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MONTHYEARML13310B9292013-12-0606 December 2013 Summary of Meeting with Constellation Energy Group, Inc., to Continue Discussions on the Proposed Risk-Informed Approach to the Resolution of Generic Letter 2004-02, Potential Impact of Debris Blockage on Emergency Recirculation During ... Project stage: Meeting RS-15-091, Response to Request for Additional Information Regarding Request for a License Amendment to Braidwood Station, Units 1 and 2, Technical Specification 3.7.9, Ultimate Heat Sink2015-03-31031 March 2015 Response to Request for Additional Information Regarding Request for a License Amendment to Braidwood Station, Units 1 and 2, Technical Specification 3.7.9, Ultimate Heat Sink Project stage: Response to RAI ML15232A5892015-09-14014 September 2015 Regulatory Audit Report for the June 11, 2015 Audit in Support of the Ultimate Heat Sink License Amendment Request Project stage: Other RS-15-266, Response to Request for Additional Information Regarding Request for a License Amendment to Technical Specification 3.7.9, Ultimate Heat Sink.2015-10-0909 October 2015 Response to Request for Additional Information Regarding Request for a License Amendment to Technical Specification 3.7.9, Ultimate Heat Sink. Project stage: Response to RAI ML15299A0132015-10-21021 October 2015 NRR E-mail Capture - (External_Sender) 1 of 2: Final Report: Five Year Post-construction Monitoring of the Unionid Community Near Braidwood Station Kankakee River Discharge Project stage: Request ML15322A3172015-11-18018 November 2015 Record of Decision Project stage: Request ML15314A8142015-11-30030 November 2015 NUREG-1437, Suppl 55, Generic Environmental Impact Statement for License Renewal of Nuclear Plants Regarding Braidwood Station, Units 1 and 2 Final Report. Project stage: Acceptance Review RS-16-049, Supplemental Information Regarding Request for a License Amendment to Technical Specification 3.7.9, Ultimate Heat Sink.2016-02-12012 February 2016 Supplemental Information Regarding Request for a License Amendment to Technical Specification 3.7.9, Ultimate Heat Sink. Project stage: Supplement ML16181A0072016-07-18018 July 2016 Environmental Assessment and Finding of No Significant Impact Related to Ultimate Heat Sink Modification Project stage: Other 2015-11-30
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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-23-114, 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-23-097, 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-23-083, 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-23-040, 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-22-093, 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 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 RS-22-008, Application to Revise Technical Specifications 5.6.5, Core Operating Limits Report to Add References to NRC-Approved Topical Report with Admin2022-01-24024 January 2022 Application to Revise Technical Specifications 5.6.5, Core Operating Limits Report to Add References to NRC-Approved Topical Report with Admin RS-22-004, Supplement to Application to Adopt TSTF-554, Revision 1, Revise Reactor Coolant Leakage Requirements2022-01-0404 January 2022 Supplement to Application to Adopt TSTF-554, Revision 1, Revise Reactor Coolant Leakage Requirements RS-21-121, Proposed Changes to Decommissioning Trust Agreements and Master Terms2021-12-15015 December 2021 Proposed Changes to Decommissioning Trust Agreements and Master Terms RS-21-117, Application for Amendment to Renewed Facility License to Remove License Condition 2.C.(12)(d)2021-12-0909 December 2021 Application for Amendment to Renewed Facility License to Remove License Condition 2.C.(12)(d) RS-21-112, Updated 10 CFR 50.46 Report of Emergency Core Cooling System Evaluation Model Changes and Errors2021-10-22022 October 2021 Updated 10 CFR 50.46 Report of Emergency Core Cooling System Evaluation Model Changes and Errors RS-21-091, Implementation of Insider Threat Program Requirements Associated with the Voluntary Security Clearance Program and Advisement of Leadership Changes2021-09-13013 September 2021 Implementation of Insider Threat Program Requirements Associated with the Voluntary Security Clearance Program and Advisement of Leadership Changes RS-21-093, R. E. Ginna, Proposed Alternative for Examinations of Examination Category C-B Steam Generator Nozzle-to-Shell Welds and Nozzle Inside Radius Sections2021-09-0101 September 2021 R. E. Ginna, Proposed Alternative for Examinations of Examination Category C-B Steam Generator Nozzle-to-Shell Welds and Nozzle Inside Radius Sections 2024-01-11
[Table view] Category:Technical Specification
MONTHYEARRS-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-23-075, 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-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. ML22145A5862022-06-0303 June 2022 R. E. Ginna Correction to Pages Issued for Amendments Nos. 225, 225, 227, 227, & 148, Respectively, Regarding Issues Identified in Westinghouse Documents 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-22-037, 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-22-008, Application to Revise Technical Specifications 5.6.5, Core Operating Limits Report to Add References to NRC-Approved Topical Report with Admin2022-01-24024 January 2022 Application to Revise Technical Specifications 5.6.5, Core Operating Limits Report to Add References to NRC-Approved Topical Report with Admin RS-21-082, Attachment 2 - Technical Basis for NSAL-09-5 Revision 1 and Applicability to Byron, Braidwood, and R.E. Ginna Non-Proprietary Version for NRC2021-08-13013 August 2021 Attachment 2 - Technical Basis for NSAL-09-5 Revision 1 and Applicability to Byron, Braidwood, and R.E. Ginna Non-Proprietary Version for NRC ML21008A4172020-12-17017 December 2020 Technical Specification Bases RS-20-068, Application to Revise Technical Specifications 3.8.1, AC Sources-Operating2020-06-26026 June 2020 Application to Revise Technical Specifications 3.8.1, AC Sources-Operating JAFP-20-0049, Application to Adopt TSTF-427, Revision 2, Allowance for Non Technical Specification Barrier Degradation on Supported System Operability2020-06-22022 June 2020 Application to Adopt TSTF-427, Revision 2, Allowance for Non Technical Specification Barrier Degradation on Supported System Operability RS-20-044, Emergency License Amendment Request for a One-Time Extension of the Steam Generator Tube Inspections2020-04-0606 April 2020 Emergency License Amendment Request for a One-Time Extension of the Steam Generator Tube Inspections ML20063L2822020-02-28028 February 2020 Application of Westinghouse Full Spectrum LOCA Evaluation Model RS-20-010, Application to Revise Technical Specifications 5.6.5, Core Operating Limits Report (COLR)2020-02-28028 February 2020 Application to Revise Technical Specifications 5.6.5, Core Operating Limits Report (COLR) JAFP-19-0067, Application to Adopt TSTF-427, Revision 2, Allowance for Non Technical Specification Barrier Degradation on Supported System Operability2019-06-27027 June 2019 Application to Adopt TSTF-427, Revision 2, Allowance for Non Technical Specification Barrier Degradation on Supported System Operability RS-19-039, Fleet License Amendment Request - Common Language for Technical Specification High Radiation Area Administrative Controls2019-06-26026 June 2019 Fleet License Amendment Request - Common Language for Technical Specification High Radiation Area Administrative Controls ML16357A5322016-12-15015 December 2016 Byron/Braidwood Nuclear Stations, Revision 16 to Updated Final Safety Analysis Report, Chapter 16, Technical Specifications RS-16-248, Braidwood Nuclear Stations, Revision 16 to Updated Final Safety Analysis Report, Technical Specifications Bases2016-12-15015 December 2016 Braidwood Nuclear Stations, Revision 16 to Updated Final Safety Analysis Report, Technical Specifications Bases ML16357A5802016-12-15015 December 2016 Byron Nuclear Stations, Revision 16 to Updated Final Safety Analysis Report, Technical Specifications Bases ML16357A5792016-12-15015 December 2016 Braidwood Nuclear Stations, Revision 16 to Updated Final Safety Analysis Report, Technical Specifications Bases RS-16-248, Byron/Braidwood Nuclear Stations, Revision 16 to Updated Final Safety Analysis Report, Chapter 16, Technical Specifications2016-12-15015 December 2016 Byron/Braidwood Nuclear Stations, Revision 16 to Updated Final Safety Analysis Report, Chapter 16, Technical Specifications RS-16-248, Byron Nuclear Stations, Revision 16 to Updated Final Safety Analysis Report, Technical Specifications Bases2016-12-15015 December 2016 Byron Nuclear Stations, Revision 16 to Updated Final Safety Analysis Report, Technical Specifications Bases ML16209A2182016-07-26026 July 2016 Application to Revise Technical Specifications to Adopt TSTF-545, Revision 3, TS Lnservice Testing Program Removal & Clarify SR Usage Rule Application to Section 5.5 Testing. RS-15-091, Response to Request for Additional Information Regarding Request for a License Amendment to Braidwood Station, Units 1 and 2, Technical Specification 3.7.9, Ultimate Heat Sink2015-03-31031 March 2015 Response to Request for Additional Information Regarding Request for a License Amendment to Braidwood Station, Units 1 and 2, Technical Specification 3.7.9, Ultimate Heat Sink RS-14-192, Application to Revise Technical Specifications to Adopt TSTF-523, Generic Letter 2008-01, Managing Gas Accumulation, Revision 2, Using the Consolidated Line Item Improvement Process2014-07-14014 July 2014 Application to Revise Technical Specifications to Adopt TSTF-523, Generic Letter 2008-01, Managing Gas Accumulation, Revision 2, Using the Consolidated Line Item Improvement Process RS-13-175, Bryon, Units 1 and 2, License Amendment Request to Revise Technical Specifications Section 3.7.2, Main Steam Isolation Valves (Msivs).2013-08-21021 August 2013 Bryon, Units 1 and 2, License Amendment Request to Revise Technical Specifications Section 3.7.2, Main Steam Isolation Valves (Msivs). RS-12-223, License Amendment Request to Revise Technical Specification 3.3.6, Containment Ventilation Isolation Instrumentation.2012-12-21021 December 2012 License Amendment Request to Revise Technical Specification 3.3.6, Containment Ventilation Isolation Instrumentation. RS-12-221, Braidwood, Units 1 and 2, Updated Final Safety Analysis Report, Revision 14, Technical Specifications Bases2012-12-14014 December 2012 Braidwood, Units 1 and 2, Updated Final Safety Analysis Report, Revision 14, Technical Specifications Bases ML13004A0692012-12-14014 December 2012 Updated Final Safety Analysis Report, Revision 14, Technical Specifications Bases RS-12-028, Application to Revise Technical Specifications to Adopt TSTF-510, Revision to Steam Generator Program Inspection Frequencies and Tube Sample Selection.2012-03-22022 March 2012 Application to Revise Technical Specifications to Adopt TSTF-510, Revision to Steam Generator Program Inspection Frequencies and Tube Sample Selection. RS-12-023, License Amendment Request to Revise Tech Specs Sections 5.5.9, Steam Generator (SG) Program, and TS 5.6.9, Steam Generator (SG) Tube Inspection Report, for Permanent Alternate Repair Criteria2012-03-20020 March 2012 License Amendment Request to Revise Tech Specs Sections 5.5.9, Steam Generator (SG) Program, and TS 5.6.9, Steam Generator (SG) Tube Inspection Report, for Permanent Alternate Repair Criteria ML1117900332011-06-23023 June 2011 Unit 1 & 2, Request for License Amendment Regarding Measurement Uncertainty Recapture (Mur) Power Uprate, Attachment 3A, Markup of Proposed Technical Specifications Bases and Technical Requirements Manual Pages RS-10-029, Application for Technical Specification Change Regarding Risk-Informed Justification for the Relocation of Specific Surveillance Frequency Requirements to Licensee Controlled Program2010-02-15015 February 2010 Application for Technical Specification Change Regarding Risk-Informed Justification for the Relocation of Specific Surveillance Frequency Requirements to Licensee Controlled Program RS-09-071, Units 1 and 2 - License Amendment Request to Revise Technical Specifications (TS) for Permanent Alternate Repair Criteria2009-06-24024 June 2009 Units 1 and 2 - License Amendment Request to Revise Technical Specifications (TS) for Permanent Alternate Repair Criteria RS-08-161, Braidwood, Units 1 and 2 - Technical Specification Bases2008-12-15015 December 2008 Braidwood, Units 1 and 2 - Technical Specification Bases ML0902704742008-12-15015 December 2008 Technical Specification Bases ML0823907072008-08-21021 August 2008 Revised T.S. Pages Oyster Creek Nuclear Generating Station and Peach Bottom Atomic Power Station, Unit 3-Correction to Facility Operating Licenses ML0735400492007-12-19019 December 2007 Letter to C. Pardee Correction of Administrative Error in Amendment - the Consistency of TSs to Exelon Generation Company Quality Assurance TR (Tac. MD3937 Thru MD3940 and MD3943 Thru MD3948). - Rplacement Pages RS-07-132, Stations, Units 1 and 2 - Application for Steam Generator Tube Alternate Repair Criteria Technical Specification Amendment2007-11-29029 November 2007 Stations, Units 1 and 2 - Application for Steam Generator Tube Alternate Repair Criteria Technical Specification Amendment RS-07-044, Amergen Energy Company, LLC, Request for Amendment to Technical Specifications Section for the Inservice Testing Program2007-08-0808 August 2007 Amergen Energy Company, LLC, Request for Amendment to Technical Specifications Section for the Inservice Testing Program RS-07-087, Request for License Amendment Related to Technical Specification 5.5.2, Primary Coolant Sources Outside Containment.2007-07-31031 July 2007 Request for License Amendment Related to Technical Specification 5.5.2, Primary Coolant Sources Outside Containment. RS-07-020, Exelon/Amergen Application to Revise Technical Specifications Regarding Control Room Envelope Habitability in Accordance with TSTF-448, Revision 3, Using the Consolidated Line Item Improvement Process2007-04-12012 April 2007 Exelon/Amergen Application to Revise Technical Specifications Regarding Control Room Envelope Habitability in Accordance with TSTF-448, Revision 3, Using the Consolidated Line Item Improvement Process RS-06-181, Byron, Dresden, LaSalle, Peach Bottom, and Quad Cities Nuclear Plants, Request for Technical Specification Change to Provide Consistency with EGC Quality Assurance Topical Report2006-12-15015 December 2006 Byron, Dresden, LaSalle, Peach Bottom, and Quad Cities Nuclear Plants, Request for Technical Specification Change to Provide Consistency with EGC Quality Assurance Topical Report ML0633404872006-11-27027 November 2006 Tech Spec Pages for Amendments 148 & 142 Regarding Reactor Coolant System Pressure and Temperature Limits Report RS-04-125, Exelon/Amergen Request for Amendment to Technical Specifications Administrative Controls to Incorporate Requirement for Control Room Envelope Integrity Program2004-11-29029 November 2004 Exelon/Amergen Request for Amendment to Technical Specifications Administrative Controls to Incorporate Requirement for Control Room Envelope Integrity Program RS-04-067, Amergen, Request for Amendment to Technical Specifications Regarding Mode Change Limitations Using the Consolidated Line Item Improvement Process2004-04-30030 April 2004 Amergen, Request for Amendment to Technical Specifications Regarding Mode Change Limitations Using the Consolidated Line Item Improvement Process ML0406200682004-03-18018 March 2004 Technical Specifications Pages, Amendment No. 130 ML0212802982002-04-30030 April 2002 Byron/Braidwood, Units 1 and 2, Tech. Spec. Pages for Amendment Nos. 128 and 123 - Review Surveillance Requirement 3.0.3 to Extend the Delay Period, Before Entering a Limiting Condition for Operation Following a Missed Surveillance RS-02-045, Request for License Amendment for Technical Specifications Surveillance Requirement for Containment Spray Nozzles2002-04-19019 April 2002 Request for License Amendment for Technical Specifications Surveillance Requirement for Containment Spray Nozzles ML0211302632002-04-19019 April 2002 Units 1 & 2, Technical Specification Pages for License Amendment Nos. 127 & 122, Will Revise the Reactor Core Safety Limit for Peak Fuel Centerline Temperature 2023-09-29
[Table view] Category:Amendment
MONTHYEARRS-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-23-075, 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-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. ML22145A5862022-06-0303 June 2022 R. E. Ginna Correction to Pages Issued for Amendments Nos. 225, 225, 227, 227, & 148, Respectively, Regarding Issues Identified in Westinghouse Documents 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-22-037, 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-20-068, Application to Revise Technical Specifications 3.8.1, AC Sources-Operating2020-06-26026 June 2020 Application to Revise Technical Specifications 3.8.1, AC Sources-Operating JAFP-20-0049, Application to Adopt TSTF-427, Revision 2, Allowance for Non Technical Specification Barrier Degradation on Supported System Operability2020-06-22022 June 2020 Application to Adopt TSTF-427, Revision 2, Allowance for Non Technical Specification Barrier Degradation on Supported System Operability RS-19-039, Fleet License Amendment Request - Common Language for Technical Specification High Radiation Area Administrative Controls2019-06-26026 June 2019 Fleet License Amendment Request - Common Language for Technical Specification High Radiation Area Administrative Controls RS-15-091, Response to Request for Additional Information Regarding Request for a License Amendment to Braidwood Station, Units 1 and 2, Technical Specification 3.7.9, Ultimate Heat Sink2015-03-31031 March 2015 Response to Request for Additional Information Regarding Request for a License Amendment to Braidwood Station, Units 1 and 2, Technical Specification 3.7.9, Ultimate Heat Sink RS-09-071, Units 1 and 2 - License Amendment Request to Revise Technical Specifications (TS) for Permanent Alternate Repair Criteria2009-06-24024 June 2009 Units 1 and 2 - License Amendment Request to Revise Technical Specifications (TS) for Permanent Alternate Repair Criteria ML0823907072008-08-21021 August 2008 Revised T.S. Pages Oyster Creek Nuclear Generating Station and Peach Bottom Atomic Power Station, Unit 3-Correction to Facility Operating Licenses ML0735400492007-12-19019 December 2007 Letter to C. Pardee Correction of Administrative Error in Amendment - the Consistency of TSs to Exelon Generation Company Quality Assurance TR (Tac. MD3937 Thru MD3940 and MD3943 Thru MD3948). - Rplacement Pages RS-07-132, Stations, Units 1 and 2 - Application for Steam Generator Tube Alternate Repair Criteria Technical Specification Amendment2007-11-29029 November 2007 Stations, Units 1 and 2 - Application for Steam Generator Tube Alternate Repair Criteria Technical Specification Amendment RS-07-087, Request for License Amendment Related to Technical Specification 5.5.2, Primary Coolant Sources Outside Containment.2007-07-31031 July 2007 Request for License Amendment Related to Technical Specification 5.5.2, Primary Coolant Sources Outside Containment. RS-06-181, Byron, Dresden, LaSalle, Peach Bottom, and Quad Cities Nuclear Plants, Request for Technical Specification Change to Provide Consistency with EGC Quality Assurance Topical Report2006-12-15015 December 2006 Byron, Dresden, LaSalle, Peach Bottom, and Quad Cities Nuclear Plants, Request for Technical Specification Change to Provide Consistency with EGC Quality Assurance Topical Report ML0633404872006-11-27027 November 2006 Tech Spec Pages for Amendments 148 & 142 Regarding Reactor Coolant System Pressure and Temperature Limits Report RS-04-125, Exelon/Amergen Request for Amendment to Technical Specifications Administrative Controls to Incorporate Requirement for Control Room Envelope Integrity Program2004-11-29029 November 2004 Exelon/Amergen Request for Amendment to Technical Specifications Administrative Controls to Incorporate Requirement for Control Room Envelope Integrity Program RS-04-067, Amergen, Request for Amendment to Technical Specifications Regarding Mode Change Limitations Using the Consolidated Line Item Improvement Process2004-04-30030 April 2004 Amergen, Request for Amendment to Technical Specifications Regarding Mode Change Limitations Using the Consolidated Line Item Improvement Process ML0406200682004-03-18018 March 2004 Technical Specifications Pages, Amendment No. 130 ML0212802982002-04-30030 April 2002 Byron/Braidwood, Units 1 and 2, Tech. Spec. Pages for Amendment Nos. 128 and 123 - Review Surveillance Requirement 3.0.3 to Extend the Delay Period, Before Entering a Limiting Condition for Operation Following a Missed Surveillance RS-02-045, Request for License Amendment for Technical Specifications Surveillance Requirement for Containment Spray Nozzles2002-04-19019 April 2002 Request for License Amendment for Technical Specifications Surveillance Requirement for Containment Spray Nozzles ML0211302632002-04-19019 April 2002 Units 1 & 2, Technical Specification Pages for License Amendment Nos. 127 & 122, Will Revise the Reactor Core Safety Limit for Peak Fuel Centerline Temperature RS-02-028, Request for License Amendment for Technical Specifications - DC Electrical Power Systems2002-03-0808 March 2002 Request for License Amendment for Technical Specifications - DC Electrical Power Systems 2023-09-29
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4300 Winfield Road Warrenville, IL 60555 Exelon Generation 630 657 2000 Office RS-15-091 10 CFR 50.90 March 31, 2015 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. 50-373 and 50-374
Subject:
Response to Request for Additional Information Regarding Request for a License Amendment to Braidwood Station, Units 1 and 2, Technical Specification 3.7.9, "Ultimate Heat Sink"
References:
- 1) Letter from D. M. Gullott (Exelon Generation Company, LLC) to U. S. Nuclear Regulatory Commission, "Request for a License Amendment to Braidwood Station, Units 1 and 2, Technical Specification 3.7.9, 'Ultimate Heat Sink,"
dated August 19, 2014
- 2) Email from J. Wiebe (NRC) to J. Krejcie (Exelon Generation Company, LLC)
Preliminary RAls for LAR Regarding Braidwood Station Technical Specification, "Ultimate Heat Sink," dated February 5, 2015 In Reference 1, Exelon Generation Company, LLC, (EGC) requested an amendment to the Technical Specifications (TS) of Facility Operating License Nos. NPF-72 and NPF-77 for Braidwood Station, Units 1 and 2. The proposed amendment would modify TS 3.7.9, "Ultimate Heat Sink (UHS)," by changing the maximum allowable temperature of the UHS from 100 °F to a maximum UHS temperature of 102 °F.
In Reference 2, the U. S. Nuclear Regulatory Commission (NRC) requested additional information related to its review of Reference 1. In a teleconference with the NRC on February 18, 2015, the NRC indicated that certain questions originally transmitted in Reference 2 no longer required a response from EGC. Additionally, the NRC indicated that certain questions contained in Reference 2 were in the process of being revised. In subsequent conversations with the NRC, EGC agreed to provide the response to question 1 from Reference 2 at this time. Attachment 1 provides the requested information for question 1.
EGC has reviewed the information supporting a finding of no significant hazards consideration that was previously provided to the NRC in Attachment 1 of Reference 1. The additional information provided in this submittal does not affect the bases for concluding that the proposed license amendment does not involve a significant hazards consideration.
March 31, 2015 U.S. Nuclear Regulatory Commission Page 2 In accordance with 10 CFR 50.91, "Notice for public comment; State consultation," paragraph (b),
a copy of this letter and its attachments are being provided to the designated State of Illinois official.
Should you have any questions concerning this letter, please contact Ms. Jessica Krejcie at (630) 657-2816.
I declare under penalty of perjury that the foregoing is true and correct. Executed on the 31st day of March 2015.
Respectfully, David M. Gullott Manager Licensing Exelon Generation Company, LLC : Response to Request for Additional Information (Non-Proprietary) cc: NRC Regional Administrator, Region III NRC Senior Resident Inspector, Braidwood Station Illinois Emergency Management Agency Division of Nuclear Safety
ATTACHMENT 1 Response to Request for Additional Information By letter dated August 19, 2014 (ADAMS Accession No. ML14231A902), Exelon Generation Company, LLC, (EGC) submitted a license amendment request to raise the Braidwood Station, Unit 1 and Unit 2, Technical Specification Surveillance Requirement 3.7.9.2 limit of 100° F to 102° F (Reference 1). During review of the submittal, the U. S. Nuclear Regulatory Commission (NRC) staff identified information that is needed to support their review. These initial questions were provided to promote early communication of needed information (Reference 2).
In Reference 2, the NRC requested additional information related to its review of Reference 1.
In a teleconference with the NRC on February 18, 2015, the NRC indicated that certain questions originally transmitted in Reference 2 no longer required a response from EGC.
Additionally, the NRC indicated that certain questions contained in Reference 2 were in the process of being revised. In subsequent conversations with the NRC, EGC agreed to provide the response to question 1 from Reference 2 at this time. The requested information for question 1 is provided below.
NRC RAI 1 On page 17 of Attachment 1 to the August 19, 2014, letter, various safety analyses were identified that were not affected by the proposed [Ultimate Heat Sink] UHS temperature increase. Provide the rationale used to support the determination that the increased UHS temperature will not affect the results of the following safety analyses.
- LOCA Analyses including Large and Small Breaks Analyses, Long-Term-Cooling/Post-LOCA Boric Acid Precipitation, and LOCA Forces
- [Updated Final Safety Analysis Report] UFSAR Chapter 15 Transient Analyses for Various Non-LOCA Events
- Low Temperature Overpressure Protection (LTOP) Analysis
- Control Systems Operability/Margin to Trip/Component Sizing Evaluations Which Reviewed the Control and Protection System Setpoints to Verify that the System Capacities do not change The information to be provided should include for each of the applicable analyses (1) a listing of equipment that are directly or indirectly supported by the UHS, and were credited in the above safety analyses, (2) the assessment of the effects of increased UHS temperature on equipment identified in above item (1), and (3) the impacts of the increased UHS temperature on the affected safety analyses. For the non-LOCA transient analyses, the requested information should be provided for each of the applicable events discussed in UFSAR Chapter 15.
EGC Response to NRC RAI 1 In response to the information requested in items (1) and (2) described above, Braidwood Station UFSAR Table 15.0-7, "Plant Systems and Equipment Credited for Transients and Accident Conditions," identifies for each transient or accident the various plant equipment credited for accident response or mitigation. As discussed in section 2.2 of the LAR (Reference 1), the UHS dissipates heat after an accident through the cooling components of the Essential Service Water (SX) system and Component Cooling Water (CC) system, the principle systems at Braidwood Station that utilize the UHS to dissipate heat. Equipment cooled by the SX and Page 1 of 8
ATTACHMENT 1 Response to Request for Additional Information CC systems were evaluated for the UHS proposed IS maximum temperature of 102°F and maximum post-accident SX inlet temperature of 105.2°F and it was found that all equipment cooled by the SX and CC systems have adequate margin at the elevated UHS temperature and are acceptable without physical modification (see section 3.5 of Reference 1).
In response to the information requested in item (3) above, EGC has reviewed the various safety analyses as listed on page 17 of Attachment 1 of the License Amendment Request (LAR)
(Reference 1) and confirmed that the analyses were not impacted by the requested increase in UHS temperature. A summary is provided below describing how the assessment was performed to evaluate the increased UHS temperature's impact on the various safety analyses.
I. LOCA Analyses including Large and Small Breaks Analyses, Long-Term-Cooling/Post-LOCA Boric Acid Precipitation, and LOCA Forces Best Estimate (BE) Lame Break Loss of Coolant Accident (LBLOCA)
The current Best-Estimate LBLOCA Analysis of Record (AOR) for Braidwood Station, Unit 1 and 2 is reported in WCAP-16841-P, "Best Estimate Analysis of the Large Break Loss-of-Coolant Accident for Byron/Braidwood Nuclear Power Plant Using the ASTRUM Methodology."
This analysis does not model the UHS or SX system. The Emergency Core Cooling System (ECCS) and Containment Spray (CS) are modeled only in their Injection Phase when these systems take suction from the Refueling Water Storage Tank (RWST). The temperature of the water in the RWST is not affected by changes in the UHS temperature.
Small Break LOCA (SBLOCA)
The NOTRUMP Evaluation Model used in the SBLOCA analysis does not explicitly model the UHS. Auxiliary Feedwater (AF) is modeled in the SBLOCA analysis with a temperature of 125 'F. The SX system is the safety related backup to the AF system. Based on the results of the UHS temperature analysis, the AF temperature could reach a maximum of 105.2°F. This maximum temperature is bounded by the temperature used in the analysis.
The temperature of the safety injection water is assumed to be at 120 °F, based on the RWST as the source. The UHS temperature change does not impact the RWST. The temperature of the recirculation water is taken as 212 °F. Design analyses that were completed in support of the proposed UHS temperature increase have calculated the Residual Heat Removal (RHR) pump discharge temperature to be below 212 'F. The CC heat exchanger has been evaluated and has been found to be able to remove the required heat load that supports the assumptions of this calculation. Thus, the assumption of the SBLOCA analysis for the recirculation water temperature of 212°F is validated.
LOCA Long-Term Cooling/Post-LOCA Boric Acid Precipitation The post-LOCA long term core cooling critical boron concentration is determined at the most reactive time in life, assuming all rods out, no Xenon condition and a post-LOCA containment recirculation sump fluid temperature in the range of 68 212 °F. The recirculation sump fluid temperature has been calculated to be below 212°F in the calculation that was performed in support of the proposed UHS temperature increase. Thus, the UHS temperature increase does not impact the post-LOCA Long Term Cooling analyses.
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ATTACHMENT 1 Response to Request for Additional Information The Braidwood Station Hot leg Switchover analysis assumes that the temperature of the ECCS cooling water is at a temperature of 212 °F. During ECCS switchover operation post-LOCA, the Safety Injection (SI) and Chemical and Volume Control System (CVCS) pumps are supplied with water from the discharge of the RHR pump. Design analyses that were completed in support of the proposed UHS temperature increase have calculated the RHR pump discharge temperature to be below 212 °F. The CC heat exchanger has been evaluated and has been found to be able to remove the required heat load that supports the assumptions of this calculation. Support equipment for the SI and CVCS pumps (i.e., oil coolers and cubicle coolers); and for the RHR pumps (i.e., cubicle cooler) have been evaluated for the higher cooling water temperatures and were found to be acceptable (reference Sections 3.5.1 and 3.5.2 of Reference 1). Thus, the assumptions of the Hot Leg Switchover analysis are maintained by the increase in UHS temperature with no impact.
LOCA Forces The duration of the LOCA Forces transient considered is extremely short, on the order of 500 milliseconds to 1 second, and therefore does not model AF injection. In addition, the analysis primarily models the primary side reactor coolant system (RCS). Only steam generator (SG) feedwater and steam temperatures are modeled outside of the primary RCS. The change in UHS temperature does not impact RCS parameters nor feedwater or steam temperatures; therefore, the UHS temperature increase has no impact on the LOCA Forces analyses.
- 2. UFSAR Chapter 15 Transient Analyses for Various Non-LOCA Events The UHS temperature has the potential to impact non-LOCA events because the SX system is the safety related back-up to the AF system. Table 1 below shows the non-LOCA events in Chapter 15 of the Braidwood Station UFSAR and the AF assumptions in the analyses.
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ATTACHMENT 1 Response to Request for Additional Information Table 1 Review of UFSAR Transient Analyses for Non-LOCA Events Non-LOCA Safety Analyses and AF Modeling Event UFSAR AF Modeling Section Assumption Feedwater System Malfunctions Causing a Reduction in 15.1.1 Not modeled Feedwater Temperature Feedwater System Malfunctions Causing an Increase in 15.1.2 Not modeled Feedwater Flow Excessive Increase in Secondary Steam Flow 15.1.3 Not modeled Inadvertent Opening of a Steam Generator Relief or Safety Valve 15.1.4 Bounded by Steam System Steam System Piping Failure at Zero Power 15.1.5 Maximum Flow Steam System Piping Failure at Full Power 15.1.6 Not modeled Loss of External Load/Turbine Trip and Events Causing a 15.2.2 to Not modeled Turbine Trip (Results of Turbine Trip bounds all events in this 15.2.5 group)
Loss of Non-emergency AC Power to the Plant Auxiliaries 15.2.6 Minimum Flow Loss of Normal Feedwater Flow 15.2.7 Minimum Flow Feedwater System Pipe Break 15.2.8 Minimum Flow Partial Loss of Forced Reactor Coolant Flow 15.3.1 Not modeled Complete Loss of Forced Reactor Coolant Flow 15.3.2 Not modeled Reactor Coolant Pump Shaft Seizure (Locked Rotor) 15.3.3 Not modeled Reactor Coolant Pump Shaft Break 15.3.4 Locked Rotor with Loss of Offsite Power 15.3.5 (Analysis of Locked Rotor bounds this group)
Uncontrolled Rod Cluster Control Assembly Bank Withdrawal 15.4.1 Not modeled From a Subcritical or Low Power Startup Condition Uncontrolled Rod Cluster Control Assembly Bank Withdrawal At 15.4.2 Not modeled Power Rod Cluster Control Assembly Misoperation (System Malfunction 15.4.3 Not modeled or Operator Error)
Chemical and Volume Control System Malfunction that Results in 15.4.6 Not modeled a Decrease in Boron Concentration in the Reactor Coolant (Boron Dilution)
Inadvertent Loading and Operation of a Fuel Assembly in an 15.4.7 Not modeled Improper Position Spectrum of Rod Cluster Control Assembly Ejection Accidents 15.4.8 Not modeled Inadvertent Operation of the Emergency Core Cooling System 15.5.1 Not modeled During Power Operation (DNBR cases only)
Chemical and Volume Control System Malfunction that Increases 15.5.2 Not modeled Reactor Coolant Inventory Radioactive Release from A Subsystem or Component 15.7.1 Not modeled (waste gas system, liquid waste system, liquid tank, fuel handling to accidents, spent fuel cask drop accident) 15.7.5 Anticipated Transients Without Scram (ATWS) 15.8 Best estimate flow modeled Page 4 of 8
ATTACHMENT 1 Response to Request for Additional Information The impact of the higher UHS temperature on the design basis events which model AF (highlighted in yellow), (i.e., when AF is supplied by the SX system), are further discussed below. Note that the AF water enthalpy is assumed to be 80.2 Btu/lbm (based on an UHS temperature of 104 °F at the suction of the pump) when conservative in the revised analyses to support an increase in UHS temperature. Additionally, the revised analyses for Loss of Non-emergency AC Power to the Plant Auxiliaries, Loss of Normal Feedwater Flow, and Feedwater System Pipebreak incorporate a 12 second AF time delay. The AF time delay is incorporated in response to an NRC Non-Cited Violation (NCV) documented in Reference 3. Analyses were reperformed in response to the NCV to properly account for the time delay in the automatic switchover of the AF pump suction source from the Condensate Storage Tank (CST) to the SX system during a hypothetical catastrophic failure of the non-seismic CST.
Steam System Piping Failure at Zero Power In the steamline break analysis the energy removal from the RCS causes a reduction in coolant temperature and pressure. The negative moderator temperature coefficient adds positive reactivity to the core. Combined with the assumption that the most reactive rod cluster control assembly is stuck in the fully withdrawn position, there is an increased possibility that the core will become critical and return to power. The analysis shows that the core is ultimately shutdown by the boric acid injection from the safety injection system.
The major break of a steamline is the most limiting cooldown transient and is analyzed at Hot Zero Power with no decay heat.
In order to maximize the impact on the RCS cooldown, thus increasing positive reactivity addition to the core, the AF system is modeled at full flow with a minimum enthalpy of 0.03 Btu/lbm. This enthalpy bounds the enthalpy for a temperature of 32 °F. The impact of the increase in the AF temperature, when supplied by the SX system at 104 °F is bounded by the assumption of 32 °F for the AF temperature in the design basis analysis.
Loss of Nonemergency AC Power to the Plant Auxiliaries / Loss of Normal Feedwater Flow The Loss of Nonemergency AC Power to the Plant Auxiliaries (LOAC) and Loss of Normal Feedwater Flow (LONF) events are analyzed primarily to show that the heat removal capacity of the AF system is such that the RCS inventory is contained in the RCS boundary and margin to pressurizer filling is maintained.
The results show that margin to pressurizer filling is maintained for the reanalyzed LONF and LOAC cases. The overall limiting LONF case models the D5 SG (Braidwood Station Unit 2) with the CST as the AF source resulting in 3.7 ft3 margin to pressurizer filling. The overall limiting LOAC case models the BWI SG (Braidwood Station Unit 1) with the SX system as the AF source resulting in 142.5 ft3 margin to pressurizer filling.
As such, it is concluded that an AF enthalpy of 80.2 Btu/lbm (based on an ultimate heat sink temperature of 104 °F) is acceptable with respect to the LONF and LOAC events.
Subcase - LOAC Analysis is the LOAC with Reactor Coolant Pump (RCP) Seal Injection A sub-case to the LOAC analysis is the LOAC with RCP Seal Injection. This analysis was performed for Braidwood Station to address the scenario in which the letdown system is Page 5 of 8
ATTACHMENT 1 Response to Request for Additional Information unavailable after the loss of offsite power, but the charging pumps (and therefore RCP seal injection) are automatically loaded on to the diesel generators and start. This analysis is supplementary to the LOAC analysis performed for the UFSAR. The analysis is performed to address concerns identified in Westinghouse Nuclear Safety Advisory Letter (NSAL)00-013 regarding the potential to damage the pressurizer safety valves (PSVs) due to a net addition of coolant to the RCS. In this analysis, water relief data is calculated and used in an analysis of the valve integrity. The objective is to demonstrate that no valve damage will occur and that the valves will remain operable.
The LOAC with RCP seal injection event is analyzed to show that the PSVs will remain operable while relieving steam and/or water over a finite number of relief cycles at a minimum water relief temperature. The analysis demonstrates that the PSVs will reseat properly following the cycling and ensures their continued operability.
The transient is analyzed for one hour because plant operating procedures ensure that appropriate actions will be taken to terminate the water relief through the PSVs within one hour after transient initiation.
The results of the limiting scenario developed show that, for one hour after transient initiation, the PSV cycling conditions are acceptable and confirm the PSVs will not be damaged. The results of the revised analysis for the higher UHS temperature show that the number of relief cycles and the minimum water relief temperature are equal or bounded by the number of cycles and minimum water temperature from the AOR. Based on these results, it is concluded that an AF enthalpy of 80.2 Btu/lbm (based on an ultimate heat sink temperature of 104 °F), is acceptable with respect to the LOAC with RCP seal injection event.
Feedwater System Pipe Break The Feedwater Line Break (FLB) event is analyzed primarily to show that core cooling is maintained in the timeframe between the initiation of the FLB with resultant loss of main feedwater and the time when the heat removal capability of the AF system is greater than the heat addition being provided by the core decay heat (plus RCP heat for the with-power case).
The time when the AF decay heat removal capability exceeds the heat generated by the primary-side is the time of transient turnaround. As shown in the analysis, core cooling is confirmed by demonstrating that the hot leg temperatures in the RCS loops with the intact SGs remain subcooled prior to turnaround.
The results for the reanalyzed cases show that no bulk boiling occurs in the RCS because margin is maintained to hot leg saturation such that core cooling is maintained.
Overall, the AOR limiting cases are those where the CST is the water source for AF. The limiting case (with offsite power available) models the D5 SG (i.e., Braidwood Station Unit 2),
resulting in 0.63 °F margin to hot leg saturation. The limiting case (with loss of offsite power) also models the D5 SG (i.e., Braidwood Station Unit 2), resulting in 24 °F margin to hot leg saturation. The results of the analysis for the higher SX temperature show a margin of 3.11 °F for the offsite power available case and 24.73 °F for the loss of offsite power case.
In summary, the analysis shows that an AF enthalpy of 80.2 Btu/lbm (based on an ultimate heat sink temperature of 104 °F), is acceptable with respect to the FLB event.
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ATTACHMENT 1 Response to Request for Additional Information Anticipated Transient without Scram (ATWS)
A best estimate AF flow is modeled in this analysis. This is a beyond design basis event, and no single failure is required to be considered in the analysis. Additionally, it is permissible to assume that non-safety grade equipment such as the ATWS Mitigating System Actuation Circuitry (AMSAC) system and SG steam dump system are operable. It is therefore not necessary to assume a concurrent failure of the CST at the time that the ATWS occurs. As such, the current analysis remains applicable and the CST failure scenario need not be considered with respect to the ATWS event. The UHS higher temperature does not have any impact on the temperature of the water in the CST.
- 3. Low Temperature Overpressure Protection (LTOP) Analysis The LTOP analysis is applicable for the RCS at low temperatures. in Operational Modes 4, 5 and 6 with the Reactor Vessel Head on the main flange.
The Braidwood Station design basis LTOP analysis addresses both a mass injection (MI) and heat injection cases (HI). The design basis MI event assumes a system failure, whereby the charging control valve fails open and, simultaneously, the letdown line flow control valve fails closed. The limiting mass input to the RCS is based on a single centrifugal charging pump in operation. The mass input rates are determined using conservative pump flow rates and a density of 62 Ibm/ft3 for the charging flow. This density corresponds to a charging water temperature of 100 °F.
The Letdown Heat Exchanger is cooled by CC. The CC heat exchanger is cooled by the SX system from the UHS. Increases in letdown water temperatures may result in higher charging temperatures. This does not impact the results of the analysis as higher charging temperatures would result in lower mass input rates.
The design basis HI event assumes the inadvertent startup of a RCP while the plant primary system is in a water solid condition and the SGs are at a temperature of 50 °F higher than the remainder of the RCS. It is assumed that both the primary side water in the SG secondary side fluids are 50 °F hotter than the remainder of the RCS. The UHS temperature increase does not have any impact on the parameters used in this analysis.
- 4. Evaluation of the Reactor Coolant Pump and Reactor Coolant Pump Motor Limits on Component Cooling and Seal Injection Temperatures An increase in UHS temperature could affect the RCPs and RCP motors if it causes the temperature of the CC serving the RCPs and RCP motors to increase, or if seal injection temperature increases. This could be a further-downstream effect of the higher UHS temperature. Limits on maximum CC and seal injection temperatures for the RCP and RCP motor are set forth in the Westinghouse RCP Instruction and Operating Book and are reflected in Braidwood Station operating procedures. The UHS temperature increase does not affect the temperature limits of the CC water or the seal injection water to the running RCPs.
The CC heat exchangers outlet temperature is limited to 105°F during normal power operation and 120°F after initiation of RHR for a normal RCS cooldown, with no accident in progress on either unit. For normal plant operation, the CC heat exchangers have been evaluated for a maximum SX supply temperature of 102 °F; the heat exchangers are able to maintain CC Page 7 of 8
ATTACHMENT 1 Response to Request for Additional Information temperature below the current established limits. CC and seal injection temperatures to the RCPs would then be within limits.
For design basis accident conditions, for the unit undergoing a normal shutdown while the opposite unit is experiencing a LOCA, the temperature of the CC water was allowed to increase to a maximum of 128 °F as part of the Measurement Uncertainty Recapture (MUR) implementation project and it is part of the Braidwood Station design basis. Evaluations have been completed for an SX temperature of 105.2 °F in support of the UHS LAR. The results of this evaluation show the CC heat exchanger outlet temperature remains below the established limit of 128 °F. Using a temperature of 105.2 °F for this case is conservative because the UHS temperature profile, for the limiting case, shows the calculated temperature remains at 102 °F and below for at least 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> into the event. This period is more than sufficient to take the unit through the normal cooldown to the RHR entry conditions.
For the LOCA Unit, the maximum CC heat exchanger temperature increases to 128 °F in order to account for the added heat load due to the ECCS Recirculation Phase. This increase does not have any impact on the RCPs because the Emergency Response Procedures have a step to stop all RCPs early in the event (Reference 1(2)BwEP-0, "Reactor Trip or Safety Injection").
In addition, the LBLOCA results in the isolation of the CC to the RCPs due to a Phase B isolation signal on a high containment pressure (20 psig).
- 5. Control Systems Operability/Margin to Trip/Component Sizing Evaluations Which Reviewed the Control and Protection System Setpoints to Verify that the System Capacities do not change An increase in the maximum UHS temperature does not impact the control systems operability because the control systems logic, setpoints and time constants, and the NSSS capacities and performance capabilities are not being revised. Furthermore, the increased UHS temperature does not affect the full power design conditions.
Therefore, the control systems performance during steady state and during and following the design basis operational transients is not affected. Since the control and protection system setpoints, time constants and capacities do not change, the increased UHS temperature will have no effect on the available plant operational margins. The control system setpoints do not need to be revised.
REFERENCES
- 1. Letter from D. M. Gullott (Exelon Generation Company, LLC) to U. S. Nuclear Regulatory Commission, "Request for a License Amendment to Braidwood Station, Units 1 and 2, Technical Specification 3.7.9, 'Ultimate Heat Sink," dated August 19, 2014
- 2. Email from J. Wiebe (NRC) to J. Krejcie (Exelon Generation Company, LLC) Preliminary RAls for LAR Regarding Braidwood Station Technical Specification, "Ultimate Heat Sink" dated February 5, 2015
- 3. Letter from US NRC to M.J. Pacilio, "Braidwood Station, Units 1 and 2, NRC Biennial Problem Identification and Resolution Inspection Report 05000456/2010006; 0500457/2010006," dated October 27, 2010.
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