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MONTHYEARML14106A1732014-04-10010 April 2014 NRR E-mail Capture - Monticello Nuclear Generating Plant Acceptance Review Relief Request RR-008 Project stage: Acceptance Review ML14258A7362014-09-11011 September 2014 NRR E-mail Capture - Monticello Nuclear Generating Plant - Final Requests for Additional Information Request RR-008 Project stage: RAI ML15013A0362015-01-23023 January 2015 Relief Request RR-008 Alternative to ASME Code, Section XI, Examination Requirements for the Reactor Pressure Vessel Shroud Support Plate Welds H8 and H9 for the Fifth 10-Year ISI Interval Project stage: Other 2014-09-11
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Category:Code Relief or Alternative
MONTHYEARML23248A2092023-09-18018 September 2023 Proposed Alternative VR-11 to the Requirements of the ASME OM Code Associated with Periodic Verification Testing of MO-2397, Reactor Water Cleanup Inboard Isolation Valve ML23123A4222023-05-16016 May 2023 Request RR 001 to Use Later Edition of ASME BPV Section XI Code for Cisi Program ML23107A2852023-04-25025 April 2023 Authorization and Safety Evaluation for Alternative Request PR-08 ML23017A2222023-01-13013 January 2023 Verbal Authorization of Proposed Alternative PR-08 Regarding Inservice Testing Requirements of Certain High Pressure Coolant Injection System Components (EPID L-2022-LLR-0088) (Email) ML22314A2162022-11-16016 November 2022 Withdrawal of Alternative Request VR-08 ML22270A2312022-09-30030 September 2022 Authorization and Safety Evaluation for Alternative Request No. VR-10 ML22208A1952022-08-0303 August 2022 Summary of July 26, 2022, Meeting with Northern States Power Company, Doing Business as Xcel Energy, Related to the Alternative Request for Excess Flow Check Valves at Monticello Nuclear Generating Plant ML22126A1052022-06-21021 June 2022 Authorization and Safety Evaluation for Alternative Request No. VR-01 ML22154A5502022-06-0707 June 2022 Authorization and Safety Evaluation for Alternative Request No. PR-02 ML22126A1342022-05-12012 May 2022 Authorization and Safety Evaluation for Alternative Request No. PR-05 ML22130A6562022-05-11011 May 2022 Authorization and Safety Evaluation for Alternative Request No. PR-04 ML22098A1272022-05-0202 May 2022 Authorization and Safety Evaluation for Alternative Request No. VR-02 ML22110A1232022-04-25025 April 2022 Withdrawal of Relief Requests PR-09 and VR-07 (Epids: L 2021 Llr 0095 and L 2022 Llr 0021) ML22018A1772022-01-21021 January 2022 Authorization and Safety Evaluation for Alternative Request No. VR-05 ML20307A0172020-11-0202 November 2020 Request for Alternative for Core Support Structure Weld Examination ML20174A5452020-07-15015 July 2020 Request for Alternative for Pressure Isolation Valve Testing ML20135G9922020-05-29029 May 2020 and Prairie Island Nuclear Generating Plant, Units 1 and 2 - 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 L-MT-18-023, 10 CFR 50.55a Request RR-012: Inservice Inspection Impracticality in Accordance with 10 CFR 50 .55a(g)(5)(iii) During the Fifth Ten-Year Interval2018-05-11011 May 2018 10 CFR 50.55a Request RR-012: Inservice Inspection Impracticality in Accordance with 10 CFR 50 .55a(g)(5)(iii) During the Fifth Ten-Year Interval ML17122A1572017-05-15015 May 2017 Request for Alternative to Use Code Case OMN-20 for the Fifth 10-Year Inservice Testing Interval ML16208A4622016-08-0303 August 2016 Safety Evaluation for Request for Alternative Associated with Reactor Pressure Vessel Internals and Components Inspection for the Fifth 10-Year Interval L-MT-15-083, 10 CFR 50.55a Request No. RR-010: Request for Approval of an Alternative to Apply the BWRVIP Guidelines in Lieu of Specific ASME Section XI Code Requirements for Reactor Pressure Vessel Internals and Components Inspection2015-11-20020 November 2015 10 CFR 50.55a Request No. RR-010: Request for Approval of an Alternative to Apply the BWRVIP Guidelines in Lieu of Specific ASME Section XI Code Requirements for Reactor Pressure Vessel Internals and Components Inspection ML15028A1522015-02-19019 February 2015 Relief Request RR-009 Regarding Relief from Examination Coverage Requirements of Section XI of the ASME Code for the Fifth 10-Year Inservice Inspection Program Interval ML15013A0362015-01-23023 January 2015 Relief Request RR-008 Alternative to ASME Code, Section XI, Examination Requirements for the Reactor Pressure Vessel Shroud Support Plate Welds H8 and H9 for the Fifth 10-Year ISI Interval ML14223A5812014-08-27027 August 2014 Alternative Request Vr 05 to the Testing Requirements of the ASME OM Code for the Fifth 10-Year Inservice Inspection Program Interval ML12244A2722012-09-26026 September 2012 Relief from the Requirements of ASME OM Code for the Fifth Ten-Year IST Program Interval (TAC Nos. ME8067, ME8088 Through ME8096) ML12180A5882012-07-12012 July 2012 Approval of ISI Relief Request RR-007 for the Fifth 10-year Interval ML1020006722010-07-28028 July 2010 Approval of Alternative to Use ASME Code Case N-705 to Address Cracks at the Standby Liquid Control Tank L-MT-10-014, Request 10 CFR 50.55a Request 17: Extension of Permanent Relief from Volumetric Examination of Reactor Pressure Vessel Circumferential Shell Welds for the Renewed Operating License Term2010-03-12012 March 2010 Request 10 CFR 50.55a Request 17: Extension of Permanent Relief from Volumetric Examination of Reactor Pressure Vessel Circumferential Shell Welds for the Renewed Operating License Term 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 L-MT-05-074, Request to Use Subsequent Edition and Addenda of ASME Code for Inservice Testing2005-07-29029 July 2005 Request to Use Subsequent Edition and Addenda of ASME Code for Inservice Testing ML0505600492005-03-0808 March 2005 Fourth 10-Year Inservice Inspection Interval Request for Relief to Use Code Case N-661 ML0436300192005-01-0606 January 2005 Relief, Fourth 10-year Inservice Inspection Interval Request for Relief No. 4, MC2222 ML0407004152004-03-25025 March 2004 Third 10-Year Interval Inservice Inspection Request for Relief RR-17, Involving Repair/Replacement Activity on the Topworks of Main Steam Safety Relief Valve (SRV) G 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 ML0320401572003-10-0303 October 2003 Relief Request No. 7, Fourth 10-Year Interval Inservice Inspection Program Plan L-MT-03-045, Request for Approval of Inservice Inspection Program Third 10-Year Interval Relief Request No. 172003-08-27027 August 2003 Request for Approval of Inservice Inspection Program Third 10-Year Interval Relief Request No. 17 ML0320605802003-08-0707 August 2003 Relief, Fourth 10-Year Interval Inservice Testing Program ML0317002092003-07-17017 July 2003 Relief Request, Nos. PR-01, PR-02, PR-03, PR-04, PR-05, and VR-02 Related to the Fourth 10-Year Interval Inservice Testing Program L-MT-03-048, Request for Authorization of Inservice Inspection Program Fourth 10-Year Interval Relief Request No. 82003-06-12012 June 2003 Request for Authorization of Inservice Inspection Program Fourth 10-Year Interval Relief Request No. 8 ML0316008642003-06-0909 June 2003 Relief Request, Fourth 10-Year Interval Inservice Inspection Program Plan Relief Request No. 5, TAC No. MB6956 ML0314001192003-05-19019 May 2003 Relief, Third 10-Year Interval Inservice Inspection Relief Request No 16, Parts a, B, and C L-MT-03-001, Relief Request No. Pr 06 for Fourth 10-Year Inservice Testing Interval - High Pressure Coolant Injection Pump Testing2003-05-0606 May 2003 Relief Request No. Pr 06 for Fourth 10-Year Inservice Testing Interval - High Pressure Coolant Injection Pump Testing 2023-09-18
[Table view] Category:Letter
MONTHYEARML24025A9362024-01-31031 January 2024 Exemption from Select Requirements of 10 CFR Part 73 (EPID L-2023-LLE-0055 (Security Notifications, Reports, and Recordkeeping and Suspicious Activity Reporting)) IR 05000263/20230042024-01-31031 January 2024 Integrated Inspection Report 05000263/2023004 ML24024A0722024-01-24024 January 2024 Independent Spent Fuel Storage Installation, Onticello, Supplement to Request for Exemption from Enhanced Weapons, Firearms Background Checks, and Security Event Notifications Implementation IR 05000263/20244012024-01-22022 January 2024 Information Request for the Cyber-Security Baseline Inspection, Notification to Perform Inspection 05000263/2024401 L-MT-23-054, Subsequent License Renewal Application Supplement 82024-01-11011 January 2024 Subsequent License Renewal Application Supplement 8 L-MT-23-047, License Amendment Request: Revision to the MNGP Pressure Temperature Limits Report to Change the Neutron Fluence Methodology and Incorporate New Surveillance Capsule Data2023-12-29029 December 2023 License Amendment Request: Revision to the MNGP Pressure Temperature Limits Report to Change the Neutron Fluence Methodology and Incorporate New Surveillance Capsule Data L-MT-23-056, Subsequent License Renewal Application Response to Request for Additional Information and Request for Confirmation of Information - Set 1 Part 22023-12-18018 December 2023 Subsequent License Renewal Application Response to Request for Additional Information and Request for Confirmation of Information - Set 1 Part 2 ML23349A0572023-12-15015 December 2023 and Independent Spent Fuel Storage Installation, Revision to Correspondence Service List for Northern States Power - Minnesota IR 05000263/20234022023-12-13013 December 2023 Security Baseline Inspection Report 05000263/2023402 L-MT-23-042, 2023 Annual Report of Changes in Emergency Core Cooling System Evaluation Models Pursuant to 10 CFR 50.462023-12-11011 December 2023 2023 Annual Report of Changes in Emergency Core Cooling System Evaluation Models Pursuant to 10 CFR 50.46 L-MT-23-052, Subsequent License Renewal Application Supplement 72023-11-30030 November 2023 Subsequent License Renewal Application Supplement 7 L-MT-23-051, Update to the Technical Specification Bases2023-11-28028 November 2023 Update to the Technical Specification Bases L-MT-23-049, Subsequent License Renewal Application Response to Request for Additional Information and Request for Confirmation of Information - Set 12023-11-21021 November 2023 Subsequent License Renewal Application Response to Request for Additional Information and Request for Confirmation of Information - Set 1 ML23319A3182023-11-15015 November 2023 Request for Exemption from Enhanced Weapons, Firearms Background Checks, and Security Event Notifications Implementation IR 05000263/20230032023-11-13013 November 2023 Integrated Inspection Report 05000263/2023003 and 07200058/2023001 L-MT-23-043, 10 CFR 50.55a(z)(1) Request Regarding OMN-17, Revision 1. VR-092023-11-13013 November 2023 10 CFR 50.55a(z)(1) Request Regarding OMN-17, Revision 1. VR-09 L-MT-23-038, License Amendment Request to Revise Monticello Technical Specification Surveillance Requirement 3.8.6.62023-11-10010 November 2023 License Amendment Request to Revise Monticello Technical Specification Surveillance Requirement 3.8.6.6 L-MT-23-046, Subsequent License Renewal Application Response to Request for Additional Information Round 2 - Set 12023-11-0909 November 2023 Subsequent License Renewal Application Response to Request for Additional Information Round 2 - Set 1 ML23291A1102023-10-23023 October 2023 Environmental Audit Summary and RCIs and RAIs ML23285A3062023-10-12012 October 2023 Implementation of the Fleet Standard Emergency Plan for the Monticello Nuclear Generating Plant and the Prairie Island Nuclear Generating Plant L-MT-23-041, Subsequent License Renewal Application Response to Request for Confirmation of Information Set 22023-10-0303 October 2023 Subsequent License Renewal Application Response to Request for Confirmation of Information Set 2 L-MT-23-037, Subsequent License Renewal Application Response to Request for Additional Information Set 32023-09-22022 September 2023 Subsequent License Renewal Application Response to Request for Additional Information Set 3 ML23262B0372023-09-19019 September 2023 Response to NRC Request for Additional Information Regarding the 2023 Monticello and Prairie Island Plant Decommissioning Funding Status Reports ML23248A2092023-09-18018 September 2023 Proposed Alternative VR-11 to the Requirements of the ASME OM Code Associated with Periodic Verification Testing of MO-2397, Reactor Water Cleanup Inboard Isolation Valve ML23256A1682023-09-13013 September 2023 Independent Spent Fuel Storage Installation and Monticello Nuclear Generating Plant - Voluntary Security Clearance Program 2023 Insider Threat Program Self-Inspection IR 05000263/20230102023-09-0707 September 2023 Commercial Grade Dedication Inspection Report 05000263/2023010 L-MT-23-036, Subsequent License Renewal Application Response to Request for Additional Information Set 2 and Supplement 62023-09-0505 September 2023 Subsequent License Renewal Application Response to Request for Additional Information Set 2 and Supplement 6 ML23214A2412023-08-31031 August 2023 Letter: Aging Management Audit - Monticello Unit 1 - Subsequent License Renewal Application IR 05000263/20230052023-08-30030 August 2023 Updated Inspection Plan for Monticello Nuclear Generating Plant (Report 05000263/2023005) L-MT-23-035, Subsequent License Renewal Application Supplement 52023-08-28028 August 2023 Subsequent License Renewal Application Supplement 5 ML23241A9732023-08-21021 August 2023 Request for Scoping Comments Concerning the Environmental Review of Monticello Nuclear Generating Plant, Unit 1, Subsequent License Renewal Application (Docket No. 50-263) L-MT-23-034, Subsequent License Renewal Application Response to Request for Additional Information Set 12023-08-15015 August 2023 Subsequent License Renewal Application Response to Request for Additional Information Set 1 ML23222A0122023-08-10010 August 2023 Independent Spent Fuel Storage Installation and Monticello Nuclear Generating Plant - Changes in Foreign Ownership, Control or Influence ML23215A1312023-08-0909 August 2023 License Renewal Regulatory Audit Regarding the Environmental Review of the Subsequent License Renewal Application IR 05000263/20230022023-08-0707 August 2023 Plantintegrated Inspection Report 05000263/2023002 L-MT-23-028, 2023 Refueling Outage 90-Day Inservice Inspection (ISI) Summary Report2023-07-31031 July 2023 2023 Refueling Outage 90-Day Inservice Inspection (ISI) Summary Report L-MT-23-032, 10 CFR 50.55a(z)(2) Request Regarding MO-2397, VR-112023-07-31031 July 2023 10 CFR 50.55a(z)(2) Request Regarding MO-2397, VR-11 ML23198A0412023-07-28028 July 2023 LRA Availability Letter ML23206A2342023-07-25025 July 2023 Independent Spent Fuel Storage Installation, and Monticello Nuclear Generating Plant, Changes in Foreign Ownership, Control or Influence ML23201A0352023-07-24024 July 2023 Notification of an NRC Biennial Licensed Operator Requalification Program Inspection and Request for Information ML23202A0032023-07-21021 July 2023 Independent Spent Fuel and Independent Spent Fuel Storage Installation, Monticello Nuclear Generating Plant, Submittal of Quality Assurance Topical Report (NSPM-1) L-MT-23-031, Subsequent License Renewal Application Supplement 4 and Responses to Request for Confirmation of Information - Set 12023-07-18018 July 2023 Subsequent License Renewal Application Supplement 4 and Responses to Request for Confirmation of Information - Set 1 ML23195A1732023-07-14014 July 2023 Revision of Standard Practice Procedures Plan IR 05000263/20235012023-07-13013 July 2023 Emergency Preparedness Inspection Report 05000263/2023501 2024-01-31
[Table view] Category:Safety Evaluation
MONTHYEARML23248A2092023-09-18018 September 2023 Proposed Alternative VR-11 to the Requirements of the ASME OM Code Associated with Periodic Verification Testing of MO-2397, Reactor Water Cleanup Inboard Isolation Valve ML23107A2852023-04-25025 April 2023 Authorization and Safety Evaluation for Alternative Request PR-08 ML23079A0742023-04-11011 April 2023 Request RR-003 to Use Later Edition of ASME Section XI Code for ISI Code of Record ML22357A1002023-03-31031 March 2023 and Prairie Island Nuclear Generating Plant, Units 1 and 2 - Issuance of Amendments Standard Emergency Plan and Consolidated Emergency Operations Facility ML23012A1562023-01-13013 January 2023 Issuance of Amendment No. 210 Re Revised Methodologies for Determining the Core Operating Limits (EPID L-2021-LLA-0144) - Non-proprietary ML22318A2152022-12-27027 December 2022 Issuance of Amendment No. 209 Ten-Year Inspection of the Diesel Generator Fuel Oil Storage Tank ML22264A1062022-10-31031 October 2022 Issuance of Amendment No. 208 Residual Heat Removal Drywell Spray Header and Nozzle Surveillance Frequency ML22270A2312022-09-30030 September 2022 Authorization and Safety Evaluation for Alternative Request No. VR-10 ML22154A5502022-06-0707 June 2022 Authorization and Safety Evaluation for Alternative Request No. PR-02 ML22126A1342022-05-12012 May 2022 Authorization and Safety Evaluation for Alternative Request No. PR-05 ML22098A1272022-05-0202 May 2022 Authorization and Safety Evaluation for Alternative Request No. VR-02 ML22018A1772022-01-21021 January 2022 Authorization and Safety Evaluation for Alternative Request No. VR-05 ML21223A2802021-10-15015 October 2021 Issuance of Amendment No. 207 Adoption of TSTF-564 Safety Limit MCPR ML21148A2742021-07-12012 July 2021 Issuance of Amendment No. 206 TSTF-505, Revision 2, Provide Risk-Informed Extended Completion Times - RITSTF Initiative 4b ML20352A3492021-01-0808 January 2021 Issuance of Amendment No. 205, Revise Technical Specifications to Adopt Technical Specifications Task Force (TSTF) Traveler TSTF-582, RPV WIC Enhancements, and TSTF-583-T, TSTF-582 Diesel Generator Variation ML20346A0972020-12-21021 December 2020 Request for Alternative for Examination of Reactor Pressure Vessel Threads in Flange ML20336A1602020-12-0909 December 2020 Request to Use a Provision of a Later Edition of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code, Section XI ML20307A0172020-11-0202 November 2020 Request for Alternative for Core Support Structure Weld Examination ML20210M0142020-09-0808 September 2020 and Prairie Island Nuclear Generating Plant, Units 1 and 2 - Issuance of Amendment Nos. 204, 231, and 219 TSTF-529 Clarify Use and Application Rules ML20153A8042020-07-31031 July 2020 Co. - Issuance of Amendments Revising Emergency Action RA3 ML20174A5452020-07-15015 July 2020 Request for Alternative for Pressure Isolation Valve Testing ML20153A4012020-06-0101 June 2020 and Prairie Island Nuclear Generating Plant, Units 1 and 2 - Relief from the Requirements of the ASME Code ML20135G9922020-05-29029 May 2020 and Prairie Island Nuclear Generating Plant, Units 1 and 2 - Relief from the Requirements of the ASME Code ML20134H9582020-05-29029 May 2020 and Prairie Island Nuclear Generating Plant, Units 1 and 2 - Relief from the Requirements of the ASME Code ML19255F5822019-10-0101 October 2019 Safety Evaluation Regarding Implementation of Hardened Containment Vents Capable of Operation Under Severe Accident Conditions Related to Order EA-13-109 (CAC No. MF4376; EPID No. L-2014-JLD-0052) ML19162A0932019-07-30030 July 2019 Issuance of Amendment No, 202 Regarding Deletion of the Note Associated with Technical Specification 3.5.1., Erccs - Operating ML19074A2692019-04-22022 April 2019 Non-Proprietary - Issuance of Amendment Revision to Technical Specifications 2.1.2 Safety Limit Minimum Critical Power Ratio ML19052A1422019-03-11011 March 2019 Correction to License Amendment No. 198 Related to Adoption of TSTF-542, Reactor Pressure Vessel Water Inventory Control ML19065A2002019-03-11011 March 2019 Correction to License Amendment No. 200 Related to Adoption of TSTF-425, Relocated Surveillance Frequencies to Licensee Control - RITSTF Initiative 5B ML19007A0902019-01-28028 January 2019 Issuance of Amendment Adoption of TSTF-425, Relocate Surveillance Frequencies to Licensee Control-RITSTF Initiative 5B ML18291B2142018-11-26026 November 2018 Issuance of Amendment Adoption of TSTF-551 Revise Secondary Containment Surveillance Requirements ML18250A0752018-10-29029 October 2018 Issuance of Amendment Adoption of TSTF-542, Reactor Pressure Vessel Water Inventory Control ML18270A0152018-10-19019 October 2018 and Prairie Island Nuclear Generating Plant, Units 1 and 2 - Relief from the Requirements of the ASME OM Code ML17345A0462018-03-0606 March 2018 Issuance of Amendment No. 197 to Adopt Changes to the Emergency Plan (CAC No. MF9560; EPID L-2017-LLA-0184) ML17319A5912017-12-10010 December 2017 Safety Evaluation Regarding Implementation of Mitigating Strategies and Reliable Spent Fuel Pool Instrumentation Related to Orders EA-12-049 and EA-12-051 ML17310B2392017-11-28028 November 2017 and Prairie Island Nuclear Generating Plant, Units 1 and 2 - Issuance of Amendment Unit Staff Qualifications (CAC Nos. MF9545, MF9546, and MF9547; EPID L-2017-LLA-0195) ML17123A3212017-06-16016 June 2017 Issuance of Amendment Adoption of TSTF-545, Revision 3, TS Inservice Testing Program Removal and Clarify SR Usage Rule Application to Section 5.5 Testing ML17122A1572017-05-15015 May 2017 Request for Alternative to Use Code Case OMN-20 for the Fifth 10-Year Inservice Testing Interval ML17103A2352017-04-25025 April 2017 Issuance of Amendment Technical Specification 5.5.11 Primary Containment Leakage Rate Testing Program ML17013A4352017-02-27027 February 2017 Issuance of Amendment Revision to Technical Specification Surveillance Requirement 3.8.4.2 ML17054C3942017-02-23023 February 2017 Non-Proprietary Issuance of Amendment Extended Flow Window ML16320A0212016-11-28028 November 2016 and Prairie Island Nuclear Generating Plant, Units 1 and 2 - Review of Changes to the Northern States Power Company Quality Assurance Topical Report ML16244A1202016-09-0606 September 2016 Generation Plant - Interim Staff Evaluation Relating to Overall Integrated Plan in Response to Phase 2 of Order EA-13-109 (Severe Accident Capable Hardened Vents) ML16208A4622016-08-0303 August 2016 Safety Evaluation for Request for Alternative Associated with Reactor Pressure Vessel Internals and Components Inspection for the Fifth 10-Year Interval ML16196A3032016-08-0101 August 2016 Issuance of Amendment Technical Specifications Surveillance Requirement 3.5.1.3 B to Correct Alternative Nitrogen System Pressure (Cac. No. MF6704) ML16125A1652016-06-21021 June 2016 Issuance of Amendment Technical Specifications Task Force (TSTF) Standard Technical Specifications (STS) Change Traveler TSTF-523, Revision 2 Generic Letter 2008-01, Managing Gas Accumulation ML15175A0162015-06-30030 June 2015 Staff Evaluation of 10 CFR 50.54(p)(2) Changes to Security Plans ML15072A1412015-06-0505 June 2015 Issuance of Amendment No. 188 Regarding Transition to Areva Atrium 10XM Fuel and Areva Safety Analysis Methods ML15154A4772015-06-0505 June 2015 Safety Evaluation Regarding License Amendment No. 188 Associated with Areva Atrium 10XM Fuel Transition (TAC No. MF2479) - (Redacted) ML14358A0392015-02-20020 February 2015 Northern States Power Company, Minnesota (NSPM) - Monticello Nuclear Generating Plant, Prairie Island Nuclear Generating Plant, Prairie Island ISFSI - Review of Changes to the NSPM Quality Assurance Topical Report 2023-09-18
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UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 January 23, 2015 Karen D. Fili Site Vice-President Northern States Power Company -Minnesota Monticello Nuclear Generating Plant 2807 West County Road 75 Monticello, MN 55362-9637
SUBJECT:
MONTICELLO NUCLEAR GENERATING PLANT-ALTERNATIVE TO THE AMERICAN SOCIETY OF MECHANICAL ENGINEERS BOILER AND PRESSURE VESSEL CODE, SECTION XI, EXAMINATION REQUIREMENTS FOR THE REACTOR PRESSURE SHROUD SUPPORT PLATE WELDS H8 AND H9 FOR THE FIFTH 10-YEAR INSERVICE INSPECTION PROGRAM INTERVAL (TAC NO. MF3551)
Dear Ms. Fili:
By letter dated February 28, 2014, as supplemented by letter dated October 10, 2014, Northern States Power Company, a Minnesota corporation (NSPM, the licensee},
doing business as Xcel Energy, submitted a relief request (RR) RR-008 to the U.S. Nuclear Regulatory Commission (NRC). Specifically, pursuant to Title 10 of the Code of Federal Regulations (1 0 CFR), Section 50.55a(a)(3)(ii},
NSPM requested relief to use a proposed alternative on the basis that complying with the specified requirement would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety. The proposed change in RR-008 would allow NSPM to revise the inspection requirements for examination coverage of shroud support plate welds H8 and H9 at the Monticello Nuclear Generating Plant (MNGP) from those based on the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (Code),Section XI, IWB-2420(b),
regarding examination of previously detected flaws. The NRC staff has reviewed the proposed alternative in RR-008 and determined, as set forth in the enclosed safety evaluation, that NSPM has adequately addressed all of the regulatory requirements set forth in 10 CFR 50.55a(a)(3)(ii),
and remains in compliance with the ASME Code requirements.
The staff concludes that the alternative provides reasonable assurance of the structural integrity of shroud support plate welds H8 and H9. Therefore, the proposed alternative specified in RR-008 is authorized in accordance with 10 CFR 50.55a(a)(3)(ii) for the fifth 1 0-year inservice inspection at MNGP that is expected to end on May 31, 2022.
K.Fili If you have any questions, please contact Terry Beltz at (301) 415-3049 or via e-mail at Terry.Beltz@nrc.gov.
Docket No. 50-263
Enclosure:
a d L. Pel on, Chief Plant Licensing Branch 111-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Staff Evaluation of Relief Request RR-008 for the Fifth 10-Year lnservice Inspection Interval cc w/encl: Distribution via ListServ UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELIEF REQUEST RR-008 REGARDING SHROUD SUPPORT PLATE WELDS H8 AND H9 FOR THE FIFTH 10-YEAR INSERVICE INSPECTION PROGRAM INTERVAL MONTICELLO NUCLEAR GENERATING PLANT NORTHERN STATES POWER COMPANY-MINNESOTA DOCKET NO. 50-263 (TAC NO. MF3551)
1.0 INTRODUCTION
By letter dated February 28, 2014 (Agencywide Document Access and Management System (ADAMS) Accession No. ML 14064A191),
as supplemented by letter dated October 24, 2014 (ADAMS Accession No. ML 14286A001),
Northern States Power Company-Minnesota (NSPM) requested relief from the requirements of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code),Section XI, IWB-2420(b).
The relief would revise inspection requirements for examination coverage of shroud support plate welds H8 and H9 at the Monticello Nuclear Generating Plant (MNGP). Specifically, pursuant to Title 10 of the Code of Federal Regulations (1 0 CFR) 50.55a(a)(3)(ii),
the licensee requested to use the proposed alternative specified in relief request RR-008 on the basis that complying with the specified requirement would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety associated with reexamination of previously detected flaws in the aforementioned welds. The alternative specified in RR-008 would change the inspection program for the fifth 1 0-year inservice inspection interval at MNGP. The U.S. Nuclear Regulatory Commission (NRC) staff's evaluation of the licensee's proposed relief follows.
2.0 REGULATORY EVALUATION lnservice inspection (lSI) of ASME Code Class 1, 2, and 3 components is performed in accordance with Section XI of the ASME Code and applicable addenda as required by 1 0 CFR, Section 50.55a(g),
except where specific relief has been granted by the NRC pursuant to 10 CFR 50.55a(g)(6)(i).
Paragraph 55a(a)(3) of 10 CFR 50 states that alternatives to the requirements of 10 CFR 50.55a(g) may be used, when authorized by the NRC, if (i) the Enclosure proposed alternatives would provide an acceptable level of quality and safety or (ii) compliance with the specified requirements would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety. Pursuant to 10 CFR 50.55a(g)(4
), ASME Code Class 1, 2, and 3 components (including supports) shall meet the requirements, except the design and access provisions and the service examination requirements, set forth in the ASME Code,Section XI, "Rules for lnservice Inspection of Nuclear Power Plant Components,"
to the extent practical within the limitations of design, geometry, and materials of construction of the components.
The regulations also require that inservice examination of components and system pressure tests conducted during the first 1 0-year lSI interval and subsequent intervals comply with the requirements in the latest edition and addenda of Section XI of the ASME Code incorporated by reference in 1 0 CFR 50.55a(b),
12 months prior to the start of the 120-month interval and subject to the limitations and modifications listed therein.
The ASME Code of record for the fifth 1 0-year lSI interval at MNGP is the 2007 Edition of the ASME Code,Section XI, with the 2008 Addenda.
3.0 TECHNICAL EVALUATION The licensee provided information in support of its request for relief from ASME Code requirements.
Pursuant to 10 CFR 50.55a(a)(3)(ii),
the licensee requests relief by demonstrating that compliance with the specified requirements would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety. This information was evaluated by the NRC staff, and the basis for disposition is documented below. 3.1 ASME Code Components The components affected by RR-008 are core shroud support plate to core shroud weld (H8) and core shroud plate to reactor pressure vessel (RPV) weld (H9) under Examination Category B-N-2, "Welded Core Support Structures and Interior Attachments to Reactor Vessels,"
Item Numbers 813.30, "Interior Attachments Beyond Beltline Region,"
and 813.40, "Core Support Structure."
3.2 ASME Code Requirements (as stated by the licensee)
The 2007 Edition of the ASME Code,Section XI, with the 2008, Addenda IWB-2420(b) states the following:
If a component is accepted for continued service in accordance with IWB-3132.3 or IWB-3142.4, the areas containing flaws or relevant conditions shall be reexamined during the next three inspection periods listed in the schedule of the Inspection Program of IWB-2400.
Alternatively, acoustic emission may be used to monitor growth of existing flaws in accordance with IWA-2234.
3.3 Licensee's Hardship in Complying with the Requirements (as stated by the licensee)
In its application, the licensee states that gaining access to the lower plenum within the reactor vessel is unusually difficult due to the inherent design configuration of the reactor vessel internals with a welded core shroud and support assembly, fuel core and core support components, core instrumentation, sparger piping in the upper vessel regions, and jet pump assemblies in the annulus region. To gain further access would require extensive disassembly of the fuel cells or jet pumps. The licensee further stated that in 2013, the general area dose rates on the refuel floor near the refuel cavity were approximately 1 to 3 millirem per hour (mrem/hr).
Applying this general area dose rate to a lower time estimate of 200 hours0.00231 days <br />0.0556 hours <br />3.306878e-4 weeks <br />7.61e-5 months <br /> for disassembly and reassembly of all 1 0 jet pump pairs, using 3 workers per shift at 2 shifts per day, the estimated dose would be 600 to 1800 mrem. 3.4 Licensee's Proposed Alternative (as stated by the licensee)
The licensee proposes to visually inspect all accessible areas of the topside and underside of both the H8 and H9 welds during each remaining refueling outage in the three periods of the MNGP Fifth lSI Interval (i.e., 2015, 2017, 2019, and 2021). NSPM intends to continue accessing the lower plenum via the jet pump inlets to perform the visual inspections.
In addition to inspecting all accessible areas of the H8 and H9 welds for changes in the general condition of the welds, NSPM will select four areas with known, distinct indications on the underside of the shroud support plate in the H8 and H9 welds to monitor for any visually apparent changes in the flaw. The areas selected will include two locations on each weld, and will be located in different quadrants of the reactor vessel. The selected locations will be mapped (by photo, video, or other effective method) and visually compared to the previous inspection.
The flaw locations will be examined for visual evidence of new branching, visual evidence of length changes (e.g. flaws that once only covered a portion of the weld now completely cross the weld, etc.), and visual evidence of any flaws that extend into the reactor vessel low alloy steel or the shroud support plate itself. The accessible topside of the welds will also be inspected to verify no cracking has penetrated through the thickness of the weld (e.g. crack-like indications on the topside that could be connected to cracking on the underside).
Based on the inspection
- results, NSPM will determine the need for additional evaluations or any resulting actions and implement them accordingly.
3.5 Licensee's Basis for the Proposed Alternative (as stated by the licensee in RR-008) The licensee's proposed alternative is based, in part, on the premise that by performing detailed mapping and monitoring of a representative sample of the flaws and investigating more refined inspection techniques, MNGP can monitor the condition of the H8 and H9 welds and continue to meet the intent of IWB-2420(b).
The primary concern related to the H8 and H9 welds is uplift of the shroud support plate in the event of a design basis loss-of-coolant accident event. The uplift of the shroud support plate would be driven by the vertical seismic loads and reactor internal pressure differences across the plate. In a 2013 evaluation, using conservative flaw profiles and consideration of the loading acting upon the shroud support plate in the reactor vessel, only 18 percent of the total weld surface is required to be free of through-wall indications to overcome the uplift loads acting on the shroud support plate. Considering such extensive flaw profiles, the evaluations demonstrate that the structural integrity of the shroud support plate and its ability to resist uplift remain intact for at least 12 years and maintain the core coolant envelope. In addition to the flaw tolerant design, hydrogen water chemistry (HWC) was implemented at MNGP in 1989, and online noble metal chemistry control was implemented in 2013. Therefore, the environment of the lower plenum is well-mitigated against flaw growth and initiation based on water chemistry controls.
Enclosure 2 of the application (ADAMS Accession No. ML 14064A 186) is an evaluation performed by Structural Integrity Associates, Inc., entitled, "Monticello Shroud Support Structure Flaw Evaluation Review and Support Plate Weld Inspection Recommendations."
The evaluation provides supporting detail and technical justifications for the licensee's determination as to whether reduced inspection coverage is technically justifiable in lieu of the implicit requirement to inspect all flawed areas of the H8 and H9 welds as described in IWB-2420(b).
This document recommends minimum inspection requirements based on the conservatisms built into the evaluations performed to date, water chemistry in the lower plenum of the reactor vessel, and crack growth potential of the flaws. The document also recommends a minimum inspection of 5 percent coverage of the bottom side of welds H8 and H9 in areas with known flaws with the objective of monitoring for unexpected change in flaw appearance.
It was noted in Enclosure 2 that, in 2013, MNGP was able to inspect 32 percent of the topside of the H8 weld, and 35 percent of the topside of the H9 weld, with no relevant indications.
Based on a review of previously inspected regions on the underside of the H8 and H9 welds, the areas accessed through the jet pump inlets will be used to meet the 5 percent coverage recommendation.
Therefore, all the performed and proposed inspections meet the minimum recommended inspection requirements stated above. 3.6 Duration of the Licensee's Proposed Alternative The proposed alternative will be used for the fifth 1 0-year lSI Interval Program for MNGP that is expected to end on May 31 , 2022. 3. 7 NRC Staff Evaluation 3.7.1 Evaluation of Hardship Section XI of the ASME Code requires the areas containing flaws or relevant conditions in the H8 and H9 welds be reexamined during the next three inspection periods listed in the schedule of the Inspection Program of IWB-2400.
As indicated in Section 3.3 of this safety evaluation, access to the lower plenum is difficult due to inherent design configuration of the RPV internals with a core shroud and support assembly, fuel core and core support components, core instrumentation, sparger piping in the upper vessel regions, and jet pump assemblies in the annulus region. As a result, the licensee's previous lSI inspections of the H8 and H9 welds were limited to EVT-1 examination of the welds at 0 degree and 180 degree at the N1A and N1 B recirculation suction nozzles, and a VT-3 examination of accessible areas of the H8 and H9 welds above and below the shroud support plate (e.g., the 2011 lSI Summary Report).
The 2011 inspection coverage was approximately 17 percent of the entire vessel circumference length. This is consistent with BWRVIP-38, "BWR Shroud Support Inspection and Flaw Evaluation Guidelines,"
in which the licensee is required to inspect 10 percent of weld circumference for the H8 and H9 welds or the weld ligaments supported by a flaw tolerance analysis.
The accessibility issue associated with the H8 and H9 welds of BWR RPVs have been recognized by the BWRVIP for a long time. To support its request pursuant to 10 CFR 50.55a(a)(3)(ii},
the licensee needs to demonstrate that complying with the specified requirement would result in hardship without a compensating increase in the level of quality and safety. The licensee states that to gain additional inspection surfaces of the welds would require disassembly of the fuel cells or jet pumps, resulting in an increase in estimated dose of 600 to 1800 mrem. Based on the above, the NRC staff determines that the inspections performed in accordance with the ASME Code,Section XI requirements, as described in Section 3.2 of this safety evaluation, would be an additional hardship for NSPM based on the estimated increase in occupational dose and overall radiological impact. 3.7.2 Evaluation of the Level of Quality and Safety To demonstrate that the hardship is without a compensating increase in the level of quality and safety, the licensee proposed an alternative consisting of: (1) a flaw evaluation to support structural integrity of the component; (2) a VT inspection of all accessible areas of the topside and underside of the H8 and H9 welds during each remaining refueling outage in the fifth lSI interval, plus monitoring of four areas with known, distinct indications on the underside of the shroud support plate in the H8 and H9 welds, to validate the flaw evaluation conclusions; and (3) a VT inspection of the accessible topside of the welds to confirm that no cracking has penetrated through the thickness of the weld. Flaw evaluation results supporting the relief request are provided in Enclosure 2 of the application, based on flaw evaluation results obtained in 2011 and 2013. These evaluations considered two hypothetical flaw configurations:
(1) distributed through-wall flaws in the uninspected region and part-through-wall in the inspected region, and (2) a 360-degree circumferential flaw with a uniform crack depth of 75 percent through wall. These flaw configurations are very conservative because operating experience indicated no through-wall cracking was found in the H8 and H9 welds, and the approved crack growth model in BWRVIP-59-A, "Evaluation of Crack Growth in BWR Nickel Base Austenitic Alloys in RPV Internals,"
suggested that crack in these welds will not reach 66 percent of the wall thickness due to residual stress profiles.
The limit load analysis is consistent with BWRVIP-76-A, "BWR Core Shroud Inspection and Flaw Evaluation Guidelines,"
and is, therefore, acceptable.
Limit load analysis is a valid methodology for the H8 and H9 welds because these welds are under low neutron fluence and are unlikely to lose their material ductility during service.
The licensee's evaluation assumed that all lateral bending moments on the core shroud are supported by the shroud support legs. In Request for Additional Information (RAI) 1, the NRC staff requested justification for this assumption.
The licensee's response dated October 10, 2014, provided an analysis showing that the shroud support legs supported 86 percent of the load. This is acceptable because the margins in Table 3, which are associated with the more realistic flaw configuration, are big enough to cover any effect due to the minor lateral moment on the H8 and H9 welds that was not considered in the analysis.
As such, the NRC staff considers RAI 1 to be resolved.
Representativeness of Table 3 results will be further discussed when the licensee's response to RAI 6 is evaluated.
On the material resistance side, RAI 3 requested the licensee provide basis for the flow stress used in the limit load analysis.
The licensee's response stated that the flow stress is defined as 3Sm where Sm is from ASME Code,Section II design stress intensity for Alloy 600 material at 550°F. Since the flow stress is from the ASME Code, the NRC staff considers RAI 3 to be resolved.
RAI 6 requested further clarification regarding the acoustic (AC) loads on the jet pumps and the core shroud in the annulus region due to a recirculation suction line break. The licensee's response stated that the 2013 analysis conservatively considered twice the AC loads because complete information regarding whether the Mode of Characteristics (MOC) was used in the original AC load calculations for MNGP was not available at that time. Since the error identified in draft GE SC 12-20, "Error in Method of Characteristics Boundary Conditions Affecting Acoustic Loads Analyses,"
dated June 10, 2013, affected only plants using the MOC code, and the licensee later confirmed that MNGP did not use the MOC code in generating the AC loads, the NRC staff determined that using the 2011 analysis results (i.e., Table 3) is appropriate.
As such, the NRC staff considers RAI 6 to be resolved.
The staff considers that other RAis not specifically discussed in this safety evaluation have been resolved by the additional details provided by the licensee in its October 10, 2014, letter. Based on the above, the NRC staff concludes that the flaw evaluation with the conservatively assumed flaw configuration in the H8 and H9 welds meets the ASME code,Section XI specified margin, and, therefore, provides reasonable assurance of structural integrity of the H8 and H9 welds. Further, since the licensee's proposed alternative inspections can validate the flaw evaluation, there is reasonable assurance of the continued structural integrity of the H8 and H9 welds.
4.0 CONCLUSION
As set forth above, the NRC staff determines that the proposed alternative provides reasonable assurance of structural integrity of shroud support plate welds H8 and H9 at MNGP. Accordingly, the NRC staff concludes that the licensee has adequately addressed all of the regulatory requirements set forth in 10 CFR 50.55a(a)(3)(ii),
and is in compliance with the ASME Code's requirements.
Therefore, the NRC staff determines that complying with the specified requirement would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety. As such, the staff authorizes relief request RR-008 for the fifth 10-year lSI interval at the Monticello Nuclear Generating Plant that is expected to end on May 31, 2022. All other ASME Code,Section XI requirements for which relief was not specifically requested and approved in the subject request for relief remain applicable, including third-party review by the Authorized Nuclear lnservice Inspector.
Principle Contributor:
S. Sheng, NRR Date of issuance:
January 23, 2015 K. Fili If you have any questions, please contact Terry Beltz at (301) 415-3049 or via e-mail at Terry.Beltz@nrc.gov.
Sincerely, IRA/ David L. Pelton, Chief Plant Licensing Branch 111-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-263
Enclosure:
Staff Evaluation of Relief Request RR-008 for the Fifth 1 0-Year I nservice Inspection Interval cc w/encl: Distribution via ListServ DISTRIBUTION:
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