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| {{#Wiki_filter:January 18, 2006Mr. Dennis L. KoehlSite Vice President Point Beach Nuclear Plant Nuclear Management Company, LLC 6590 Nuclear Road Two Rivers, WI 54241-9516 | | {{#Wiki_filter:January 18, 2006 Mr. Dennis L. Koehl Site Vice President Point Beach Nuclear Plant Nuclear Management Company, LLC 6590 Nuclear Road Two Rivers, WI 54241-9516 |
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| ==SUBJECT:== | | ==SUBJECT:== |
| POINT BEACH NUCLEAR PLANT, UNIT 1 - REACTOR VESSEL CLOSUREHEAD PENETRATION FLAW CHARACTERIZATION RELIEF REQUEST MR 02-018-2, REVISION 2 (TAC NO. MC7482) | | POINT BEACH NUCLEAR PLANT, UNIT 1 - REACTOR VESSEL CLOSURE HEAD PENETRATION FLAW CHARACTERIZATION RELIEF REQUEST MR 02-018-2, REVISION 2 (TAC NO. MC7482) |
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| ==Dear Mr. Koehl:== | | ==Dear Mr. Koehl:== |
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| By letter to the Nuclear Regulatory Commission (NRC) dated July 1, 2005, the NuclearManagement Company, LLC (NMC), submitted Relief Request MR 02-018-2, Revision 2 for Point Beach Nuclear Plant (PBNP), Unit 1. The relief request pertains to relief from the requirement to characterize flaws that may exist in the remnants of the control rod drivemechanism nozzle J-groove welds after the repair of a reactor vessel head penetration. Relief Request MR 02-018-2 was originally authorized on September 10, 2003. In its safety evaluation (SE) of the initial relief request, the NRC staff limited its authorization to cases wherethere was no overlap of the new Alloy 52 weld material onto any portion of the remnant J-groove weld. MR 02-018-2, Revision 1, submitted on May 13, 2004, intended by NMC to provide the technical basis for eliminating this restriction, was authorized by the NRC staff on July 16, 2004. By letter dated July 1, 2005, NMC submitted Revision 2 to Relief Request MR-02-018-2, specifically to correct an error in the reactor pressure vessel (RPV) upper head temperature for PBNP, Unit 1. The NRC staff has completed its review of Relief Request MR-02-018-2, Revision 2 asdocumented in the enclosed SE. Our SE concluded that your proposed alternative provides anacceptable level of quality and safety. Therefore, the licensee's proposed alternative is authorized pursuant to 10 CFR 50.55a(a)(3)(i) for the remainder of the operati ng cycl e #29 atPBNP, Unit 1, which ended on November 24, 2005. All other American Society of Mechanical Engineers Boiler and Pressure Vessel Code, Section XI requirements for which relief was notspecifically requested and approved in this relief request remain applicable, including third party review by the Authorized Nuclear Inservice Inspector. | | By letter to the Nuclear Regulatory Commission (NRC) dated July 1, 2005, the Nuclear Management Company, LLC (NMC), submitted Relief Request MR 02-018-2, Revision 2 for Point Beach Nuclear Plant (PBNP), Unit 1. The relief request pertains to relief from the requirement to characterize flaws that may exist in the remnants of the control rod drive mechanism nozzle J-groove welds after the repair of a reactor vessel head penetration. |
| D. Koehl- 2 -If you have any questions concerning this matter, please contact Mr. F. Lyon of my staff at (301) 415-2296.Sincerely, /RA/Timothy Kobetz, Acting ChiefPlant Licensing Branch III-1 Division of Operating Reactor Licensing Office of Nuclear Reactor RegulationDocket No. 50-266 | | Relief Request MR 02-018-2 was originally authorized on September 10, 2003. In its safety evaluation (SE) of the initial relief request, the NRC staff limited its authorization to cases where there was no overlap of the new Alloy 52 weld material onto any portion of the remnant J-groove weld. MR 02-018-2, Revision 1, submitted on May 13, 2004, intended by NMC to provide the technical basis for eliminating this restriction, was authorized by the NRC staff on July 16, 2004. By letter dated July 1, 2005, NMC submitted Revision 2 to Relief Request MR-02-018-2, specifically to correct an error in the reactor pressure vessel (RPV) upper head temperature for PBNP, Unit 1. |
| | The NRC staff has completed its review of Relief Request MR-02-018-2, Revision 2 as documented in the enclosed SE. Our SE concluded that your proposed alternative provides an acceptable level of quality and safety. Therefore, the licensee's proposed alternative is authorized pursuant to 10 CFR 50.55a(a)(3)(i) for the remainder of the operating cycle #29 at PBNP, Unit 1, which ended on November 24, 2005. All other American Society of Mechanical Engineers Boiler and Pressure Vessel 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. |
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| | D. Koehl If you have any questions concerning this matter, please contact Mr. F. Lyon of my staff at (301) 415-2296. |
| | Sincerely, |
| | /RA/ |
| | Timothy Kobetz, Acting Chief Plant Licensing Branch III-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-266 |
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| ==Enclosure:== | | ==Enclosure:== |
| As stated cc w/encl: See next page D. Koehl- 2 -If you have any questions concerning this matter, please contact Mr. F. Lyon of my staff at (301) 415-2296.Sincerely, Timothy Kobetz, Acting ChiefPlant Licensing Branch III-1 Division of Operating Reactor Licensing Office of Nuclear Reactor RegulationDocket No. 50-266 | | As stated cc w/encl: See next page |
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| | D. Koehl If you have any questions concerning this matter, please contact Mr. F. Lyon of my staff at (301) 415-2296. |
| | Sincerely, Timothy Kobetz, Acting Chief Plant Licensing Branch III-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-266 |
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| ==Enclosure:== | | ==Enclosure:== |
| As stated cc w/encl: See next pageDISTRIBUTION | | As stated cc w/encl: See next page DISTRIBUTION: |
| :PUBLICRidsOgcRpTSteingass LPLIII-1 R/FRidsAcrsAcnwMailCenterRidsNrrDciCpnbRidsNrrDorlLplERidsNrrPMFLyon RidsRgn3MailCenterRidsNrrLADClarkeDORL DPRAMuniz DWeaverADAMS Accession Number: ML052700197*memo date 8/12/05
| | PUBLIC RidsOgcRp TSteingass LPLIII-1 R/F RidsAcrsAcnwMailCenter RidsNrrDciCpnb RidsNrrDorlLplE RidsNrrPMFLyon RidsRgn3MailCenter RidsNrrLADClarke DORL DPR AMuniz DWeaver ADAMS Accession Number: ML052700197 |
| **previously concurredNRR-028OFFICELPLIII-1/PELPLIII-1/PMLPLIII-1/LACPNB/BCOGCLPLIII-1/BC (A)NAMEAMunizCLyonDClarke**TChan*JZorn**TKobetzDATE1/13/061/17/0612/29/058/12/051/5/061/18/06OFFICIAL RECORD COPY Point Beach Nuclear Plant, Unit 1 cc: | | *memo date 8/12/05 |
| Jonathan Rogoff, EsquireVice President, Counsel & Secretary Nuclear Management Company, LLC 700 First Street Hudson, WI 54016Mr. F. D. KuesterPresident & Chief Executive Officer WE Generation 231 West Michigan Street Milwaukee, WI 53201Regulatory Affairs ManagerPoint Beach Nuclear Plant Nuclear Management Company, LLC 6610 Nuclear Road Two Rivers, WI 54241Mr. Ken DuveneckTown Chairman Town of Two Creeks 13017 State Highway 42 Mishicot, WI 54228ChairmanPublic Service Commission of Wisconsin P.O. Box 7854 Madison, WI 53707-7854Regional Administrator, Region IIIU.S. Nuclear Regulatory Commission 801 Warrenville RoadLisle, IL 60532-4351Resident Inspector's OfficeU.S. Nuclear Regulatory Commission 6612 Nuclear Road Two Rivers, WI 54241Mr. Jeffery KitsembelElectric Division Public Service Commission of Wisconsin P.O. Box 7854 Madison, WI 53707-7854Nuclear Asset ManagerWisconsin Electric Power Company 231 West Michigan Street Milwaukee, WI 53201Michael B. SellmanPresident and Chief Executive Officer Nuclear Management Company, LLC 700 First Street Hudson, MI 54016Douglas E. CooperSenior Vice President - Group Operations Palisades Nuclear Plant Nuclear Management Company, LLC 27780 Blue Star Memorial Highway Covert, MI 49043Site Director of OperationsNuclear Management Company, LLC 6610 Nuclear Road Two Rivers, WI 54241 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATIONRELIEF REQUEST MR 02-018-2, REVISION 2 POINT BEACH NUCLEAR STATION, UNIT 1NUCLEAR MANAGEMENT COMPANY, LLCDOCKET NO. 50-266 | | **previously concurred NRR-028 OFFICE LPLIII-1/PE LPLIII-1/PM LPLIII-1/LA CPNB/BC OGC LPLIII-1/BC (A) |
| | NAME AMuniz CLyon DClarke** TChan* JZorn** TKobetz DATE 1/13/06 1/17/06 12/29/05 8/12/05 1/5/06 1/18/06 OFFICIAL RECORD COPY |
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| | Point Beach Nuclear Plant, Unit 1 cc: |
| | Jonathan Rogoff, Esquire Mr. Jeffery Kitsembel Vice President, Counsel & Secretary Electric Division Nuclear Management Company, LLC Public Service Commission of Wisconsin 700 First Street P.O. Box 7854 Hudson, WI 54016 Madison, WI 53707-7854 Mr. F. D. Kuester Nuclear Asset Manager President & Chief Executive Officer Wisconsin Electric Power Company WE Generation 231 West Michigan Street 231 West Michigan Street Milwaukee, WI 53201 Milwaukee, WI 53201 Michael B. Sellman Regulatory Affairs Manager President and Chief Executive Officer Point Beach Nuclear Plant Nuclear Management Company, LLC Nuclear Management Company, LLC 700 First Street 6610 Nuclear Road Hudson, MI 54016 Two Rivers, WI 54241 Douglas E. Cooper Mr. Ken Duveneck Senior Vice President - Group Operations Town Chairman Palisades Nuclear Plant Town of Two Creeks Nuclear Management Company, LLC 13017 State Highway 42 27780 Blue Star Memorial Highway Mishicot, WI 54228 Covert, MI 49043 Chairman Site Director of Operations Public Service Commission Nuclear Management Company, LLC of Wisconsin 6610 Nuclear Road P.O. Box 7854 Two Rivers, WI 54241 Madison, WI 53707-7854 Regional Administrator, Region III U.S. Nuclear Regulatory Commission 801 Warrenville Road Lisle, IL 60532-4351 Resident Inspector's Office U.S. Nuclear Regulatory Commission 6612 Nuclear Road Two Rivers, WI 54241 |
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| | SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELIEF REQUEST MR 02-018-2, REVISION 2 POINT BEACH NUCLEAR STATION, UNIT 1 NUCLEAR MANAGEMENT COMPANY, LLC DOCKET NO. 50-266 |
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| | ==1.0 INTRODUCTION== |
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| | By letter to the Nuclear Regulatory Commission (NRC, Commission) dated July 1, 2005 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML051940343), Nuclear Management Company, LLC (NMC, the licensee), submitted Relief Request MR-02-018-2, Revision 2, to correct an error in the reactor pressure vessel (RPV) upper head operating temperature for Point Beach Nuclear Plant (PBNP), Unit 1. Relief Request MR 02-018-2 was originally authorized on September 10, 2003 (ADAMS Accession No. ML032310402). In its safety evaluation (SE) of the initial relief request, the NRC staff limited its authorization to cases where there was no overlap of the new Alloy 52 weld material onto any portion of the remnant J-groove weld. MR 02-018-2, Revision 1, submitted on May 13, 2004 (ADAMS Accession No. ML041410464), intended by NMC to provide the technical basis for eliminating this restriction, was authorized by the NRC staff on July 16, 2004 (ADAMS Accession No. ML041760065). By letter dated July 1, 2005, NMC submitted Revision 2 to its Relief Request MR-02-018-2, specifically to correct an error in the RPV upper head temperature for PBNP, Unit 1. Therefore, this SE will address only the effects of the increase in RPV head operating temperature on the NRC SE dated July 16, 2004. |
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| | ==2.0 REGULATORY EVALUATION== |
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| | Alternatives to requirements may be authorized or relief granted by the NRC pursuant to Title 10 of Code of Federal Regulations (10 CFR) 50.55a(a)(3)(i), 10 CFR 50.55a(a)(3)(ii), or 10 CFR 50.55a(g)(6)(i). In proposing alternatives or requesting relief, the licensee must demonstrate that: (1) the proposed alternatives provide an acceptable level of safety; or (2) compliance would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety; or (3) conformance is impractical for the facility. Pursuant to 10 CFR 50.55a(g)(4)(iv), inservice inspection of items may meet the requirements set forth in subsequent editions and addenda of the American Society of Mechanical Engineers (ASME) |
| | Boiler and Pressure Vessel Code (Code) that are incorporated by reference in 10 CFR 50.55a(b), subject to the limitations and modifications listed therein, and subject to Commission approval. Portions of editions or addenda may be used provided that all related requirements of the respective editions or addenda are met. In its letter dated July 1, 2005, pursuant to 10 CFR 50.55a(a)(3)(i), the licensee requested a revision to the relief authorized by the NRC staffs SE dated July 16, 2004, from the requirements of the 1998 Edition of the ASME Code, Section XI, IWA-3300(b), IWB-3142.4, and IWB-3420, which require characterization of flaw(s) existing in the remnant of the J-groove weld(s) that will remain in service for PBNP, Unit 1, |
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| | reactor vessel closure heads if a control rod drive mechanism nozzle must be partially removed and a new pressure boundary deposited over a portion of the J-groove weld remnant. |
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| ==1.0 INTRODUCTION== | | ==3.0 TECHNICAL EVALUATION== |
| By letter to the Nuclear Regulatory Commission (NRC, Commission) dated July 1, 2005(Agencywide Documents Access and Management System (ADAMS) Accession No. ML051940343), Nuclear Management Company, LLC (NMC, the licensee), submitted Relief Request MR-02-018-2, Revision 2, to correct an error in the reactor pressure vessel (RPV) upper head operating temperature for Point Beach Nuclear Plant (PBNP), Unit 1. Relief Request MR 02-018-2 was originally authorized on September 10, 2003 (ADAMS Accession No. ML032310402). In its safety evaluation (SE) of the initial relief request, the NRC stafflimited its authorization to cases where there was no overlap of the new Alloy 52 weld material onto any portion of the remnant J-groove weld. MR 02-018-2, Revision 1, submitted on May 13, 2004 (ADAMS Accession No. ML041410464), intended by NMC to provide the technical basis for eliminating this restriction, was authorized by the NRC staff on July 16, 2004(ADAMS Accession No. ML041760065). By letter dated July 1, 2005, NMC submitted Revision 2 to its Relief Request MR-02-018-2, specifically to correct an error in the RPV upper head temperature for PBNP, Unit 1. Therefore, this SE will address only the effects of theincrease in RPV head operating temperature on the NRC SE dated July 16, 2004.
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| ==2.0 REGULATORY EVALUATION==
| | The NRC staff notes the revised calculations submitted by the licensee show a slightly higher crack growth rate due to the higher operating temperature, resulting in a change to the calculated operating time from 1.39 effective full power years (EFPY) to 1.31 EFPY. This equates to a change from 16.68 months to 15.72 months for a net result of 0.96 month shorter duration of time until a hypothetical flaw reaches the toe of the J-groove weld in the weld overlap region. The conservative assumptions in the revised calculations indicate that an actual flaw would require more than 1.31 EFPY to grow through the J-groove weld. PBNP, Unit 1 is not expected to accumulate greater than 1.41 EFPY prior to the plant shutdown due to operational constraints. The delta in RPV operating temperatures is sufficiently low to not challenge the conservative assumptions in the calculations that formed part of the basis for NRC authorization of the current and previous revisions of this relief request. The NRC staff concludes that the previously granted reliefs remain in effect for this repair situation. |
| Alternatives to requirements may be authorized or relief granted by the NRC purs uant to Title 10 of Code of Federal Regulations (10 CFR) 50.55a(a)(3)(i), 10 CFR 50.55a(a)(3)(ii), or 10 CFR 50.55a(g)(6)(i). In proposing alternatives or requesting relief, the licensee must demonstrate that: (1) the proposed alternatives provide an acceptable level of safety; or (2) compliance would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety; or (3) conformance is impractical for the facility. Purs uant to 10 CFR 50.55a(g)(4)(iv), inservice inspection of items may meet the requirements set forth in subsequent editions and addenda of the American Society of Mechanical Engineers (ASME)
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| Boiler and Pressure Vessel Code (Code) that are incorporated by reference in 10 CFR50.55a(b), subject to the limitations and modifications listed therein, and subject to Commission approval. Portions of editions or addenda may be used provided that all related requirements of the respective editions or addenda are met. In its letter dated July 1, 2005, pursuant to 10 CFR 50.55a(a)(3)(i), the licensee requested a revision to the relief authorized by the NRCstaff's SE dated July 16, 2004, from the requirements of the 1998 Edition of the ASME Code,Section XI, IWA-3300(b), IWB-3142.4, and IWB-3420, which require characterization of flaw(s) existing in the remnant of the J-groove weld(s) that will remain in service for PBNP, Unit 1, reactor vessel closure heads if a control rod drive mechanism nozzle must be partially removedand a new pressure boundary deposited over a portion of the J-groove weld remnant.
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| ==3.0 TECHNICAL EVALUATION== | | ==4.0 CONCLUSION== |
| The NRC staff notes the revised calculations submitted by the licensee show a slightly hi ghercrack growth rate due to the higher operating temperature, resulting in a change to the calculated operating time from 1.39 effective full power years (EFPY) to 1.31 EFPY. This equates to a change from 16.68 months to 15.72 months for a net result of 0.96 month shorterduration of time until a hypothetical flaw reaches the toe of the J-groove weld in the weldoverlap region. The conservative assumptions in the revised calculations indicate that an actual flaw would require more than 1.31 EFPY to grow through the J-groove weld. PBNP, Unit 1 is not expected to accumulate greater than 1.41 EFPY prior to the plant shutdown due tooperational constraints. The delta in RPV operating temperatures is sufficiently low to not challenge the conservative assumptions in the calculations that formed part of the basis for NRC authorization of the current and previous revisions of this relief request. The NRC staffconcludes that the previously granted reliefs remain in effect for this repair situation.
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| ==4.0 CONCLUSION==
| | The NRC staff concludes that Relief Request MR-02-018-2, Revision 2 is acceptable because the proposed alternative authorized in the previous revisions has not changed, the delta in RPV operating temperatures is sufficiently low to not challenge the conservative assumptions in the calculations made to support continued safe operation, and the removal from service of the RPV head during the Autumn 2005 outage. Based on the information provided in the licensee's submittal, the NRC staff concludes that the alternative proposed in MR-02-018-2, Revision 2 provides an acceptable level of quality and safety. Therefore, the licensee's proposed alternative is authorized pursuant to 10 CFR 50.55a(a)(3)(i) for the remainder of operating cycle |
| The NRC staff concludes that Relief Request MR-02-018-2, Revision 2 is acceptable becausethe proposed alternative authorized in the previous revisions has not changed, the delta in RPVoperating temperatures is sufficiently low to not challenge the conservative assumptions in the calculations made to support continued safe operation, and the removal from service of the RPV head during the Autumn 2005 outage. Based on the information provided in the licensee'ssubmittal, the NRC staff concludes that the alternative proposed in MR-02-018-2, Revision 2provides an acceptable level of quality and safety. Therefore, the licensee's proposed alternative is authorized pursuant to 10 CFR 50.55a(a)(3)(i) for the remainder of operating cycle#29 at PBNP, Unit 1, which ended on November 24, 2005.All other ASME Code, Section XI requirements for which relief was not specifically requestedand approved in this relief request remain applicable, including third party review by the Authorized Nuclear Inservice Inspector.Principal Contributor: T. Steingass Date: January 19, 2006}} | | #29 at PBNP, Unit 1, which ended on November 24, 2005. |
| | 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. |
| | Principal Contributor: T. Steingass Date: January 19, 2006}} |
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Category:Code Relief or Alternative
MONTHYEARML24068A2492024-04-22022 April 2024
– Authorization and Safety Evaluation for Alternative Request No. I6-RR-01
ML24071A0912024-04-22022 April 2024
Issuance of Relief Request I6-RR-03 - Extension of the Unit 2 Steam Generator Primary Nozzle Dissimilar Metal Welds Sixth 10-Year Inservice Inspection Program Interval
ML23279A0672023-11-0909 November 2023
Issuance of Relief Request I6 RR 02 - Examination of the Unit 2 Steam Generator Feedwater Nozzle Extension to Nozzle Weld Sixth 10 Year Inservice Inspection Program Interval
ML20036F2612020-03-0404 March 2020
Approval of Relief Request 1-RR-13 and 2-RR-13 Regarding Extension of Inspection Interval for Point Beach Unit 1 and Unit 2 Reactor Pressure Vessel Welds from 10 to 20 Years
ML19339H7472019-12-13013 December 2019
Approval of Relief Request 2-RR-17 Regarding Steam Generator Primary Nozzle Dissimilar Metal Welds Inspection Interval
ML18106B1212018-04-25025 April 2018
Proposed Alternative to Use Encoded Phased Array Ultrasonic Examination Techniques
L-2017-121, Proposed Alternative for the Use of Encoded Phased Array Ultrasonic Examination Techniques in Lieu of Radiography2017-07-24024 July 2017
Proposed Alternative for the Use of Encoded Phased Array Ultrasonic Examination Techniques in Lieu of Radiography
ML16330A1182016-12-15015 December 2016
NextEra Fleet - Safety Evaluation for Proposed Alternative to the American Society of Mechanical Engineers Operation and Maintenance Code by Adoption of Approved Code Case OMN-20, Inservice Test Frequency (CAC Nos. MF8195 Through MF8201)
ML16063A0582016-03-22022 March 2016
Approval of Relief Request 2-RR-11; Steam Generator Nozzle to Safe-End Dissimilar Metal (DM) Weld Inspection
ML15246A3052015-09-16016 September 2015
Evaluation of Relief Request RR-10 - Examination of Feedwater Nozzle Extension to Nozzle Weld Fifth 10-Year Inservice Program Interval
ML15127A2912015-05-20020 May 2015
Relief Request RR-8, Relief from the Requirements of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code for Examination of Buried Components
NRC 2015-0025, Requests Relief from Performing Inservice Testing (ISI) of Relief Valve 1CC-00763B2015-05-14014 May 2015
Requests Relief from Performing Inservice Testing (ISI) of Relief Valve 1CC-00763B
ML15099A0182015-05-0707 May 2015
Relief Request RR-9, Proposed Alternative from the Requirements of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code for System Leakage Test
ML14343A0512014-12-10010 December 2014
Relief from the Requirements of the ASME Code for Re-Examination of the Reactor Pressure Vessel a Inlet Nozzle Weld for the Fifth Ten-Year Inservice Inspection Program Interval
ML13329A0312013-12-20020 December 2013
Relief from the Requirements of ASME B&PV Code, Section XI, for the Fourth 10-Year ISI Interval (RR-4L1)
ML13079A1412013-03-19019 March 2013
CFR 50.55a Request, Relief Request RR-4L3 Inservice Inspection Impracticality Examination Limitations Due to Configuration Fourth Ten-Year Inservice Inspection Program Interval
NRC 2013-0020, CFR 50.55a Request, Relief Request RR-4L3 Inservice Inspection Impracticality Examination Limitations Due to Configuration Fourth Ten-Year Inservice Inspection Program Interval2013-03-19019 March 2013
CFR 50.55a Request, Relief Request RR-4L3 Inservice Inspection Impracticality Examination Limitations Due to Configuration Fourth Ten-Year Inservice Inspection Program Interval
ML13064A4252013-03-18018 March 2013
Relief Request 1-RR-4 Re-Examination of the Unit 1 RPV Indication on the a Inlet Nozzle Weld
ML12286A1042012-11-15015 November 2012
Evaluation of Relief Requests RR-2 & RR-3 (ME7974 & ME7975)
ML0617103642006-07-0303 July 2006
Monticello Nuclear Generating Plant, Palisades Nuclear Plant, Point Beach Nuclear Plant Units 1 and 2, Prairie Island Nuclear Generating Plant, Units 1 and 2 - Use of ASME Code Case N-513-2
ML0527001972006-01-18018 January 2006
Relief Request - Reactor Vessel Closure Head Penetration Flaw Characterization Relief Request MR 02-018-2, Revision 2
ML0526503122005-09-27027 September 2005
Relief Requests - the Previsions of ASME Section XI, IWA-5244, Buried Components, RR-1-26 and RR-2-34
NRC 2005-0084, Reactor Vessel Closure Head Penetration Flaw Characterization Relief Request MR 02-018-2, Revision 22005-07-0101 July 2005
Reactor Vessel Closure Head Penetration Flaw Characterization Relief Request MR 02-018-2, Revision 2
NRC 2005-0016, Relief Request from the Provisions of ASME Section Xl, IWA-4422.2.2, Defect Removal Followed by Welding or Brazing, Relief Request 162005-02-0404 February 2005
Relief Request from the Provisions of ASME Section Xl, IWA-4422.2.2, Defect Removal Followed by Welding or Brazing, Relief Request 16
NRC 2005-0015, Relief Request from the Provisions of ASME Section XI, IWA-5244, Buried Components, Relief Request 152005-01-21021 January 2005
Relief Request from the Provisions of ASME Section XI, IWA-5244, Buried Components, Relief Request 15
ML0408601612004-04-0101 April 2004
Request for Relief VRR 03-01 on a One Time Basis for Performing Inservice Testing of Relief Valve 1RH-861C
ML0401607382004-02-26026 February 2004
Prairie, Units 1 and 2, Kewaunee, Point Beach, Units 1 and 2, Palisades, Re Request for Alternatives to ASME Section XI, Appendix Viii, Supplement 10
ML0323104022003-09-10010 September 2003
Relief, MR 02 018-2 Pertaining to Reactor Vessel Closure Head Penetration Repair
ML0306203282003-03-21021 March 2003
Relief, Use of Code Case N-600 for the Fourth 10-Year Interval, MB5403 and MB5404
ML0225401092003-03-21021 March 2003
Relief Request No 8 - Requirement for Scheduling of Components for Examination
ML0302101262003-02-27027 February 2003
Relief, Alternative to Examine All Three Vessels of the Regenerative Heat Exchanger, MB5401 & MB5402
ML0225300232002-10-0808 October 2002
Relief, Granted for ASME Code Section XI, Relief Request 6 Regarding Evaluation of Leakage with Bolting In-Place
ML0225300062002-10-0808 October 2002
Code Relief, ASME Code Section XI, Relief Request 5 Regarding Visual Examination of Insulted Bolting on Borated Systems
NRC 2002-0073, Reactor Vessel Closure Head Penetration Repair Relief Requests MR 02-018-1 and MR 02-018-22002-08-28028 August 2002
Reactor Vessel Closure Head Penetration Repair Relief Requests MR 02-018-1 and MR 02-018-2
2024-04-22
[Table view]
Category:Letter
MONTHYEARIR 05000266/20240032024-11-0505 November 2024
Integrated Inspection Report 05000266/2024003 and 05000301/2024003
ML24295A1142024-10-31031 October 2024
Suppl Env Audit Summary Letter W/Enclosure 1
L-2024-176, Annual 10 CFR 140.21 Licensee Guarantees of Payment of Deferred Premiums2024-10-30030 October 2024
Annual 10 CFR 140.21 Licensee Guarantees of Payment of Deferred Premiums
ML24295A0862024-10-21021 October 2024
Notification of NRC Baseline Inspection and Request for Information Inspection Report 0500026/2025002
L-2024-160, Core Operating Limits Report (COLR) Unit 2 Reload Cycle 41 (U2C41)2024-10-21021 October 2024
Core Operating Limits Report (COLR) Unit 2 Reload Cycle 41 (U2C41)
L-2024-169, Supplement to License Amendment Request to Adopt Common Emergency Plan with Site- Specific Annexes2024-10-15015 October 2024
Supplement to License Amendment Request to Adopt Common Emergency Plan with Site- Specific Annexes
L-2024-118, Fleet License Amendment Request to Relocate Staff Qualifications from Technical Specifications to the Quality Assurance Topical Report (FPL-1)2024-10-0808 October 2024
Fleet License Amendment Request to Relocate Staff Qualifications from Technical Specifications to the Quality Assurance Topical Report (FPL-1)
L-2024-120, LAR 301 - Revise Renewed Facility Operating Licenses to Adopt an Alternative Approach for Addressing Seismic Risk in 10 CFR 50.69 Risk-Informed Categorization2024-10-0808 October 2024
LAR 301 - Revise Renewed Facility Operating Licenses to Adopt an Alternative Approach for Addressing Seismic Risk in 10 CFR 50.69 Risk-Informed Categorization
L-2024-158, Supplemental Information to License Amendment Request to Adopt Common Emergency Plan with Site-Specific Annexes2024-09-25025 September 2024
Supplemental Information to License Amendment Request to Adopt Common Emergency Plan with Site-Specific Annexes
IR 05000266/20240112024-09-18018 September 2024
Biennial Problem Identification and Resolution Inspection Report 05000266/2024011 and 05000301/2024011
L-2024-136, Supplement to License Amendment Request to Adopt Common Emergency Plan with Site-Specific Annexes2024-09-16016 September 2024
Supplement to License Amendment Request to Adopt Common Emergency Plan with Site-Specific Annexes
L-2024-145, Notification of Deviation from EPRI MRP-227, Revision 1-A Baffle Bolt Inspection Frequency2024-09-0909 September 2024
Notification of Deviation from EPRI MRP-227, Revision 1-A Baffle Bolt Inspection Frequency
ML24207A0202024-08-28028 August 2024
Response to Request for Re-Engagement Regarding the Subsequent License Renewal Environmental Review for Point Beach Nuclear Plant, Units 1 and 2 (Docket Numbers: 50-0266 and 50-0301)
IR 05000266/20240052024-08-21021 August 2024
Updated Inspection Plan for Point Beach Nuclear Plant, Units 1 and 2 (Report 05000266/2024005 and 05000301/2024005)
IR 05000266/20240022024-08-13013 August 2024
Integrated Inspection Report 05000266/2024002 and 05000301/2024002
L-2024-131, Response to Request for Additional Information Regarding License Amendment Request 300, Modify Containment Average Air Temperature Requirements2024-08-0909 August 2024
Response to Request for Additional Information Regarding License Amendment Request 300, Modify Containment Average Air Temperature Requirements
ML24163A0012024-08-0505 August 2024
LTR-24-0119-1-1 Response to Nh Letter Regarding Review of NextEras Emergency Preparedness Amendment Review
ML24214A3092024-08-0202 August 2024
Confirmation of Initial License Examination
L-2024-113, License Amendment Request 294, Application to Revise Technical Specifications to Adopt TSTF- 577, Revised Frequencies for Steam Generator Tube Inspections2024-07-24024 July 2024
License Amendment Request 294, Application to Revise Technical Specifications to Adopt TSTF- 577, Revised Frequencies for Steam Generator Tube Inspections
ML24194A1802024-07-24024 July 2024
– Revision to the Reactor Vessel Material Surveillance Capsule Withdrawal Schedule
L-2024-125, Notice of Intent to Provide Supplemental Information to License Amendment Request to Adopt Common Emergency Plan with Site-Specific Annexes2024-07-24024 July 2024
Notice of Intent to Provide Supplemental Information to License Amendment Request to Adopt Common Emergency Plan with Site-Specific Annexes
IR 05000266/20244012024-07-23023 July 2024
Public - Point Beach Nuclear Plant Cyber Security Inspection Report 05000266/2024401 and 05000301/2024401
ML24193A2432024-07-12012 July 2024
– Interim Audit Summary Report in Support of Review of License Amendment Requests Regarding Fleet Emergency Plan
L-2024-116, Preparation and Scheduling of Operator Licensing Examinations2024-07-11011 July 2024
Preparation and Scheduling of Operator Licensing Examinations
L-2024-114, Quality Assurance Topical Report (FPL-1 Revision 31 Annual Submittal2024-07-10010 July 2024
Quality Assurance Topical Report (FPL-1 Revision 31 Annual Submittal
IR 05000266/20240102024-07-10010 July 2024
Age-Related Degradation Inspection Report 05000266/2024010 and 05000301/2024010
L-2024-105, License Amendment Request 300, Modify Containment Average Air Temperature Requirements2024-06-26026 June 2024
License Amendment Request 300, Modify Containment Average Air Temperature Requirements
L-2024-107, Schedule for Subsequent License Renewal Environmental Review2024-06-25025 June 2024
Schedule for Subsequent License Renewal Environmental Review
ML24176A2242024-06-24024 June 2024
Notification of an NRC Biennial Licensed Operator Requalification Program Inspection and Request for Information
L-2024-102, Official Service List Update2024-06-19019 June 2024
Official Service List Update
ML24149A2862024-06-12012 June 2024
NextEra Fleet - Proposed Alternative Frr 23-01 to Use ASME Code Case N-752-1, Risk-Informed Categorization and Treatment for Repair/Replacement Activities in Class 2 and 3 Systems Section X1, Division 1 (EPID L-2023-LLR-0009) - Letter
L-2024-093, Steam Generator Divider Plate Assemblies Bounding Analysis Evaluation for Aging Management Commitment 14 Revision2024-06-10010 June 2024
Steam Generator Divider Plate Assemblies Bounding Analysis Evaluation for Aging Management Commitment 14 Revision
IR 05000266/20244202024-06-0505 June 2024
Security Baseline Inspection Report 05000266/2024420 and 05000301/2024420
ML24149A1922024-05-28028 May 2024
Notification of NRC Baseline Inspection and Request for Information
ML24141A1382024-05-20020 May 2024
Information Request to Support the NRC Annual Baseline Emergency Action Level and Emergency Plan Changes Inspection.Docx
ML24127A0632024-05-0606 May 2024
Supplemental Information to License Amendment Request to Adopt Common Emergency Plan with Site-Specific Annexes
IR 05000266/20243012024-04-29029 April 2024
NRC Initial License Examination Report 05000266/2024301; 05000301/2024301
L-2024-067, Annual Monitoring Report2024-04-26026 April 2024
Annual Monitoring Report
ML24116A0402024-04-23023 April 2024
Periodic Update of the Updated Final Safety Analysis Report
ML24071A0912024-04-22022 April 2024
Issuance of Relief Request I6-RR-03 - Extension of the Unit 2 Steam Generator Primary Nozzle Dissimilar Metal Welds Sixth 10-Year Inservice Inspection Program Interval
IR 05000266/20240012024-04-11011 April 2024
Integrated Inspection Report 05000266/2024001 and 05000301/2024001
L-2024-030, Supplement to Relief Request CISl-03-01 for Relief Concerning Containment Unbonded Post-Tensioning System Inservice Inspection Requirements2024-03-27027 March 2024
Supplement to Relief Request CISl-03-01 for Relief Concerning Containment Unbonded Post-Tensioning System Inservice Inspection Requirements
L-2024-043, Revised Reactor Vessel Materials Surveillance Capsule Withdrawal Schedules2024-03-25025 March 2024
Revised Reactor Vessel Materials Surveillance Capsule Withdrawal Schedules
L-2024-011, And Point Beach, Units 1 and 2 - 10 CFR 50.46 Annual Reporting of Changes to, or Errors in Emergency Core Cooling System Models or Applications2024-03-13013 March 2024
And Point Beach, Units 1 and 2 - 10 CFR 50.46 Annual Reporting of Changes to, or Errors in Emergency Core Cooling System Models or Applications
IR 05000266/20230062024-02-28028 February 2024
Annual Assessment Letter for Point Beach Nuclear Plant, Units 1 and 2 (Report 05000266/2023006 and 05000301/2023006)
ML24053A3732024-02-22022 February 2024
Operator Licensing Examination Approval Point Beach, March 2024
L-2024-020, Refueling Outage Owners Activity Report (OAR-1) Unit 1 for Inservice Inspections2024-02-22022 February 2024
Refueling Outage Owners Activity Report (OAR-1) Unit 1 for Inservice Inspections
ML24036A2652024-02-0505 February 2024
Notice of Inspection and Request for Information for the NRC Age-Related Degradation Inspection: Inspection Report 05000266/2024010 and 05000301/2024010
IR 05000266/20230042024-02-0101 February 2024
Integrated Inspection Report 05000266/2023004 and 05000301/2023004
ML24030A0352024-01-30030 January 2024
Information Request for the Cyber-Security Baseline Inspection, Notification to Perform Inspection
2024-09-09
[Table view]
Category:Safety Evaluation
MONTHYEARML24194A1802024-07-24024 July 2024
– Revision to the Reactor Vessel Material Surveillance Capsule Withdrawal Schedule
ML24200A1612024-07-19019 July 2024
Alternative CISI-03-01
ML24149A2862024-06-12012 June 2024
NextEra Fleet - Proposed Alternative Frr 23-01 to Use ASME Code Case N-752-1, Risk-Informed Categorization and Treatment for Repair/Replacement Activities in Class 2 and 3 Systems Section X1, Division 1 (EPID L-2023-LLR-0009) - Letter
ML24068A2492024-04-22022 April 2024
– Authorization and Safety Evaluation for Alternative Request No. I6-RR-01
ML23352A2752024-01-23023 January 2024
Issuance of Amendment Nos. 274 and 276 Regarding Revision to Technical Specification 5.5.17, Pre-Stressed Concrete Containment Tendon Surveillance Program
ML23279A0672023-11-0909 November 2023
Issuance of Relief Request I6 RR 02 - Examination of the Unit 2 Steam Generator Feedwater Nozzle Extension to Nozzle Weld Sixth 10 Year Inservice Inspection Program Interval
ML23208A0952023-08-28028 August 2023
Issuance of Amendment Nos. 273 and 275 Regarding Revising Licensing Basis to Address Generic Safety Issue-191 and Respond to Generic Letter 2004-02 Using a Risk Informed Approach
ML23160A0642023-08-21021 August 2023
Issuance of Amendment Nos. 272 and 274 Regarding Revision to Use Beacon Power Distribution Monitoring System
ML23103A1332023-06-0101 June 2023
Issuance of Amendment Nos. 271 and 273 Regarding Technical Specifications to Adopt TSTF-505, Revision 2, Provide Risk-Informed Extended Completion Times - RITSTF Initiative 4b
ML22193A1142022-09-12012 September 2022
Issuance of Amendment Nos. 270 and 272 Elimination of the Requirements to Maintain the Post-Accident Sampling System
ML22140A1272022-05-25025 May 2022
Subsequent License Renewal Application Safety Evaluation Revision 1 Public
ML22041A3342022-02-23023 February 2022
Transmittal Letter for Point Beach Final SE for SLRA Review to AA La 2-9 (3)
ML22054A1082022-02-23023 February 2022
Subsequent License Renewal Application Safety Evaluation Public
ML21148A2552021-07-21021 July 2021
Issuance of Amendment Nos. 269 and 271 Technical Specification Changes to Implement New Surveillance Methods for Transient Heat Flux Hot Channel Factor
ML20363A1762021-02-23023 February 2021
Issuance of Amendment Nos. 268 and 270 Regarding Tornado Missile Protection Licensing Basis
ML20241A0582020-09-25025 September 2020
Issuance of Amendment No. 267 for One-Time Extension of License Condition 4.I, Containment Building Construction Truss (EPID L-2020-LLA-0180 (COVID-19))
ML20036F2612020-03-0404 March 2020
Approval of Relief Request 1-RR-13 and 2-RR-13 Regarding Extension of Inspection Interval for Point Beach Unit 1 and Unit 2 Reactor Pressure Vessel Welds from 10 to 20 Years
ML19357A1952020-02-10010 February 2020
Unit No.1; & Turkey Point Nuclear Generating Unit Nos. 3 & 4 - Issuance of Amendments Nos. 265, 268, 164, 290, and 284 Revise Technical Specifications to Adopt TSTF-563
ML20015A1232020-02-0606 February 2020
Proposed Alternative to Use Encoded Phased Array Ultrasonic Examination Techniques
ML19339H7472019-12-13013 December 2019
Approval of Relief Request 2-RR-17 Regarding Steam Generator Primary Nozzle Dissimilar Metal Welds Inspection Interval
ML19064A9042019-04-25025 April 2019
Issuance of Amendments to Extend Containment Leakage Rate Test Frequency
ML19052A5442019-03-27027 March 2019
Issuance of Amendments 264 and 267 to Adopt TSTF-547, Clarification of Rod Position Requirements
ML18289A3782018-11-26026 November 2018
Issuance of Amendments to Adopt Title 10 of Code of Federal Regulations 50.69, Risk Informed Categorization and Treatment of Structures, Systems and Components for Nuclear Power Reactors
ML18079A0452018-06-13013 June 2018
Issuance of Amendments Revision to the Point Beach Nuclear Plant Emergency Action Level Scheme (CAC Nos. MF9859 and MF9860 EPID L-2017-LLS-0278)
ML18106B1212018-04-25025 April 2018
Proposed Alternative to Use Encoded Phased Array Ultrasonic Examination Techniques
ML17159A7782017-07-27027 July 2017
Issuance of Amendment to Approve H*: Alternate Repair Criteria for Steam Generator Tube Sheet Expansion Region
ML17027A0782017-04-0707 April 2017
Issuance of Amendments Regarding Technical Specifications for Inservice Testing Programs (CAC Nos. MF8202 Through MF8209)
ML17039A3002017-02-22022 February 2017
Issuance of Amendments -Removal of Completed License Conditions and Changes to the Ventilation Filter Testing Program
ML16330A1182016-12-15015 December 2016
NextEra Fleet - Safety Evaluation for Proposed Alternative to the American Society of Mechanical Engineers Operation and Maintenance Code by Adoption of Approved Code Case OMN-20, Inservice Test Frequency (CAC Nos. MF8195 Through MF8201)
ML16241A0002016-09-23023 September 2016
Mitigating Strategies and Spent Fuel Pool Instrumentation Safety Evaluation
ML16196A0932016-09-0808 September 2016
Issuance of Amendment Regarding Transition to a Risk-Informed, Performance-Based Fire Protection Program in Accordance with 10 CFR 50.48 (C)
ML16118A1542016-06-17017 June 2016
Issuance of Amendments
ML16063A0582016-03-22022 March 2016
Approval of Relief Request 2-RR-11; Steam Generator Nozzle to Safe-End Dissimilar Metal (DM) Weld Inspection
ML16035A5092016-03-0909 March 2016
Correction of Typographical Error in Safety Evaluation Associated with License Amendment Nos. 238 and 242
ML15293A4572015-11-25025 November 2015
Issuance of Amendments for the Steam Generator Technical Specifications, to Reflect Adoption of TSTF-510
ML15246A3052015-09-16016 September 2015
Evaluation of Relief Request RR-10 - Examination of Feedwater Nozzle Extension to Nozzle Weld Fifth 10-Year Inservice Program Interval
ML15195A2012015-07-28028 July 2015
Issuance of Amendments Regarding Relocation of Surveillance Frequencies to Licensee Control
ML15155A5392015-07-14014 July 2015
Issuance of Amendments Concerning Extension of Cyber Security Plan Milestone 8
ML15161A5352015-06-24024 June 2015
Relief Request VR-01; Alternatives to Certain Inservice Testing Requirements of the American Society of Mechanical Engineers (ASME) Code of Operation and Maintenance of Nuclear Power Plants
ML15127A2912015-05-20020 May 2015
Relief Request RR-8, Relief from the Requirements of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code for Examination of Buried Components
ML15099A0182015-05-0707 May 2015
Relief Request RR-9, Proposed Alternative from the Requirements of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code for System Leakage Test
ML15014A2492015-01-27027 January 2015
Issuance of Amendments to Revise Technical Specifications to Adopt Technical Specifications Task Force - 523, Generic Letter 2008-01, Managing Gas Accumulation (Tac Nos. MF4353 and MF4354)
ML14343A0512014-12-10010 December 2014
Relief from the Requirements of the ASME Code for Re-Examination of the Reactor Pressure Vessel a Inlet Nozzle Weld for the Fifth Ten-Year Inservice Inspection Program Interval
ML14293A0022014-10-21021 October 2014
Issuance of Safety Evaluation Regarding Relief Request RR-5
ML14126A3782014-06-30030 June 2014
Issuance of License Amendment Nos. 250 and 254 Regarding Change to Technical Specification 5.6.5, Reactor Coolant System Pressure and Temperature Limits Report
ML14058B0292014-05-0909 May 2014
Issuance of Amendment Nos. 249 and 253 Regarding Use of Optimized Zirlo Fuel Cladding Material
ML14014A2052014-01-30030 January 2014
Issuance of Relief Request Regarding Risk-Informed Inservice Inspection Program for the Fifth 10-Year Inservice Inspection Interval
ML13329A0422013-12-20020 December 2013
Relief from the Requirements of ASME Code, Section XI, for the Fourth 10-Year Inservice Inspection Interval (RR-4L3)
ML13329A0312013-12-20020 December 2013
Relief from the Requirements of ASME B&PV Code, Section XI, for the Fourth 10-Year ISI Interval (RR-4L1)
ML13346A0402013-12-18018 December 2013
Relief from the Requirements of ASME Code, Section XI, for the Fourth 10-Year Inservice Inspection Interval (RR-4L2)
2024-07-24
[Table view] |
Text
January 18, 2006 Mr. Dennis L. Koehl Site Vice President Point Beach Nuclear Plant Nuclear Management Company, LLC 6590 Nuclear Road Two Rivers, WI 54241-9516
SUBJECT:
POINT BEACH NUCLEAR PLANT, UNIT 1 - REACTOR VESSEL CLOSURE HEAD PENETRATION FLAW CHARACTERIZATION RELIEF REQUEST MR 02-018-2, REVISION 2 (TAC NO. MC7482)
Dear Mr. Koehl:
By letter to the Nuclear Regulatory Commission (NRC) dated July 1, 2005, the Nuclear Management Company, LLC (NMC), submitted Relief Request MR 02-018-2, Revision 2 for Point Beach Nuclear Plant (PBNP), Unit 1. The relief request pertains to relief from the requirement to characterize flaws that may exist in the remnants of the control rod drive mechanism nozzle J-groove welds after the repair of a reactor vessel head penetration.
Relief Request MR 02-018-2 was originally authorized on September 10, 2003. In its safety evaluation (SE) of the initial relief request, the NRC staff limited its authorization to cases where there was no overlap of the new Alloy 52 weld material onto any portion of the remnant J-groove weld. MR 02-018-2, Revision 1, submitted on May 13, 2004, intended by NMC to provide the technical basis for eliminating this restriction, was authorized by the NRC staff on July 16, 2004. By letter dated July 1, 2005, NMC submitted Revision 2 to Relief Request MR-02-018-2, specifically to correct an error in the reactor pressure vessel (RPV) upper head temperature for PBNP, Unit 1.
The NRC staff has completed its review of Relief Request MR-02-018-2, Revision 2 as documented in the enclosed SE. Our SE concluded that your proposed alternative provides an acceptable level of quality and safety. Therefore, the licensee's proposed alternative is authorized pursuant to 10 CFR 50.55a(a)(3)(i) for the remainder of the operating cycle #29 at PBNP, Unit 1, which ended on November 24, 2005. All other American Society of Mechanical Engineers Boiler and Pressure Vessel 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.
D. Koehl If you have any questions concerning this matter, please contact Mr. F. Lyon of my staff at (301) 415-2296.
Sincerely,
/RA/
Timothy Kobetz, Acting Chief Plant Licensing Branch III-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-266
Enclosure:
As stated cc w/encl: See next page
D. Koehl If you have any questions concerning this matter, please contact Mr. F. Lyon of my staff at (301) 415-2296.
Sincerely, Timothy Kobetz, Acting Chief Plant Licensing Branch III-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-266
Enclosure:
As stated cc w/encl: See next page DISTRIBUTION:
PUBLIC RidsOgcRp TSteingass LPLIII-1 R/F RidsAcrsAcnwMailCenter RidsNrrDciCpnb RidsNrrDorlLplE RidsNrrPMFLyon RidsRgn3MailCenter RidsNrrLADClarke DORL DPR AMuniz DWeaver ADAMS Accession Number: ML052700197
- previously concurred NRR-028 OFFICE LPLIII-1/PE LPLIII-1/PM LPLIII-1/LA CPNB/BC OGC LPLIII-1/BC (A)
NAME AMuniz CLyon DClarke** TChan* JZorn** TKobetz DATE 1/13/06 1/17/06 12/29/05 8/12/05 1/5/06 1/18/06 OFFICIAL RECORD COPY
Point Beach Nuclear Plant, Unit 1 cc:
Jonathan Rogoff, Esquire Mr. Jeffery Kitsembel Vice President, Counsel & Secretary Electric Division Nuclear Management Company, LLC Public Service Commission of Wisconsin 700 First Street P.O. Box 7854 Hudson, WI 54016 Madison, WI 53707-7854 Mr. F. D. Kuester Nuclear Asset Manager President & Chief Executive Officer Wisconsin Electric Power Company WE Generation 231 West Michigan Street 231 West Michigan Street Milwaukee, WI 53201 Milwaukee, WI 53201 Michael B. Sellman Regulatory Affairs Manager President and Chief Executive Officer Point Beach Nuclear Plant Nuclear Management Company, LLC Nuclear Management Company, LLC 700 First Street 6610 Nuclear Road Hudson, MI 54016 Two Rivers, WI 54241 Douglas E. Cooper Mr. Ken Duveneck Senior Vice President - Group Operations Town Chairman Palisades Nuclear Plant Town of Two Creeks Nuclear Management Company, LLC 13017 State Highway 42 27780 Blue Star Memorial Highway Mishicot, WI 54228 Covert, MI 49043 Chairman Site Director of Operations Public Service Commission Nuclear Management Company, LLC of Wisconsin 6610 Nuclear Road P.O. Box 7854 Two Rivers, WI 54241 Madison, WI 53707-7854 Regional Administrator, Region III U.S. Nuclear Regulatory Commission 801 Warrenville Road Lisle, IL 60532-4351 Resident Inspector's Office U.S. Nuclear Regulatory Commission 6612 Nuclear Road Two Rivers, WI 54241
SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELIEF REQUEST MR 02-018-2, REVISION 2 POINT BEACH NUCLEAR STATION, UNIT 1 NUCLEAR MANAGEMENT COMPANY, LLC DOCKET NO. 50-266
1.0 INTRODUCTION
By letter to the Nuclear Regulatory Commission (NRC, Commission) dated July 1, 2005 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML051940343), Nuclear Management Company, LLC (NMC, the licensee), submitted Relief Request MR-02-018-2, Revision 2, to correct an error in the reactor pressure vessel (RPV) upper head operating temperature for Point Beach Nuclear Plant (PBNP), Unit 1. Relief Request MR 02-018-2 was originally authorized on September 10, 2003 (ADAMS Accession No. ML032310402). In its safety evaluation (SE) of the initial relief request, the NRC staff limited its authorization to cases where there was no overlap of the new Alloy 52 weld material onto any portion of the remnant J-groove weld. MR 02-018-2, Revision 1, submitted on May 13, 2004 (ADAMS Accession No. ML041410464), intended by NMC to provide the technical basis for eliminating this restriction, was authorized by the NRC staff on July 16, 2004 (ADAMS Accession No. ML041760065). By letter dated July 1, 2005, NMC submitted Revision 2 to its Relief Request MR-02-018-2, specifically to correct an error in the RPV upper head temperature for PBNP, Unit 1. Therefore, this SE will address only the effects of the increase in RPV head operating temperature on the NRC SE dated July 16, 2004.
2.0 REGULATORY EVALUATION
Alternatives to requirements may be authorized or relief granted by the NRC pursuant to Title 10 of Code of Federal Regulations (10 CFR) 50.55a(a)(3)(i), 10 CFR 50.55a(a)(3)(ii), or 10 CFR 50.55a(g)(6)(i). In proposing alternatives or requesting relief, the licensee must demonstrate that: (1) the proposed alternatives provide an acceptable level of safety; or (2) compliance would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety; or (3) conformance is impractical for the facility. Pursuant to 10 CFR 50.55a(g)(4)(iv), inservice inspection of items may meet the requirements set forth in subsequent editions and addenda of the American Society of Mechanical Engineers (ASME)
Boiler and Pressure Vessel Code (Code) that are incorporated by reference in 10 CFR 50.55a(b), subject to the limitations and modifications listed therein, and subject to Commission approval. Portions of editions or addenda may be used provided that all related requirements of the respective editions or addenda are met. In its letter dated July 1, 2005, pursuant to 10 CFR 50.55a(a)(3)(i), the licensee requested a revision to the relief authorized by the NRC staffs SE dated July 16, 2004, from the requirements of the 1998 Edition of the ASME Code,Section XI, IWA-3300(b), IWB-3142.4, and IWB-3420, which require characterization of flaw(s) existing in the remnant of the J-groove weld(s) that will remain in service for PBNP, Unit 1,
reactor vessel closure heads if a control rod drive mechanism nozzle must be partially removed and a new pressure boundary deposited over a portion of the J-groove weld remnant.
3.0 TECHNICAL EVALUATION
The NRC staff notes the revised calculations submitted by the licensee show a slightly higher crack growth rate due to the higher operating temperature, resulting in a change to the calculated operating time from 1.39 effective full power years (EFPY) to 1.31 EFPY. This equates to a change from 16.68 months to 15.72 months for a net result of 0.96 month shorter duration of time until a hypothetical flaw reaches the toe of the J-groove weld in the weld overlap region. The conservative assumptions in the revised calculations indicate that an actual flaw would require more than 1.31 EFPY to grow through the J-groove weld. PBNP, Unit 1 is not expected to accumulate greater than 1.41 EFPY prior to the plant shutdown due to operational constraints. The delta in RPV operating temperatures is sufficiently low to not challenge the conservative assumptions in the calculations that formed part of the basis for NRC authorization of the current and previous revisions of this relief request. The NRC staff concludes that the previously granted reliefs remain in effect for this repair situation.
4.0 CONCLUSION
The NRC staff concludes that Relief Request MR-02-018-2, Revision 2 is acceptable because the proposed alternative authorized in the previous revisions has not changed, the delta in RPV operating temperatures is sufficiently low to not challenge the conservative assumptions in the calculations made to support continued safe operation, and the removal from service of the RPV head during the Autumn 2005 outage. Based on the information provided in the licensee's submittal, the NRC staff concludes that the alternative proposed in MR-02-018-2, Revision 2 provides an acceptable level of quality and safety. Therefore, the licensee's proposed alternative is authorized pursuant to 10 CFR 50.55a(a)(3)(i) for the remainder of operating cycle
- 29 at PBNP, Unit 1, which ended on November 24, 2005.
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.
Principal Contributor: T. Steingass Date: January 19, 2006
