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MONTHYEARRS-11-050, Third 10-Year Inservice Inspection Interval Relief Request I3R-08, Alternative Requirements to ASME Section XI Appendix VII (Supplements 2 and 10), Examinations of Class 1 Pressure Retaining Welds Conducted from the Inside.2011-04-11011 April 2011 Third 10-Year Inservice Inspection Interval Relief Request I3R-08, Alternative Requirements to ASME Section XI Appendix VII (Supplements 2 and 10), Examinations of Class 1 Pressure Retaining Welds Conducted from the Inside. Project stage: Supplement ML11147A1872011-05-20020 May 2011 NRR E-mail Capture - Draft Supplemental Information Request for Braidwood Station, Units 1 and 2 - Relief Request I3R-08: Alternative Requirements to ASME Code Requirements for Class 1 Pressure Retaining Welds Project stage: Draft Other ML1112605322011-05-25025 May 2011 Unacceptable with Opportunity to Supplement Relief Request I3R-08, Alternative Requirements to ASME Code Requirements for Class 1 Pressure Retaining Welds Project stage: Other RS-11-087, Supplemental Information Supporting Relief Request 13R-08: Alternative Requirements to ASME Code Requirements for Class 1 Pressure Retaining Welds2011-06-0606 June 2011 Supplemental Information Supporting Relief Request 13R-08: Alternative Requirements to ASME Code Requirements for Class 1 Pressure Retaining Welds Project stage: Supplement ML1117200082011-06-17017 June 2011 NRR E-mail Capture - Braidwood Station, Units 1 and 2 - Acceptance Review of Relief Request I3R-08, Alternative Requirements to ASME Code for Class 1 Pressure Retaining Welds Project stage: Acceptance Review ML11266A0032011-09-22022 September 2011 NRR E-mail Capture - Draft RAI Braidwood - Relief Request I3R-08 (TAC Nos. ME6024 & ME6025) Project stage: Draft RAI ML1127005992011-10-12012 October 2011 Request for Additional Information Request for Relief I3R-08 Project stage: RAI RS-11-173, Response to Request for Additional Information Related to Relief Request I3R-082011-11-0202 November 2011 Response to Request for Additional Information Related to Relief Request I3R-08 Project stage: Response to RAI ML12052A3232012-02-21021 February 2012 E-Mail to Lisa Schofield Project stage: Other ML12058A0052012-02-24024 February 2012 Request for Additional Information E-Mail Project stage: RAI ML1206006032012-03-0909 March 2012 Request for Additional Information Regarding Request for Relief I3R-08 (TAC Nos. ME6024, and ME6025) Project stage: RAI RS-12-040, Response to Request for Additional Information Related to Braidwood Station Relief Request 13R-082012-03-12012 March 2012 Response to Request for Additional Information Related to Braidwood Station Relief Request 13R-08 Project stage: Response to RAI ML12108A1232012-04-19019 April 2012 Safety Evaluation in Support of the Third 10-Year Inservice Inspection Interval Request for Relief 13R-08 (Tac Nos. ME6024 and ME6025) Project stage: Approval 2011-06-06
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Category:Letter
MONTHYEARML24263A1272024-09-23023 September 2024 – Request for Additional Information (EPID 2023-LLA-0136) - Non-Proprietary IR 05000456/20240112024-09-12012 September 2024 Biennial Problem Identification and Resolution Inspection Report 05000456/2024011 and 05000457/2024011 ML24164A0032024-09-10010 September 2024 Issuance of Amendment Nos. 235 and 235 Revision of Technical Specifications for the Ultimate Heat Sink IR 05000456/20240052024-08-29029 August 2024 Updated Inspection Plan and Assessment Follow-Up Letter for Braidwood Station, Units 1 and 2 (Report 05000456/2024005 and 05000457/2024005) ML24227A0522024-08-29029 August 2024 Audit Plan for LAR to Remove ATWS Mtc Limit ML24225A1112024-08-13013 August 2024 Notification of NRC Fire Protection Team Inspection Request for Information ML24222A6772024-08-0909 August 2024 Response to Request for Additional Information for Application to Revise Technical Specifications to Adopt TSTF-591-A, Revise Risk Informed Completion Time (RICT) Program Revision 0 and Revise 10 CFR 50.69 License Condition IR 05000456/20240022024-08-0808 August 2024 Integrated Inspection Report 05000456/2024002 and 05000457/2024002 ML24172A1252024-07-26026 July 2024 Environmental Assessment and Finding of No Significant Impact Related to a Requested Increase in Ultimate Heat Sink Temperature (EPID L-2024-LLA-0075) - Transmittal Letter ML24179A3262024-07-23023 July 2024 LTR - Constellation - SG Welds and Nozzles (L-2023-LLR-0053, L-2023-LLR-0054, L-2023-LLR-0055, L-2023-LLR-0056) 05000456/LER-2024-001, Submittal of LER 2024-001-00 for Braidwood Station, Unit 1, Trip on Low Steam Generator Level Due to Failure to Verify Isolation Valves Were Open2024-07-0303 July 2024 Submittal of LER 2024-001-00 for Braidwood Station, Unit 1, Trip on Low Steam Generator Level Due to Failure to Verify Isolation Valves Were Open ML24163A3922024-06-25025 June 2024 Individual Notice of Consideration of Issuance of Amendments to Renewed Facility Operating Licenses, Proposed No Significant Hazards Consideration Determination, & Opportunity for a Hearing (EPID L-2024-LLA-0075)- Letter RS-24-061, Constellation Energy Generation, LLC, Response to NRC Regulatory Issue Summary 2024-01, Preparation and Scheduling of Operator Licensing Examinations2024-06-14014 June 2024 Constellation Energy Generation, LLC, Response to NRC Regulatory Issue Summary 2024-01, Preparation and Scheduling of Operator Licensing Examinations ML24164A2132024-06-13013 June 2024 Information Request to Support Upcoming Problem Identification and Resolution (Pi&R) Inspection at Braidwood Nuclear Plant RS-24-057, License Amendment to Technical Specification 3.7.9, Ultimate Heat Sink2024-06-0404 June 2024 License Amendment to Technical Specification 3.7.9, Ultimate Heat Sink IR 05000456/20240102024-05-31031 May 2024 License Renewal Phase 1 Report 05000456/2024010 RS-24-044, License Amendment to Transition to Framatome Gaia Fuel and Exemptions to 10 CFR 50.46 and 10 CFR 50 Appendix K. Attachments 1 to 11 Enclosed2024-05-28028 May 2024 License Amendment to Transition to Framatome Gaia Fuel and Exemptions to 10 CFR 50.46 and 10 CFR 50 Appendix K. Attachments 1 to 11 Enclosed RS-24-043, Application to Remove Power Distribution Monitoring System (Pdms) Details from Technical Specifications2024-05-24024 May 2024 Application to Remove Power Distribution Monitoring System (Pdms) Details from Technical Specifications ML24079A0762024-05-23023 May 2024 Issuance of Amendments to Adopt TSTF 264 ML24142A3352024-05-21021 May 2024 Quad Cities—Information Request to Support the NRC Annual Baseline Emergency Action Level and Emergency Plan Changes RS-24-055, 2023 Corporate Regulatory Commitment Change Summary Report2024-05-17017 May 2024 2023 Corporate Regulatory Commitment Change Summary Report ML24136A0132024-05-15015 May 2024 2023 Annual Radiological Environmental Operating Report ML24136A2452024-05-15015 May 2024 Notification of NRC Baseline Inspection and Request for Information; Inspection Report 05000457/2024004 ML24128A1212024-05-0707 May 2024 Response to Braidwood and Dresden FOF Dates Change Request (2025) ML24122A6522024-05-0101 May 2024 Submittal of 2023 Annual Radioactive Effluent Release Report RS-24-041, Alternative Request to Utilize Code Case OMN-32, Alternative Requirements for Range and Accuracy of Pressure, Flow, and Differential Pressure Instruments Used in Pump Tests2024-04-30030 April 2024 Alternative Request to Utilize Code Case OMN-32, Alternative Requirements for Range and Accuracy of Pressure, Flow, and Differential Pressure Instruments Used in Pump Tests IR 05000456/20243012024-04-29029 April 2024 NRC Initial License Examination Report 05000456/2024301; 05000457/2024301 RS-24-026, License Amendment to Braidwood Station, Units 1 and 2, and Byron Station, Units 1 and 2, Technical Specifications 5.6.5, Core Operating Limits Report (COLR)2024-04-25025 April 2024 License Amendment to Braidwood Station, Units 1 and 2, and Byron Station, Units 1 and 2, Technical Specifications 5.6.5, Core Operating Limits Report (COLR) ML24116A0052024-04-25025 April 2024 Transmittal of Braidwood Station, Unit 1, Core Operating Limits Report, Braidwood Unit 1 Cycle 25 IR 05000456/20240012024-04-24024 April 2024 Integrated Inspection Report 05000456/2024001 and 05000457/2024001 ML24113A1272024-04-22022 April 2024 Audit Plan in Support of Review of LAR Revision of TS 3.7.15, 3.7.16, and 4.3.1 (EPID: L-2023-LLA-0136) (Non-Proprietary) IR 05000457/20240902024-04-19019 April 2024 Final Significance Determination for 2b Auxiliary Feedwater Pump Diesel Engine Fuel Oil Dilution Issue - NRC Inspection Report 05000457/2024090 ML24103A2042024-04-12012 April 2024 Constellation Energy Generation, LLC, Application to Revise Technical Specifications to Adopt TSTF-591-A, Revise Risk Informed Completion Time (RICT) Program Revision 0 and Revise 10 CFR 50.69 License Condition RS-24-034, Annual 10 CFR 50.46 Report of Emergency Core Cooling System Evaluation Model Changes and Errors2024-04-10010 April 2024 Annual 10 CFR 50.46 Report of Emergency Core Cooling System Evaluation Model Changes and Errors ML24094A2692024-04-0303 April 2024 Notification of Deviation from Pressurized Water Reactor Owners Group (PWROG) Report, WCAP-17451-P, Revision 2, Reactor Internals Guide Tube Wear Westinghouse Domestic Fleet Operational Projections RS-24-002, Constellation Energy Generation, LLC - Annual Property Insurance Status Report2024-04-0101 April 2024 Constellation Energy Generation, LLC - Annual Property Insurance Status Report ML24057A0372024-03-26026 March 2024 Proposed Alternative from Certain Requirements Contained in the American Society of Mechanical Engineers Boiler and Pressure Vessel Code, Section XI RS-24-024, Response to Request for Additional Information Regarding Proposed Alternative for Examination of Pressurizer Circumferential and Longitudinal Shell-to-Head Welds and Nozzle-to-Vessel Welds2024-03-22022 March 2024 Response to Request for Additional Information Regarding Proposed Alternative for Examination of Pressurizer Circumferential and Longitudinal Shell-to-Head Welds and Nozzle-to-Vessel Welds ML24067A3252024-03-0707 March 2024 U.S. Department of Energy, Office of Legacy Management, 2023 Annual Site Inspection and Monitoring Report for Uranium Mill Tailings Radiation Control Act Title I Disposal Sites ML24066A0122024-03-0606 March 2024 Operator Licensing Examination Approval Braidwood Station, Units 1 and 2, March 2024 IR 05000456/20244012024-03-0505 March 2024 Cyber Security Inspection Report 05000456/2024401 and 05000457/2024401 (Public) IR 05000456/20230062024-02-28028 February 2024 Annual Assessment Letter for Braidwood Station, Units 1 and 2 (Report 05000456/2023006 and 05000457/2023006) ML24057A3022024-02-26026 February 2024 Regulatory Conference Supplemental Information ML24047A2382024-02-20020 February 2024 Regulatory Conference to Discuss Risk Associated with 2b Auxiliary Feedwater Pump Diesel Engine Fuel Oil Leak RS-24-013, Response to Request for Additional Information Regarding Proposed Alternative to Distribution Requirements of ASME Code Table IWC-2411-1 for the Steam Generators2024-02-13013 February 2024 Response to Request for Additional Information Regarding Proposed Alternative to Distribution Requirements of ASME Code Table IWC-2411-1 for the Steam Generators IR 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 2024-09-23
[Table view] Category:Safety Evaluation
MONTHYEARML24164A0032024-09-10010 September 2024 Issuance of Amendment Nos. 235 and 235 Revision of Technical Specifications for the Ultimate Heat Sink ML24194A0222024-07-22022 July 2024 Issuance of Relief Proposed Alternative Request Associated with Pressurizer Examinations ML24057A0372024-03-26026 March 2024 Proposed Alternative from Certain Requirements Contained in the American Society of Mechanical Engineers Boiler and Pressure Vessel Code, Section XI ML23277A0032023-12-11011 December 2023 Issuance of Amendments Regarding Adoption of TSTF-370 ML23241A9092023-09-19019 September 2023 Enclosure 2 - Non-Proprietary - Review of License Renewal Commitment Number 10 Safety Evaluation 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 ML23114A2522023-04-28028 April 2023 Request to Use a Provision of a Later Edition of the ASME Boiler & Pressure Vessel Code, Section XI ML22364A0242023-03-0101 March 2023 R. E. Ginna Nuclear Power Plant Issuance of Amendments Nos. 231, 231, 232, 232, and 154 Regarding Adoption of TSTF-246 ML22293B8052022-11-30030 November 2022 Constellation Energy Generation, Llc_Fleet - Request to Authorize Use Honeywell Mururoa V4F1 R Supplied Air Suits ML22307A2462022-11-10010 November 2022 Proposed Alternative to the Requirements of the American Society of Mechanical Engineers Boiler & Pressure Vessel Code (EPIDs L-2021-LLR-0035 and L-2021-LLR-0036) ML22210A0312022-08-30030 August 2022 Issuance of Amendments Nos. 230, 230, 230, and 230, Respectively, Regarding Adoption of Technical Specifications Task Force Traveler (TSTF) 501, Revision 1, Relocate Stored Fuel Oil and Lube Oil Volume Values to Licensee Control ML22173A1812022-08-11011 August 2022 Issuance of Amendment No. 229 to Remove License Condition ML22173A2142022-08-10010 August 2022 Issuance of Amendments Nos. 228 and 228 Revision of Technical Specifications for the Ultimate Heat Sink ML22095A2702022-05-12012 May 2022 Issuance of Amendment Nos. 227, 227, 229, 229, and 245, Respectively, Regarding Adoption of TSTF 273, Safety Function Determination Program Clarifications ML22090A0862022-04-29029 April 2022 Amendments to Adopt TSTF-541,Rev.2,Add Exceptions to Surveillance Requirements for Valves,Dampers Locked in Actuated Position ML22026A4892022-03-22022 March 2022 Issuance of Amendment Nos. 225, 225, 227, 227, and 148, Respectively, Regarding Issues Identified in Westinghouse Documents (EPID L-2021-LLA-0066) Nonproprietary ML21347A0382022-01-13013 January 2022 Issuance of Amendments to Revise Reactor Coolant Leakage Requirements ML21277A2482021-11-16016 November 2021 Letter with Enclosure 4, Safety Evaluation for Transfer of Licenses and Draft Conforming License Amendments (Public Version) ML21280A0782021-10-27027 October 2021 Proposed Alternative to Use of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code Case N-893 (Epids L-2020-LLR-0147 and L-2020-LLR-0148) ML21230A2062021-09-0303 September 2021 Proposed Alternative to Use ASME OM Code Case OMN-28 ML21216A2202021-08-0505 August 2021 Proposed Alternative to Eliminate Certain Documentation Requirements for Pressure Retaining Bolting ML21134A0062021-08-0303 August 2021 Proposed Alternatives to the Requirements of the American Society of Mechanical Engineers Boiler & Pressure Vessel Code (Epids L-2020-LLR-0099 and L-2020-LLR-0100) ML21154A0462021-07-13013 July 2021 Issuance of Amendments Nos. 222 and 222 Revision of Technical Specifications for the Ultimate Heat Sink ML21166A1682021-06-25025 June 2021 ML21060B2812021-04-0202 April 2021 Issuance of Amendments Nos. 221, 221, 224, and 224 Regarding Technical Specifications 3.8.1, AC Sources-Operating ML21054A0082021-03-10010 March 2021 Issuance of Amendment Nos. 220 and 220 One-Time Deferral of Steam Generator Tube Inspections ML21063A0162021-03-0808 March 2021 Relief from the Requirements of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code, Alternative to 10 CFR 50.55a(z)(2) ML21039A6362021-02-17017 February 2021 R. E. Ginna - Proposed Alternative to Use the American Society of Mechanical Engineers Boiler and Pressure Vessel Code Case N-885 ML21028A6732021-02-0303 February 2021 Request to Use a Provision of Later Edition of the ASME Boiler and Pressure Vessel Code, Section XI (EPID L-2020-LLR-011 ML20317A0012020-12-28028 December 2020 Non-Proprietary, Issuance of Amendment Nos. 219, 219, 223, and 223, Revise Loss-of-Coolant Accident Methodology in TS 5.6.5, Core Operating Limits Report (COLR) ML20287A1302020-11-0505 November 2020 Review of Quality Assurance Program Changes ML20269A2002020-09-30030 September 2020 Request to Use a Provision of Later Edition of the ASME Boiler and Pressure Vessel Code, Section XI (EPID L2020-LLR-0117 ML20245E4192020-09-24024 September 2020 Issuance of Amendments Nos. 218 and 218 Revision of Technical Specifications for the Ultimate Heat Sink ML20163A0462020-09-18018 September 2020 Issuance of Amendments Nos. 217, 217, 221, and 221, Revise Technical Specification 5.6.6 to Allow Use of Areva Np Topical Report BAW-2308 ML20167A0072020-09-11011 September 2020 R. E. Ginna - Issuance of Amendment Nos. 216, 216, 220, 220, and 143 - Adoption of TSTF-567, Rev. 1, Add Containment Sump Technical Specifications to Address GSI-191 Issues ML20149K6982020-09-10010 September 2020 Issuance of Amendment Nos. 215, 215, 219, and 219 Permanent Extension of Type a and Type C Leak Rate Test Frequencies ML20232A1712020-09-0101 September 2020 Request to Use Alternative Code Case OMN-26 ML20153A8042020-07-31031 July 2020 Co. - Issuance of Amendments Revising Emergency Action RA3 ML20141L6362020-07-10010 July 2020 Issuance of Amendments Based on TSTF-427,Allowance for Nontechnical Specification Barrier Degradation on Supported System Operability,Rev 2 ML20134H9402020-07-0808 July 2020 Issuance of Amendments Revising the High Radiation Area Administrative Controls ML20118C4292020-06-0909 June 2020 Issuance of Amendments Revision of Technical Specifications for the Ultimate Heat Sink ML20133K0932020-05-14014 May 2020 Relief from the Requirements of the ASME Code ML20111A0002020-05-0101 May 2020 Issuance of Amendment No. 209, Revision Technical Specification 5.5.9, Steam Generator (SG) Program, for One-Time Revision to Frequency of SG Tube Inspections (Exigent Circumstances) ML20113F0412020-04-30030 April 2020 Proposed Alternative to the Submittal Schedule for Certain Reports (COVID-19) ML20021A0702020-04-0606 April 2020 Issuance of Amendments to Delete License Conditions for Decommissioning Trusts ML19331A7252020-02-14014 February 2020 Issuance of Amendments Revising Emergency Action Levels ML20028E3992020-02-0404 February 2020 Proposed Alternative to Use ASME Code Case N-879 ML19269C5342019-09-27027 September 2019 Proposed Alternative to Use Encoded Phased Array Ultrasonic Examination Techniques ML19240B1122019-09-0909 September 2019 Proposed Alternative to the Requirements of the American Society of Mechanical Engineers Code 2024-09-10
[Table view] Category:Code Relief or Alternative
MONTHYEARML24179A3262024-07-23023 July 2024 LTR - Constellation - SG Welds and Nozzles (L-2023-LLR-0053, L-2023-LLR-0054, L-2023-LLR-0055, L-2023-LLR-0056) ML24194A0222024-07-22022 July 2024 Issuance of Relief Proposed Alternative Request Associated with Pressurizer Examinations ML22307A2462022-11-10010 November 2022 Proposed Alternative to the Requirements of the American Society of Mechanical Engineers Boiler & Pressure Vessel Code (EPIDs L-2021-LLR-0035 and L-2021-LLR-0036) 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-22-036, 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 ML21280A0782021-10-27027 October 2021 Proposed Alternative to Use of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code Case N-893 (Epids L-2020-LLR-0147 and L-2020-LLR-0148) ML21230A2062021-09-0303 September 2021 Proposed Alternative to Use ASME OM Code Case OMN-28 ML21216A2202021-08-0505 August 2021 Proposed Alternative to Eliminate Certain Documentation Requirements for Pressure Retaining Bolting ML21134A0062021-08-0303 August 2021 Proposed Alternatives to the Requirements of the American Society of Mechanical Engineers Boiler & Pressure Vessel Code (Epids L-2020-LLR-0099 and L-2020-LLR-0100) RS-21-008, Request for Alternative: One-Time Deferral of Follow-Up Inspection for Reactor Pressure Vessel Head Penetration Nozzles with Mitigated Alloy 600/82/182 Peened Surfaces in Accordance with 10 CFR 50.55a(z)(2)2021-01-25025 January 2021 Request for Alternative: One-Time Deferral of Follow-Up Inspection for Reactor Pressure Vessel Head Penetration Nozzles with Mitigated Alloy 600/82/182 Peened Surfaces in Accordance with 10 CFR 50.55a(z)(2) ML20269A2002020-09-30030 September 2020 Request to Use a Provision of Later Edition of the ASME Boiler and Pressure Vessel Code, Section XI (EPID L2020-LLR-0117 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 ML20113F0412020-04-30030 April 2020 Proposed Alternative to the Submittal Schedule for Certain Reports (COVID-19) ML20099D9552020-04-17017 April 2020 Request to Use Provisions in the 2013 Edition of the ASME Boiler and Pressure Vessel Code for Performing Non-Destructive Examinations RS-20-020, Relief Request for Alternative Frequency to Supplemental Indication Requirements of 10 CFR 50.55a(b)(3)(xi)2020-02-28028 February 2020 Relief Request for Alternative Frequency to Supplemental Indication Requirements of 10 CFR 50.55a(b)(3)(xi) ML19269C5342019-09-27027 September 2019 Proposed Alternative to Use Encoded Phased Array Ultrasonic Examination Techniques ML19136A3862019-06-0505 June 2019 Relief from the Requirements of the American Society of Mechanical Engineers Code ML19141A0202019-06-0505 June 2019 Relief from the Requirements of the American Society of Mechanical Engineers Code ML19155A0602019-06-0505 June 2019 Relief from the Requirements of the American Society of Mechanical Engineers Code ML19161A2572019-06-0404 June 2019 BWR Fleet Msv/Srv - Testing Frequency Relief Request NRC Pre-Application Meeting June 4, 2019 JAFP-19-0023, Relief Request Associated with the Use of Encoded Phased Array Ultrasonic Examination Techniques in Lieu of Radiography for Ferritic and Austenitic Piping Welds2019-02-15015 February 2019 Relief Request Associated with the Use of Encoded Phased Array Ultrasonic Examination Techniques in Lieu of Radiography for Ferritic and Austenitic Piping Welds RS-19-015, Relief Request 14-RS-1 to Implement Code Case OMN-132019-01-31031 January 2019 Relief Request 14-RS-1 to Implement Code Case OMN-13 ML18347B4192019-01-17017 January 2019 Relief Request I4R-03, Relief from ASME Requirements for the Fourth 10-Year Inservice Inspection Interval Related to Degraded Canopy Seal Welds Associated with Control Rod Drive Mechanism ML18318A3342019-01-17017 January 2019 Relief from the Requirements of the American Society of Mechanical Engineers Code ML18331A0372019-01-17017 January 2019 Relief from the Requirements of the American Society of Mechanical Engineers Code ML18305A3602018-12-0606 December 2018 Relief from the Requirements of the American Society of Mechanical Engineers Code RS-18-125, Proposed Alternative Requirements for the Repair and Examination of Reactor Pressure Vessel Head Penetration Nozzles for the Fourth Lnservice Inspection Interval in Accordance with 10 CFR 50.55a(z)(1)2018-10-11011 October 2018 Proposed Alternative Requirements for the Repair and Examination of Reactor Pressure Vessel Head Penetration Nozzles for the Fourth Lnservice Inspection Interval in Accordance with 10 CFR 50.55a(z)(1) RS-18-123, Request for Alternative Follow-up Inspection for Reactor Pressure Vessel Head Penetration Nozzles with Mitigated Alloy 600/82/182 Peened Surfaces in Accordance with 10 CFR 50.55a(z)(2)2018-09-24024 September 2018 Request for Alternative Follow-up Inspection for Reactor Pressure Vessel Head Penetration Nozzles with Mitigated Alloy 600/82/182 Peened Surfaces in Accordance with 10 CFR 50.55a(z)(2) JAFP-18-0052, Proposed Alternative to Utilize Code Case N-8792018-05-30030 May 2018 Proposed Alternative to Utilize Code Case N-879 JAFP-18-0053, Proposed Alternative to Utilize Code Cases N-878 and N-8802018-05-30030 May 2018 Proposed Alternative to Utilize Code Cases N-878 and N-880 RS-17-168, Request for Relief for Extension of Examination Interval for Reactor Pressure Vessel Head Penetration Nozzles with Mitigated Alloy 600/82/182 Peened Surfaces in Accordance with 10 CFR 50.55a(z)(1)2017-12-20020 December 2017 Request for Relief for Extension of Examination Interval for Reactor Pressure Vessel Head Penetration Nozzles with Mitigated Alloy 600/82/182 Peened Surfaces in Accordance with 10 CFR 50.55a(z)(1) RS-17-168, Braidwood Station, Unit 2, Request for Relief for Extension of Examination Interval for Reactor Pressure Vessel Head Penetration Nozzles with Mitigated Alloy 600/82/182 Peened Surfaces in Accordance with 10 CFR 50.55a(z)(1)2017-12-20020 December 2017 Braidwood Station, Unit 2, Request for Relief for Extension of Examination Interval for Reactor Pressure Vessel Head Penetration Nozzles with Mitigated Alloy 600/82/182 Peened Surfaces in Accordance with 10 CFR 50.55a(z)(1) ML17249A2982017-11-13013 November 2017 Relief from the Requirements of the ASME Code (CAC No. MF9597; EPID L-2017-LLR-0021) ML17170A0132017-06-26026 June 2017 Proposed Alternative to Eliminate Examination of Threads in Reactor Pressure Vessel Flange (CAC Nos. MF8712-MF8729 and MF9548) ML17095A2682017-03-31031 March 2017 Submittal of Request for Relief for Extension of Examination Interval for Reactor Pressure Vessel Head Penetration Nozzles with Mitigated Alloy 600/82/182 Peened Surfaces: Attachment 1, Relief Request ML17054C2552017-03-15015 March 2017 Request for Relief from Requirements of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code) ML16230A2372016-09-0606 September 2016 Fleet Request for Proposed Alternative to Use ASME Code Case N-513-4 (CAC Nos. MF7301-MF7322) ML16162A2112016-06-29029 June 2016 Request for Use of Alternative ML16109A3372016-04-27027 April 2016 Relief from the Requirements of the ASME Code (CAC Nos. MF6715, MF6716, MF6717, and MF6718) ML14303A5062014-12-10010 December 2014 Relief from the Requirements of the ASME Code to Extend the Reactor Vessel Inservice Inspection Interval ML13016A5152013-01-30030 January 2013 Relief from the Requirements of the ASME Code for the Third 10-Year Interval of Inservice Inspection ML12121A6372012-05-10010 May 2012 Request to Use Code Case N-789 ML12108A1232012-04-19019 April 2012 Safety Evaluation in Support of the Third 10-Year Inservice Inspection Interval Request for Relief 13R-08 (Tac Nos. ME6024 and ME6025) ML1113306532011-06-0606 June 2011 Unacceptable with Opportunity to Supplement Alt. to ASME Code Requirements for Repair of Reactor Vessel Head Penetrations (TACs ME6071, ME6072, ME6073, and ME6074) ML1105909212011-03-0303 March 2011 Relief Request from ASME Code Case N-729-1 Requirements for Examination of Reactor Vessel Head Penetration Welds ML1011905272010-05-11011 May 2010 Relief Request 13R-06 for Detailed Visual Examination During Appendix J Pnuematic Leakage Testing ML1012301792010-05-10010 May 2010 Relief Request I3R-03 for Examination of Structural Weld Overlays ML1008504952010-03-26026 March 2010 Application Accepted - Braidwood & Byron Relief Request Re. ASME Code Case N-729-1 (TACs ME3510 - ME3513) RS-10-046, Request for Relief from ASME Code Case N-729-1 Requirements for Examination of Reactor Vessel Head Penetration Welds2010-03-12012 March 2010 Request for Relief from ASME Code Case N-729-1 Requirements for Examination of Reactor Vessel Head Penetration Welds ML1004806992010-03-0808 March 2010 Relief Request 12R-50 for Second 10-Year Inservice Inspection Interval 2024-07-23
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UNITED NUCLEAR REGULATORY WASHINGTON, D.C. 20555*0001 April 19, 2012 Mr. Michael President and Chief Nuclear Officer Exelon Nuclear 4300 Winfield Road Warrenville, IL 60555 BRAIDWOOD STATION, UNITS 1 AND 2 -SAFETY EVALUATION IN SUPPORT OF THE THIRD 10-YEAR INSERVICE INSPECTION INTERVAL REQUEST FOR RELIEF 13R-08 (TAC NOS. ME6024 AND ME6025)
Dear Mr. Pacilio:
By letter to the U.S. Nuclear Regulatory Commission (NRC) dated April 11,2011, and supplemented by letters dated June 6,2011, November 2,2011, and March 12,2012 (Agencywide Documents Access and Management System (ADAMS) Accession Numbers ML 111020263, ML 111580106, ML 113070349, and ML 120730196 respectively), Exelon Generation Company, LLC (the licensee) requested relief from the requirements of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code for Braidwood Station, Units 1 and 2 (Braidwood). Specifically, pursuant to Title 10 of the Code of Federal Regulations (10 CFR) 50.55a(g)(5)(iii), the licensee requested relief from the depth-sizing uncertainty qualification requirement for ultrasonic examinations conducted from the inside diameter of pipes, i.e., root mean square error not greater than 0.125 inches, contained in ASME Code Cases N-695, "Qualification Requirements for Dissimilar Metal Piping Welds,Section XI, Division 1," and N-696, "Qualification Requirements for Appendix VIII Piping Examinations Conducted From the Inside Surface,Section XI, Division 1," due to impracticality. The licensee initially requested that relief be granted for the remainder of the third 10-year inservice inspection interval of Braidwood, however, in its March 12, 2012 submittal, the licensee revised its request to include only examinations conducted at Braidwood, Unit 1 prior to the end of refueling outage A 1 R16 which is scheduled to begin on April 16, 2012 and examinations conducted at Braidwood, Unit 2 prior to the end of refueling outage A2R16 which is scheduled to begin in the fall of 2012. The NRC staff determined that granting relief pursuant to 10 CFR 50.55a(g)(6)(i) is authorized by law and will not endanger life or property or the common defense and security, and is otherwise in the public interest giving due consideration to the burden upon the licensee that could result if the requirements were imposed on the facility. Accordingly, the NRC staff concludes that the licensee has adequately addressed all of the regulatory requirements set forth in 10 CFR 50.55a. Therefore, the NRC staff grants the use of the alternate depth-sizing qualification (0.189 inch for ASME Code Case N-695 welds and 0.245 inch for ASME Code Case N-696 welds), subject to the licensee providing, for NRC staff review and approval prior to the expiration of the relief, information identified in the safety evaluation (SE) should a crack requiring depth-sizing be identified. This relief is granted for Braidwood Units 1 and 2, until prior to startup following refueling outage A1 R16 and A2R16, respectively. The NRC staffs SE is enclosed. Please contact the Braidwood Project Manager, Brenda Mozafari at (301) 415-2020 if you have any questions on this action.
Sincerely,erman, Chief Plant Licensing Branch 111-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. 50-456 and Safety cc w/encl: Distribution via Listserv UNITED NUCLEAR REGULATORY WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION ON THE THIRD 10-YEAR INSERVICE INSPECTION INTERVAL REQUEST FOR RELIEF 13R-08, REVISION 0 EXELON GENERATION COMPANY, LLC BRAIDWOOD STATION, UNITS 1 AND 2 DOCKET NUMBERS 50-456 AND 50-457 1.0 INTRODUCTION By letter to the U.S. Nuclear Regulatory Commission (NRC, the CommiSSion) dated April 11, 2011, and supplemented by letters dated June 6, 2011, November 2, 2011, and March 12, 2012 (Agencywide Documents Access and Management System (ADAMS) Accession Numbers ML 111020263, ML 111580106, ML113070349, and ML 120730196, respectively), Exelon Generation Company, LLC (the licensee) requested relief from the requirements of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code for Braidwood Station, Units 1 and 2 (Braidwood). Specifically, pursuant to Title 10 of the Code ofFederal Regulations (CFR), Section 50.55a(g)(5)(iii), the licensee requested relief from the depth-sizing uncertainty qualification requirement for ultrasonic (UT) examinations conducted from the inside diameter (ID) of pipes, i.e., root mean square (RMS) error not greater than 0.125 inches, contained in ASME Code Cases N-695, "Qualification Requirements for Dissimilar Metal Piping Welds,Section XI, Division 1," and N-696, "Qualification Requirements for Appendix VIII Piping Examinations Conducted From the Inside Surface,Section XI, Division 1," due to impracticality. The licensee initially requested that relief be granted for the remainder of the third 10-year inservice inspection interval (lSI) for Braidwood, however, in its March 12,2012 submittal, the licensee revised its request to include only examinations conducted at Braidwood, Unit 1 prior to the end of refueling outage A1R16 which is scheduled to begin on April 16.2012 and examinations conducted at Braidwood. Unit 2 prior to the end of refueling outage A2R16 which is scheduled to begin in the fall of 2012. 2.0 REGULATORY REQUIREMENTS In its letter dated June 6. 2011, the licensee requests relief from the depth-sizing acceptance criteria contained in ASME Code Cases N-695 and N-696 pursuant to 10 CFR 50.55a(g}(5}(iii).
-ASME Code Cases N-695 and N-696 are accepted for use in NRC Regulatory Guide (RG) 1.147, Revision 16, "Inservice Inspection Code Case Acceptability ASME Section XI, Division 1," and incorporated by reference in 10 CFR 50.55a(b). Section 50.55a(g)(4)(ii) of 10 CFR states, in part, that inservice examination of components must comply with the latest edition of the Code incorporated by reference in 10 CFR 50.55a(b) or the optional ASME Code cases listed in RG 1.147. Section 50.55a(g)(5}(iii} of 10 CFR states, in part, that licensees may determine that conformance with certain code requirements is impractical and that the licensee shall notify the Commission and submit information in support of the determination. Section 50.55a(g}(6}(i) of 10 CFR states, in part that the Commission will evaluate determinations under 10 CFR 50.55a(g)(5) that code requirements are impractical and that the Commission may grant such relief and may impose such alternative requirements as it determines is authorized by law and will not endanger life or property. Based on the above, and subject to the following technical evaluation, the NRC staff finds that regulatory authority exists for the licensee to request and the Commission to grant the relief requested by the licensee. 3.0 TECHNICAL EVALUATION 3.1 Component Descriptions The licensee stated that this relief request addresses nozzle-to-safe end dissimilar metal welds and adjacent safe end-to-pipe or safe end-to-elbow austenitic welds at the locations listed in Tables 1 and 2 below. Table Dissimilar Metal Category B-F Item Number B5.10/RI-ISI Category R-A Item Number R1.15 Welds Component Number Unit Description 1 RV-01-022 1 Reactor Pressure Vessel (RPV) Nozzle-to-Safe End Hot leg, loop C 1 RV-01-023 1 RPV Nozzle-to-Safe End Cold leg, loop C 1 RV-01-024 1 RPV Nozzle-to-Safe End Cold leg, loop D 'I RV-01-025 1 RPV Nozzle-to-Safe End Hot leg, loop D 1 RV-01-026 1 RPV Nozzle-to-Safe End Hot leg, loop A 1 RV-01-027 1 RPV Nozzle-to-Safe End Cold leg, loop A 'I RV-01-028 1 RPV Nozzle-to-Safe End Cold leg, loop B 1 RV-OI-029 1 RPV Nozzle-to-Safe End Hot leg, loop B 2RV-01-022 2 RPV Nozzle-to-Safe End Hot leg, loop D 2RV-01-023 2 RPV Nozzle-to-Safe End Cold leg, loop D 2RV-01-024 2 RPV Nozzle-to-Safe End Cold leg, loop C 2RV-01-025 2 RPV Nozzle-to-Safe End Hot leg, loop C 2RV-01-026 2 RPV Nozzle-to-Safe End Hot leg, loop B 2RV-01-027 2 RPV Nozzle-to-Safe End Cold leg, loop B
-3 2RV-01-028 2 RPV Nozzle-to-Safe End Cold leg, loop A 2RV-01-029 2 RPV Nozzle-to-Safe End Hot leg, loop A Table 2: Austenitic Category 8-J Item Number 89.11 RI-ISI Category R-A Item Number R1.20 Welds Component Number Unit Description 1 RV-01-030 1 RPV Safe End-to-Pipe Hot leg, loop C 1 RV-01-031 1 RPV Safe End-to-Elbow Cold leg, loop C 1 RV-01-032 1 RPV Safe End-to-Elbow Cold leg, loop D 'I RV-01-033 1 RPV Safe End-to-Pipe Hot leg, loop D 1 RV-01-034 1 RPV Safe End-to-Pipe Hot leg, loop A 1 RV-01-035 1 RPV Safe End-to-Elbow Cold leg, loop A 'I RV-01-036 1 RPV Safe End-to-Elbow Cold leg, loop 8 1 RV-01-037 1 RPV Safe End-to-Pipe Hot leg, loop 8 2RV-01-030 2 RPV Safe End-to-Pipe Hot leg, loop D 2RV-01-031 2 RPV Safe End-to-Elbow Cold leg, loop D RPV Safe End-to-Elbow Cold leg, loop C 2RV-01-032 2 2RV-01-033 2 RPV Safe End-to-Pipe Hot leg, loop C 2RV-01-034 2 RPV Safe End-to-Pipe Hot leg, loop 8 2RV-01-035 2 RPV Safe End-to-Elbow Cold leg, loop 8 2RV-01-036 2 RPV Safe End-to-Elbow Cold leg, loop A 2RV-01-037 2 RPV Safe End-to-Pipe Hot leg, loop A 3.1 Applicable ASME Code and Addenda The applicable ASME Code edition and addenda is ASME Code Section XI, "Rules for Inservice Inspection of Nuclear Power Plant Components," 2001 edition through the 2003 addenda. In addition, as required by 10 CFR 50.55a, ASME Code Section XI, 2001 edition is used for Appendix VIII, "Performance Demonstration for Ultrasonic Examination Systems." 3.2 Applicable ASME Code Requirements Volumetric examinations of the components contained in Tables 1 and 2 are mandated in ASME Code Section XI Table IW8-2500 Category 8-F, Item 8F5.10. Requirements for UT examinations used to conduct these inspections are contained in ASME Code Section XI Paragraph IWA-2232 and ASME Code Section XI Appendix I Paragraph 1-2220. These paragraphs refer to ASME Code Section XI Appendix VIII including Supplements 2 and 10. ASME Code Case N-695, "Qualification Requirements for Dissimilar Metal Piping Welds,Section XI, Division 1," provides alternative requirements to Appendix VIII, Supplement 10. It is unconditionally approved for use in NRC RG 1.147, Revision 16. Paragraph 3.3(c) of ASME Code Case N-695 states "Examination procedures, equipment, and personnel are qualified for depth-sizing when the RMS error of the flaw depth measurements, as compared to the true flaw depths, do not exceed 0.125 in. (3 mm)."
ASME Code Case N-696, "Qualification Requirements for Appendix VIII Piping Examinations Conducted From the Inside Surface,Section XI, Division 1," provides alternative requirements to Appendix VIII Supplements 2, 3, and 10 examinations conducted from the inside surface. ASME Code Case N-696 is unconditionally approved for use in NRC RG 1.147, Revision 16. Supplement 3 is not applicable to the welds referenced under this relief request. Paragraph 3.3(d) of ASME Code Case N-696 states, "Supplement 2 or Supplement 3 examination procedures, equipment, and personnel are qualified for depth-sizing when the flaw depths estimated by UT examinations, as compared to true depths, do not exceed 0.125 in. (3 mm) RMS, when they are combined with a successful Supplement 10 qualification." 3.4 Impracticality of Compliance The licensee stated that although examination vendors have qualified for detection and length sizing of flaws in accordance with the ASME Code requirements for examinations from the 10 surface, none have met the established RMS error of 0.125-inch for flaw indication depth-sizing. The lSI vendor that will be used in the upcoming inspections at Braidwood has demonstrated the ability to depth-size flaw indications in dissimilar metal welds with a RMS error of 0.189 inch instead of the 0.125-inch RMS error required by Appendix VIII Supplement 10 (ASME Code Case N-695) and an RMS error of 0.245-inch instead of the 0.125 inch RMS error required by the combined Appendix VIII Supplement 2 and 10 qualification (ASME Code Case N-696). The NRC staff also notes that inspection of these welds from the outer diameter (00) of the pipe is often precluded by the geometry of the weld or the presence of obstructions. The NRC staff further notes that, due to the location of these welds, 00 inspections are normally associated with significant dose to the inspector. 3.5 Licensee's Proposed Alternative For welds governed by ASME Code Case N-695, the licensee proposed using an alternative depth-sizing RMS error qualification criterion of 0.189 inches. To compensate for the greater uncertainty in the depth of a measured flaw which will result from this acceptance criterion, the licensee proposed to add to the measured depth of any cracks detected a value equal to the difference between the proposed acceptance criterion (0.189 inches) and the acceptance criterion contained in the code case {0.125 inches}, in this case 0.064 inches. For welds governed by ASME Code Case N-696, the licensee proposed using an alternative depth-sizing RMS error qualification criterion of 0.245 inches. To compensate for the greater uncertainty in the depth of a measured flaw which will result from this acceptance criterion, the licensee proposed to add to the measured depth of any cracks detected a value equal to the difference between the proposed acceptance criterion (0.245 inches) and the acceptance criterion contained in the code case (0.125 inches), in this case 0.120 inches. 3.6 Duration of Relief In its original request the licensee requested relief for the remainder of the third 10-Year lSI Interval for Braidwood, Units 1 and 2, which is currently scheduled to end July 28, 2018, for Unit 1, and on October 16, 2018, for Unit 2. In its letter dated March 12, 2012, the licensee revised its request to include only examinations conducted at Braidwood, Unit 1 prior to the end of refueling outage A1 R16 which is scheduled to begin in April 2012 and examinations conducted at Braidwood, Unit 2 prior to the end of refueling outage A2R16 which is scheduled to begin in the fall of 2012.
-3.7 NRC Staff Evaluation As described above, the licensee has requested relief from the requirements of ASME Code Case N-695, "Qualification Requirements for Dissimilar Metal Piping Welds" and ASME Code Case N-696, "Qualification Requirements for Appendix VIII Piping Examinations Conducted From the Inside Surface,Section XI, Division 1." These ASME Code Cases require that procedures used to inspect welds from the ID be qualified by performance demonstration. The acceptance criterion established by the code case is an RMS error of not greater than 0.125 inches. The licensee has reported and the NRC staff has confirmed that attempts have been made to qualify ID UT inspection procedures since 2002 and that, to date, no inspection vendor has been able to meet the acceptance criteria established by ASME Code cases despite the fact that numerous individuals from several companies have attempted to do so. The NRC staff finds that this repeated inability to qualify ID UT inspection techniques in accordance with ASME Code Cases N-695 and N-696 constitutes an impracticality as described in 10 CFR 50.55a(g)(5)(iii). Use of an alternate approach, such as an OD examination represents a burden to the licensee as the welds under consideration may be inaccessible from the OD or inspectors may be subject to significant radiation dose in performing the inspections. Section 50.55a(g)(6)(i) to 10 CFR permits the Commission to grant relief in cases of impracticality when it determines (in part) that such relief will not endanger life or property. Given the fact that there is always some uncertainty in measuring cracks, and given that relaxation of the ASME Code Cases N-695 and N-696 acceptance criteria may increase the uncertainty of crack measurements, there is a possibility that, for a given crack, the depth measured using a technique which fails to meet the ASME Code Case acceptance criteria will underestimate the depth of the crack to a greater extent than a similar measurement using a procedure which meets the acceptance criteria. This situation could be viewed as a decreasing assurance that structural integrity or leak tightness of the subject welds will be maintained and, therefore, a failure to meet the criteria contained in 10 CFR 50.55a(g)(6)(i). To address the issue of increased potential for undersizing of flaws by inspection procedures which do not meet the ASME Code Cases N-695 and N-696 acceptance criterion, the licensee has proposed to increase the measured depth of the crack by the difference between the actual RMS error obtained for the procedure employed and the code case acceptance criterion. In this case, the RMS error for the procedure employed is 0.189 inches for ASME Code Case N-695 welds and 0.245 inches for ASME Code Case N-696 welds. For example, a crack in an ASME Code Case N-696 weld which measured 0.2 inches deep would be treated as though it was 0.32 inches deep: 0.2" + (0.245" -0.125") =0.32" While this approach provides some additional margin of safety when compared to not adjusting the crack length, the NRC staff did not have sufficient evidence to confirm that the adjustment to the crack depth was sufficient to provide reasonable assurance that cracks measured using procedures with larger RMS error values would not be undersized when compared to those meeting the RMS error criteria. This lack of assurance was primarily based on the difficulty of sizing cracks when the ID surfaces of the weld are rough and the fact that the testing program was not designed to address this issue.
In November of 2011, the NRC staff reviewed the Performance Demonstration Initiative program (administered by Electric Power Research Institute) data used to qualify welds for ASME Code Cases. Based on this review, the NRC staff was able to determine: When using inspection procedures which do not meet the AS ME Code Case N-695 or N-696 acceptance criterion, a significant number of data points are undersized by greater than 0.064 inches (0.189" -0.125") for ASME Code Case N-695 welds and 0.12 inches for ASME Code Case N-696 welds (0.245" 0.125") indicating that reasonable assurance that cracks will not be significantly undersized does not exist when using this approach. When using inspection procedures which do not meet the ASME Code Case N-695 or N-696 acceptance criterion, only a very few data points are undersized by greater than twice the RMS error of the procedure employed (in this case 2 x 0.189" =0.378" for ASME Code Case N-695 welds and 2 x 0.245" =0.490" for ASME Code Case N-696 Welds). The addition of twice the RMS error to the measured depths of any flaws found during the ID inspections provides reasonable assurance that cracks will not be significantly undersized. Due to the qualification procedures used in ASME Code Case N-696, for some service providers, the reported RMS error value may understate the actual RMS error for austenitic stainless steel welds (Supplement 2 welds). Based on these findings the NRC staff issued a request for additional information (RAI) on March 9, 2012 (ADAMS Accession No. ML 120600603). In this RAI the NRC staff proposed two alternative approaches to the issue of crack depth-sizing that could potentially be found to be acceptable. These were: Adjust the measured depth of the crack upward by 0.378 inches (2 x 0.189) for ASME Code Case N-695 welds and 0.490 inches (2 x 0.245 inch) for ASME Code Case N-696 welds; or Provide a commitment such that if cracks are identified for which sizing is required during the scheduled examinations, a flaw evaluation will be submitted to the NRC for review and approval. Specific information to be included in the flaw evaluation, as requested in the RAI, includes: Provide the measured flaw size. Provide the personal qualification RMS error for the analyst who measured the flaw size and justification for adjusting the depth of the measured crack by less than twice the RMS error. Calculate the RMS error for Supplement 2 and 10 examinations separately. Conduct eddy current testing to determine if the flaw is surface breaking. Provide the ID surface profile of the weld, pipe, nozzle, and safe-end (as applicable) in the region at and surrounding the transducer locations used to depth-size the detected flaw.
Describe the suspected flaw-degradation mechanism and the process used to determine the degradation mechanism. In its response, dated March 12, 2012, the licensee agreed that as part of the relief request to provide the information requested in items 1, 4, 5, and 6 above. The licensee did not agree to provide the information requested in item 2 because the licensee believes that disclosure of individual qualification results could compromise the integrity of the qualification. The NRC staff finds that the risks to the integrity of the qualification program outweigh the benefit of obtaining this information. Item 3 is not necessary as the Supplement 2 RMS error was found in a relief request response using the same vendor dated July 7,2005 (ADAMS Accession No. ML051960005). The NRC staff withdraws the requests for items 2 and 3 of the March 9,2012 RAI. Based on the concerted efforts by the industry to meet the acceptance criteria contained in ASME Code Case N-695 and the difficulties associated with other inspection methods, the NRC staff finds that meeting the 0.125 inch acceptance criterion in ASME Code Cases N-695 and N-696 is impractical and represents a burden to the licensee. This relief is subject to the licensee providing the information listed above in items 1, 3, 4, and 5, for the subject indicated flaws. On that basis, and subsequent to NRC staff review and approval of information submitted, the NRC staff finds that this alternative provides reasonable assurance of structural integrity or leak tightness of the subject component and therefore "will not endanger life or property" as required by 10 CFR 50.55a(g)(6)(i). CONCLUSION As set forth above, the NRC staff determines that granting relief pursuant to 10 CFR 50.55a(g)(6)(i) is authorized by law and will not endanger life or property or the common defense and security, and is otherwise in the public interest giving due consideration to the burden upon the licensee that could result if the requirements were imposed on the facility. Accordingly, the NRC staff concludes that the licensee has adequately addressed all of the regulatory requirements set forth in 10 CFR 50.55a. Therefore, the NRC staff grants relief as specified above, i.e., the use of the alternate depth-sizing qualification (0.189 inch for ASME Code Case N-695 welds and 0.245 inch for ASME Code Case N-696 welds), subject to the licensee providing, for NRC staff review and approval prior to the expiration of the relief, the information identified above as items 1, 4, 5, and 6 should a crack requiring depth-sizing be identified. This relief is granted for Braidwood, Units 1 and 2, until prior to startup following refueling Cycle A 1 R 16 and A2R 16, respectively. All other ASME Code,Section XI requirements for which relief was not specifically requested and approved in the subject request for relief remain applicable, including third-party review by the Authorized Nuclear Inservice Inspector. Principal Contributor: S. Cumblidge, NRR Date: April 19, 2012 Therefore, the NRC staff grants the use of the alternate depth-sizing qualification (0.189 inch for ASME Code Case N-695 welds and 0.245 inch for ASME Code Case N-696 welds), subject to the licensee providing, for NRC staff review and approval prior to the expiration of the relief, information identified in the safety evaluation (SE) should a crack requiring depth-sizing be identified. This relief is granted for Braidwood Units 1 and 2, until prior to startup following refueling outage A1 R16 and A2R16, respectively. The NRC staffs SE is enclosed. Please contact the Braidwood Project Manager, Brenda Mozafari at (301) 415-2020 if you have any questions on this action.
Sincerely,IRAJ Jacob I. Zimmerman, Chief Plant Licensing Branch 111-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. 50-456 and 50-457
Enclosure:
Safety Evaluation cc w/encl: Distribution via Listserv DISTRIBUTION: PUBLIC LPL3-2 RIF RidsNrrPMBraidwood Resource RidsNrrLAKGoldstein Resource RidsOgcRp Resource RidsNrrDorlLpl3-2 Resource RidsAcrsAcnw_MailCTR Resource RidsNrrDeEpnb Resource RidsRgn3MailCenter Resource OFFICE LPL 3-2/PM LPL3-2/LA DE/EPNB/BC* LPL3-2/BC LPL NAME MMahoney BMozafari KGoldstein DAiley for TLupold JZimmerman BMozafari DATE 04/18/12 04/19/12 04/18/12 4113112 04/19/12 04/19/12 OFFICIAL RECORD