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Category:Annual Operating Report
MONTHYEARRS-24-034, Annual 10 CFR 50.46 Report of Emergency Core Cooling System Evaluation Model Changes and Errors2024-04-10010 April 2024 Annual 10 CFR 50.46 Report of Emergency Core Cooling System Evaluation Model Changes and Errors ML24071A1152024-03-31031 March 2024 U.S. Department of Energy, Office of Legacy Management, 2023 Annual Site Inspection and Monitoring Report for Uranium Mill Tailings Radiation Control Act Title I Disposal Sites BW220029, Annual Radiological Environmental Operating Report, 1 January Through 31 December 20212022-05-31031 May 2022 Annual Radiological Environmental Operating Report, 1 January Through 31 December 2021 ML22118A3642022-04-28028 April 2022 2021 Annual Radioactive Effluent Release Report (ARERR) RS-21-046, Annual 10 CFR 50.46 Report of Emergency Core Cooling System Evaluation Model Changes and Errors2021-04-0808 April 2021 Annual 10 CFR 50.46 Report of Emergency Core Cooling System Evaluation Model Changes and Errors BW200037, Annual Radiological Environmental Operating Report2020-05-31031 May 2020 Annual Radiological Environmental Operating Report ML20121A0302020-04-30030 April 2020 Annual Radiological Environmental Operating Report - Part 2 of 4 BYRON 2020-0031, Annual Radiological Environmental Operating Report - Part 4 of 42020-04-30030 April 2020 Annual Radiological Environmental Operating Report - Part 4 of 4 ML20121A0312020-04-30030 April 2020 Annual Radiological Environmental Operating Report - Part 3 of 4 ML20121A0292020-04-30030 April 2020 Annual Radiological Environmental Operating Report - Part 1 of 4 RS-20-043, 10 CFR 50.46 Report of Emergency Core Cooling System Evaluation Model Changes and Errors2020-04-0909 April 2020 10 CFR 50.46 Report of Emergency Core Cooling System Evaluation Model Changes and Errors ML19282A4032019-10-0909 October 2019 10 CFR 72.48 Evaluation Summary Report RS-19-042, Submittal of Annual 1O CFR 50.46 Report of Emergency Core Cooling System Evaluation Model Changes and Errors2019-04-0909 April 2019 Submittal of Annual 1O CFR 50.46 Report of Emergency Core Cooling System Evaluation Model Changes and Errors BYRON 2018-0042, Submittal of 2017 Annual Radiological Environmental Operating Report2018-05-0303 May 2018 Submittal of 2017 Annual Radiological Environmental Operating Report RS-18-039, Stations - Annual 10 CFR 50.46 Report of Emergency Core Cooling System Evaluation Model Changes and Errors2018-04-0909 April 2018 Stations - Annual 10 CFR 50.46 Report of Emergency Core Cooling System Evaluation Model Changes and Errors ML17132A3892017-05-12012 May 2017 2016 Annual Radiological Environmental Operating Report BYRON 2015-0054, 2014 Annual Radiological Environmental Operating Report (AREOR)2015-05-13013 May 2015 2014 Annual Radiological Environmental Operating Report (AREOR) RS-15-104, Annual Property Insurance Status Report2015-04-0101 April 2015 Annual Property Insurance Status Report BYRON 2014-0051, Annual Radioactive Effluent Release Report2014-04-24024 April 2014 Annual Radioactive Effluent Release Report ML13137A1052013-05-14014 May 2013 Submittal of 2012 Annual Radiological Environmental Operating Report ML13134A0492013-04-30030 April 2013 Submittal of 2012 Radioactive Effluent Release Report RS-13-090, and Byron Station - Annual 10 CFR 50.46 Report of Emergency Core Cooling System Evaluation Model Changes and Errors2013-04-0505 April 2013 and Byron Station - Annual 10 CFR 50.46 Report of Emergency Core Cooling System Evaluation Model Changes and Errors RS-14-029, Annual Report on the Meteorological Monitoring Program at the Braidwood Nuclear Power Station 20122012-12-31031 December 2012 Annual Report on the Meteorological Monitoring Program at the Braidwood Nuclear Power Station 2012 ML14030A2942012-12-31031 December 2012 Annual Report on the Meteorological Monitoring Program at the Braidwood Nuclear Power Station 2012 ML12342A3882012-12-17017 December 2012 Report of Changes, Tests and Experiments ML12137A3042012-05-14014 May 2012 Submittal of 2011 Annual Radiological Environmental Operating Report BYRON 2012-0051, Annual Dose Report for 20112012-04-27027 April 2012 Annual Dose Report for 2011 ML12130A1982012-04-26026 April 2012 2011 Radioactive Effluent Release Report RS-12-054, Submittal of Annual 10 CFR 50.46 Report of the Emergency Core Cooling System Evaluation Model Changes and Errors2012-04-0606 April 2012 Submittal of Annual 10 CFR 50.46 Report of the Emergency Core Cooling System Evaluation Model Changes and Errors ML14030A2932011-12-31031 December 2011 Annual Report on the Meteorological Monitoring Program at the Braidwood Nuclear Power Station 2011 RS-14-029, Annual Report on the Meteorological Monitoring Program at the Braidwood Nuclear Power Station 20112011-12-31031 December 2011 Annual Report on the Meteorological Monitoring Program at the Braidwood Nuclear Power Station 2011 BYRON 2011-0067, 2010 Annual Radioactive Effluent Release Report2011-04-29029 April 2011 2010 Annual Radioactive Effluent Release Report ML14030A2922011-04-22022 April 2011 Transmittal of Annual Report on the Meteorological Monitoring Program at the Braidwood Nuclear Power Station 2010 RS-14-029, Transmittal of Annual Report on the Meteorological Monitoring Program at the Braidwood Nuclear Power Station 20102011-04-22022 April 2011 Transmittal of Annual Report on the Meteorological Monitoring Program at the Braidwood Nuclear Power Station 2010 RS-11-052, Units 1 and 2 - Annual 10 CFR 50.46 Report of the Emergency Core Cooling System Evaluation Model Changes and Errors2011-04-0606 April 2011 Units 1 and 2 - Annual 10 CFR 50.46 Report of the Emergency Core Cooling System Evaluation Model Changes and Errors ML1013904752010-05-14014 May 2010 Annual Radiological Environmental Operating Report ML1012301402010-04-30030 April 2010 2009 Annual Radiological Effluent Release Report BYRON 2010-0029, Annual Radioactive Effluent Release Report2010-04-30030 April 2010 Annual Radioactive Effluent Release Report BYRON 2010-0034, Annual Dose Report for 20092010-04-30030 April 2010 Annual Dose Report for 2009 RS-10-075, Annual 10 CFR 50.46 Report of the Emergency Core Cooling System Evaluation Model Changes and Errors2010-04-0808 April 2010 Annual 10 CFR 50.46 Report of the Emergency Core Cooling System Evaluation Model Changes and Errors BYRON 2010-0017, Regulatory Commitment Change Summary Report2010-03-0101 March 2010 Regulatory Commitment Change Summary Report BYRON 2010-0039, Annual Radiological Environmental Operating Report, 1 January Through 31 December 20092009-12-31031 December 2009 Annual Radiological Environmental Operating Report, 1 January Through 31 December 2009 RS-09-051, Annual 10 CFR 50 .46 Report of the Emergency Core Cooling System Evaluation Model Changes and Errors2009-04-0909 April 2009 Annual 10 CFR 50 .46 Report of the Emergency Core Cooling System Evaluation Model Changes and Errors ML14030A2902008-12-31031 December 2008 Annual Report on the Meteorological Monitoring Program at the Braidwood Nuclear Power Station 2008 RS-14-029, Annual Report on the Meteorological Monitoring Program at the Braidwood Nuclear Power Station 20082008-12-31031 December 2008 Annual Report on the Meteorological Monitoring Program at the Braidwood Nuclear Power Station 2008 RS-08-053, Annual & Thirty Day Reports of the Emergency Core Cooling System Evaluation Model Changes & Errors Required by 10 CFR 50.46, Acceptance Criteria for Emergency Core Cooling Systems..2008-04-11011 April 2008 Annual & Thirty Day Reports of the Emergency Core Cooling System Evaluation Model Changes & Errors Required by 10 CFR 50.46, Acceptance Criteria for Emergency Core Cooling Systems.. RS-07-052, Annual Report of the Emergency Core Cooling System Evaluation Model Changes and Errors Required by 10 CFR 50.46, Acceptance Criteria for Emergency Core Cooling Systems for Light-Water Nuclear Power Reactors.2007-04-13013 April 2007 Annual Report of the Emergency Core Cooling System Evaluation Model Changes and Errors Required by 10 CFR 50.46, Acceptance Criteria for Emergency Core Cooling Systems for Light-Water Nuclear Power Reactors. ML0634902862006-12-13013 December 2006 Report of Changes, Tests and Experiments ML0614303732006-04-30030 April 2006 Annual Radiological Environmental Operating Report ML0435705032004-12-16016 December 2004 10 CFR 50.59 Biennial Report 2024-04-10
[Table view] Category:Letter type:RS
MONTHYEARRS-24-064, Nuclear Radiological Emergency Plan Document Revisions2024-06-28028 June 2024 Nuclear Radiological Emergency Plan Document Revisions RS-24-066, Inservice Testing Interval Extension for Essential Service Water Valves Due to Required Use of Containment Chiller2024-06-25025 June 2024 Inservice Testing Interval Extension for Essential Service Water Valves Due to Required Use of Containment Chiller RS-24-061, Constellation Energy Generation, LLC, Response to NRC Regulatory Issue Summary 2024-01, Preparation and Scheduling of Operator Licensing Examinations2024-06-14014 June 2024 Constellation Energy Generation, LLC, Response to NRC Regulatory Issue Summary 2024-01, Preparation and Scheduling of Operator Licensing Examinations RS-24-057, License Amendment to Technical Specification 3.7.9, Ultimate Heat Sink2024-06-0404 June 2024 License Amendment to Technical Specification 3.7.9, Ultimate Heat Sink RS-24-044, License Amendment to Transition to Framatome Gaia Fuel and Exemptions to 10 CFR 50.46 and 10 CFR 50 Appendix K. Attachments 1 to 11 Enclosed2024-05-28028 May 2024 License Amendment to Transition to Framatome Gaia Fuel and Exemptions to 10 CFR 50.46 and 10 CFR 50 Appendix K. Attachments 1 to 11 Enclosed RS-24-043, Application to Remove Power Distribution Monitoring System (Pdms) Details from Technical Specifications2024-05-24024 May 2024 Application to Remove Power Distribution Monitoring System (Pdms) Details from Technical Specifications RS-24-055, 2023 Corporate Regulatory Commitment Change Summary Report2024-05-17017 May 2024 2023 Corporate Regulatory Commitment Change Summary Report RS-24-041, Alternative Request to Utilize Code Case OMN-32, Alternative Requirements for Range and Accuracy of Pressure, Flow, and Differential Pressure Instruments Used in Pump Tests2024-04-30030 April 2024 Alternative Request to Utilize Code Case OMN-32, Alternative Requirements for Range and Accuracy of Pressure, Flow, and Differential Pressure Instruments Used in Pump Tests RS-24-026, License Amendment to Braidwood Station, Units 1 and 2, and Byron Station, Units 1 and 2, Technical Specifications 5.6.5, Core Operating Limits Report (COLR)2024-04-25025 April 2024 License Amendment to Braidwood Station, Units 1 and 2, and Byron Station, Units 1 and 2, Technical Specifications 5.6.5, Core Operating Limits Report (COLR) RS-24-034, Annual 10 CFR 50.46 Report of Emergency Core Cooling System Evaluation Model Changes and Errors2024-04-10010 April 2024 Annual 10 CFR 50.46 Report of Emergency Core Cooling System Evaluation Model Changes and Errors RS-24-002, Constellation Energy Generation, LLC - Annual Property Insurance Status Report2024-04-0101 April 2024 Constellation Energy Generation, LLC - Annual Property Insurance Status Report RS-24-024, Response to Request for Additional Information Regarding Proposed Alternative for Examination of Pressurizer Circumferential and Longitudinal Shell-to-Head Welds and Nozzle-to-Vessel Welds2024-03-22022 March 2024 Response to Request for Additional Information Regarding Proposed Alternative for Examination of Pressurizer Circumferential and Longitudinal Shell-to-Head Welds and Nozzle-to-Vessel Welds RS-24-013, Response to Request for Additional Information Regarding Proposed Alternative to Distribution Requirements of ASME Code Table IWC-2411-1 for the Steam Generators2024-02-13013 February 2024 Response to Request for Additional Information Regarding Proposed Alternative to Distribution Requirements of ASME Code Table IWC-2411-1 for the Steam Generators RS-24-004, Proposed Alternative to the Distribution Requirements of ASME Code Table IWC-2411-1 for the Steam Generators2024-01-11011 January 2024 Proposed Alternative to the Distribution Requirements of ASME Code Table IWC-2411-1 for the Steam Generators RS-23-118, Supplemental Information Letter for Part 73 Exemption Request - Responses to Request for Confirmatory Information2023-11-10010 November 2023 Supplemental Information Letter for Part 73 Exemption Request - Responses to Request for Confirmatory Information RS-23-117, Supplemental Information Letter for Part 73 Exemption Request - Responses to Request for Confirmatory Information2023-11-10010 November 2023 Supplemental Information Letter for Part 73 Exemption Request - Responses to Request for Confirmatory Information RS-23-114, Proposed Alternative for Examination of Pressurizer Circumferential and Longitudinal Shell-to-Head Welds and Nozzle-to-Vessel Welds2023-11-0101 November 2023 Proposed Alternative for Examination of Pressurizer Circumferential and Longitudinal Shell-to-Head Welds and Nozzle-to-Vessel Welds RS-23-100, Request for Exemption from Enhanced Weapons, Firearms Background Checks, and Security Event Notifications Implementation2023-10-13013 October 2023 Request for Exemption from Enhanced Weapons, Firearms Background Checks, and Security Event Notifications Implementation RS-23-097, Constellation Energy Generation, LLC, Advisement of Leadership Changes and Submittal of Updated Standard Practice Procedures Plans2023-10-12012 October 2023 Constellation Energy Generation, LLC, Advisement of Leadership Changes and Submittal of Updated Standard Practice Procedures Plans RS-23-108, Proposed Alternative for Examinations of Examination Categories B-B, B-D, and C-A Steam Generator Pressure Retaining Welds and Full Penetration Welded Nozzles2023-10-11011 October 2023 Proposed Alternative for Examinations of Examination Categories B-B, B-D, and C-A Steam Generator Pressure Retaining Welds and Full Penetration Welded Nozzles RS-23-105, Proposed Alternative for Examinations of Examination Category C-B Steam Generator Nozzle-to-Shell Welds and Nozzle Inside Radius Sections2023-10-10010 October 2023 Proposed Alternative for Examinations of Examination Category C-B Steam Generator Nozzle-to-Shell Welds and Nozzle Inside Radius Sections RS-23-093, License Amendment to Braidwood Station, Units 1 and 2, and Byron Station, Units 1 and 2, Technical Specifications 3.7.15, Spent Fuel Pool Boron Concentration, 3.7.16, Spent Fuel.2023-09-29029 September 2023 License Amendment to Braidwood Station, Units 1 and 2, and Byron Station, Units 1 and 2, Technical Specifications 3.7.15, Spent Fuel Pool Boron Concentration, 3.7.16, Spent Fuel. RS-23-094, Relief Request I4R-24, Alternative for Post-Peening Reexamination Frequency for Reactor Pressure Vessel Head Penetration Nozzle Number 752023-09-29029 September 2023 Relief Request I4R-24, Alternative for Post-Peening Reexamination Frequency for Reactor Pressure Vessel Head Penetration Nozzle Number 75 RS-23-091, Relief Request I4R-25, Alternative Requirements for Reactor Pressure Vessel Inservice Inspection Intervals2023-09-26026 September 2023 Relief Request I4R-25, Alternative Requirements for Reactor Pressure Vessel Inservice Inspection Intervals RS-23-083, Withdrawal - Proposed Alternatives Related to the Steam Generators2023-06-27027 June 2023 Withdrawal - Proposed Alternatives Related to the Steam Generators RS-23-077, Response to NRC Regulatory Issue Summary 2023-01, Preparation and Scheduling of Operator Licensing Examinations2023-06-16016 June 2023 Response to NRC Regulatory Issue Summary 2023-01, Preparation and Scheduling of Operator Licensing Examinations RS-23-074, Supplement to Application for License Amendment to Braidwood Station, Units 1 and 2, Technical Specification 3.7.9, Ultimate Heat Sink2023-06-0909 June 2023 Supplement to Application for License Amendment to Braidwood Station, Units 1 and 2, Technical Specification 3.7.9, Ultimate Heat Sink RS-23-075, Application for Technical Specification Improvement to Extend the Completion Time for Condition B of Technical Specification 3.5.1, Accumulators, Using the Consolidated Line Item Improvement Process2023-06-0707 June 2023 Application for Technical Specification Improvement to Extend the Completion Time for Condition B of Technical Specification 3.5.1, Accumulators, Using the Consolidated Line Item Improvement Process RS-23-050, Supplement to Application to Revise Technical Specifications to Adopt TSTF-577, Revised Frequencies for Steam Generator Tube.2023-05-22022 May 2023 Supplement to Application to Revise Technical Specifications to Adopt TSTF-577, Revised Frequencies for Steam Generator Tube. RS-23-056, Response to Request for Additional Information to Braidwood Station, Unit 1, and Byron Station, Unit 1, for Steam Generator License Renewal Response to Commitment 102023-04-20020 April 2023 Response to Request for Additional Information to Braidwood Station, Unit 1, and Byron Station, Unit 1, for Steam Generator License Renewal Response to Commitment 10 RS-23-055, Annual 10 CFR 50.46 Report of Emergency Core Cooling System Evaluation Model Changes and Errors2023-04-10010 April 2023 Annual 10 CFR 50.46 Report of Emergency Core Cooling System Evaluation Model Changes and Errors RS-23-052, License Amendment to Braidwood Station, Units 1 and 2, Technical Specification 3.7.9, Ultimate Heat Sink2023-03-24024 March 2023 License Amendment to Braidwood Station, Units 1 and 2, Technical Specification 3.7.9, Ultimate Heat Sink RS-23-049, Constellation Energy Generation, LLC, Report on Status of Decommissioning Funding for Reactors and Independent Spent Fuel Storage Installations2023-03-23023 March 2023 Constellation Energy Generation, LLC, Report on Status of Decommissioning Funding for Reactors and Independent Spent Fuel Storage Installations RS-23-045, Constellation Energy Generation, LLC Submittal of Fitness for Duty Performance Data Reports for 2022 Per 10 CFR 26.717(c) & 10 CFR 26.2032023-02-28028 February 2023 Constellation Energy Generation, LLC Submittal of Fitness for Duty Performance Data Reports for 2022 Per 10 CFR 26.717(c) & 10 CFR 26.203 RS-23-038, Response to Request for Additional Information for the Byron Proposal to Reinsert an Accident Tolerant Fuel Lead Test Assembly2023-02-27027 February 2023 Response to Request for Additional Information for the Byron Proposal to Reinsert an Accident Tolerant Fuel Lead Test Assembly RS-23-040, Constellation Energy Generation, LLC, Supplemental Information - Proposed Alternatives Related to the Steam Generators2023-02-21021 February 2023 Constellation Energy Generation, LLC, Supplemental Information - Proposed Alternatives Related to the Steam Generators RS-23-003, Constellation Energy Generation, LLC, Summary of Changes to Quality Assurance Topical Report, NO-AA-10, and Decommissioning Quality Assurance Program, NO-DC-102023-01-31031 January 2023 Constellation Energy Generation, LLC, Summary of Changes to Quality Assurance Topical Report, NO-AA-10, and Decommissioning Quality Assurance Program, NO-DC-10 RS-22-123, Response to Request for Additional Information Regarding Application to Revise Technical Specifications to Adopt TSTF-577, Revised Frequencies for Steam Generator2022-12-0707 December 2022 Response to Request for Additional Information Regarding Application to Revise Technical Specifications to Adopt TSTF-577, Revised Frequencies for Steam Generator RS-22-126, Constellation Energy Generation, LLC - Request to Use Provisions of a Later Edition of the ASME Boiler and Pressure Vessel Code, Section XI2022-11-30030 November 2022 Constellation Energy Generation, LLC - Request to Use Provisions of a Later Edition of the ASME Boiler and Pressure Vessel Code, Section XI RS-22-118, Withdrawal of License Amendment Request to Revise Braidwood Station, Units 1 and 2, Technical Specification 3.7.9, Ultimate Heat Sink2022-10-31031 October 2022 Withdrawal of License Amendment Request to Revise Braidwood Station, Units 1 and 2, Technical Specification 3.7.9, Ultimate Heat Sink RS-22-110, Supplemental Information - Proposed Alternatives Related to the Steam Generators and Request.2022-09-20020 September 2022 Supplemental Information - Proposed Alternatives Related to the Steam Generators and Request. RS-22-097, License Amendment Request to Reinsert an Accident Tolerant Fuel Lead Test Assembly2022-08-31031 August 2022 License Amendment Request to Reinsert an Accident Tolerant Fuel Lead Test Assembly RS-22-093, Advisement of Leadership Changes for Constellation Energy Generation, LLC and Submittal of Updated Standard Practice Procedures Plans2022-08-18018 August 2022 Advisement of Leadership Changes for Constellation Energy Generation, LLC and Submittal of Updated Standard Practice Procedures Plans RS-22-086, R. E. Ginna Nuclear Power Plant - Application to Revise Technical Specifications to Adopt TSTF-577, Revised Frequencies for Steam.2022-08-10010 August 2022 R. E. Ginna Nuclear Power Plant - Application to Revise Technical Specifications to Adopt TSTF-577, Revised Frequencies for Steam. RS-22-084, Response to Request for Additional Information Proposed Alternative for Examinations of Examination Categories B-B, B-D, and C-A Steam Generator .2022-06-17017 June 2022 Response to Request for Additional Information Proposed Alternative for Examinations of Examination Categories B-B, B-D, and C-A Steam Generator . RS-22-072, Response to Request for Additional Information Regarding License Amendment to Revise Braidwood Station, Units 1 and 2, Technical Specification 3.7.9, Ultimate Heat Sink2022-06-0606 June 2022 Response to Request for Additional Information Regarding License Amendment to Revise Braidwood Station, Units 1 and 2, Technical Specification 3.7.9, Ultimate Heat Sink RS-22-075, License Amendment to Braidwood Station, Units 1 and 2, Technical Specification 3.7.9, Ultimate Heat Sink2022-06-0202 June 2022 License Amendment to Braidwood Station, Units 1 and 2, Technical Specification 3.7.9, Ultimate Heat Sink RS-22-074, Response to Request for Additional Information - Proposed Alternative for Examinations of Examination Category C-B Steam Generator Nozzle-to-Shell Welds .2022-05-20020 May 2022 Response to Request for Additional Information - Proposed Alternative for Examinations of Examination Category C-B Steam Generator Nozzle-to-Shell Welds . RS-22-073, Supplemental Information Regarding License Amendment Request -Adoption of Technical Specification Task Force (TSTF) Traveler TSTF-501, Revision 1, Relocate Stored Fuel Oil and Lube.2022-05-19019 May 2022 Supplemental Information Regarding License Amendment Request -Adoption of Technical Specification Task Force (TSTF) Traveler TSTF-501, Revision 1, Relocate Stored Fuel Oil and Lube. RS-22-037, License Amendment Request - Adoption of Technical Specification Task Force (TSTF) Traveler TSTF-501, Revision 1, Relocate Stored Fuel Oil and Lube Oil Volume Values to Licensee Control2022-04-21021 April 2022 License Amendment Request - Adoption of Technical Specification Task Force (TSTF) Traveler TSTF-501, Revision 1, Relocate Stored Fuel Oil and Lube Oil Volume Values to Licensee Control 2024-06-04
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n Nuclear RS-11-0S2 10 CFR 50.46 April 6, 2011 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 Braidwood Station, Units 1 and 2 Facility Operating License Nos. NPF-72 and NPF-77 NRC Docket Nos. STN 50-456 and STN 50-457 Byron Station, Units 1 and 2 Facility Operating License Nos. NPF-37 and NPF-66 NRC Docket Nos. STN 50-454 and STN 50-455
Subject:
Annual 10 CFR 50.46 Report of the Emergency Core Cooling System Evaluation Model Changes and Errors
Reference:
- 1. Letter from J. L. Hansen (Exelon Generation Company, LLC) to U. S. NRC, "ECCS Evaluation Model Error - 10 CFR 50.46 Reports,"
dated March 15,2011.
In accordance with 10 CFR 50.46, "Acceptance criteria for emergency core cooling systems for light-water nuclear power reactors," paragraph (a)(3)(ii), Exelon Generation Company, LLC, (EGC) is submitting the attached information to fulfill the annual reporting requirements for the Braidwood and Byron Stations, Units 1 and 2.
In Reference 1, EGC submitted a report to fulfill the 30-day reporting requirement in accordance with 10 CFR 50.46 for the new large break Best-Estimate LOCA (BELOCA) Automated Statistical Treatment of Uncertainty Method (ASTRUM) analysis in support of Braidwood and Byron Stations, Units 1 and 2. Updated assessment notes were provided with the PCT values for the limiting large break loss-of-coolant accident (LBLOCA) evaluations regarding the new BELOCA ASTRUM analysis. The small break loss-of-coolant accident (SBLOCA) analyses were not affected. provides peak cladding temperature (PCT) information for the limiting loss-of-coolant accident (LOCA) evaluations for the Braidwood and Byron Stations. Attachment 2 contains the assessment notes, which provides a detailed description for each change reported.
The Attachments indicate that there has been no net PCT change for the Braidwood and Byron Stations, Units 1 and 2.
April 6, 2010 U. S. Nuclear Regulatory Commission Page 2 There are no regulatory commitments contained in this letter. Should you have any questions concerning this letter, please contact Ms. Lisa A. Schofield at (630) 657-2815 .
. Hansen Manager - Licensing Exelon Generation Company, LLC Attachments:
1: Braidwood and Byron Stations, Units 1 and 2 - 10 CFR 50.46 Report 2: Braidwood and Byron Stations, Units 1 and 2 - 10 CFR 50.46 Report Assessment Notes cc:
NRC Regional Administrator, Region III NRC Senior Resident Inspector, Braidwood Station NRC Senior Resident Inspector, Byron Station NRR Project Manager, Braidwood and Byron Stations Illinois Emergency Management Agency - Division of Nuclear Safety
ATTACHMENT 1 Braidwood and Byron Stations, Units 1 and 2 10 CFR 50.46 Report PLANT NAME: Braidwood Station, Unit 1 ECCS EVALUATION MODEL: Small Break Loss-of-Coolant Accident (SBLOCA)
REPORT REVISION DATE: January 24, 2011 CURRENT OPERATING CYCLE: A1C16 ANALYSIS OF RECORD Evaluation Model: NOTRUMP Calculation: Westinghouse CN-US-00-208, December 2000 Fuel: VANTAGE+ 17 x 17 Limiting Fuel Type: VANTAGE+ 17 x 17 Limiting Single Failure: Loss of one train of ECCS flow Limiting Break Size and Location: 211 break in the bottom of the Cold Leg Reference Peak Cladding Temperature (PCT): 1624.0 of MARGIN ALLOCATION A. PRIOR LOSS..OF-COOlANT ACCIDENT (LOCA) MODEL ASSESSMENTS 10 CFR 50.46 report dated June 11, 2001 (see note 1) ~PCT = 0 of 10 CFR 50.46 report dated April 18, 2002 (see note 2) ~PCT= 0 of 10 CFR 50.46 report dated April 14, 2003 (see note 3) ~PCT= 0 of 10 CFR 50.46 report dated April 14, 2004 (see note 4) ~PCT = 35 of 10 CFR 50.46 report dated April 14, 2005 (see note 5) ~PCT= 0 of 10 CFR 50.46 report dated April 14, 2006 (see note 6) ~PCT = 0 of 10 CFR 50.46 report dated April 13, 2007 (see note 7) ~PCT= 0 of 10 CFR 50.46 report dated June 22, 2007 (see note 9) ~PCT = 0 of 10 CFR 50.46 report dated November 19, 2007 (see note 10) ~PCT = 90 of 10 CFR 50.46 report dated April 11 ,2008 (see note 11) ~PCT= 0 of 10 CFR 50.46 report dated April 9, 2009 (see note 12) ~PCT = 0 of 10 CFR 50.46 report dated April 8, 2010 (see note 13) ~PCT = 0 of Net PCT PCT = 1749.0 of B. CURRENT LOCA MODEL ASSESSMENTS Urania-Gadolinia Pellet Thermal Conductivity Calculation (see note 15) ~PCT = 0 of Pellet Crack and Dish Volume Calculation (see note 16) ~PCT = 0 of Treatment of Vessel Average Temperature Uncertainty (see note 17) ~PCT = 0 of General Code Maintenance (see note 18) ~PCT = 0 of Total PCT change from current assessments L ~PCT = 0 of Cumulative PCT change from current assessments L I~PCTI = 0 OF Net PCT PCT =1749.0 of Page 1 of 8
ATTACHMENT 1 Braidwood and Byron Stations, Units 1 and 2 10 CFR 50.46 Report PLANT NAME: Braidwood Station, Unit 1 ECCS EVALUATION MODEL: Large Break LossMof-Coolant Accident (LBLOCA)
REPORT REVISION DATE: January 24, 2011 CURRENT OPERATING CYCLE: A1C16 ANALYSIS OF RECORD Evaluation Model: ASTRUM (2004)
Calculation: Westinghouse WCAp M16841-P, November 2007 Fuel: VANTAGE+ 17 x 17 Limiting Fuel Type: VANTAGE+ 17 x 17 Limiting Single Failure: Loss of one train of ECCS flow Limiting Break Size and Location: Guillotine break in the Cold Leg Reference PCT: 1913.0 of MARGIN ALLOCATION A. PRIOR LOCA MODEL ASSESSMENTS I 10 CFR 50.46 report dated March 15, 2011 (see note 14)
Net PCT PCT = 1913.0 of B. CURRENT LOCA MODEL ASSESSMENTS No 2010 model changes, corrections, or enhancements ~PCT = 0 of Total PCT change from current assessments L ~PCT = 0 OF Cumulative peT change from current assessments L I,1PCT 1= 0 OF Net PCT PCT =1913.0 of Page 2 of 8
ATTACHMENT 1 Braidwood and Byron Stations, Units 1 and 2 10 CFR 50.46 Report PLANT NAME: Braidwood Station, Unit 2 ECCS EVALUATION MODEL: SBLOCA REPORT REVISION DATE: January 24,2011 CURRENT OPERATING CYCLE: A2C15 ANALYSIS OF RECORD Evaluation Model: NOTRUMP Calculation: Westinghouse CN-LiS-00-208, December 2000 Fuel: VANTAGE+ 17 x 17 Limiting Fuel Type: VANTAGE+ 17 x 17 Limiting Single Failure: Loss of one train of ECCS flow Limiting Break Size and Location: 2" break in the bottom of the Cold Leg Reference PCT: 1627.0 of MARGIN ALLOCATION A. PRIOR LOCA MODEL ASSESSMENTS 10 CFR 50.46 report dated June 11, 2001 (see note 1) ~PCT = 3 of 10 CFR 50.46 report dated April 18, 2002 (see note 2) ~PCT = 0 of 10 CFR 50.46 report dated April 14, 2003 (see note 3) ~PCT = 0 of 10 CFR 50.46 report dated April 14, 2004 (see note 4) ~PCT = 35 of 10 CFR 50.46 report dated April 14, 2005 (see note 5) ~PCT= 0 of 10 CFR 50.46 report dated April 14, 2006 (see note 6) ~PCT = 0 of 10 CFR 50.46 report dated April 13, 2007 (see note 7) ~PCT= 0 of 10 CFR 50.46 report dated June 22, 2007 (see note 9) ~PCT = 0 of 10 CFR 50.46 report dated April 11 ,2008 (see note 11) ~PCT = 90 of 10 CFR 50.46 report dated April 9, 2009 (see note 12) ~PCT = 0 of 10 CFR 50.46 report dated April 8, 2010 (see note 13) ~PCT = 0 OF Net PCT PCT =1755.0 OF B. CURRENT LOCA MODEL ASSESSMENTS Pellet Crack and Dish Volume Calculation (see note 16) ~PCT = 0 OF Treatment of Vessel Average Temperature Uncertainty ~PCT = 0 OF (see note 17)
General Code Maintenance (see note 18) ~PCT = 0 OF Total PCT change from current assessments L ~PCT = 0 OF Cumulative PCT change from current assessments L I~PCT 1= 0 OF Net PCT PCT = 1755.0 OF Page 3 of 8
ATIACHMENT1 Braidwood and Byron Stations, Units 1 and 2 10 CFR 50.46 Report PLANT NAME: Braidwood Station, Unit 2 ECCS EVALUATION MODEL: LBLOCA REPORT REVISION DATE: January 24,2011 CURRENT OPERATING CYCLE: A2C15 ANALYSIS OF RECORD Evaluation Model: ASTRUM (2004)
Calcu lation: Westinghouse WCAP 16841 P, November 2007 M M Fuel: VANTAGE+ 17 x 17 Limiting Fuel Type: VANTAGE+ 17 x 17 Limiting Single Failure: Loss of one train of ECCS flow Limiting Break Size and Location: Guillotine break in the Cold Leg Reference PCT: 2041.0 of MARGIN ALLOCATION A. PRIOR LOCA MODEL ASSESSMENTS I 10 CFR 50.46 report dated March 15, 2011 (see note 14)
Net PCT PCT = 2041.0 of B. CURRENT LOCA MODEL ASSESSMENTS No 2010 model changes, corrections, or enhancements ~PCT = 0 of Total PCT change from current assessments L 6PCT = 0 OF Cumulative PCT change from current assessments L I~PCT 1= 0 OF Net peT PCT = 2041.0 of Page 4 of 8
ATTACHMENT 1 Braidwood and Byron Stations, Units 1 and 2 10 CFR 50.46 Report PLANT NAME: Byron Station, Unit 1 ECCS EVALUATION MODEL: SBLOCA REPORT REVISION DATE: January 24,2011 CURRENT OPERATING CYCLE: B1C18*
- Cycle 18 scheduled to start April 2011 ANAL VSIS OF RECORD Evaluation Model: NOTRUMP Calculation: Westinghouse CN-L\S-00-208, December 2000 Fuel: VANTAGE+ 17 x 17 Limiting Fuel Type: VANTAGE+ 17 x 17 Limiting Single Failure: Loss of one train of ECCS flow Limiting Break Size and Location: 2" break in the bottom of the Cold Leg Reference PCT: 1624.0 of MARGIN ALLOCATION A. PRIOR LOCA MODEL ASSESSMENTS 10 CFR 50.46 report dated June 11, 2001 (see note 1) ~PCT= O°F 10 CFR 50.46 report dated April 18, 2002 (see note 2) ~PCT = 0 of 10 CFR 50.46 report dated April 14, 2003 (see note 3) ~PCT = 0 of 10 CFR 50.46 report dated April 14, 2004 (see note 4) ~PCT = 35 of 10 CFR 50.46 report dated April 14, 2005 (see note 5) ~PCT = 0 of 10 CFR 50.46 report dated April 14, 2006 (see note 6) ~PCT = 0 of 10 CFR 50.46 report dated April 13, 2007 (see note 7) ~PCT= O°F 10 CFR 50.46 report dated June 22, 2007 (see note 9) ~PCT = 0 of 10 CFR 50.46 report dated April 11, 2008 (see note 11) ~PCT = 90 OF 10 CFR 50.46 report dated April 9, 2009 (see note 12) ~PCT = 0 OF 10 CFR 50.46 report dated April 8, 2010 (see note 13) ~PCT = 0 OF Net PCT PCT =1749.0 OF B. CURRENT LOCA MODEL ASSESSMENTS Pellet Crack and Dish Volume Calculation (see note 16) ~PCT = 0 OF Treatment of Vessel Average Temperature Uncertainty ~PCT = 0 OF (see note 17)
General Code Maintenance (see note 18) ~PCT = 0 OF Total PCT change from current assessments L ~PCT = 0 OF Cumulative PCT change from current assessments L I~PCT 1= 0 OF Net PCT PCT = 1749.0 OF Page 5 of 8
ATTACHMENT 1 Braidwood and Byron Stations, Units 1 and 2 10 CFR 50.46 Report PLANT NAME: Byron Station, Unit 1 ECCS EVALUATION MODEL: LBLOCA REPORT REVISION DATE: January 24,2011 CURRENT OPERATING CYCLE: B1C18*
- Cycle 18 scheduled to start April 2011 ANALYSIS OF RECORD Evaluation Model: ASTRUM (2004)
Calculation: Westinghouse WCAP-16841, November 2007 Fuel: VANTAGE+ 17 x 17 Limiting Fuel Type: VANTAGE+ 17 x 17 Limiting Single Failure: Loss of one train of ECCS flow Limiting Break Size and Location: Guillotine break in the Cold Leg Reference PCT: 1913.0 of MARGIN ALLOCATION A. PRIOR LOCA MODEL ASSESSMENTS
\10 CFR 50.46 report dated March 15, 2011 (see note 14)
Net PCT PCT = 1913.0 of B. CURRENT LOCA MODEL ASSESSMENTS No 2010 model changes, corrections, or enhancements LlPCT = 0 of Total PCT change from current assessments L LlPCT = 0 of Cumulative peT change from current assessments L ILlPCT 1= 0 OF Net PCT PCT =1913.0 of Page 6 of 8
ATIACHMENT1 Braidwood and Byron Stations, Units 1 and 2 10 CFR 50.46 Report PLANT NAME: Byron Station, Unit 2 ECCS EVALUATION MODEL: SBLOCA REPORT REVISION DATE: January 24,2011 CURRENT OPERATING CYCLE: B2C16 ANALYSIS OF RECORD Evaluation Model: NOTRUMP Calculation: Westinghouse CN-US-00-208, December 2000 Fuel: VANTAGE+ 17 x 17 Limiting Fuel Type: VANTAGE+ 17 x 17 Limiting Single Failure: Loss of one train of ECCS flow Limiting Break Size and Location: 211 break in the bottom of the Cold Leg Reference PCT:
MARGIN ALLOCATION A. PRIOR LOCA MODEL ASSESSMENTS 10 CFR 50.46 report dated June 11, 2001 (see note 1) ~PCT = 3 of 10 CFR 50.46 report dated April 18, 2002 (see note 2) ~PCT = 0 of 10 CFR 50.46 report dated April 14, 2003 (see note 3) ~PCT = 0 of 10 CFR 50.46 report dated April 14, 2004 (see note 4) ~PCT = 35 of 10 CFR 50.46 report dated April 14, 2005 (see note 5) ~PCT= 0 of 10 CFR 50.46 report dated April 14,2006 (see note 6) ~PCT = 0 of 10 CFR 50.46 report dated April 13, 2007 (see note 7) ~PCT = 0 of 10 CFR 50.46 report dated May 10, 2007 (see note 8) ~PCT = 90 of 10 CFR 50.46 report dated June 22, 2007 (see note 9) ~PCT = 0 of 10 CFR 50.46 report dated April 11 ,2008 (see note 11) ~PCT = 0 of 10 CFR 50.46 report dated April 9, 2009 (see note 12) ~PCT = 0 of 10 CFR 50.46 report dated April 8, 2010 (see note 13) ~PCT = 0 of Net PCT PCT =1755.0 of B. CURRENT LOCA MODEL ASSESSMENTS Pellet Crack and Dish Volume Calculation (see note 16) ~PCT= O°F Treatment of Vessel Average Temperature Uncertainty ~PCT = 0 of (see note 17)
General Code Maintenance (see note 18) ~PCT = 0 of Total PCT change from current assessments L ~PCT = 0 OF Cumulative PCT change from current assessments L I~PCT 1= 0 of Net PCT PCT = 1755.0 of Page 7 of 8
ATTACHMENT 1 Braidwood and Byron Stations, Units 1 and 2 10 CFR 50.46 Report PLANT NAME: Byron Station, Unit 2 ECCS EVALUATION MODEL: LBLOCA REPORT REVISION DATE: January 24,2011 CURRENT OPERATING CYCLE: B2C16 ANALYSIS OF RECORD Evaluation Model: ASTRUM (2004)
Calculation: Westinghouse WCAP-16841-P, November 2007 Fuel: VANTAGE+ 17 x 17 Limiting Fuel Type: VANTAGE+ 17 x 17 Limiting Single Failure: Loss of one train of ECCS flow Limiting Break Size and Location: Guillotine break in the Cold Leg Reference PCT: 2041.0 of MARGIN ALLOCATION A. PRIOR LOCA MODEL ASSESSMENTS 110 CFR 50.46 report dated March 15, 2011 (see note 14)
Net PCT PCT =2041.0 of B. CURRENT LOCA MODEL ASSESSMENTS No 2010 model changes, corrections, or enhancements APCT= 0 OF Total PCT change from current assessments L APCT = 0 of Cumulative PCT change from current assessments L IAPCT 1= 0 OF Net PCT PCT =2041.0 of Page 8 of 8
ATTACHMENT 2 Braidwood and Byron Stations, Units 1 and 2 10 CFR 50.46 Report Assessment Notes
- 1. Prior Loss-of-Coolant Accident (LOCA) Model Assessment The 10 CFR 50.46 report dated June 11, 2001, reported a new small break loss-of-coolant accident (SBLOCA) analysis to support operations at uprated power conditions. The same report assessed the impact from annular axial blankets on SBLOCA analysis, which determined a 0 OF PCT penalty each for Braidwood Unit 1 and Byron Unit 1 and resulted in a 3 of PCT penalty each for Braidwood Unit 2 and Byron Unit 2. Evaluations for plant conditions and SBLOCA model changes that resulted in 0 OF PCT change were reported.
- 2. Prior LOCA Model Assessment The 10 CFR 50.46 report dated April 18, 2002, reported evaluations for SBLOCA model changes, which resulted in 0 of PCT change.
- 3. Prior LOCA Model Assessment The 10 CFR 50.46 report dated April 14, 2003, reported evaluations for SBLOCA model changes, which resulted in 0 OF PCT change.
- 4. Prior LOCA Model Assessment The 10 CFR 50.46 report dated April 14, 2004, reported evaluations for an SBLOCA assessment related to NOTRUMP bubble rise/drift flux model inconsistency corrections, which resulted in 35 of PCT assessment.
- 5. Prior LOCA Model Assessment The 10 CFR 50.46 report dated April 14, 2005, reported evaluations for SBLOCA model changes, which resulted in a 0 OF PCT change. Assembly N10S was reconstituted with two stainless steel filler rods during Braidwood Unit 1 Refueling Outage 11. This assembly was reloaded into the core and was used during Braidwood Unit 1 Cycle 12 operation. The introduction of up to five stainless steel filler rods was evaluated and shown to have no impact on SBLOCA analysis. The estimated PCT effect was 0 of. This assembly was discharged during Reload 12.
- 6. Prior LOCA Model Assessment The 10 CFR 50.46 report dated April 14, 2006, reported evaluations for SBLOCA NOTRUMP General Code Maintenance, which resulted in 0 of PCT change.
- 7. Prior LOCA Model Assessment The 10 CFR 50.46 report dated April 13, 2007, reported evaluations for SBLOCA model changes and errors. The report documented general code maintenance for NOTRUMP, AXIOM lead test assembly evaluation and NOTRUMP refined break spectrum, which resulted in 0° F PCT impact.
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ATTACHMENT 2 Braidwood and Byron Stations, Units 1 and 2 10 CFR 50.46 Report Assessment Notes
- 8. Prior LOCA Model Assessment The 30-day 10 CFR 50.46 report dated May 10, 2007, applicable to Byron Unit 2 reported an assessment of the Emergency Core Cooling System (ECCS), which evaluated changes in ECCS flow during the recirculation phase due to Generic Safety Issue (GSI) 191 related safety injection (SI) throttle valve replacements. The evaluation of recirculation phase ECCS flow changes relative to impact on the current Analysis of Record was performed for the SBLOCA. Based on the NOTRUMP and SBLOCA calculations performed for Byron Unit 2, a conservative, bounding PCT assessment of +90 of was applied to the current Byron Unit 2 SBLOCA PCT.
- 9. Prior LOCA Model Assessment A 30-day report was submitted to the NRC dated June 22,2007, to report an error in the HOT SPOT Gode, which did not impact the SBLOCA analysis. The 30-day report also reported minor errors with the reactor vessel data collections that potentially affected the vessel inlet and outlet fluid volume, metal mass, and surface area. The corrected values were evaluated for impact, and a 0 of penalty was assessed for the SBLOGA analysis for the Braidwood and Byron Stations, Units 1 and 2.
- 10. Prior LOGA Model Assessment The 30-day 10 CFR 50.46 report dated November 19, 2007, applicable to Braidwood Unit 1 reported an assessment of the EGGS, which evaluated changes in EGGS flow during the recirculation phase due to GSI-191 related SI throttle valve replacements. The evaluation of recirculation phase EGGS flow changes relative to impact on the current Analysis of Record was performed for the SBLOGA. Based on the NOTRUMP and SBLOCA calculations performed for Braidwood Unit 1, a conservative, bounding PCT assessment of +90 of was applied to the Braidwood Unit 1 SBLOGA PCT.
11 . Prior LOCA Model Assessment The 10 CFR 50.46 report dated April 11, 2008, reported evaluations for LOCA model changes and errors. Applicable to Braidwood Unit 2 and Byron Unit 1, the ECCS assessment evaluated changes in ECCS flow during the recirculation phase due to GSI-191 related SI throttle valve replacements. A conservative, bounding PCT assessment of +90 of was applied to the Braidwood Unit 2 and Byron Unit 1 SBLOCA PCTs. The report also documented general code maintenance for SBLOCA and evaluation for pump weir resistance modeling for SBLOCA analysis, which resulted in 0 of PCT impact.
- 12. Prior LOCA Model Assessment The 10 CFR 50.46 report dated April 9, 2009, reported evaluations for LOCA model changes and errors. The report documents general code maintenance for SBLOCA, errors in reactor vessel lower plenum surface area calculations, discrepancies in metal mass from drawings, and an evaluation of Areva Lead Use Assemblies (LUA). All have a 0 of PGT penalty associated with them.
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ATTACHMENT 2 Braidwood and Byron Stations, Units 1 and 2 10 CFR 50.46 Report Assessment Notes
- 13. Prior LOCA Model Assessment The 10 CFR 50.46 report dated April 8, 2010, reported no evaluations for SBLOCA model changes, which resulted in 0 of PCT change.
- 14. Prior LOCA Model Assessment The 30-day 10 CFR 50.46 report dated March 15, 2011, reported a new large break BELOCA (ASTRUM) analysis to support operations for the Braidwood and Byron Stations, Units 1 and 2. The same report assessed the impact from several errors, issues, and code enhancements. Each of these errors / issues / code enhancements resulted in 0 of PCT impact with a net 0 of PCT impact.
- 15. Urania-Gadolinia Pellet Thermal Conductivity Calculation (SBLOCA)
Two errors were discovered in the pellet thermal conductivity calculation for urania-gadolinia pellets in the SBLOCA code. First, the calculation did not include the terms required to adjust for pellet densities other than 95% of the theoretical density. Second, the conversion from Fahrenheit to Rankine used an adder of 459 instead of 459.67. These errors have been corrected and were evaluated for impact on SBLOCA analysis for Braidwood Station, Unit 1; a PCT assessment of 0 OF was applied to the Braidwood Station, Unit 1.
- 16. Pellet Crack and Dish Volume Calculation (SBLOCA)
Two errors were discovered in the calculation of the normalized peJlet crack and dish volumes in the SBLOCA code. First, an incorrect operator was used to select between two tables of normalized volume vs. linear heat generation rate. Second, the normalized volume at 18 kW/ft was incorrectly programmed in one of the tables as 1.58 instead of 1.59. These errors were evaluated for impact on existing SBLOCA analysis; based on a combination of SBLOCA sensitivity calculations and engineering judgment, a 0 of PCT penalty was assessed for the Braidwood and Byron Stations, Units 1 and 2.
- 17. Treatment of Vessel Average Temperature Uncertainty (SBLOCA)
Historically, the overall vessel average temperature uncertainty calculated by Westinghouse considered only II_II instrument uncertainties, corresponding to the indicated temperature being lower than the actual temperature. This uncertainty was then applied as a "+1-"
uncertainty in some LOCA analyses, rather than using specific "+11 and "_" uncertainties.
This discrepancy was evaluated for impact on the SBLOCA analysis results, and a 0 of PCT penalty was assessed for the Braidwood and Byron Stations, Units 1 and 2.
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ATTACHMENT 2 Braidwood and Byron Stations, Units 1 and 2 10 CFR 50.46 Report Assessment Notes
- 18. General Code Maintenance (SBLOCA)
Various changes have been made to enhance the usability of the codes and to help preclude errors in analyses. This includes items such as modifying input variable definitions, units, and defaults; improving the input diagnostic checks; enhancing the code output; optimizing active coding; and eliminating inactive coding. These changes resulted in 0 of PCT impact with a net 0 OF PCT impact.
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