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Category:Code Relief or Alternative
MONTHYEARML23312A0882023-11-0808 November 2023 Acceptance of Request to Revise Alternatives 1-RR-5-10 and 2-RR-5-10 ML23130A3872023-05-18018 May 2023 Request 1-RR-5-15 and 2-RR-5-15 to Use Later Edition of ASME Section XI Code for Performance of Repair/Replacement Activities ML22270A3252022-09-30030 September 2022 (Prairie Island), Unit 2 - Authorization and Safety Evaluation for Alternative Request No. RR-08 ML20276A0032020-10-0202 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 ML20135G9922020-05-29029 May 2020 and Prairie Island Nuclear Generating Plant, Units 1 and 2 - Relief from the Requirements of the ASME Code ML19282A5412019-11-0505 November 2019 Relief from the Requirements of the ASME Code L-PI-19-002, 10 CFR 50.55a Requests Nos. 1-RR-5-10 and 2-RR-5-10, Proposed Alternative to Reactor Vessel Inservice Inspection (ISI) Intervals for Prairie Island, Unit 1 and Unit 22019-06-13013 June 2019 10 CFR 50.55a Requests Nos. 1-RR-5-10 and 2-RR-5-10, Proposed Alternative to Reactor Vessel Inservice Inspection (ISI) Intervals for Prairie Island, Unit 1 and Unit 2 ML19046A1662019-02-25025 February 2019 Relief from the Requirements of the ASME Code ML18270A0152018-10-19019 October 2018 and Prairie Island Nuclear Generating Plant, Units 1 and 2 - Relief from the Requirements of the ASME OM Code ML16246A2002016-10-0404 October 2016 Request for Relief 1RR-4-11 and 2-RR-4-11 Associated with Certain Inservice Inspection of Component Welds for the Fourth 10-Year Interval ML15196A2132015-08-17017 August 2015 Requests 1-RR-4-9 and 2-RR-4-9 Associated with the Fourth Ten-Year Interval for the Inservice Inspection Program ML15196A2272015-08-12012 August 2015 Requests 1-RR-5-3 and 2-RR-5-3 Associated with the Fifth Ten-Year Interval for the Inservice Inspection Program ML15196A2212015-08-12012 August 2015 Requests 1-RR-5-1 and 2-RR-5-1 Associated with the Fifth Ten-Year Interval for the Inservice Inspection Program ML15079A0032015-05-0404 May 2015 Relief Requests 1-RR-5-5 and 2-RR-5-5 for Fifth 10-Year Interval for the Inservice Inspection Program ML15079A0022015-05-0404 May 2015 Relief Requests 1-RR-5-2 and 2-RR-5-2 for Fifth 10-Year Interval for the Inservice Inspection Program ML14329A1852014-12-0505 December 2014 Relief Requests for Fifth 10-Year Inservice Testing Program Interval (Tac Nos. MF3928 and MF3929) L-PI-14-117, Withdrawal of 10 CFR 50.55a Request RR-01 Associated with the Fifth Ten-Year Interval for the Inservice Test Program (TAC Nos. MF3928 and Mf 3929)2014-11-24024 November 2014 Withdrawal of 10 CFR 50.55a Request RR-01 Associated with the Fifth Ten-Year Interval for the Inservice Test Program (TAC Nos. MF3928 and Mf 3929) ML0807905632008-05-0909 May 2008 Relief Request 1-RR-4-6 for Piping Weld Examination Coverage for the Fourth Inservice Inspection Interval 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 ML0617200152006-07-0303 July 2006 Relief, Evaluation of Relief Request 2-RR-4-6 for Reduced Examination Volume for Class 2 RHR Heat Exchanger Shell- to-Flange Weld L-HU-05-028, ASME Section XI, ISI Request for Relief 21 & 222005-12-19019 December 2005 ASME Section XI, ISI Request for Relief 21 & 22 L-PI-05-038, Withdrawal of 4th Ten-Year Inspection Interval Lnservice Lnspection Program Relief Request No. 1-RR-4-12005-05-0505 May 2005 Withdrawal of 4th Ten-Year Inspection Interval Lnservice Lnspection Program Relief Request No. 1-RR-4-1 ML0509601872005-04-27027 April 2005 Evaluation of Relief Request Nos. 1-RR-4-2, 1-RR-4-3, 2-RR-4-3, 1-RR-4-4, and 2-RR-4-4 for the Fourth 10-Year Inservice Inspection Interval ML0507504282005-04-0707 April 2005 Evaluation of Relief Request to Use Code Case N-661. TAC MC3883 and TAC MC3884 ML0422200422004-10-18018 October 2004 Evaluation of Relief Request No. 16 for the Unit 2 3rd 10-year Interval Inservice Inspection Program ML0418902032004-07-27027 July 2004 Code Relief, Evaluation of Relief Request No. 20, Revision 1 - for the Third 10-Year Inservice Inspection Interval ML0411301922004-05-0303 May 2004 Relief, Limited Examinations Associated with the PINGP Unit 1, Third 10-year Inservice Inspection (ISI) Interval, MB7975 ML0409802342004-04-12012 April 2004 Prairie, Units 1 and 2, Relief Request Nos. 19 and 20 Associated with the 10-Year Interval Inservice Inspection Interval L-PI-04-044, Request for Relief No. 20, Revision 0, for Units 1 & 2, 3rd Ten Year Inservice Inspection Interval2004-03-30030 March 2004 Request for Relief No. 20, Revision 0, for Units 1 & 2, 3rd Ten Year Inservice Inspection Interval 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 ML0332806022004-01-13013 January 2004 Prairie, Units 1 and 2, Relief, No. 15, Evaluation of Relief Request No. 15 for the Third 10-Year Interval Inservice Inspection Program ML0323301462003-10-0202 October 2003 Relief, Re the Fourth and Successive Inservice Inspection and Inservice Testing Program Intervals ML0321300802003-08-13013 August 2003 Relief, Re Using VT-1 Visual Examinations During Third 10-Year Interval Inservice Inspection Program ML0320506712003-07-30030 July 2003 Evaluation of Relief Request No. 18 to Perform a Visual Examination in Lieu of Volumetric Examination of Reactor Vessel Nozzle Inner Radius Sections Per Code Case N-648-1, MB8363 and MB8364 ML0316402462003-06-17017 June 2003 Relief Request, Third 10-Year Interval Inservice Inspection Program ML0232905782002-11-16016 November 2002 Request for Relief No. 9 for the Unit 2 Third 10-Year Interval Inservice Inspection Program ML0230102092002-11-0707 November 2002 Relief, Third 10-Year Interval Inservice Inspection Program ML0226202392002-10-0101 October 2002 Relief, Evaluation of Relief Request Associated with the Third 10-Year Interval Inservice Inspection Program (MB5388, 5399, 5390, & 5391) ML0212904282002-06-11011 June 2002 Relief Request, Related to the First Interval Inservice Inspection Program for Metal Containment ML0216105082002-05-31031 May 2002 Request for Relief No. 12 for the Unit 2 3rd 10-Year Interval Inservice Inspection Program 2023-05-18
[Table view] Category:Letter type:L
MONTHYEARL-PI-23-034, Response to Request for Additional Information Regarding License Amendment Request to Revise Technical Specification 3.7.8, Cooling Water (Cl) System,2024-01-0202 January 2024 Response to Request for Additional Information Regarding License Amendment Request to Revise Technical Specification 3.7.8, Cooling Water (Cl) System, L-PI-23-035, Response to Request for Additional Information Regarding License Amendment Request to Revise Technical Specifications Section 5.6.6, Reactor Coolant System (RCS) Pressure and Temperature Limits Report2023-12-20020 December 2023 Response to Request for Additional Information Regarding License Amendment Request to Revise Technical Specifications Section 5.6.6, Reactor Coolant System (RCS) Pressure and Temperature Limits Report L-PI-23-033, Supplement to License Amendment Request to Revise Technical Specification 3.8.1, Surveillance Requirement 3.8.1.2, Note 32023-12-0505 December 2023 Supplement to License Amendment Request to Revise Technical Specification 3.8.1, Surveillance Requirement 3.8.1.2, Note 3 L-PI-23-025, License Amendment Request to Revise Technical Specification 3.8.1, Surveillance Requirement 3.8.1.2, Note 32023-09-28028 September 2023 License Amendment Request to Revise Technical Specification 3.8.1, Surveillance Requirement 3.8.1.2, Note 3 L-PI-23-023, Baffle Former Bolts Alternate Aging Management Strategy2023-09-11011 September 2023 Baffle Former Bolts Alternate Aging Management Strategy L-PI-23-018, License Amendment Request to Revise ISFSI Technical Specification 4.4 to Allow Use of a Code Alternative to ASME Code, NB-5130, Examination of Weld Edge Preparation Surfaces, for the TN-40HT2023-07-14014 July 2023 License Amendment Request to Revise ISFSI Technical Specification 4.4 to Allow Use of a Code Alternative to ASME Code, NB-5130, Examination of Weld Edge Preparation Surfaces, for the TN-40HT L-PI-23-006, License Amendment Request to Revise Technical Specification 3.7.8 Required Actions2023-06-22022 June 2023 License Amendment Request to Revise Technical Specification 3.7.8 Required Actions L-PI-23-016, 2022 10 CFR 50.46 LOCA Annual Report2023-06-14014 June 2023 2022 10 CFR 50.46 LOCA Annual Report L-PI-23-010, Annual Report of Individual Monitoring2023-04-27027 April 2023 Annual Report of Individual Monitoring L-PI-23-007, Supplement to Application to Revise Technical Specification Section 5.6.6, Reactor Coolant System (RCS) Pressure and Temperature Limits Report (PTLR)2023-03-28028 March 2023 Supplement to Application to Revise Technical Specification Section 5.6.6, Reactor Coolant System (RCS) Pressure and Temperature Limits Report (PTLR) L-PI-23-005, CFR 50.55a Requests Nos. 1-RR-5-15 and 2-RR-5-15, Proposed Use of Subsequent ASME Code Edition and Addenda in Accordance with 10 CFR 50.55a(g)(4)(iv)2023-03-0303 March 2023 CFR 50.55a Requests Nos. 1-RR-5-15 and 2-RR-5-15, Proposed Use of Subsequent ASME Code Edition and Addenda in Accordance with 10 CFR 50.55a(g)(4)(iv) L-PI-23-001, Day Steam Generator Tube Inspection Report2023-01-30030 January 2023 Day Steam Generator Tube Inspection Report L-PI-22-047, Resubmittal of Prairie Island Nuclear Generating Plant (PINGP) 2018 Unit 1 180-Day Steam Generator Tube Inspection Report2022-12-21021 December 2022 Resubmittal of Prairie Island Nuclear Generating Plant (PINGP) 2018 Unit 1 180-Day Steam Generator Tube Inspection Report L-PI-22-020, Application to Revise Technical Specification Section 5.6.6, Reactor Coolant System (RCS) Pressure and Temperature Limits Report (PTLR)2022-12-0202 December 2022 Application to Revise Technical Specification Section 5.6.6, Reactor Coolant System (RCS) Pressure and Temperature Limits Report (PTLR) L-PI-22-040, Supplement to Application to Revise Technical Specifications to Adopt TSTF-577, Revised Frequencies for Steam Generator Tube Inspections2022-10-0606 October 2022 Supplement to Application to Revise Technical Specifications to Adopt TSTF-577, Revised Frequencies for Steam Generator Tube Inspections L-PI-22-037, Updated Approach for Prairie Island Unit 1 and Unit 2 Baffle Former Bolts2022-09-20020 September 2022 Updated Approach for Prairie Island Unit 1 and Unit 2 Baffle Former Bolts L-PI-22-032, CFR 50.46 LOCA Annual Report2022-06-16016 June 2022 CFR 50.46 LOCA Annual Report L-PI-22-033, Response to Request for Additional Information Prairie Island Nuclear Generating Plant, Units 1 and 2, Amendment to Adopt 24-Month Operating Cycles2022-06-10010 June 2022 Response to Request for Additional Information Prairie Island Nuclear Generating Plant, Units 1 and 2, Amendment to Adopt 24-Month Operating Cycles L-PI-22-003, Application to Revise Technical Specifications to Adopt TSTF-577, Revised Frequencies for Steam Generator Tube Inspections2022-06-0707 June 2022 Application to Revise Technical Specifications to Adopt TSTF-577, Revised Frequencies for Steam Generator Tube Inspections L-PI-22-024, Supplement to Application for License Amendment to Implement 24-Month Operating Cycle2022-03-0707 March 2022 Supplement to Application for License Amendment to Implement 24-Month Operating Cycle L-PI-21-047, Response to Request for Additional Information 24-Month Cycle Amendment Prairie Island Nuclear Generating Plant, Units 1 and 22021-12-0707 December 2021 Response to Request for Additional Information 24-Month Cycle Amendment Prairie Island Nuclear Generating Plant, Units 1 and 2 L-PI-21-045, Response to Request for Additional Information Cooling Water System License Amendment Request2021-11-0404 November 2021 Response to Request for Additional Information Cooling Water System License Amendment Request L-PI-21-029, Exigent License Amendment Request to Revise Technical Specification 3.7.8 to Allow a One-Time Extension of the Completion Time of Required Action B.12021-10-0707 October 2021 Exigent License Amendment Request to Revise Technical Specification 3.7.8 to Allow a One-Time Extension of the Completion Time of Required Action B.1 L-PI-21-006, License Amendment Request to Revise Technical Specification 3.3.1, Reactor Trip System (RTS) Instrumentation, to Incorporate Installed Bypass Test Capability for the Power Range RTS Functions2021-10-0202 October 2021 License Amendment Request to Revise Technical Specification 3.3.1, Reactor Trip System (RTS) Instrumentation, to Incorporate Installed Bypass Test Capability for the Power Range RTS Functions L-PI-21-032, Response to Request for Additional Information Amendment Request to Adopt TSTF-471 and 571-T for Prairie Island2021-09-30030 September 2021 Response to Request for Additional Information Amendment Request to Adopt TSTF-471 and 571-T for Prairie Island L-PI-21-016, Application for License Amendment to Implement 24-Month Operating Cycle2021-08-0606 August 2021 Application for License Amendment to Implement 24-Month Operating Cycle L-PI-21-027, 2020 10 CFR 50.46 LOCA Annual Report2021-06-28028 June 2021 2020 10 CFR 50.46 LOCA Annual Report L-PI-21-023, Independent Spent Fuel Storage Installation - 2020 Annual Radiological Environmental Monitoring Program Report2021-05-14014 May 2021 Independent Spent Fuel Storage Installation - 2020 Annual Radiological Environmental Monitoring Program Report L-PI-21-007, Application to Revise Technical Specifications to Adopt TSTF-471, Eliminate Use of the Term Core Alterations in Actions and Notes2021-04-19019 April 2021 Application to Revise Technical Specifications to Adopt TSTF-471, Eliminate Use of the Term Core Alterations in Actions and Notes L-PI-20-050, Request for a One-Time Exemption from 10 CFR 50, Appendix E, Biennial Emergency Preparedness Exercise Requirements, Due to COVID-19 Pandemic2020-10-0707 October 2020 Request for a One-Time Exemption from 10 CFR 50, Appendix E, Biennial Emergency Preparedness Exercise Requirements, Due to COVID-19 Pandemic L-PI-20-051, Supplement to Application to Revise Technical Specifications to Adopt TSTF-547, Clarification of Rod Position Requirements2020-09-28028 September 2020 Supplement to Application to Revise Technical Specifications to Adopt TSTF-547, Clarification of Rod Position Requirements L-PI-20-026, Response to Request for Additional Information: License Amendment Request to Revise Technical Specifications to Adopt Risk Informed Completion Times TSTF-505, Revision 2, Provide Risk-Informed Extended Completion Times - RITSTF Initiativ2020-09-0101 September 2020 Response to Request for Additional Information: License Amendment Request to Revise Technical Specifications to Adopt Risk Informed Completion Times TSTF-505, Revision 2, Provide Risk-Informed Extended Completion Times - RITSTF Initiative 4 L-PI-20-035, = Request for Revision to Reactor Vessel Material Surveillance Capsule Removal Schedule2020-07-28028 July 2020 = Request for Revision to Reactor Vessel Material Surveillance Capsule Removal Schedule L-PI-20-023, Independent Spent Fuel Storage Installation, Response to Request for Additional Information: License Amendment Request to Expand the Storage Capacity of the Independent Spent Fuel Storage Installation (ISFSI)2020-06-10010 June 2020 Independent Spent Fuel Storage Installation, Response to Request for Additional Information: License Amendment Request to Expand the Storage Capacity of the Independent Spent Fuel Storage Installation (ISFSI) L-PI-20-014, Supplement to License Amendment Request: Expand the Storage Capacity of the Independent Spent Fuel Storage Installation (ISFSI2020-04-29029 April 2020 Supplement to License Amendment Request: Expand the Storage Capacity of the Independent Spent Fuel Storage Installation (ISFSI L-PI-20-004, License Amendment Request to Revise Technical Specifications (TS) to Remove Note I from Limiting Condition for Operating (LCO) 3.4.12 and LCO 3.4.132020-03-30030 March 2020 License Amendment Request to Revise Technical Specifications (TS) to Remove Note I from Limiting Condition for Operating (LCO) 3.4.12 and LCO 3.4.13 L-PI-20-001, License Amendment Request to Address Issues Identified in Westinghouse Nuclear Safety Advisory Letter NSAL-09-5, Revision 1, and NSAL-15-12020-01-29029 January 2020 License Amendment Request to Address Issues Identified in Westinghouse Nuclear Safety Advisory Letter NSAL-09-5, Revision 1, and NSAL-15-1 L-PI-19-041, Application to Revise Technical Specifications to Adopt TSTF-547, Clarification of Rod Position Requirements2019-12-23023 December 2019 Application to Revise Technical Specifications to Adopt TSTF-547, Clarification of Rod Position Requirements L-PI-19-031, License Amendment Request: Revise Technical Specifications to Adopt Risk Informed Completion Times TSTF-505, Revision 2, Provide Risk-Informed Extended Completion Times - RITSTF Initiative 4b2019-12-16016 December 2019 License Amendment Request: Revise Technical Specifications to Adopt Risk Informed Completion Times TSTF-505, Revision 2, Provide Risk-Informed Extended Completion Times - RITSTF Initiative 4b L-PI-19-040, License Amendment Request: Revise Technical Specification 5.5.14 to Permanently Extend Containment Leakage Rate Test Frequency2019-10-0707 October 2019 License Amendment Request: Revise Technical Specification 5.5.14 to Permanently Extend Containment Leakage Rate Test Frequency L-PI-19-038, Submittal of Revised Pressure and Temperature Limits Report2019-09-19019 September 2019 Submittal of Revised Pressure and Temperature Limits Report L-PI-19-037, Response to Request for Additional Information: 10 CFR 50.55a Requests Nos. 1-RR-5-10 and 2-RR-5-10, Proposed Alternative to Reactor Vessel Inservice Inspection (ISI) Intervals2019-09-16016 September 2019 Response to Request for Additional Information: 10 CFR 50.55a Requests Nos. 1-RR-5-10 and 2-RR-5-10, Proposed Alternative to Reactor Vessel Inservice Inspection (ISI) Intervals L-PI-19-025, Request to Approve Site-Specific Probabilistic Risk Assessment (PRA) Model for Flowserve N-Seal Abeyance Seal and Dynamic Testing for the Prairie Island Nuclear Generating Plant (PINGP)2019-08-27027 August 2019 Request to Approve Site-Specific Probabilistic Risk Assessment (PRA) Model for Flowserve N-Seal Abeyance Seal and Dynamic Testing for the Prairie Island Nuclear Generating Plant (PINGP) L-PI-19-029, Supplement to Response to Request for Additional Information: Application to Adopt 10 CFR 50.69, Risk-Informed Categorization and Treatment of Structures, Systems and Components For...2019-08-0505 August 2019 Supplement to Response to Request for Additional Information: Application to Adopt 10 CFR 50.69, Risk-Informed Categorization and Treatment of Structures, Systems and Components For... L-PI-19-002, 10 CFR 50.55a Requests Nos. 1-RR-5-10 and 2-RR-5-10, Proposed Alternative to Reactor Vessel Inservice Inspection (ISI) Intervals for Prairie Island, Unit 1 and Unit 22019-06-13013 June 2019 10 CFR 50.55a Requests Nos. 1-RR-5-10 and 2-RR-5-10, Proposed Alternative to Reactor Vessel Inservice Inspection (ISI) Intervals for Prairie Island, Unit 1 and Unit 2 L-PI-19-014, Response to Request for Additional Information: Application to Adopt 10 CFR 50.69, Risk-Informed Categorization and Treatment of Structures, Systems and Components for Nuclear Power Reactors2019-04-29029 April 2019 Response to Request for Additional Information: Application to Adopt 10 CFR 50.69, Risk-Informed Categorization and Treatment of Structures, Systems and Components for Nuclear Power Reactors L-PI-19-003, Request for Revision to Reactor Vessel Material Surveillance Capsule Removal Schedule for Prairie Island Nuclear Generating Plant (PINGP)2019-02-0404 February 2019 Request for Revision to Reactor Vessel Material Surveillance Capsule Removal Schedule for Prairie Island Nuclear Generating Plant (PINGP) L-PI-19-006, Emergency License Amendment Request Regarding One-Time Extension for Technical Specification Completion Time Requirements2019-01-29029 January 2019 Emergency License Amendment Request Regarding One-Time Extension for Technical Specification Completion Time Requirements L-PI-19-005, Online Reference Portal for NRC Review of License Amendment Request to Implement 10 CFR 50.692019-01-15015 January 2019 Online Reference Portal for NRC Review of License Amendment Request to Implement 10 CFR 50.69 L-PI-18-063, Response to Request for Additional Information: Revise License Condition Associated with Implementation of NFPA 8052018-12-0606 December 2018 Response to Request for Additional Information: Revise License Condition Associated with Implementation of NFPA 805 2024-01-02
[Table view] |
Text
- NMC Committed to Nudcea r
Praire Island Nuclear Generating Plant Operated by Nuclear Management Company, LLC L-PI-04-044 MAR 3 0 203 10 CFR 50.55a U S Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555 Prairie Island Nuclear Generating Plant Docket Nos. 50-282 and 50-306 License Nos. DPR-42 and DPR-60 Request for Relief No. 20, Revision 0, for Units I and 2 3rd Ten Year Inservice Inspection Interval Pursuant to 10 CFR 50.55a(a)(3)(i), Prairie Island Nuclear Generating Plant requests NRC approval of the attached request for the 3rd Ten-Year In-Service Inspection Interval. These requirements are associated with the performance of Reactor Vessel Automated Examinations including compliance with ASME Boiler and Pressure Vessel Code Section Xl Appendix Vil Qualification of Nondestructive Examination Personnel For Ultrasonic Examination.
Relief Request No. 20 addresses the use of 0.189" root mean square error (RMSE) depth sizing requirements in lieu of 0.125" RMSE sizing requirements as required per American Society of Mechanical Engineers Boiler and Pressure Vessel Code Section Xl, Appendix VIII Supplement 10 Paragraph 3.2(b). This relief request provides requests to use an alternative examination requirement to a previously NRC-approved relief request submitted by the Nuclear Management Company (NMC) on September 17, 2003. Prairie Island Nuclear Generating Plant, as part of the NMC, submitted a Relief Request For an "Alternative to ASME Section Xi, Appendix Vil, Supplement 10."
On February 26, 2004, the Nuclear Regulatory Commission approved the Request for Relief for Appendix VIII Supplement 10 for the Prairie Island Nuclear Generating Plant, 3 rd Ten Year Interval (TAC Nos. MC0820 and MC0821). The attached request only addresses the use of a 0.189 RMSE value in lieu of 0.125 RMSE. All other provisions of the previously approved relief request for alternatives to Appendix Vil, Supplement 10 will be followed during the examinations.
The details of the 10 CFR 50.55a(a)(3)(i) request are enclosed in the attached relief request for Prairie Island Unit 1 and Unit 2 (contained in one document). Nuclear Management Company requests approval by September 1, 2004 to support the AjJ4v-1717 Wakonade Drive East
- Welch, Minnesota 55089-9642 Telephone: 651.388.1121
i.: 1 USNRC NUCLEAR MANAGEMENT COMPANY, LLC L-PI-04-044 Page 2 refueling outage on Unit 1. The proposed alternative was approved for the Virgil C.
Summer Nuclear Station by NRC letter dated February 3, 2004 (TAC No. MC0108).
This letter contains no new commitments and no revisions to existing commitments.
Please contact Jack Leveille (651-388-1121, Ext. 4142) if you have any questions related to this letter.
ice Presid t, Yrairie Island Nuclear Generating Plant cc: Regional Administrator, USNRC, Region IlIl Project Manager, Prairie Island Nuclear Generating Plant, USNRC, NRR NRC Resident Inspector- Prairie Island Nuclear Generating Plant Chief Boiler Inspector, State of Minnesota P. Fisher, Hartford Insurance
Attachment:
(one document)
Prairie Island Unit 1 - RELIEF REQUEST NUMBER: 20 (Rev. 0)
Prairie Island Unit 2 - RELIEF REQUEST NUMBER: 20 (Rev. 0) 1717 Wakonade Drive East
- Welch, Minnesota 55089-9642 Telephone: 651.388.1121
Nuclear Management Company Inservice Inspection Prairie Island Unit 1 and Unit 2 3"d Interval Examination Plan Prairie Island Unit 1 - RELIEF REQUEST NUMBER: 20 (Addendum 1)
Prairie Island Unit 2 - RELIEF REQUEST NUMBER: 20 (Addendum 1)
Alternative to Appendix VIII - Supplement 10 SYSTEM/COMPONENT(S) FOR WHICH RELIEF REQUEST WILL BE USED Code Class: Class 1
Reference:
ASME,Section XI, 1989 Edition with no Addenda Examination Category: B-F Item Number: B5.10
Description:
Dissimilar Metal Pressure Retaining Piping Welds subject to examination using procedures, personnel, and equipment qualified to ASME Section XI, Appendix VIII, Supplement 10 criteria.
Component Numbers: See Table I CODE REQUIREMENTS:
A volumetric examination of dissimilar metal pressure retaining piping welds is required per 1989 Edition per ASME Section XI Table IWB-2500-1, Examination Categories B-F, Item No.
B5.10. In addition, 10 CFR 50.55a requires these examinations be performed using procedures, personnel, and equipment qualified to the criteria of ASME Section XI, 1995 Edition, 1996 Addenda, Appendix VIII, Supplement 10.
PROPOSED ALTERNATIVE:
Pursuant to 10 CFR 50.55a(a)(3)(i), relief is requested. As discussed below, this will provide an acceptable level of quality and safety. The inspection vendor to be employed by Prairie Island Nuclear Generating Plant for the remote automated Reactor Vessel examinations did not achieve the 0.125" RMS Appendix VIII, Supplement 10 acceptance tolerance during procedure qualification. Prairie Island Nuclear Generating Plant proposes using the vendor (WESDYNE) achieved sizing error of 0.189" RMS.
The proposed procedure to address sizing of the flaws that may be found during the examination is to add to the measured flaw size the difference between the achieved sizing error and the
'0.125" RMS Appendix VIII Supplement 10 acceptance criteria.
In addition, as a supplementary examination to the volumetric examination, Prairie Island proposes to perform profilometry measurements of the weld inside diameter (ID) profile and to perform a visual inspection (VT-1) of the ID in those areas where an ultrasonic scanning limitation may be encountered.
The Ultrasonic Weld Profiling is a tool to assist in the evaluation of poor transducer contact as a result of surface roughness and as a method to provide permanent record of the actual surface condition of the nozzle to pipe/elbow dissimilar metal weld joint. The data is taken with focused Revision 0 Page 1 of 3 Date 3/30/2004
Nuclear Management Company Inservice Inspection Prairie Island Unit 1 and Unit 2 3rd Interval Examination Plan immersion transducers mounted on the ends of the examination array. The scan stroke is long enough to allow for a surface profile up to 3 inches on either side of the examination volume at approximately 1 degree increments around the nozzle bore. When translated through software, the profile data is presented on a section view scale showing the weld joint and geometry such as root protrusion and counter-bore. Perhaps the most important feature allows the examiner to transpose transducer position along the scan line to evaluate the effect of the geometry on the exam beam To support nozzle bore examinations and provide a confirmation mechanism for surface conditions, a remote enhanced visual examination (EVT) is planned for all four primary nozzles.
The ROS PT-10 cameras are positioned near the ultrasonic end-effector. Lighting conditions, focus and zoom are adjustable and the angular/radial position data is transposed on the visual screen for interpretation/correlation with other data. The EVT is capable of a I-mil wire resolution. Visual data will be recorded directly to DVD.
BASIS FOR REQUESTING RELIEF On September 17, 2003 Prairie Island Nuclear Generating Plant, as part of the Nuclear Management Company, submitted a Relief Request For Alternative to ASME Section XI, Appendix VIII, Supplement 10. On February 26, 2004, the Nuclear Regulatory Commission approved the Request for Relief for Appendix VIII Supplement 10 for the Nuclear Management Company Fleet Relief Request (TAC Nos. MCO820, MC0821, Prairie Island Nuclear Generating Plant, 3' Ten Year Interval).
The 1995 Edition with 1996 Addenda of the ASME Code,Section XI, Appendix VIII Supplement 10, Paragraph 3.2(b), states that the examination procedures, equipment, and personnel are qualified for depth sizing when the RMS (root mean square) error of the flaw depth measurements, as compared to the true flaw depths, is less than or equal to 0.125". An RMS of 0.189" is based on actual vendor demonstrated in-process field qualifications and is the optimum value that could be achieved. Prairie Island Nuclear Generating Plant believes the use of 0.189" RMSE as an adjustment to the measured flaw will ensure a conservative bounding flaw depth value.
Revision 0 Page 2 of 3 Date 3/30/2004
Nuclear Management Company Inservice Inspection Prairie Island Unit 1 and Unit 2 3rd Interval Examination Plan IMPLEMENTATION SCHEDULE:
The proposed alternative is requested for the remainder of the 3rd 10 Year Interval of the Inservice Inspection Program for Prairie Island Unit 1 and Unit 2.
REFERENCE:
By letter dated February 3, 2004, the NRC Staff authorized similar relief for Virgil C. Summer Nuclear Station, Docket No. 50-395 (Relief Request RR-II-20 [TAC No. MC0 108])
Table 1. Dissimilar Metal Pressure Retaining Piping Welds Within Scope of Request
'SI Unit Summary Component Component No. No. Identification Description 1 300518 W-5 Safe End to Nozzle 1 300538 W-5 Safe End to Nozzle 1 300923 W-7 Safe End to Nozzle 2 501130 W- 5 Safe End To Nozzle 2 501151 W- 5 Safe End To Nozzle 2 501484 W-1 1 Safe End to Nozzle 2 501499 W-11 Safe End To Nozzle Revision 0 Page 3 of 3 Date 3/30/2004