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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:Response to Request for Additional Information (RAI)
MONTHYEARRS-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-23-056, 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-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-22-084, 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-22-074, 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-22-027, 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 ML21320A2422021-11-16016 November 2021 Stations - Response to Request for Additional Information - Proposed Alternative for Examination of Pressurizer Circumferential and Longitudinal Shell-to-Head Welds and Nozzle-to-Vessel Welds JAFP-21-0087, Response to Request for Additional Information Regarding Request for Approval of Transfer of Licenses and Conforming Amendments2021-09-16016 September 2021 Response to Request for Additional Information Regarding Request for Approval of Transfer of Licenses and Conforming Amendments ML21225A0062021-08-13013 August 2021 Response to Request for Additional Information Regarding License Amendment Request to Address the Issues Identified in Westinghouse Documents NSAL-09-5, Rev.1 and NSAL-15-1, Rev. 0 NMP2L2773, Company - Response to Request for Additional Information2021-06-30030 June 2021 Company - Response to Request for Additional Information JAFP-21-0044, Response to Request for Additional Information Regarding Request for Approval of Transfer of Licenses and Conforming Amendments2021-06-11011 June 2021 Response to Request for Additional Information Regarding Request for Approval of Transfer of Licenses and Conforming Amendments JAFP-21-0032, Response to Request for Additional Information - Proposed Alternative Concerning ASME Section XI Repair/Replacement Documentation for Replacement of Pressure Retaining Bolting2021-04-20020 April 2021 Response to Request for Additional Information - Proposed Alternative Concerning ASME Section XI Repair/Replacement Documentation for Replacement of Pressure Retaining Bolting RS-21-015, Response to Request for Additional Information Related to Relief Request for Braidwood I4R-11 and Byron I4R-182021-02-10010 February 2021 Response to Request for Additional Information Related to Relief Request for Braidwood I4R-11 and Byron I4R-18 RS-21-019, Response to Request for Additional Information Regarding Application for Revision to TS 5.5.9, Steam Generator (SG) Program, for a One-Time Deferral of Steam Generator Tube Inspections2021-02-0909 February 2021 Response to Request for Additional Information Regarding Application for Revision to TS 5.5.9, Steam Generator (SG) Program, for a One-Time Deferral of Steam Generator Tube Inspections RS-20-017, Response to Request for Additional Information - Proposed Alternative to Utilize Code Case N-8852021-02-0101 February 2021 Response to Request for Additional Information - Proposed Alternative to Utilize Code Case N-885 RS-20-147, Response to Request for Additional Information Related to Exemption from Part 26 Requirements2020-11-23023 November 2020 Response to Request for Additional Information Related to Exemption from Part 26 Requirements RS-20-143, Response to Request for Additional Information Regarding Application to Revise Technical Specifications 3.8.1, AC Sources- Operating2020-11-20020 November 2020 Response to Request for Additional Information Regarding Application to Revise Technical Specifications 3.8.1, AC Sources- Operating ML20188A2642020-07-0606 July 2020 Clinton Power Station, R.E. Ginna Station, Limerick Station, Nine Mile Point Station & Peach Bottom Station - Proposed Alternative to Utilize Code Case OMN-26 - Response to Request for Additional Information RS-20-057, Response to Request for Additional Information Regarding Emergency License Amendment Request for a One-Time Extension of the Steam Generator Tube Inspections2020-04-13013 April 2020 Response to Request for Additional Information Regarding Emergency License Amendment Request for a One-Time Extension of the Steam Generator Tube Inspections RS-20-012, Response to Request for Additional Information Regarding Relief Request Associated with the Third Ten-Year Inservice Inspection Program Interval2020-01-29029 January 2020 Response to Request for Additional Information Regarding Relief Request Associated with the Third Ten-Year Inservice Inspection Program Interval NMP2L2711, Byron Station; Calvert Cliffs; Clinton Power Station; LaSalle County Station; Limerick Generating Station; and Nine Mile Point Nuclear Station - Proposed Alternative to Utilize Code Case N-8792019-10-16016 October 2019 Byron Station; Calvert Cliffs; Clinton Power Station; LaSalle County Station; Limerick Generating Station; and Nine Mile Point Nuclear Station - Proposed Alternative to Utilize Code Case N-879 RS-19-100, Response to Request for Additional Information Regarding Application to Revise Technical Specifications to Adopt Risk Informed Completion Times TSTF-505, Revision 2, Provide Risk-Informed Extended.2019-10-0707 October 2019 Response to Request for Additional Information Regarding Application to Revise Technical Specifications to Adopt Risk Informed Completion Times TSTF-505, Revision 2, Provide Risk-Informed Extended. RS-19-088, Response to Request for Additional Information Regarding Application to Revise Braidwood Station and Byron Station Technical Specifications to Adopt Risk Informed...2019-09-0505 September 2019 Response to Request for Additional Information Regarding Application to Revise Braidwood Station and Byron Station Technical Specifications to Adopt Risk Informed... JAFP-19-0057, Response to Request for Additional Information - Proposed Alternative to Utilize Code Cases N-878 and N-8802019-06-0404 June 2019 Response to Request for Additional Information - Proposed Alternative to Utilize Code Cases N-878 and N-880 RS-19-051, Response to Request for Additional Information for the License Amendment Request to Utilize Tvel TVS-K Lead Test Assemblies2019-05-30030 May 2019 Response to Request for Additional Information for the License Amendment Request to Utilize Tvel TVS-K Lead Test Assemblies NMP1L3279, Response to Request for Additional Information - Proposed Alternatives to Utilize Code Cases N-878 and N-880 for Plants2019-05-0101 May 2019 Response to Request for Additional Information - Proposed Alternatives to Utilize Code Cases N-878 and N-880 for Plants JAFP-19-0006, Response to Request for Additional Information - Proposed Alternative to Utilize Code Cases N-878 and N-8802019-01-0808 January 2019 Response to Request for Additional Information - Proposed Alternative to Utilize Code Cases N-878 and N-880 RS-18-143, Supplemental Information Supporting License Amendment Requests for Approval of Changes to Emergency Plan Staffing Requirements2018-11-29029 November 2018 Supplemental Information Supporting License Amendment Requests for Approval of Changes to Emergency Plan Staffing Requirements RS-18-094, Response to Request for Additional Information Regarding Braidwood Station Fourth Lnservice Inspection Interval Relief Request 14R-062018-07-31031 July 2018 Response to Request for Additional Information Regarding Braidwood Station Fourth Lnservice Inspection Interval Relief Request 14R-06 RS-18-095, Response to Request for Additional Information Regarding Braidwood Station Fourth Lnservice Inspection Interval Relief Request 14R-032018-07-31031 July 2018 Response to Request for Additional Information Regarding Braidwood Station Fourth Lnservice Inspection Interval Relief Request 14R-03 RS-18-097, Response to Request for Additional Information and Supplemental Information-License Amendment.2018-07-27027 July 2018 Response to Request for Additional Information and Supplemental Information-License Amendment. RS-18-084, Response to Request for Additional Information Regarding Braidwood TORMIS Amendment Request2018-07-0909 July 2018 Response to Request for Additional Information Regarding Braidwood TORMIS Amendment Request RS-18-083, Response to Request for Additional Information Regarding Braidwood Station Fourth Inservice Inspection Interval Relief Request 14R-052018-06-22022 June 2018 Response to Request for Additional Information Regarding Braidwood Station Fourth Inservice Inspection Interval Relief Request 14R-05 RS-18-072, Response to Request for Additional Information Regarding 50.69 Amendment2018-06-13013 June 2018 Response to Request for Additional Information Regarding 50.69 Amendment RS-18-061, Response to Request for Additional Information Regarding Decommissioning Funding Plans for Independent Spent Fuel Storage Installations (Isfsis)2018-05-0202 May 2018 Response to Request for Additional Information Regarding Decommissioning Funding Plans for Independent Spent Fuel Storage Installations (Isfsis) RS-18-041, Response to Request for Additional Information Regarding Request for Relief (I4R-07) for Extension of Examination Interval for Reactor Pressure Vessel Head Penetration Nozzles with Mitigated Alloy 600/82/182 Peened ...2018-04-12012 April 2018 Response to Request for Additional Information Regarding Request for Relief (I4R-07) for Extension of Examination Interval for Reactor Pressure Vessel Head Penetration Nozzles with Mitigated Alloy 600/82/182 Peened ... RS-17-138, Stations - Response to Request for Additional Information Related to License Amendment Request to Incorporate Changes Supported by TSTF-547, Clarification of Rod Position Requirements'2017-10-25025 October 2017 Stations - Response to Request for Additional Information Related to License Amendment Request to Incorporate Changes Supported by TSTF-547, Clarification of Rod Position Requirements' RS-17-084, Response to Request for Additional Information for Byron Station and Braidwood Station Post Peening Relief Requests2017-07-14014 July 2017 Response to Request for Additional Information for Byron Station and Braidwood Station Post Peening Relief Requests RS-17-085, Response to Request for Additional Information Regarding Amendments to Remove Obsolete License and Technical Specification Items2017-06-29029 June 2017 Response to Request for Additional Information Regarding Amendments to Remove Obsolete License and Technical Specification Items ML17178A0762017-06-26026 June 2017 Security Plan, Training and Qualification Plan, and Safeguards Contingency Plan, Revision 15; Supplemental Information Letter RS-17-044, Response to Request for Additional Information Regarding Proposed Alternative for the Use of Encoded Phased Array Ultrasonic Examination Techniques in Lieu of Radiography2017-03-13013 March 2017 Response to Request for Additional Information Regarding Proposed Alternative for the Use of Encoded Phased Array Ultrasonic Examination Techniques in Lieu of Radiography RS-16-213, Response to Request for Additional Information Regarding Relief Request 13R-17 to Extend the Reactor Pressure Vessel Inservice Inspection Interval2016-11-21021 November 2016 Response to Request for Additional Information Regarding Relief Request 13R-17 to Extend the Reactor Pressure Vessel Inservice Inspection Interval RS-16-231, Revised Response to Request for Additional Information for License Amendment Request for a One-Time Extension of the Essential Service Water (SX) Train Completion Time to Support 2A SX Pump Repair2016-11-14014 November 2016 Revised Response to Request for Additional Information for License Amendment Request for a One-Time Extension of the Essential Service Water (SX) Train Completion Time to Support 2A SX Pump Repair RS-16-222, Response to Request for Additional Information Related to License Amendment Request for a One-Time Extension of the Essential Service Water (SX) Train Completion Time to Support 2A SX Pump Repair2016-10-28028 October 2016 Response to Request for Additional Information Related to License Amendment Request for a One-Time Extension of the Essential Service Water (SX) Train Completion Time to Support 2A SX Pump Repair RS-16-218, Response to Request for Additional Information License Amendment Request for a One-Time Extension of the Essential Service Water Train Completion Time to Support 2A SX Pump Repair2016-10-26026 October 2016 Response to Request for Additional Information License Amendment Request for a One-Time Extension of the Essential Service Water Train Completion Time to Support 2A SX Pump Repair RA-16-049, Response to Request for Additional Information Regarding Requests to Withhold Emergency Preparedness Documents from Public Disclosure2016-05-26026 May 2016 Response to Request for Additional Information Regarding Requests to Withhold Emergency Preparedness Documents from Public Disclosure RS-16-112, Response to Request for Additional Information Regarding Request for a License Amendment to Technical Specification 3.7.9, Ultimate Heat Sink.2016-04-29029 April 2016 Response to Request for Additional Information Regarding Request for a License Amendment to Technical Specification 3.7.9, Ultimate Heat Sink. RS-16-079, Response to Request for Additional Information Regarding the Request for Exemption from 10 CFR 50.55a, Codes and Standards, Paragraph (h)(2), Protection Systems, Including Change to Request NRC Approval Under 10 CFR 50.55a..2016-04-11011 April 2016 Response to Request for Additional Information Regarding the Request for Exemption from 10 CFR 50.55a, Codes and Standards, Paragraph (h)(2), Protection Systems, Including Change to Request NRC Approval Under 10 CFR 50.55a.. RS-16-078, Clarification to Response to Requests for Additional Information for Relief for Alternate Requirements for Repair of Reactor Vessel Head Penetrations with Nozzles Having Pressure-Retaining Partial-Penetration J-Groove Welds2016-04-0404 April 2016 Clarification to Response to Requests for Additional Information for Relief for Alternate Requirements for Repair of Reactor Vessel Head Penetrations with Nozzles Having Pressure-Retaining Partial-Penetration J-Groove Welds 2023-04-20
[Table view] |
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ExeLon G Warrenville, IL 60555 www.exeloncorp.com RS-16-213 10 CFR 50.55a November 21, 2016 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-001 Braidwood Station, Units 1 and 2 Renewed Facility Operating License Nos. NPF-72 and NFP-77 NRC Docket Nos. STN-50-456 and STN 50-457
Subject:
Response to Request for Additional Information Regarding Relief Request 13R-17 to Extend the Reactor Pressure Vessel Inservice Inspection Interval
References:
- 1) Letter from D. M. Gullott (Exelon Generation Company, LLC) to NRC, "Relief Request 13R-17 to Extend the Reactor Pressure Vessel Inservice Inspection Interval," dated July 21, 2016
- 2) Email form J. Wiebe (NRC) to J. A. Bauer (Exelon Generation Company, LLC), "Preliminary RAls for Braidwood Station Relief Request 13R-17," dated October 6, 2016
- 3) WCAP-16168-NP-A, Revision 3, "Risk-Informed Extension of the Reactor Vessel In-Service Inspection Interval," dated October 2011
- 4) Letter from R. A. Nelson (NRC) to W. A. Nowinowski (PWROG), "Revised Final Safety Evaluation by the Office of Nuclear Reactor Regulation Regarding Pressurized Water Reactor Owners Group Topical Report WCAP-16168-NP-A, Revision 2, 'Risk-Informed Extension of the Reactor Vessel In-Service Inspection Interval,"' dated July 26, 2011 In the Reference 1, Exelon Generation Company, LLC (EGC) requested approval of Relief Request 13R-17 in accordance with 10 CFR 50.55x, "Codes and standards," paragraph (z)(1).
Relief Request 13R-17 is associated with the Third 10-year Inservice Inspection (ISI) Program Interval for Braidwood Station, Units 1 and 2. The third interval of the Braidwood Station ISI program complies with the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (B&PV) Code,Section XI, "Rules for Inservice Inspection of Nuclear Power Plant Components," 2001 Edition through the 2003 Addenda. The Braidwood Station Third 10-year ISI intervals are currently scheduled to end on July 28, 2018 for Unit 1 and October 16, 2018 for Unit 2.
November 21, 2016 U. S. Nuclear Regulatory Commission Page 2 NRC approval of Relief Request 13R-17 submitted in Reference 1 is requested based on the justification provided in topical report WCAP-16168-NP-A, Revision 3 (Reference 3). This topical report demonstrated that extending the ASME B&PV Code,Section XI ISI interval from the current 10 years to 20 years for reactor pressure vessel (RPV) pressure containing welds is an alternative inspection interval that provides an acceptable level of quality and safety. NRC approval of WCAP-16168 is documented in Reference 4.
In Reference 2, the NRC requested additional information related to Relief Request 13R-17. It was agreed that the response to the request for additional information would be submitted to the NRC on or before November 28, 2016. The subject response is presented in Attachment 1.
As noted in Reference 1, EGC requested approval of this relief request by July 21, 2017.
There are no regulatory commitments contained in this letter. Should you have any questions concerning this letter, please contact Joseph A. Bauer at (630) 657-2804.
Respectfully, David M. Gullott Manager Licensing Exelon Generation Company, LLC : Response to Request for Additional Information Regarding Relief Request 13R-17 to Extend the Reactor Pressure Vessel Inservice Inspection Interval cc: NRC Regional Administrator, Region III NRC Senior Resident Inspector, Braidwood Station NRR Project Manager, Braidwood Station Illinois Emergency Management Agency Division of Nuclear Safety
ATTACHMENT 1 Response to Request for Additional Information Regarding Relief Request 13R-17 to Extend the Reactor Pressure Vessel Inservice Inspection Interval Page 1 of 7 In the Reference 1, Exelon Generation Company, LLC (EGC) requested approval of Relief Request 13R-17 in accordance with 10 CFR 50.55x, "Codes and standards," paragraph (z)(1).
Relief Request 13R-17 is associated with the Third 10-year Inservice Inspection (ISI) Program Interval for Braidwood Station, Units 1 and 2. The third interval of the Braidwood Station ISI program complies with the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (B&PV) Code,Section XI, "Rules for Inservice Inspection of Nuclear Power Plant Components," 2001 Edition through the 2003 Addenda. The Braidwood Station Third 10-year ISI intervals are currently scheduled to end on July 28, 2018 for Unit 1 and October 16, 2018 for Unit 2.
NRC approval of Relief Request 13R-17 submitted in Reference 1 is requested based on the justification provided in topical report WCAP-16168-NP-A, Revision 3 (Reference 3). This topical report demonstrated that extending the ASME B&PV Code,Section XI ISI interval from the current 10 years to 20 years for reactor pressure vessel (RPV) pressure containing welds is an alternative inspection interval that provides an acceptable level of quality and safety.
NRC approval of WCAP-16168 is documented in Reference 4.
In Reference 2, the NRC requested additional information related to Relief Request 13R-17. It was agreed that the response to the request for additional information would be submitted to the NRC on or before November 28, 2016. The subject response is presented below.
NRC Request for Additional Information In reviewing the Exelon Generation Company, LLC's (Exelon's) submittal dated July 21, 2016 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML16203A081), related to request for alternative 13R-17, for the Braidwood Station (Braidwood),
Units 1 and 2, the NRC staff has determined that the following information is needed in order to complete its review:
RAI-1
Request for Alternative 13R-17 is based on WCAP-16168, "Risk-Informed Extension of the Reactor Vessel In-Service Inspection Interval" (ADAMS Accession No. ML113060207). The revised safety evaluation (SE) for WCAP-16168 dated July 26, 2011 (ADAMS Accession Number ML111600303), states in required plant-specific information Item 3 that, "The 20-year inspection interval is a maximum interval. "
The licensee stated in its July 21, 2016, request for alternative 13R-17, "Upon approval of this relief request, these required [reactor pressure vessel (RPV) weld] examinations would be performed during the Fourth ISI Interval for each Unit in 2027 plus or minus one refueling outage. " The licensee also states that, these dates are consistent with those provided in Pressurized Water Reactor Owners Group (PWROG) letter OG-06-356 (ADAMS Accession No. ML082210245) and the latest implementation plan provided in PWROG letter OG-10-238 (ADAMS Accession Number ML11153A033). If this request for an alternative is approved, the third ISI interval will be extended to July 27, 2028 for Unit 1 and October 15, 2028 for Unit 2. As such, the proposed inspection date of 2027, plus or minus one refueling outage, could be within
ATTACHMENT 1 Response to Request for Additional Information Regarding Relief Request 13R-17 to Extend the Reactor Pressure Vessel Inservice Inspection Interval Page 2 of 7 the implementation plan dates but potentially be beyond the requested extended intervals, which would put them beyond the 20 year maximum interval. Confirm that the required RPV inspections will be performed before the approved extension dates of July 27, 2028 for Unit 1 and October 15, 2028 for Unit 2.
Response to RAI-1 Based on the current refueling outage dates projection, Braidwood Station, Unit 1 is scheduled to have a refueling outage in Spring 2027 (Al R26); and Unit 2 is scheduled to have a refueling outage in Fall 2027 (A2R26). As noted above, conducting the required weld examinations in 2027 plus one refueling outage (i.e., Fall 2028 (Al R27) for Unit 1; and Spring 2029 (A2R27) for Unit 2) would extend the examination date beyond the 20 year maximum interval (i.e., July 27, 2028 for Unit 1 and October 15, 2028 for Unit 2).
EGC confirms that the subject weld examinations will be performed no later than the completion of the Spring 2027 refueling outage (Al R26) for Unit 1; and Fall 2027 refueling outage (A2R26) for Unit 2.
RAI-2
The licensee in Tables 3-1 and 3-2 of the July 21, 2016, request for alternative provides information pertaining to RPV inspections for Braidwood, Units 1 and 2. It was stated in both tables that the indications (3 for Unit 1 and 2 for Unit 2) are acceptable per Table I WB-2510-1 of Section Xl of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code. Also, Table 3-2 listed flaw numbers under "Scaled maximum number of forging flaws" along with the detected number of forging flaws.
- a. Provide the dimensions of the 5 indications mentioned above to allow the staff to confirm that they are acceptable per Table IWB-2510-1.
- b. Provide information on the methodology used to determine the scaled maximum number of forging flaws.
Response to RAI-2a The dimensions of the three Unit 1 indications, identified in Table 3-1; and the two Unit 2 indications, identified in Table 3-2, are presented below.
Braidwood Unit 1 (Three Indications)
Indication 1 (Weld Number 1 RV-01-004), Upper Shell to Intermediate Circumferential (Circ)
Weld The indication was classified as a planar subsurface indication 0.5" long, 0.57" from the reactor pressure vessel (RPV) outside diameter (OD) surface with a through wall extent of 0.19" in an area with a design thickness of 8.63".
ATTACHMENT 1 Response to Request for Additional Information Regarding Relief Request 13R-17 to Extend the Reactor Pressure Vessel Inservice Inspection Interval Page 3 of 7 Indication 2 (Weld Number 1 RV-01-003), Intermediate Circ Weld to Lower Circ Weld The indication was classified as a planar subsurface indication 1.5" long, 3.89" from the RPV OD surface with a through wall extent of 0.16" in an area with a design thickness of 8.63".
Indication 3 (Weld Number 1 RV-01-003), Intermediate Circ Weld to Lower Circ Weld The indication was classified as a planar subsurface indication 1.0" long, 3.44" from the RPV inside diameter (ID) surface with a through wall extent of 0.27" in an area with a design thickness of 8.63".
Braidwood Unit 2 (Two Indications)
Indication 1 (Weld Number 2RV-01-004), Upper Shell to Intermediate Circ Weld The indication was classified as a planar subsurface indication 4.5" long, 0.51" from the RPV OD surface with a through wall extent of 0.32" in an area with a design thickness of 8.63" (actual thickness is 9.09", but the more conservative thickness was used for the evaluation).
Indication 2 (Weld Number 2RV-01-003), Intermediate Circ Weld to Lower Circ Weld The indication was classified as a planar subsurface indication 2.1" long, 0.1" from the RPV ID surface with an assigned through wall extent of 0.125" (based on the Ultrasonic Testing (UT) data, the indication did not have any through wall extent; when this occurs the procedure assigns the resolution capability of the transducer as the through wall extent).
The thickness based on design information equals 8.63".
Response to RAI-2b The methodology used to determine the scaled maximum number of forging flaws is discussed in a Westinghouse Electric Company, LLC (i.e., Westinghouse) technical report prepared for Braidwood Station, Units 1 and 2. This report addressed the implementation of NRC-approved technical report, WCAP-16168-NP-A, "Risk-Informed Extension of the Reactor Vessel In-Service Inspection Interval," Revision 3, dated October 2011 ( Reference 3). The methodology is described below.
Reactor vessel inservice inspection data must be reviewed to confirm that satisfactory inspections have been performed on the reactor vessel and that the flaw distributions used in the pilot plant analyses are bounding. In accordance with the analyses in NUREG-1874, "Recommended Screening Limits for Pressurized Thermal Shock (PTS)," March 2010, which is the probabilistic fracture mechanics basis for the Alternate PTS Rule (i.e., 10 CFR 50.61 a, "Alternate fracture toughness requirements for protection against pressurized thermal shock events"), the only flaws of concern are those that are within the inner 3/8th of the reactor vessel thickness and in the beltline region adjacent to the reactor core.
ATTACHMENT 1 Response to Request for Additional Information Regarding Relief Request 13R-17 to Extend the Reactor Pressure Vessel Inservice Inspection Interval Page 4 of 7 These flaws must meet the allowable sizes in Table IWB-3510-1 of Section XI of the ASME Code. Furthermore, only those flaws, found in an ASME Section XI, Appendix VIII inspection, in the inner 1/1 0th thickness or 1 inch, whichever is greater, must meet the allowable flaw limits of Tables 4.5-1 and 4.5-2.
The following process is followed for each flaw (recordable indication) identified in one of the welds that is at least partially in the beltline region:
- To determine if the flaw is in the inner 3/8th thickness of the reactor vessel: The flaw "S" dimension is compared to the RV thickness (at the location of the flaw) multiplied by 0.375.
- To determine if the limits of Tables 4.5-1 and 4.5-2 apply, the flaw "S" dimension is compared to the greater of 1/1 0th of the reactor vessel thickness or 1 inch.
- To determine if the flaw is in the weld or the adjacent forging material (only required if the ISI report does not explicitly state the location of the flaw):
o Axial Welds - The difference between the angular position of the flaw and the weld centerline is determined. This difference in angular position is used along with the reactor vessel inside diameter and flaw depth to determine the offset of the flaw from the weld centerline. This offset is then compared to Y2 of the weld width. If the offset is greater than 1/2 of the weld width, the flaw is determined to be in the forging material. If the offset is less than % of the weld width, the flaw is determined to be in the weld material.
o Circumferential Welds The offset or difference between the height of the flaw and the weld centerline is determined. This offset is then compared to'/2 of the weld width. If the offset is greater than 1/2 of the weld width, the flaw is determined to be in the forging material. If the offset is less than % of the weld width, the flaw is determined to be in the weld material.
To determine the orientation of the flaw: For axial welds, ISI scan direction UP/DN yields a parallel scan and CW/CCW yields a perpendicular scan and, for circumferential welds, ISI scan direction UP/DN yields a perpendicular scan and CW/CCW yields a parallel scan. The designations are defined as follows: UP is Up, DN is Down, CW is Clockwise, and CCW is Counter-Clockwise. The direction of the ultrasonic beam/scan was then used to determine the orientation of the flaw. A scan perpendicular to an axial weld or parallel to a circumferential weld is capable of detecting flaws that are oriented axially with respect to the RV. A scan parallel to an axial weld or perpendicular to a circumferential weld is capable of detecting flaws that are oriented circumferentially with respect to the RV.
The flaws that are determined to be in the beltline region and within the inner 1/10th thickness or 1 inch of the reactor vessel, whichever is greater, are then to be evaluated against the limits of Tables 4.5-1 and 4.5-2. The limits of these tables are expressed in
ATTACHMENT 1 Response to Request for Additional Information Regarding Relief Request 13R-17 to Extend the Reactor Pressure Vessel Inservice Inspection Interval Page 5 of 7 terms of number of flaws per 1000 inches of length or square inches of area.
Therefore, in order to make use of Table 4.5-1, the total length of welds inspected in the beltline is determined based on the plant-specific dimensions of the inspected beltline region. This weld length is used to determine plant-specific weld flaw limits. In order to make use of Table 4.5-2, the total area of forgings or plates inspected in the beltline region is also determined. This is also done using the plant-specific dimensions including the RV thickness and diameter along with the requirements for examination volume in Figures IWB-2500-1 and IWB-2500-2 of Section XI of the ASME Code. This area is used to determine plant-specific forging flaw limits.
T J)le 4.55-1: _a llosvabl e :L; uinber of Flaws in Welds (Refereuce 1' Through-Wall Extent (TWE) of Fl ilv Maximum number of flans per 1000 (in) inches of meld Iength in the inspection volume that are greater than or equal to
- Tli'E.,11N and less than TCt"E,,wx TWEMtN '1 : M.tti 0 0.075 No Limit 0.075 0.475 1,66.70 0.125 0.475 90.80 0.175 0.475 22 82 0.225 0.475 S.66 0.275 0.47: 4.01 0.325 0.475 3.01 0.375 0.475 1.49 0.425 0.475 1.00 0.?' ntinibe 0.00 Table 4.~-2, -Ulowable Number of Raves in Plates or Foi oii.g i.Reference 5)
Maximum number of flat -. per 1000 Through-Wall Extent (TI E) of F1,%vv square inches of inside surface area in
[ia} the inspection volume that are greater than or equal to TATMIN and less thin TWEMLV TWEif Ax TANEMA\
0 0.075 No Linut 0.375 S_C 5 0.1-1. 0.375 3.15 0.171 0.375 0.83 0.'.)25 0.375 0.29 0.27:: 0.375 O.CS 0.32: 0.375 0_C1 0.3 75 Ianaite 0.00 Note that the data in Tables 4.5-1 and 4.5-2 above are taken from 10 CFR 50.61 a (i.e.,
Reference 5).
ATTACHMENT 1 Response to Request for Additional Information Regarding Relief Request 13R-17 to Extend the Reactor Pressure Vessel Inservice Inspection Interval Page 6 of 7 The data in Table 5.5.2-2 below (excerpted from the Westinghouse report) was used to calculate the total forging area (in the beltline) of 9401 in2.
Table 5.5 '-- , Ins ection :area - Braidwood Unit "
Inside Diameter of RV(including cladding thickness 173 In CladdinLt Thickness 0.19 in' "
Inside Diameter of Weld Inspection Volume 173.38 m KV base metal thickness 8.63 in!"
'-"umber of Circumferential Melds 2 C ircuw.ferennalWeld Length 544.7 in Total Forging Area (in beltline'I 9401 in' N ore:
(a) V. ahre repotted in 3raLdaood [:nit 2 ISI repeat ($efi. ascr 11).
The total forging area (in the beltline) is then multiplied by each of the flaw acceptance criteria values in Table 4.5-2 to determine the values shown in Table 5.5.2-3 below.
able -5,5.2-3: :Uternate PTS Rule kllowable -umber of Flaws in Plates and For n-, Scaled for Braidwood Emit 2 Through-Wall Extent, Scaled Maximum number of flaws per 9401 square-inches of inside TWT (in) surface area in the inspection volume that are greater than or equal to Number of Flaws I %TMIN and less than TVi .,,AA.Y. This flaw density does not include (AxialrCirc.)
r11" EMIN MENAX underclad Cracks in for?_,tne1.
0 C.10755 ido 7 Unit 0 0.075 0.375 76 1 (011) 0.1~= 0.375 30 1 (0r1) 0.1"1: 0375 S 0
- 0. _ 25 0.375 0 0."5 0375 1 0 0.3'5 0375 1 0 0.375 1 Ia6n:ite C C The sizes of the identified flaws were then compared to the acceptance criteria, in terms of Through-Wall Extent. No Unit 1 flaws were required to be evaluated using the 10 CFR 50.61 a flaw acceptance criteria as shown in Table 5.5.1-1 below. One Unit 2 flaw was determined to be within the inner 1 inch of the Unit 2 vessel and thus required an evaluation using the 10 CFR 50.61 a flaw acceptance criteria as shown in Table 5.5.2-1 below.
ATTACHMENT 1 Response to Request for Additional Information Regarding Relief Request 13R-17 to Extend the Reactor Pressure Vessel Inservice Inspection Interval Page 7 of 7 Table 5.5.1-1: Reactor ISI Information for Belthne 171awrl Braidwood Unit 1 Inner Flaw Limits v to a TWE Loca tion fine' Inner (1.10)t FI ate Evaluation (in)(` ,Forsing.fi'eld) (3.8)t" Orientation or 1"" Regnired" 1RV 004 0.19 F igune Yes Cu:c. No 1RV-01-003 - 0.16 Forging Yes Cue. No 027 Feld Ye. Circ. No
`late.
i a) 23 dimension from Table 4.6.1-3 Table Reactor Vessel ISI Information for Belthile Flan-1 Braidwood Unit Inner wFlan Limit, Indication Tai-E Location Inner Fla
'A"eld ISI No, Be'ltline" (1.10)t E ~*aluation No. (in)( ` (Forging:~i'1'eld) (3:8)t" orientatiun or 1"" Required"
.004 1 C 33 weld Ye:, 'era 'tio CUZ No
.2 - RV-01-003 I -, - _ Forg= ye. Y'e ~ 1_"_ CuZ Ye::
Note:
(a) 2a dimension from Table 4.6.2-3 This flaw (in weld 2-RV-01-003) was found to be acceptable in accordance with the Alternate PTS Rule flaw acceptance criteria; therefore, Braidwood Station Unit 2 demonstrated that the flaw distributions used in the pilot plant analyses were bounding.
REFERENCES
- 1. Letter from D. M. Gullott (Exelon Generation Company, LLC) to NRC, " Relief Request 13R-17 to Extend the Reactor Pressure Vessel Inservice Inspection Interval," dated July 21, 2016
- 2. Email form J. Wiebe (NRC) to J. A. Bauer (Exelon Generation Company, LLC),
"Preliminary RAls for Braidwood Station Relief Request 13R-17," dated October 6, 2016
- 3. WCAP-16168-NP-A, Revision 3, "Risk-Informed Extension of the Reactor Vessel In-Service Inspection Interval," dated October 2011
- 4. Letter from R. A. Nelson (NRC) to W. A. Nowinowski (PWROG), "Revised Final Safety Evaluation by the Office of Nuclear Reactor Regulation Regarding Pressurized Water Reactor Owners Group Topical Report WCAP-16168-NP-A, Revision 2, 'Risk-Informed Extension of the Reactor Vessel In-Service Inspection Interval,"' dated July 26, 2011
- 5. 10 CFR 50.61 a, "Alternate fracture toughness requirements for protection against pressurized thermal shock events"