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MONTHYEARML18121A0312018-04-30030 April 2018 NRR E-mail Capture - Monticello Nuclear Generating Plant - Acceptance of Requested Licensing Action License Amendment Implementation of the Provisions of 10 CFR 50.69 Project stage: Acceptance Review L-MT-19-018, Response to Request for Additional Information: Application to Adopt 10 CFR 50.69, Risk- Informed Categorization & Treatment of Structures, Systems and Components for Nuclear Power Reactors2019-03-13013 March 2019 Response to Request for Additional Information: Application to Adopt 10 CFR 50.69, Risk- Informed Categorization & Treatment of Structures, Systems and Components for Nuclear Power Reactors Project stage: Response to RAI L-MT-19-030, Supplement to a Response for a Request for Additional Information: Application to Adopt 10 CFR 50.69, Risk-Informed Categorization and Treatment of Structures, Systems and Components for Nuclear Power Reactors.2019-05-15015 May 2019 Supplement to a Response for a Request for Additional Information: Application to Adopt 10 CFR 50.69, Risk-Informed Categorization and Treatment of Structures, Systems and Components for Nuclear Power Reactors. Project stage: Supplement ML19176A4212019-08-29029 August 2019 Issuance of Amendment No. 203 Adoption of 10 CFR 50.69 Project stage: Approval ML19305A8012019-11-12012 November 2019 Correction Letter for Amendment No. 203 to Renewed Operating License No. DPR-22 Project stage: Other 2019-03-13
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Category:Letter type:L
MONTHYEARL-MT-24-028, Response to RCI for RR-017 ISI Impracticality2024-08-28028 August 2024 Response to RCI for RR-017 ISI Impracticality L-MT-24-022, – Preparation and Scheduling of Operator Licensing Examinations2024-07-0909 July 2024 – Preparation and Scheduling of Operator Licensing Examinations L-MT-24-019, Submittal of ASME Section XI, Section IWB-3720 Analytical Evaluation in Accordance with 10 CFR 50.55a(b)(2)(xliii)2024-06-10010 June 2024 Submittal of ASME Section XI, Section IWB-3720 Analytical Evaluation in Accordance with 10 CFR 50.55a(b)(2)(xliii) L-MT-24-017, Response to Request for Additional Information Regarding License Amendment Request to Revise Monticello Technical Specification Surveillance Requirement 3.8.6.6 (EPID-L-2023-LLA-01602024-06-0404 June 2024 Response to Request for Additional Information Regarding License Amendment Request to Revise Monticello Technical Specification Surveillance Requirement 3.8.6.6 (EPID-L-2023-LLA-0160 L-MT-24-015, Response to Request for Additional Information - Alternative Request VR-09 for OMN-172024-05-16016 May 2024 Response to Request for Additional Information - Alternative Request VR-09 for OMN-17 L-MT-24-013, 2023 Annual Radiological Environmental Operating Report2024-05-14014 May 2024 2023 Annual Radiological Environmental Operating Report L-MT-24-016, 2023 Annual Report of Individual Monitoring for the Monticello Nuclear Generating Plant (MNGP)2024-05-0808 May 2024 2023 Annual Report of Individual Monitoring for the Monticello Nuclear Generating Plant (MNGP) L-MT-24-002, Submittal of Revision 41 to the Updated Safety Analysis Report2024-04-17017 April 2024 Submittal of Revision 41 to the Updated Safety Analysis Report L-MT-24-004, Submittal of 10 CFR 72.48 Report for Monticello Nuclear Generating Plant (MNGP)2024-04-17017 April 2024 Submittal of 10 CFR 72.48 Report for Monticello Nuclear Generating Plant (MNGP) L-MT-24-006, Subsequent License Renewal Application Annual Update 1 and Supplement 92024-02-29029 February 2024 Subsequent License Renewal Application Annual Update 1 and Supplement 9 L-MT-24-007, Response to NRC Request for Additional Information on Proposed Inservice Inspection Alternative RR-002 for the Sixth Ten-Year ISI Interval2024-02-15015 February 2024 Response to NRC Request for Additional Information on Proposed Inservice Inspection Alternative RR-002 for the Sixth Ten-Year ISI Interval L-MT-24-001, 10 CFR 50.55a Request RR-017 - Inservice Inspection (ISI) Impracticality in Accordance with 10 CFR 50.55a(g)(5)(iii) During the Fifth Ten-Year Interval2024-01-30030 January 2024 10 CFR 50.55a Request RR-017 - Inservice Inspection (ISI) Impracticality in Accordance with 10 CFR 50.55a(g)(5)(iii) During the Fifth Ten-Year Interval 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 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-23-032, 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-23-028, 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-23-031, 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 2024-08-28
[Table view] Category:Response to Request for Additional Information (RAI)
MONTHYEARL-MT-24-017, Response to Request for Additional Information Regarding License Amendment Request to Revise Monticello Technical Specification Surveillance Requirement 3.8.6.6 (EPID-L-2023-LLA-01602024-06-0404 June 2024 Response to Request for Additional Information Regarding License Amendment Request to Revise Monticello Technical Specification Surveillance Requirement 3.8.6.6 (EPID-L-2023-LLA-0160 L-MT-24-015, Response to Request for Additional Information - Alternative Request VR-09 for OMN-172024-05-16016 May 2024 Response to Request for Additional Information - Alternative Request VR-09 for OMN-17 ML24092A2152024-03-27027 March 2024 Email: Requests for Information to Support Deis ML24088A2152024-02-20020 February 2024 Feb. 20, 2024 Xcel Energy Subsequent License Renewal Application Response to Request for Additional Information and Request for Confirmation of Information for Monticello Nuclear Generating Plant L-MT-24-007, Response to NRC Request for Additional Information on Proposed Inservice Inspection Alternative RR-002 for the Sixth Ten-Year ISI Interval2024-02-15015 February 2024 Response to NRC Request for Additional Information on Proposed Inservice Inspection Alternative RR-002 for the Sixth Ten-Year ISI Interval 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-23-049, 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-23-046, 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-23-037, 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 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-23-034, 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-23-005, 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) ML22220A2722022-08-0808 August 2022 Response to a Request for Additional Informational Regarding the Monticello Fuel Oil Storage Tank Inspection 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) ML22161A9152022-06-10010 June 2022 and Prairie Island Nuclear Generating Plant, Units 1 and 2 - Response to a Request for Additional Information Xcel Energy Amendment Request to Create a Common Emergency Plan and Emergency Operations. 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-21-017, Response to Request for Additional Information (RAI) Related to License Amendment Request to Implement Technical Specifications Task Force Traveler TSTF-505, Revision 22021-04-20020 April 2021 Response to Request for Additional Information (RAI) Related to License Amendment Request to Implement Technical Specifications Task Force Traveler TSTF-505, Revision 2 L-MT-20-036, Response to Request for Additional Information Related to License Amendment Request to Implement Technical Specification Task Force Traveler TSTF-505, Revision 2, Provide Risk-Informed Extended Completion Times2020-12-21021 December 2020 Response to Request for Additional Information Related to License Amendment Request to Implement Technical Specification Task Force Traveler TSTF-505, Revision 2, Provide Risk-Informed Extended Completion Times L-MT-20-024, Response to Request for Additional Information (RAI) Monticello 10 CFR 50.55a Request: Request RR-016 Associated with the Fifth Ten-Year Inservice Inspection (ISI) Interval2020-07-20020 July 2020 Response to Request for Additional Information (RAI) Monticello 10 CFR 50.55a Request: Request RR-016 Associated with the Fifth Ten-Year Inservice Inspection (ISI) Interval L-MT-20-015, Response to Request for Additional Information (RAI) Long-Term Replacement Steam Dryer Inspection Plan2020-06-0808 June 2020 Response to Request for Additional Information (RAI) Long-Term Replacement Steam Dryer Inspection Plan ML20045E8942020-02-14014 February 2020 Response to a Request for Additional Information for Proposed 10 CFR 50.55a(z)(2) Alternatives to Utilize ASME Code Case N-786-3, Alternative Requirements for Sleeve Reinforcement of Class 2 and 3 Moderate-Energy Carbon Steel Piping, and AS L-MT-19-030, Supplement to a Response for a Request for Additional Information: Application to Adopt 10 CFR 50.69, Risk-Informed Categorization and Treatment of Structures, Systems and Components for Nuclear Power Reactors.2019-05-15015 May 2019 Supplement to a Response for a Request for Additional Information: Application to Adopt 10 CFR 50.69, Risk-Informed Categorization and Treatment of Structures, Systems and Components for Nuclear Power Reactors. L-MT-19-024, Response to a Request for Additional Information for Removal of a Note Associated with Technical Specification 3.5.12019-04-18018 April 2019 Response to a Request for Additional Information for Removal of a Note Associated with Technical Specification 3.5.1 L-MT-19-018, Response to Request for Additional Information: Application to Adopt 10 CFR 50.69, Risk- Informed Categorization & Treatment of Structures, Systems and Components for Nuclear Power Reactors2019-03-13013 March 2019 Response to Request for Additional Information: Application to Adopt 10 CFR 50.69, Risk- Informed Categorization & Treatment of Structures, Systems and Components for Nuclear Power Reactors L-MT-18-058, Response to Request for Additional Information: Application for Technical Specification Change Regarding Risk-Informed Justification for the Relocation of Specific Surveillance Frequency Requirements to ...2018-10-23023 October 2018 Response to Request for Additional Information: Application for Technical Specification Change Regarding Risk-Informed Justification for the Relocation of Specific Surveillance Frequency Requirements to ... L-MT-18-041, Response to Request for Additional Information Re Request for Permanent Exemption from 10CFR50, App R, III.G.2.a Requirements for Exposed Structural Steel2018-07-20020 July 2018 Response to Request for Additional Information Re Request for Permanent Exemption from 10CFR50, App R, III.G.2.a Requirements for Exposed Structural Steel L-MT-18-032, Response to Request for Additional Information Regarding Application to Revise Technical Specifications to Adopt TSTF-542, Reactor Pressure Vessel Water Inventory Control, and Supplement (EPID: L-2017-LLA-03602018-06-0101 June 2018 Response to Request for Additional Information Regarding Application to Revise Technical Specifications to Adopt TSTF-542, Reactor Pressure Vessel Water Inventory Control, and Supplement (EPID: L-2017-LLA-0360 ML18131A2232018-05-11011 May 2018 Prairie and Monticello - Response to Request for Additional Information Regarding Proposed Alternative to Utilize Code Case N-513-4, Evaluation Criteria for Temporary Acceptance of Flaws in Moderate Energy Class 2 or 3 Piping Section XI, Di L-MT-18-013, Response to Request for Additional Information Regarding Exemption Request for Nonconforming Dye Penetrant Examinations of Dry Shielded Canisters (Dscs) 11 Through 152018-04-0505 April 2018 Response to Request for Additional Information Regarding Exemption Request for Nonconforming Dye Penetrant Examinations of Dry Shielded Canisters (Dscs) 11 Through 15 L-MT-17-071, Response to Request for Additional Information Regarding Risk-Informed Request for Exemption from 10CFR50, Appendix R, III.G.2 Requirements for Multiple Spurious Operations of Drywell Spray Motor-Operated Valves2017-11-20020 November 2017 Response to Request for Additional Information Regarding Risk-Informed Request for Exemption from 10CFR50, Appendix R, III.G.2 Requirements for Multiple Spurious Operations of Drywell Spray Motor-Operated Valves L-MT-17-066, Supplemental Information for the Notification of Full Compliance of Required Action for NRC Order EA-12-049 Mitigation Strategies for Beyond-Design-Basis External Events2017-09-28028 September 2017 Supplemental Information for the Notification of Full Compliance of Required Action for NRC Order EA-12-049 Mitigation Strategies for Beyond-Design-Basis External Events L-MT-17-065, License Amendment Request to Revise the Emergency Action Level Scheme - Supplement and Response to Requests for Additional Information2017-09-25025 September 2017 License Amendment Request to Revise the Emergency Action Level Scheme - Supplement and Response to Requests for Additional Information L-MT-17-063, Response to Request for Additional Information Regarding License Amendment Request to Revise Emergency Plan Staff Augmentation Response Times2017-09-20020 September 2017 Response to Request for Additional Information Regarding License Amendment Request to Revise Emergency Plan Staff Augmentation Response Times L-MT-17-025, High Frequency Supplement to Seismic Hazard Screening Report, Response to NRC Request for Information Pursuant to 10CFR50.54(f) Regarding Recommendation 2.1 of the Near-Term Task Force Review of Insights from the Fukushima Dai-Ichi Accid2017-04-11011 April 2017 High Frequency Supplement to Seismic Hazard Screening Report, Response to NRC Request for Information Pursuant to 10CFR50.54(f) Regarding Recommendation 2.1 of the Near-Term Task Force Review of Insights from the Fukushima Dai-Ichi Accident L-MT-17-022, Response to Second Round PRA Related RAIs for the License Amendment Request for a Permanent Extension of the 10 CFR 50 Appendix J Containment Type a Test Interval2017-03-29029 March 2017 Response to Second Round PRA Related RAIs for the License Amendment Request for a Permanent Extension of the 10 CFR 50 Appendix J Containment Type a Test Interval L-MT-17-007, Part 3 Response to Probabilistic Risk Assessment (PRA) Related Requests for Additional Information: License Amendment Request for a Permanent Extension of the 10 CFR 50 Appendix J Containment Type a Test Interval2017-02-0707 February 2017 Part 3 Response to Probabilistic Risk Assessment (PRA) Related Requests for Additional Information: License Amendment Request for a Permanent Extension of the 10 CFR 50 Appendix J Containment Type a Test Interval L-MT-17-002, Part 2 Response to Probabilistic Risk Assessment (PRA) Related Requests for Additional Information: License Amendment Request for a Permanent Extension of the 10 CFR 50Appendix J Containment Type a Test Interval2017-01-31031 January 2017 Part 2 Response to Probabilistic Risk Assessment (PRA) Related Requests for Additional Information: License Amendment Request for a Permanent Extension of the 10 CFR 50Appendix J Containment Type a Test Interval L-MT-16-062, Part 1 Response to Probabilistic Risk Assessment Related Requests for Additional Information: License Amendment Request for a Permanent Extension of the 10 CFR 50 Appendix J Containment Type a Test Interval2016-12-16016 December 2016 Part 1 Response to Probabilistic Risk Assessment Related Requests for Additional Information: License Amendment Request for a Permanent Extension of the 10 CFR 50 Appendix J Containment Type a Test Interval L-MT-16-058, Supplement to License Amendment Request to Revise Battery Charger Surveillance Requirement 3.8.4.22016-11-22022 November 2016 Supplement to License Amendment Request to Revise Battery Charger Surveillance Requirement 3.8.4.2 ML16288A0972016-10-14014 October 2016 and Monticello - Response to Generic Letter 2016-01, Monitoring of Neutron-Absorbing Materials in Spent Fuel Pools L-MT-16-044, Response to Request for Additional Information: License Amendment Request for a Permanent Extension of the 10 CFR 50 Appendix J Containment Type a Test Interval2016-10-10010 October 2016 Response to Request for Additional Information: License Amendment Request for a Permanent Extension of the 10 CFR 50 Appendix J Containment Type a Test Interval L-MT-16-045, Response to Request for Additional Information: License Amendment Request to Revise Battery Charger Surveillance Requirement 3.8.4.22016-10-0303 October 2016 Response to Request for Additional Information: License Amendment Request to Revise Battery Charger Surveillance Requirement 3.8.4.2 L-MT-16-041, License Amendment Request for Areva Extended Flow Window Supplement to Address Power Distribution Uncertainties2016-09-14014 September 2016 License Amendment Request for Areva Extended Flow Window Supplement to Address Power Distribution Uncertainties L-MT-16-038, Seventh Six-Month Status Report in Response to March 12, 2012 Commission Order Modifying Licenses with Regard to Requirements for Mitigation Strategies for Beyond-Design-Basis External Events (Order No. EA-12-049)2016-08-19019 August 2016 Seventh Six-Month Status Report in Response to March 12, 2012 Commission Order Modifying Licenses with Regard to Requirements for Mitigation Strategies for Beyond-Design-Basis External Events (Order No. EA-12-049) L-MT-16-039, Response to Nuclear Security and Incident Response Requests for Additional Information Regarding Changes to the Monticello Nuclear Generation Plant Physical Security Plan (Revision 16) Pursuant to 10 CFR 50.54(p)(2)2016-08-15015 August 2016 Response to Nuclear Security and Incident Response Requests for Additional Information Regarding Changes to the Monticello Nuclear Generation Plant Physical Security Plan (Revision 16) Pursuant to 10 CFR 50.54(p)(2) ML16221A2742016-07-31031 July 2016 ANP-3435NP, Revision 2, Areva Responses to RAI-8 and RAI-32 from Srxb and Snpb on MNGP EFW Lar. L-MT-16-026, Response to Request for Additional Information for Approval of Alternative to Apply the BWRVIP Guidelines in Lieu of Specific ASME Section XI Code Requirements for Reactor Pressure Vessel Internals and Components Inspection2016-06-0202 June 2016 Response to Request for Additional Information for Approval of Alternative to Apply the BWRVIP Guidelines in Lieu of Specific ASME Section XI Code Requirements for Reactor Pressure Vessel Internals and Components Inspection L-MT-16-028, Transmittal of Extended Power Uprate, Extended Steam Dryer - Response to Requests for Additional Information2016-05-18018 May 2016 Transmittal of Extended Power Uprate, Extended Steam Dryer - Response to Requests for Additional Information 2024-06-04
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fl Xcel Energy*
RESPONSIBLE BY NATURE~
2807 West County Road 75 Monticello, MN 55362 May 15, 2019 L-MT-19-030 10 CFR 50.90 10 CFR 50.69 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 Supplement to a Response for a Request for Additional Information: Application to Adopt 10 CFR 50.69, Risk-Informed Categorization and Treatment of Structures, Systems and Components for Nuclear Power Reactors (EPID L-2018-LLA-0076)
References:
- 1) Letter from NSPM to the NRC, Application to Adopt 10 CFR 50.69, Risk-Informed Categorization and Treatment of Structures, Systems and Components for Nuclear Power Reactors, (L-MT-18-010) dated March 28, 2018 (ADAMS Accession No. ML18087A323)
- 2) Email from NRC to NSPM, Request for Additional Information RE:
Monticello License Amendment Request to Adopt 10 CFR 50.69, dated January 31, 2019 (ADAMS Accession No. ML19031A913)
- 3) Letter from NSPM to the NRC, Response to Request for Additional Information: Application to Adopt 10 CFR 50.69, Risk-Informed Categorization and Treatment of Structures, Systems and Components for Nuclear Power Reactors (EPID L-2018-LLA-0076), (L-MT-19-018) dated March 13, 2019 (ADAMS Accession No. ML19072A298)
- 4) NRC Notice of Public Meeting with NSPM, Public Teleconference with Xcel Energy to Discuss the Monticello Nuclear Generating Plant License Amendment Request to Adopt Section 50.69 of Title 10 of the Code of Federal Regulations (EPID L-2018-LLA-0076), dated April 26, 2019 (ADAMS Accession No. ML19116A125)
On March 28, 2018, the Northern States Power Company, a Minnesota corporation, doing business as Xcel Energy (hereafter NSPM), requested a license amendment (Reference 1) to adopt 10 CFR 50.69, Risk-Informed Categorization and Treatment of Structures, Systems and Components [SSCs] for Nuclear Power Reactors, for the Monticello Nuclear Generating Plant (MNGP). The categorization process being implemented through this change is
Document Control Desk Page 2 consistent with Nuclear Energy Institute (NEI) Report NEI 00-04, "10 CFR 50.69 SSC Categorization Guideline", as endorsed by Regulatory Guide 1.201, Revision 1, "Guidelines for Categorizing Structures, Systems, and Components in Nuclear Power Plants According to their Safety Significance". The NRG identified the need for additional information and issued a Request for Additional Information (RAI) on January 31, 2019 (Reference 2). On March 13, 2019, NSPM provided a response to the NRG RAI (Reference 3).
On May 8, 2019, a public teleconference (Reference 4) was held with the NRG to discuss NSPM's response to the RAI. NSPM is providing this supplement to address an NRG request for clarification with respect to RAI 02 discussed during the call. This supplement supersedes and completely replaces the response provided for RAI 02 within the March 13, 2019, letter.
In addition, the enclosure includes a revision to the proposed license condition submitted in response to RAI 07 in the March 13, 2019, letter. This revision supersedes and completely replaces the prior proposed license condition.
The information provided in this letter does not alter the evaluations performed in accordance with 10 CFR 50.92 in Reference 1.
Should you have questions regarding this letter, please contact Mr. Richard Loeffler at (612) 342-8981.
Summary of Commitments This letter makes no new commitments and no revisions to existing commitments.
I declare under penalty of perjury, that the foregoing is true and correct.
Executed on May.!_!, 2019.
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hristopher 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 State of Minnesota
ENCLOSURE MONTICELLO NUCLEAR GENERATING PLANT SUPPLEMENT TO A RESPONSE FOR A REQUEST FOR ADDITIONAL INFORMATION APPLICATION TO ADOPT 10 CFR 50.69 RISK-INFORMED CATEGORIZATION AND TREATMENT OF STRUCTURES, SYSTEMS AND COMPONENTS FOR NUCLEAR POWER REACTORS (8 pages follow)
L-MT-19-030 NSPM Enclosure Supplement to a Response for a Request for Additional Information Application to Adopt 10 CFR 50.69 Risk-Informed Categorization and Treatment of Structures, Systems and Components for Nuclear Power Reactors
1.0 BACKGROUND
On March 28, 2018, the Northern States Power Company, a Minnesota corporation, doing business as Xcel Energy (hereafter NSPM), requested a license amendment (Reference 1) to adopt 10 CFR 50.69, Risk-Informed Categorization and Treatment of Structures, Systems and Components [SSCs] for Nuclear Power Reactors, for the Monticello Nuclear Generating Plant (MNGP). Section 3.1.1 of the license amendment request (LAR) states that NSPM will implement the risk categorization process in accordance with Nuclear Energy Institute (NEI) 00-04, 10 CFR 50.69 SSC Categorization Guideline, (Reference 2) as endorsed by Regulatory Guide (RG) 1.201, Revision 1, Guidelines for Categorizing Structures, Systems, and Components in Nuclear Power Plants According to their Safety Significance, (Reference 3).
The NRC identified the need for additional information and issued a Request for Additional Information (RAI) on January 31, 2019 (Reference 4). On March 13, 2019, NSPM provided a response to the NRC RAI (Reference 5).
On May 8, 2019, a public teleconference (Reference 6) was held with the NRC to discuss NSPMs response to the RAI. NSPM is providing this supplement to address an NRC request for clarification with respect to RAI 02 discussed during the call. This supplement supersedes and completely replaces the response provided for RAI 02 within the March 13, 2019 letter.
Also, this enclosure includes a revision to the proposed license condition submitted in response to RAI 07 in the March 13, 2019, letter. This revision supersedes and completely replaces the prior proposed license condition.
2.0 REVISED RESPONSES TO THE REQUEST FOR ADDITIONAL INFORMATION RAI 02 - Identified Key Assumptions and Sources of Uncertainties Paragraphs 50.69(c)(1)(i) and (ii) of 10 CFR require that a licensees PRA be of sufficient quality and level of detail to support the SSC categorization process, and that all aspects of the integrated, systematic process used to characterize SSC importance must reasonably reflect the current plant configuration and operating practices, and applicable plant and industry operational experience. The guidance in NEI 00-04 specifies sensitivity studies to be conducted for each PRA model to address uncertainty. The sensitivity studies are performed to ensure that assumptions and sources of uncertainty (e.g., human error, common cause failure, and maintenance probabilities) do not mask the SSC(s) importance. Regulatory Guide
[RG] 1.174, Revision 3, cites NUREG-1855, Revision 1, as related guidance. In Section B Page 1 of 8
L-MT-19-030 NSPM Enclosure of RG 1.174, Revision 3, the guidance acknowledges specific revisions of NUREG-1855 to include changes associated with expanding the discussion of uncertainties.
In Section 4.1 of the LAR, Monticello identifies RG 1.174, Revision 3, as an applicable regulatory requirement/criteria. Contrary to Section 4.1 of the LAR, Section 3.2.7 of the LAR states that guidance in NUREG-1855, Revision 0, Guidance on the Treatment of Uncertainties Associated with PRAs in Risk-Informed Decisionmaking, and Electric Power Research Institute (EPRI) TR-1016737, Treatment of Parameter and Modeling Uncertainty for Probabilistic Risk Assessments, was used to identify, characterize, and screen model uncertainties. Attachment 6 of the LAR identifies five assumptions and sources of uncertainty applicable to either the IEPRA (includes internal flood) or FPRA models.
NUREG-1855 has been updated to Revision 1 as of March 2017 (ADAMS Accession No. ML17062A466). The NRC staff notes that NUREG-1855, Revision 1, provides guidance in stages A through E for how to treat uncertainties associated with PRA models in risk-informed decisionmaking. Revision 1 of NUREG-1855 cites EPRI TR-1026511, Practical Guidance on the Use of Probabilistic Risk Assessment in Risk-Informed Applications with a Focus on the Treatment of Uncertainty. Considering these observations provide the following:
- a. A detailed summary of the process used to identify the key assumptions and sources of uncertainty presented in Attachment 6 of the LAR. The discussion should include:
- i. How the process is consistent with NUREG-1855, Revision 1, or other NRC-accepted methods (e.g., NUREG-1855, Revision 0). If deviating from the current guidance provided in NUREG-1855, Revision 1, provide a basis to justify the appropriateness of any deviations for use in the 10 CFR 50.69 categorization process (e.g., exclusion/consideration of EPRI TR-1026511).
ii. A brief description of how the key assumptions and sources of uncertainties provided in Attachment 6 of the LAR were identified from the initial comprehensive list of PRA model(s) (i.e., base model) uncertainties and assumptions, including those associated with plant-specific features, modeling choices, and generic industry concerns. This can include an identification of the sources of plant-specific and applicable generic modeling uncertainties identified in the uncertainty analyses for the base IEPRA (includes internal flood) and the base FPRA and include a disposition for each of the assumptions and/or uncertainties addressing their impact for the 10 CR 50.69 risk application. For any source of uncertainty or assumption judged not to be key to the application, provide discussion for why it is not pertinent to the application and therefore does not need to be addressed (i.e., sensitivity studies performed).
- b. If the process used to identify, characterize, and assess the key assumption(s) and the treatment for the sources of uncertainty provided in Attachment 6 of the LAR cannot be justified for use in the 50.69 categorization process, provide the results of an updated assessment of the key assumptions, sources of uncertainty, and treatment of the sources Page 2 of 8
L-MT-19-030 NSPM Enclosure of uncertainty performed in accordance with NUREG-1855, Revision 1, and NEI 00-04, Revision 0. For the treatment of the sources of uncertainty (e.g., sensitivity studies to be performed) include a detailed description of the sensitivity study and how the sensitivity study is bounding to address the specific key assumption and/or source of uncertainty.
NSPM Response At the time of the submittal of the LAR, the sources of uncertainty evaluation for the internal events PRA had considered both plant-specific sources of uncertainty and the generic uncertainties identified in Electric Power Research Institute (EPRI) TR-1016737 (Reference 7).
The fire PRA considered the plant-specific uncertainty sources, but did not specifically address the EPRI generic sources as provided in EPRI TR-1026511 (Reference 8).
Since the time of the LAR submittal, the internal events and fire PRAs have been updated to include the EPRI-identified generic sources of uncertainty as documented in EPRI TR-1016737 and TR-1026511. Both modeling uncertainty and completeness uncertainty sources were examined. Each PRA includes an evaluation of the sources of uncertainty for the base case models using the approach that is consistent with the ASME/ANS RA-Sa-2009 (Reference 9) requirements for identification and characterization of uncertainties and assumptions. This evaluation meets the intent of steps C-1 and E-1 of NUREG-1855, Revision 1 (Reference 10).
At the time of the original LAR submittal, the identification of those base PRA uncertainties that were important for 10 CFR 50.69 categorization was performed based on expert judgement.
To enhance the traceability of this evaluation, an additional review was performed. The approach used for this review is similar to that used at the Prairie Island Nuclear Generating Plant for its 10 CFR 50.69 LAR (Reference 11). The updated evaluation process includes a review of the Internal Events and Fire PRA Uncertainty Notebooks to determine which uncertainties could impact the 10 CFR 50.69 categorization process results. This evaluation meets the intent of the screening portion of steps C-2 and E-2 of NUREG-1855, Revision 1.
The ultimate goal in assessing model uncertainty is to determine whether (and the degree to which) the risk metric results challenge or exceed the quantitative acceptance guidelines for the application, due to sources of model uncertainty and related assumptions. For 10 CFR 50.69 categorization, the acceptance guidelines are actually threshold values for Fussell Vesely (F-V) and Risk Achievement Worth (RAW) for each system, structure, and component (SSC) being categorized, above which the SSC is categorized as high safety significant (HSS), and below which the SSC is categorized as low safety significant (LSS). As described in Step E-2 of the NUREG-1855, Revision 1, each relevant uncertainty/assumption requires some sort of sensitivity analysis, and each sensitivity performed to evaluate an uncertainty/assumption involves some change to the PRA results. Since any change to the PRA results has the potential to change the F-V and RAW importance measures for all components, every relevant uncertainty/assumption has the potential to challenge the acceptance guidelines. That is, since RAW and F-V are relative importance measures, any change to any part of the model will generate a new set of cutsets and potentially impact the Page 3 of 8
L-MT-19-030 NSPM Enclosure RAW and F-V for every SSC. Thus, the only way to evaluate the impact of a sensitivity is to quantify the sensitivity case and compare the F-V and RAW values for all SSCs against the base case F-V and RAW values to determine if any exceed the HSS threshold in the sensitivity case that did not previously do so.
As a result of the updated evaluation of the uncertainties, Attachment 6 has been revised, replacing in total what was provided in the LAR.
Attachment 6: Disposition of Key Assumptions/Sources of Uncertainty Assumption/Uncertainty Discussion Disposition Very small loss of coolant The impact of this assumption A sensitivity study will be accidents (LOCAs) are will need to be assessed for performed that addresses defined as those for which specific risk applications, very small LOCAs in flow rates are less than can including 10 CFR 50.69. accordance with NEI 00-04, be made up by normal Particularly, any applications Table 5-2, to determine if makeup systems such as where small LOCA initiators there are any changes in the Control Rod Drive Hydraulic could be significant HSS/LSS determination.
System (~100 gpm). The contributors may be affected mitigating success criteria for by this assumption.
this break size is identical to that for a transient initiator in The drywell coolers are not which decay heat makeup credited in the PRA model; rates are required, the only therefore, there is no risk difference in plant response impact.
being that a high drywell pressure may occur. This would result in trip of drywell coolers and the Residual Heat Removal Service Water System due to load shed, but would affect no other mitigating systems. Therefore, this event is considered to be encompassed by the Reactor Trip or Turbine Trip initiating event and no new initiating event is created for the purpose of evaluating very small LOCAs.
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L-MT-19-030 NSPM Enclosure Assumption/Uncertainty Discussion Disposition A minimum value for a single HEP values and their Sensitivity studies in pre or post initiator Human dependence have the ability to accordance with NEI 00-04, Error Probability (HEP) was significantly impact model Table 5-2, will be performed assumed to be 1.0 E-5. This results; therefore, this is to evaluate the potential value is reserved for operator considered an uncertainty. impact of variations in HEP actions which only take a few values.
minutes but have over ten hours to perform. An independent or dependent HEP combination minimum value was assumed to be 1E-6.
While the walkdown sheets Per EPRI report (Pipe This item does not represent were used whenever possible Rupture Frequencies for a key source of uncertainty to obtain source systems, Internal Flooding Probabilistic for 50.69 calculations.
pipe sizes, and pipe lengths Risk Assessments, Revision 3, for the various walkdown EPRI-3002000079 zones, there are zones in the (Reference 12), April 2013) plant for which no such data internal event flood exists. Such cases required frequencies are directly the estimation of the proportional to the pipe necessary equipment based lengths. The internal flood on P&ID information and the events where pipe lengths analysts experience at other could not be validated with a similar plants. The walkdown walkdown were reviewed and notebook documents the data found to have reasonable that was estimated for this pipe length estimates based analysis. on room size. Furthermore, their risk contribution was insignificant.
Isometric drawings were used to estimate pipe lengths for high-risk floods.
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L-MT-19-030 NSPM Enclosure RAI 07 - Proposed License Condition NSPM Response NSPM is approved to implement 10 CFR 50.69 using the approaches for categorization of Risk Informed Safety Class (RISC)-1, RISC-2, RISC-3, and RISC-4 structures, systems, and components (SSCs) using: Probabilistic Risk Assessment (PRA) models to evaluate risk associated with internal events, including internal flooding and internal fire, with the shutdown safety assessment process to assess shutdown risk; the Arkansas Nuclear One, Unit 2 (ANO-2) passive categorization method to assess passive component risk for Class 2 and Class 3 SSCs and their associated supports; and the results of non-PRA evaluations that are based on the IPEEE Screening Assessment for External Hazards, i.e., seismic margin analysis (SMA) to evaluate seismic risk, and a screening of other external hazards (e.g., external flooding and high winds) updated using the external hazard screening significance criteria identified in ASME/ANS PRA Standard RA-Sa-2009, as endorsed in RG 1.200, Revision 2; as specified in MNGP License Amendment No. [XXX] dated [DATE].
Prior NRC approval, under 10 CFR 50.90, is required for a change to the categorization approach specified above (e.g., change from a seismic margins approach to a seismic probabilistic risk assessment approach).
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L-MT-19-030 NSPM Enclosure
3.0 REFERENCES
- 1. Letter from NSPM to the NRC, Application to Adopt 10 CFR 50.69, Risk-Informed Categorization and Treatment of Structures, Systems and Components for Nuclear Power Reactors, (L-MT-18-010) dated March 28, 2018 (ADAMS Accession No. ML18087A323 )
- 2. Nuclear Energy Institute (NEI) 00-04, 10 CFR 50.69 SSC Categorization Guideline, Revision 0, dated July 2005 (ADAMS Accession No. ML052910035)
- 3. NRC Regulatory Guide (RG) 1.201, Revision 1, Guidelines for Categorizing Structures, Systems, and Components in Nuclear Power Plants According to their Safety Significance, dated May 2006 (ADAMS Accession No. ML061090627)
- 4. Email from NRC to NSPM, Request for Additional Information RE: Monticello License Amendment Request to Adopt 10 CFR 50.69, dated January 31, 2019 (ADAMS Accession No. ML19031A913)
- 5. Letter from NSPM to the NRC, Response to Request for Additional Information:
Application to Adopt 10 CFR 50.69, Risk-Informed Categorization and Treatment of Structures, Systems and Components for Nuclear Power Reactors (EPID L-2018-LLA-0076), (L-MT-19-018) dated March 13, 2019 (ADAMS Accession No. ML19072A298)
- 6. NRC Notice of Public Meeting with NSPM, Public Teleconference with Xcel Energy to Discuss the Monticello Nuclear Generating Plant License Amendment Request to Adopt Section 50.69 of Title 10 of the Code of Federal Regulations (EPID L-2018-LLA-0076),
dated April 26, 2019 (ADAMS Accession No. ML19116A125)
- 7. Electric Power Research Institute (EPRI) Technical Report TR-1016737, Treatment of Parameter and Model Uncertainty for Probabilistic Risk Assessments, dated December 2008
- 8. EPRI Technical Report TR-1026511, Practical Guidance on the Use of PRA in Risk-Informed Applications with a Focus on the Treatment of Uncertainty, dated December 2012
- 9. American Society of Mechanical Engineering (ASME) Standard ASME/ANS RA-Sa-2009, Addenda to ASME/ANS RA-S-2008 Standard for Level 1/Large Early Release Frequency Probabilistic Risk Assessment for Nuclear Power Plant Applications, dated February 2, 2009 Page 7 of 8
L-MT-19-030 NSPM Enclosure
- 10. NRC NUREG-1855, Guidance on the Treatment of Uncertainties Associated with PRAs in Risk-Informed Decision Making Final Report, Revision 1, dated March 2017 (ADAMS Accession No. ML17062A466)
- 11. Letter from NSPM to the NRC, Application to Adopt 10 CFR 50.69, Risk-Informed Categorization and Treatment of Structures, Systems and Components for Nuclear Power Reactors, (L-PI-18-012) dated July 20, 2018 (ADAMS Accession No. ML18204A393)
- 12. EPRI Technical Report 3002000079, Pipe Rupture Frequencies for Internal Flooding Probabilistic Risk Assessments, Revision 3, dated April 2013 Page 8 of 8