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Category:Code Relief or Alternative
MONTHYEARML20302A0802020-10-30030 October 2020 Request to Use a Provision of a Later Edition of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code, Section XI ML19291A0182019-11-12012 November 2019 Request for Relief I4R-07, Utilize Code Case N-513-4 - Evaluation Criteria for Temporary Acceptance of Flaws in Moderate Energy Class 2 or 3 Piping of American Society of Mechanical Engineers Boiler and Pressure Code Section XI ML18334A0132018-12-12012 December 2018 Request for Relief from the Requirements of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code, Code Case N-666-1 ML18283A0492018-10-18018 October 2018 Request for Relief I4R-6 from ASME Code Visual Examination Requirements for Reactor Vessel Head Penetration Nozzle Weld Specified by Code Case N-729-4 ML17202E9192017-08-0202 August 2017 Relief Request 14R-05 from Certain Pressure Test Requirements of the ASME Code for Reactor Pressure Vessel Leak-Off Lines for the Fourth 10 Year Inservice Inspection Interval ML16320A0322016-11-30030 November 2016 Request for Relief I4R-, Alternative Risk-Informed Methodology in Selecting Class 1 and 2 Piping Welds, for the Fourth 10-Year Inservice Inspection Interval ML16231A1972016-08-23023 August 2016 Relief Request I4R-02, from Requirements of ASME Code Table IWF-2500-1 VT-3 Examination for Class 1 Supports, for the Fourth 10-Year Inservice Inspection Interval ML15226A3542015-08-21021 August 2015 Relief Request, Alternative to ASME Code Case N-579, Use of Nonstandard Nuts, Class 1, 2, and 3, Mc, CS Components and Supports Construction Section III, Division1, Excess Letdown Heat Exchanger ML15216A2292015-08-10010 August 2015 Request for Relief Nos. 4VR-01 and 4GR-01 Related to ASME OM Code Requirements for Set Pressure Measurement Accuracy for Relief Valves and Frequency Specification, Fourth 10-Year Inservice Testing Program ML15190A2022015-07-13013 July 2015 Relief Request 13R-12, Extension of the Third Inservice Inspection Program Interval to Perform Reactor Vessel Stud Hole Ligament Examinations ML15134A0022015-05-15015 May 2015 Relief Request Nos. 4PR-01 and 4PR-02 and Withdrawal of 4VR-02, Alternative to Requirements of ASME Code Case OMN-21 and OM Code ISTB-3510(b)(1) for Pumps, Fourth 10-Year Inservice Testing Program ML15040A0202015-02-12012 February 2015 Relief Request I3R-10, Alternative from Pressure Test Requirements of ASME Code, Section XI, IWC-5220, Third 10-Year Inservice Inspection Interval ML15028A1762015-02-0909 February 2015 Correction to Relief Requests I3R-08, Reactor Pressure Vessel (RPV) Interior Attachments and I3R-09, RPV Pressure-Retaining Welds, Exam Interval Extensions, Third 10-year Inservice Inspection Interval (TAC MF3321 & MF3322) ML15023A2202015-01-28028 January 2015 Relief Request 13R-11, Alternative from Pressure Test Requirements of ASME Code Section XI IWC-5220 for the Third 10-Year Inservice Inspection Interval ML14321A8642014-12-10010 December 2014 Relief Requests 13R-08, Reactor Pressure Vessel (RPV) Interior Attachments and 13R-09, RPV Pressure-Retaining Welds, Exam Interval Extensions, Third 10-year Inservice Inspection Interval ML12353A2412013-01-0404 January 2013 Relief Request I3R-07 from ASME Code Case N-729-1 for Examination of Reactor Vessel Head Penetration Welds, for Remainder of Third 10-Year Inservice Inspection Interval ET 12-0010, 10CFR50.55a Request Number I3R-07, Relief from ASME Code Case N-729-1 Requirements for Examination of Reactor Vessel Head Penetration Welds2012-07-0202 July 2012 10CFR50.55a Request Number I3R-07, Relief from ASME Code Case N-729-1 Requirements for Examination of Reactor Vessel Head Penetration Welds ML0717001842007-07-19019 July 2007 Correction to Authorization of Relief Request 13R-05 - Alternatives to Structural Weld Overlay Requirements ML0706705142007-04-0303 April 2007 Authorization of Relief Request 13R-05, Alternative to Structural Weld Overlay Requirements ML0702605382007-02-21021 February 2007 Relief Request, Third 10-Year Interval Inservice Inspection Program Relief Request I3R-01 ML0634700822006-12-27027 December 2006 Request for Relief I2R-37 and I2R-38 for the Second 10-year Interval Inservice Inspection, MD0291 and MD0292 ML0630705902006-11-20020 November 2006 Relief Request I2R-34 for the Second 10-year Interval Inservice Inspection ML0630706022006-11-20020 November 2006 Relief Request I2R-36 for the Second 10-year Interval Inservice Inspection ML0630705872006-11-20020 November 2006 Relief Request I2R-35 for the Second 10-Year Interval Inservice Inspection ET 06-0029, CFR 50.55a Request, Use of Alternative Ultrasonic Examination Method in Lieu of the Radiography Required by ASME Section III, Subarticle NC-52222006-09-0101 September 2006 CFR 50.55a Request, Use of Alternative Ultrasonic Examination Method in Lieu of the Radiography Required by ASME Section III, Subarticle NC-5222 ML0619304072006-08-0404 August 2006 Relief Request 3PR-04 for the Third 10-Year Inservice Testing Program ET 06-0027, Response to Request for Additional Information Regarding 10 CFR 50.55a Requests I2R-34, I2R-35, and I2R-362006-07-12012 July 2006 Response to Request for Additional Information Regarding 10 CFR 50.55a Requests I2R-34, I2R-35, and I2R-36 ML0613200612006-06-16016 June 2006 Relief, ASME Code Requirements Hardship, TAC MD0299 ML0613901352006-06-0202 June 2006 Relief, Relief Request I3R-03 for the Third 10-Year Interval Inservice Inspection and Examination of Snubbers ML0611403722006-05-10010 May 2006 Third 10-Year Interval Inservice Inspection Program Relief Request I3R-02 ML0606900442006-04-0404 April 2006 Relief, ASME Code Inspection Requirements for Section XI, Class 1, Table IWB-2500-1, Examination Category B-D, Item No. B3.90, Nozzles-to-Vessel Welds for the Second 10-Year Interval ML0605504472006-03-21021 March 2006 Relief Requests for the Third 10-year Pump and Valve Inservice Testing Program ML0534101282005-11-29029 November 2005 10 CFR 50.55a Request I1R-51 Regarding ASME Section XI Requirements for Examination Coverage of the Reactor Pressure Vessel Nozzle-to-Vessel Welds, and Correction to Relief Requests I2R-03 and I2R-21 for Wcnoc'S Second Inservice Inspection ET 05-0027, CFR 50.55a Request Number I2R-33 Regarding ASME Section XI Requirements for Examination Coverage of the Reactor Pressure Vessel Nozzle-to-Vessel Welds2005-11-22022 November 2005 CFR 50.55a Request Number I2R-33 Regarding ASME Section XI Requirements for Examination Coverage of the Reactor Pressure Vessel Nozzle-to-Vessel Welds ML0513803962005-05-17017 May 2005 Gs, Relief, Relief Requests 12R-29 Through 12R-32 Pertaining to Implementation of ASME Code, Section XI Requirements for Examination of Welds ML0311306362003-04-23023 April 2003 Relief Request No. I2R-23, Limited Examination on Feedwater Nozzle to Steam Generator Weld ML0309202432003-04-0202 April 2003 Relief Request No. 12R-26 Related to Limited Examination on Austenitic Stainless Steel Piping Welds with Single Side Access, MB4080 ET 02-0048, Supplemental Information for Inservice Inspection Program Alternative for Limited Examination on Feedwater Nozzle to Steam Generator Shell Weld, Relief Request L2R-232002-11-0404 November 2002 Supplemental Information for Inservice Inspection Program Alternative for Limited Examination on Feedwater Nozzle to Steam Generator Shell Weld, Relief Request L2R-23 ML0225405752002-10-0404 October 2002 Relief Request, Inservice Inspection Interval ML0133904582002-02-0707 February 2002 Relief, Request to Use Code Case N-597 (Tac No MB2453) 2020-10-30
[Table view] Category:Letter
MONTHYEARML24025A0992024-01-25025 January 2024 Withdrawal of Request for Exemption from Specific 10 CFR Part 73 Requirements ML24018A0792024-01-22022 January 2024 Individual Notice of Consideration of Issuance of Amendment to Renewed Facility Operating License, Proposed Nshcd and Opportunity for a Hearing (EPID L-2024-LLA-0007) (Letter) ML24018A1382024-01-18018 January 2024 Inservice Inspection Request for Information ML24018A2482024-01-18018 January 2024 License Amendment Request to Modify the Implementation Date of License Amendment No. 238 IR 05000482/20230042024-01-11011 January 2024 Integrated Inspection Report 05000482/2023004 ML23356A0722024-01-0404 January 2024 Supplemental Information Needed for Acceptance of Requested Licensing Actions Exemption from Specific 10 CFR Part 73N Requirements (EPID L-2023-LLE-0048) (Redacted Version) WO 23-0002, Operating Corp., Summary of Actions Implemented for EA-18-165, Confirmatory Order, NRC Inspection Report 05000482/2019010 and NRC Investigation Report 4-2018-0082023-12-26026 December 2023 Operating Corp., Summary of Actions Implemented for EA-18-165, Confirmatory Order, NRC Inspection Report 05000482/2019010 and NRC Investigation Report 4-2018-008 ML23348A3662023-12-18018 December 2023 Notification of an NRC Fire Protection Baseline Inspection (NRC Inspection Report 05000482 2024012) and Request for Information ML23334A2502023-11-30030 November 2023 Request for Exemption from Enhanced Weapons, Firearms Background Checks, and Security Event Notifications Implementation ML23299A2662023-11-29029 November 2023 Issuance of Amendment No. 238 Modified Implementation Date of License Amendment No. 237 ML23331A4972023-11-27027 November 2023 Supplement to License Amendment Request to Modify the 90-Day Implementation of License Amendment No. 237 ML23325A2112023-11-20020 November 2023 Submittal of Request for Exemption from Specific Provisions in 10 CFR 73.55 ML23320A2772023-11-16016 November 2023 License Amendment Request to Revise Ventilation Filter Testing Program Criteria in Technical Specification 5.5.11.b and Administrative Correction of Absorber in Technical Specification 5.5.11 ML23311A2082023-11-0909 November 2023 Reassignment of U.S. Nuclear Regulatory Commission Branch Chief in the Division of Operating Reactor Licensing for Plant Licensing Branch IV ML23305A3472023-11-0101 November 2023 Revision of One Form That Implements the Radiological Emergency Response Plan (RERP) ML23292A3592023-10-19019 October 2023 License Amendment Request to Modify the 90-Day Implementation of License Amendment No. 237 ML23292A3572023-10-19019 October 2023 Operating Corporation, Request for Correction in Response to Issuance of Amendment No. 237 and Corresponding Safety Evaluation ML23277A2772023-10-11011 October 2023 Regulatory Audit Summary Concerning Review of Request Number CI3R-01 for Proposed Alternative Inspection Frequency for Containment Unbonded Post Tensioning System Components ML23284A2392023-10-11011 October 2023 NRC Initial Operator Licensing Examination Approval 05000482/2023301 ML23276B4552023-10-0303 October 2023 Technical Specification 5.6.8 Post Accident Monitoring (PAM) Report ML23264A8682023-09-21021 September 2023 Independent Spent Fuel Storage Installation Registration of Dry Storage Canister WC-EOS-37PTH-008-B4L, WC-EOS-37PTH-007-B4L, and WC-EOS-37PTH-002-B4L ML23256A2062023-09-14014 September 2023 Withdrawal of Requested Licensing Action Exemption from Specific 10 CFR Part 73 Requirements ML23165A2502023-08-31031 August 2023 Issuance of Amendment No. 237 Request for Deviation from Fire Protection Requirements ML23243A0002023-08-31031 August 2023 Wolf Generating Station - Withdrawal of Request for Exemption from Specific 10 CFR Part 73 Requirements IR 05000482/20230052023-08-23023 August 2023 Updated Inspection Plan for Wolf Creek Nuclear Operating Corporation, Unit 1 (Report 05000482/2023005) - Mid-Cycle Letter 2023 ML23232A0012023-08-19019 August 2023 Revision of One Procedure That Implements the Radiological Emergency Response Plan (Rerp). Includes EPP 06-009, Rev. 13 ML23221A3932023-08-0909 August 2023 Submittal of Request for Exemption from Specific 10 CFR Part 73 Requirements ML23220A4222023-08-0808 August 2023 Supplement to License Amendment Request (LAR) for Deviation from Fire Protection Program Requirements ML23201A1212023-08-0707 August 2023 Issuance of Amendment No. 236 Revision to Technical Specifications to Adopt TSTF-554, Revise Reactor Coolant Leakage Requirements ML23130A2902023-07-26026 July 2023 Issuance of Amendment No. 235 Revision to Technical Specifications to Adopt TSTF-577, Revised Frequencies for Steam Generator Tube Inspections IR 05000482/20230022023-07-11011 July 2023 Integrated Inspection Report 05000482/2023002 ML23167A0042023-06-26026 June 2023 Audit Plan to Support Review of Request for Alternative Containment Inservice Inspection Frequency IR 05000482/20234012023-06-21021 June 2023 Security Baseline Inspection Report 05000482/2023401 ML23171B1312023-06-20020 June 2023 Additional Supplement to License Amendment Request to Adopt TSTF-577-A, Revision 1, Revised Frequencies for Steam Generator Tube Inspections ML23151A4162023-05-31031 May 2023 Revision of One Form That Implements the Radiological Emergency Response Plan (RERP) for Wolf Creek Generating Station (WCGS) ET 23-0006, CFR 50.55a Request Number CI3R-01 for the Third Containment Inservice Inspection Program Interval for Proposed Alternative Frequency to Containment Unbonded Post-Tensioning System Components2023-05-17017 May 2023 CFR 50.55a Request Number CI3R-01 for the Third Containment Inservice Inspection Program Interval for Proposed Alternative Frequency to Containment Unbonded Post-Tensioning System Components WO 23-0014, Response to NRC Regulatory Issue Summary 2023-01, Preparation and Scheduling of Operator Licensing Examinations2023-05-17017 May 2023 Response to NRC Regulatory Issue Summary 2023-01, Preparation and Scheduling of Operator Licensing Examinations ML23132A1532023-05-12012 May 2023 Transmittal of 2022 Annual Financial Reports WM 23-0009, Financial Assurance Requirements for Decommissioning Nuclear Power Reactors 10 CFR 50.75(f)(1)2023-05-10010 May 2023 Financial Assurance Requirements for Decommissioning Nuclear Power Reactors 10 CFR 50.75(f)(1) ML23135A1172023-05-0808 May 2023 Redacted Version of Revision 36 to Updated Safety Analysis Report ML23117A2112023-04-27027 April 2023 Annual Exposure Report (2022) ML23117A1342023-04-27027 April 2023 2022 Annual Radioactive Effluent Release Report Report 46 2024-01-04
[Table view] Category:Safety Evaluation
MONTHYEARML23299A2662023-11-29029 November 2023 Issuance of Amendment No. 238 Modified Implementation Date of License Amendment No. 237 ML23256A2882023-09-20020 September 2023 Authorization and Safety Evaluation for Alternative Request No. I4R-08 ML23165A2502023-08-31031 August 2023 Issuance of Amendment No. 237 Request for Deviation from Fire Protection Requirements ML23201A1212023-08-0707 August 2023 Issuance of Amendment No. 236 Revision to Technical Specifications to Adopt TSTF-554, Revise Reactor Coolant Leakage Requirements ML23130A2902023-07-26026 July 2023 Issuance of Amendment No. 235 Revision to Technical Specifications to Adopt TSTF-577, Revised Frequencies for Steam Generator Tube Inspections ML22252A1512022-11-0404 November 2022 Issuance of Amendment No. 234 Diesel Generator Completion Time Extension for Technical Specification 3.8.1, AC Sources - Operating ML22199A2942022-08-16016 August 2022 Issuance of Amendment No. 233 Removal of Table of Contents from the Technical Specifications ML22069A0562022-05-18018 May 2022 Issuance of Amendment No. 232 Regarding Revision to the Emergency Plan Related to On-Shift Staffing ML22021B5982022-02-23023 February 2022 Issuance of Amendment No. 231 Revision of Technical Specification 3.3.2, Engineered Safety Feature Actuation System (ESFAS) Instrumentation ML21210A2472021-09-0303 September 2021 Issuance of Amendment No. 230 Revision of Technical Specification 3.6.3 and Surveillance Requirement 3.6.3.1 to Allow Use of a Blind Flange ML21095A1922021-07-20020 July 2021 Issuance of Amendment No. 229 Change to Owner Licensee Names ML21061A0782021-04-23023 April 2021 Issuance of Amendment No. 228 Regarding Replacement of Engineered Safety Features Transformers with New Transformers That Have Active Automatic Load Tap Changers ML21053A1172021-04-0808 April 2021 1 - Issuance of Amendment No. 227 TS Change Risk-Informed Justification for the Relocation of Specific Surveillance Frequency Requirements to a Licensee Controlled Program Based on TSTF-425 ML20276A1492020-12-0707 December 2020 Issuance of Amendment No. 226 Extension of Type a and Type C Leak Rate Test Frequencies ML20302A0802020-10-30030 October 2020 Request to Use a Provision of a Later Edition of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code, Section XI ML20111A3372020-04-27027 April 2020 Request from Relief from Requirements of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code Case N-666-1 ML19353C5002020-02-27027 February 2020 Issuance of Amendment No. 224 Revision to Technical Specification 3.3.5, - Loss of Power (LOP) Diesel Generator (DG) Start Instrumentation, - ML19291A0182019-11-12012 November 2019 Request for Relief I4R-07, Utilize Code Case N-513-4 - Evaluation Criteria for Temporary Acceptance of Flaws in Moderate Energy Class 2 or 3 Piping of American Society of Mechanical Engineers Boiler and Pressure Code Section XI ML19182A3452019-08-19019 August 2019 Issuance of Amendment No 222 Revise Technical Specifications to Adopt TSTF Traveler TSTF-529, Clarify Use and Application Rules ML19100A1222019-05-31031 May 2019 Issuance of Amendment No. 221 License Amendment Request for Transition to Westinghouse Core Design and Safety Analyses Including Adoption of Alternative Source Term ML19052A5462019-04-0101 April 2019 Issuance of Amendment No. 220 Revision to the Emergency Plan ML18334A0132018-12-12012 December 2018 Request for Relief from the Requirements of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code, Code Case N-666-1 ML18283A0492018-10-18018 October 2018 Request for Relief I4R-6 from ASME Code Visual Examination Requirements for Reactor Vessel Head Penetration Nozzle Weld Specified by Code Case N-729-4 ML18040A6662018-03-12012 March 2018 Letter, Order, and Safety Evaluation Order Approving Indirect Transfer of Control of Renewed Facility Operating License No. NPF-42 ML17166A4092017-08-28028 August 2017 Issuance of Amendment No. 218, Request to Adopt Emergency Action Level (EAL) Scheme Pursuant to Nuclear Energy Institute (NEI) 99-01, Revision 6 ML17144A0092017-08-0202 August 2017 Safety Evaluation Regarding Implementation of Mitigating Strategies and Reliable Spent Fuel Pool Instrumentation Related to Orders EA-12-049 and EA-12-051 ML17202E9192017-08-0202 August 2017 Relief Request 14R-05 from Certain Pressure Test Requirements of the ASME Code for Reactor Pressure Vessel Leak-Off Lines for the Fourth 10 Year Inservice Inspection Interval ML17024A2412017-03-24024 March 2017 Issuance of Amendment No. 217 Revision to the Cyber Security Plan Implementation Schedule ML16320A0322016-11-30030 November 2016 Request for Relief I4R-, Alternative Risk-Informed Methodology in Selecting Class 1 and 2 Piping Welds, for the Fourth 10-Year Inservice Inspection Interval ML16179A2932016-08-0303 August 2016 Issuance of Amendment No. 216, Revise Technical Specification (TS) 4.2.1 and TS 5.6.5 to Allow Use of Optimized Zirlo as Approved Fuel Rod Cladding ML16179A4432016-08-0202 August 2016 Safety Evaluation, Request for Exemption from 10 CFR 50.46 and Appendix K to Allow Use of Optimized Zirlo as Approved Fuel Rod Cladding ML16081A1942016-04-15015 April 2016 Issuance of Amendment No. 215, Revise Technical Specification 3.8.1, AC Sources - Operating, Surveillance Requirements 3.8.1.10 and 3.8.1.14 Consistent with TSTF-276-A, Revise DG Full Load Rejection Test ML15183A0522015-09-11011 September 2015 Issuance of Amendment No. 214, Request to Revise Fire Protection Program Related to Alternative Shutdown Capability as Described in Updated Safety Analysis Report ML15203A0052015-08-28028 August 2015 Redacted, Issuance of Amendment No. 213, Revise Technical Specification 5.6.5, Core Operating Limits Report (Colr), for Large-Break Loss-of-Coolant Accident Analysis Methodology (Astrum) ML15169A2132015-07-28028 July 2015 Issuance of Amendment No. 212, Revise Technical Specifications to Adopt TSTF-523, Revision 2, Generic Letter 2008-01, Managing Gas Accumulation ML15190A2022015-07-13013 July 2015 Relief Request 13R-12, Extension of the Third Inservice Inspection Program Interval to Perform Reactor Vessel Stud Hole Ligament Examinations ML15134A0022015-05-15015 May 2015 Relief Request Nos. 4PR-01 and 4PR-02 and Withdrawal of 4VR-02, Alternative to Requirements of ASME Code Case OMN-21 and OM Code ISTB-3510(b)(1) for Pumps, Fourth 10-Year Inservice Testing Program ML15040A0202015-02-12012 February 2015 Relief Request I3R-10, Alternative from Pressure Test Requirements of ASME Code, Section XI, IWC-5220, Third 10-Year Inservice Inspection Interval ML15023A2202015-01-28028 January 2015 Relief Request 13R-11, Alternative from Pressure Test Requirements of ASME Code Section XI IWC-5220 for the Third 10-Year Inservice Inspection Interval ML14321A8642014-12-10010 December 2014 Relief Requests 13R-08, Reactor Pressure Vessel (RPV) Interior Attachments and 13R-09, RPV Pressure-Retaining Welds, Exam Interval Extensions, Third 10-year Inservice Inspection Interval ML14209A0232014-08-14014 August 2014 Issuance of Amendment No. 210, Modify License Condition Related to Approval of Revised Cyber Security Plan Implementation Milestone 8 Completion Date ML14156A2462014-08-0707 August 2014 Issuance of Amendment No. 209, Revise Technical Specification 5.6.5, Core Operating Limits Report (Colr), to Replace Westinghouse Methodologies ML14157A0822014-07-0101 July 2014 Issuance of Amendment No. 208, Adopt TSTF-522-A, Revision 0, Revise Ventilation System to Operate for 10 Hours Per Month, Using Consolidated Line Item Improvement Process ML13282A5342013-12-0606 December 2013 Issuance of Amendment No. 207, Revise Technical Specification 3.4.12, Low Temperature Overpressure Protection (LTOP) System, to Reflect Mass Input Transient Analysis ML13282A1472013-12-0202 December 2013 Issuance of Amendment No. 206, Revise Technical Specification 3.8.1, AC Sources - Operating, Surveillance Requirement 3.8.1.10 to Increase Voltage Limit for Diesel Generator Load Rejection Test ML13197A2102013-08-23023 August 2013 Issuance of Amendment No. 205, Revise Fire Protection License Condition and the Updated Safety Analysis Report for a Deviation from Appendix R for Volume Control Tank Outlet Valves ML13109A2342013-07-12012 July 2013 Safety Evaluation Regarding Wolf Creek Generating Station - Loss of Offsite Power and Augmented Inspection, Issue for Resolution 2012-004 ML13149A1712013-06-0303 June 2013 Safety Assessment in Response to Information Request Pursuant to 10 CFR 50.54(f) - Recommendation 9.3 Communications Assessment ML13078A1552013-03-20020 March 2013 License Amendment Request to Change Diesel Generator Surveillance Requirements ML12353A2412013-01-0404 January 2013 Relief Request I3R-07 from ASME Code Case N-729-1 for Examination of Reactor Vessel Head Penetration Welds, for Remainder of Third 10-Year Inservice Inspection Interval 2023-09-20
[Table view] |
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UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 January 28, 2015 Mr. Adam C. Heflin President, Chief Executive Officer, and Chief Nuclear Officer Wolf Creek Nuclear Operating Corporation Post Office Box 411 Burlington, KS 66839
SUBJECT:
WOLF CREEK GENERATING STATION- REQUEST FOR RELIEF NO. 13R-11 FOR THE THIRD 10-YEAR INSERVICE INSPECTION PROGRAM INTERVAL (TAC NO. MF4304)
Dear Mr. Heflin:
By letter dated June 26, 2014, as supplemented by letter dated August 21, 2014, Wolf Creek Nuclear Operating Corporation (WCNOC, the licensee) proposed an alternative to the inservice inspection (lSI) interval requirements of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code),Section XI, lSI Program, for the Wolf Creek Generating Station (WCGS).
Pursuant to Title 10 of the Code of Federal Regulations (1 0 CFR) paragraph 50.55a(a)(3)(ii)
(retitled paragraph 50.55a(z)(2) by 79 FR 65776, dated November 5, 2014), relief request 13R-11 proposed an alternative to the pressure test requirements of ASME Code Section XI, paragraph IWC-5220 for the Class 2 piping and components in the reactor vessel flange leak-off lines connected to the reactor pressure vessel running to isolation valve BBHV8032.
The U.S. Nuclear Regulatory Commission (NRC) staff has reviewed the subject request and concludes, as set forth in the enclosed safety evaluation, that the licensee has adequately addressed all of the regulatory requirements set forth in 10 CFR 50.55a(z)(2) for 13R-11.
Therefore, the NRC authorizes the use of relief request 13R-11 at WCGS for the remainder of the third lSI interval, which ends on September 2, 2015.
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 lnservice Inspector.
A. Heflin The detailed results of the NRC staff review are provided in the enclosed safety evaluation. If you have any questions concerning this matter, please call Mr. F. Lyon of my staff at (301) 415-2296 or by electronic mail at fred.lyon@nrc.gov.
Sincerely, Eric R. Oesterle, Acting Chief Plant Licensing Branch IV-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-482 Enclosure Safety Evaluation cc w/encl: Distribution via Listserv
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION THIRD 10-YEAR INSERVICE INSPECTION PROGRAM INTERVAL REQUEST FOR RELIEF NO. 13R-11 WOLF CREEK NUCLEAR OPERATING CORPORATION WOLF CREEK GENERATING STATION DOCKET NO. 50-482
1.0 INTRODUCTION
By letter dated June 26, 2014, as supplemented by letter dated August 21, 2014 (Agencywide Documents Access and Management System (ADAMS) Accession Nos. ML14182A091 and ML14239A495, respectively), Wolf Creek Nuclear Operating Corporation (WCNOC, the licensee) proposed an alternative to the inservice inspection (lSI) interval requirements of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code),
Section XI, lSI Program, for the Wolf Creek Generating Station (WCGS).
Pursuant to Title 10 of the Code of Federal Regulations (1 0 CFR) paragraph 50.55a(a)(3)(ii)
(retitled paragraph 50.55a(z)(2) by 79 FR 65776, dated November 5, 2014), relief request 13R-11 proposed an alternative to the pressure test requirements of ASME Code,Section XI, paragraph IWC-5220 for the Class 2 piping and components in the reactor vessel flange leak-off lines connected to the reactor pressure vessel (RPV) running to isolation valve BBHV8032. The licensee proposed an alternative system leakage testing for the Class 2 RPV head flange seal leak-off line piping on the basis that complying with the specified requirement would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety.
The relief request is for the remainder of the third 10-year lSI interval, which ends on September 2, 2015.
2.0 REGULATORY EVALUATION
Pursuant to 10 CFR 50.55a(g)(4), the ASME Code Class 1, 2, and 3 components (including supports) must meet the requirements, except the design and access provisions and the preservice examination requirements, set forth in the ASME Code,Section XI, "Rules for lnservice Inspection of Nuclear Power Plant Components," to the extent practical within the limitations of design, geometry, and materials of construction of the components.
Pursuant to 10 CFR 50.55a(z), alternatives to the requirements of paragraph (g) of 10 CFR 50.55a may be used when authorized by the Director, Office of Nuclear Reactor Enclosure
Regulation. A proposed alternative must be submitted and authorized prior to implementation.
The licensee must demonstrate (1) the proposed alternative would provide an acceptable level of quality and safety; or (2) compliance with the specified requirements of this section would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety.
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 NRC to authorize the alternative requested by the licensee.
3.0 TECHNICAL EVALUATION
3.1 Component Affected The component affected is ASME Code Class 2. In accordance with IWC-2500 (Table IWC-2500-1 ), this component is classified as Examination Category C-H, Item Number C7.10.
The licensee requested relief for the Class 2 RPV flange seal leak-off lines piping. In the June 26, 2014, letter, the licensee stated that the material of construction of the leak-off piping is Type 304 stainless steel.
3.2 Applicable Code Edition and Addenda The Code of record for the third 10-year lSI interval is the 1998 Edition through 2000 Addenda of the ASME Code,Section XI.
3.3 Duration of Relief Request The licensee submitted RR 13R-11 for the remainder of the third 10-year lSI interval, which will end on September 2, 2015.
3.4 ASME Code Requirement The ASME Code,Section XI, IWC-2500, Table IWC-2500-1, Examination Category C-H, requires the system leakage test be conducted according to IWC-5220 and the associated VT-2 visual examination according to IWA-5240 during each inspection period. As required by IWC-5221, the system leakage test shall be conducted at the system pressure obtained while the system, or portion of the system, is in service performing its normal operating function or at the system pressure developed during a test conducted to verify system operability (e.g., to demonstrate system safety function or satisfy technical specification surveillance requirements).
3.5 Proposed Alternative and Basis for Use The licensee proposed an alternative to IWC-5221. To conduct the system leakage test of the RPV leak-off piping, the licensee proposed to subject the piping to the static pressure head, developed from the elevation of at least 23 feet of normal refueling water above the reactor vessel closure flange when the reactor cavity is flooded for refueling, for at least 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. In the
August 21, 2014, letter, the licensee stated that as part of the system leakage test, it will perform the VT -2 visual examinations of the accessible insulated portion of the piping in accordance with IWA-5242 and the accessible non-insulated portion of the piping in accordance with IWA-5241 (a). The licensee will also perform the VT-2 visual examinations of the inaccessible portion of the piping in accordance with IWA-5241(b). In addition, the licensee will perform a supplemental visual examination for boric acid residue indicative of leakage from the leak-off piping when the piping can be made accessible later in the refueling outage after drain down of the refueling cavity when access to the reactor vessel nozzle gallery is made available.
This supplemental visual examination will include opening of the mirror insulation that covers a portion of the inaccessible leak-off piping in the nozzle gallery to allow direct performance of the visual examination.
The licensee stated that the leak-off piping is separated from the reactor coolant pressure boundary by metallic 0-ring seal. The pressure openings for the leak-off piping are located on the RPV flange mating surface. Failure of the inner 0-ring seal is the only condition under which the leak-off piping could be pressurized. Therefore, the leak-off piping is not expected to be pressurized during the system pressure test following a refueling outage or during normal operation. During operating cycle, if the inner 0-ring seal should leak, it will be identified by an increase in temperature above ambient temperature. This piping has a temperature indication and a high temperature alarm in the Control Room which is monitored by the operator. This piping also collects leakage which is routed to the reactor coolant drain tank.
The licensee stated that the leak-off piping would only function as a Class 2 pressure boundary if the inner 0-ring seal fails, thereby, pressurizing the line. If any significant pipe through-wall leakage were to occur in this piping during this time of pressurization, it would exhibit boric acid accumulation that would be identified by the boron trace residue during the VT-2 visual examination following the proposed leakage test. This piping is also subjected to the VT-2 visual examination during the reactor coolant system (RCS) pressure test at the end of each refueling outage for indications of leakage.
In the August 21, 2014, letter, the licensee stated that it has not detected any degradation of the RPV leak-off piping during performance of the VT-2 visual examinations at the end of each refueling outage, which includes accessible portions of the leak-off piping after pressurization from the static head of the flooded refueling cavity. There has not been any degradation or evidence of leakage in these leak-off piping identified in the spring of 2011 when the mirror insulation was removed from the leak-off piping inside the reactor vessel main loop nozzle gallery during walk down and work planning for reactor vessel nozzle mitigation. The licensee further stated that during plant heat up following a refueling outage in January 1988, it identified leakage from the inner 0-ring seal, subsequently cooled down the plant, and replaced the 0-rings. This was the only time the RPV leak-off piping at WCGS experienced elevated pressure since plant construction, and the licensee did not identify any evidence of degradation and leakage in this piping as a result of the inner 0-ring leakage and subsequent pressurization of piping.
The licensee also stated that in an unlikely event, if a through-wall leak would occur in the leak-off piping concurrent with leak or failure of the RPV flange inner 0-ring seal during normal operation, it would result in unidentified RCS leakage that is controlled by Technical Specification 3.4.13, "RCS Operational LEAKAGE." Leakage detection systems have been
designed to aid Control Room operators in differentiating between possible sources of detected leakage within the containment and identifying the physical location of the leak. The RCS leakage detection systems consist of the sump level and flow monitoring system, the containment air particulate monitoring system, the containment cooler condensate measuring system, and the containment humidity monitoring system. The sump level and flow monitoring system indicates leakage by monitoring increases in sump level. The containment cooler condensate measuring system and the containment humidity measuring system detect leakage from the release of steam or water to the containment atmosphere. The air particulate gas monitoring system detects leakage from the release of radioactive materials to the containment atmosphere.
3.6 Basis for Hardship The licensee stated that the configuration of leak-off piping would pose personnel and equipment safety concerns if the pressure testing would be performed at the ASME Code required RCS operating pressure. With the reactor vessel head removed, plugs would need to be installed in the reactor vessel flange face to act as a pressure boundary for the system leakage test of the lines and removed after the test. The installation of the plugs and subsequent activities would cause personnel to incur additional radiological dose due to additional time for personnel at the reactor vessel flange. The handling of a very small diameter plug over the reactor vessel would present a foreign material exclusion issue if accidently dropped into the reactor pool. The use of an alternative test rig to test those isolated portions of piping at the full RCS operating pressure would have to include application of a compatible pressurized medium. This would result in exposing personnel stationed near pressurized vent or drain valves to unnecessary safety hazards in the event of a leak from the non-class test pressure rig connections. A break at any connection of the test rig under such conditions (temporary non-code connections under the RCS test pressure) would pose personnel safety hazards. The configuration also precludes pressurizing the line externally with the reactor vessel head installed. The closure head contains concentric grooves that hold the 0-ring seal.
The 0-ring is held in place by a series of retainer clips that are housed in recessed cavities in the flange face. If a pressure test was to be performed with the reactor vessel head installed, the 0-ring would be pressurized in a direction opposite to its design function. This test pressure would result in a net inward force on the 0-ring that would tend to push it into the recessed cavity that houses the retainer clips. The thin 0-ring material could be damaged by the inward force.
In the August 21, 2014, letter, the licensee stated that conducting the ASME Code required system leakage test before removing the RPV head in the beginning of a refueling outage would require the use of a hydro pump test skid and a non-class test pressure skid connection. This could expose personnel setting up and conducting the test, or stationed near the pressurized vent or drain valves, to unnecessary additional radiation dose and personnel safety hazards in the event of a leak or break of a non-class test pressure skid connections.
The licensee further stated that an accurate estimate of personnel radiation exposure for activities to facilitate the performance of the ASME Code required system leakage test of the leak-off piping (e.g., the installation of the plugs and/or non-class test pressure rig connections, subsequent testing, and removal after testing) cannot be made because these activities has never been performed in the past. However, based on past refueling outage radiation dose
rates survey, the estimated dose rates at the reactor vessel flange are 3 to 4 roentgen equivalent man per hour (rem/h).
- 3. 7 NRC Staff Evaluation The NRC staff has evaluated RR 13R-11 pursuant to 10 CFR 50.55a(z)(2). The NRC staff focuses on whether compliance with the specified requirements of 10 CFR 50.55a(g), or portions thereof, would result in hardship or unusual difficulty, without a compensating increase in the level of quality and safety.
Hardship The NRC staff found that requiring the licensee to comply with IWC-5221 and conduct system leakage test of the RPV head flange seal leak-off lines piping at the RCS operating pressure would result in hardship. The basis for the hardship is as follows. To conduct the ASME Code required system leakage test of the leak-off piping when the reactor head is removed during refueling, the licensee would have to modify the existing RPV head flange taps to install plugs and/or non-class test pressure skid connections to facilitate for pressurizing the piping by use of a hydro pump test skid. The activities associated with installing the plugs and/or the test connections, pressurizing the piping to the RCS pressure and conducting the ASME Code required system leakage test, and removing the plugs after completion of test would cause personnel to incur additional radiation dose, and could introduce foreign materials into the reactor pool as well as the lines. Pressurizing the non-class test pressure skid connections to the RCS operating pressure would create personnel safety hazards in the event of a leak or break in any of the non-class test pressure rig connections. Pressurizing the lines to conduct the ASME Code required system leakage test when the RPV head is installed would not be possible due to design and configuration of the RPV head flange taps and the inner 0-ring. The inner 0-ring is designed to withstand pressure in one direction only, pressurizing in the opposite direction could damage the inner 0-ring, and even result in unsuccessful test. Externally pressurizing the lines to conduct the ASME Code required system leakage test at the beginning of refueling outage when the RPV head is on would not be possible because the entire containment would have limited access due to high radiation level. Furthermore, pressurizing the inner 0-ring in the opposite direction could damage the 0-ring and result in unsuccessful test. Therefore, the NRC staff determined that concerns from the Foreign Material Exclusion program and an as low as is reasonably achievable criteria constitute a hardship.
Test Pressure In evaluating the licensee's proposed alternative, the NRC staff assessed whether it appeared that the licensee used the highest achievable test pressure to conduct system leakage testing and the manner in which the licensee adequately preformed the testing and the associated VT-2 visual examinations of the piping for leakage. The NRC staff found that the licensee will use the highest pressure that is obtainable without major modifications to existing configuration of the lines to test the RPV leak-off piping for leakage. Specifically, the licensee's proposed system leakage test will subject the piping to the static pressure head developed from the elevation of 23 feet of refueling water above the vessel flange during the refueling cavity flood-up which eliminates a need for major design modifications to existing configurations of both the vessel flange and the leak-off lines. By performing the VT-2 visual examination of the insulated area of
the piping (according to IWA-5242) and the non-insulated area of the piping (according to IWA-4241 ), the licensee will be able to detect any leakage if it originated from an existing flaw in the piping and its welded connections after maintaining the static test pressure. As a supplement to IWA-5241 and IWA-5242, the licensee will visually examine the inaccessible area of the piping for boric acid residue when the piping can be made accessible later in the refueling outage after drain down of the refueling cavity when access to the reactor vessel nozzle gallery is made available. This supplemental examination will include opening of the mirror insulation that covers a portion of the inaccessible leak-off piping in the nozzle gallery to allow direct VT-2 visual examinations. Therefore, the NRC staff found that the licensee's alternative system leakage test subjects the piping under consideration to a test pressure that is as high as reasonably achievable.
Safety Significance of Alternative Test Pressure In addition to the analysis described above, the NRC staff evaluated the safety significance of performance of the system leakage test at an alternative reduced pressure. The NRC staff notes that the leak-off piping is made of stainless steel. The degradation mechanism could be fatigue and stress-corrosion cracking. However, fatigue crack is known to have relatively slow growth and field experience has shown that stress-corrosion cracking under normal operating conditions is not expected to be a problem. Significant degradation would likely be detected by the system leakage test performed under proposed maximum obtainable static pressure head.
The NRC staff notes that if in an unlikely event, these piping developed a through wall flaw and a leak, the WCGS existing reactor coolant leakage detection systems will be able to identify the leakage during normal operation, and the licensee will take appropriate corrective actions in accordance with the plant technical specifications. Therefore, the NRC staff determined that based on the alternative system leakage testing that subject this piping to the maximum obtainable static pressure head and the performance of the ASME Code required VT-2 visual examinations, it is reasonable to conclude that if significant service-induced degradation had occurred, evidence of it would have be detected either by the examinations that the licensee performed or the RCS leakage detection systems.
Therefore, the NRC staff concludes that the proposed system leakage testing using the proposed test pressure is adequate to provide a reasonable assurance of structural integrity and leak-tightness of the RPV flange seal leak-off lines piping.
4.0 CONCLUSION
As set forth above, the NRC staff determines that the proposed alternative provides reasonable assurance of structural integrity and leak-tightness of the RPV head flange seal leak-off lines piping. The NRC staff finds that complying with the specified ASME Code requirement would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety. Accordingly, the NRC staff concludes that the licensee has adequately addressed all of the regulatory requirements set forth in 10 CFR 50.55a(z)(2). Therefore, the NRC staff authorizes the use of RR 13R-11 for the remainder of the third 10-year lSI interval, which ends on September 2, 2015.
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 lnservice Inspector.
Principal Contributor: A. Rezai, NRR/DE/EPNB Date: January 28, 2015
ML15023A220 *email dated OFFICE NRR/DORULPL4-1/PM NRR/DORLILPL4-1/LA NRR/DE/EPNB/BC NRR/DORULPL4-1/BC(A)
NAME Flyon JBurkhardt DAiley* EOesterle DATE 1/28/15 1/27/15 1/22/15 1/28/15