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MONTHYEAR0CAN091302, Assistance in Obtaining Information on Dams2013-09-30030 September 2013 Assistance in Obtaining Information on Dams Project stage: Request L-PI-14-024, Extension Request Regarding the Flooding Hazard Reevaluation Report Required by NRC Request for Information Pursuant to Title 10 of the Code of Federal Regulations 50.54(0 Regarding Recommendation 2.1, Flooding, of the Near-Term Task For2014-03-10010 March 2014 Extension Request Regarding the Flooding Hazard Reevaluation Report Required by NRC Request for Information Pursuant to Title 10 of the Code of Federal Regulations 50.54(0 Regarding Recommendation 2.1, Flooding, of the Near-Term Task Force Project stage: Request L-14-113, Extension Request Regarding the Flood Hazard Reevaluation Report Required by NRC Request for Information Pursuant to 10 CFR 50.54(f) Regarding the Flooding Aspects of Recommendation 2.1 of the Near-Term Task Force (NTTF) Review of Insights.2014-03-11011 March 2014 Extension Request Regarding the Flood Hazard Reevaluation Report Required by NRC Request for Information Pursuant to 10 CFR 50.54(f) Regarding the Flooding Aspects of Recommendation 2.1 of the Near-Term Task Force (NTTF) Review of Insights. Project stage: Request 0CAN031401, Required Response 2 to March 12, 2012, Request for Information (Rfi), Enclosure 2, Recommendation 2.1, Flood Hazard Evaluation Report Arkansas Nuclear One - Units 1 and 22014-03-12012 March 2014 Required Response 2 to March 12, 2012, Request for Information (Rfi), Enclosure 2, Recommendation 2.1, Flood Hazard Evaluation Report Arkansas Nuclear One - Units 1 and 2 Project stage: Request 0CAN041402, Required Response 2 to March 12, 2012, Request for Information (Rfi), Enclosure 2, Recommendation 2.1, Flood Hazard Evaluation Report2014-04-14014 April 2014 Required Response 2 to March 12, 2012, Request for Information (Rfi), Enclosure 2, Recommendation 2.1, Flood Hazard Evaluation Report Project stage: Request ML14168A0182014-05-16016 May 2014 Xcel Energy Responses to NRC Questions Regarding Flood Hazard Reevaluation Report Extension Requests Project stage: Other ML14153A0242014-05-16016 May 2014 Xcel Energy'S Presentation Slides Supporting a May 16, 2014, Public Meeting to Discuss the Flood Hazard Reevaluation Schedule for the Monticello and Prairie Island Nuclear Generating Plants Project stage: Meeting ML14153A0252014-06-20020 June 2014 Meeting Summary - Public Meeting on May 16, 2014, to Discuss Xcel Energy'S Flood Hazard Reevaluation Extension Request for the Monticello and Prairie Island Nuclear Generating Plants Project stage: Meeting ML14162A3712014-06-26026 June 2014 5/22/2014 Summary of Meeting with Entergy Operations, Inc. to Discuss Flooding Hazard Reevaluation Extension Request for Arkansas Nuclear One, Units 1 and 2; Near-Term Task Force Recommendation 2.1 Project stage: Meeting ML14171A5292014-07-17017 July 2014 Relaxation of Response Due Date Regarding Flooding Hazard Reevaluations for Recommendation 2.1 of the Near-Term Task Force Review of the Insights from the Fukushima Dai-Ichi Accident Project stage: Approval ML14171A1662014-07-29029 July 2014 Relaxation of Response Due Date Regarding Flooding Hazard Reevaluations for Recommendation 2.1 of the Near-Term Task Force Review of the Insights from the Fukushima Dai-Ichi Accident Project stage: Approval ML15155A6312015-06-0808 June 2015 Nuclear Regulatory Commission Plan for the Audit of Duke Energy Carolinas, Llc'S Flood Hazard Reevaluation Report Submittal Relating to the Near-Term Task Force Recommendation 2.1 Flooding for Catawba Nuclear Station, Units 1 and 2 Project stage: Other 2014-05-16
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Category:Letter type:L
MONTHYEARL-PI-23-034, Response to Request for Additional Information Regarding License Amendment Request to Revise Technical Specification 3.7.8, Cooling Water (Cl) System,2024-01-0202 January 2024 Response to Request for Additional Information Regarding License Amendment Request to Revise Technical Specification 3.7.8, Cooling Water (Cl) System, L-PI-23-035, Response to Request for Additional Information Regarding License Amendment Request to Revise Technical Specifications Section 5.6.6, Reactor Coolant System (RCS) Pressure and Temperature Limits Report2023-12-20020 December 2023 Response to Request for Additional Information Regarding License Amendment Request to Revise Technical Specifications Section 5.6.6, Reactor Coolant System (RCS) Pressure and Temperature Limits Report L-PI-23-033, Supplement to License Amendment Request to Revise Technical Specification 3.8.1, Surveillance Requirement 3.8.1.2, Note 32023-12-0505 December 2023 Supplement to License Amendment Request to Revise Technical Specification 3.8.1, Surveillance Requirement 3.8.1.2, Note 3 L-PI-23-025, License Amendment Request to Revise Technical Specification 3.8.1, Surveillance Requirement 3.8.1.2, Note 32023-09-28028 September 2023 License Amendment Request to Revise Technical Specification 3.8.1, Surveillance Requirement 3.8.1.2, Note 3 L-PI-23-023, Baffle Former Bolts Alternate Aging Management Strategy2023-09-11011 September 2023 Baffle Former Bolts Alternate Aging Management Strategy L-PI-23-018, License Amendment Request to Revise ISFSI Technical Specification 4.4 to Allow Use of a Code Alternative to ASME Code, NB-5130, Examination of Weld Edge Preparation Surfaces, for the TN-40HT2023-07-14014 July 2023 License Amendment Request to Revise ISFSI Technical Specification 4.4 to Allow Use of a Code Alternative to ASME Code, NB-5130, Examination of Weld Edge Preparation Surfaces, for the TN-40HT L-PI-23-006, License Amendment Request to Revise Technical Specification 3.7.8 Required Actions2023-06-22022 June 2023 License Amendment Request to Revise Technical Specification 3.7.8 Required Actions L-PI-23-016, 2022 10 CFR 50.46 LOCA Annual Report2023-06-14014 June 2023 2022 10 CFR 50.46 LOCA Annual Report L-PI-23-010, Annual Report of Individual Monitoring2023-04-27027 April 2023 Annual Report of Individual Monitoring L-PI-23-007, Supplement to Application to Revise Technical Specification Section 5.6.6, Reactor Coolant System (RCS) Pressure and Temperature Limits Report (PTLR)2023-03-28028 March 2023 Supplement to Application to Revise Technical Specification Section 5.6.6, Reactor Coolant System (RCS) Pressure and Temperature Limits Report (PTLR) L-PI-23-005, CFR 50.55a Requests Nos. 1-RR-5-15 and 2-RR-5-15, Proposed Use of Subsequent ASME Code Edition and Addenda in Accordance with 10 CFR 50.55a(g)(4)(iv)2023-03-0303 March 2023 CFR 50.55a Requests Nos. 1-RR-5-15 and 2-RR-5-15, Proposed Use of Subsequent ASME Code Edition and Addenda in Accordance with 10 CFR 50.55a(g)(4)(iv) L-PI-23-001, Day Steam Generator Tube Inspection Report2023-01-30030 January 2023 Day Steam Generator Tube Inspection Report L-PI-22-047, Resubmittal of Prairie Island Nuclear Generating Plant (PINGP) 2018 Unit 1 180-Day Steam Generator Tube Inspection Report2022-12-21021 December 2022 Resubmittal of Prairie Island Nuclear Generating Plant (PINGP) 2018 Unit 1 180-Day Steam Generator Tube Inspection Report L-PI-22-020, Application to Revise Technical Specification Section 5.6.6, Reactor Coolant System (RCS) Pressure and Temperature Limits Report (PTLR)2022-12-0202 December 2022 Application to Revise Technical Specification Section 5.6.6, Reactor Coolant System (RCS) Pressure and Temperature Limits Report (PTLR) L-PI-22-040, Supplement to Application to Revise Technical Specifications to Adopt TSTF-577, Revised Frequencies for Steam Generator Tube Inspections2022-10-0606 October 2022 Supplement to Application to Revise Technical Specifications to Adopt TSTF-577, Revised Frequencies for Steam Generator Tube Inspections L-PI-22-037, Updated Approach for Prairie Island Unit 1 and Unit 2 Baffle Former Bolts2022-09-20020 September 2022 Updated Approach for Prairie Island Unit 1 and Unit 2 Baffle Former Bolts L-PI-22-032, CFR 50.46 LOCA Annual Report2022-06-16016 June 2022 CFR 50.46 LOCA Annual Report L-PI-22-033, Response to Request for Additional Information Prairie Island Nuclear Generating Plant, Units 1 and 2, Amendment to Adopt 24-Month Operating Cycles2022-06-10010 June 2022 Response to Request for Additional Information Prairie Island Nuclear Generating Plant, Units 1 and 2, Amendment to Adopt 24-Month Operating Cycles L-PI-22-003, Application to Revise Technical Specifications to Adopt TSTF-577, Revised Frequencies for Steam Generator Tube Inspections2022-06-0707 June 2022 Application to Revise Technical Specifications to Adopt TSTF-577, Revised Frequencies for Steam Generator Tube Inspections L-PI-22-024, Supplement to Application for License Amendment to Implement 24-Month Operating Cycle2022-03-0707 March 2022 Supplement to Application for License Amendment to Implement 24-Month Operating Cycle L-PI-21-047, Response to Request for Additional Information 24-Month Cycle Amendment Prairie Island Nuclear Generating Plant, Units 1 and 22021-12-0707 December 2021 Response to Request for Additional Information 24-Month Cycle Amendment Prairie Island Nuclear Generating Plant, Units 1 and 2 L-PI-21-045, Response to Request for Additional Information Cooling Water System License Amendment Request2021-11-0404 November 2021 Response to Request for Additional Information Cooling Water System License Amendment Request L-PI-21-029, Exigent License Amendment Request to Revise Technical Specification 3.7.8 to Allow a One-Time Extension of the Completion Time of Required Action B.12021-10-0707 October 2021 Exigent License Amendment Request to Revise Technical Specification 3.7.8 to Allow a One-Time Extension of the Completion Time of Required Action B.1 L-PI-21-006, License Amendment Request to Revise Technical Specification 3.3.1, Reactor Trip System (RTS) Instrumentation, to Incorporate Installed Bypass Test Capability for the Power Range RTS Functions2021-10-0202 October 2021 License Amendment Request to Revise Technical Specification 3.3.1, Reactor Trip System (RTS) Instrumentation, to Incorporate Installed Bypass Test Capability for the Power Range RTS Functions L-PI-21-032, Response to Request for Additional Information Amendment Request to Adopt TSTF-471 and 571-T for Prairie Island2021-09-30030 September 2021 Response to Request for Additional Information Amendment Request to Adopt TSTF-471 and 571-T for Prairie Island L-PI-21-016, Application for License Amendment to Implement 24-Month Operating Cycle2021-08-0606 August 2021 Application for License Amendment to Implement 24-Month Operating Cycle L-PI-21-027, 2020 10 CFR 50.46 LOCA Annual Report2021-06-28028 June 2021 2020 10 CFR 50.46 LOCA Annual Report L-PI-21-023, Independent Spent Fuel Storage Installation - 2020 Annual Radiological Environmental Monitoring Program Report2021-05-14014 May 2021 Independent Spent Fuel Storage Installation - 2020 Annual Radiological Environmental Monitoring Program Report L-PI-21-007, Application to Revise Technical Specifications to Adopt TSTF-471, Eliminate Use of the Term Core Alterations in Actions and Notes2021-04-19019 April 2021 Application to Revise Technical Specifications to Adopt TSTF-471, Eliminate Use of the Term Core Alterations in Actions and Notes L-PI-20-050, Request for a One-Time Exemption from 10 CFR 50, Appendix E, Biennial Emergency Preparedness Exercise Requirements, Due to COVID-19 Pandemic2020-10-0707 October 2020 Request for a One-Time Exemption from 10 CFR 50, Appendix E, Biennial Emergency Preparedness Exercise Requirements, Due to COVID-19 Pandemic L-PI-20-051, Supplement to Application to Revise Technical Specifications to Adopt TSTF-547, Clarification of Rod Position Requirements2020-09-28028 September 2020 Supplement to Application to Revise Technical Specifications to Adopt TSTF-547, Clarification of Rod Position Requirements L-PI-20-026, Response to Request for Additional Information: License Amendment Request to Revise Technical Specifications to Adopt Risk Informed Completion Times TSTF-505, Revision 2, Provide Risk-Informed Extended Completion Times - RITSTF Initiativ2020-09-0101 September 2020 Response to Request for Additional Information: License Amendment Request to Revise Technical Specifications to Adopt Risk Informed Completion Times TSTF-505, Revision 2, Provide Risk-Informed Extended Completion Times - RITSTF Initiative 4 L-PI-20-035, = Request for Revision to Reactor Vessel Material Surveillance Capsule Removal Schedule2020-07-28028 July 2020 = Request for Revision to Reactor Vessel Material Surveillance Capsule Removal Schedule L-PI-20-023, Independent Spent Fuel Storage Installation, Response to Request for Additional Information: License Amendment Request to Expand the Storage Capacity of the Independent Spent Fuel Storage Installation (ISFSI)2020-06-10010 June 2020 Independent Spent Fuel Storage Installation, Response to Request for Additional Information: License Amendment Request to Expand the Storage Capacity of the Independent Spent Fuel Storage Installation (ISFSI) L-PI-20-014, Supplement to License Amendment Request: Expand the Storage Capacity of the Independent Spent Fuel Storage Installation (ISFSI2020-04-29029 April 2020 Supplement to License Amendment Request: Expand the Storage Capacity of the Independent Spent Fuel Storage Installation (ISFSI L-PI-20-004, License Amendment Request to Revise Technical Specifications (TS) to Remove Note I from Limiting Condition for Operating (LCO) 3.4.12 and LCO 3.4.132020-03-30030 March 2020 License Amendment Request to Revise Technical Specifications (TS) to Remove Note I from Limiting Condition for Operating (LCO) 3.4.12 and LCO 3.4.13 L-PI-20-001, License Amendment Request to Address Issues Identified in Westinghouse Nuclear Safety Advisory Letter NSAL-09-5, Revision 1, and NSAL-15-12020-01-29029 January 2020 License Amendment Request to Address Issues Identified in Westinghouse Nuclear Safety Advisory Letter NSAL-09-5, Revision 1, and NSAL-15-1 L-PI-19-041, Application to Revise Technical Specifications to Adopt TSTF-547, Clarification of Rod Position Requirements2019-12-23023 December 2019 Application to Revise Technical Specifications to Adopt TSTF-547, Clarification of Rod Position Requirements L-PI-19-031, License Amendment Request: Revise Technical Specifications to Adopt Risk Informed Completion Times TSTF-505, Revision 2, Provide Risk-Informed Extended Completion Times - RITSTF Initiative 4b2019-12-16016 December 2019 License Amendment Request: Revise Technical Specifications to Adopt Risk Informed Completion Times TSTF-505, Revision 2, Provide Risk-Informed Extended Completion Times - RITSTF Initiative 4b L-PI-19-040, License Amendment Request: Revise Technical Specification 5.5.14 to Permanently Extend Containment Leakage Rate Test Frequency2019-10-0707 October 2019 License Amendment Request: Revise Technical Specification 5.5.14 to Permanently Extend Containment Leakage Rate Test Frequency L-PI-19-038, Submittal of Revised Pressure and Temperature Limits Report2019-09-19019 September 2019 Submittal of Revised Pressure and Temperature Limits Report L-PI-19-037, Response to Request for Additional Information: 10 CFR 50.55a Requests Nos. 1-RR-5-10 and 2-RR-5-10, Proposed Alternative to Reactor Vessel Inservice Inspection (ISI) Intervals2019-09-16016 September 2019 Response to Request for Additional Information: 10 CFR 50.55a Requests Nos. 1-RR-5-10 and 2-RR-5-10, Proposed Alternative to Reactor Vessel Inservice Inspection (ISI) Intervals L-PI-19-025, Request to Approve Site-Specific Probabilistic Risk Assessment (PRA) Model for Flowserve N-Seal Abeyance Seal and Dynamic Testing for the Prairie Island Nuclear Generating Plant (PINGP)2019-08-27027 August 2019 Request to Approve Site-Specific Probabilistic Risk Assessment (PRA) Model for Flowserve N-Seal Abeyance Seal and Dynamic Testing for the Prairie Island Nuclear Generating Plant (PINGP) L-PI-19-029, Supplement to Response to Request for Additional Information: Application to Adopt 10 CFR 50.69, Risk-Informed Categorization and Treatment of Structures, Systems and Components For...2019-08-0505 August 2019 Supplement to Response to Request for Additional Information: Application to Adopt 10 CFR 50.69, Risk-Informed Categorization and Treatment of Structures, Systems and Components For... L-PI-19-002, 10 CFR 50.55a Requests Nos. 1-RR-5-10 and 2-RR-5-10, Proposed Alternative to Reactor Vessel Inservice Inspection (ISI) Intervals for Prairie Island, Unit 1 and Unit 22019-06-13013 June 2019 10 CFR 50.55a Requests Nos. 1-RR-5-10 and 2-RR-5-10, Proposed Alternative to Reactor Vessel Inservice Inspection (ISI) Intervals for Prairie Island, Unit 1 and Unit 2 L-PI-19-014, Response to Request for Additional Information: Application to Adopt 10 CFR 50.69, Risk-Informed Categorization and Treatment of Structures, Systems and Components for Nuclear Power Reactors2019-04-29029 April 2019 Response to Request for Additional Information: Application to Adopt 10 CFR 50.69, Risk-Informed Categorization and Treatment of Structures, Systems and Components for Nuclear Power Reactors L-PI-19-003, Request for Revision to Reactor Vessel Material Surveillance Capsule Removal Schedule for Prairie Island Nuclear Generating Plant (PINGP)2019-02-0404 February 2019 Request for Revision to Reactor Vessel Material Surveillance Capsule Removal Schedule for Prairie Island Nuclear Generating Plant (PINGP) L-PI-19-006, Emergency License Amendment Request Regarding One-Time Extension for Technical Specification Completion Time Requirements2019-01-29029 January 2019 Emergency License Amendment Request Regarding One-Time Extension for Technical Specification Completion Time Requirements L-PI-19-005, Online Reference Portal for NRC Review of License Amendment Request to Implement 10 CFR 50.692019-01-15015 January 2019 Online Reference Portal for NRC Review of License Amendment Request to Implement 10 CFR 50.69 L-PI-18-063, Response to Request for Additional Information: Revise License Condition Associated with Implementation of NFPA 8052018-12-0606 December 2018 Response to Request for Additional Information: Revise License Condition Associated with Implementation of NFPA 805 2024-01-02
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Xcel Energy Prairie Island Nuclear Generating Plant 1717 Wakonade Drive East Welch, MN 55089 L-PI-14-024 MAR 10 2014 10 CFR 50.54(f)
U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 Prairie Island Nuclear Generating Plant Units 1 and 2 Docket Nos. 50-282 and 50-306 Renewed Operating License Nos. DPR-42 and DPR-60 Extension Request Regarding the Flooding Hazard Reevaluation Report Required by NRC Request for Information Pursuant to Title 10 of the Code of Federal Regulations 50.54(0 Regarding Recommendation 2.1, Flooding, of the Near-Term Task Force Review of Insights from the Fukushima Dai-ichi Accident
References:
- 1. NRC Letter, "Request for Information Pursuant to Title 10 of the Code of Federal Regulations 50.54(f) Regarding Recommendations 2.1, 2.3, and 9.3, of the Near-Term Task Force Review of Insights from the Fukushima Dai-ichi Accident," dated March 12, 2012 (ADAMS Accession No. ML12053A340).
- 2. NRC Letter, "Prioritization of Response Due Dates for Request for Information Pursuant to Title 10 of the Code of Federal Regulations 50.54(f) Regarding Flooding Hazard Reevaluations for Recommendation 2.1 of the Near-Term Task Force Review of Insights from the Fukushima Dai-ichi Accident," dated May 11, 2012 (ADAMS Accession No. ML12097A509).
- 3. NRC Letter, "Supplemental Information Related to Request for Information Pursuant to Title 10 of the Code of Federal Regulations 50.54(f) Regarding Flooding Hazard Reevaluations for Recommendation 2.1 of the Near-Term Task Force Review of Insights from the Fukushima Dai-ichi Accident," dated March 1, 2013 (ADAMS Accession No. ML13044A561 ).
On March 12, 2012, the Nuclear Regulatory Commission (NRC) Staff issued the Reference 1 letter to all NRC power reactor licensees and holders of construction permits in active or deferred status. Enclosure 2 of Reference 1 contains specific
Document Control Desk Page 2 Requested Actions, Requested Information, and Required Responses associated with Near-Term Task Force (NTTF) Recommendation 2.1 for flooding hazards. Enclosure 2 of Reference 1 directed reevaluation of flooding hazards at sites and indicated that a Hazard Reevaluation Report (HRR) would be due, within one to three years from the date of the Reference 1 letter. On May 11, 2012, the NRC issued the Reference 2 letter that contained the NRC's prioritization plan and due dates for licensees' submittal of HRRs. The Prairie Island Nuclear Generating Plant (PINGP), Units 1 and 2, operated by Northern States Power Company, a Minnesota corporation (NSPM), d/b/a Xcel Energy, were identified as Category 2 sites in Reference 2 and were required to submit the HRR by March 12, 2014.
In Reference 3, the NRC provided supplemental information that states incomplete HRRs that only contain an analysis of some flooding hazard mechanisms would not be of substantive benefit for staff review and would not be acceptable. Therefore, Reference 3 recommends licensees not submit partial reports, but instead submit an extension request.
The purpose of this letter is to request an extension of the March 12, 2014 due date for submittal of the HRR for PINGP, Units 1 and 2. The Enclosure to this letter contains the extension request. The extension request was prepared using the guidance in Reference 3 and includes the reasons for the delay, a proposed schedule for the submittal of a complete HRR, and the basis for acceptability of the revised schedule.
If there are any questions or if additional information is needed, please contact Jennie Wike, Licensing Engineer, at 612-330-5788.
Document Control Desk Page 3 Summary of Commitments This letter makes one new commitment and makes no revisions to existing commitments.
- NSPM will submit the required flood HRR for PINGP, Units 1 and 2, within 10 months of receiving the US Army Corps of Engineers' final information.
I declare under penalty of perjury that the foregoing is true and correct.
Executed on
- Jap Scott Sharp Director Site Operations, Prairie Island Nuclear Generating Plant Northern States Power Company - Minnesota cc: Administrator, Region Ill, USNRC Director of Nuclear Reactor Regulation (NRR), USNRC NRR Project Manager, Prairie Island Nuclear Generating Plant, US NRC Senior Resident Inspector, Prairie Island Nuclear Generating Plant, USNRC
ENCLOSURE PRAIRIE ISLAND NUCLEAR GENERATING PLANT, UNITS 1 AND 2 Extension Request Regarding the Flooding Hazard Reevaluation Report Required by NRC Request for Information Pursuant to Title 10 of the Code of Federal Regulations 50.54(f) Regarding Recommendation 2.1, Flooding, of the Near-Term Task Force Review of Insights from the Fukushima Dai-ichi Accident Request For Extension: Reason for the Delay, Proposed Schedule, and Basis for Acceptability Page 1 of 5
Enclosure NSPM Request For Extension 1.0 Introduction On March 12, 2012, the Nuclear Regulatory Commission (NRC) Staff issued the Reference 1 letter to all NRC power reactor licensees and holders of construction permits in active or deferred status. Enclosure 2 of Reference 1 contains specific Requested Actions, Requested Information, and Required Responses associated with Near-Term Task Force (NTTF) Recommendation 2.1 for flooding hazards. Enclosure 2 of Reference 1 directed reevaluation of flooding hazards at sites and indicated that a Hazard Reevaluation Report (HRR) would be due within one to three years from the date of Reference 1 letter. On May 11, 2012, the NRC issued the Reference 2 letter that contained the NRC's prioritization plan and due dates for licensees' submittal of HRRs.
The Prairie Island Nuclear Generating Plant (PINGP), Units 1 and 2, operated by Northern States Power Company, a Minnesota corporation (NSPM), d/b/a Xcel Energy, were identified as Category 2 sites in Reference 2 and were required to submit the HRR by March 12, 2014.
In Reference 3, the NRC provided supplemental information stating incomplete HRRs that only contain an analysis of some flooding hazard mechanisms would not be of substantive benefit for staff review and would not be acceptable. Therefore, Reference 3 recommends licensees not submit partial reports, but instead submit an extension request.
The purpose of this Enclosure is to request an extension to the March 12, 2014 due date for the submittal of the HRR for PINGP, Units 1 and 2. As recommended in Reference 3, the reasons for the delay, the proposed schedule for the submittal of the complete HRR, and the basis for the acceptability of the revised schedule are described below.
2.0 Reasons for the Delay Completion of Site-Specific Analyses:
NSPM plans to follow the hierarchical hazard approach (HHA) concept described in the Reference 1 letter to refine site specific data. The HHA is a progressively refined, stepwise estimation of site-specific hazards that evaluates the safety of Systems, Structures and Components (SSCs) with the most conservative plausible assumptions consistent with available data. The HHA process starts with the most conservative simplifying assumptions that maximize the hazards from the probable maximum event for each natural flood-causing phenomenon expected to occur in the vicinity of a proposed site. If the site is not inundated by floods from any of the phenomena to an elevation critical for safe operation of the SSCs, a conclusion that the SSCs are not susceptible to flooding would be valid, and no further flood-hazard assessment would Page 2 of 5
Enclosure NSPM Request For Extension be needed. However, if the level of assessed hazards results in an adverse effect or exposure to any safety related SSC, a more site-specific hazard assessment should be performed for the probable maximum event. Several iterations of the flood hazard assessment, each based on inclusion of additional site-specific data, may be needed to demonstrate that the assessed hazards from the probable maximum event are still based on conservative assumptions yet do not adversely affect the safety-related SSCs.
NSPM is working to refine the hazards assessment using more site-specific inputs and assumptions. The extended due date provides the additional time required for NSPM to perform the analytical refinements per the methodology described above.
US Army Corps of Engineers (USAGE) Dam Information:
Additional detailed technical information is needed to facilitate evaluation of the dams as part of the HHA approach. NSPM understands that the NRC is interfacing with the USAGE to obtain dam information and NSPM has submitted a request for technical assistance (Reference 4). NSPM intends to use the information from the USAGE evaluations as part of the above described analytical refinements.
3.0 Proposed Submittal Schedule NSPM will submit the required flood HRR for PINGP, Units 1 and 2, within 10 months of receiving the USAGE's final information. This timing supports incorporation of the USAGE's information on dams and the site-specific analytical refinements into the HRR.
NSPM continues to validate and improve site readiness in regard to implementation of existing flood protection features and mitigating strategies. These efforts include improvement of flooding procedures, simulation of flood protection actions to validate effectiveness, and improvement of site flood response readiness through pre-staging of materials.
4.0 Basis for Acceptability of the Revised Schedule NSPM's proposed submittal schedule extension is acceptable based on the discussion below:
The NRC's March 12, 2012, 50.54(f) cover letter states that the current regulatory approach and the resultant plant capabilities provide confidence that an accident with consequences similar to the Fukushima accident is unlikely to occur in the United States. The NRC letter concluded that continued plant operation and the continuation of licensing activities do not pose an imminent risk to public health and safety.
Page 3 of 5
Enclosure NSPM Request For Extension The events being analyzed in the PINGP HRR are beyond the plant's design and licensing basis as noted in Reference 3. In accordance with Enclosure 2 of the March 12, 2012, 50.54(f) letter item 1.d, interim evaluations and actions will be taken or planned to address any reevaluated higher flooding hazards relative to the design basis.
These interim evaluations and actions, if any, will be included in the complete PINGP HRR as part of the required response.
NSPM's position is that, assessed qualitatively, the reevaluated flooding hazard is an unlikely event and is not likely to occur within the extension request timeframe. The table below compares the Historical Maximum Observed River Elevation, the 1000 year flood near the site, the current design bases probable maximum flood, and the flood protection elevation for the PINGP site, which demonstrates the margin between the predicted maximum flood and PINGP's current flood protection.
Table - Licensing Basis Flood Elevations Elevation Description Elevation Historical Maximum Peak Stage 688ft (Approximate) 1 USACE Estimated 1000-year Flood 691.8 ft Elevation 2 Plant Grade (Approximate) 695ft Design Bases Probable Maximum Flood 703.6 ft Flood Protection Elevation (top of 705.0 ft substructure and/or superstructure)
Note 1: Maximum flood of record occurred in 1965 at Lock and Dam #3, approximately 1 mile downstream of the site.
Note 2: 1000 year peak stage at Lock and Dam #3, approximately 1 mile downstream of the site.
5.0
References:
- 1. NRC Letter, "Request for Information Pursuant to Title 10 of the Code of Federal Regulations 50.54(f) Regarding Recommendations 2.1, 2.3, and 9.3, of the Near-Term Task Force Review of Insights from the Fukushima Dai-ichi Accident," dated March 12, 2012 (ADAMS Accession No. ML12053A340).
- 2. NRC Letter, "Prioritization of Response Due Dates for Request for Information Pursuant to Title 10 of the Code of Federal Regulations Page 4 of 5
Enclosure NSPM Request For Extension 50.54(f) Regarding Flooding Hazard Reevaluations for Recommendation 2.1 of the Near-Term Task Force Review of Insights from the Fukushima Dai-ichi Accident," dated May 11, 2012 (ADAMS Accession No. ML12097A509).
- 3. NRC Letter, "Supplemental Information Related to Request for Information Pursuant to Title 10 of the Code of Federal Regulations 50.54(f)
Regarding Flooding Hazard Reevaluations for Recommendation 2.1 of the Near-Term Task Force Review of Insights from the Fukushima Dai-ichi Accident," dated March 1, 2013 (ADAMS Accession No. ML13044A561).
- 4. NSPM Letter, "Request for NRC Assistance to Obtain Information on Dams from the U.S. Army Corps of Engineers (USAGE)," dated March 4, 2014.
Page 5 of 5