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Category:Annual Operating Report
MONTHYEARBW220029, 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 ML20121A0292020-04-30030 April 2020 Annual Radiological Environmental Operating Report - Part 1 of 4 ML20121A0302020-04-30030 April 2020 Annual Radiological Environmental Operating Report - Part 2 of 4 ML20121A0312020-04-30030 April 2020 Annual Radiological Environmental Operating Report - Part 3 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 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 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 ML14030A2902008-12-31031 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 ML0413902102004-05-14014 May 2004 2003 Annual Radiological Environmental Operating Report BYRON 2004-0043, 2003 Occupational Radiation Exposure Report2004-04-30030 April 2004 2003 Occupational Radiation Exposure Report 2022-05-31
[Table view] Category:Letter type:RS
MONTHYEARRS-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-22-057, Constellation Radiological Emergency Plan Addendum Revision2022-04-21021 April 2022 Constellation Radiological Emergency Plan Addendum Revision 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 RS-22-051, Constellation Energy Generation, LLC - Update to Correspondence Addressees and Service Lists2022-04-12012 April 2022 Constellation Energy Generation, LLC - Update to Correspondence Addressees and Service Lists RS-22-047, Annual 10 CFR 50.46 Report of Emergency Core Cooling System Evaluation Model Changes and Errors2022-04-0808 April 2022 Annual 10 CFR 50.46 Report of Emergency Core Cooling System Evaluation Model Changes and Errors RS-22-050, Supplemental Information - Proposed Alternative for Examination of Pressurizer Circumferential and Longitudinal Shell-to-Head Welds and Nozzle-to-Vessel Welds2022-04-0808 April 2022 Supplemental Information - Proposed Alternative for Examination of Pressurizer Circumferential and Longitudinal Shell-to-Head Welds and Nozzle-to-Vessel Welds RS-22-049, Constellation Energy Generation, LLC, Supplemental Information to Correct Typographical Errors in Constellation'S Application to Revise Technical Specifications to Adopt TSTF-541 Revision 2, Add Exceptions to Surveillance Requirements for2022-04-0404 April 2022 Constellation Energy Generation, LLC, Supplemental Information to Correct Typographical Errors in Constellation'S Application to Revise Technical Specifications to Adopt TSTF-541 Revision 2, Add Exceptions to Surveillance Requirements for V RS-22-045, Constellation Energy Generation, LLC, Response to NRC Regulatory Issue Summary 2022-01, Preparation and Scheduling of Operator Licensing Examinations2022-03-25025 March 2022 Constellation Energy Generation, LLC, Response to NRC Regulatory Issue Summary 2022-01, Preparation and Scheduling of Operator Licensing Examinations RS-22-014, Application to Adopt TSTF-246, RTS Instrumentation, 3.3.1 Condition F Completion Time2022-03-24024 March 2022 Application to Adopt TSTF-246, RTS Instrumentation, 3.3.1 Condition F Completion Time RS-22-041, Withdrawal of License Amendment Request to Revise Technical Specifications 5.6.5, Core Operating Limits Report (COLR) to Add References to NRC Approved Topical Report with Administrative Changes2022-03-22022 March 2022 Withdrawal of License Amendment Request to Revise Technical Specifications 5.6.5, Core Operating Limits Report (COLR) to Add References to NRC Approved Topical Report with Administrative Changes RS-22-036, Supplemental Information - Proposed Alternative for Examination of Pressurizer Circumferential and Longitudinal Shell-to-Head Welds and Nozzle-to-Vessel Welds2022-03-10010 March 2022 Supplemental Information - Proposed Alternative for Examination of Pressurizer Circumferential and Longitudinal Shell-to-Head Welds and Nozzle-to-Vessel Welds RS-22-023, Constellation Energy Generation, LLC, Executed Trust Fund Agreement Amendment and Subordinate Trust Agreement2022-02-23023 February 2022 Constellation Energy Generation, LLC, Executed Trust Fund Agreement Amendment and Subordinate Trust Agreement RS-22-027, Constellation, Response to Request for Additional Information Regarding Application to Revise Technical Specifications to Adopt TSTF-541 Revision 2, Add Exceptions to Surveillance Requirements for Valves and Dampers Locked in the Actuated2022-02-23023 February 2022 Constellation, Response to Request for Additional Information Regarding Application to Revise Technical Specifications to Adopt TSTF-541 Revision 2, Add Exceptions to Surveillance Requirements for Valves and Dampers Locked in the Actuated P RS-22-019, Constellation Energy Generation, LLC - Update to Correspondence Addressees and Service Lists2022-02-16016 February 2022 Constellation Energy Generation, LLC - Update to Correspondence Addressees and Service Lists RS-22-015, Notification of Completion of License Transfer and Request to Continue Processing Pending NRC Actions Previously Requested by Exelon Generation Company, LLC2022-02-0101 February 2022 Notification of Completion of License Transfer and Request to Continue Processing Pending NRC Actions Previously Requested by Exelon Generation Company, LLC 2024-01-11
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Exe~ Ir m Exelor Generation 4300 Vd ;nfield Road wvvw.exeloncorp_com Nuclear Warrenville, IL 60555 10 CFR 50.46 RS-07-052 April 13, 2007 United States Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D .C. 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 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" 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, is submitting the annual report of the Emergency Core Cooling System Evaluation Model changes and errors for Byron Station, Units 1 and 2, and Braidwood Station, Units 1 and 2.
Attachment 1, "Peak Cladding Temperature Rack-Up Sheets," provides updated information regarding the peak cladding temperature for the limiting small break and large break loss-of-coolant accident analyses evaluations for the Byron and Braidwood Stations . Attachment 2, "Assessment Notes," contains a detailed description for each change or error reported .
Please contact Mr. David Chrzanowski at (630) 657-2816 should you have any questions concerning this report .
PWLfx K . A~~
Respectfully, Patrick R. Simpson Manager - Licensing Attachment 1 : Peak Cladding Temperature Rack-Up Sheets Attachment 2: Assessment Notes
Attachment 1 Peak Cladding Temperature Rack-Up Sheets BRAIDWOOD STATION UNITS 1 AND 2 Docket Nos. 50-456 and 50-457 License Nos. NPF-72 and NPF-77 and BYRON STATION UNITS 1 AND 2 Docket Nos. 50-454 and 50-455 License Nos. NPF-37 and NPF-66 10 CFR 50.46, "Acceptance criteria for emergency core cooling systems for light-water nuclear power reactors" Report of the Emergency Core Cooling System Evaluation Model Changes and Errors Assessments
Attachment 1 Peak Cladding Temperature Rack-up Sheets Byron and Braidwood Units 1 and 2 10CFR50 .46 Report PLANT NAME: Braidwood Station Unit 1 ECCS EVALUATION MODEL: Small Break Loss of Coolant Accident (SBLOCA)
REPORT REVISION DATE: 03/07/07 CURRENT OPERATING CYCLE : 13 ANALYSIS OF RECORD (AOR)
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 Heat Flux Hot Channel Factor (FQ) = 2 .60 Nuclear Enthalpy Rise Hot Channel Factor (FNAH) = 1 .70 Steam Generator Tube Plugging (SGTP) = 5%
Limiting Break Size : 2" Low Tavg Notes : Zr-4/ZIRLO Clad Fuel Reference Peak Cladding Temperature (PCT) PCT = 1624 .0°F MARGIN ALLOCATION A. PRIOR LOSS OF COOLANT ACCIDENT (LOCA) MODEL ASSESSMENTS 10CFR50.46 report dated June 11, 2001 see note 1 APCT = 0°F 10CFR50.46 report dated Aril 18, 2002 see note 2 APCT = 0°F 10CFR50.46 report dated Aril 14, 2003 see note 3 APCT = 0°F 10CFR50.46 report dated Aril 14, 2004 see note 4 APCT = 35°F 10CFR50.46 report dated Aril 14, 2005 see note 56) APCT = 0°F 10CFR50.46 report dated April 14, 2006 (see note APCT = 0°F NET PCT PCT = 1659 .0°F B. CURRENT LOCA MODEL ASSESSMENTS General Code Maintenance (see note 8) ( APCT = 0°F NOTRUMP Refined Break Spectrum (see note 11) - - APCT = 0°F NET PCT PCT = 1659 .0°F
Attachment 1 Peak Cladding Temperature Rack-up Sheets Byron and Braidwood Units 1 and 2 10CFR50.46 Report PLANT NAME : Braidwood Station Unit 1 ECCS EVALUATION MODEL: Large Break Loss of Coolant Accident (LBLOCA)
REPORT REVISION DATE: 03/07/07 CURRENT OPERATING CYCLE: 13 AOR Evaluation Model: CQD (1996)
Calculation: Westinghouse CN-LIS-00-7, September 2000 Fuel : VANTAGE+ 17 x 17 Limiting Fuel Type: VANTAGE+ 17 x 17 Limiting Single Failure: Loss of one train of ECCS flow Heat Flux Hot Channel Factor (FQ) = 2.60 Nuclear Enthalpy Rise Hot Channel Factor (FNAH) = 1 .70 Steam Generator Tube Plugging (SGTP) = 5%
Limiting Break Size : Guillotine Notes: Zr-4/ZIRLO Clad Fuel Reference PCT PCT = 2044.0°F MARGIN ALLOCATION A. PRIOR LOCA MODEL ASSESSMENTS 10CFR50.46 report dated June 11, 2001 see note 1 APCT = 12°F 10CFR50.46 report dated Aril 18, 2002 see note 2 APCT = 0°F 10CFR50.46 re p ort dated April 14, 2003 see note 3 APCT = 0°F 10CFR50.46 report dated Aril 14, 2004 see note 4 APCT = 0°F 10CFR50.46 report dated Aril 14, 2005 see note 5 APCT = 5°F 10CFR50.46 report dated Aril 14, 2006 see note 6 APCT = 0°F Axial Power Shape Distribution Violation (see note 7) APCT = 80°F NET PCT PCT = 2141 .0°F CURRENT LOCA MODEL ASSESSMENTS I Containment Heat Sink Addition (See note 9) 4PCT = 0°F NET PCT PCT = 2141 .0°F
Attachment 1 Peak Cladding Temperature Rack-up Sheets Byron and Braidwood Units 1 and 2 10CFR50.46 Report PLANT NAME: Braidwood Station Unit 2 ECCS EVALUATION MODEL: SBLOCA REPORT REVISION DATE: 03/07/07 CURRENT OPERATING CYCLE: 13 AOR 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 Heat Flux Hot Channel Factor (FQ) = 2 .60 Nuclear Enthalpy Rise Hot Channel Factor (FNAH) = 1 .70 Steam Generator Tube Plugging (SGTP) = 10%
Limiting Break: 2" Low Tavg Notes : Zr-4/ZIRLO Clad Fuel Reference PCT PCT = 1627 .0°F MARGIN ALLOCATION A. PRIOR LOCA MODEL ASSESSMENTS 10CFR50.46 report dated June 11, 2001 see note 1 APCT = 3°F 10CFR50 .46 report dated Aril 18, 2002 see note 2 APCT = 0°F 10CFR50.46 report dated Aril 14, 2003 see note 3 APCT = 0°F 10CFR50.46 report dated Aril 14, 2004 see note 4 APCT= 35°F 10CFR50.46 report dated Aril 14, 2005 see note 5 APCT = 0°F 1 OCFR50.46 report dated April 14, 2006 (see note 6) APCT = 0°F NET PCT PCT = 1665 .0°F B. CURRENT LOCA MODEL ASSESSMENTS General Code Maintenance (see note 8) OPCT = 0°F NOTRUMP Refined Break Spectrum (see note 11) APCT = 0°F NET PCT PCT = 1665.0°F
Attachment 1 Peak Cladding Temperature Rack-up Sheets Byron and Braidwood Units 1 and 2 10CFR50.46 Report PLANT NAME : Braidwood Station Unit 2 ECCS EVALUATION MODEL: LBLOCA REPORT REVISION DATE : 03/07/07 CURRENT OPERATING CYCLE: 13 AOR Evaluation Model : CQD (1996)
Calculation : Westinghouse CN-LIS-00-7, September 2000 Fuel : VANTAGE+ 17 x 17 Limiting Fuel Type: VANTAGE+ 17 x 17 Limiting Single Failure: Loss of one train of ECCS flow Heat Flux Hot Channel Factor (FQ) = 2 .60 Nuclear Enthalpy Rise Hot Channel Factor (FNAH) = 1 .70 Steam Generator Tube Plugging (SGTP) = 10%
Limiting Break Size : Guillotine Notes : Zr-4/ZIRLO Clad Fuel Reference PCT PCT = 2088.0°F MARGIN ALLOCATION A. PRIOR LOCA MODEL ASSESSMENTS 10CFR50 .46 report dated June 11, 2001 see note 1 APCT = 12°F 10CFR50 .46 report dated Aril 18, 2002 see note 2 APCT = 0°F 10CFR50 .46 report dated Aril 14, 2003 see note 3 APCT = 0°F 10CFR50 .46 report dated Aril 14, 2004 see note 4 APCT = 0°F 10CFR50.46 report dated Aril 14, 2005 see note 5 APCT = 5°F
[ 1 OCFR50.46 report dated April 14, 2006 (see note 6) APCT = 0°F NET PCT PCT = 2105.0°F B. CURRENT LOCA MODEL ASSESSMENTS Axial Power Shape Distribution Violation see note 7 APCT = 8°F Containment Heat Sink Addition (See note 9) APCT = 0°F NET PCT PCT = 2113.0°F
Attachment 1 Peak Cladding Temperature Rack-up Sheets Byron and Braidwood Units 1 and 2 10CFR50.46 Report PLANT NAME: Boron Station Unit 1 ECCS EVALUATION MODEL: SBLOCA REPORT REVISION DATE: 03/07/07 CURRENT OPERATING CYCLE: 15 AOR 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 Heat Flux Hot Channel Factor (FQ) = 2.60 Nuclear Enthalpy Rise Hot Channel Factor (FNAH) = 1 .70 Steam Generator Tube Plugging (SGTP) = 5%
Limiting Break: 2" Low Tavg Notes : Zr-4/ZIRLO Clad Fuel Reference PCT PCT = 1624.0°F MARGIN ALLOCATION A. PRIOR LOSS OF COOLANT ACCIDENT (LOCA) MODEL ASSESSMENTS 10CFR50 .46 report dated June 11, 2001 see note 1 APCT = 0°F 10CFR50 .46 report dated Aril 18, 2002 see note 2 APCT = 0°F 10CFR50 .46 report dated Aril 14, 2003 see note 3 APCT = 0°F 10CFR50.46 report dated Aril 14, 2004 see note 4 APCT = 35°F 10CFR50 .46 report dated Aril 14, 2005 see note 5 APCT = 0°F 1 OCFR50.46 report dated April 14, 2006 (see note 6) APCT = 0°F NET PCT PCT = 1659.0°F B. CURRENT LOCA MODEL ASSESSMENTS General Code Maintenance see note 8 APCT = 0°F AXIOM LTAs See note 10 APCT = 0°F NOTRUMP Refined Break Spectrum (see note 11) APCT = 0°F NET PCT PCT = 1659 .0°F
Attachment 1 Peak Cladding Temperature Rack-up Sheets Byron and Braidwood Units 1 and 2 10CFR50.46 Report PLANT NAME: Boron Station Unit 1 ECCS EVALUATION MODEL: LBLOCA REPORT REVISION DATE: 03/07/07 CURRENT OPERATING CYCLE : 15 AOR Evaluation Model: CQD (1996)
Calculation : Westinghouse CN-LIS-00-7, September 2000 Fuel : VANTAGE+ 17 x 17 Limiting Fuel Type: VANTAGE+ 17 x 17 Limiting Single Failure : Loss of one train of ECCS flow Steam Generator Tube Plugging (SGTP) = 5%
Heat Flux Hot Channel Factor (FQ) = 2.60 Nuclear Enthalpy Rise Hot Channel Factor (FNAH) = 1 .70 Limiting Break Size : Guillotine Notes: Zr-4/ZIRLO Clad Fuel Reference PCT PCT = 2044.0°F MARGIN ALLOCATION A. PRIOR LOCA MODEL ASSESSMENTS 10CFR50 .46 report dated June 11, 2001 see note 1 APCT = 12°F 10CFR50 .46 report dated Aril 18, 2002 see note 2 APCT = 0°F 10CFR50 .46 report dated Aril 14, 2003 see note 3 APCT = 0°F 10CFR50.46 report dated Aril 14, 2004 see note 4 APCT = 0°F 10CFR50.46 report dated Aril 14, 2005 see note 5 APCT = 5°F 10CFR50.46 rel wik0;- ier r?~ ~ril 14, 2006 (see note 6) APCT = 0° F NET PCT PCT = 2061 .0°F B. CURRENT LOCA MODEL ASSESSMENTS Axial Power Shape Distribution Violation see note 7 APCT = 80°F Containment Heat Sink Addition See note 9 APCT = 0°F AXIOM LTAs (See note 10) APCT = 0°F NET PCT PCT = 2141 .0°F
Attachment 1 Peak Cladding Temperature Rack-up Sheets Byron and Braidwood Units 1 and 2 10CFR50.46 Report PLANT NAME: Boron Station Unit 2 ECCS EVALUATION MODEL: SBLOCA REPORT REVISION DATE : 03/07/07 CURRENT OPERATING CYCLE : 14*
- Cycle 14 to start April 2007 AOR 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 Heat Flux Hot Channel Factor (FQ) = 2 .60 Nuclear Enthalpy Rise Hot Channel Factor (FNAH) = 1 .70 Steam Generator Tube Plugging (SGTP) = 10%
Limiting Break: 2" Low Tavg Notes: Zr-4/ZIRLO Clad Fuel Reference PCT PCT = 1627 .0°F MARGIN ALLOCATION A. PRIOR LOCA MODEL ASSESSMENTS 10CFR50.46 report dated June 11, 2001 see note 1 APCT = 3°F 10CFR50.46 report dated Aril 18, 2002 see note 2 APCT = 0°F 10CFR50.46 report dated Aril 14, 2003 see note 3 APCT = 0°F 10CFR50.46 report dated Aril 14, 2004 see note 4 APCT = 35°F 10CFR50.46 report dated Aril 14, 2005 see note 5 APCT = 0°F 10CFR50.46 14, 2006 (see note 6 -- APCT = 0°F NET PCT PCT = 1665.0°F B. CURRENT LOCA MODEL ASSESSMENTS General Code Maintenance see note 8 APCT = 0°F NOTRUMP Refined Break Spectrum (see note 11) APCT = 0°F NET PCT PCT = 1665 .0°F
Attachment 1 Peak Cladding Temperature Rack-up Sheets Byron and Braidwood Units 1 and 2 10CFR50.46 Report PLANT NAME: Byron Station Unit 2 ECCS EVALUATION MODEL: LBLOCA REPORT REVISION DATE: 03/07/07 CURRENT OPERATING CYCLE : 14*
- Cycle 14 to start April 2007 AOR Evaluation Model: CQD (1996)
Calculation: Westinghouse CN-LIS-00-7, September 2000 Fuel : VANTAGE+ 17 x 17 Limiting Fuel Type: VANTAGE+ 17 x 17 Limiting Single Failure: Loss of one train of ECCS flow Heat Flux Hot Channel Factor (FQ) = 2.60 Nuclear Enthalpy Rise Hot Channel Factor (FNAH) = 1 .70 Steam Generator Tube Plugging (SGTP) = 10%
Limiting Break Size : Guillotine Notes : Zr-4/ZIRLO Clad Fuel Reference PCT PCT = 2088 .0°F MARGIN ALLOCATION A. PRIOR LOCA MODEL ASSESSMENTS 10CFR50 .46 report dated June 11, 2001 see note 1 APCT = 12°F 10CFR50 .46 report dated Aril 18, 2002 see note 2 APCT = 0°F 10CFR50 .46 report dated Aril 14, 2003 see note 3 APCT = 0°F 10CFR50 .46 report dated Aril 14, 2004 see note 4 APCT = 0°F 10CFR50 .46 report dated Aril 14, 2005 see note 5 APCT = 5°F 10CFR50 .46 report dated April 14, 2006 (see note 6) APCT = 0°F NET PCT PCT = 2105.0°F B. CURRENT LOCA MODEL ASSESSMENTS Axial Power Shape Distribution Violation (see note 7) APCT = 8°F Containment Heat Sink Addition (See note 9) APCT = 0°F NET PCT PCT = 2113.0°F
Attachment 2 Assessment Notes BRAIDWOOD STATION UNITS 1 AND 2 Docket Nos . 50-456 and 50-457 License Nos . NPF-72 and NPF-77 and BYRON STATION UNITS 1 AND 2 Docket Nos. 50-454 and 50-455 License Nos . NPF-37 and NPF-66 10 CFR 50.46, "Acceptance criteria for emergency core cooling systems for light-water nuclear power reactors" Report of the Emergency Core Cooling System Evaluation Model Changes and Errors Assessments
Attachment 2 Assessment Notes Byron and Braidwood Units 1 and 2 10 CFR 50.46 Report 1 . Prior LOCA Model Assessment The 10CFR50.46 report dated June 11, 2001, reported new LBLOCA and SBLOCA analyses to support operations at uprated power conditions. The same report assessed the impact from a decay heat uncertainty error in Monte Carlo calculations on LBLOCA analysis and the impact from annular axial blankets on SBLOCA analysis . Evaluations for plant conditions and LBLOCA and SBLOCA model changes which resulted in 0°F PCT change were reported . Cycle specific evaluations related to axial power shape distribution envelope violation were reported for the applicable operating cycles .
- 2. Prior LOCA Model Assessment The 10CFR50.46 report dated April 18, 2002, reported evaluations for LBLOCA and SBLOCA model changes which resulted in 0°F PCT change . Cycle specific evaluations related to axial power shape distribution envelope violation were reported for the applicable operating cycles .
3 . Prior LOCA Model Assessment The 10CFR50 .46 report dated April 14, 2003, reported evaluations for LBLOCA and SBLOCA model changes which resulted in 0°F PCT change . Cycle specific evaluations related to axial power shape distribution envelope violation were reported for the applicable operating cycles .
- 4. Prior LOCA Model Assessment The 10CFR50.46 report dated April 14, 2004, reported evaluations for LBLOCA model changes which resulted in 0 °F PCT change. A SBLOCA assessment related to NOTRUMP bubble rise/drift flux model inconsistency corrections, which resulted in 35°F PCT assessment, was reported . Cycle specific evaluations related to axial power shape distribution envelope violation were reported for the applicable operating cycles .
- 5. Prior LOCA Model Assessment The 10CFR50.46 report dated April 14, 2005, reported evaluations for LBLOCA model changes, which resulted in a 5°F PCT change due to the Revised Blowdown Heatup Uncertainty Calculation. Assembly N10S was reconstituted with two stainless steel filler rods during Braidwood Unit 1 Refueling Outage 11 . This assembly is reloaded into the core and is in use during Braidwood Unit 1 Cycle 12 operation . The introduction of up to five stainless steel filler rods has been evaluated and shown to have no impact on LBLOCA and SBLOCA analyses . The estimated PCT effect is 0°F. This assembly will be discharged during Reload 12 and will not be resident in the core for Braidwood Unit 1 Cycle 13. Cycle specific evaluations related to axial power shape distribution envelope violation were reported for the applicable operating cycles .
Attachment 2 Assessment Notes Byron and Braidwood Units 1 and 2 10 CFR 50.46 Report 6 . Prior LOCA Model Assessment The 10CFR50.46 report dated April 14, 2006, reported evaluations for LBLOCA model changes which resulted in a 0°F PCT change due to the Revised Iteration Algorithm for Calculating the Average Fuel Temperature, Pellet Radial Profile, Improved Automation of End of Blowdown Time, Thermodynamic Properties from THERMO, Vessel Unheated Conductor Noding and Containment Relative Humidity Assumption . General Code Maintenance for SBLOCA model resulted in 0°F PCT change . Cycle specific evaluations related to axial power shape distribution envelope violation were reported for the applicable operating cycles .
7 . Axial power Shape Distribution Envelope Violation (PMID,PBOT)
The LBLOCA analysis is performed based on assuming an axial power shape distribution envelope (PMID, PBOT), where PMID is the power in the middle one-third of the core ; and PBOT is the power in the lower one-third of the core. The envelope is pertinent to the BELOCA analysis and is presented as Figure 11-1 of WCAP-15585, "Best Estimate Analysis of the Large Break Loss of Coolant Accident for the Byron/Braidwood Units 1 and 2 Nuclear Plant," November 2000 . For every reload cycle Westinghouse verifies that the envelope remains limiting. If there is a violation then a PCT penalty is calculated .
For Braidwood Unit 1 Cycle 13 there was a violation and a PCT penalty of 80°F was calculated . This penalty has been reported in 10CFR50.46 report dated April 14, 2006.
For Braidwood Unit 2 Cycle 13 there was a violation and a PCT penalty of 8°F was calculated .
For Byron Unit 1 Cycle 15 there was a violation and a PCT penalty of 80°F was calculated, which was first reported for Byron Unit 1 Cycle 13 in the 30-day 50.46 notification to the NRC, dated November 7, 2003 . The assessment for Byron Unit 1 Cycle 15 does not affect the net PCT.
For Byron Unit 2 Cycle 14 there was a violation and a PCT penalty of 8°F was calculated .
For Byron Unit 2 Cycle 14, Westinghouse found two types of violations . For the violations outside of the sampling range shown in Figure 11-1 of WCAP-15585 but inside the response surface shown in Figure 9 .2-1 of WCAP-15585, a PCT penalty of 8 °F .
This is the same type of violation as the cycles discussed above.
The second type of violation is for power shapes slightly outside the response surface shown in Figure 9 .2-1 of WCAP-15585. These violations were determined to be non-limiting power shapes and were evaluated by extrapolating the power distribution response surface in order to predict PCT. In all cases the predicted PCT was non-limiting as compared to the LBLOCA PCT reported in the Attachment 1 LBLOCA PCT sheet. The Westinghouse reload methodology approved by the NRC, WCAP-9272-P-A, "Westinghouse Reload Safety Evaluation Methodology, March 1978,"
allows evaluation of key parameters slightly out of bounds using conservative quantitative evaluation . The extrapolation methodology used in this evaluation is 2of3
Attachment 2 Assessment Notes Byron and Braidwood Units 1 and 2 10 CFR 50.46 Report consistent with the methodology for extrapolating FQ and FAH described in WCAP-12945-P-A .
- 8. General Code Maintenance (NOTRUMP)
Various changes in code input and output format have been made to enhance usability and help preclude errors in analyses . This includes both input changes (e .g ., more relevant input variables defined and more common input values used as defaults) and input diagnostics designed to preclude unreasonable values from being used, as well as various changes to code output which have no effect on calculated results . In addition, various updates were made to eliminate inactive coding, improve active coding, and enhance commenting, both for enhanced usability and to facilitate code debugging when necessary. These changes represent Discretionary Changes that will be implemented on a forward-fit basis, in accordance with Section 4.1 .1 of WCAP-13451 . The nature of these changes leads to an estimated PCT impact of 0°F.
- 9. Containment Heat Sink Addition Addition of heat sink materials in containment was considered for LBLOCA PCT penalties for Byron and Braidwood Units 1 and 2. These additions included GSI-191, "Assessment of Debris Accumulation on Pressurized-Water Reactor (PWR) Sump Performance," sump screens and related materials and outage pre-staging materials such as lead shielding. The PCT effect was zero degrees.
- 10. AXIOM LTAs Byron Unit 1 Cycle 15 implemented Westinghouse AXIOM Lead Test Assemblies (LTAs) . The LTA program tests fuel rods clad with an advanced Westinghouse alloy.
The PCT effect was zero degrees.
11 . NOTRUMP Refined Break Spectrum The Nuclear Regulatory Commission (NRC) questioned the break spectrum analyzed in the NOTRUMP evaluation model (EM) . The NRC was concerned that the resolution of the break spectrum used in the NOTRUMP EM (1 .5, 2, 3, 4, and 6 inch cases) may not be fine enough to capture the worst break with regard to limiting peak clad temperature as per 10CFR50.46. That is, the plant could be SBLOCA limited with regard to overall LOCA results. For plants with SBLOCA PCTs less than 1700 F, no explicit, plant-specific break spectrum calculations were performed. Based on generic sensitivity studies, the estimated PCT impact is 0°F.