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Category:Letter
MONTHYEARIR 05000456/20230042024-02-0202 February 2024 Integrated Inspection Report 05000456/2023004 and 05000457/2023004 ML24025C7242024-01-29029 January 2024 Notification of NRC Baseline Inspection and Request for Information; Inspection Report 05000456/2024002; 05000457/2024002 IR 05000457/20230112024-01-25025 January 2024 2B Auxiliary Feedwater Pump Diesel Fuel Oil Dilution Report 05000457/2023011 and Preliminary Greater than Green Finding and Apparent Violation ML24018A0362024-01-17017 January 2024 Paragon Energy Solutions, Defect with Detroit Diesel/Mtu Fuel Injectors P/N R5229660 Cat Id 0001390618 RS-24-004, Proposed Alternative to the Distribution Requirements of ASME Code Table IWC-2411-1 for the Steam Generators2024-01-11011 January 2024 Proposed Alternative to the Distribution Requirements of ASME Code Table IWC-2411-1 for the Steam Generators ML23348A2162023-12-15015 December 2023 Exemption from Select Requirements of 10 CFR Part 73 (EPID L-2023-LLE-0030 (Security Notifications, Reports, and Recordkeeping and Suspicious Activity Reporting)) IR 05000456/20200232023-12-15015 December 2023 Baseline Security Inspection Document; 05000456/2023/402; 05000457/2023/402 ML23277A0032023-12-11011 December 2023 Issuance of Amendments Regarding Adoption of TSTF-370 ML23339A0452023-12-0505 December 2023 Request for Information for an NRC Post-Approval Site Inspection for License Renewal Inspection Report 05000546/2024010 ML23313A1552023-12-0101 December 2023 Review of the Fall 2022 Steam Generator Tube Inspection Report ML23331A8922023-11-22022 November 2023 Supplement - Braidwood Security Rule Exemption Request ISFSI Docket No. Reference 05000457/LER-2023-001, Submittal of LER 2023-001-00 for Braidwood Station, Unit 2, Train B Auxiliary Feedwater Pump Was Inoperable Due to Degraded Oil in the Crank Case2023-11-17017 November 2023 Submittal of LER 2023-001-00 for Braidwood Station, Unit 2, Train B Auxiliary Feedwater Pump Was Inoperable Due to Degraded Oil in the Crank Case ML23321A0442023-11-17017 November 2023 Notification of Deviation from Electric Power Research Institute (EPRI) Topical Report MRP-227, Revision 1-A, Materials Reliability Program: Pressurized Water Reactor Internals Inspection and Evaluation Guideline 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 ML23317A1192023-11-10010 November 2023 Constellation Energy Generation, LLC - 2023 Annual Report - Guarantees of Payment of Deferred Premiums 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 IR 05000456/20234012023-10-18018 October 2023 Security Baseline Inspection Report 05000456/2023401 and 05000457/2023401 IR 05000456/20230102023-10-18018 October 2023 Functional Engineering Inspection Commercial Grade Dedication Report 05000456/2023010 and 05000457/2023010 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. ML23226A0062023-09-19019 September 2023 Review of License Renewal Commitment Number 10 Submittal ML23180A1692023-09-11011 September 2023 Calvert Cliff Units 1 & 2, and R.E. Ginna Plant - Withdrawal of Proposed Alternatives to American Society of Mechanical Engineers (ASME) Requirements (Epids L-2022-LRR-0074, 0076, 0079, 0091, 0092, 0093 and 0094) IR 05000456/20230052023-08-30030 August 2023 Updated Inspection Plan for Braidwood Station Report 05000456/2023005 and 05000457/2023005 ML23234A2462023-08-25025 August 2023 Confirmation of Initial License Examination IR 05000456/20230022023-08-0303 August 2023 Integrated Inspection Report 05000456/2023002 and 05000457/2023002 ML23188A1292023-07-26026 July 2023 Issuance of Amendment Nos. 233 and 233 Adoption of TSTF-577, Revised Frequencies for Steam Generator Tube Inspections, Revision 1 ML23087A0762023-07-13013 July 2023 Issuance of Amendment Nos. 232 and 232 Revision of Technical Specifications for the Ultimate Heat Sink ML23191A8442023-07-10010 July 2023 05000456; 05000457 Notification of an NRC Biennial Licensed Operator Requalification Program Inspection and Request for Information ML23178A2422023-06-28028 June 2023 Reassignment of the U.S. Nuclear Regulatory Commission Branch Chief in the Division of Operating Reactor Licensing for Plant Licensing Branch III 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 ML23110A1152023-06-12012 June 2023 Environmental Assessment and Finding of No Significant Impact Related to a Requested Increase in Ultimate Heat Sink Temperature (EPID L-2023-LLA-0042) (Letter) 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. ML23138A1342023-05-18018 May 2023 Information Meeting with a Question and Answer Session to Discuss NRC 2022 End-Of-Cycle Plant Performance Assessment of Braidwood Station and Byron Station ML23132A0472023-05-12012 May 2023 Submittal of 2022 Annual Radiological Environmental Operating Report ML23130A0072023-05-10010 May 2023 Submittal of Core Operating Limits Report Cycle 24, Rev. 16 IR 05000456/20230012023-05-0808 May 2023 Integrated Inspection Report 05000456/2023001 and 05000457/2023001 ML23114A2522023-04-28028 April 2023 Request to Use a Provision of a Later Edition of the ASME Boiler & Pressure Vessel Code, Section XI ML23118A0202023-04-28028 April 2023 Submittal of 2022 Annual Radioactive Effluent Release Report ML23110A3202023-04-21021 April 2023 Information Request to Support the NRC Annual Baseline Emergency Action Level and Emergency Plan Changes Inspection 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 ML23095A1292023-04-0505 April 2023 Steam Generator Tube Inspection Report for Refueling Outage 23 ML23094A1352023-04-0404 April 2023 Request for Information for Nrc Commercial Grade Dedication Inspection Inspection Report 05000456/2023010 05000457/2023010 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 2024-02-02
[Table view] Category:Report
MONTHYEARBW230054, Attachment 2: MDMP Deviation Form2023-11-17017 November 2023 Attachment 2: MDMP Deviation Form ML23321A0452023-11-17017 November 2023 EC 639996 (Byron), Revision 1 and 640160 (Braidwood), Revision 0, Technical Evaluation for NEI 03-08 Deviation of Baffle-Former Bolts Volumetric Examinations for Byron and Braidwood 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 BW220062, Pressure and Temperature Limits Report (Ptlr), Revision 92022-10-20020 October 2022 Pressure and Temperature Limits Report (Ptlr), Revision 9 NMP1L3469, Constellation Energy Company, LLC, Request for Use of Honeywell Mururoa V4F1 R Supplied Air Suits2022-06-30030 June 2022 Constellation Energy Company, LLC, Request for Use of Honeywell Mururoa V4F1 R Supplied Air Suits RS-22-071, License Renewal Response to Commitment 10 - Evaluation of Possible PWSCC Crack Initiation and Propagation in the Steam Generator Channel Head Assembly and Tube-to-Tubesheet Welds2022-06-0707 June 2022 License Renewal Response to Commitment 10 - Evaluation of Possible PWSCC Crack Initiation and Propagation in the Steam Generator Channel Head Assembly and Tube-to-Tubesheet Welds 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 ML21349A1062021-12-15015 December 2021 Justification for the Deviation from MRP 2019-008, Technical Evaluation 635273 BW210065, Pressure and Temperature Limits Report, Revision 82021-10-27027 October 2021 Pressure and Temperature Limits Report, Revision 8 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-21-093, 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 BW210047, ER-BR-330-1008, Revision 0, Snubber Program Plan for the Fourth 10-Year Interval2021-06-30030 June 2021 ER-BR-330-1008, Revision 0, Snubber Program Plan for the Fourth 10-Year Interval RS-21-056, Proposed Alternative for Examination of Pressurizer Circumferential and Longitudinal Shell-to-Head Welds and Nozzle-to-Vessel Weld2021-05-12012 May 2021 Proposed Alternative for Examination of Pressurizer Circumferential and Longitudinal Shell-to-Head Welds and Nozzle-to-Vessel Weld RS-21-001, Revised Proposed Alternative to Utilize Code Cases N-878 and N-880 for Carbon Steel Piping2021-01-0404 January 2021 Revised Proposed Alternative to Utilize Code Cases N-878 and N-880 for Carbon Steel Piping RS-20-154, Application for Revision to TS 5.5.9, Steam Generator (SG) Program for a One-Time Deferral of Steam Generator Tube Inspections2020-12-16016 December 2020 Application for Revision to TS 5.5.9, Steam Generator (SG) Program for a One-Time Deferral of Steam Generator Tube Inspections ML20303A1752020-10-23023 October 2020 Proposed Relief Request from Section XI Repair/Replacement Documentation for Bolting Replacement of Pressure Retaining Bolting ML20195B1622020-06-30030 June 2020 Attachment 11 - SG-SGMP-17-25-NP, Revision 1, Foreign Object Limits Analysis for the Byron and Braidwood Unit 2 Steam Generators June 2020 ML19228A0232019-08-15015 August 2019 Proposed Alternative to Utilize Code Case N-879 ML18348A9792018-12-14014 December 2018 Transmittal of 10 CFR 50.59 Summary Report ML18348A9722018-12-12012 December 2018 Submittal of Analytical Evaluation in Accordance with ASME Code Section XI RS-17-048, Annual 10 CFR 50.46 Report of Emergency Core Cooling System Evaluation Model Changes and Errors2017-04-0707 April 2017 Annual 10 CFR 50.46 Report of Emergency Core Cooling System Evaluation Model Changes and Errors ML17024A0362016-12-31031 December 2016 Operating Data Report for 2016 RS-16-223, Report of Full Compliance with March 12, 2012 Commission Order Modifying Licenses with Regard to Requirements for Mitigation Strategies for Beyond-Design-Basis External Events (Order Number EA-12-049)2016-12-0707 December 2016 Report of Full Compliance with March 12, 2012 Commission Order Modifying Licenses with Regard to Requirements for Mitigation Strategies for Beyond-Design-Basis External Events (Order Number EA-12-049) RS-16-174, High Frequency Supplement to Seismic Hazard Screening Report, Response to NRC Request for Information Pursuant to 10 CFR 50.54(f) Regarding Recommendation 2.1 of the Near-Term Task Force Review...2016-11-0303 November 2016 High Frequency Supplement to Seismic Hazard Screening Report, Response to NRC Request for Information Pursuant to 10 CFR 50.54(f) Regarding Recommendation 2.1 of the Near-Term Task Force Review... ML17170A1472016-10-0707 October 2016 Areva, 51-9263014-000, PWSCC Evaluation of Uhp Cavitation Peening for Byron and Braidwood Reactor Vessel Head Penetrations. ML16356A0202016-10-0707 October 2016 51-9263014-000, PWSCC Evaluation of Uhp Cavitation Peening for Byron and Braidwood Reactor Vessel Head Penetrations. ML17360A1742016-10-0707 October 2016 Attachment 6: Areva Document 51-9263014-000, PWSCC Evaluation of Uhp Cavitation Peening for Byron and Braidwood Reactor Vessel Head Penetrations (Non-Proprietary) RS-17-039, Attachment 6, Areva Document 51-9263014-000, PWSCC Evaluation of Uhp Cavitation Peening for Byron and Braidwood Reactor Vessel Head Penetrations.2016-10-0707 October 2016 Attachment 6, Areva Document 51-9263014-000, PWSCC Evaluation of Uhp Cavitation Peening for Byron and Braidwood Reactor Vessel Head Penetrations. ML17095A2692016-10-0707 October 2016 Attachment 6, Areva Document 51-9263014-000, PWSCC Evaluation of Uhp Cavitation Peening for Byron and Braidwood Reactor Vessel Head Penetrations. ML16236A2082016-08-23023 August 2016 Report of Backfit Appeal Review Panel ML16214A1992016-08-11011 August 2016 an Assessment of Core Damage Frequency for Byron/Braidwood Backit Appeal Review RS-16-099, Mitigating Strategies Flood Hazard Assessment (Msfha) Submittal2016-06-30030 June 2016 Mitigating Strategies Flood Hazard Assessment (Msfha) Submittal ML16250A5182016-04-30030 April 2016 Technical Evaluation Report Related to the Exelon Generation Company, LLC, License Amendment Request to Braidwood Station, Units 1 and 2, Technical Specification 3.7.9, Ultimate Heat Sink. Docket Nos. Stn 50-456 & 457 RS-16-073, Submittal of Annual 10 CFR 50.46 Report of Emergency Core Cooling System Evaluation Model Changes and Errors2016-04-0707 April 2016 Submittal of Annual 10 CFR 50.46 Report of Emergency Core Cooling System Evaluation Model Changes and Errors RS-16-057, Supplement 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 Welds2016-03-15015 March 2016 Supplement 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 ML16014A1882016-01-22022 January 2016 Staff Assessment of Information Provided Pursuant to Title 10 of the Code of Federal Regulations Part 50 Section 50.54(f) Seismic Hazard Reevaluations for Recommendation 2.1of the Near-Term Task Force Review of Insights ML15344A1592015-12-10010 December 2015 Submittal of Pressure and Temperature Limits Reports (Ptlrs), Revision 8 and Braidwood, Unit 2 - Pressure and Temperature Limits Reports (Ptlrs), Revision 7 ML15322A3172015-11-18018 November 2015 Record of Decision ML15237A3822015-10-15015 October 2015 Pressure and Temperature Limits Report for Measurement Uncertainty Recapture Power Uprate RS-15-259, Final Report: Five Year Post-Construction Monitoring of the Unionid Community Near the Braidwood Station Kankakee River Discharge.2015-09-30030 September 2015 Final Report: Five Year Post-Construction Monitoring of the Unionid Community Near the Braidwood Station Kankakee River Discharge. RS-15-129, Westinghouse Report CCE-15-27, Revision 1, Braidwood Units 1 and 2 - Responses to NRC Request for Additional Information (Rai)Regarding Ultimate Heat Sink Temperature Increase License Amendment Request, April 20152015-04-30030 April 2015 Westinghouse Report CCE-15-27, Revision 1, Braidwood Units 1 and 2 - Responses to NRC Request for Additional Information (Rai)Regarding Ultimate Heat Sink Temperature Increase License Amendment Request, April 2015 ML14349A6572014-12-15015 December 2014 CFR 50.59 Changes, Tests, and Experiments, Paragraph (d)(2), Summary Report RS-14-348, Comments on the Safety Evaluation Report with Open Items, Related to the License Renewal Application2014-12-15015 December 2014 Comments on the Safety Evaluation Report with Open Items, Related to the License Renewal Application RS-14-277, Proposed Alternative to Utilize Code Case N-513-4, Evaluation Criteria for Temporary Acceptance of Flaws in Moderate Energy Class 2 or 3 Piping Section XI, Division 12014-09-24024 September 2014 Proposed Alternative to Utilize Code Case N-513-4, Evaluation Criteria for Temporary Acceptance of Flaws in Moderate Energy Class 2 or 3 Piping Section XI, Division 1 ML14141A1332014-06-30030 June 2014 Staff Assessment of the Flooding Walkdown Report Supporting Implementation of Near-Term Task Force Recommendation 2.3 Related to the Fukushima DAI-ICHI Nuclear Power Plant Accident ML14178B2222014-06-24024 June 2014 Technical Review of TIA 2013-02, Single Spurious Assumptions for Braidwood and Byron Stations Safe-Shutdown Methodology ML14101A4452014-06-0404 June 2014 Staff Assessment of the Seismic Walkdown Report Supporting Implementation of Near-Term Task Force Recommendation 2.3 Related to the Fukushima Dai-Ichi Nuclear Power Plant Accident (Tac No. MF0095) ML14101A3522014-06-0404 June 2014 Staff Assessment of the Seismic Walkdown Report Supporting Implementation of Near-Term Task Force Recommendation 2.3 Related to the Fukushima Dai-ichi Nuclear Power Plant Accident ML14127A1742014-05-0707 May 2014 Startup Report for the Measurement Uncertainty Recapture Power Uprate ML14120A0392014-04-24024 April 2014 Units 1 & 2 - License Amendment Request to Install New Low Degraded Voltage Relays & Timers on the 4.16 Kv Engineered Safety Features (ESF) Buses 2023-04-20
[Table view] Category:Miscellaneous
MONTHYEARBW230054, Attachment 2: MDMP Deviation Form2023-11-17017 November 2023 Attachment 2: MDMP Deviation Form ML23321A0452023-11-17017 November 2023 EC 639996 (Byron), Revision 1 and 640160 (Braidwood), Revision 0, Technical Evaluation for NEI 03-08 Deviation of Baffle-Former Bolts Volumetric Examinations for Byron and Braidwood RS-22-071, License Renewal Response to Commitment 10 - Evaluation of Possible PWSCC Crack Initiation and Propagation in the Steam Generator Channel Head Assembly and Tube-to-Tubesheet Welds2022-06-0707 June 2022 License Renewal Response to Commitment 10 - Evaluation of Possible PWSCC Crack Initiation and Propagation in the Steam Generator Channel Head Assembly and Tube-to-Tubesheet Welds 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 ML21349A1062021-12-15015 December 2021 Justification for the Deviation from MRP 2019-008, Technical Evaluation 635273 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 ML20303A1752020-10-23023 October 2020 Proposed Relief Request from Section XI Repair/Replacement Documentation for Bolting Replacement of Pressure Retaining Bolting RS-17-048, Annual 10 CFR 50.46 Report of Emergency Core Cooling System Evaluation Model Changes and Errors2017-04-0707 April 2017 Annual 10 CFR 50.46 Report of Emergency Core Cooling System Evaluation Model Changes and Errors ML17024A0362016-12-31031 December 2016 Operating Data Report for 2016 RS-16-174, High Frequency Supplement to Seismic Hazard Screening Report, Response to NRC Request for Information Pursuant to 10 CFR 50.54(f) Regarding Recommendation 2.1 of the Near-Term Task Force Review...2016-11-0303 November 2016 High Frequency Supplement to Seismic Hazard Screening Report, Response to NRC Request for Information Pursuant to 10 CFR 50.54(f) Regarding Recommendation 2.1 of the Near-Term Task Force Review... RS-17-039, Attachment 6, Areva Document 51-9263014-000, PWSCC Evaluation of Uhp Cavitation Peening for Byron and Braidwood Reactor Vessel Head Penetrations.2016-10-0707 October 2016 Attachment 6, Areva Document 51-9263014-000, PWSCC Evaluation of Uhp Cavitation Peening for Byron and Braidwood Reactor Vessel Head Penetrations. ML17170A1472016-10-0707 October 2016 Areva, 51-9263014-000, PWSCC Evaluation of Uhp Cavitation Peening for Byron and Braidwood Reactor Vessel Head Penetrations. ML17095A2692016-10-0707 October 2016 Attachment 6, Areva Document 51-9263014-000, PWSCC Evaluation of Uhp Cavitation Peening for Byron and Braidwood Reactor Vessel Head Penetrations. ML16236A2082016-08-23023 August 2016 Report of Backfit Appeal Review Panel ML16214A1992016-08-11011 August 2016 an Assessment of Core Damage Frequency for Byron/Braidwood Backit Appeal Review RS-16-099, Mitigating Strategies Flood Hazard Assessment (Msfha) Submittal2016-06-30030 June 2016 Mitigating Strategies Flood Hazard Assessment (Msfha) Submittal RS-16-073, Submittal of Annual 10 CFR 50.46 Report of Emergency Core Cooling System Evaluation Model Changes and Errors2016-04-0707 April 2016 Submittal of Annual 10 CFR 50.46 Report of Emergency Core Cooling System Evaluation Model Changes and Errors ML16014A1882016-01-22022 January 2016 Staff Assessment of Information Provided Pursuant to Title 10 of the Code of Federal Regulations Part 50 Section 50.54(f) Seismic Hazard Reevaluations for Recommendation 2.1of the Near-Term Task Force Review of Insights ML15344A1592015-12-10010 December 2015 Submittal of Pressure and Temperature Limits Reports (Ptlrs), Revision 8 and Braidwood, Unit 2 - Pressure and Temperature Limits Reports (Ptlrs), Revision 7 ML15322A3172015-11-18018 November 2015 Record of Decision ML15237A3822015-10-15015 October 2015 Pressure and Temperature Limits Report for Measurement Uncertainty Recapture Power Uprate RS-15-129, Westinghouse Report CCE-15-27, Revision 1, Braidwood Units 1 and 2 - Responses to NRC Request for Additional Information (Rai)Regarding Ultimate Heat Sink Temperature Increase License Amendment Request, April 20152015-04-30030 April 2015 Westinghouse Report CCE-15-27, Revision 1, Braidwood Units 1 and 2 - Responses to NRC Request for Additional Information (Rai)Regarding Ultimate Heat Sink Temperature Increase License Amendment Request, April 2015 ML14349A6572014-12-15015 December 2014 CFR 50.59 Changes, Tests, and Experiments, Paragraph (d)(2), Summary Report RS-14-348, Comments on the Safety Evaluation Report with Open Items, Related to the License Renewal Application2014-12-15015 December 2014 Comments on the Safety Evaluation Report with Open Items, Related to the License Renewal Application ML14141A1332014-06-30030 June 2014 Staff Assessment of the Flooding Walkdown Report Supporting Implementation of Near-Term Task Force Recommendation 2.3 Related to the Fukushima DAI-ICHI Nuclear Power Plant Accident ML14101A4452014-06-0404 June 2014 Staff Assessment of the Seismic Walkdown Report Supporting Implementation of Near-Term Task Force Recommendation 2.3 Related to the Fukushima Dai-Ichi Nuclear Power Plant Accident (Tac No. MF0095) ML14101A3522014-06-0404 June 2014 Staff Assessment of the Seismic Walkdown Report Supporting Implementation of Near-Term Task Force Recommendation 2.3 Related to the Fukushima Dai-ichi Nuclear Power Plant Accident ML14127A1742014-05-0707 May 2014 Startup Report for the Measurement Uncertainty Recapture Power Uprate ML14120A0392014-04-24024 April 2014 Units 1 & 2 - License Amendment Request to Install New Low Degraded Voltage Relays & Timers on the 4.16 Kv Engineered Safety Features (ESF) Buses ML14066A4792014-03-0404 March 2014 Clarification of Licensing Basis Assumptions for a Natural Circulation Cooldown Event ML14059A1242014-02-28028 February 2014 Pressure and Temperature Limits Reports (Ptlrs) IR 05000272/20132012013-01-31031 January 2013 Exelon Generation Co., LLC, U.S. Nuclear Regulatory Commission (NRC) Office of Investigations (01) Investigation; Summary of 01 Report No. 3-2010-034; NRC Inspection Report Conference Letter - Licensee IR 05000456/20132022013-01-31031 January 2013 Exelon Generation Co., LLC, U.S. Nuclear Regulatory Commission (NRC) Office of Investigations (01) Investigation; Summary of 01 Report No. 3-2010-034; NRC Inspection Report Conference Letter - Licensee ML13008A2192013-01-31031 January 2013 U.S. Nuclear Regulatory Commission (NRC) Office of Investigations (01) Investigation; Summary of 01 Report No. 3-2010-034; NRC Inspection Report Conference Letter - Licensee IR 05000454/20132022013-01-31031 January 2013 Exelon Generation Co., LLC, U.S. Nuclear Regulatory Commission (NRC) Office of Investigations (01) Investigation; Summary of 01 Report No. 3-2010-034; NRC Inspection Report Conference Letter - Licensee IR 05000461/20132012013-01-31031 January 2013 Exelon Generation Co., LLC, U.S. Nuclear Regulatory Commission (NRC) Office of Investigations (01) Investigation; Summary of 01 Report No. 3-2010-034; NRC Inspection Report Conference Letter - Licensee IR 05000237/20132022013-01-31031 January 2013 Exelon Generation Co., LLC, U.S. Nuclear Regulatory Commission (NRC) Office of Investigations (01) Investigation; Summary of 01 Report No. 3-2010-034; NRC Inspection Report Conference Letter - Licensee IR 05000373/20132012013-01-31031 January 2013 Exelon Generation Co., LLC, U.S. Nuclear Regulatory Commission (NRC) Office of Investigations (01) Investigation; Summary of 01 Report No. 3-2010-034; NRC Inspection Report Conference Letter - Licensee IR 05000289/20132012013-01-31031 January 2013 Exelon Generation Co., LLC, U.S. Nuclear Regulatory Commission (NRC) Office of Investigations (01) Investigation; Summary of 01 Report No. 3-2010-034; NRC Inspection Report Conference Letter - Licensee IR 05000254/20132012013-01-31031 January 2013 Exelon Generation Co., LLC, U.S. Nuclear Regulatory Commission (NRC) Office of Investigations (01) Investigation; Summary of 01 Report No. 3-2010-034; NRC Inspection Report Conference Letter - Licensee IR 05000277/20132012013-01-31031 January 2013 Exelon Generation Co., LLC, U.S. Nuclear Regulatory Commission (NRC) Office of Investigations (01) Investigation; Summary of 01 Report No. 3-2010-034; NRC Inspection Report Conference Letter - Licensee IR 05000219/20132012013-01-31031 January 2013 Exelon Generation Co., LLC, U.S. Nuclear Regulatory Commission (NRC) Office of Investigations (01) Investigation; Summary of 01 Report No. 3-2010-034; NRC Inspection Report Conference Letter - Licensee IR 05000352/20132012013-01-31031 January 2013 Exelon Generation Co., LLC, U.S. Nuclear Regulatory Commission (NRC) Office of Investigations (01) Investigation; Summary of 01 Report No. 3-2010-034; NRC Inspection Report Conference Letter - Licensee ML12349A3632012-12-14014 December 2012 10 CFR 50.59 Summary Report for June 19, 2010 Through June 18, 2012 ML12339A2172012-11-16016 November 2012 12Q0108.10-R-001, Revision 0, Seismic Walkdown Report in Response to the 50.54(f) Information Request Regarding Fukushima Near-Term Task Force Recommendation 2.3: Seismic for the Braidwood Station, Unit 1. Part 1 of 5 ML12339A2182012-11-16016 November 2012 12Q0108.10-R-002, Revision 0, Seismic Walkdown Report in Response to the 50.54(f) Information Request Regarding Fukushima Near-Term Task Force Recommendation 2.3: Seismic for the Braidwood Station, Unit 2. Part 5 of 5 ML12339A2192012-11-16016 November 2012 12Q0108.10-R-001, Revision 0, Seismic Walkdown Report in Response to the 50.54(f) Information Request Regarding Fukushima Near-Term Task Force Recommendation 2.3: Seismic for the Braidwood Station, Unit 1. Part 2 of 5 ML12339A2202012-11-16016 November 2012 12Q0108.10-R-001, Revision 0, Seismic Walkdown Report in Response to the 50.54(f) Information Request Regarding Fukushima Near-Term Task Force Recommendation 2.3: Seismic for the Braidwood Station, Unit 1. Part 3 of 5 ML12339A2212012-11-16016 November 2012 12Q0108.10-R-001, Revision 0, Seismic Walkdown Report in Response to the 50.54(f) Information Request Regarding Fukushima Near-Term Task Force Recommendation 2.3: Seismic for the Braidwood Station, Unit 1. Part 4 of 5 ML12339A2222012-11-16016 November 2012 12Q0108.10-R-001, Revision 0, Seismic Walkdown Report in Response to the 50.54(f) Information Request Regarding Fukushima Near-Term Task Force Recommendation 2.3: Seismic for the Braidwood Station, Unit 1. Part 5 of 5 2023-11-17
[Table view] |
Text
April 12, 2012 BW120036 U. S. Nuclear Regulatory Commission ATIN: Document Control Desk Washington, DC 20555-0001 Braidwood Station, Unit 2 Facility Operating License No. NPF-77 NRC Docket No. STN 50-457
Subject:
Submittal of Analytical Evaluation in Accordance with ASME Code Section XI In accordance with the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code,Section XI, "Rules for Inservice Inspection of Nuclear Power Plant Components,"
2001 Edition through 2003 Addenda, IWC-3125, "Review by Authorities," paragraph (b), Braidwood Station is sUbmitting an analytical evaluation associated with the Unit 2 Residual Heat Removal (RHR) Mixing Tee examined during the Braidwood Unit 2 spring 2011 refueling outage (A2R15).
The mixing tee is the point where the RHR heat exchanger bypass line ties into the RHR heat exchanger discharge line. The potential for thermal fatigue exists at this location because the cooled flow from the heat exchanger mixes with the warmer flow from the bypass line.
During A2R15, Braidwood inspected the mixing tee at the discharge of the 2A RHR heat exchanger in accordance with the guidelines of Electric Power Research Institute (EPRI) Material Reliability Program (MRP)-192, "Materials Reliability Program Assessment of RHR Mixing Tee Thermal Fatigue in PWR Plants." The inspection found an indication at the toe (tee side) of weld 2RH-03-28. The size of the indication exceeds the ASME Section XI acceptance standards of IWB-3514, which was used in accordance with the provision of IWC-3514. As allowed by the ASME Section XI, a flaw evaluation in accordance with IWC-3640 was completed to determine the allowable flaw size and the time required for the flaw to grow to the maximum allowable size.
The evaluation has determined that the detected indication meets the ASME Section XI acceptance criteria. Based on the analysis of the differential temperature of interest, the time period for the flaw to grow to the maximum allowable size is more than sufficient to cover the remaining operating life of Braidwood Unit 2.
April 12, 2012 U. S. Nuclear Regulatory Commission Page 2 The nondestructive examination report concluded that no craze cracking was detected. The highly localized nature of the flaw and the absence of craze cracking are not indicative of fatigue due to thermal mixing effect. The indication can most likely be attributed to the geometrical imperfection discontinuity at the weld root that occurred during the original weld fabrication.
There are no regulatory commitments contained in this letter.
If you have any questions or require additional information, please contact Mr. Chris VanDenburgh, Regulatory Assurance Manager, at 815-417-2800.
Sincerely, Daniel J. Enrig,"'--. ...J Site Vice President Braidwood Station
Attachment:
Flaw Tolerance Evaluation of Braidwood Station Unit 2 RHR Mixing Tee Weld cc: NRC Project Manager- Braidwood Station Illinois Emergency Management Agency - Division of Nuclear Safety US NRC Regional Administrator, Region III US NRC Senior Resident Inspector (Braidwood Station)
ExelcnSM CC-AA-309-1001 Revision 6 Nuclear oeSlgn Analysis . M' ajar ReVlslon Cover Sh eet Design Analysis (Major Revision) I Last Page No.6 19 Analysis No.: 1 1100643.301 Revision: 2 0
Title:
3 Flaw Tolerance Evaluation of Braidwood Station Unit 2 RHR Mixing Tee Weld EC/ECR No.:
- 384283 Revision: 5 0 Station(s): 7 Braidwood Component(s): 14 Unit No.: 8 2 2RH03AA Discipline: 9 MEDC Descrip. Code/Keyword: 10 M04 Safety/QA Class: 11 SR System Code: 12 RH Structure: 13 N/A CONTROLLED DOCUMENT REFERENCES 15 Document No.: FromITo Document No.: FromITo Is this Design Analysis Safeguards Information? I. YesD No~ If yes, see SY-AA-101-106 Does this Design Analysis contain Unverified Assumptions? 17 YesD No~ If yes, ATI/AR#:
This Design Analysis SUPERCEDES: 18 N/A in its entirety.
Description of Revision (list changed pages when all pages of original analysis were not changed): 19 Original Issue-This design analysis documents the flaw tolerance evaluation that has been performed by Structural Integrity for the RHR mixing tee weld at the discharge of the 2A RHR heat exchanger on line 2RH03AA, weld
- 2RH-03-28.
Preparer: 20 G.A. Miessi See attached calculation Print Name Si n Name Date Method of Review: 21 Detailed Review ~ Alternate Calculations (attached) D Testing D Reviewer: 22 S. Tang/D. Harris See attached calculation Print Name Sign Name Date Review Notes: 23 Independent review'R!' Peer review D (For External Analyses Only)
External Approver: 2' G.A. Miessi See attached calculation Exelon Reviewer: 25 Print Name GIIJYl}NNJ PrintPtMII Name eI UV,VIlJ111
/
Jign la11d!£ Siqn Name Date k.-.30-Z011 Date Independent ard Party Review Reqd? 26 Yes~ NoD fZ~
Independent fQZ (;.. VI11;cr~
Reviewer Guy Deboo ne.1'm..~c.otJ 04-('3 0 t 11 Print Name
,~ Sign Name oA OA.{ ;~atI11 L-.
Exelon Approver: 27
- 12. .J. ~l~ '7
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Print Name Sian Name Date
Structural Integrity Associates, Inc.
L 10 P G PR Fla alu tion of raid\ ood tation nit 2 RHR ing Tee Id p
Braidw d tation, nil2 Flaw leran evaluation of the raidwood nit 2 RHR i ing T e Id Docum nt fJ td c i ion ription Re i ion Pa o t - 19 Origin II ue
~Atr-tan Tang T 04/30/11 Oa id . Ham o H 04/30/1 t Page 1 of 19 F0306-01Rl
Structural Integrity Associates, Inc.
Table of Contents
1.0 INTRODUCTION
4 2.0 METHODOLOGY 4 3.0 ASSUMPTIONS 4 4.0 DESIGN INPUTS 5 4.1 Location and Characterization of Reported Indication 5 4.2 Dimensions and Loads 5 4.3 Temperature-Time Histories 5 5.0 CALCULATIONS 10 5.1 Evaluation of Code Allowable Flaw Size 10 5.2 Crack Growth Analysis 11 5.3 Consideration of SCC for Flaws in Austenitic Welds 11 6.0 RESULTS 14 6.1 Allowable Flaw Sizes 14 6.2 Fatigue Crack Growth 14
7.0 CONCLUSION
18
8.0 REFERENCES
19 File No.: 1100643.301 Page 2 of19 Revision: 0 F0306-01Rl
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List of Tables Table 4-1: Dimensions and Operating Conditions 6 Table 4-2: Stresses from Piping Analysis 7 Table 4-3: Number of Hours Spent at L1T above 144°F 7 Table 6-1: Allowable Part Through-Wall Flaw Size 15 Table 6-2: Time for Reported Flaw to Grow to Code Allowable at a RHR HX L1Tof162°F 16 List of Figures Figure 4-1. RHR Heat Exchanger 2A Inlet-Outlet Temperature Difference (A2RI5 Coo1down) 8 Figure 4-2. RHR Heat Exchanger IB Inlet-Outlet Temperature Difference (AIRI5 Startup) 9 Figure 5-1: IGSCC Resistance in Weld Metal may be predicted by the Combined Influence of Carbon Content and Percent Ferrite 13 Figure 6-1. Crack Growth Results 17 File No.: 1100643.301 Page 3 of 19 Revision: 0 F0306-01Rl
e StflJcturallntegrify Associates, Inc.
1.0 INTRODUCTION
This calculation package is aimed at performing mixing tee thermal fatigue crack growth evaluations of the piping tee connection weld in the residual heat removal (RHR) line at the Braidwood Station, Unit 2.
The guidelines ofMRP-l92 [1] and ASME Code,Section XI [2] are used to perform a crack growth evaluation to determine the time required at the limiting frequency for a flaw to grow to the maximum allowable flaw size.
2.0 METHODOLOGY The first step in evaluating the stress intensity factor is to determine the temperature in the pipe wall for a sinusoidal temperature variation of the coolant inside the pipe. The steady state temperature solution is obtained using a plate model with convection at one surface and insulation on the opposite surface. The steady state temperature is obtained in closed-form solution and used in conjunction with the classical thermo-elastic solution for a cylinder with an axisymmetric temperature distribution through the wall.
The resulting stresses are then combined with influence functions to obtain the stress intensity factors as a function of crack size.
Then, the calculated stress intensity factors are used in fatigue crack growth analyses based on the crack growth equations and material property data obtained from References 2 through 4.
3.0 ASSUMPTIONS The assumptions in this calculation package are as follows:
- The thermal problem is idealized as a slab. The thickness of the pipe is small relative to the pipe diameter (Rlt = 13.4). Thus, treating the pipe wall as a slab is reasonable for this evaluation. A comparison of figures on Pages 101 and 200 of Carslaw and Jaeger [12] shows that the temperatures near the surface at short times are very similar for a slab and solid cylinder subject to a step change at the surface. The differences between the slab and cylinder would be even smaller for a hollow cylinder.
- The stresses in a cylinder are obtained for an axisymmetric through-wall temperature distribution as obtained for a slab, using the assumptions oflinear elasticity.
- The convection heat transfer coefficient is fixed at a value determined to provide agreement on crack initiation times observed in service.
- The temperature fluctuation is assumed to occur at a certain frequency, and fracture mechanics calculations are performed to determine the size of a crack that will just grow to an ASME Code allowable crack size as a function of time for selected frequencies. The worst frequency is then identified as the one that produces the fastest average crack growth rate, da/dt, for a crack growing from the initial size to the allowable size.
File No.: 1100643.301 Page 4 of 19 Revision: 0 F0306-01RI
e Structural/nlsgr;'y Associates, Inc.
- The weld residual stresses are assumed equal to the yield stress at the ID of the pipe with a pure bending through-wall distribution per the recommendations of Reference 13.
4.0 DESIGN INPUTS 4.1 Location and Characterization of Reported Indication The indication was reported at Weld No. 2RH-03-28 on Line No. 2RH03AA-8" of the RHR "A" Train
[5]. The indication was characterized as a planar circumferential flaw with a depth of 0.14" and a length of 0.70" [5] and found unacceptable per ASME Code,Section XI, IWB -3500 acceptance criteria.
4.2 Dimensions and Loads The design input pertaining to piping section dimensions and their respective limiting loadings are obtained from Reference 5. The design inputs required in calculating the allowable flaw size and the subsequent crack growth analysis are provided in Tables 4-1 and 4-2. The stresses shown in Table 4-2 were calculated using the design equations from the original code of construction [6]; thus, the stress intensification factor and 0.75 factor included in those equation must be considered in calculating the nominal stresses. The stresses used to calculate the allowable flaw size were determined in Spreadsheet HBR WD RHR Allowable Flaw Size. xis. "
The heat transfer coefficient used in this evaluation is equal to the critical value of 6400 BtuIhr- ft2- OF obtained from Reference 7.
4.3 Temperature-Time Histories Plant monitoring data at the locations of the thermal mixing are provided in Reference 5 for the Braidwood Station, Unit 2. The temperature-time histories from Reference 5 provide the time-history of the temperature difference, ~ T, between the inlet and the outlet of the RHR heat exchanger during the Braidwood Station Unit 2 plant shutdown from the current refueling outage, as illustrated in Figure 4-1.
Similarly, the temperature difference, ~ T, between the inlet and the outlet of the RHR heat exchanger during the Braidwood Station Unit 1 plant startup is illustrated in Figure 4-2. This temperature data is assumed to be similar to the startup data for Unit 2. A summary of the data presented in Table 4-3 shows the maximum number of hours spent at ~T above 144°F, the temperature threshold established in MRP-192 [1] and the maximum ~ T that was reached during the event. This maximum ~ T is conservatively used for the entire crack growth period.
File No.: 1100643.301 Page 5 of 19 Revision: 0 F0306-01Rl
e Structural Integrity Associates, Inc.
Table 4-1: Dimensions and Operating Conditions Node 1310 Material SA-312 TP 304 Pipe OD (in) 8.625 Pipe Wall Thickness (in) 0.322 Maximum Operating Temperature (OF) 350 Maximum Operating Pressure (psi) 570 Limiting ~ T (OF) 162 File No.: 1100643.301 Page 6 of 19 Revision: 0 F0306-01Rl
Table 4-2: Stresses from Piping Analysis.
Pressure Moment load (ksi) (ksi)
Equation 8 3.574 0.973 Equation 9 3.574 2.2 Equation 10 0 11.471 Equation 11 3.574 12.444 Pipe Break 3.574 13.716 Faulted 3.574 3.307 Notes:
(1) The DW stress is increased by 10% and carried to the other equations per Reference 5.
(2) A stress intensification factor of 1.8 was applied to the weld location in the piping analysis (3) Given that the stresses were calculated using the equations from Reference 6, the 0.75 factor included in those equations must be considered in calculating the nominal stresses.
Table 4-3: Number of Hours Spent at AT above 144°F Maximum Inlet - Outlet Transient Date Time (hrs) Temperature Difference (OF)
A2R15 Shutdown Sprin~ 2011 0.17 162 A1R15 Startup Fall 2010 0.0 104 File No.: 1100643.301 Page 7 of 19 Revision: 0 F0306-01RI
2ARH HX~T *A2R15 Cooldown 04:00~5:00 180 - --
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File No.: 1100643.301 Page 8 of 19 Revision: 0 F0306*01RI
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5.0 CALCULATIONS The fracture mechanics analyses are performed to determine the ASME Code [2] allowable flaw sizes and evaluate the time required for a crack to grow to the maximum allowable value. Design inputs from Section 4.0 were used to perform the crack growth evaluation of the Braidwood Unit 2 RHR mixing tees and the adjacent welds. The output files associated with these evaluations have been archived as supporting files.
5.1 Evaluation of Code Allowable Flaw Size The piping section under consideration is an austenitic stainless steel (SA-312 TP 304) and hence an elastic-plastic fracture mechanics (EPFM) failure criterion is used in this evaluation. The EPFM analysis was performed to bound the mixing tee and the weld location which are conservatively assumed to be flux welds. The technical approach consists of determining the critical flaw size (circumferential extent and through wall depth) in the pipe that will cause the pipe to fracture by ductile crack extension.
The stress ratios were calculated as follows:
For combined loading, StressRatlO. =-Z O"r
( 0"m + 0"b + _e_
0" SFb J (la) and, for membrane stress, ZSF 0" Stress Ratio = m m (lb)
O"r where, Z =1.30[1 + 0.01O(NPS - 4)] for SA-312 TP-304 (conservatively assuming that the weld is SMAW or SAW flux weld) 0"m and O"b are the primary membrane and primary bending stresses, respectively.
O"e is the secondary bending stress.
0"r is the flow stress (43.3 ksi at the operating temperature of 350°F [4)).
SF;, is the safety factor for bending stress.
SFm is the safety factor for membrane stress.
The tables of ASME Code,Section XI, Appendix C [2] are used to determine the allowable flaw depth-to-thickness ratio for each service leveL Spreadsheet "B WRD RHR_Allowable Flaw Size.xls " is used for the allowable flaw size calculations.
File No.: 1100643.301 Page 10 of19 Revision: 0 F0306-01Rl
5.2 Crack Growth Analysis Fatigue crack growth relations for austenitic materials in LWR environment are obtained from Reference
- 3. The fatigue crack growth relation for austenitic stainless steel in PWR water is taken from Reference 3 with modifications to agree with Reference 2. As stated at the bottom of Table 2, Page 7 of Reference 3, the total fatigue crack growth rate is composed of an air and an environmental term:
dal total = dN dN dal air + dN dal env (2)
Each of the contributors is given as follows:
-
da I air = CairS(R)LlK3.3 dN (3) dNda Ienv
= Cenv [S/R\':1 1]0.5 T°.5 LlK1.65 R
TR is the rise time of the cycle (in seconds), which will be taken to be half the cycle period. LlK is Kma:c Kmin . The S term depends on the R-ratio, which is Kmin/Kmax.
I R~O S(R)= 1+1.8R R~0.79 (4)
{
- 43.35 + 57.97 R R > 0.79 The constants 1.8 and 0.79 in Equation 4 are in Reference 2, but the values are 1.8 and 0.8 in Reference
- 3. The Reference 2 values are used, because they exhibit the required behavior of a continuous S at the transition point (0.79).
The full range of K is considered (i.e. the negative stress intensity factor is included.) The constants Cair and Cenv for austenitic stainless steel are given in Reference 3 for crack growth rates in meters/cycle and stress intensity factors in MPa-m 1l2
- The maximum operating temperature of 350°F [5] is used in the evaluation. Also, it should be noted that the frequency effect on the crack growth per cycle (rise time, TR ) appears in Equation 3.
5.3 Consideration of SCC for Flaws in Austenitic Welds Sensitization is a term that describes the precipitation of chromium carbides at the grain boundaries of austenitic stainless steel and nickel-base alloys and the subsequent susceptibility of these alloys to intergranular corrosion in aqueous media following certain heat treatments such as welding or furnace File No.: 1100643.301 Page 11 of 19 Revision: 0 F0306-01Rl
post weld heat treatments (PWHTs). The precipitation of chromium-rich carbides (e.g., Cr23C6 in austenitic stainless steel and Cr7C3 in Alloy 600) along grain boundaries depletes the region adjacent to the boundaries of chromium and induces susceptibility to intergranular corrosion due to the creation of a small anodic area surrounded by a much larger cathodic area. The most common example of sensitization is the intergranular corrosion (lOA) or intergranular stress corrosion cracking (IOSCC) susceptibility of the HAZ near the weld.
The superior resistance of duplex stainless steels to sensitization and the high resistance of the weld material to lOA and IOSCC have been known for over 70 years [8]. Many studies have demonstrated that the resistance of two-phase, austenitic-ferritic stainless steel weld metal and castings is a strong function of microstructure. Specifically, in work performed on wrought duplex stainless steels, the resistance to sensitization was shown to be a function of chemistry (e.g., carbon and chromium), as well as the amount and distribution of ferrite [8, 9].
Figure 5-1 presents material failure / non-failure data on a graph of carbon versus ferrite for various types of specimens (e.g., full size pipes, constant extension rate, variable-load and constant load) exposed an environment of high purity water <<1 IlS/cm) with 6 +/- 2 ppm dissolved oxygen at a temperature 550°F (288°C) [9]. This plot represents the traditional approach of evaluating various casting heats. The results of this extensive test program revealed that for welded applications, a control on ferrite of 5% is recommended, and in furnace-sensitized applications, 12% ferrite will assure resistance to IOSCC.
These measurements should be made after the mill solution heat treatment.
The U. S. Nuclear Regulatory Commission (NRC) considers weld metal and castings to be resistant to IOSCC: "Low carbon weld metal, including types 308L, 316L, 309L and similar grades, with a maximum carbon content of 0.035% and a minimum of7.5% ferrite (or 7.5 FN) as deposited" are considered resistant to IOSCC [10, 11]. The NRC further states that "welds joining resistant material that meet the ASME Boiler and Pressure Vessel Code requirement of 5% ferrite (or 5 FN), but are below 7.5% ferrite (or 7.5 FN) may be sufficiently resistant, depending on carbon content and other factors.
These will be evaluated on an individual case basis." Since these data represent IOSCC at 550°F (288°C), results at lower temperatures demonstrate even better resistance to stress corrosion cracking in these aggressive oxygenated environments.
Thus, since the Braidwood Station, Unit 2 RHR mixing tee flaw is located in the weld, SCC crack propagation is not anticipated due to the high SCC resistance of the duplex microstructure of the weld metal.
File No.: 1100643.301 Page 12 of 19 Revision: 0 F0306-01Rl
0.10,...--------------------------------------..
CLOSED SYMBOL - IGSCC 0.0& UHl:S Rl:'RESENT LOWER BOUNDARY Of' STRESS CORROSION FAILURES OPEN SYMBOL - NO IGSCC HAI.F-FII.I.ED SYMIlOI. IGSCC - AT LEAST OftE SAMPlE CROSS-HATCHED SYMBOL - MINOR ENVIROftMENTAl INFLUENCE 0.08
<> 6:n"c 11150"Fln4h t> 10ll0"F/.
0.04 d AS*WE I.DEO. AW + L1'$. AW + SHT V STULITE HAROSVRFACED 0.03 t>
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°0~--~-~--~-~ -~-~~-~--~-~:--~-~~-~--~-~~-~30 FERRITt -I Figure 5-1: IGSCC Resistance in Weld Metal may be predicted by the Combined Influence of Carbon Content and Percent Ferrite [9]
File No.: 1100643.301 Page 13 of 19 Revision: 0 F0306-01Rl
6.0 RESULTS 6.1 Allowable Flaw Sizes ASME Code allowable flaw sizes have been evaluated for semi-elliptical flaws with an aspect ratio al2e, of 0.2, where a is the crack depth and e is the half crack length. The results of the allowable flaw size calculations are presented in Table 6-1. The table presents the calculated allowable semi-elliptical flaw depth for various flaw lengths for the different service levels. It can be seen that the most limiting service level is Service Level D with the pipe break load. The corresponding allowable flaw size is 75%
of the wall thickness, i.e., 0.242 inches, for a flaw length up to 20% of the pipe circumference, i.e., 5.4 inches.
Also, the total piping stress to be used as the mean stress in the fatigue crack growth evaluation, in addition to the room temperature yield stress of 30 ksi, is shown in Table 6-1.
6.2 Fatigue Crack Growth The crack growth analysis is performed for a range of cyclic temperature frequencies and results are reported herein for the frequency that results in the minimum computed time to grow from the initial flaw size to the maximum allowable flaw depth. Calculations are performed with a fixed crack aspect ratio of 0.2. The results of the fatigue crack growth evaluations for the RHR mixing tees and adjacent welds are presented in Figure 6-1 and summarized in Table 6-2.
Figures 6-1 shows the predicted time for a semi-elliptical crack to grow to the calculated ASME Code allowable crack depth for all service levels. It can be observed from the output and Figure 6-1 that at a
~T = 162°F, it would take approximately 3692 hours0.0427 days <br />1.026 hours <br />0.0061 weeks <br />0.0014 months <br /> for the flaw to reach the Code allowable depth of 0.242 inch for the bounding thermal cycling frequency of 0.135 Hz.
File No.: 1100643.301 Page 14 of19 Revision: 0 F0306-01RI
Table 6-1: Allowable Part Through-Wall Flaw Size Allowable Flaw Size Calculation Circumferential Flaw Dimensjons
= Ro (in) 4.31 Ri (in) 3.99 tnom (in) 0.32 (in Z
3
)
16.8 NPS 8
Zfactor 1.352 p MX MY MZ cr Load (psi) (in-Ib) (in-Ib) (in-Ib) (ksi)
Pressure --- --- --- --- 3.574 OW --- --- --- --- 0.793 Thermal --- --- --- --- 6.373 Occasional --- --- --- --- 1.829 Pipe Break --- --- --- --- 10.160 Faulted --- --- --- --- 2.695 Total 11.776
Type 304 SS @ 350 F Service cr m crb cr. SFm SF b S. Su crf Stress Ratio Applicability Level (ksi) (ksil (ksi) Iksil (ksi) (ksi) Comb Memb Check A 3.574 1.829 6.373 2.7 2.3 21.55 65.1 43.3 .30 YES B 3.574 1.829 6.373 2.4 2.0 21.55 65.1 43.3 0.268 0.27 YES 0 3.574 10.160 6.373 1.3 1.4 21.55 65.1 43.3 0.571 0.14 YES 0 3.574 2.695 6.373 1.3 1.4 21.55 65.1 43.3 0.338 0.14 YES 11.776
~ Allowable Flaw Depth-to-Thickness Ratio Ratio of Flaw Length to Pipe Circumference, l(nD 0 I 0.1 0.2 0.3 0.4 I 0.5 0.6 0.75 Service Flaw Length, 11 (degree)
Level 0 36 72 108 144 180 216 270 A 0.75 0.75 0.75 0.75 0.74 0.672 0.62 0.58 B 0.75 0.75 0.75 0.75 0.75 0.70 0.65 0.61 0 0.75 0.75 0.75 0.69 0.56 0.49 0.44 0.41 0 0.75 0.75 0.75 0.75 0.75 0.71 0.66 0.61
- Total Piping Stress (Pressure Stress + DW + Thermal + Seismic)
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Table 6-2: Time for Reported Flaw to Grow to Code Allowable at a RHR HX AT of 162°F Crack Aspect Ratio (a/I) = 0.2, Limiting Frequency = 0.15 Hz Temp Time Amp a c da/dN (Hours) (OF) (in) (in) (in/cycle) 19 81 0.143 0.3576 3.21E-07 57 81 0.1487 0.3717 2.90E-07 95 81 0.1538 0.3845 2.63E-07 133 81 0.1584 0.396 2.38E-07 171 81 0.1626 0.4065 2.17E-07 209 81 0.1664 0.4161 1.98E-07 247 81 0.1699 0.4249 1.82E-07 285 81 0.1732 0.4329 1.67E-07 323 81 0.1761 0.4403 1.54E-07 361 81 0.1789 0.4472 1.43E-07 399 81 0.1814 0.4535 1.32E-07 437 81 0.1838 0.4594 1.23E-07 475 81 0.186 0.4649 1.15E-07 513 81 0.188 0.47 1.08E-07 551 81 0.1899 0.4748 1.01E-07 589 81 0.1917 0.4794 9.50E-08 3420 81 0.2409 0.6024 8.57E-09 3439 81 0.241 0.6026 8.48E-09 3458 81 0.2411 0.6028 8.39E-09 3477 81 0.2412 0.6029 8.30E-09 3496 81 0.2413 0.6031 8.21E-09 3572 81 0.2416 0.6039 7.88E-09 3591 81 0.2416 0.6041 7.80E-09 3610 81 0.2417 0.6042 7.72E-09 3667 81 0.2419 0.6048 7.50E-09 3686 81 0.242 0.6049 7.42E-09 3692 81 0.242 0.605 7.40E-09 File No.: 1100643.301 Page 16 of19 Revision: 0 F0306-01Rl
RHR Mixing Tee Crack Growth 0.3 0.25 0.2 g
...c-
.r:.
(lJ 0.15 Q
~
U u
...
III 0.1 0.05 0
0 500 1000 1500 2000 2500 3000 3500 4000 Time (Hours)
Figure 6-1. Crack Growth Results File No.: 1100643.301 Page 17 of 19 Revision: 0 F0306*01Rl
7.0 CONCLUSION
A thermal fatigue crack growth evaluation has been performed for the RHR mixing tee at the Braidwood Station Unit 2. The calculated time it would take for the semi-elliptical, 43% wall thickness (0.14 inch) deep flaw to grow to the ASME Code allowable flaw depth, at the maximum ilT of 162°F, has been determined to be at least 3692 hours0.0427 days <br />1.026 hours <br />0.0061 weeks <br />0.0014 months <br />. This allowable fatigue crack growth time far exceeds the number of hours the thermal mixing locations would be at ilT above 144°F in multiple cycles of operation.
The evaluation documented herein includes many conservative assumptions including the use of the maximum ilT of 162°F for the entire fatigue crack growth period which reduces the time it takes a postulated flaw to reach the ASME Code allowable size.
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8.0 REFERENCES
- 1. Materials Reliability Program: Assessment ofRHR Mixing Tee Thermal Fatigue in PWR Plants (MRP-192), EPRI, Palo Alto, CA: 2006. 1013305.
- 2. ASME Boiler and Pressure Vessel Code,Section XI, 2001 Edition with Addenda through 2003.
- 3. W. J. Shack and T. F. Kassner, "Review of Environmental Effects on Fatigue Crack Growth of Austenitic Steels," NUREG/CR-6l76 with errata, September 1993.
- 4. ASME Boiler and Pressure Vessel Code,Section II, Part D, 2001 Edition with Addenda through 2003.
- 5. Braidwood Station, Transmittal of Design Information (TODI), TODI # DIT-BRW-20ll-0029, Revision 0 dated April 27, 2011 and Revision 1 dated April 29, 2011, SI File No. 1100643.201.
- 6. ASME Boiler and Pressure Vessel Code,Section III, Subsection NC, 1974 Edition with Addenda through 1976.
- 7. R. O. McGill, D. O. Harris, K. Wolfe, "Method for Predicting Mixing Tee Thermal Fatigue in Carbon Steel Based on Austenitic Stainless Steel Operating Experience," Proceedings of the 2009 ASME Pressure Vessel and Piping Division Conference, PVP 2009-77975, July 26-30, Prague, Czech Republic.
- 8. H. Menendez, J. S. Chen and T. M. Devine, "The Influence of Microstructure on the Sensitization Behavior of Duplex Stainless Steel Welds," paper 562 presented at Corrosion 89, NACE, New Orleans, Louisiana, April 17-21, 1989.
- 9. N. R. Hughes, W. L. Clarke and D. E. Delwiche, "Intergranular Stress Corrosion Cracking Resistance of Austenitic Stainless Steel Castings," Stainless Steel Castings. ASTM STP 756, V. G.
Behal and A. S. Melilli, Eds. American Society for Testing and Materials, 1982, p. 26.
- 10. W. S. Hazelton and W. H. Koo, "Technical Report on Material Selection and Processing Guidelines for BWR Coolant Pressure Boundary Piping," NUREG-0313, Rev. 2, US Nuclear Regulatory Commission, January 1988.
- 11. US NRC Generic Letter 88-01, "NRC Position on IGSCC in BWR Austenitic Stainless Steel,"
January 25, 1988.
- 12. H.S. Carslaw and J.e. Jaeger, Conduction ofHeat in Solids, second edition, Clarendon Press, Oxford, 1959
- 13. EPRI Report, "Alloy 82/182 Pipe Butt Weld Safety Evaluation for US PWR Plant Designs:
Westinghouse and CE Plant Designs," EPRI, Palo Alto, CA, TP-I00l491.
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