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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] Category:Annual Operating Report
MONTHYEARL-22-038, 2021 Annual 10 CFR 50.46 Report of Changes to or Errors in Emergency Core Cooling System Evaluation Models2022-04-26026 April 2022 2021 Annual 10 CFR 50.46 Report of Changes to or Errors in Emergency Core Cooling System Evaluation Models L-19-075, Submittal of 2018 Annual 10 CFR 50.46 Report of Changes to or Errors in Emergency Core Cooling System Evaluation Models2019-04-15015 April 2019 Submittal of 2018 Annual 10 CFR 50.46 Report of Changes to or Errors in Emergency Core Cooling System Evaluation Models ML18122A1772018-04-21021 April 2018 Annual Radioactive Effluent Release Report, and Annual Radiological Environmental Operating Report ML18122A1792018-04-21021 April 2018 Annual Non-Radiological Environmental Operating Report L-17-114, Report of Facility Changes, Tests and Experiments2017-11-21021 November 2017 Report of Facility Changes, Tests and Experiments L-17-040, Transmittal of 2016 Annual 10 CFR 50.46 Report of Changes to or Errors in Emergency Core Cooling System Evaluation Models2017-06-20020 June 2017 Transmittal of 2016 Annual 10 CFR 50.46 Report of Changes to or Errors in Emergency Core Cooling System Evaluation Models ML17124A1282017-02-16016 February 2017 2016 Annual Environmental Operating Report (Non-Radiological) L-16-350, Independent Spent Fuel Storage Installation Changes, Tests, and Experiments2016-12-16016 December 2016 Independent Spent Fuel Storage Installation Changes, Tests, and Experiments L-16-096, Transmittal of 2015 Annual 10 CFR 50.46 Report of Changes to or Errors in Emergency Core Cooling System Evaluation Models and Commitment Change Related to Revised Peak Cladding Temperature Analysis2016-06-0909 June 2016 Transmittal of 2015 Annual 10 CFR 50.46 Report of Changes to or Errors in Emergency Core Cooling System Evaluation Models and Commitment Change Related to Revised Peak Cladding Temperature Analysis L-15-282, Report of Facility Changes, Tests, and Experiments for June - October 20152015-11-23023 November 2015 Report of Facility Changes, Tests, and Experiments for June - October 2015 L-14-202, Report of Facility Changes, Tests and Experiments for the Period of June 1, 2012 to May 31, 20142014-07-0101 July 2014 Report of Facility Changes, Tests and Experiments for the Period of June 1, 2012 to May 31, 2014 ML14147A0282014-04-16016 April 2014 Radioactive Effluent Release Report, and 2013 Annual Radiological Environmental Operating Report, Part 2 of 2 ML14147A0272014-04-16016 April 2014 Radioactive Effluent Release Report, and 2013 Annual Radiological Environmental Operating Report, Part 1 of 2 L-12-349, Report of Facility Changes, Tests and Experiments for Period October 29, 2010 Through September 24, 20122012-12-0303 December 2012 Report of Facility Changes, Tests and Experiments for Period October 29, 2010 Through September 24, 2012 ML12124A1072012-04-30030 April 2012 Radioactive Effluent Release Report, Section 3 and RTL#A9.630F, 2011 Annual Environmental Operating Report, Enclosure 3 ML12123A7242012-04-30030 April 2012 Radioactive Effluent Release Report, Rtl# A9.621B - 1/2-ODC-1.01, Revision 13, ODCM: Index, Matrix and History of ODCM Changes ML12123A7252012-04-30030 April 2012 Radioactive Effluent Release Report, Rtl# A9.621B - 1/2-ODC-2.01, Revision 9, ODCM: Liquid Effluents ML12123A7272012-04-30030 April 2012 Radioactive Effluent Release Report, Rtl# A9.621B - 1/2-ODC-2.03, Revision 4, ODCM: Radiological Environmental Monitoring Program L-12-085, Submittal of 2011 Radioactive Effluent Release Report and 2011 Annual Radiological Environmental Operating Report and 2011 Annual Environmental Operating Report (Non-Radiological)2012-04-30030 April 2012 Submittal of 2011 Radioactive Effluent Release Report and 2011 Annual Radiological Environmental Operating Report and 2011 Annual Environmental Operating Report (Non-Radiological) ML12123A7312012-04-30030 April 2012 Radioactive Effluent Release Report, Rtl# A9.621B - 1/2-ODC-3.03, Revision 11, ODCM: Controls for RETS and REMP Programs ML12123A7302012-04-30030 April 2012 Radioactive Effluent Release Report, Rtl# A9.621B - 1/2-ODC-3.02, Revision 2, ODCM: Bases for ODCM Controls ML12123A7292012-04-30030 April 2012 Radioactive Effluent Release Report, Rtl# A9.621B - 1/2-ODC-3.01, Revision 1, ODCM: Dispersion Calculation Procedure and Source Term Inputs ML12123A7262012-04-30030 April 2012 Radioactive Effluent Release Report, Rtl# A9.621B - 1/2-ODC-2.02, Revision 3, ODCM: Gaseous Effluents ML12123A7282012-04-30030 April 2012 Radioactive Effluent Release Report, Rtl# A9.621B - 1/2-ODC-2.04, Revision 1, ODCM: Information Related to 40 CFR 190 ML1112505302011-04-25025 April 2011 Beaver Valley - Rtl# A9.690E, 2010 Annual Radiological Environmental Operating Report, Enclosure 3 to L-11-041 ML1112505142011-04-25025 April 2011 Beaver Valley - Rtl# A9.690E, 2010 Radioactive Effluent Release Report and 2010 Annual Radiological Environmental Operating Report, Enclosure 1 to L-11-041 ML1112505352011-03-23023 March 2011 Rtl# A9.630F, 2010 Annual Environmental Operating Report Non-Radiological, Enclosure B to L-11-041 ML1013104142010-04-0101 April 2010 Rtl# A9.690E, 2009 Radioactive Effluent Release Report and 2009 Annual Radiological Environmental Operating Report. ML1013104162010-02-25025 February 2010 Rtl# A9.630F, 2009 Annual Environmental Operating Report (Non-Radiological). ML0912004172009-04-0303 April 2009 2008 Radioactive Effluent Release Report and 2008 Annual Radiological Environmental Operating Report L-09-098, Annual Environmental Operating Report, Non-Radiological2009-03-30030 March 2009 Annual Environmental Operating Report, Non-Radiological L-08-340, Transmittal of 10 CFR 50.46 Report of Changes or Errors in ECCS Evaluation Models2008-11-26026 November 2008 Transmittal of 10 CFR 50.46 Report of Changes or Errors in ECCS Evaluation Models ML0812808582008-04-16016 April 2008 Rtl# A9.690E, 2007 Radioactive Effluent Release Report and 2007 Annual Radiological Environmental Operating Report, Cover Page Through Enclosure 2, Attachment 2 L-07-062, Units 1 and 2, 2006 Firstenergy Corp. Annual Report2007-04-0909 April 2007 Units 1 and 2, 2006 Firstenergy Corp. Annual Report L-06-079, Annual Radioactive Effluent Release Report for 2005, and Annual Radiological Environmental Operating Report for 20052006-05-0101 May 2006 Annual Radioactive Effluent Release Report for 2005, and Annual Radiological Environmental Operating Report for 2005 L-05-148, Commitment Changes and Report of Facility Changes, Tests, and Experiments for October 12, 2003 Through April 11, 20052005-08-29029 August 2005 Commitment Changes and Report of Facility Changes, Tests, and Experiments for October 12, 2003 Through April 11, 2005 L-05-072, Attachment 2, Beaver Valley Power Station - Units 1 & 2, Annual Radioactive Effluent Release Report 2004, Offsite Dose Calculation Manual Changes2005-04-27027 April 2005 Attachment 2, Beaver Valley Power Station - Units 1 & 2, Annual Radioactive Effluent Release Report 2004, Offsite Dose Calculation Manual Changes L-03-180, Commitment Changes and Report of Facility Changes, Tests and Experiments2003-10-29029 October 2003 Commitment Changes and Report of Facility Changes, Tests and Experiments L-03-039, Firstenergy - Retrospective Premium Guarantee2003-03-0606 March 2003 Firstenergy - Retrospective Premium Guarantee L-02-101, 10 CFR 50.46 Report of Changes or Errors in ECCS Evaluation Models2002-09-25025 September 2002 10 CFR 50.46 Report of Changes or Errors in ECCS Evaluation Models L-02-092, Transmittal of Commitment Changes & Report of Facility Changes, Tests & Experiments for Beaver Valley Power Station Unit No. 1 & Unit No. 22002-08-30030 August 2002 Transmittal of Commitment Changes & Report of Facility Changes, Tests & Experiments for Beaver Valley Power Station Unit No. 1 & Unit No. 2 2022-04-26
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Beaver Valley Power Station P.O. Box 4 Shippingport, PA 15077 Firs/Energy Nuclear Operating Company Paul A. Harden 724-682-5234 Site Vice President Fax: 724-643-8069 December 3, 2012 L-12-349 10 CFR 50.59(d)(2)
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 Report of Facility Changes, Tests and Experiments In accordance with 10 CFR 50.59(d)(2), the FirstEnergy Nuclear Operating Company hereby submits the attached Report of Facility Changes, Tests and Experiments for Beaver Valley Power Station, Unit No.2. The report covers the period of October 29, 2010 through September 24, 2012.
There are no regulatory commitments contained in this letter. If there are any questions or if additional information is required, please contact Mr. Thomas A. Lentz, Manager - Fleet Licensing, at (330) 315-6810.
Attachment:
Beaver Valley Power Station, Unit No.2, Report of Facility Changes, Tests and Experiments, October 29, 2010 through September 24, 2012 cc: NRC Region I Administrator NRC Resident Inspector NRC Project Manager Director BRP/DEP Site BRP/DEP Representative
Attachment L-12-349 Page 1 of 5 Beaver Valley Power Station, Unit No.2, Report of Facility Changes, Tests and Experiments October 29, 2010 through September 24, 2012 Evaluation No. : 10-04767, Revision 0 Revise BVPS-2 UFSAR Section 9.1.2.2, "Spent Fuel Pool Storage" Activity
Description:
Beaver Valley Power Station, Unit No.2 (BVPS-2), Updated Final Safety Analysis Report (UFSAR), Section 9.1.2.2, stated that "the Spent Fuel Pool is sized to accommodate the storage of a minimum of one full core in the event the reactor must be emptied of fuel at any time during BVPS-2 life." This statement implies that full core offload capacity, termed "full core reserve storage capability," is a requirement to ensure that the spent fuel pool (SFP) must have sufficient space to accommodate the offload of all 157 assemblies from the BVPS-2 reactor core at any time. .
Full core storage capability is not an NRC licensing or safety requirement, as determined by a review of both the industry and the specific BVPS-2 licensing bases.
The current UFSAR 9.1.2.2 statement is correct, only if some or all of the spent fuel discharged over the operating life of BVPS was stored in some type of long term depository external to the site. Construction of such a facility was an unstated assumption at the time the nuclear fleet of the United States was first constructed and was thus implicit in the BVPS-2 initial licensing process in 1987. However, such a facility has never been constructed, rendering the background information incorrect. Increasing the storage capacity of individual plants must be employed instead. The proposed clarification acknowledges that maintaining full core offload storage capacity is simply a prudent management practice, rather than a licensing bases requirement. This full offload capacity may be temporarily lost due to the need for additional regulatory review and approval of on-site storage facilities, when the existing SFPcapacity is challenged.
The statement in Section 9.1.2.2 indicating the SFP can accept a full core offload at any time in BVPS-2 plant life was deleted, with an additional paragraph inserted explaining the position discussed above.
Summary of Evaluation:
There are no regulatory requirements either in the Code of Federal Regulations or the NUREGs (including NUREG 0800, Standard Review Plan) to have the ability to perform full core offloads. The statement that BVPS-2 would maintain that capability over its entire initial licensed operating life of 40 years was incorrect, since the same discussion in UFSAR Section 9.1.2.2 notes that the spent fuel pool storage capability is 1,088 assemblies, which is insufficient to support approximate 26 total 1.5 year-long operating cycles (40 years/1.5 years per operating cycle) averaging 57 assemblies per refueling (average based on a review of refueling data through the fall 2009 refueling outage [2R14]). Furthermore, there is no design analysis crediting the ability to perform full core offload to mitigate any identified design basis accident, and no discussion of such a mitigation strategy in the BVPS-2 emergency operating procedures. Additionally, the beyond design bases mitigation
Attachment L-12-349 Page 2 of 5 strategies of the severe accident management guidelines do not discuss use of full core offload.
A review of the safety evaluation report for BVPS-2, as well as the standard review plan, leads to the conclusion that the discussion of core offload was meant to demonstrate that, should full core offloads be performed, the existing SFP cooling capability was sufficient to accommodate the heat load. BVPS-2 determined the existing SFP cooling was adequate by analysis, which the NRC accepted.
Evaluation No. : 11-00626, Revision 1
Title:
Assessment of Beaver Valley Unit 2 Containment Response for Design Basis Accidents for Containment Atmospheric Conversion Project Activity
Description:
Calculation titled "Assessment of Beaver Valley Unit 2 Containment Response for Design Basis Accidents for Containment Atmospheric Conversion Project," Revision 2, is the approved calculation for the containment integrity analysis and recirculation spray pump net positive suction head for large break loss-of-coolantaccident (LBLOCA) and main steam line break (MSLB). The mass and energy (M&E) release data for both LBLOCA and MSLB were developed and supplied by Westinghouse. Some inputs to the Westinghouse M&E calculations vary on a cycle to cycle basis and were confirmed to be bounding by the reload process. One input parameter affecting the MSLB M&E data is the fuel moderator density coefficient (MDC). For BVPS-2 Cycle 16 (April 2011 through October 2012), it was determined that the MDC assumed in the MSLB M&E analysis would not be bounding for the hot zero (HZP) power case. Therefore, a decision was made to increase the assumed MDC and calculate new M&E data for this case to be used in the containment analysis.
The bounding 0 percent power case was chosen and assumed single failure of a main steam isolation valve to close. This case was designated as case 16M in the containment calculation. This addendum re-analyzed case 16M to determine the resulting changes in containment pressure, temperature and liner temperature. Associated UFSAR updates were also covered by this evaluation. The changes included the Section 6.2.1.1.3.7 discussion of MSLB results and Tables 6.2-10 and 6.2-11, as well as Figures 6.2-12 through 6.2-15. A change to the Technical Specification Bases, Sections 3.6.5, 3.6.6, and 3.6.7 was also required and covered by this evaluation.
Summary of Evaluation:
The hot zero power MSLB M&E release analysis incorporated input changes, which had the effect of increasing the return to power following the event. This resulted in an increase in the steam release to containment, which resulted in higher containment pressure and temperature. The peak containment pressure and containment temperature and liner temperature remained within acceptance limits. There was no impact on the reactor departure from nucleate boiling ratio (DNBR) analysis and the steam release for offsite and control room dose results were not changed. The changes did not introduce the possibility of a new accident or malfunction. There was no change in the method of evaluation, and no fission product barriers were challenged. All questions were answered no; therefore, a license amendment was not required.
Attachment L-12-349 Page 3 of 5 Evaluation No. : 11-05225, Revision 0 Pressurizer Surge Line LBB/SWOL Analysis (BVPS-2)
Activity
Description:
In 1987, a leak-before-break (LBB) evaluation was performed for the Beaver Valley Power Station, Unit 2 (BVPS-2) pressurizer surge line (PSL) and approved by the Nuclear Regulatory Commission (NRC). In 1988, the original LBB analysis was updated to evaluate the effects of thermal stratification, and subsequently approved by the NRC. An updated PSL LBB evaluation was also performed for the 9.4 percent power uprate program.
To mitigate primary water stress corrosion cracking (PWSCC), BVPS-2 applied a structural weld overlay (SWOL) at the Alloy 82/182 weld location of the PSL to the pressurizer nozzle.
The SWOL analysis is documented in WCAP-16612-P, Revision 0, "Beaver Valley Unit 2 Pressurizer Safety/Relief, Spray, and Surge Nozzles Structural Weld Overlay Qualification,"
September 2006. The subsequent LBB evaluation results are documented in WCAP-17394-P, Revision 0, "Leak-Before-Break Analysis, Update for the Beaver Valley Unit 2 Pressurizer Surge Line," December, 2011.
The NRC, in RIS 2010-07, "Regulatory Requirements for Application of Weld Overlays and Other Mitigation Techniques in Piping Systems Approved for Leak-Before-Break," states that any planned changes to an approved LBB analysis must be evaluated under 10 CFR 50.59.
Summary of Evaluation:
The original LBB evaluation methodology did not include PWSCC effects for leak rate calculations of the PSL because the PWSCC issue for Alloy 82/182 welds had not been identified at that time. The updated methodology, including SWOL, was approved by the NRC for the Waterford Steam Electric Station, Unit No.3, via Amendment 232. Its application to BVPS-2 was consistent with the NRC approval. Therefore, the methodology change is acceptable under 10 CFR 50.59, and a License Amendment was not required.
Evaluation No. : 12-03459, Revision 0 BVPS-2 Yard Excavation Between Pipe Trench and Electrical Manhole Activity
Description:
A through-wall leak was identified on an 8-inch service water line. The leak was located in a Beaver Valley Power Station, Unit No.2 (BVPS-2) pipe trench between the service building and safeguards building. To support repair of the pipe leak, a yard excavation was made between the west wall of the pipe trench and the east wall of an adjacent electrical manhole. During the yard excavation activities, missile protection for safety-related electrical components necessary for safe shutdown was degraded. This equipment was maintained operable with the reduced missile protection. The excavation was performed per an approved work order.
Attachment L-12-349 Page 4 of 5 The excavation permitted the implementation of a temporary modification that cut an opening in the pipe trench wall, allowing access for the pipe repair. The excavation uncovered the west side of the pipe trench and the east side of an adjacent electrical manhole, and was approximately 13 feet deep. It extended to the bottom of the pipe trench.
About 1'-2" below the bottom of the pipe trench was an electrical duct. The duct runs east from the electrical manhole and passes under the pipe trench.
Excavation also required temporary removal of a portion of a yard storm water drainage pipe. This pipe drains storm water from the trough adjacent to the service building roll-up door and empties into a yard catch basin.
The work order provided measures to limit the amount of storm water entering the excavation and the roll-up door trough. A sump pump in the excavation was available to remove any storm water entering the excavation.
The electrical manhole, the electrical duct and the pipe trench are safety-related structures.
The yard storm water drainage pipe is non-safety related, non-seismic.
Summary of Evaluation:
The requirement of the BVPS-2 UFSAR, Section 3.5.1.4, and safety evaluation report section 3.5.2 to maintain protection of structures, systems and components important to safety from tornado-generated missiles could not be maintained during the excavation activities. During excavation, one safety-related duct bank and one safety-related electrical manhole was uncovered and the 5 feet of soil missile protection was not maintained as defined in UFSAR Section 2.2.3.2. Soil cover depth is defined in a BVPS-2 document titled "Structural Design Criteria." The contingency plan included staging fill material in the immediate vicinity of the excavation with equipment available to refill the excavation within one hour of any tornado watch in accordance with a procedure titled "Acts of Nature -
Tornado or High Wind Condition." Refilling the excavated hole restored the UFSAR described tornado-generated missile protection.
There have been several previous probabilistic risk assessment (PRA) evaluations for similar activities at BVPS-2. A PRA evaluation was completed for the risk analysis of the BVPS 1 and BVPS 2 equipment hatch missile shield removal. This evaluation used the latest PRA model for Unit 2. This PRA model also utilized missile probability in the evaluation per UFSAR Section 3.3.2. It concluded that the probability of a tornado-generated missile hitting the exposed equipment hatch and causing damage is 4.20E-9.
The area of the unprotected equipment hatch used is 170 square feet.
The trench created by the work order had a smaller area - about 100 square feet. As such, the PRA evaluation bounded the proposed excavation activity and by comparison, plant operation proceeded since it was highly unlikely that core damage would occur as a result of the excavation.
FirstEnergy Nuclear Operating Company concluded the proposed activity could proceed without obtaining a license amendment.
Attachment L-12-349 Page 5 of 5 Evaluation No. : 12-03475, Revision 0 Reactor Coolant Pump Shutdown Seal Activity
Description:
The proposed activity was to replace the reactor coolant pump (RCP) number 1 seal inserts in the three Beaver Valley Power Station, Unit 2 (BVPS-2) RCPs with a modified design called the SHIELD Shutdown Seal (SDS). This included replacement of the existing number 1 runner retainer sleeve and retainer sleeve adapter with a shutdown seal sleeve and a shutdown seal sleeve adapter.
With one exception, all potentially affected UFSAR-described system, structure, or component design functions were screened-out. The RCP design function, to provide core cooling flow during normal operating conditions, was screened-in for further evaluation given it could be adversely impacted if the SDS were to inadvertently actuate.
The BVPS-2 updated safety analysis report was reviewed; changes were suggested. The BVPS-2 Technical Specifications and Bases were also reviewed; no changes were required. FirstEnergy Nuclear Operating Company (FENOC) concluded the propo$ed activity could proceed without obtaining a license amendment.
Summary of Evaluation:
The SDS was analyzed, evaluated and tested to the extent that installation and use of the SDS was acceptable.
NOTE: For the period of this report, only one of three RCPs were modified with the SDS.
The remaining two RCPs will be modified during future outages.