Similar Documents at Byron |
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Category:Code Relief or Alternative
MONTHYEARML22307A2462022-11-10010 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) 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 ML20245E5062020-09-0909 September 2020 Relief from the Requirements of the ASME Code (COVID-19) 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 ML19269C5342019-09-27027 September 2019 Proposed Alternative to Use Encoded Phased Array Ultrasonic Examination Techniques ML19161A2572019-06-0404 June 2019 BWR Fleet Msv/Srv - Testing Frequency Relief Request NRC Pre-Application Meeting June 4, 2019 ML19035A2942019-02-25025 February 2019 Relief from the Requirements of the ASME Code 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 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 JAFP-18-0052, Proposed Alternative to Utilize Code Case N-8792018-05-30030 May 2018 Proposed Alternative to Utilize Code Case N-879 ML17325B5712018-01-10010 January 2018 Non-Proprietary, Request for Relief for Extension of Examination Interval for Reactor Pressure Vessel Head Penetration Nozzles with Mitigated Alloy 600/82/182 Peened Services (CAC No. MF9854; EPID L-2017-LLR-0042) ML17249A2412017-09-19019 September 2017 Relief from the Requirements of the ASME Code RS-17-083, Inservice Inspection Program Third Interval Relief Requests 13R-12 and 13R-152017-06-29029 June 2017 Inservice Inspection Program Third Interval Relief Requests 13R-12 and 13R-15 ML17170A0132017-06-26026 June 2017 Proposed Alternative to Eliminate Examination of Threads in Reactor Pressure Vessel Flange (CAC Nos. MF8712-MF8729 and MF9548) ML17062A4282017-03-0606 March 2017 Request for Relief from the Requirements of the ASME Code ML17047A0382017-02-24024 February 2017 Request I4R-11 Relief from the Requirements of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code) ML16327A3962016-12-20020 December 2016 Request I4R-01, Relief from the Requirements of the ASME Code ML16327A3182016-12-20020 December 2016 Request I4R-06, Relief from the Requirements of the ASME Code ML16327A4212016-12-19019 December 2016 Requested Relief from the Requirements of the Asme Code (CAC Nos. MF7641 and MF7642) ML16306A3662016-11-0808 November 2016 Request I4R-08, Relief from the Requirements of the 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) ML15173A4362015-06-30030 June 2015 Relief from the Requirements of the ASME Code RS-15-170, Submittal of Relief Requests Associated with the Fourth Inservice Testing Interval2015-06-22022 June 2015 Submittal of Relief Requests Associated with the Fourth Inservice Testing Interval ML12121A6372012-05-10010 May 2012 Request to Use Code Case N-789 ML1129907832012-02-0101 February 2012 Inservice Inspection Relief Request I3R-19: Alternative Requirements for the Repair of Reactor Vessel Head Penetrations 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) ML1113302792011-05-25025 May 2011 Partial Withdrawal of Relief Request for Alternative to ASME Code Repairs ML1105909212011-03-0303 March 2011 Relief Request from ASME Code Case N-729-1 Requirements for Examination of Reactor Vessel Head Penetration Welds 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 ML1002102312010-01-28028 January 2010 Relief Request 13R-16 for Reactor Pressure Vessel Head Penetration Examination Frequency ML1001206082010-01-15015 January 2010 Withdrawal of Relief Request Related to Code Case N-729-1, Table 1 ML0804400772008-02-27027 February 2008 Inservice Inspection Program Second Interval RR 12R-21, TAC MD4097 & MD4098 ML0732303122008-01-15015 January 2008 Relief, Inservice Inspection Program Second Interval Relief Requests I2R-22, I2R-23, I2R-25, and I2R-53, ML0712900112007-05-23023 May 2007 Evaluation of Proposed Alternatives for Inservice Inspection Examination Requirements ML0706800992007-03-29029 March 2007 Requests for Relief, Evaluation of Proposed Alternatives for Inservice Inspection Examination Requirements for Third 10-Year Interval Inservice Inspection Program Plan, MD3661, MD3662, and MD3663 ML0635303222007-03-19019 March 2007 Relief, Evaluation of Relief Request I3R-08 Pertaining to Structural Weld Overlays ML0705204802007-03-0909 March 2007 Relief, Evaluation of Relief Request 13R-06 for Control Rod Drive Canopy Seal Welds ML0625101692007-01-29029 January 2007 Evaluation of Relief Requests 13R-08 Pertaining to Structural Weld Overlays ML0632600742007-01-16016 January 2007 Evaluation of Inservice Inspection Program Relief Requests 13R-07 and 12R-46 Pertaining to Essential Service Water Buried Piping ML0622303512006-09-0707 September 2006 Relief Request from the ASME OM Code for the Third Inservice Test Interval 2022-09-20
[Table view] Category:Letter
MONTHYEARIR 05000454/20230042024-02-0202 February 2024 Integrated Inspection Report 05000454/2023004 and 05000455/2023004 ML24022A2722024-01-23023 January 2024 Request for Information for an NRC Post-Approval Site Inspection for License Renewal Inspection Report 05000454/2024011 ML24018A0362024-01-17017 January 2024 Paragon Energy Solutions, Defect with Detroit Diesel/Mtu Fuel Injectors P/N R5229660 Cat Id 0001390618 ML23320A1762023-12-13013 December 2023 Exemption from Select Requirements of 10 CFR Part 73 (EPID L-2023-LLE-0027 (Security Notifications, Reports, and Recordkeeping and Suspicious Activity Reporting)) ML23277A0032023-12-11011 December 2023 Issuance of Amendments Regarding Adoption of TSTF-370 ML23318A5102023-12-0101 December 2023 Relief from the Requirements of the ASME Code IR 05000454/20233012023-11-27027 November 2023 NRC Initial License Examination Report 05000454/2023301 and 05000455/2023301 BYRON 2023-0065, Unit 2 - 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 Guideline2023-11-17017 November 2023 Unit 2 - 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 BYRON 2023-0063, Day Inservice Inspection Report for Interval 4, Period 3, (B2R24)2023-11-16016 November 2023 Day Inservice Inspection Report for Interval 4, Period 3, (B2R24) IR 05000454/20230032023-11-13013 November 2023 Integrated Inspection Report 05000454/2023003 and 05000455/2023003 and Exercise of Enforcement Discretion RS-23-117, Supplemental Information Letter for Part 73 Exemption Request - Responses to Request for Confirmatory Information2023-11-10010 November 2023 Supplemental Information Letter for Part 73 Exemption Request - Responses to Request for Confirmatory Information ML23317A1192023-11-10010 November 2023 Constellation Energy Generation, LLC - 2023 Annual Report - Guarantees of Payment of Deferred Premiums ML23313A0852023-11-0909 November 2023 Submittal of 2023-301 Byron Initial License Examination Post-Examination Comments 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 ML23297A2352023-10-26026 October 2023 Information Request for the Cyber Security Baseline Inspection, Notification to Perform Inspection RS-23-100, Request for Exemption from Enhanced Weapons, Firearms Background Checks, and Security Event Notifications Implementation2023-10-13013 October 2023 Request for Exemption from Enhanced Weapons, Firearms Background Checks, and Security Event Notifications Implementation RS-23-097, Constellation Energy Generation, LLC, Advisement of Leadership Changes and Submittal of Updated Standard Practice Procedures Plans2023-10-12012 October 2023 Constellation Energy Generation, LLC, Advisement of Leadership Changes and Submittal of Updated Standard Practice Procedures Plans RS-23-108, Proposed Alternative for Examinations of Examination Categories B-B, B-D, and C-A Steam Generator Pressure Retaining Welds and Full Penetration Welded Nozzles2023-10-11011 October 2023 Proposed Alternative for Examinations of Examination Categories B-B, B-D, and C-A Steam Generator Pressure Retaining Welds and Full Penetration Welded Nozzles RS-23-105, Proposed Alternative for Examinations of Examination Category C-B Steam Generator Nozzle-to-Shell Welds and Nozzle Inside Radius Sections2023-10-10010 October 2023 Proposed Alternative for Examinations of Examination Category C-B Steam Generator Nozzle-to-Shell Welds and Nozzle Inside Radius Sections ML23278A0272023-10-0505 October 2023 Operator Licensing Examination Approval - Byron Station, October 2023 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. ML23251A1622023-09-29029 September 2023 Steam Generator Tube Inspection Reports to Reflect TSTF-577 Reporting Requirements RS-23-094, Relief Request I4R-24, Alternative for Post-Peening Reexamination Frequency for Reactor Pressure Vessel Head Penetration Nozzle Number 752023-09-29029 September 2023 Relief Request I4R-24, Alternative for Post-Peening Reexamination Frequency for Reactor Pressure Vessel Head Penetration Nozzle Number 75 RS-23-091, Relief Request I4R-25, Alternative Requirements for Reactor Pressure Vessel Inservice Inspection Intervals2023-09-26026 September 2023 Relief Request I4R-25, Alternative Requirements for Reactor Pressure Vessel Inservice Inspection Intervals 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) ML23242A3282023-09-0101 September 2023 Amendment No. 233 Correction IR 05000454/20235012023-08-31031 August 2023 Emergency Preparedness Biennial Exercise Inspection Report 05000454/2023501 and 05000455/2023501 IR 05000454/20230052023-08-30030 August 2023 Updated Inspection Plan for Byron Station (Report 05000454/2023005 and 05000455/2023005) IR 05000454/20230022023-08-0303 August 2023 Integrated Inspection Report 05000454/2023002 and 05000455/2023002 ML23209A7242023-07-31031 July 2023 Request for Information on the NRC Quadrennial Comprehensive Engineering Team Inspection: Inspection Report 05000454/2024010 and 05000455/2024010 ML23192A0362023-07-25025 July 2023 Review of the Spring 2022 Steam Generator Tube Inspection Report ML23122A3022023-07-20020 July 2023 Issuance of Amendments Technical Specifications 2.1.1 and 4.2.1 to Allow a Previously Irradiated Accident Tolerant Fuel Lead Test Assembly to Be Further Irradiated in Unit No. 2 IR 05000454/20234022023-07-18018 July 2023 Baseline Security Inspection Document; 05000454/2023402; 05000455/2023402 ML23198A0372023-07-17017 July 2023 Information Request to Support Upcoming Problem Identification and Resolution (PIR) Inspection at Byron Nuclear Plant 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 ML23172A1172023-06-22022 June 2023 Notification of NRC Fire Protection Team Inspection Request for Information 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 ML23159A1302023-06-0808 June 2023 Day Inservice Inspection Report for Interval 4, Period 3, (B1R25) ML23159A1542023-06-0808 June 2023 2022 Regulatory Commitment Change Summary Report 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 ML23157A2832023-06-0606 June 2023 Notification of NRC Baseline Inspection and Request for Information BYRON 2023-0029, Response to Request for Additional Information Regarding Steam Generator Tube Inspection Reports to Reflect TSTF-577 Reporting Requirements2023-06-0101 June 2023 Response to Request for Additional Information Regarding Steam Generator Tube Inspection Reports to Reflect TSTF-577 Reporting Requirements 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 ML23003A7882023-05-11011 May 2023 Report for December 12-16, 2022, Regulatory Audit Regarding Reinsertion of a High Burnup Accident Tolerant Fuel Lead Test Assembly 2024-02-02
[Table view] Category:Safety Evaluation
MONTHYEARML23277A0032023-12-11011 December 2023 Issuance of Amendments Regarding Adoption of TSTF-370 ML23318A5102023-12-0101 December 2023 Relief from the Requirements of the ASME Code ML23241A9092023-09-19019 September 2023 Enclosure 2 - Non-Proprietary - Review of License Renewal Commitment Number 10 Safety Evaluation ML23122A3022023-07-20020 July 2023 Issuance of Amendments Technical Specifications 2.1.1 and 4.2.1 to Allow a Previously Irradiated Accident Tolerant Fuel Lead Test Assembly to Be Further Irradiated in Unit No. 2 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 ML22305A6992022-12-28028 December 2022 Issuance of Amendment Nos. 231 and 231 Adoption of TSTF-577, Revised Frequencies for Steam Generator Tube Inspections, Revision 1 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 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 ML21076A3712021-08-17017 August 2021 Approval of Certified Fuel Handler Training and Retraining Program 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) 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 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 ML21013A0052021-02-0404 February 2021 Issuance of Amendments to Adopt Technical Specifications Task Traveler TSTF-568, Revise Applicability of BWR/4 TS 3.6.2.5 and TS 3.6.3.2 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 ML20246G8652020-09-21021 September 2020 Issuance of Amendment Nos. 222 and 222 One-Time Extension of Unit No. 2 Steam Generator Inspections (COVID-19) 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 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 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 ML19190A0812019-08-28028 August 2019 Issuance of Amendments Regarding Limiting Condition of Operation for Inoperability of Snubbers ML19192A2442019-07-18018 July 2019 Proposed Alternative to Use ASME Code Cases N-878 and N-880 ML19038A0172019-04-0303 April 2019 Issuance of Amendment No. 207 Regarding Use of Accident Tolerant Fuel Lead Test Assemblies ML19036A5862019-03-21021 March 2019 Issuance of Amendments to Revise the Emergency Response Organization Staffing Requirements ML19035A2942019-02-25025 February 2019 Relief from the Requirements of the ASME Code ML18302A2272018-12-12012 December 2018 Issuance of Amendments No. 200 and 205 Regarding Axial Flux Difference Technical Specifications ML18264A0922018-10-22022 October 2018 Issuance of Amendment Nos. 198, 198, 204, and 204, Respectively, Regarding Adoption of Title 10 of the Code of Federal Regulations Section 50.69 CAC Nos. MG0201, MG0202, MG0203, and MG0204; EPID-L 2017-LLA-0285) ML18206A2822018-08-0202 August 2018 Issuance of Amendments to Relocate the Staff Qualification Requirements ML18065A5292018-06-27027 June 2018 Issuance of Amendments Regarding Clarification of Rod Position Requirements (CAC Nos. MF9951-9954; EPID L-2017-LLA-0249) 2023-09-19
[Table view] |
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March 9, 2007 Mr. Christopher M. Crane President and Chief Nuclear Officer Exelon Generation Company, LLC 4300 Winfield Road Warrenville, IL 60555
SUBJECT:
BYRON STATION, UNIT NOS. 1 AND 2 - EVALUATION OF RELIEF REQUEST I3R-06 FOR CONTROL ROD DRIVE CANOPY SEAL WELDS (TAC NOS.
MD3863 AND MD3864)
Dear Mr. Crane:
By letter dated February 14, 2006, Exelon Generating Company, LLC (the licensee) requested relief from the requirements of the 2001 Edition through the 2003 Addenda of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code),Section XI, Subarticle IWA-4000, regarding the contingency repair or replacement of control rod drive mechanism canopy seal welds at Byron Station, Unit Nos. 1 and 2. The licensee proposed to use the ASME Code,Section XI, Code Case N-504-2, Alternative Rules for Repair of Classes 1, 2, and 3 Austenitic Stainless Steel Piping, for weld repair and to perform an enhanced visual examination and a pressure test.
The Nuclear Regulatory Commission (NRC) staff concludes that complying with the ASME Code-specified repair method and the surface examination of the canopy seal welds would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety. The NRC staff concludes further that the proposed alternative is acceptable because the repair or replacement method and associated post-weld examination and pressure testing will provide reasonable assurance of the welds structural integrity. Therefore, pursuant to Section 50.55a(a)(3)(ii) of Title 10 of the Code of Federal Regulations (10 CFR), the licensees proposed alternative described in relief request I3R-06, Revision 0, is authorized for the third 10-year inservice inspection interval.
C. Crane The licensees submittal dated February 14, 2006, also included Relief Requests I3R-02 through I3R-05. Relief Request I3R-02 will be evaluated in a separate letter and I3R-03 through I3R-05 were withdrawn by letter dated January 24, 2007.
Sincerely,
/RA/
Russell A. Gibbs, Chief Plant Licensing Branch III-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. STN 50-454 and STN 50-455
Enclosure:
Safety Evaluation cc w/encl: See next page
C. Crane The licensees submittal dated February 14, 2006, also included Relief Requests I3R-02 through I3R-05. Relief Request I3R-02 will be evaluated in a separate letter and I3R-03 through I3R-05 were withdrawn by letter dated January 24, 2007.
Sincerely,
/RA/
Russell A. Gibbs, Chief Plant Licensing Branch III-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. STN 50-454 and STN 50-455
Enclosure:
Safety Evaluation cc w/encl: See next page DISTRIBUTION:
PUBLIC RidsNrrDciCpnb LPL3-2 R/F RidsOgcRp RidsNrrDorlLpl3-2 RidsAcrsAcnwMailCenter RidsNrrPMCGratton JTsao, NRR RidsNrrLAEWhitt RidsRgn3MailCenter RidsNrrDorlDpr RidsSecyMailCenter TBloomer, EDO Accession Number: ML070520480 NRR-028 ** See SE dated *no legal objection OFFICE LPL3-2/PM LPL3-2/LA DCI/CPNB/BC OGC
- LPL3-2/BC NAME CGratton:mw EWhitt TChan** SHamrick RGibbs DATE 3/9/07 3/9/07 2/06/07 3/6/07 3/9/07 OFFICIAL RECORD COPY
Byron Station, Units 1 and 2 cc:
Regional Administrator, Region III Attorney General U.S. Nuclear Regulatory Commission 500 S. Second Street 2443 Warrenville Road, Suite 210 Springfield, IL 62701 Lisle, IL 60532-4351 Plant Manager - Byron Station Illinois Emergency Management Agency Exelon Generation Company, LLC Division of Disaster Assistance & 4450 N. German Church Road Preparedness Byron, IL 61010-9794 110 East Adams Street Springfield, IL 62701-1109 Site Vice President - Byron Station Exelon Generation Company, LLC Document Control Desk - Licensing 4450 N. German Church Road Exelon Generation Company, LLC Byron, IL 61010-9794 4300 Winfield Road Warrenville, IL 60555 Senior Vice President - Operations Support Exelon Generation Company, LLC Mr. Dwain W. Alexander, Project Manager 4300 Winfield Road Westinghouse Electric Corporation Warrenville, IL 60555 P.O. Box 355 Pittsburgh, PA 15230 Chairman Will County Board of Supervisors Howard A. Learner Will County Board Courthouse Environmental Law and Policy Joliet, IL 60434 Center of the Midwest 35 East Wacker Drive Director - Licensing and Regulatory Affairs Suite 1300 Exelon Generation Company, LLC Chicago, IL 60601-2110 4300 Winfield Road Warrenville, IL 60555 U.S. Nuclear Regulatory Commission Byron Resident Inspectors Office Manager Regulatory Assurance - Byron 4448 North German Church Road Exelon Generation Company, LLC Byron, IL 61010-9750 4450 N. German Church Road Byron, IL 61010-9794 Ms. Lorraine Creek RR 1, Box 182 Associate General Counsel Manteno, IL 60950 Exelon Generation Company, LLC 4300 Winfield Road Chairman, Ogle County Board Warrenville, IL 60555 P.O. Box 357 Oregon, IL 61061 Vice President - Regulatory Affairs Exelon Generation Company, LLC Mrs. Phillip B. Johnson 4300 Winfield Road 1907 Stratford Lane Warrenville, IL 60555 Rockford, IL 61107
Byron Station, Units 1 and 2 cc:
Manager Licensing - Braidwood, Byron and LaSalle Exelon Generation Company, LLC 4300 Winfield Road Warrenville, IL 60555 Senior Vice President - Midwest Operations Exelon Generation Company, LLC 4300 Winfield Road Warrenville, IL 60555 Mr. Barry Quigley 3512 Louisiana Rockford, IL 61108
SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO RELIEF REQUEST NO. I3R-06, REV. 0 THIRD 10-YEAR INTERVAL INSERVICE INSPECTION PROGRAM PLAN EXELON GENERATION COMPANY, LLC BYRON STATION, UNIT NOS. 1 AND 2 DOCKET NOS. STN 50-454 AND STN 50-455
1.0 INTRODUCTION
By letter dated February 14, 2006 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML063530333), Exelon Generation Company, LLC (the licensee) requested relief from the requirements of the 2001 Edition through the 2003 Addenda of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code),
Section XI, Subarticle IWA-4000 regarding the contingency repair or replacement of control rod drive mechanism (CRDM) canopy seal welds at Byron Station, Unit Nos. 1 and 2 (Byron). The licensee submitted Relief Request I3R-06 for the third 10-year inservice inspection interval which began on January 16, 2006, and will end on January 15, 2016.
The licensee proposed to repair leaking CRDM canopy seal welds using the guidelines of ASME Code Case N-504-2, Alternative Rules for Repair of Classes 1, 2, and 3 Austenitic Stainless Steel Piping, which establishes acceptability of a repair by increasing the weld thickness. The licensee also proposed an enhanced visual examination and pressure test in lieu of the ASME Code-required surface examination for final acceptance of the repaired weld.
The licensee contended that the ASME Code-required repair and the surface examination of the seal welds would expose personnel to a high radiation dose and, therefore, would result in a hardship or unusual difficulty without a compensating increase in the level of quality and safety.
By letter dated September 16, 2003 (ADAMS Accession No. ML032410022), the Nuclear Regulatory Commission (NRC) staff approved an identical relief request for the repair of the CRDM canopy seal welds for the second inspection interval at Byron.
2.0 REGULATORY EVALUATION
The inservice inspection of ASME Code Class 1, Class 2, and Class 3 components is to be performed in accordance with Section XI of the ASME Code and applicable edition and addenda as required by Section 50.55a(g) of Title 10 of the Code of Federal Regulations (10 CFR), except where specific written relief has been granted by the Commission pursuant to 10 CFR 50.55a(g)(6)(i). Section 50.55a(a)(3) states in part that alternatives to the requirements of paragraph (g) may be used, when authorized by the NRC, if the licensee demonstrates that:
(i) the proposed alternatives would provide an acceptable level of quality and safety, or (ii)
compliance with the specified requirements would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety.
Pursuant to 10 CFR 50.55a(g)(4), ASME Code Class 1, 2, and 3 components (including supports) shall meet the requirements, except the design and access provisions and the pre-service examination requirements, set forth in the ASME Code,Section XI, "Rules for Inservice Inspection (ISI) of Nuclear Power Plant Components," to the extent practical within the limitations of design, geometry, and materials of construction of the components. The regulations require that inservice examination of components and system pressure tests conducted during the first 10-year interval and subsequent intervals comply with the requirements in the latest edition and addenda of Section XI of the ASME Code incorporated by reference in 10 CFR 50.55a(b), 12 months prior to the start of the 10-year interval, subject to the limitations and modifications listed therein. The Code of record for the third 10-year ISI interval at Byron is the 2001 Edition through the 2003 Addenda of Section XI of the ASME Code.
The licensee submitted Relief Request I3R-06 pursuant to 10 CFR 50.55a(a)(3)(i), which requires that the proposed alternative provide an acceptable level of quality and safety. In addition to the requirements of 10 CFR 50.55a(a)(3)(i), the NRC staff evaluated this submittal pursuant to 10 CFR 50.55a(a)(3)(ii), which requires the licensee to demonstrate that compliance with the ASME Code requirements would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety.
3.0 TECHNICAL EVALUATION
3.1 Relief Request I3R-06, Revision 0 3.1.1 ASME Code Component Affected Reactor CRDM canopy seal welds - Class 1 appurtenance to the reactor vessel.
3.1.2 Applicable Code Edition and Addenda The third 10-year ISI program is based on the ASME Code,Section XI, 2001 Edition through the 2003 Addenda. The CRDM assemblies were designed and fabricated to the ASME Code,Section III, 1974 Edition through the Summer 1974 Addenda.
3.1.3 Applicable Code Requirements Subarticle IWA-4000 of the ASME Code,Section XI, requires that repairs be performed in accordance with the owners original construction Code of the component or system, or later editions and addenda of the Code. The canopy seal weld is described in the ASME Code,Section III, and a repair to this weld would require the following activities:
- a. Excavation of the rejectable indications,
- b. A surface examination of the excavated areas,
- c. Re-welding and restoration to the original configuration and materials, and
- d. Final surface examination
3.1.4 Reason for the Request The licensee identified the principal issues of this relief request as the excavation of the existing CRDM canopy seal weld, the accompanying radiation dose received during the excavation and examination activities, and the weld material used for the repair or replacement.
The licensee stated that due to the nature of the flaw in the subject canopy seal weld, the excavation of the leaking portion of the weld would result in a cavity that would extend completely through wall. A liquid penetrant examination (PT) of this cavity is required to verify the removal of the rejectable flaw or to verify that the flaw is removed or reduced to an acceptable size. This PT examination would deposit the penetrant materials onto the inner surfaces of the subject weld. The penetrant material would not be readily removable prior to re-welding due to the inaccessibility of the inside surface of the weld. The remnant penetrant material in the weld would introduce contaminants to the new weld metal and reduce the quality of the repair weld. In addition, the configuration of the canopy assembly would prevent the establishment and maintenance of an adequate back-purge during the welding process and would further reduce the quality of the repaired weld.
The licensee stated that the CRDM canopy seal weld is located above the reactor vessel closure head, which is highly congested and subject to high ASME Code requirements would be contrary to the intent of the as low as reasonably achievable (ALARA) radiological controls program. Most of the repair activities would be performed remotely using robotic equipment.
This will reduce the radiation exposure to personnel involved in the welding process. However, the required excavation and PT examinations would necessitate hands-on access to the canopy weld and are estimated to result in a total occupational ASME Code-required excavation and examinations are activities that would not be needed if the relief is granted from these requirements and thus, represent the estimated occupational radiation dose savings.
Subarticle IWA-4200 of the ASME Code,Section XI, requires that the repair material conform to the original design specification or the ASME Code,Section III. In the proposed alternative, the replacement weld material would have the same resistance to stress corrosion cracking as the original weld material. Use of the original material does not guarantee that the repaired component will continue to maintain leakage integrity throughout the intended life of the item.
The licensee will use applicable portions of ASME Code Case N-504-2 as guidance for repair by weld overlay to provide a new leakage barrier. In lieu of performing PT examinations of CRDM seal weld repairs or replacement, the licensee will perform a 5X or better magnification visual examination after repair welding is completed.
In addition, the licensee will use Alloy 52 nickel-based weld material rather than austenitic stainless steel as required by ASME Code Case N-504-2. The licensee selected Alloy 52 nickel-based weld repair material based on its resistance to stress corrosion cracking.
Consequently, the ferrite requirements of Code Case N-504-2 do not apply. The suitability of the replacement material has been evaluated and is determined to be compatible with the existing component and will provide a leakage barrier for the remainder of the intended life of the CRDM.
The licensee stated that industry experience with failure analyses performed on leaking canopy seal welds removed from service at other plants has attributed the majority of the cases to transgranular stress corrosion cracking (SCC). The size of the opening where the leakage
occurs has been extremely small, normally a few thousandths of an inch. The crack orientations vary, but often radiate outward such that a pinhole appears on the surface, as opposed to a long crack. The SCC results from exposure of a susceptible material to residual stress, which is often concentrated by weld discontinuities, and to a corrosive environment. A corrosive environment can form with water being trapped in the cavity behind the seal weld that is mixed with air initially in the cavity, resulting in a higher oxygen content than is in the bulk primary coolant.
3.1.5 Proposed Alternative and Basis for Use Following the guidance of ASME Code Case N-504-2, the licensee will not remove the flaw(s) in the CRDM canopy seal weld, but will deposit Alloy 52 weld material on the surface of the existing seal weld. The licensee will also perform an analysis of the repaired weldment using Paragraph (g) of the code case as guidance to assure that the remnant flaw will not propagate unacceptably. The proposed canopy seal weld is not a structural weld, nor a pressure-retaining weld, but provides a seal to prevent reactor coolant leakage if the mechanical joint leaks. The weld buildup is considered a repair in accordance with Subarticle IWA-4110 of the ASME Code,Section XI. Applicability of the original code of construction or design specification is mandated because the weld is performed on an appurtenance to a pressure-retaining component. The alternative CRDM canopy seal weld repair uses a gas tungsten arc welding (GTAW) process controlled remotely.
The licensee will examine the repaired weld visually using methods and personnel qualified to the standards of ASME VT-1 requirements. The visual examination will be performed using the welding equipment video camera with 5X or better magnification within several inches of the weld. The visual examination is qualified to ensure the identification of flaws so that an adequate margin of safety is maintained. The examination technique will be demonstrated to resolve a 0.001-inch thick wire against the surface of the weld. The repaired weld will be examined for quality of workmanship and discontinuities will be evaluated and dispositioned to ensure the adequacy of the new leakage barrier.
The automated GTAW weld repair and alternate VT-1 examination methods result in significantly lower radiation exposure because the equipment is remotely operated after setup.
The licensee will perform a post-maintenance pressure test at nominal temperature and pressure to determine potential leakage with a VT-2 visual examination.
The licensee will document the repair or replacement activities, using the process described in this relief request, on the required NIS-2 forms. This relief request will be identified on the NIS-2 forms in lieu of an adopted or invoked ASME Code case. The repair documents will be reviewed by the Authorized Nuclear Inspector, and maintained in accordance with the requirements for archiving permanent plant records.
3.1.6 Duration of Proposed Alternative Relief is requested for the third inspection interval which began on January 16, 2006 and will end on January 15, 2016.
3.2 NRC Staff Evaluation In lieu of ASME Code-required weld repair, the licensee proposed to perform the repair of leaking CRDM canopy seal welds using the applicable provisions of ASME Code Case N-504-2, which establishes acceptability of a repair by increasing the weld thickness. The code case allows deposition of one or more layers of the weld overlay to seal unacceptable indications in the area to be repaired without excavation. The multiple layers of weld metal provide a redundant CRDM nozzle-to-canopy seal. The code case further requires a stress analysis of the repaired weldment to assure that the existing flaw will not propagate unacceptably for the design life of the repair. The analysis will consider potential flaw growth due to fatigue and stress corrosion cracking which is most likely the mechanism that initiated the flaw. This analysis will establish a critical flaw size that can be used as a benchmark to qualify the VT-1 examination method to ensure the capability of detecting sufficiently small flaw size to assure that an adequate margin of safety is maintained. The seal weld itself is neither a structural weld nor a pressure-retaining weld (i.e., the weld does not support any loads). The NRC staff finds the proposed alternative repair method acceptable because the proposed weld repair is applicable for the purpose of preventing leakage and not for the purpose of supporting loads.
The licensee proposed to use Alloy 52 nickel-base weld repair material in place of austenitic stainless steel as required by Code Case N-504-2. The NRC staff finds that this material is acceptable for use due to its resistance to stress corrosion cracking as documented in the Electric Power Research Institute Report, Materials Reliability Program: Resistance to Primary Water Stress Corrosion Cracking of Alloy 690, 52, and 152 in Pressurized Water Reactors (MRP-111), 1009801, (ADAMS Accession No. ML0416805460).
In lieu of surface examination, the licensee would conduct a remote visual examination using a video camera with 5X magnification and 0.001 inch resolution within several inches of the weld.
The visual resolution of this video camera system has greater capability than that of the ASME Code-required direct VT-1 visual examination of resolving a wire segment as narrow as 1/32-inch black line on a 18 percent neutral gray card. Therefore, the resolution and consistency of the licensees enhanced visual examination technique are much greater than that provided by an ASME Code VT-1 examination. Based on the capability of the remote visual examination system to resolve flaws of a size 0.001 inch in width, reasonable assurance of detecting surface flaws is provided. Therefore, the NRC staff has determined that the remote visual examination is acceptable in lieu of the ASME Code-required surface examination.
In addition, the adequacy of the seal is verified with a routine system leakage test that is performed at normal operating temperature and pressure, and held at such conditions for a ASME Code-required soak time prior to returning to the system to service. The licensee will perform a VT-2 examination during the pressure test to observe any potential leakage.
As the licensee stated above, the radiological dose associated with strict compliance with the ASME Code-required repair would be high because the canopy seal weld is located in a high radiation area. Plant personnel will receive the radiological dose during the manual excavation of the flaws, PT of the excavated areas of the weld, and final PT of new weld. According to the licensee, the estimated total occupational radiation dose is 1.688 person-Rem per CRDM
canopy seal weld. The NRC staff agrees with the licensee that this radiological dose is significant and that compliance with the specified ASME Code requirements would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety.
The NRC staff finds that proposed alternative will result in reducing radiation dose exposure and that the proposed alternative repair will provide acceptable leakage integrity for the CRDM canopy seal weld while maintaining reasonable assurance of structural integrity.
4.0 CONCLUSION
Based on the above evaluation, the NRC staff concludes that the ASME Code-required repair or replacement, and surface examination of the CRDM canopy seal welds would result in a hardship or unusual difficulty without a compensating increase in the level of quality and safety.
The NRC staff concludes further that the proposed alternative described in Section 3.1.5 above is acceptable for the repair or replacement of the CRDM canopy seal welds because the repair or replacement method and associated post-weld examination and pressure testing will provide reasonable assurance of the weld integrity. Therefore, pursuant to 10 CFR 50.55a(a)(3)(ii), the NRC staff authorizes use of the proposed alternative as stated in Relief Request I3R-06, Rev. 0 for the third inservice inspection intervals at Byron.
All other ASME Code,Section XI requirements for which relief was not specifically requested and approved in this relief request remain applicable, including third party review by the Authorized Nuclear Inservice Inspector.
Principle Contributor: J. Tsao Date: March 9, 2007