emergency Core Cooling System Evaluation Models In accordance with Title 10 of the Code of Federal Regulations, Part 50, Section 50.46(a)(3)(ii),
FirstEnergy Nuclear Operating Company (FENOC) provides the attached report as annual notification of changes or errors in emergency core cooling system evaluation models or the application of the models for the Beaver Valley Power Station, Unit Nos. 1 (BVPS-1) and 2 (BVPS-2).
Current information for both large and small break loss-of-coolant accident (LOCA) transients is provided to satisfy 10 CFR 50.46 reporting requirements.
The attachments provide a summary list and description of each change or error in the acceptable evaluation models or the application of the models that affects the peak fuel cladding temperature (PCT) calculation for various LOCA transients, as well as the estimated PCT effects of the change or error. The changes or errors communicated to FENOC since the previous annual report are listed in the attachment and result in PCTs for the large and small break LOCA transients as follows:
BVPS-1 large break LOCA-1840 degrees Fahrenheit CF) BVPS-1 small break LOCA-1895°F BVPS-2 large break LOCA-1839°F BVPS-2 small break LOCA-191 rF Beaver Valley Power Station, Unit Nos. 1 and 2 L-15-187 Page2 There are no regulatory commitments contained in this submittal.
If there are any questions or if additional information is required, please contact Mr. Thomas A. Lentz, Manager-Fleet Licensing, at (330) 315-6810.
Sincerely, Eric A. Larson Attachments:
1 Summary of Peak Fuel Cladding Temperature (PCT) Effects for Beaver Valley Power Station (BVPS) Loss-of-Coolant Accident (LOCA) Transients 2 Descriptions of Emergency Core Cooling System (ECCS) Evaluation Model Changes or Errors cc: NRC Region I Administrator NRC Resident Inspector NRC Project Manager Director BRP/DEP Site BRP/DEP Representative Attachment 1 L-15-187 Summary of Peak Fuel Cladding Temperature (PCT) Effects for Beaver Valley Power Station (BVPS) Loss-of-Coolant Accident (LOCA) Transients Page 1 of 1 Beaver Valley Power Station Unit No. 1 (BVPS-1)
Large Break LOCA Description PCT Effect Attachment 2 (OF) Description (page) General Computer Code Maintenance 0 1 Errors in Decay Group Uncertainty Factors 0 3 Beaver Valley Units 1 and 2 Change of Modeled 0 5 Pressurizer Level Program BVPS-1 Small Break LOCA Description PCT Effect Attachment 2 (OF) Description (page) General Computer Code Maintenance 0 1 Fuel Rod Gap Conductance Error 0 6 Radiation Heat Transfer Model Error 0 7 SBLOCTA Pre-DNB Cladding Surface Heat Transfer 0 8 Coefficient Calculation Beaver Valley Power Station Unit No. 2_1BVPS-2l Large Break LOCA Description PCT Effect Attachment 2 _ _(_OF) Description (pagel General Computer Code Maintenance 0 1 Revised Total Computer Code Uncertainty in Best 0 2 Estimate Large Break LOCA Monte Carlo Simulations Containment Model Input Error Corrections 7 4 Beaver Valley Units 1 and 2 Change of Modeled 0 5 Pressurizer Level Program BVPS-2 Small Break LOCA Description PCT Effect Attachment 2 JoF) Desc riQti on _{p_age l General Computer Code Maintenance 0 1 Fuel Rod Gap Conductance Error 0 6 Radiation Heat Transfer Model Error 0 7 SBLOCTA Pre-DNB Cladding Surface Heat Transfer 0 8 Coefficient Calculation Attachment 2 L-15-187 Descriptions of Emergency Core Cooling System (ECCS) Evaluation Model Changes or Errors Page 1 of 8 GENERAL COMPUTER CODE MAINTENANCE Background Various changes have been made to enhance the usability of computer codes and to streamline future analyses.
Examples of these changes include:
modifying input variable definitions, units and defaults; improving the input diagnostic checks; enhancing the computer code output; optimizing active coding; and eliminating inactive coding. These changes represent discretionary changes that will be implemented on a forward-fit basis in accordance with Section 4.1.1 of WCAP-13451, "Westinghouse Methodology for Implementation of 10 CFR 50.46 Reporting,"
October 1992. Affected Evaluation Models 1. 1996 Westinghouse Best Estimate Large Break loss-of-coolant accident (LOCA) Evaluation Model (Applicable to BVPS Unit No.2 [BVPS-2])
- 2. 2004 Westinghouse Realistic Large Break LOCA Evaluation Model Using Automated Statistical Treatment of Uncertainty Method (ASTRUM)
(Applicable to BVPS Unit No. 1 [BVPS-1])
- 3. 1985 Westinghouse Small Break LOCA Evaluation Model with NOTRUMP (a nodal transient small break and general network computer code) (Applicable to BVPS-1 and BVPS-2) Estimated Effect The nature of these changes leads to an estimated peak fuel cladding temperature (PCT) impact of 0 degrees Fahrenheit
(°F) for both BVPS-1 and BVPS-2.
Attachment 2 L-15-187 Page 2 of 8 REVISED TOTAL COMPUTER CODE UNCERTAINTY IN BEST ESTIMATE LARGE BREAK LOCA MONTE CARLO SIMULATIONS Background As part of the WCOBRAITRAC validation basis for use in Best-Estimate (BE) Break Loss-of-Coolant Accident (LBLOCA)
- analyses, simulations of many separate effects tests and integral effects tests were performed.
In the Westinghouse 1996 BE LBLOCA evaluation model (the code qualification document evaluation model), the simulations of the test data that included core heat transfer measurements were used in the uncertainty methodology in two ways: 1. To develop heat transfer multiplier distributions that are used in the HOTSPOT computer code for local hot fuel rod calculations.
- 2. To determine the minimum code uncertainties for blowdown and reflood that are used in the MONTECF computer code for the Monte Carlo uncertainty calculations.
The effects of several changes and error corrections to the WCOBRAITRAC computer code were previously reported in 2013, including effects of revised heat transfer multiplier distributions.
As a result of previously reported changes and error corrections made to WCOBRAITRAC, the total code uncertainty values utilized in the MONTECF computer code for both blowdown and reflood were recalculated.
The recalculated total code uncertainty values are lower than the previous values. Resolution of this issue represents a non-discretionary change in accordance with Section 4.1.2 of WCAP-13451.
Affected Evaluation Model 1. 1996 Westinghouse Best Estimate Large Break LOCA Evaluation Model (Applicable to BVPS-2) Estimated Effect The total code uncertainty value is essentially a minimum limit for the superposition correction in the Monte Carlo simulations performed with the MONTECF computer code. As a result, a reduction in the minimum superposition correction would be expected to produce a small benefit, and will thus be conservatively assigned an estimated PCT impact of ooF for BVPS-2.
Attachment 2 L-15-187 Page 3 of 8 ERRORS IN DECAY GROUP UNCERTAINTY FACTORS Background Errors in the calculation of decay heat were discovered in the WCOBRA!TRAC computer code. The decay group uncertainty factors for each fissile isotope are provided in Table 8-14 of WCAP-16009-P-A, "Realistic Large Break LOCA Evaluation Methodology Using Automated Statistical Treatment of Uncertainty Method (ASTRUM),"
January 2005. The uncertainty factors for 239Pu were applied to 238U, and those for 238U were applied to 239Pu. This error causes an over-prediction of the uncertainty in decay power from 239Pu and an under-prediction of the uncertainty in decay power from 238U. Further, the decay group uncertainty factor for decay group 6 of 235U was erroneously coded as 2.5 percent instead of 2.25 percent.
Correction of these errors impacts the application of the sampled decay heat uncertainty, which may result in small changes to the decay heat power. These issues have been evaluated to estimate the impact on ASTRUM LBLOCA analysis results.
The resolution of these issues represents a closely-related group of non-discretionary changes in accordance with Section 4.1.2 of WCAP-13451.
Affected Evaluation Model 1. 2004 Westinghouse Realistic Large Break LOCA Evaluation Model Using ASTRUM (Applicable to BVPS-1) Estimated Effect Based on inspection of the limiting cases, it was concluded that the decay heat power was conservatively modeled for the limiting case, resulting in an estimated PCT impact of ooF for BVPS-1.
Attachment 2 L-15-187 Page 4 of 8 CONTAINMENT MODEL INPUT ERROR CORRECTIONS Background Several input errors were discovered in the containment model for the BVPS-2 BE LBLOCA analysis.
The input errors included an incorrect specification of the number of ECCS spill table entries, a typo in the heat transfer surface area of one heat sink, and an incorrect thickness input for one of the heat sinks. These input errors have been evaluated to estimate the impact on the BE LBLOCA analysis results.
The resolution of these issues represents a non-discretionary change in accordance with Section 4.1.2 of WCAP-13451.
Affected Evaluation Model 1. 1996 Westinghouse Best Estimate Large Break LOCA Evaluation Model (Applicable to BVPS-2) Estimated Effect The correction of these errors resulted in a lower calculated containment
- pressure, and the plant-specific evaluation resulted in an estimated PCT increase of rF for BVPS-2.
Attachment 2 L-15-187 Page 5 of 8 BEAVER VALLEY UNITS 1 AND 2 CHANGE OF MODELED PRESSURIZER LEVEL PROGRAM Background It was discovered that the pressurizer level program modeled in the LBLOCA analyses of record for BVPS-1 and BVPS-2 was changed when implemented at the plant. Modeling the as-implemented program results in a change to the assumed pressurizer water volume range in the LBLOCA analyses.
The change in pressurizer water volume range resulting from the as-implemented program has been evaluated.
This change represents a non-discretionary change in accordance with Section 4.1.2 of WCAP-13451
. Affected Evaluation Models 1. 1996 Westinghouse Best Estimate Large Break LOCA Evaluation Model (Applicable to BVPS-2) 2. 2004 Westinghouse Realistic Large Break LOCA Evaluation Model Using ASTRUM (Applicable to BVPS-1) Estimated Effect The plant-specific evaluation determined that the change in pressurizer water volume range was insignificant, leading to an estimated PCT impact of ooF for both BVPS-1 and BVPS-2.
Attachment 2 L-15-187 Page 6 of 8 FUELRODGAPCONDUCTANCEERROR Background An error was identified in the fuel rod gap conductance model in the NOTRUMP computer code. The error is associated with the use of an incorrect temperature in the calculation of the cladding emissivity term. This error corresponds to a discretionary change as described in Section 4.1.2 of WCAP-13451.
Affected Evaluation Model 1. 1985 Westinghouse Small Break LOCA Evaluation Model with NOTRUMP (Applicable to BVPS-1 and BVPS-2) Estimated Effect The estimated effect was determined based on a combination of engineering judgment of the phenomena and physics of a small break LOCA and sensitivity calculations performed with the advanced plant version of NOTRUMP.
It was concluded that this error has a negligible effect on small break LOCA analysis
- results, leading to an estimated PCT impact of ooF for both BVPS-1 and BVPS-2.
Attachment 2 L-15-187 Page 7 of 8 RADIATION HEAT TRANSFER MODEL ERROR Background Two errors were discovered in the calculation of the radiation heat transfer coefficient within the fuel rod model of the NOTRUMP computer code. First, existing logic did not preclude non-physical negative or large (negative or positive) radiation heat transfer coefficients from being calculated.
These erroneous calculations occurred when the vapor temperature exceeded the cladding surface temperature or when the predicted temperature difference was less than 1 °F. Second, a temperature term incorrectly used degrees Fahrenheit instead of Rankine.
These errors represent a closely related group of non-discretionary changes in accordance with Section 4.1.2 of WCAP-13451.
Affected Evaluation Model 1. 1985 Westinghouse Small Break LOCA Evaluation Model with NOTRUMP (Applicable to BVPS-1 and BVPS-2) Estimated Effect The estimated effect was determined based on a combination of engineering judgment of the phenomena and physics of a small break LOCA and sensitivity calculations performed with the advanced plant version of NOTRUMP.
It was concluded that this error has a negligible effect on small break LOCA analysis
- results, leading to an estimated PCT impact of 0°F for both BVPS-1 and BVPS-2.
Attachment 2 L-15-187 Page 8 of 8 SBLOCTA PRE-DNB CLADDING SURFACE HEAT TRANSFER COEFFICIENT CALCULATION Background Two errors were discovered in the pre-departure from nucleate boiling (pre-ON B) cladding surface heat transfer coefficient calculation in the SBLOCTA computer code (cladding heat-up calculations).
The first error is a result of inconsistent time units (hours versus seconds) in the parameters used for the calculation of the Reynolds and Prandtl numbers, and the second error relates to an incorrect diameter used to develop the area term in the cladding surface heat flux calculation.
Both of these issues impact the calculation of the pre-DNB convective heat transfer coefficient, representing a closely related group of non-discretionary changes to the evaluation model as described in Section 4.1.2 of WCAP-13451.
Affected Evaluation Model 1. 1985 Westinghouse Small Break LOCA Evaluation Model with NOTRUMP (Applicable to BVPS-1 and BVPS-2) Estimated Effect These errors have been corrected in the SBLOCTA computer code. Because this condition occurred prior to DNB, it was judged that these errors had no direct impact on the cladding heat-up related to the core uncover period. A series of validation tests confirmed that these errors have a negligible effect on small break LOCA analysis
- results, leading to an estimated PCT impact of ooF for both BVPS-1 and BVPS-2.