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MONTHYEARET 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 Project stage: Other ML12216A0272012-08-0303 August 2012 Acceptance Review Email, Relief Request I3R-07 from ASME Code Case N-7229 for Examination of Reactor Vessel Head Penetration Welds, for Remainder of Third 10-Year Inservice Inspection Interval Project stage: Acceptance Review ML12248A0802012-09-0404 September 2012 Email, Request for Additional Information, 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 Project stage: RAI ET 12-0024, Response to Request for Additional Information Regarding 10 CFR 50.55a Request Number I3R-07, Relief from ASME Code Case N-729-1 Requirements for Examination of Reactor Vessel Head Penetration Welds2012-10-15015 October 2012 Response to Request for Additional Information Regarding 10 CFR 50.55a Request Number I3R-07, Relief from ASME Code Case N-729-1 Requirements for Examination of Reactor Vessel Head Penetration Welds Project stage: Response to RAI 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 Project stage: Other 2012-08-03
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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 type:ET
MONTHYEARET 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 ET 23-0005, 10 CFR 50.55a Request Number I4R-08 for the Fourth Inservice Inspection Program Interval, Relief for Extension of Follow Up Examination Requirements for Reactor Pressure Vessel Head Penetration Nozzles with Mitigated Alloy 600/82/182 Peen2023-03-16016 March 2023 10 CFR 50.55a Request Number I4R-08 for the Fourth Inservice Inspection Program Interval, Relief for Extension of Follow Up Examination Requirements for Reactor Pressure Vessel Head Penetration Nozzles with Mitigated Alloy 600/82/182 Peened ET 23-0003, License Amendment Request (LAR) for Removal of the Power Range Neutron Flux Rate - High Negative Rate Trip Function from Technical Specifications2023-03-0101 March 2023 License Amendment Request (LAR) for Removal of the Power Range Neutron Flux Rate - High Negative Rate Trip Function from Technical Specifications ET 23-0002, Supplement to License Amendment Request to Adopt TSTF-577-A, Revision 1, Revised Frequencies for Steam Generator Tube Inspections2023-02-0707 February 2023 Supplement to License Amendment Request to Adopt TSTF-577-A, Revision 1, Revised Frequencies for Steam Generator Tube Inspections ET 23-0004, Response to Requests for Additional Information (RAI) Regarding License Amendment Request (LAR) for Deviation from Fire Protection Program Requirements2023-01-26026 January 2023 Response to Requests for Additional Information (RAI) Regarding License Amendment Request (LAR) for Deviation from Fire Protection Program Requirements ET 22-0006, Operating Corp. - Application to Revise Technical Specifications to Adopt TSTF-577-A, Revision 1, Revised Frequencies for Steam Generator Tube Inspections2022-12-0101 December 2022 Operating Corp. - Application to Revise Technical Specifications to Adopt TSTF-577-A, Revision 1, Revised Frequencies for Steam Generator Tube Inspections ET 22-0010, License Amendment Request (LAR) for Deviation from Fire Protection Program Requirements2022-08-0202 August 2022 License Amendment Request (LAR) for Deviation from Fire Protection Program Requirements ET 22-0012, Supplement to License Amendment Request - Diesel Generator Completion Time Extension for Technical Specification 3.8.1, AC Sources - Operating2022-07-12012 July 2022 Supplement to License Amendment Request - Diesel Generator Completion Time Extension for Technical Specification 3.8.1, AC Sources - Operating ET 22-0011, Withdrawal of 10 CFR 50.55a Request I4R-08, Relief for Extension of Follow-up Examination and Visual Examination Requirements for Reactor Pressure Vessel Head Penetration Nozzles with Mitigated Alloy 600/82/182 Peened Surface2022-05-31031 May 2022 Withdrawal of 10 CFR 50.55a Request I4R-08, Relief for Extension of Follow-up Examination and Visual Examination Requirements for Reactor Pressure Vessel Head Penetration Nozzles with Mitigated Alloy 600/82/182 Peened Surface ET 22-0005, Response to Request for Additional Information Regarding License Amendment Request - Diesel Generator Completion Time Extension for Technical Specification 3.8.1, AC Sources - Operating2022-04-13013 April 2022 Response to Request for Additional Information Regarding License Amendment Request - Diesel Generator Completion Time Extension for Technical Specification 3.8.1, AC Sources - Operating ET 22-0003, CFR 50.55a Request I4R-09 for the Fourth Inservice Inspection Program Interval, Relief from Examination of Reactor Vessel Flange Threads2022-04-0606 April 2022 CFR 50.55a Request I4R-09 for the Fourth Inservice Inspection Program Interval, Relief from Examination of Reactor Vessel Flange Threads ET 22-0002, Operating Corp 10 CFR 50.55a Request Number I4R-08 for the Fourth Inservice Inspection Program Interval, Relief for Extension of Follow Up Examination and Visual Examination Requirements for Reactor Pressure Vessel Head .2022-04-0404 April 2022 Operating Corp 10 CFR 50.55a Request Number I4R-08 for the Fourth Inservice Inspection Program Interval, Relief for Extension of Follow Up Examination and Visual Examination Requirements for Reactor Pressure Vessel Head . ET 22-0004, Operating Corp - Response to Request for Confirmation of Information (RCI) Regarding Steam Generator Tube Inspection Report2022-03-15015 March 2022 Operating Corp - Response to Request for Confirmation of Information (RCI) Regarding Steam Generator Tube Inspection Report ET 22-0001, Removal of the Table of Contents from the Technical Specifications2022-01-12012 January 2022 Removal of the Table of Contents from the Technical Specifications ET 21-0017, Response to Request for Additional Information Regarding License Amendment Request - Revision to Technical Specification 3.3.2, Engineered Safety Feature Actuation System (ESFAS) Instrumentation2021-12-22022 December 2021 Response to Request for Additional Information Regarding License Amendment Request - Revision to Technical Specification 3.3.2, Engineered Safety Feature Actuation System (ESFAS) Instrumentation ET 21-0009, Results of the Steam Generator Tube Inservice Inspection During the 24th Refueling Outage2021-11-0101 November 2021 Results of the Steam Generator Tube Inservice Inspection During the 24th Refueling Outage ET 21-0015, Withdrawal of License Amendment Request for a Risk-Informed Resolution to GSI-1912021-10-20020 October 2021 Withdrawal of License Amendment Request for a Risk-Informed Resolution to GSI-191 ET 21-0012, Supplement to License Amendment Request for a Risk-Informed Resolution to GSI-1912021-10-11011 October 2021 Supplement to License Amendment Request for a Risk-Informed Resolution to GSI-191 ET 21-0004, License Amendment Request - Diesel Generator Completion Time Extension for Technical Specification 3.8.1, AC Sources - Operating2021-09-29029 September 2021 License Amendment Request - Diesel Generator Completion Time Extension for Technical Specification 3.8.1, AC Sources - Operating ET 21-0010, License Amendment Request - Revision to Technical Specification 3.3.2, Engineered Safety Feature Actuation System (ESFAS) Instrumentation2021-09-29029 September 2021 License Amendment Request - Revision to Technical Specification 3.3.2, Engineered Safety Feature Actuation System (ESFAS) Instrumentation ET 21-0005, Operating Corp., License Amendment Request for a Risk-Informed Resolution to GSI-1912021-08-12012 August 2021 Operating Corp., License Amendment Request for a Risk-Informed Resolution to GSI-191 ET 20-0013, Response to Request for Additional Information Re Application for Technical Specification Change Re Risk-Informed Justification for Relocation of Specific Surveillance Frequency Requirements to Licensee Controlled Program (TSTF-415)2020-10-26026 October 2020 Response to Request for Additional Information Re Application for Technical Specification Change Re Risk-Informed Justification for Relocation of Specific Surveillance Frequency Requirements to Licensee Controlled Program (TSTF-415) ET 20-0011, Request to Use a Provision of a Later Edition of the ASME Boiler and Pressure Vessel Code, Section XI2020-10-0101 October 2020 Request to Use a Provision of a Later Edition of the ASME Boiler and Pressure Vessel Code, Section XI ET 20-0007, License Amendment Request for Replacement of Engineered Safety Features Transformers with New Transformers That Have Active Automatic Load Tap Changers2020-06-0808 June 2020 License Amendment Request for Replacement of Engineered Safety Features Transformers with New Transformers That Have Active Automatic Load Tap Changers ET 20-0008, Operating Corporation Update for Full Implementation of Open Phase Detection System2020-05-20020 May 2020 Operating Corporation Update for Full Implementation of Open Phase Detection System ET 20-0004, Application for Technical Specification Change Regarding Risk-Informed Justification for the Relocation of Specific Surveillance Frequency Requirements to a Licensee Controlled Program (TSTF-425)2020-04-27027 April 2020 Application for Technical Specification Change Regarding Risk-Informed Justification for the Relocation of Specific Surveillance Frequency Requirements to a Licensee Controlled Program (TSTF-425) ET 20-0002, Response to Request for Additional Information Regarding Utilizing Code Case N-666-1, Weld Overlay of Class 1, 2, and 3 Socket Welded Connections, Section XI, Division 12020-01-29029 January 2020 Response to Request for Additional Information Regarding Utilizing Code Case N-666-1, Weld Overlay of Class 1, 2, and 3 Socket Welded Connections, Section XI, Division 1 ET 19-0021, Errata for Supplemental Response to Request for Additional Information Revision to Technical Specification 3.3.5, Loss of Power (LOP) Diesel Generator (DG) Start Instrumentation2019-12-0909 December 2019 Errata for Supplemental Response to Request for Additional Information Revision to Technical Specification 3.3.5, Loss of Power (LOP) Diesel Generator (DG) Start Instrumentation ET 19-0020, Supplemental Response to Request for Additional Information Revision to Technical Specification 3.3.5, Loss of Power (LOP) Diesel Generator (DG) Start Instrumentation2019-11-13013 November 2019 Supplemental Response to Request for Additional Information Revision to Technical Specification 3.3.5, Loss of Power (LOP) Diesel Generator (DG) Start Instrumentation ET 19-0019, Response to Request for Additional Information Revision to Technical Specification 3.3.5, Loss of Power (LOP) Diesel Generator (DG) Start Instrumentation.2019-09-10010 September 2019 Response to Request for Additional Information Revision to Technical Specification 3.3.5, Loss of Power (LOP) Diesel Generator (DG) Start Instrumentation. ET 19-0018, Supplement to License Amendment Request to Revise Technical Specification 3.3.5, Loss of Power (LOP) Diesel Generator (DG) Start Instrumentation.2019-08-22022 August 2019 Supplement to License Amendment Request to Revise Technical Specification 3.3.5, Loss of Power (LOP) Diesel Generator (DG) Start Instrumentation. ET 19-0014, lnserv1ce Inspection (ISI) Program Relief Request Number 14R-07, to Utilize Code Case N-513-4, Evaluation Criteria for Temporary Acceptance of Flaws in Moderate Energy Class 2 or 3 Piping Section XI, D1v1s1on 12019-08-15015 August 2019 lnserv1ce Inspection (ISI) Program Relief Request Number 14R-07, to Utilize Code Case N-513-4, Evaluation Criteria for Temporary Acceptance of Flaws in Moderate Energy Class 2 or 3 Piping Section XI, D1v1s1on 1 ET 19-0002, License Amendment Request to Revise Technical Specification 3.3.5, Loss of Power (LOP) Diesel Generator (DG) Start Instrumentation.2019-03-18018 March 2019 License Amendment Request to Revise Technical Specification 3.3.5, Loss of Power (LOP) Diesel Generator (DG) Start Instrumentation. ET 19-0008, Response to Request for Additional Information Related to Thermal Conductivity Degradation for License Amendment Request to Revise Technical Specifications to Transition to Westinghouse Core Design and Safety Analysis Including Adoption.2019-03-0505 March 2019 Response to Request for Additional Information Related to Thermal Conductivity Degradation for License Amendment Request to Revise Technical Specifications to Transition to Westinghouse Core Design and Safety Analysis Including Adoption. ET 19-0003, License Amendment Request to Revise Technical Specification 3.6.3 and Surveillance Requirement 3.6.3.1 to Remove Use of a Blind Flange2019-02-25025 February 2019 License Amendment Request to Revise Technical Specification 3.6.3 and Surveillance Requirement 3.6.3.1 to Remove Use of a Blind Flange ET 19-0001, Application to Revise Technical Specifications to Adopt TSTF-529, Clarify Use and Application Rules, Revision 42019-01-23023 January 2019 Application to Revise Technical Specifications to Adopt TSTF-529, Clarify Use and Application Rules, Revision 4 ET 18-0032, Operating Corporation Change of Date for Full Implementation of Open Phase Detection System2018-12-0707 December 2018 Operating Corporation Change of Date for Full Implementation of Open Phase Detection System ET 18-0035, Operating Corp., Supplemental Response to RAI for License Amendment Request to Revise Technical Specifications to Transition to Westinghouse Core Design and Safety Analysis Including Adoption of Alternative Source Term2018-12-0606 December 2018 Operating Corp., Supplemental Response to RAI for License Amendment Request to Revise Technical Specifications to Transition to Westinghouse Core Design and Safety Analysis Including Adoption of Alternative Source Term ET 18-0029, (WCGS) - Results of the Twenty First Steam Generator Tube Inservice Inspection2018-10-31031 October 2018 (WCGS) - Results of the Twenty First Steam Generator Tube Inservice Inspection ET 18-0018, Supplement to License Amendment Request to Revise Technical Specifications to Transition to Westinghouse Core Design and Safety Analysis Including Adoption of Alternative Source Term2018-06-19019 June 2018 Supplement to License Amendment Request to Revise Technical Specifications to Transition to Westinghouse Core Design and Safety Analysis Including Adoption of Alternative Source Term ET 18-0016, Response to Generic Letter 2016-01, Monitoring of Neutron Absorbing Materials in Spent Fuel Pools Request for Supplemental Information2018-05-29029 May 2018 Response to Generic Letter 2016-01, Monitoring of Neutron Absorbing Materials in Spent Fuel Pools Request for Supplemental Information ET 18-0014, Response to Request for Additional Information Regarding the License Amendment Request for Addition of New Technical Specification 3.7.20, Class 1E Electrical Equipment Air Conditioning (A/C) System.2018-05-29029 May 2018 Response to Request for Additional Information Regarding the License Amendment Request for Addition of New Technical Specification 3.7.20, Class 1E Electrical Equipment Air Conditioning (A/C) System. ET 18-0013, Relief Request Number I4R-06, Request for Relief from ASME Code Case N-729-4 for Reactor Vessel Head Penetration Nozzle Weld2018-05-0202 May 2018 Relief Request Number I4R-06, Request for Relief from ASME Code Case N-729-4 for Reactor Vessel Head Penetration Nozzle Weld ET 18-0011, (WCGS) - Guarantee of Payment of Deferred Premiums2018-04-30030 April 2018 (WCGS) - Guarantee of Payment of Deferred Premiums ET 18-0012, Operating Corp., Supplement to License Amendment Request to Revise Technical Specifications to Transition to Westinghouse Core Design and Safety Analysis Including Adoption of Alternative Source Term2018-04-19019 April 2018 Operating Corp., Supplement to License Amendment Request to Revise Technical Specifications to Transition to Westinghouse Core Design and Safety Analysis Including Adoption of Alternative Source Term ET 18-0010, Financial Protection Levels2018-03-29029 March 2018 Financial Protection Levels ET 18-0007, Supplement to License Amendment Request for Addition of New Technical Specification 3.7.20, Class 1E Electrical Equipment Air Conditioning (A/C) System.2018-02-15015 February 2018 Supplement to License Amendment Request for Addition of New Technical Specification 3.7.20, Class 1E Electrical Equipment Air Conditioning (A/C) System. ET 18-0005, Withdrawal of License Amendment Request for Revision to the Emergency Plan2018-02-0505 February 2018 Withdrawal of License Amendment Request for Revision to the Emergency Plan ET 18-0004, Supplement to License Amendment Request to Revise Technical Specifications to Transition to Westinghouse Core Design and Safety Analysis Including Adoption of Alternative Source Term2018-01-29029 January 2018 Supplement to License Amendment Request to Revise Technical Specifications to Transition to Westinghouse Core Design and Safety Analysis Including Adoption of Alternative Source Term ET 17-0025, Response to Request for Additional Information Regarding License Amendment Request to Revise Technical Specifications to Transition to Westinghouse Core Design and Safety Analyses Including Adoption of Alternative Source Term2017-11-14014 November 2017 Response to Request for Additional Information Regarding License Amendment Request to Revise Technical Specifications to Transition to Westinghouse Core Design and Safety Analyses Including Adoption of Alternative Source Term 2023-05-17
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W0LF CREEK NUCLEAR OPERATING CORPORATION John P. Broschak July 2, 2012 Vice President Engineering ET 12-0010 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555
Subject:
Docket 50-482: 10 CFR 50.55a Request Number 13R-07, Relief from ASME Code Case N-729-1 Requirements for Examination of Reactor Vessel Head Penetration Welds Gentlemen:
Pursuant to 10 CFR 50.55a(a)(3)(ii), Wolf Creek Nuclear Operating Corporation (WCNOC) hereby requests Nuclear Regulatory Commission (NRC) approval of 10 CFR 50.55a Request Number 13R-07 (attached) for the Third Ten-Year Interval of WCNOC's Inservice Inspection (ISI) Program. The attached 10 CFR 50.55a Request (13R-07) requests relief from certain ASME Code Case N-729-1 requirements for examination of reactor vessel head penetration welds.
The Code of Federal Regulations 10 CFR 50.55a(g)(6)(ii)(D)(1) requires that examinations of the reactor vessel head be performed in accordance with ASME Code Case N-729-1 subject to conditions specified in paragraphs 10 CFR 50.55a(g)(6)(ii)(D)(2) through (6). The vendor chosen by WCNOC to perform these examinations is unable to meet required examination coverage below the J-groove weld on two control rod drive mechanism (CRDM) penetrations.
Both of these CRDM penetrations are configured such that the volumetric examination distance required by N-729-1 cannot be met. The attachment to this letter, 10 CFR 50.55a Request 13R-07, documents the ultrasonic coverage limitations.
An extended forced outage to effect repairs in conjunction with a longer than expected Refueling Outage 18 has resulted in a decision to move the scheduled start dates of Refueling Outages 19 and 20. Therefore, examination of reactor vessel head penetration welds has been moved from Refueling Outage 20 to Refueling Outage 19 to maintain compliance with the required examination frequency. WCNOC requests approval of the attached 10 CFR 50.55a Request 13R-07 by January 14, 2013, to support planning for Refueling Outage 19, which is now scheduled to begin February 4, 2013.
P.O. Box 411 / Burlington, KS 66839 / Phone: (620) 364-8831 An Equal Opportunity Employer M/F/HCNET
ET 12-0010 Page 2 of 2 There are no commitments contained within this letter. If you have any questions concerning this matter, please contact me at (620) 364-4085, or Mr. Gautam Sen at (620) 364-4175.
Sincerely, John P. Broschak JPB/rlt
Attachment:
10 CFR 50.55a Request Number 13R-07 cc: E. E. Collins (NRC), w/a J. R. Hall (NRC), w/a N. F. O'Keefe (NRC), w/a Senior Resident Inspector (NRC), w/a
Attachment to ET 12-0010 Page 1 of 10 Wolf Creek Nuclear Operating Corporation 10 CFR 50.55a Request 13R-07 Request for Relief from the Requirements of ASME Code Case N-729-1
Attachment to ET 12-0010 Page 2 of 10 10 CFR 50.55a Request 13R-07 Request for Relief from the Requirements of ASME Code Case N-729-1 Proposed Alternative in Accordance with 10 CFR 50.55a(a)(3)(ii)
Hardship or Unusual Difficulty Without Compensating Increase in Level of Quality or Safety
- 1. ASME Code Components Affected Code Class: 1
Reference:
ASME Code Case N-729-1 / 10 CFR 50.55a(g)(6)(ii)(D)
Item No.: B4.20
Description:
UNS N06600 Nozzles and UNS N06082 or UNS W86182 Partial-Penetration Welds in Head.
Reactor vessel head control rod drive mechanism (CRDM) penetration nozzle base material and J-groove weld that attaches the nozzle base material to the underside of the head for penetration nozzles 77 and 78.
- 2. Applicable Code Edition and Addenda ASME Code Section Xl, 1998 Edition through 2000 Addenda, as augmented by ASME Code Case N-729-1 (Reference 1), "Alternative Examination Requirements for PWR Reactor Vessel Upper Heads with Nozzles Having Pressure-Retaining Partial-Penetration WeldsSection XI, Division 1," as amended by 10 CFR 50.55a(g)(6)(ii)(D).
- 3. Applicable Code Requirement 10 CFR 50.55a(g)(6)(ii)(D)(1) requires that examinations of the reactor vessel head be performed in accordance with ASME Code Case N-729-1 subject to the conditions specified in paragraphs 10 CFR 50.55a(g)(6)(ii)(D)(2) through (6).
Paragraph -2500 of Code Case N-729-1 states, in part:
If obstructions or limitations prevent examination of the volume or surface required by Figure 2 for one or more nozzles, the analysis procedure of Appendix I shall be used to demonstrate the adequacy of the examination volume or surface for each such nozzle. If Appendix I is used, the evaluation shall be submitted to the regulatory authority having jurisdiction at the plant site.
Figure 2 in ASME Code Case N-729-1, as referenced by paragraph -2500, requires that the volumetric or surface examination coverage distance below the toe of the J-groove weld (i.e. dimension "a") be 1.5 inches for incidence angle, e, less than or equal to 30
Attachment to ET 12-0010 Page 3 of 10 degrees; 1 inch for incidence angle, 8, greater than 30 degrees; or to the end of the tube, whichever is less. These coverage requirements are applicable to Wolf Creek Generating Station (WCGS) reactor vessel head penetrations as shown in Table 1.
Table 1: WCGS Reactor Vessel Head Penetration Coverage Requirements Penetration Numbers Incidence Angle, 0 Required Coverage, "a" (degrees) (inches) 1 to 29 -*30 1.5 30 to 78 > 30 1.0
- 4. Reason for Request Due to physical configuration of certain reactor vessel head penetration nozzles, full examination volume required by ASME Code Case N-729-1 Table 1 cannot be achieved for reactor vessel head penetration nozzles 77 and 78, therefore, use of Mandatory Appendix I is requested in accordance with 10 CFR 50.55a(g)(6)(ii)(D)(6).
Reactor vessel head CRDM penetrations at WCGS have two styles of ends, referred to as Type "X" and Type "Y" (Figure 1). Penetrations 1 through 73 are Type "Y" that are essentially a smooth wall cylinder with a radius at the outer diameter and inner diameter.
Penetrations 74 through 78 have a threaded outside diameter and an internal taper.
The design of reactor vessel head penetration nozzles 74 through 78, referred to as Type "X", (Figure 1) includes a threaded section, approximately 1.19 inch in length at the bottom of the nozzles. These penetrations are located at the 48.7 degree location. The dimensional configuration at this location is such that the distance from the lowest point at the toe of the J-groove weld to the top of the threaded region could be less than the required coverage dimension "a" shown in Figure 2 of ASME Code Case N-729-1.
Therefore, deviation from the required inspection coverage is sought for reactor vessel head penetrations 77 and 78, as the required coverage for these two penetrations cannot be obtained.
For the initial examinations of reactor vessel head penetration welds performed in accordance with Reference 6, a similar request was previously submitted for inability to examine the required examination volume (References 3 and 4). This previous request was approved by the NRC in Reference 5.
- 5. Proposed Alternative and Basis for Use As an alternative to the volumetric and surface examination coverage requirements shown as dimension "a" in Figure 2 of ASME Code Case N-729-1, WCGS proposes the use of attainable ultrasonic examination distances shown in Table 2. The required examination coverage dimension for the other penetrations will be met or exceeded.
Attachment to ET 12-0010 Page 4 of 10 Table 2: WCGS Inspection Coverage Obtained for CRDM Penetrations Having Limited Coverage Penetration No. 0 (degrees) N-729-1 Required Inspection Coverage Exam Coverage Obtained (inches)
(inches) 77 48.7 1.0 0.6 78 48.7 1.0 0.88 Appendix I of ASME Code Case N-729-1 provides the analysis procedure for evaluation of an alternative examination area or volume to that specified in Figure 2 of Code Case N-729-1 if impediments prevent examination of the complete zone. Section 1-1000 of ASME Code Case N-729-1 requires, for alternative examination zones below the J-groove weld, that analyses shall be performed using at least the stress analysis method (Section I-2000) or the deterministic fracture mechanics analysis method (Section 1-3000) to demonstrate that the applicable criteria are satisfied. The techniques described in Section 1-2000 were validated in WCAP-16589-P (Reference 2). Although not required, the deterministic fracture mechanics analysis described in Section 1-3000 was also validated in Reference 2. This analysis does not fully meet the requirements stated in 1-3200(a)
Method 1 in that Reference 2 used the crack growth formula in the Electric Power Research Institute (EPRI) report, "Materials Reliability Program (MRP) Crack Growth Rates for Evaluating Primary Stress Corrosion Cracking (PWSCC) of Thick Wall Alloy 600Material (MRP-55), Revision 1."
5.1 Stress Analysis in Accordance with ASME Code Case N-729-1 Section 1-2000 Section 1-2000 of ASME Code Case N-729-1 requires that plant-specific analysis demonstrate that the hoop and axial stresses remain below 20 kips per square inch (ksi) (tensile) over the entire region outside the alternative examination zone but within the examination zone defined in Figure 2 of the Code Case.
The distance below the J-groove weld that requires examination, as determined by the point at which the CRDM penetration hoop stress distribution for the operating stress levels is less than 20 (ksi) tension, was obtained from Appendix A of Reference 2. Note that hoop stresses during steady state operation are much greater than the axial stresses.
The hoop stress distribution plots for penetrations 77 and 78 are provided in Figure 2 of this submittal. The hoop stress distribution plots in Figure 2 indicate that the minimum achievable inspection coverage below the bottom of the J-groove weld insures stresses remain below 20 ksi tensile over the entire region outside the alternative examination zone but within the examination zone defined in Figure 2 of ASME Code Case N-729-1. The hoop stress distribution plots display the downhill side as this is more limiting. Also, stress distribution plots shown are for the inside and outside surface. Table 3 summarizes the distance from below the toe of the downhill side J-groove weld to where both the inside and outside surface hoop stress drops below 20 ksi for penetrations 77 and 78.
Attachment to ET 12-0010 Page 5 of 10 Table 3: Distance Below Toe of Downhill Side J-Groove Weld Where Hoop Stress is Less Than 20 KSI Penetration Nozzle Source Distance Below Toe of No. Downhill Side J-Groove Weld Where Hoop Stress
< 20 ksi (inch) 77 and 78 Figure 2 0.30 5.2 Deterministic Fracture Mechanics Analysis in Accordance with ASME Code Case N-729-1 Section 1-3200, Method I A fracture mechanics analysis was performed and documented in Reference 2. As previously stated, this analysis is not required and does not fully meet the requirements stated in 1-3200(a) Method 1 because the analysis used the crack growth formula in EPRI MRP-55. The analysis does demonstrate that a potential axial crack in the unexamined zone will not grow to the toe of the J-groove weld prior to the examination frequency specified in Table 1 of ASME Code Case N-729-1.
The fracture mechanics analysis was performed using input from the previously discussed stress analysis. The results of the analysis are shown as flaw tolerance charts, which can be used to determine minimum required inspection coverage. This insures that any flaws initiated below the weld, in the region of the penetration nozzle not being inspected, would not reach the bottom of the weld before the next inspection. The flaw tolerance chart for penetrations 77 and 78 is presented in Figure 3.
The flaw tolerance chart in Figure 3 demonstrates that a postulated through-wall flaw at the bottom edge of the proposed alternative examination zone will not grow to the toe of the J-groove weld within an inspection interval of four refueling cycles. The crack growth prediction shows greater than six effective full power years (EFPY) of operation required to grow the postulated flaw to the toe of the weld. Additionally, the assumed initial upper extremity locations of axial through-wall flaws are conservative based on achievable inspection coverage, because the assumed upper crack extremities are located within the achievable inspection zone.
Examination of portions of the nozzle significantly below the J-groove weld is not pertinent to the phenomena of concern, which include leakage through the J-groove weld and circumferential cracking in the nozzle above the J-groove weld. In all cases, the measured coverage is adequate to allow WCGS to continue to operate prior to the hypothetical flaws reaching the J-groove weld. In accordance with 10 CFR 50.55a(g)(6)(ii)(D) requirements, the next required examination would be completed prior to potential flaw propagation into the J-groove welds.
Attachment to ET 12-0010 Page 6 of 10 5.3 Surface Examination 10 CFR 50.55a(g)(6)(ii)(D)(3) states in part that "if a surface examination is being substituted for a volumetric examination on a portion of a penetration nozzle that is below the toe of the J-groove weld, the surface examination shall be of the inside and outside wetted surface of the penetration nozzle not examined volumetrically."
To reduce personnel radiation exposure, the nozzles are typically inspected using remotely operated volumetric examination equipment. Although dye penetrant testing of threaded surfaces is possible, it is not practical. The threaded outside diameter (OD) makes a dye penetrant examination on the lower section of the penetration impractical because of excessive bleed out from the threads. Eddy current examination would similarly not be effective due to the threaded configuration. Additionally, the radiation levels under the reactor vessel head are estimated to be 10,000 millirem (mRem)/hr at the bottom of the CRDM nozzles resulting in an exposure of approximately 2500 mRem per nozzle. These dose rates are consistent with dose rates measured at The Seabrook Station during 2006, as reported in a Seabrook Station request for relief dated October 27, 2011 (Reference 7). The reactor vessel head configuration at Seabrook Station is similar that at WCGS. At Seabrook Station, radiation levels under the reactor vessel head were measured during their previous inspection in 2006 and ranged from 7000 mRem/hr to 10,000 mRem/hr at the bottom of the CRDM nozzles, resulting in an exposure of approximately 1750 to 2500 mRem per nozzle to perform surface examination.
Therefore, no alternative is proposed for the two CRDM nozzles with limited examination coverage below the J-groove weld.
- 6. Duration of Proposed Alternative The alternative requirements of this request will be applied for the remaining duration of the current 3 rd 10-year Inservice Inspection (ISI) Interval.
- 7. Precedents Similar relief requests have been granted to the following plants:
- NRC Safety Evaluation dated December 22, 2009, for San Onofre Nuclear Generating Station, Units 2 and 3, "Relief Request ISI-3-29, Request for Relief from Inspection Requirements of ASME Code Case N-729-1 for Control Element Drive Mechanism Penetrations (TAC Nos. ME0768 and ME0769)"
(ML093441035)
- NRC Safety Evaluation dated March 3, 2011, for Braidwood Station Units 1 and 2, and Byron Station Units I and 2, "Relief Request from ASME Code Case N-729-1 Requirements for Examination of Reactor Vessel Head Penetration Welds (TAC Nos. ME3510, ME3511, ME3512 and ME3513)" (ML110590921)
Attachment to ET 12-0010 Page 7 of 10
- 8. References
- 1. ASME Code Case N-729-1, "Alternative Examination Requirements for PWR Reactor Vessel Upper Heads with Nozzles Having Pressure-Retaining Partial-Penetration Welds,Section XI, Division 1," March 28, 2006.
- 2. WCAP-16589-P, Revision 0, "Structural Integrity Evaluation of Reactor Vessel Upper Head Penetrations to Support Continued Operation: Wolf Creek," August 2006.
- 3. WCNOC letter ET 06-0035 from T. J. Garrett, WCNOC, to USNRC, "Relaxation Request from the First Revised NRC Order EA-03-009 Regarding Requirements for Nondestructive Examination of Nozzles Below the J-Groove," October 5, 2006.
- 4. WCNOC letter ET 06-0048 from T. J. Garrett, WCNOC, to USNRC, "Additional Information Related to the First Revised NRC Order EA-03-009 Regarding Requirements for Nondestructive Examination of Nozzles Below the J-Groove,"
November 1, 2006.
- 5. NRC letter from D. Terao, USNRC, to R. A. Muench, WCNOC, "Wolf Creek Generating Station - Request to Relax Nondestructive Examination of Reactor Pressure Vessel Head Penetration Nozzles in First Revised Order EA-03-009 (TAC NO. MD3210)," December 7, 2006.
- 6. NRC letter EA-03-009, "Issuance Of First Revised NRC Order (EA-03-009)
Establishing Interim Inspection Requirements For Reactor Pressure Vessel Heads At Pressurized Water Reactors," February 20, 2004.
- 7. Nextera Energy letter SBK-L-11192, "Seabrook Station Request for Relief from ASME Code Case N-729-1 Requirements for Examination of Reactor Vessel Head Penetration Welds," October 27, 2011.
Attachment to ET 12-0010 Page 8 of 10 Figure 1 WCGS Reactor Vessel Head Penetration Ends
.203- 007 SEE DETAL "D" (D-TYPE "X" END TYPE "Y" END Details of the threaded and tapered portions of Penetrations 74, 75, 76, 77, and 78 referred to as "Type X" and the end of Penetrations 1 through 73, referred to as "Type Y."
Attachment to ET 12-0010 Page 9 of 10 Figure 2 Hoop Stress Distribution Downhill Side (48.7* CRDM Penetration Nozzle) 80.OOD 70,000 Ii, a 60.000 50.000 40,000 ------------------------
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Attachment to ET 12-0010 Page 10 of 10 Figure 3 Crack Growth Prediction for WCGS for Through-Wall Longitudinal Flaws Located in the 48.70 Row of CRDM Penetrations, Downhill Side
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