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Category:Letter type:L
MONTHYEARL-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 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 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 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 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 L-MT-23-035, Subsequent License Renewal Application Supplement 52023-08-28028 August 2023 Subsequent License Renewal Application Supplement 5 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 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 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 L-MT-23-030, Subsequent License Renewal Application Supplement 32023-07-0404 July 2023 Subsequent License Renewal Application Supplement 3 L-MT-23-025, Subsequent License Renewal Application Supplement 22023-06-26026 June 2023 Subsequent License Renewal Application Supplement 2 L-MT-23-019, Submittal of 2022 Annual Radiological Environmental Operating Report2023-05-10010 May 2023 Submittal of 2022 Annual Radiological Environmental Operating Report L-MT-23-020, Submittal of 2022 Annual Radioactive Effluent Release Report2023-05-10010 May 2023 Submittal of 2022 Annual Radioactive Effluent Release Report L-MT-23-021, Core Operating Limits Report (COLR) for the Monticello Nuclear Generating Plant for Cycle 322023-05-0202 May 2023 Core Operating Limits Report (COLR) for the Monticello Nuclear Generating Plant for Cycle 32 L-MT-23-017, 2022 Annual Report of Individual Monitoring for the Monticello Nuclear Generating Plant (MNGP)2023-04-18018 April 2023 2022 Annual Report of Individual Monitoring for the Monticello Nuclear Generating Plant (MNGP) L-MT-23-010, Subsequent License Renewal Application Supplement 12023-04-0303 April 2023 Subsequent License Renewal Application Supplement 1 L-MT-23-013, Core Operating Limits Report (COLR) for Cycle 31, Revision 32023-03-28028 March 2023 Core Operating Limits Report (COLR) for Cycle 31, Revision 3 L-MT-23-012, Core Operating Limits Report (COLR) for Monticello Nuclear Generating Plant Cycle 31, Revision 22023-03-17017 March 2023 Core Operating Limits Report (COLR) for Monticello Nuclear Generating Plant Cycle 31, Revision 2 L-MT-23-008, 10CFR50.55a Request to Use Later Edition of ASME Section XI for ISI Code of Record (RR-003)2023-02-0707 February 2023 10CFR50.55a Request to Use Later Edition of ASME Section XI for ISI Code of Record (RR-003) L-MT-23-004, CFR 50.55a Request RR-001, Request to Use a Provision of a Later Edition of the ASME Boiler and Pressure Vessel Code, Section XI for the Monticello Third Interval Containment Inservice Inspection Program2023-01-23023 January 2023 CFR 50.55a Request RR-001, Request to Use a Provision of a Later Edition of the ASME Boiler and Pressure Vessel Code, Section XI for the Monticello Third Interval Containment Inservice Inspection Program L-MT-23-005, Response to the NRC Request for Additional Information Regarding the 50.55a Request Pr 08, HPCI Pump Quarterly Testing (EPID Number L-2022-LLR-0088)2023-01-0606 January 2023 Response to the NRC Request for Additional Information Regarding the 50.55a Request Pr 08, HPCI Pump Quarterly Testing (EPID Number L-2022-LLR-0088) L-MT-22-049, Industry Groundwater Protection Initiative Special Report2022-12-15015 December 2022 Industry Groundwater Protection Initiative Special Report L-MT-22-052, L-MT-22-052 Monticello Nuclear Generating Plant 10 CFR 50.55a Request No. Pr 08, Request for HPCI Pump Quarterly Alternative2022-12-15015 December 2022 L-MT-22-052 Monticello Nuclear Generating Plant 10 CFR 50.55a Request No. Pr 08, Request for HPCI Pump Quarterly Alternative L-MT-22-046, 2022 Annual Report of Changes in Emergency Core Cooling System Evaluation Models Pursuant to 10 CFR 50.462022-12-13013 December 2022 2022 Annual Report of Changes in Emergency Core Cooling System Evaluation Models Pursuant to 10 CFR 50.46 L-MT-22-048, Update to the Monticello Technical Specification Bases2022-11-28028 November 2022 Update to the Monticello Technical Specification Bases L-MT-22-047, Withdrawal of Request for Relief from ASME OM Code for the Sixth Inservice Testing Interval2022-11-10010 November 2022 Withdrawal of Request for Relief from ASME OM Code for the Sixth Inservice Testing Interval L-MT-22-045, Letter Submitting Post-Exam Package2022-11-0404 November 2022 Letter Submitting Post-Exam Package L-MT-22-030, Sixth Interval Inservice Testing (1ST) Plan2022-09-0606 September 2022 Sixth Interval Inservice Testing (1ST) Plan L-MT-22-037, Supplement to 10 CFR 50.55a Request Associated with the Monticello Sixth Inservice Testing Ten-Year Interval, Alternative VR-10, Excess Flow Check Valve Testing Frequency2022-08-29029 August 2022 Supplement to 10 CFR 50.55a Request Associated with the Monticello Sixth Inservice Testing Ten-Year Interval, Alternative VR-10, Excess Flow Check Valve Testing Frequency L-MT-22-007, Response to Request for Additional Information for the Monticello Nuclear Generating Plant Alternative Request VR-08 (EPID: L-MT-22-007)2022-07-22022 July 2022 Response to Request for Additional Information for the Monticello Nuclear Generating Plant Alternative Request VR-08 (EPID: L-MT-22-007) L-MT-22-026, Changes to the Emergency Plan2022-07-19019 July 2022 Changes to the Emergency Plan L-MT-22-024, Response to a Request for Additional Information for the Monticello Nuclear Generating Plant Related to the Amendment to Adopt Advanced Framatome Methodologies2022-06-0606 June 2022 Response to a Request for Additional Information for the Monticello Nuclear Generating Plant Related to the Amendment to Adopt Advanced Framatome Methodologies L-MT-22-022, Response to a Request for Additional Information: Monticello Alternative VR-10, Excess Flow Check Valve Testing Frequency2022-05-25025 May 2022 Response to a Request for Additional Information: Monticello Alternative VR-10, Excess Flow Check Valve Testing Frequency L-MT-22-017, 2021 Annual Radiological Environmental Operating Report2022-05-11011 May 2022 2021 Annual Radiological Environmental Operating Report L-MT-22-018, 2021 Annual Radioactive Effluent Release Report2022-05-11011 May 2022 2021 Annual Radioactive Effluent Release Report L-MT-22-016, 2021 Annual Report of Individual Monitoring2022-04-28028 April 2022 2021 Annual Report of Individual Monitoring L-MT-22-019, Withdrawal of Requests for Relief from ASME OM Code for the Sixth Inservice Testing Interval2022-04-18018 April 2022 Withdrawal of Requests for Relief from ASME OM Code for the Sixth Inservice Testing Interval L-MT-22-010, License Amendment Request to Revise Technical Specification 3.6.1.8 Residual Heat Removal (RHR) Drywell Spray Header and Nozzle Surveillance Frequency2022-03-18018 March 2022 License Amendment Request to Revise Technical Specification 3.6.1.8 Residual Heat Removal (RHR) Drywell Spray Header and Nozzle Surveillance Frequency L-MT-22-012, Special Report for the Bypass of the Offgas Treatment Storage System2022-03-15015 March 2022 Special Report for the Bypass of the Offgas Treatment Storage System L-MT-22-008, 10 CFR 50.55a Request Associated with the Monticello Sixth Inservice Testing Ten-Year Interval Alternative Related to Excess Flow Check Valve Testing Frequency (L-MT-22-008)2022-03-0707 March 2022 10 CFR 50.55a Request Associated with the Monticello Sixth Inservice Testing Ten-Year Interval Alternative Related to Excess Flow Check Valve Testing Frequency (L-MT-22-008) L-MT-22-006, 10 CFR 50.55a Request Associated with the Monticello Sixth Inservice Testing Ten-Year Interval OMN-26 (L-MT-22-006)2022-02-18018 February 2022 10 CFR 50.55a Request Associated with the Monticello Sixth Inservice Testing Ten-Year Interval OMN-26 (L-MT-22-006) 2024-01-11
[Table view] Category:Report
MONTHYEARML23172A1112023-06-21021 June 2023 SLRA - Requests for Confirmation of Information - Set 1 L-MT-22-024, Response to a Request for Additional Information for the Monticello Nuclear Generating Plant Related to the Amendment to Adopt Advanced Framatome Methodologies2022-06-0606 June 2022 Response to a Request for Additional Information for the Monticello Nuclear Generating Plant Related to the Amendment to Adopt Advanced Framatome Methodologies L-MT-22-004, Technical Specification 5.6.4 Post Accident Monitoring Report2022-01-20020 January 2022 Technical Specification 5.6.4 Post Accident Monitoring Report ML21211A5962021-06-30030 June 2021 Attachments 9a, B, 10a, B, 11a, 12a and 13a and 13b, Including ANP-3933NP, Revision 0, Monticello ATWS-I Evaluation for Atrium 11 Fuel, Affidavit L-MT-21-016, Nuclear Material Transaction Report2021-03-25025 March 2021 Nuclear Material Transaction Report L-MT-20-002, 10 CFR 50.55a(z)(1) Request RR-015: Proposed Alternative for Examination of ASME Section XI, Examination Category B-G-1, Item Number B6.40, Threads in Flange2020-01-31031 January 2020 10 CFR 50.55a(z)(1) Request RR-015: Proposed Alternative for Examination of ASME Section XI, Examination Category B-G-1, Item Number B6.40, Threads in Flange ML19308A0562019-10-0808 October 2019 Enclosure 1 - Notification of Changes to the Emergency Response Data System (ERDS) L-MT-19-020, Report of Full Compliance with Phase 1 and Phase 2 of June 6, 2013 Commission Order Modifying Licenses with Regard to Reliable Hardened Containment Vents Capable of Operation Under Severe Accident Conditions ..2019-04-25025 April 2019 Report of Full Compliance with Phase 1 and Phase 2 of June 6, 2013 Commission Order Modifying Licenses with Regard to Reliable Hardened Containment Vents Capable of Operation Under Severe Accident Conditions .. L-MT-17-060, Steam Dryer Visual Inspection Report2017-08-10010 August 2017 Steam Dryer Visual Inspection Report L-MT-17-055, Seismic Mitigating Strategies Assessment (MSA) Report for the Reevaluated Seismic Hazard Information - NEI 12-06, Revision 4, Appendix H, H.4.4, Path 42017-07-26026 July 2017 Seismic Mitigating Strategies Assessment (MSA) Report for the Reevaluated Seismic Hazard Information - NEI 12-06, Revision 4, Appendix H, H.4.4, Path 4 ML17088A5272017-03-31031 March 2017 NSIR Dpcp Confirmatory Review Report, 10 CFR 50.54(p)(2) Changes to the Security Plan, Revision 16, Northern States Power Company, Monticello Nuclear Generating Plant L-MT-17-015, Mitigating Strategies Flood Hazard Assessment (MSA) Submittal2017-03-28028 March 2017 Mitigating Strategies Flood Hazard Assessment (MSA) Submittal ML16302A2462016-10-28028 October 2016 Spent Fuel Pool Evaluation Supplemental Report, Response to NRC Request for Information Pursuant to 10 CFR 50.54(f) Regarding Recommendation 2.1 of the Near-Term Task Force Review of Insights from Fukushima.. ML16221A2752016-07-31031 July 2016 ANP-3274NP, Revision 2, Analytical Methods for Monticello ATWS-I. L-MT-16-037, ANP-3284NP, Revision 1, Results of Analysis and Benchmarking of Methods for Monticello ATWS-I.2016-07-31031 July 2016 ANP-3284NP, Revision 1, Results of Analysis and Benchmarking of Methods for Monticello ATWS-I. L-MT-16-010, Enclosure 2 Areva Report ANP-3295NP, Rev. 3, Monticello Licensing Analysis for EFW (Epu/Mella+).2016-02-29029 February 2016 Enclosure 2 Areva Report ANP-3295NP, Rev. 3, Monticello Licensing Analysis for EFW (Epu/Mella+). ML15274A4742015-09-29029 September 2015 Enclosure 4, Engineering Evaluation, EC 25987, Calculation Framework for the Extended Flow Window Stability (Efws) Setpoints L-MT-15-065, Enclosure 8, Areva Report ANP-3424NP, Revision 1 to Areva Responses to RAI from Scvb on MNGP EFW LAR (Non-Proprietary) and Enclosure 9, Areva Affidavits2015-09-29029 September 2015 Enclosure 8, Areva Report ANP-3424NP, Revision 1 to Areva Responses to RAI from Scvb on MNGP EFW LAR (Non-Proprietary) and Enclosure 9, Areva Affidavits ML15274A4752015-09-29029 September 2015 Enclosure 6, Areva Report ANP-3435NP, Revision 0 to Are VA Responses to RAI-8 and RAI-32 from Srxb and Snpb on MNGP EFW LAR (Non-Proprietary) ML15175A3362015-07-0808 July 2015 Staff Assessment of Information Provided Pursuant to Title 10 of the Code of Federal Regulations Part 50, Section 50.54(f), Seismic Hazard Reevaluations for Recommendation 2.1 of the Near Term Task Force ML15169A3582015-06-11011 June 2015 PSP Rev 14 Technical Review 06-11-15 ML16035A1842015-01-30030 January 2015 Document 51-9234641-001, Technical Report of the Demonstration of UT NDE Procedure 54-UT-114-000, Phased Array Ultrasonic Examination of Dry Storage Canister Lid Welds. ML16035A1862015-01-30030 January 2015 54-PQ-114-001, Technical Justification, Phased Array Ultrasonic Examination of Dry Storage Canister Lid Welds. Part 2 of 3 ML16035A1852015-01-30030 January 2015 54-PQ-114-001, Technical Justification, Phased Array Ultrasonic Examination of Dry Storage Canister Lid Welds. Part 1 of 3 L-MT-14-103, Areva Report ANP-3376NP, Rev. 0, Supplement to Xcel Energy License Amendment Request for Areva Extended Flow Window, Enclosure 2 to L-MT-14-1032014-12-31031 December 2014 Areva Report ANP-3376NP, Rev. 0, Supplement to Xcel Energy License Amendment Request for Areva Extended Flow Window, Enclosure 2 to L-MT-14-103 ML16005A1102014-09-25025 September 2014 Redacted 2014 Decommissioning Cost Analysis for the Prairie Island Nuclear Generating Plant ML14269A3252014-09-0505 September 2014 LTR-BWR-ENG-14-037-NP, Alternate Power Ascension Process. ML14269A3232014-09-0505 September 2014 LTR-BWR-ENG-14-034-NP, Investigation Into the Cause of Exceeding the Level 1 (L1) and Level 2 (L2) Limit Curves Generated Based on 2011 Monticello Main Steam Line Strain Gauge Data. ML14204A6232014-07-18018 July 2014 Enclosure 3, Attachment - Response to the Us NRC Request for Additional Information Relative to the Monticello Replacement Steam Dryer Accoustic/Structural Ananlyses Set #7 ML14176A9612014-06-24024 June 2014 Submittal of Non-Proprietary BWROG Technical Product, BWROGTP-11-006 - ECCS Containment Walkdown Procedure, Rev 1 (January 2011), as Formally Requested During the Public Meeting Held on April 30, 2014 ML18100A1902014-05-22022 May 2014 Enclosure 8 - Structural Integrity Associates, Inc. Report 130415.403, Revision 2, Assessment of Monticello Spent Fuel Canister Closure Plate Welds Based on Welding Video Records ML14136A2892014-05-12012 May 2014 14C4229-RPT-001, Revision 3, Monticello Nuclear Generating Plant Seismic Hazard and Screening Report. L-MT-14-041, ANP-3304NP, Rev. 0, Areva Response to NRC Follow-Up on Srxb RAI-6: ASME Overpressure Analysis.2014-05-0909 May 2014 ANP-3304NP, Rev. 0, Areva Response to NRC Follow-Up on Srxb RAI-6: ASME Overpressure Analysis. ML14069A0052014-04-11011 April 2014 Staff Assessment of the Seismic Walkdown Report Supporting Implementation of NTTF Recommendation 2.3 Related to the Fukushima Dai-ichi Nuclear Power Plant Accident ML14014A2682014-02-28028 February 2014 Staff Review and Evaluation of the Fifth 10-Year Interval Inservice Inspection Program Plan ML14064A1862014-02-25025 February 2014 Structural Integrity Associates, Inc., Evaluation File No. 1301525.301, Monticello Shroud Support Structure Flaw Evaluation Review and Support Plate Weld Inspection Recommendations L-MT-14-003, Enclosures 2 & 3 to L-MT-14-003, ANP-3286NP, Rev. 0, Responses to RAI from Srxb on MNGP Transition to Areva Fuel & Areva Atrium 10XM Fuel Transition - Response to Requests for Additional Information2014-01-31031 January 2014 Enclosures 2 & 3 to L-MT-14-003, ANP-3286NP, Rev. 0, Responses to RAI from Srxb on MNGP Transition to Areva Fuel & Areva Atrium 10XM Fuel Transition - Response to Requests for Additional Information ML13358A3732013-12-31031 December 2013 Enclosure 1, Reload 26 Cycle 27 Supplemental Reload Licensing Report, 000N0154-SRLR, Revision 5, Extended Power Uprate and Maximum Extended Load Line Limit Plus L-MT-13-126, Enclosure 3, GE-MNGP-AEP-3306, Enclosure 2 GEH Response to Mella+ Eicb RAI and Affidavit2013-12-20020 December 2013 Enclosure 3, GE-MNGP-AEP-3306, Enclosure 2 GEH Response to Mella+ Eicb RAI and Affidavit ML13282A1422013-10-0404 October 2013 GE-MNGP-AEP-3304R1, Enclosure 2, GEH Response to Mella + RAI 2 L-MT-13-096, GE-MNGP-AEP-3304R1, Enclosure 4, NEDC-33435 Corrected Pages2013-10-0404 October 2013 GE-MNGP-AEP-3304R1, Enclosure 4, NEDC-33435 Corrected Pages ML13248A3462013-08-29029 August 2013 WCAP-17252-NP, Revision 4 - Acoustic Loads Definition for the Monticello Steam Dyer Replacement Project, Enclosure 12 ML13248A3452013-08-29029 August 2013 WCAP-17251-NP, Revision 2 - Monticello Replacement Steam Dryer Four-Line Acoustic Subscale Testing Report, Enclosure 11 ML13248A3442013-08-29029 August 2013 WCAP-17548-NP, Revision 2 - Signal Processing Performed on Monticello MSL Strain Gauge and Rsd Instrumentation Data, Enclosure 10 ML13248A3482013-08-29029 August 2013 WCAP-17549-NP, Revision 2 - Monticello Replacement Steam Dryer Structural Evaluation for High-Cycle Acoustic Loads Using ACE, Enclosure 13 L-MT-13-091, WCAP-17716-NP, Revision 1 - Benchmarking of the Acoustic Circuit Enhanced Revision 2.0 for the Monticello Steam Dryer Replacement Project, Enclosure 142013-08-29029 August 2013 WCAP-17716-NP, Revision 1 - Benchmarking of the Acoustic Circuit Enhanced Revision 2.0 for the Monticello Steam Dryer Replacement Project, Enclosure 14 ML13200A1922013-07-31031 July 2013 ANP-3092(NP), Rev. 0, Monticello Thermal-Hydraulic Design Report for Atrium 10XM Fuel Assemblies. ML13200A1962013-07-31031 July 2013 ANP-3211(NP), Rev. 1, Monticello EPU LOCA Break Spectrum Analysis for Atrium 10XM Fuel. ML13205A0852013-07-12012 July 2013 Enclosure 2 - Westinghouse Electric Co. Letter LTR-A&SA-13-14, NP-Attachment, Revision 0, Responses to the U.S. NRC Request for Additional Information Relative to the Monticello Replacement Steam Dryer Acoustic/Structural Analyses Set #5 ML13200A1902013-06-30030 June 2013 ANP-3224NP, Rev. 2, Applicability of Areva Np BWR Methods to Monticello. 2023-06-21
[Table view] Category:Technical
MONTHYEARL-MT-22-024, Response to a Request for Additional Information for the Monticello Nuclear Generating Plant Related to the Amendment to Adopt Advanced Framatome Methodologies2022-06-0606 June 2022 Response to a Request for Additional Information for the Monticello Nuclear Generating Plant Related to the Amendment to Adopt Advanced Framatome Methodologies L-MT-22-004, Technical Specification 5.6.4 Post Accident Monitoring Report2022-01-20020 January 2022 Technical Specification 5.6.4 Post Accident Monitoring Report ML21211A5962021-06-30030 June 2021 Attachments 9a, B, 10a, B, 11a, 12a and 13a and 13b, Including ANP-3933NP, Revision 0, Monticello ATWS-I Evaluation for Atrium 11 Fuel, Affidavit L-MT-21-016, Nuclear Material Transaction Report2021-03-25025 March 2021 Nuclear Material Transaction Report L-MT-20-002, 10 CFR 50.55a(z)(1) Request RR-015: Proposed Alternative for Examination of ASME Section XI, Examination Category B-G-1, Item Number B6.40, Threads in Flange2020-01-31031 January 2020 10 CFR 50.55a(z)(1) Request RR-015: Proposed Alternative for Examination of ASME Section XI, Examination Category B-G-1, Item Number B6.40, Threads in Flange ML19308A0562019-10-0808 October 2019 Enclosure 1 - Notification of Changes to the Emergency Response Data System (ERDS) L-MT-19-020, Report of Full Compliance with Phase 1 and Phase 2 of June 6, 2013 Commission Order Modifying Licenses with Regard to Reliable Hardened Containment Vents Capable of Operation Under Severe Accident Conditions ..2019-04-25025 April 2019 Report of Full Compliance with Phase 1 and Phase 2 of June 6, 2013 Commission Order Modifying Licenses with Regard to Reliable Hardened Containment Vents Capable of Operation Under Severe Accident Conditions .. L-MT-17-055, Seismic Mitigating Strategies Assessment (MSA) Report for the Reevaluated Seismic Hazard Information - NEI 12-06, Revision 4, Appendix H, H.4.4, Path 42017-07-26026 July 2017 Seismic Mitigating Strategies Assessment (MSA) Report for the Reevaluated Seismic Hazard Information - NEI 12-06, Revision 4, Appendix H, H.4.4, Path 4 L-MT-16-010, Enclosure 2 Areva Report ANP-3295NP, Rev. 3, Monticello Licensing Analysis for EFW (Epu/Mella+).2016-02-29029 February 2016 Enclosure 2 Areva Report ANP-3295NP, Rev. 3, Monticello Licensing Analysis for EFW (Epu/Mella+). ML15274A4752015-09-29029 September 2015 Enclosure 6, Areva Report ANP-3435NP, Revision 0 to Are VA Responses to RAI-8 and RAI-32 from Srxb and Snpb on MNGP EFW LAR (Non-Proprietary) L-MT-15-065, Enclosure 8, Areva Report ANP-3424NP, Revision 1 to Areva Responses to RAI from Scvb on MNGP EFW LAR (Non-Proprietary) and Enclosure 9, Areva Affidavits2015-09-29029 September 2015 Enclosure 8, Areva Report ANP-3424NP, Revision 1 to Areva Responses to RAI from Scvb on MNGP EFW LAR (Non-Proprietary) and Enclosure 9, Areva Affidavits ML15274A4742015-09-29029 September 2015 Enclosure 4, Engineering Evaluation, EC 25987, Calculation Framework for the Extended Flow Window Stability (Efws) Setpoints ML16035A1842015-01-30030 January 2015 Document 51-9234641-001, Technical Report of the Demonstration of UT NDE Procedure 54-UT-114-000, Phased Array Ultrasonic Examination of Dry Storage Canister Lid Welds. ML16035A1862015-01-30030 January 2015 54-PQ-114-001, Technical Justification, Phased Array Ultrasonic Examination of Dry Storage Canister Lid Welds. Part 2 of 3 ML16035A1852015-01-30030 January 2015 54-PQ-114-001, Technical Justification, Phased Array Ultrasonic Examination of Dry Storage Canister Lid Welds. Part 1 of 3 L-MT-14-103, Areva Report ANP-3376NP, Rev. 0, Supplement to Xcel Energy License Amendment Request for Areva Extended Flow Window, Enclosure 2 to L-MT-14-1032014-12-31031 December 2014 Areva Report ANP-3376NP, Rev. 0, Supplement to Xcel Energy License Amendment Request for Areva Extended Flow Window, Enclosure 2 to L-MT-14-103 ML16005A1102014-09-25025 September 2014 Redacted 2014 Decommissioning Cost Analysis for the Prairie Island Nuclear Generating Plant ML14269A3252014-09-0505 September 2014 LTR-BWR-ENG-14-037-NP, Alternate Power Ascension Process. ML14269A3232014-09-0505 September 2014 LTR-BWR-ENG-14-034-NP, Investigation Into the Cause of Exceeding the Level 1 (L1) and Level 2 (L2) Limit Curves Generated Based on 2011 Monticello Main Steam Line Strain Gauge Data. ML14176A9612014-06-24024 June 2014 Submittal of Non-Proprietary BWROG Technical Product, BWROGTP-11-006 - ECCS Containment Walkdown Procedure, Rev 1 (January 2011), as Formally Requested During the Public Meeting Held on April 30, 2014 ML18100A1902014-05-22022 May 2014 Enclosure 8 - Structural Integrity Associates, Inc. Report 130415.403, Revision 2, Assessment of Monticello Spent Fuel Canister Closure Plate Welds Based on Welding Video Records L-MT-14-041, ANP-3304NP, Rev. 0, Areva Response to NRC Follow-Up on Srxb RAI-6: ASME Overpressure Analysis.2014-05-0909 May 2014 ANP-3304NP, Rev. 0, Areva Response to NRC Follow-Up on Srxb RAI-6: ASME Overpressure Analysis. ML14069A0052014-04-11011 April 2014 Staff Assessment of the Seismic Walkdown Report Supporting Implementation of NTTF Recommendation 2.3 Related to the Fukushima Dai-ichi Nuclear Power Plant Accident ML14064A1862014-02-25025 February 2014 Structural Integrity Associates, Inc., Evaluation File No. 1301525.301, Monticello Shroud Support Structure Flaw Evaluation Review and Support Plate Weld Inspection Recommendations L-MT-14-003, Enclosures 2 & 3 to L-MT-14-003, ANP-3286NP, Rev. 0, Responses to RAI from Srxb on MNGP Transition to Areva Fuel & Areva Atrium 10XM Fuel Transition - Response to Requests for Additional Information2014-01-31031 January 2014 Enclosures 2 & 3 to L-MT-14-003, ANP-3286NP, Rev. 0, Responses to RAI from Srxb on MNGP Transition to Areva Fuel & Areva Atrium 10XM Fuel Transition - Response to Requests for Additional Information ML13358A3732013-12-31031 December 2013 Enclosure 1, Reload 26 Cycle 27 Supplemental Reload Licensing Report, 000N0154-SRLR, Revision 5, Extended Power Uprate and Maximum Extended Load Line Limit Plus ML13200A1962013-07-31031 July 2013 ANP-3211(NP), Rev. 1, Monticello EPU LOCA Break Spectrum Analysis for Atrium 10XM Fuel. ML13200A1922013-07-31031 July 2013 ANP-3092(NP), Rev. 0, Monticello Thermal-Hydraulic Design Report for Atrium 10XM Fuel Assemblies. ML13200A1952013-06-30030 June 2013 ANP-3213(NP), Rev. 1, Monticello Fuel Transition Cycle 28 Reload Licensing Analysis (Epu/Mellla). ML13200A1902013-06-30030 June 2013 ANP-3224NP, Rev. 2, Applicability of Areva Np BWR Methods to Monticello. ML13191B1282013-06-26026 June 2013 NEDO-33820, Revision 0, Monticello Nuclear Generating Plant Upper Shelf Energy Evaluation for Plate C2220 Material for 54 EFPY, Enclosure 5 ML13200A2002013-05-31031 May 2013 ANP-3139(NP), Rev. 1, Nuclear Fuel Design Report Monticello Cycle 28 Atrium 10XM Fuel. ML13200A1992013-05-31031 May 2013 ANP-3221NP, Rev. 0, Fuel Rod Thermal-Mechanical Design for Monticello Atrium 10XM Fuel Assemblies, Cycle 28. ML13200A1972013-05-31031 May 2013 ANP-3212(NP), Rev. 0, Monticello EPU LOCA-ECCS Analysis MAPLHGR Limits for Atrium 10XM Fuel. ML13200A1942013-05-31031 May 2013 ANP-3215(NP), Rev. 0, Monticello Fuel Transition Cycle 28 Fuel Cycle Design (Epu/Mellla). ML13092A3492013-03-29029 March 2013 Enclosure 10 to L-MT-13-029 - WCAP-17548-NP, Revision 1, Signal Processing Performed on Monticello MSL Strain Gauge and Rsd Instrumentation Data. L-MT-13-029, Enclosure 14 to L-MT-13-029 - WCAP-17716-NP, Revision 0, Benchmarking of the Acoustic Circuit Enhanced Revision 2.0 for the Monticello Steam Dryer Replacement Project.2013-03-29029 March 2013 Enclosure 14 to L-MT-13-029 - WCAP-17716-NP, Revision 0, Benchmarking of the Acoustic Circuit Enhanced Revision 2.0 for the Monticello Steam Dryer Replacement Project. ML13092A3502013-03-29029 March 2013 Enclosure 11 to L-MT-13-029 - WCAP-17251-NP, Revision 1, Monticello Replacement Steam Dryer Four-Line Acoustic Subscale Testing Report. ML13092A3512013-03-29029 March 2013 Enclosure 12 to L-MT-13-029 - WCAP-17252-NP, Revision 3, Acoustic Loads Definition for the Monticello Steam Dryer Replacement Project. ML13092A3522013-03-29029 March 2013 Enclosure 13 to L-MT-13-029 - WCAP-17549-NP, Revision 1, Monticello Replacement Steam Dryer Structural Evaluation for High-Cycle Acoustic Loads Using Ace. ML13037A2012013-01-31031 January 2013 NEDO-33800, Rev. 0, Monticello Nuclear Generating Plant Automatic Depressurization System Bypass Timer Extended Power Uprate. ML13200A1892012-11-30030 November 2012 ANP-2637, Rev. 4, Boiling Water Reactor Licensing Methodology Compendium. ML13200A1912012-10-31031 October 2012 ANP-3119NP, Rev. 0, Mechanical Design Report for Monticello Atrium 10XM Fuel Assemblies. ML12300A2202012-10-31031 October 2012 Attachment 10: BWROG-TP-12-018, Revision 0, Task 3 - Pump Operation at Reduced Npsha Conditions (Cvds Pump) L-MT-12-082, Extended Power Uprate and Maximum Extended Load Line Limit Analysis Plus License Amendment Requests: Supplement to Address Secy 11-0014 Use of Containment Accident Pressure2012-09-28028 September 2012 Extended Power Uprate and Maximum Extended Load Line Limit Analysis Plus License Amendment Requests: Supplement to Address Secy 11-0014 Use of Containment Accident Pressure ML12307A4342012-08-31031 August 2012 ANP-3113(NP), Rev. 0, Monticello Nuclear Plant Spent Fuel Storage Pool Criticality Safety Analysis for Atrium 10XM Fuel. ML12300A2242012-08-31031 August 2012 Attachment 13: BWROG-TP-12-014, Revision 0, Task 6 - Npshr Test Instrument Inaccuracy Effect on Published Results (Cvds Pump) ML12300A2222012-08-31031 August 2012 Attachment 12: BWROG-TP-12-013, Revision 0, Task 5 - Effects of Non-Condensible Gases on Seals (Cvds Pump) ML12300A2212012-08-31031 August 2012 Attachment 11: BWROG-TP-12-012, Revision 0, Task 4 - Operation in the Maximum Erosion Rate Zone (Cvds Pump) ML12300A2192012-08-31031 August 2012 Attachment 9: BWROG-TP-12-011, Revision 0, Task 2 - Equation for Pump Speed Correction (Cvds Pump) 2022-06-06
[Table view] |
Text
(l Xcel Energy*
RESP O N $ I 8 L E 8 Y H AT URE* 2807 West County Road 75 Monticello, MN 55362 January 31, 2020 L-MT-20-002 10 CFR 50.55a ATTN: Document Control Desk U.S. Nuclear Regulatory Commission Washington, DC 20555-0001 Monticello Nuclear Generating Plant Docket No. 50-263 Renewed Facility Operating License No. DPR-22 10 CFR 50.55a(z)(1) Request RR-015: Proposed Alternative for Examination of ASME Section XI, Examination Category B-G-1, Item Number B6.40, Threads in Flange Pursuant to 10 CFR 50.55a(z)(1), Alternatives to codes and standards requirements, the Northern States Power Company, a Minnesota corporation, doing business as Xcel Energy (hereafter NSPM), requests approval of an alternative to the examination requirements of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code,Section XI, Examination Category B-G-1, Item Number B6.40, Threads in Flange, on the basis that the proposed alternative provides an acceptable level of quality and safety.
Specifically, NSPM is requesting an alternative to the volumetric examination of the threads in the reactor pressure vessel flange that is required every interval. The basis for this alternative is provided in the enclosure.
The MNGP is currently operating in the fifth 10-year interval of the Inservice Inspection (ISI)
Program in compliance with the ASME Code,Section XI, 2007 Edition with the 2008 Addenda.
NSPM is submitting this 10 CFR 50.55a(z)(1) request for the remainder of the fifth 10-year ISI interval for MNGP, scheduled to end on May 31, 2022. NSPM requests approval of this alternative by February 28, 2021.
If there are any questions or if additional information is needed, please contact Mr. Richard Loeffler at (612) 342-8981 or Richard.Loeffler@xenuclear.com.
Document Control Desk Page2 Summary of Commitments This letter makes no new commitments and no revisions to existing commitments.
Christopher R. Church Site Vice President, Monticello Nuclear Generating Plant Northern States Power Company - Minnesota Enclosure cc: Administrator, Region Ill, USNRC Project Manager, Monticello, USNRC Resident Inspector, Monticello, USNRC
L-MT-20-002 NSPM Enclosure 10 CFR 50.55a(z)(1) Request RR-015 Proposed Alternative for Examination of ASME Section XI, Examination Category B-G-1, Item Number B6.40, Threads in Flange
1.0 ASME Code Components Affected
Code Class: ASME Section XI Code Class 1 Examination Category: B-G-1, Pressure Retaining Bolting, Greater than 2 in. (50 mm) in Diameter Item Number: B6.40, Threads in Flange
Description:
Reactor Vessel Threads in Flange 2.0 Applicable Code Edition and Addenda American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code,Section XI, Rules for lnservice Inspection and Testing of Components of Light-Water Cooled Plants, 2007 Edition with the 2008 Addenda. The Monticello Nuclear Generating Plant (MNGP) Fifth 10-Year Inservice Inspection (ISI) Interval began on September 1, 2012, and is scheduled to end on May 31, 2022.
ASME Section XI, Appendix VIII requirements are implemented as required by, and as modified by, 10 CFR 50.55a. Procedures and personnel are qualified to the Performance Demonstration Initiative (PDI). The PDI Program document meets the requirements of 10 CFR 50.55a up through the 2013 Edition of ASME Section XI.
3.0 Applicable Code Requirement The Reactor [Pressure] Vessel (RPV) Threads in Flange, Examination Category B-G-1, Item Number B6.40, are examined using a volumetric examination technique with 100 percent of the flange threaded stud holes examined every ISI interval. The examination area is the one-inch area around each RPV stud hole, as shown on ASME Section XI Figure IWB-2500-12.
4.0 Proposed Alternative and Reason for the Request Pursuant to 10 CFR 50.55a, Codes and standards, paragraph (z)(1), Alternatives to codes and standards requirements, Northern States Power Company (NSPM) is requesting an alternative to the requirement under Section XI of the ASME Code, Examination Category B-G-1, Item Number B6.40, Threads in Flange, to perform in-service ultrasonic examinations of the RPV flange threads every interval. The Page 1 of 13
L-MT-20-002 NSPM Enclosure proposed 10 CFR 50.55a(z)(1) alternative is to eliminate this inspection requirement.
The basis for elimination of these examinations is Electric Power Research Institute (EPRI) Technical Report (TR) No. 3002010354, entitled, Reactor Pressure Vessel (RPV) Threads in Flange Examination Requirements (Reference 1). Approval of this proposed alternative eliminates performance of unnecessary volumetric inspections of the RPV flange threads at the Monticello Nuclear Generating Plant (MNGP). This 10 CFR 50.55a(z)(1) alternative provides for more efficient conduct of plant refueling outages while maintaining an acceptable level of quality and safety.
Licensees within the United States and internationally have worked with EPRI and in 2017 produced the above referenced final report (TR No. 3002010354). The final report includes a survey of inspection results from 168 nuclear units that responded (including domestic and international units), a review of operating experience related to RPV flange / bolting, and a flaw tolerance evaluation. The conclusion from this evaluation was that these ASME Code Section XI examinations had not been identifying any service-induced degradation and the associated impact on worker exposure, personnel safety, critical path time, and additional time at reduced water inventory was not commensurate with performance.
Potential Degradation Mechanisms An evaluation of potential degradation mechanisms that could impact flange/threads reliability is described in the 2017 EPRI report. Potential types of degradation evaluated included pitting, intergranular attack, corrosion fatigue, stress corrosion cracking, crevice corrosion, velocity phenomena, dealloying corrosion, general corrosion, stress relaxation, creep, mechanical wear, and mechanical / thermal fatigue. Other than the potential for mechanical / thermal fatigue, there are no active degradation mechanisms identified for the threads in flange component.
The final EPRI report also notes a general conclusion from ASME's Risk-Based Inspection: Development of Guidelines, (Reference 2) that when a component item has no active degradation mechanism present and a preservice inspection has confirmed that the inspection volume is in good condition (i.e., contains no flaws /
indications), then subsequent in-service inspections do not provide additional value going forward. As explained in the final EPRI report, the RPV flange ligaments have not only received the required pre-service examinations, but more than 10,000 ISIs have been carried out with no relevant findings.
To address the potential for mechanical / thermal fatigue, the EPRI report documents a stress analysis and flaw tolerance evaluation of the flange thread area to assess mechanical / thermal fatigue potential. The evaluation consists of two parts. In the first part, a stress analysis is performed considering all applicable loads on the threads in flange component. In the second part, the stresses at the critical locations of the component are used in a fracture mechanics evaluation to determine the allowable flaw Page 2 of 13
L-MT-20-002 NSPM Enclosure size for the component as well as how much time it will take for a postulated initial flaw to grow to the allowable flaw size using guidelines in the ASME Code,Section XI, IWB-3600, Analytical Evaluation of Flaws.
Stress Analysis As discussed in the EPRI report, a stress analysis was performed to determine the stresses at critical regions of the threads in flange component as input to a flaw tolerance evaluation. A bounding finite element model was developed to represent a typical threads in flange component for the fleet of nuclear plants. To create a representative geometry for the finite element model, a PWR design was selected as a representative geometry because of its higher design pressure and temperature. The largest RPV diameter of the PWRs was used along with the largest bolts and the highest number of bolts. The larger and more numerous bolt configuration results in less flange material between bolt holes, whereas the larger RPV diameter results in higher pressure and thermal stresses.
The details of the RPV parameters for MNGP as compared to the bounding values used in this evaluation are shown in Tables 1, 2, and 3. Table 1 provides a comparison of basic dimensions and loads. As this table shows, not all MNGP parameters are bounded by the parameters evaluated in the EPRI report; however, the preload stress for MNGP is bounded by that specified within the EPRI report. Specifically, the preload stress in the EPRI report is 42,338 psi whereas for MNGP the preload stress is 36,589 psi. Also, the design pressure (1,250 psi) for MNGP, which is a BWR, is considerably less than that of the PWR design pressure (2,500 psi) used in the analysis. Considering the preload stress and design pressure, the stress per stud for MNGP is smaller than that used in the analysis of the EPRI report. Therefore, the MNGP stress is bounded by the stress used in the analysis described in the EPRI report. The dimensions of the analyzed geometry are shown in Figure 1 in the attachment.
Table 2 provides a comparison of thread specifications. As this table shows, the flange hole diameter discussed within the EPRI report analysis is larger than the flange hole diameter at the MNGP. The larger hole diameter results in a smaller remaining ligament between stud holes and is therefore conservative. The pitch of the threads used in the analysis in the EPRI report is identical to the pitch of the threads at MNGP.
The depth of the threads at MNGP is slightly greater than that presented in the EPRI report analysis, resulting in lower stress in the threads at MNGP. Therefore, the MNGP configuration is bounded by the configuration used in the EPRI report analysis.
Table 3 provides a comparison of fracture toughness parameters between the controlling stud preload conditions used in the analysis and those at MNGP. As this table shows, since the Reference Temperature for Nil Ductility Transition (RTNDT) for the MNGP flange (10°F), is much lower than the RTNDT used in the analysis (60°F), the Page 3 of 13
L-MT-20-002 NSPM Enclosure fracture toughness for the MNGP is much greater than that used in the analysis discussed in the EPRI report.
In summary, the comparisons shown in Tables 1, 2, and 3 demonstrate that the stress analysis discussed in the EPRI report is applicable to MNGP.
Table 1: Comparison of MNGP Parameters to Bounding Values Used in the Analysis Nominal Flange No. of Minimum RPV Inside Diameter Thickness Design Preload Studs No. of Diameter at Plant of Flange at Flange Pressure Stress Currently Studs Flange Hole Hole Hole (psig) (psi)
Installed Evaluated (inches)
(inches) (inches)
MNGP 64 64 6.0 206.375 10.625 1, 250 36,589 Bounding Values 54 NA 7.0 173 16 2,500 42,338 Used in Analysis Table 2: Comparison of MNGP Flange Thread Parameters to Bounding Values Used in the Analysis Flange Thread Flange Thread Flange Thread Depth Plant Pitch Specification (inches)
(thread/inch)
MNGP 6-8UN-2B 8 0.067500 Bounding Values 7-8N-2B 8 0.065000 Used in Analysis Table 3: Comparison of MNGP Fracture Toughness Parameters to Bounding Values Used in the Analysis Flange Temp Flange RTNDT Flange T- RTNDT Flange Fracture Plant During Bolt (oF) (oF) Toughness (ksiin)
Preload (oF)
MNGP 10 60 50 89.6 Bounding Values Used 60 60 0 53.9 in Analysis The analytical model for the stress analysis discussed in the EPRI report is shown in Figures 2 and 3 of the attachment. The loads considered in the analysis consisted of:
Page 4 of 13
L-MT-20-002 NSPM Enclosure
- A design pressure of 2500 psig at an operating temperature of 600°F was applied to all internal surfaces exposed to internal pressure.
- Bolt/stud preload - The preload on the bounding geometry is calculated as:
Ppreload = C x P x ID2 = 1.1 x 2500 x 1732 = 42,338 psi (291.9 MPa)
S x D2 54 x 62 Where, Ppreload = Preload pressure to be applied on modeled bolt (psi)
P = Internal pressure (psi)
ID = Largest inside diameter of RPV (inch)
C = Bolt-up contingencies (+10%)
S = Least number of studs D = Smallest stud diameter (inch)
- Thermal stresses - The only significant transient affecting the bolting flange is heat-up/cooldown. This transient typically consists of a steady 100°F/hour ramp up to the operating temperature with a corresponding pressure ramp up to the operating pressure.
The ANSYS finite element analysis program was used to determine the stresses in the threads in flange component for the three loads described above.
Flaw Tolerance Evaluation A flaw tolerance evaluation was performed using the results of the stress analysis described in the EPRI report to determine how long it would take an initial postulated flaw to reach the ASME Code,Section XI allowable flaw size. A linear elastic fracture mechanics evaluation consistent with ASME Code,Section XI, IWB-3600 was performed.
Stress intensity factors (Ks) at four flaw depths of 360° inside-surface-connected, partial-through-wall circumferential flaws are calculated using finite element analysis (FEA) techniques with the model described above. The maximum stress intensity factor (K) values around the bolt hole circumference for each flaw depth (a) are extracted and used to perform the crack growth calculations. The circumferential flaw is modeled to start between the 10th and 11th flange threads from the top end of the flange because that is where the largest tensile axial stress occurs. The modeled flaw depth-to-wall thickness ratios (a/t) are 0.02, 0.29, 0.55, and 0.77, as measured in any direction from the stud hole. This creates an ellipsoidal flaw shape around the circumference of the flange as shown in Figure 4 of the attachment for the flaw model with a/t = 0.77 a/t crack Page 5 of 13
L-MT-20-002 NSPM Enclosure model. The crack tip mesh for the other flaw depths follows the same pattern. When preload is not being applied, the stud, stud threads, and flange threads are not modeled. The model is otherwise unchanged between load cases.
The maximum K results are summarized in Table 4 for four crack depths. Because the crack tip varies in depth around the circumference, the maximum K from all locations at each crack size is conservatively used for the K-vs.-a profile.
Table 4: Maximum K vs. a/t K at Crack Depth (ksiin)
Load 0.02 a/t 0.29 a/t 0.55 a/t 0.77 a/t Case 1 Preload 11.2 17.4 15.5 13.9 Case 2 Preload + Heatup + Pressure 13.0 19.8 16.1 16.3 As shown in Table 4, two load cases were considered in the evaluation.
- 1. Preload
- 2. Preload + Heatup + Pressure Case 1 involving only Preload is limiting since the operating temperature for this case is 60°F while the operating temperature for Case 2 is approximately 550°F. The value of (T - RTNDT) for Case 1 for the MNGP RPV flange is 50°F as shown in Table 3.
From the equations in paragraph A-4200, Fracture Toughness, of the ASME Code,Section XI, Appendix A, the corresponding value of fracture toughness, KIc, is 89.6 ksiin for the MNGP flange. As discussed in the Safety Evaluation attached to Reference 4, for evaluation of a postulated flaw such as that considered in the analysis, a structural factor of 2, consistent with Section XI, Appendix G, Fracture Toughness Criteria for Protection Against Failure, of the ASME Code, can be used instead of the 10 for detected flaws in IWB-3610 of ASME Code,Section XI. If the structural factor of 2 is used, the allowable KIc is 44.8 ksiin. This is much greater than the maximum stress intensity factor (KI) for the bolt-up condition of 17.4 ksiin obtained in the analysis shown in Table 4, indicating considerable margin.
Page 6 of 13
L-MT-20-002 NSPM Enclosure As Table 4 shows, the allowable stress intensity factor is not exceeded for all crack depths up to the deepest analyzed flaw of a/t = 0.77. Hence the allowable flaw depth of the 360° circumferential flaw is at least 77% of the thickness of the flange. The allowable flaw depth is assumed to be equal to the deepest modeled crack for the purposes of this analysis.
For the crack growth evaluation, an initial postulated flaw size of 0.2 inch (5.08 mm) is chosen consistent with the ASME Code,Section XI IWB-3500, Acceptance Standards, for flaws. The deepest flaw analyzed is a/t = 0.77 because of the inherent limits of the model. Two load cases are considered for fatigue crack growth: heatup/cooldown and bolt preload. The heat-up/cooldown load case includes the stresses due to thermal and internal pressure loads and is conservatively assumed to occur 50 times per year. The bolt preload is assumed to be present and constant during the load cycling of the heat-up/cooldown load case. The bolt preload load case is conservatively assumed to occur five times per year, and these cycles do not include thermal or internal pressure. The resulting crack growth was determined to be negligible due to the small delta K and the relatively low number of cycles associated with the transients evaluated. Because the crack growth is insignificant, the allowable flaw size will not be reached and the integrity of the component is not challenged for at least 80 years (original 40-year design life plus additional 40 years of plant life extension).
The stress analysis and flaw tolerance evaluation presented above show that the threads in flange component at MNGP is very flaw tolerant and can operate for 80 years without exceeding ASME Code,Section XI safety margins. This clearly demonstrates that the threads in flange component examinations can be eliminated without affecting the safety of the RPV.
Operating Experience Review Summary As discussed above, the results of the survey discussed in the EPRI report confirmed that the RPV threads in flange examinations are adversely impacting outage activities, such as worker exposure, personnel safety, and critical path time while not identifying any service induced degradation. Specifically, for the U.S. fleet, a total of 94 nuclear units have responded and none of these units have identified any type of degradation.
As can be seen in Table 5 below, the data is encompassing. The 94 units represent data from 33 BWRs and 61 PWRs. For the BWR units, a total of 3,793 examinations were conducted; and for the PWR units, a total of 6,869 examinations were conducted with no service induced degradation identified. The BWR response data includes information from all of the plant designs in operation in the United States and includes the BWR-2, -3, -4, -5, and -6 designs. The PWR response data includes the 2-loop, 3-loop, and 4-loop designs and each of the PWR NSSS (Nuclear Steam Supply System) designs (i.e., Babcock and Wilcox, Combustion Engineering, and Westinghouse).
Page 7 of 13
L-MT-20-002 NSPM Enclosure Table 5: Summary of Survey Results - United States Fleet Number of Number of Plant Type Number of Units Reportable Examinations Indications BWR 33 3,793 0 PWR 61 6,869 0 Total 94 10,662 0 Related RPV Assessments In addition to the examination history and flaw tolerance discussed above, the EPRI report discusses studies conducted in response to the issuance of the Anticipated Transient Without Scram (ATWS) Rule by the NRC. This rule was issued to require design changes to reduce expected ATWS frequency and consequences. Many studies have been conducted to understand the ATWS phenomena and key contributors to successful response to an ATWS event. In particular, the reactor coolant system (RCS) and its individual components were reviewed to determine weak links.
As an example, even though significant structural margin was identified in NRC SECY-83-293, Amendments to 10CFR50 Related to Anticipated Transients Without Scram (ATWS) Events, dated July 19, 1983, for PWRs, the ASME Service Level C pressure of 3200 psig was assumed to be an unacceptable plant condition. While a higher ASME service level might be defensible for major RCS components, other portions of the RCS could deform to the point of inoperability. Additionally, there was the concern that steam generator tubes might fail before other RCS components with a resultant bypass of containment. The key take-away for these studies is that the RPV flange ligament was not identified as a weak link and other RCS components were significantly more limiting. Thus, there is substantial structural margin associated with the RPV flange.
As the EPRI report discusses, the RPV threads in flange are performing with very high reliability, based on operating and examination experience. This is due to the robust design and a relatively benign operating environment (e.g., the number and magnitude of transients are small, and the threads are generally not in contact with primary water at plant operating temperatures/pressures, etc.). The robust design is manifested in that plant operation has been allowed at several plants, even with a bolt/stud assumed to be out of service. As such, significant degradation of multiple bolts/threads would be needed prior to any RCS leakage.
Page 8 of 13
L-MT-20-002 NSPM Enclosure Control of Non-Service Induced Degradation To protect against non-service related degradation, NSPM uses detailed procedures for RPV disassembly and reassembly to ensure protection and care of the studs and flange, including threads in flange. Tensioning and detensioning are performed in multiple passes. Protective covers and guide caps are used on the studs and the RPV flange. Removed studs and nuts are stored out of the vessel work area until needed for reassembly activities. Clearances and alignment are verified and observed to ensure stud damage does not occur when moving the RPV head and internals.
Threads, stud holes, and the flange surfaces are cleaned and threads are lubricated prior to reassembly. All activities are performed and documented in each step of their applicable procedures. These controlled maintenance activities provide further assurance that degradation is mitigated and detected prior to returning the reactor to service.
5.0 Proposed Alternative and Basis for Use Pursuant to 10 CFR 50.55a, Codes and standards, paragraph (z)(1), Alternatives to codes and standards requirements, NSPM is requesting an alternative to the requirement under Section XI of the ASME Code, Examination Category B-G-1, Item Number B6.40, Threads in Flange, to perform in-service ultrasonic examinations of the RPV flange threads every interval. The proposed alternative is to eliminate this inspection requirement.
EPRI report TR No. 3002010354, Reactor Pressure Vessel (RPV) Threads in Flange Examination Requirements, provides the technical basis for the elimination of the RPV threads in flange examination requirement at the MNGP as described therein when coupled with the plant-specific information provided as the basis of this request. This report was developed because evidence had suggested that there have been no occurrences of service-induced degradation and there are negative impacts on worker exposure, personnel safety, and outage critical path time from performance of these examinations. Approval of this proposed alternative will eliminate performance of unnecessary volumetric inspections of the RPV flange threads at the MNGP.
Since there is reasonable assurance that the proposed alternative is an acceptable alternate approach to the performance of the ultrasonic examinations, NSPM requests authorization to use the proposed alternative pursuant to 10 CFR 50.55a(z)(1) on the basis that application of the alternative provides an acceptable level of quality and safety.
Page 9 of 13
L-MT-20-002 NSPM Enclosure 6.0 Duration of the Proposed Alternative This 10 CFR 50.55a(z)(1) alternative will be applied for the duration of the inservice inspection interval defined in Section 2 of this request or until such time as the NRC approves an applicable alternative in Regulatory Guide 1.147, Inservice Inspection Code Case Acceptability, ASME Section XI, Division 1, or another document.
7.0 Precedent The NRC has authorized use of an alternate to examination of the RPV threads in flange for several utilities whose plants include both BWRs and PWRs based on the earlier version of the EPRI report (Reference 9). A partial list is provided below. Donald C. Cook Nuclear Plant, Units 1 and 2, (Reference 8), has been more recently authorized based upon the current version of the EPRI report. It should be noted that the MNGP is a BWR-3 design similar to several BWR-3s listed below (i.e., Dresden Units 2 and 3, and Quad Cities Units 1 and 2) for which alternate examinations have been authorized by the NRC.
Southern Nuclear Operating Company, Inc. (Reference 3)
- Vogtle Electric Generating Plant, Units 1and 2 (PWRs)
- Joseph M. Farley Nuclear Plant, Unit 1 (PWR)
Exelon Generation (Reference 4)
- Braidwood Station, Units 1 and 2 (PWR)
- Dresden, Units 2 and 3 (BWR)
- Byron Station Units, 1 and 2 (PWR)
- Quad Cities, Units 1 and 2 (BWR)
- Peach Bottom, Units 2 and 3 (BWR)
- Three Mile Island, Unit 1 (PWR)
- Limerick, Units 1 and 2 (BWR)
- Calvert Cliff Nuclear Power Plant, Units
- Nine Mile Point, Units 1 and 2 (BWR) 1 and 2 (PWR)
- Clinton Power Station (BWR)
Duke Energy (Reference 5)
- Catawba Nuclear Station, Unit 2 (PWR)
- Brunswick Steam Electric Plant, Unit 1 (BWR)
- Shearon Harris Nuclear Power Plant,
- H. B. Robinson Steam Electric Plant, Unit 1 (PWR) Unit 2 (PWR)
- McGuire Nuclear Station, Units 1 and 2
- Oconee Nuclear Station, Units 1, 2 (PWR) and 3 (PWR)
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L-MT-20-002 NSPM Enclosure Dominion Nuclear Connecticut, Inc. (References 6 and 7)
- Millstone Power Station, Units 2 and 3 (PWR)
- North Anna Power Station, Units 1 and 2 (PWR)
Indiana Michigan Power Company (Reference 8)
- Donald C. Cook Nuclear Plant, Units 1 and 2 (PWRs)
Page 11 of 13
L-MT-20-002 NSPM Enclosure 8.0 References
- 1. Electric Power Research Institute (EPRI) Technical Report (TR)
No. 3002010354, Reactor Pressure Vessel (RPV) Threads in Flange Examination Requirements, Final Report, dated December 2017
- 2. American Society of Mechanical Engineers, Risk-Based Inspection:
Development of Guidelines, Volume 2-Part 1 and Volume 2-Part 2, Light Water Reactor (LWR) Nuclear Power Plant Components. CRTD-Vols. 20-2 and 20-4, ASME Research Task Force on Risk-Based Inspection Guidelines, Washington, D.C., 1992 and 1998
- 3. Letter from M. T. Markley (NRC) to C. R. Pierce (Southern Nuclear Operating Co. Inc.), Vogtle Electric Generating Plant, Units 1 and 2, and Joseph M. Farley Nuclear Plant, Unit 1 - Alternative to Inservice Inspection Regarding Reactor Pressure Vessel Threads Inflange Inspection (CAC Nos. MF8061, MF8062, MF8070), dated January 26, 2017 (ADAMS Accession No. ML17006A109)
- 4. Letter from D. J. Wrona (NRC) to B. C. Hanson (Exelon Generating Company, LLC), Braidwood Station, Units 1 and 2; Byron Station, Unit Nos. 1 and 2; Calvert Cliffs Nuclear Power Plant, Units 1 and 2; Clinton Power Station, Unit No. 1; Dresden Nuclear Power Station, Units 2 and 3; Limerick Generation Station, Units 1 and 2; Nine Mile Point Nuclear Station, Units 1 and 2; Peach Bottom Atomic Power Station, Units 2 and 3; Quad Cities Nuclear Power Station, Units 1 and 2; R. E. Ginna Nuclear Power Plant; and Three Mile Island Nuclear Station, Unit 1 - Proposed Alternative to Eliminate Examination of Threads in Reactor Pressure Vessel Flange (CAC Nos. MF8712-MF8729 and MF9548),
dated June 26, 2017 (ADAMS Accession No. ML17170A013)
- 5. Letter from U. Shoop (NRC) to S. Capps (Duke Energy), Brunswick Steam Electric Plant, Unit No. 1; Catawba Nuclear Station, Unit No. 2; Shearon Harris Nuclear Power Plant, Unit No. 1; McGuire Nuclear Station, Unit Nos. 1 and 2; Oconee Nuclear Station, Unit Nos. 1, 2 and 3; and H. B. Robinson Steam Electric Plant, Unit No. 2 - Alternative to Inservice Inspection Regarding Reactor Pressure Vessel Threads in Flange Inspection (CAC Nos. MF9513 - MF9521; EPID L-2017-LLR-0019, dated December 26, 2017 (ADAMS Accession No. ML17331A086)
- 6. Letter from J. G. Danna (NRC) to D. G. Stoddard (Dominion Nuclear Connecticut, Inc.), Millstone Power Station, Unit Nos. 2 and 3 - Alternative Requests RR-04-24 and IR-3-30 for Elimination of the Reactor Pressure Vessel Page 12 of 13
L-MT-20-002 NSPM Enclosure Threads in Flange Examination (CAC Nos. MF8468 and MF8469), dated May 25, 2017 (ADAMS Accession No. ML17132A187)
- 7. Letter from M. T. Markley (NRC) to D. G. Stoddard (Dominion), North Anna Power Station, Units 1 and 2 - Proposed Inservice Inspection Alternatives N1-14-NDE-009 and N2-14-NDE-004 (CAC Nos. MF9298 and MF9299; EPID L-2016-LLR-0018), dated December 6, 2017 (ADAMS Accession No. ML17132A663)
- 8. Letter from D. J. Wrona (NRC) to J. P. Gebbie (Indiana Michigan Power Company), Donald C. Cook Nuclear Plant, Units 1 and 2 - Proposed Alternative Request for Elimination of the Reactor Pressure Vessel Threads in Flange Examination EPID L-2018-LLR-0084), dated December 11, 2018, (ADAMS Accession No. ML18337A394)
- 9. EPRI TR No. 300200626, Nondestructive Evaluation: Reactor Pressure Vessel Threads in Flange Examination Requirements, dated March 2016 (ADAMS Accession No. ML16221A068)
Page 13 of 13
ATTACHMENT 1 MONTICELLO NUCLEAR GENERATING PLANT 10 CFR 50.55a(z)(1) REQUEST RR-015 PROPOSED ALTERNATIVE FOR EXAMINATION OF ASME SECTION XI, EXAMINATION CATEGORY B-G-1, ITEM NUMBER B6.40, THREADS IN FLANGE FIGURES 1-4 (4 Pages Follow)
Figure 1 Modeled Dimensions R86.5J/ ! 8 .5" 12.0" I*
17 .Ou 7.0" 16.0" L
R83 .75" R4.5u
- 10. 75" R85 .69u I in. = 25.4 lllill
Figure 2 Finite Element Model Showing Bolt and Flange Connection ELENE11ITS RE.DiL NUM J>.ID Vessel Flanqe
Figure 3 Finite Element Model Mesh with Detail at Thread Location I
I I
.....-I 1
I I
1, I
I I
I I
I I
IIJ,
Figure 4 Cross Section of Circumferential Flaw with Crack Tip Element Inserted After 10th Thread from Top of Flange