Letter Sequence Request |
---|
|
|
MONTHYEARML12361A0062012-11-26026 November 2012 (Mns), Units 1 and 2, Flooding Walkdown Information Requested by 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 Project stage: Request ML13190A2722013-06-20020 June 2013 Response to NRC Request for Information Pursuant to 10 CFR 50.54(f) Re Seismic Aspects of Recommendation 2.3 of the Near Term Task Force Review of Insights from the Fukushima Dai-ichi Accident Project stage: Response to RAI ML13192A1552013-07-0101 July 2013 Update to Seismic Walkdown Information Submitted in November 27, 2012 Project stage: Request ML14083A5862014-03-12012 March 2014 EPRI to NEI - Fleet Seismic Core Damage Frequency Estimates for Central and Eastern United States Power Plants Using New Site-Specific Seismic Hazard Estimates: Attachment 1 Project stage: Request MNS-14-029, Seismic Hazard and Screening Report (CEUS Sites), Response to NRC 10 CFR 50.54(f) 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.2014-03-20020 March 2014 Seismic Hazard and Screening Report (CEUS Sites), Response to NRC 10 CFR 50.54(f) 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. Project stage: Request RC-14-0048, Seismic Hazard and Screening Report (CEUS Sites), 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 the Fukushima Dai-Ichi Accident2014-03-26026 March 2014 Seismic Hazard and Screening Report (CEUS Sites), 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 the Fukushima Dai-Ichi Accident Project stage: Response to RAI ML14093A0522014-03-31031 March 2014 Seismic Hazard and Screening Report (CEUS Sites), Response to NRC 10 CFR 50.54(f) RAI Per 10 CFR 50.54(f) Re Recommendations 2.1, 2.3 and 9.3 of the Near-Term Task Force Review of Insights from the Fukushima Dai-ichi Accident Project stage: Other RNP-RA/14-0013, Seismic Hazard Evaluation, Response to NRC 10 CFR 50.54(f) 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.2014-03-31031 March 2014 Seismic Hazard Evaluation, Response to NRC 10 CFR 50.54(f) 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. Project stage: Request ONS-2014-046, Seismic Hazard and Screening Report (CEUS Sites), Response to NRC 10 CFR 50.54(f) Request for Information Pursuant to Title 10 of the Code of Federal Regulations 50.54(f) Regarding Recommendations 2.1, 2.3..2014-03-31031 March 2014 Seismic Hazard and Screening Report (CEUS Sites), Response to NRC 10 CFR 50.54(f) Request for Information Pursuant to Title 10 of the Code of Federal Regulations 50.54(f) Regarding Recommendations 2.1, 2.3.. Project stage: Request ML14268A5162014-10-23023 October 2014 Request for Additional Information Associated with Near-Term Task Force Recommendation 2.1, Seismic Re-evaluation Related to Southeastern Catalog Changes Project stage: RAI RNP-RA/14-0118, Response to Request for Additional Information Associated with Near-Term Task Force Recommendation 2.1, Seismic Re-Evaluations Related to Southeastern Catalog Changes2014-11-12012 November 2014 Response to Request for Additional Information Associated with Near-Term Task Force Recommendation 2.1, Seismic Re-Evaluations Related to Southeastern Catalog Changes Project stage: Response to RAI ONS-2014-147, Response to the Nrc'S Request for Additional Information Dated October 23, 2014, Related to Southeastern Catalog Changes and Seismic Re-Evaluations2014-11-14014 November 2014 Response to the Nrc'S Request for Additional Information Dated October 23, 2014, Related to Southeastern Catalog Changes and Seismic Re-Evaluations Project stage: Request ML14303A1922014-11-25025 November 2014 Request for Additional Information Associated with Near-Term Task Force Recommendation 2.1, Seismic Hazard and Screening Report Project stage: RAI MNS-14-097, Expedited Seismic Evaluation Process Report (CEUS Sites), Response to NRC Request for Information Pursuant to Title 10 of the Code of Federal Regulations 50.54(f) Regarding Recommendations 2.1 of the Near-Term Task Force Review.2014-12-17017 December 2014 Expedited Seismic Evaluation Process Report (CEUS Sites), Response to NRC Request for Information Pursuant to Title 10 of the Code of Federal Regulations 50.54(f) Regarding Recommendations 2.1 of the Near-Term Task Force Review. Project stage: Response to RAI ONS-2014-161, Submittal of the Expedited Seismic Evaluation Process Report (CEUS Sites)2014-12-19019 December 2014 Submittal of the Expedited Seismic Evaluation Process Report (CEUS Sites) Project stage: Request CNS-14-130, Expedited Seismic Evaluation Process Report (CEUS Sites), 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 the Fukushima Dai-ichi Accident2014-12-31031 December 2014 Expedited Seismic Evaluation Process Report (CEUS Sites), 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 the Fukushima Dai-ichi Accident Project stage: Response to RAI CNS-14-132, Request for Additional Information Associated with Near-Term Task Force Recommendation 2.1, Seismic Hazard and Screening Report2015-01-0808 January 2015 Request for Additional Information Associated with Near-Term Task Force Recommendation 2.1, Seismic Hazard and Screening Report Project stage: Request ML15096A5132015-04-27027 April 2015 Staff Assessment of Information Provided Pursuant to Title 10 of the Code of Federal Regulations Part 50, Section 50.54(f), Seismic Hazard Revaluations Relating to Recommendation 2.1 of the NTTF Review Project stage: Other RC-15-0071, Response to NRC Request for Additional Information Regarding Seismic Evaluation Related to Southeastern Catalog Changes2015-04-28028 April 2015 Response to NRC Request for Additional Information Regarding Seismic Evaluation Related to Southeastern Catalog Changes Project stage: Response to RAI RNP-RA/15-0031, Response to NRC Request for Additional Information Regarding Seismic Re-Evaluations Related to Southeastern Catalog Changes2015-04-29029 April 2015 Response to NRC Request for Additional Information Regarding Seismic Re-Evaluations Related to Southeastern Catalog Changes Project stage: Response to RAI ONS-2015-055, Second Response to Request for Additional Information Dated October 23, 2014, Related to Southeastern Catalog Changes and Seismic Re-Evaluations2015-04-30030 April 2015 Second Response to Request for Additional Information Dated October 23, 2014, Related to Southeastern Catalog Changes and Seismic Re-Evaluations Project stage: Response to RAI ML15182A0672015-07-20020 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 Review of Insights From.. Project stage: Approval ML15201A0082015-07-22022 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 Review Project stage: Other RNP-RA/15-0069, Addendum to Submittal of Revision to Seismic Hazard Evaluation to Include New Ground Motion Response Spectra (GMRS) Using New Geotechnical Data and Shear-Wave Testing2015-08-29029 August 2015 Addendum to Submittal of Revision to Seismic Hazard Evaluation to Include New Ground Motion Response Spectra (GMRS) Using New Geotechnical Data and Shear-Wave Testing Project stage: Request ML15226A1852015-09-0808 September 2015 Nuclear Regulatory Commission Plan for the Audit of Duke Energy Carolinas, LLC Interim Evaluations Associated with Reevaluated Seismic Hazard Implementing Near-Term Task Force Recommendation 2.1 - Seismic Project stage: Other ML15280A1992015-10-19019 October 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 Review Project stage: Other ML16034A3252015-12-31031 December 2015 Revision 0 to Summary of Methodology Used for Seismic Capacity Evaluation of GMRS Mitigation Components Project stage: Other ONS-2016-007, Supplemental Information Related to the Expedited Seismic Evaluation Process (ESEP) Report in Response to the NRC Audit Described in Letter Dated September 2, 20152016-01-27027 January 2016 Supplemental Information Related to the Expedited Seismic Evaluation Process (ESEP) Report in Response to the NRC Audit Described in Letter Dated September 2, 2015 Project stage: Supplement 2014-03-26
[Table View] |
|
---|
Category:Letter
MONTHYEARML24023A0392024-01-22022 January 2024 NEI Comments on the Information Collection Renewal for Domestic Licensing of Special Nuclear Material, Docket Id NRC-2023-0118 ML23355A1972023-12-14014 December 2023 NEI, Comments on NRC Draft Resolution of SFAQ 2022-02, SAE Program Requirements ML23219A1672023-10-25025 October 2023 Response Letter to Fee Exemption Request for Pre-Submittal Activities, Review, and Endorsement of NEI 20-07 ML23270B9002023-09-27027 September 2023 NEI Letter Request for an Extension of Comment Period on Proposed Revision to Standard Review Plan Section 15.0, Introduction - Transient and Accident Analyses, Docket Id NRC 2023 0079 ML23268A0102023-09-22022 September 2023 NEI, Fee Exemption Request for Endorsement, Review and Meeting to Discuss Draft Nuclear Energy Institute Technical Report NEI 23-01, Operator Cold License Training Plan for Advanced Nuclear Reactors ML23241A8612023-08-25025 August 2023 Consolidated Industry Comments to NRC Regulatory Issue Summary 2023-02, Scheduling Information for the Licensing of Accident Tolerant, Increased Enrichment, and Higher Burnup Fuels ML23236A4992023-08-24024 August 2023 Industry Feedback on Region II Fuel Cycle Facility Construction Oversight Workshop Held August 15, 2023, and Suggested Topics for Additional Public Meetings in Fall 2023 ML23256A1622023-08-0101 August 2023 Incoming NEI Letter Dated August 1, 2023 Regarding Increase in Fees 2023-2025 ML23206A0292023-07-24024 July 2023 Incoming Fee Exemption Request for Pre-Submittal Activities, Review, and Endorsement of NEI 20-07 ML23143A1232023-06-22022 June 2023 NRC Fee Waiver Request for Draft NEI 23-01 ML23200A1662023-05-30030 May 2023 NEI Proposed Metrics for a Performance-Based Emergency Preparedness Program ML23116A0732023-05-25025 May 2023 Letter to Hillary Lane in Response to a Request for a Fee Exemption for NEI 23-03 ML23135A7332023-05-0909 May 2023 NEI Comments on NRC Safety Culture Program Effectiveness Review ML23110A6752023-04-18018 April 2023 04-18-23_NRC_Fee Waiver for NEI 23-03 ML23110A6782023-04-18018 April 2023 Request for Review and Endorsement of NEI 23-03, Supplemental Guidance for Application of 10 CFR 50.59 to Digital Modifications at Non-Power Production or Utilization Facilities ML23110A6762023-04-18018 April 2023 04-18-23_NRC_NEI 23-03 Review + Endorse ML23107A2302023-03-31031 March 2023 NEI Letter, to Andrea Veil, NRC, Regarding Industry Recommendations for a 10 CFR 50.46a/c Combined Rulemaking ML23138A1662023-03-24024 March 2023 Transmittal of NEI 22-05 Revision a, Technology Inclusive Risk Informed Change Evaluation (Tirice) Guidance for the Evaluation of Changes to Facilities Utilizing NEI 18-04 and NEI 21-07 ML23083B4622023-03-24024 March 2023 Transmittal of NEI 22-05 Revision a, Technology Inclusive Risk Informed Change Evaluation (Tirice) Guidance for the Evaluation of Changes to Facilities Utilizing NEI 18-04 and NEI 21-07 ML23060A3272023-03-0101 March 2023 NEI, Wireless Cyber Security Guidance ML23060A2142023-03-0101 March 2023 NEI, Request for NRC Endorsement of NEI White Paper, Enabling a Remote Response by Members of an Emergency Response Organization, Revision 0 ML23023A2752023-01-23023 January 2023 Request for Extension of Comment Period from the Nuclear Energy Institute on PRM-50-124 - Licensing Safety Analysis for Loss-of-Coolant Accidents ML22348A1122023-01-17017 January 2023 Letter to Richard Mogavero Response to Fee Exemption NEI 08-09 Revision 7 ML22353A6082023-01-11011 January 2023 U.S. Nuclear Regulatory Commission Report of the Regulatory Audit of the NEI-Proposed Aging Management Program Revision to Selective Leaching Program (XI.M33) ML22349A1012022-12-12012 December 2022 LTR-22-0343 Ellen Ginsberg, Sr. Vice President, General Counsel and Secretary, Nuclear Energy Institute, Expresses Concerns Related to Issuance of Regulatory Issue Summary 2022-02; Operational Leakage ML22336A0372022-11-16016 November 2022 Fee Exemption Request for NEI 08-09 Revision 7 - Changes to NEI 08-09 Cyber Security Plan for Nuclear Power Reactors ML22321A3152022-11-16016 November 2022 NEI Letter with Comments on Significance Determination Process Timeliness Review ML22298A2262022-10-25025 October 2022 Endorsement of NEI 15-09, Cyber Security Event Notifications, Revision 1, Dated October 2022 ML22298A2302022-10-17017 October 2022 Submittal of NEI 22-03, Draft Revision 0, Nuclear Generation Quality Assurance Program Description ML22207B6512022-07-26026 July 2022 NEI, Full Fee Exemption Request for Industry Guidance Proposal - Weather Related Administrative Controls During Transient Outdoor Dry Cask Operations ML22195A1662022-07-14014 July 2022 NEI, Draft G of NEI 99-01, Development of Emergency Action Levels for Non-Passive Reactors, Revision 7 ML22195A0202022-07-13013 July 2022 07-13-22 NRC Fee Exemption Request for NEI 21-05 Review ML22195A0672022-07-13013 July 2022 Fee Exemption Request for Review and Meeting to Discuss Draft Nuclear Energy Institute Technical Report NEI 21-05, Reporting Guidance for Licensees with Risk-Informed Licensing Bases ML22159A2772022-06-28028 June 2022 Response Letter to Richard Mogavero for Fee Exemption for the Nuclear Regulatory Commission Review Ad Endorsement of NEI 15-09, Revision 1 ML22153A2782022-06-0202 June 2022 Nie, Fee Exemption Request for Endorsement of NEI 15-09, Cyber Security Event Notifications, Revision 1, Dated May 2022 ML22154A2962022-06-0202 June 2022 LTR from R. Mogavero to M. Sampson Dated Jun 2 2022 Endorsement of NEI 15-09 Cyber Security Event Notifications Rev 1 Dated May 2022 ML22152A2712022-06-0101 June 2022 Digital Instrumentation and Control Common Cause Failure Policy Considerations, Revision 1 ML22143A9362022-05-20020 May 2022 May 13, 2022, Public Meeting on Draft Regulatory Issue Summary Operational Leakage, 87 Fed. Reg. 2361 (Jan. 14, 2022) (Docket Id NRC-2021-0173) ML22110A1752022-05-0303 May 2022 NRC Response to the Nuclear Energy Institute April 1, 2022, Letter, Regarding the Nrc'S CUI Implementation Plan ML22109A2082022-04-0808 April 2022 April 8, 2022, NEI White Paper on Digital Instrumentation and Control Common Cause Policy Considerations Version 2.0 ML22110A1782022-04-0101 April 2022 April 1, 2022, Letter from NEI Regarding Nrc'S Controlled Unclassified Information Program Implementation ML22048A5812022-02-16016 February 2022 NEI 22-02: Guidelines for Weather-Related Administrative Controls for Short Duration Outdoor Dry Cask Storage Operations ML22019A2922022-01-12012 January 2022 NEI, Submittal of Proposed Revisions to Aging Management Programs XI.M33, Selective Leaching and XI.E3, Inaccessible Power Cables Not Subject to 10 CFR 50.49 Environmental Qualification Requirements ML21343A2922021-12-0808 December 2021 NEI, Transmittal of NEI 21-07 Revision 0-B, Technology Inclusive Guidance for Non-Light Water Reactor Safety Analysis Report: for Applicants Utilizing NEI 18-04 Methodology ML21337A3802021-12-0303 December 2021 NEI Technical Report NEI 17-06 - Guidance on Using Iec 61508 SIL Certification to Support the Acceptance of Commercial Grade Digital Equipment for Nuclear Safety Related Applications, Revision 1 ML21319A3522021-11-10010 November 2021 NRC NEI Fee Waiver Request Changes to NEI 10-04 and NEI 13-10, Dated November 10, 2021 ML21306A3652021-10-29029 October 2021 NEI Letter from D. Young to NRC S. Atack to Cease Work on Draft D of NEI 20-05, Methodological Approach and Considerations for a Technical Analysis to Demonstrate Compliance with the Eligibility Criteria of 10 CFR 73.55(a)(7) ML21342A1682021-10-29029 October 2021 Letter from W. Gross to S. Atack, Endorsement of Nuclear Energy Institute 10-04, Identifying Systems and Assets Subject to the Cyber Security Rule, Revision 3, Dated October 29, 2021 ML21342A2032021-10-29029 October 2021 Letter from W. Gross to S. Atack, Endorsement of Nuclear Energy Institute 13-10, Cyber Security Control Assessments, Revision 7, Dated October 29, 2021 ML22081A2002021-10-29029 October 2021 NEI Backfitting Concerns with NRCs Developing Position on Protection of Dry Storage Systems from Natural Phenomena During Short Term Operations 2024-01-22
[Table view] Category:Report
MONTHYEARML23290A1252023-10-17017 October 2023 NEI - NEI 99-02, Rev. 8, Draft Regulatory Assessment Performance Indicator Guideline ML23290A1472023-10-17017 October 2023 NEI 99-02 Rev 8 Draft 9 29 2023 Redline Version ML23157A1062023-06-0606 June 2023 NEI 19-01, Rev 1, Safety and Economic Benefits of Accident Tolerant Fuel ML23129A0282023-05-0202 May 2023 20230502, NEI Issues for Event Notification Implementation Workshops ML23110A6762023-04-18018 April 2023 04-18-23_NRC_NEI 23-03 Review + Endorse ML23138A1662023-03-24024 March 2023 Transmittal of NEI 22-05 Revision a, Technology Inclusive Risk Informed Change Evaluation (Tirice) Guidance for the Evaluation of Changes to Facilities Utilizing NEI 18-04 and NEI 21-07 ML23083B4622023-03-24024 March 2023 Transmittal of NEI 22-05 Revision a, Technology Inclusive Risk Informed Change Evaluation (Tirice) Guidance for the Evaluation of Changes to Facilities Utilizing NEI 18-04 and NEI 21-07 ML23125A3202023-03-0101 March 2023 Guidance for the Evaluation of Changes to Facilities Utilizing NEI 18-04 and NEI 21-07 Revision a 5-4-23 Tirice Response to NRC Comments ML22298A2282022-10-25025 October 2022 NEI 15-09, Rev. 1, Cybersecurity Event Notifications ML22297A2482022-10-20020 October 2022 NEI Comments on 10-20-2022 CCF Meeting Feedback and Comments ML23072A0632022-09-30030 September 2022 (Draft) NEI White Paper Remediation of Vulnerabilities Identified in CDAs - 08302022R0 ML22195A1692022-07-31031 July 2022 NEI, Draft G of NEI 99-01, Development of Emergency Action Levels for Non-Passive Reactors, Revision 7 - Change Summary ML22195A1672022-07-14014 July 2022 NEI, Draft G of NEI 99-01, Development of Emergency Action Levels for Non-Passive Reactors, Revision 7 ML22195A1682022-07-14014 July 2022 NEI, Marked-Up to Draft G of NEI 99-01, Development of Emergency Action Levels for Non-Passive Reactors, Revision 7 ML22109A2082022-04-0808 April 2022 April 8, 2022, NEI White Paper on Digital Instrumentation and Control Common Cause Policy Considerations Version 2.0 ML22048A5812022-02-16016 February 2022 NEI 22-02: Guidelines for Weather-Related Administrative Controls for Short Duration Outdoor Dry Cask Storage Operations ML22019A2922022-01-12012 January 2022 NEI, Submittal of Proposed Revisions to Aging Management Programs XI.M33, Selective Leaching and XI.E3, Inaccessible Power Cables Not Subject to 10 CFR 50.49 Environmental Qualification Requirements ML21343A2822021-12-31031 December 2021 Redline/Strikeout Version of NEI 21-07 Rev 0-B, Technology Inclusive Guidance for Non-Light Water Reactors Safety Analysis Report Content for Applicants Using the NEI 18-04 Methodology ML21343A2922021-12-0808 December 2021 NEI, Transmittal of NEI 21-07 Revision 0-B, Technology Inclusive Guidance for Non-Light Water Reactor Safety Analysis Report: for Applicants Utilizing NEI 18-04 Methodology ML21337A3802021-12-0303 December 2021 NEI Technical Report NEI 17-06 - Guidance on Using Iec 61508 SIL Certification to Support the Acceptance of Commercial Grade Digital Equipment for Nuclear Safety Related Applications, Revision 1 ML21305A0012021-11-30030 November 2021 NEI 17-06 Rev 0 Draft B, Guidance on Using Iec 61508 SIL Certification to Support the Acceptance of Commercial Grade Digital Equipment for Nuclear Safety Related Applications (Staff Comments Incorporated ML21342A2032021-10-29029 October 2021 Letter from W. Gross to S. Atack, Endorsement of Nuclear Energy Institute 13-10, Cyber Security Control Assessments, Revision 7, Dated October 29, 2021 ML21342A1682021-10-29029 October 2021 Letter from W. Gross to S. Atack, Endorsement of Nuclear Energy Institute 10-04, Identifying Systems and Assets Subject to the Cyber Security Rule, Revision 3, Dated October 29, 2021 ML21274A0312021-10-0101 October 2021 NRC Draft Detailed Comments Related to NEI 21-07, Revision 0, Technology Inclusive Guidance for Non-Light Water Reactors Safety Analysis Report Content for Applicants Using the NEI 18-04 Methodology ML21278A4722021-09-30030 September 2021 NEI 20-07, Rev. Draft, Guidance for Addressing Common Cause Failure in High Safety-Significant Safety-Related Digital I&C Systems ML21257A2352021-08-19019 August 2021 Rulemaking: Proposed Rule: Advanced Reactor Physical Security, Email Exchange Between Nrc and NEI Draft Pages from NEI-20-05 Rev. E ML21250A3782021-08-0202 August 2021 NEI 21-07, Revision 0, Technology Inclusive Guidance for Non-Light Water Reactors SAR Content for Applicants Using the NEI 18-04 Methodology ML21130A5962021-05-31031 May 2021 NEI, Guidance for Addressing Software Common Cause Failure in High Safety-Significant Safety Related Digital I&C Systems - Draft C ML21130A5972021-05-31031 May 2021 NEI 20-07, Guidance for Addressing Software Common Cause Failure in High Safety-Significant Safety Related Digital I&C Systems - Draft C ML21130A5982021-05-0707 May 2021 NEI Responses to NRC Staff Comments on NEI 20-07 Draft B ML21125A2842021-05-0505 May 2021 Transmittal of NEI 20-09: Performance of PRA Peer Reviews Using the Asme/Ans Advanced Non-Light Water Reactor Standard ML21110A0662021-04-20020 April 2021 Nei'S Comparison Table Between NEI 20-07 Sdos and NRC RGs and Endorsed IEEE Stds R2 ML21085A5552021-03-25025 March 2021 NEI 20-09 -Performance of PRA Peer Reviews Using the Asme/Ans Advanced Non-LWR PRA Standard, March 2021 ML21089A0902021-03-18018 March 2021 NEI Comments on Renewal of Performance Indicators Information Collection, March 18, 2021 ML21049A0572021-03-0202 March 2021 Rulemaking: Proposed Rule: NRC Markup of NEI-20-05 Draft B Comments on Methodological Approach and Considerations for a Technical Analysis to Demonstrate Compliance with the Performance Criteria of 10 CFR 73.55(a)(7) ML20339A4852020-11-23023 November 2020 NEI 20-09 - NRC Comments Resolved November 2020 ML20322A3392020-11-17017 November 2020 NEI ROP White Paper Modification of the Description of Unplanned Scrams with Complications for Nov 18 2020 ROP Public Meeting ML21050A0902020-08-31031 August 2020 Staff Detailed Comments - NEI 20_07 Draft Revision B -February 2021 ML20245E1472020-08-31031 August 2020 Attachment 1 - NEI Guidelines for the Implementation of the Risk-Informed Process for Evaluations Integrated Decision-Making Panel ML20245E5612020-08-31031 August 2020 Guidance for Addressing Software Common Cause Failure in High Safety-Significant Safety Related Digital I&C Systems ML20302A1152020-08-24024 August 2020 NEI 20-09 - Nlwr PRA Peer Review Rev1 August 2020 ML20211L7142020-07-24024 July 2020 Industry Position Regarding Safety Margin: Dispositioning Degraded or Failed Management Measures Above and Beyond Regulatory Requirements, and Meeting Performance Criteria; Follow Up to May 6, 2020 Letter on Smarter Program Inspection Prior ML20155K6852020-06-0101 June 2020 Availability of NEI 20-09, Performance of PRA Peer Reviews Using the Asme/Ans Advanced Non-LWR Standard, for NRC Review and Endorsement ML20154K5662020-06-0101 June 2020 Availability of NEI 20-09, Performance of PRA Peer Reviews Using the Asme/Ans Advanced Non-LWR Standard, for NRC Review and Endorsement ML20129J8592020-05-31031 May 2020 NEI 96-07, Appendix D Revision 1, Draft M, May 2020 ML20129J8582020-05-30030 May 2020 NEI 96-07, Appendix D, Revision 1, Draft M, May 2020 with Redline Strike ML20141L7882020-05-20020 May 2020 NEI - Comments on Draft Micro-Reactor Applications COL-ISG-029, Environmental Considerations Associated with Micro-Reactors ML20139A1902020-05-14014 May 2020 05-14-20 Changes to NEI 10-04 and NEI 13-10 Guidance for Identifying and Protecting Digital Assets Associated with Safety-Related and Important-to-Safety Functions ML20134J0332020-05-13013 May 2020 Submittal of Response to Request for Additional Information (RAI) for NEI 14-05A, Revision 1, Guidelines for the Use of Accreditation in Lieu of Commercial Grade Surveys for Procurement of Laboratory Calibration and Test Services ML20135H1682020-05-13013 May 2020 Request for NRC Endorsement of NEI 96-07, Appendix D, Rev 1 2023-06-06
[Table view] Category:Technical
MONTHYEARML23290A1472023-10-17017 October 2023 NEI 99-02 Rev 8 Draft 9 29 2023 Redline Version ML23290A1252023-10-17017 October 2023 NEI - NEI 99-02, Rev. 8, Draft Regulatory Assessment Performance Indicator Guideline ML23157A1062023-06-0606 June 2023 NEI 19-01, Rev 1, Safety and Economic Benefits of Accident Tolerant Fuel ML23110A6762023-04-18018 April 2023 04-18-23_NRC_NEI 23-03 Review + Endorse ML23083B4622023-03-24024 March 2023 Transmittal of NEI 22-05 Revision a, Technology Inclusive Risk Informed Change Evaluation (Tirice) Guidance for the Evaluation of Changes to Facilities Utilizing NEI 18-04 and NEI 21-07 ML23138A1662023-03-24024 March 2023 Transmittal of NEI 22-05 Revision a, Technology Inclusive Risk Informed Change Evaluation (Tirice) Guidance for the Evaluation of Changes to Facilities Utilizing NEI 18-04 and NEI 21-07 ML23125A3202023-03-0101 March 2023 Guidance for the Evaluation of Changes to Facilities Utilizing NEI 18-04 and NEI 21-07 Revision a 5-4-23 Tirice Response to NRC Comments ML22298A2282022-10-25025 October 2022 NEI 15-09, Rev. 1, Cybersecurity Event Notifications ML22195A1692022-07-31031 July 2022 NEI, Draft G of NEI 99-01, Development of Emergency Action Levels for Non-Passive Reactors, Revision 7 - Change Summary ML22195A1682022-07-14014 July 2022 NEI, Marked-Up to Draft G of NEI 99-01, Development of Emergency Action Levels for Non-Passive Reactors, Revision 7 ML22195A1672022-07-14014 July 2022 NEI, Draft G of NEI 99-01, Development of Emergency Action Levels for Non-Passive Reactors, Revision 7 ML22048A5812022-02-16016 February 2022 NEI 22-02: Guidelines for Weather-Related Administrative Controls for Short Duration Outdoor Dry Cask Storage Operations ML21343A2822021-12-31031 December 2021 Redline/Strikeout Version of NEI 21-07 Rev 0-B, Technology Inclusive Guidance for Non-Light Water Reactors Safety Analysis Report Content for Applicants Using the NEI 18-04 Methodology ML21343A2922021-12-0808 December 2021 NEI, Transmittal of NEI 21-07 Revision 0-B, Technology Inclusive Guidance for Non-Light Water Reactor Safety Analysis Report: for Applicants Utilizing NEI 18-04 Methodology ML21337A3802021-12-0303 December 2021 NEI Technical Report NEI 17-06 - Guidance on Using Iec 61508 SIL Certification to Support the Acceptance of Commercial Grade Digital Equipment for Nuclear Safety Related Applications, Revision 1 ML21305A0012021-11-30030 November 2021 NEI 17-06 Rev 0 Draft B, Guidance on Using Iec 61508 SIL Certification to Support the Acceptance of Commercial Grade Digital Equipment for Nuclear Safety Related Applications (Staff Comments Incorporated ML21342A1682021-10-29029 October 2021 Letter from W. Gross to S. Atack, Endorsement of Nuclear Energy Institute 10-04, Identifying Systems and Assets Subject to the Cyber Security Rule, Revision 3, Dated October 29, 2021 ML21342A2032021-10-29029 October 2021 Letter from W. Gross to S. Atack, Endorsement of Nuclear Energy Institute 13-10, Cyber Security Control Assessments, Revision 7, Dated October 29, 2021 ML21274A0312021-10-0101 October 2021 NRC Draft Detailed Comments Related to NEI 21-07, Revision 0, Technology Inclusive Guidance for Non-Light Water Reactors Safety Analysis Report Content for Applicants Using the NEI 18-04 Methodology ML21278A4722021-09-30030 September 2021 NEI 20-07, Rev. Draft, Guidance for Addressing Common Cause Failure in High Safety-Significant Safety-Related Digital I&C Systems ML21257A2352021-08-19019 August 2021 Rulemaking: Proposed Rule: Advanced Reactor Physical Security, Email Exchange Between Nrc and NEI Draft Pages from NEI-20-05 Rev. E ML21250A3782021-08-0202 August 2021 NEI 21-07, Revision 0, Technology Inclusive Guidance for Non-Light Water Reactors SAR Content for Applicants Using the NEI 18-04 Methodology ML21130A5972021-05-31031 May 2021 NEI 20-07, Guidance for Addressing Software Common Cause Failure in High Safety-Significant Safety Related Digital I&C Systems - Draft C ML21130A5962021-05-31031 May 2021 NEI, Guidance for Addressing Software Common Cause Failure in High Safety-Significant Safety Related Digital I&C Systems - Draft C ML21125A2842021-05-0505 May 2021 Transmittal of NEI 20-09: Performance of PRA Peer Reviews Using the Asme/Ans Advanced Non-Light Water Reactor Standard ML21110A0662021-04-20020 April 2021 Nei'S Comparison Table Between NEI 20-07 Sdos and NRC RGs and Endorsed IEEE Stds R2 ML21085A5552021-03-25025 March 2021 NEI 20-09 -Performance of PRA Peer Reviews Using the Asme/Ans Advanced Non-LWR PRA Standard, March 2021 ML21049A0572021-03-0202 March 2021 Rulemaking: Proposed Rule: NRC Markup of NEI-20-05 Draft B Comments on Methodological Approach and Considerations for a Technical Analysis to Demonstrate Compliance with the Performance Criteria of 10 CFR 73.55(a)(7) ML20339A4852020-11-23023 November 2020 NEI 20-09 - NRC Comments Resolved November 2020 ML20322A3392020-11-17017 November 2020 NEI ROP White Paper Modification of the Description of Unplanned Scrams with Complications for Nov 18 2020 ROP Public Meeting ML20245E5612020-08-31031 August 2020 Guidance for Addressing Software Common Cause Failure in High Safety-Significant Safety Related Digital I&C Systems ML20245E1472020-08-31031 August 2020 Attachment 1 - NEI Guidelines for the Implementation of the Risk-Informed Process for Evaluations Integrated Decision-Making Panel ML21050A0902020-08-31031 August 2020 Staff Detailed Comments - NEI 20_07 Draft Revision B -February 2021 ML20302A1152020-08-24024 August 2020 NEI 20-09 - Nlwr PRA Peer Review Rev1 August 2020 ML20211L7142020-07-24024 July 2020 Industry Position Regarding Safety Margin: Dispositioning Degraded or Failed Management Measures Above and Beyond Regulatory Requirements, and Meeting Performance Criteria; Follow Up to May 6, 2020 Letter on Smarter Program Inspection Prior ML20155K6852020-06-0101 June 2020 Availability of NEI 20-09, Performance of PRA Peer Reviews Using the Asme/Ans Advanced Non-LWR Standard, for NRC Review and Endorsement ML20154K5662020-06-0101 June 2020 Availability of NEI 20-09, Performance of PRA Peer Reviews Using the Asme/Ans Advanced Non-LWR Standard, for NRC Review and Endorsement ML20129J8592020-05-31031 May 2020 NEI 96-07, Appendix D Revision 1, Draft M, May 2020 ML20129J8582020-05-30030 May 2020 NEI 96-07, Appendix D, Revision 1, Draft M, May 2020 with Redline Strike ML20139A1902020-05-14014 May 2020 05-14-20 Changes to NEI 10-04 and NEI 13-10 Guidance for Identifying and Protecting Digital Assets Associated with Safety-Related and Important-to-Safety Functions ML20135H1682020-05-13013 May 2020 Request for NRC Endorsement of NEI 96-07, Appendix D, Rev 1 ML20134J0332020-05-13013 May 2020 Submittal of Response to Request for Additional Information (RAI) for NEI 14-05A, Revision 1, Guidelines for the Use of Accreditation in Lieu of Commercial Grade Surveys for Procurement of Laboratory Calibration and Test Services ML20126G3742020-04-30030 April 2020 NEI 96-07, Appendix D, Revision 1, Draft F, April 2020 with Redline Strike ML20126G3752020-04-30030 April 2020 NEI 96-07, Appendix D, Revision 1, Draft F, April 2020 ML20115E4132020-04-24024 April 2020 Endorsement of NEI White Paper, Changes to NEI 10-04 and NEI 13-10 Guidance for Identifying and Protecting Digital Assets Associated with the Balance of Plant, Dated April 2020 ML20114E1952020-04-22022 April 2020 NEI 96-07 Appendix D Revision 1, Draft B, April 2020 with Red-line Strike ML20107D8942020-04-13013 April 2020 Email from D. Young Draft B of NEI 20-05, Methodological Approach and Considerations for a Security Assessment to Demonstrate Compliance with the Performance Criteria of 10 CFR 73.55(TBD) ML20104A3072020-04-10010 April 2020 Draft NEI 20-05, Methodological Approach and Considerations for a Security Assessment to Demonstrate Compliance with the Performance Criteria of 10 CFR 73.55(TBD) ML20092J9202020-04-0101 April 2020 NEI 96-07 Appendix D Rev 1 February 2020 ML20092J9162020-04-0101 April 2020 NEI 96-07 Appendix D Rev 1 February 2020_Red Line 2023-06-06
[Table view] |
Text
ATTACHMENT 1 March 11, 2014 Anthony R. Pietrangelo Senior Vice President and Chief Nuclear Officer, Nuclear Generation Nuclear Energy Institute 1201 F Street, NW, Suite 1100 Washington, DC 20004
Subject:
Fleet Seismic Core Damage Frequency Estimates for Central and Eastern U.S. Nuclear Power Plants Using New Site-Specific Seismic Hazard Estimates
Dear Mr. Pietrangelo:
The Electric Power Research Institute (EPRI) has recently completed site-specific seismic hazard evaluations for nuclear plants in the central and eastern United States (CEUS) using the guidance in Electric Power Research Institute (EPRI) 1025287 (EPRI 2013a). To provide perspective regarding the safety implications of these new seismic hazard estimates, EPRI has performed an initial assessment of the changes in the seismic core-damage frequency relative to earlier fleet-wide estimates.
A description of the fleet evaluation is attached.
If you have questions or would like to discuss this evaluation, please contact John Richards at 704-595-2707 or jrichards@epri.com.
Sincerely, Stuart Lewis Program Manager Risk and Safety Management RSM-031114-077 Attachment
Fleet Seismic Core Damage Frequency Estimates for Central and Eastern U.S. Nuclear Power Plants Using New Site-Specific Seismic Hazard Estimates Electric Power Research Institute Project Manager J. Richards This evaluation was prepared by Simpson Gumpertz & Heger Inc., under contract to the Electric Power Research Institute.
The principal authors are G. Hardy, T. Graf, F. Grant, and Y. Tang.
1 BACKGROUND Following the accident at the Fukushima Daiichi Nuclear Power Plant resulting from the March 11, 2011, Great Tohoku Earthquake and subsequent tsunami, the U.S. Nuclear Regulatory Commission (USNRC) established a Near Term Task Force (NTTF) to conduct a systematic review of USNRC processes and regulations, and to determine if the agency should make additional improvements to its regulatory system. The NTTF developed a set of recommendations intended to clarify and strengthen the regulatory framework for protection against natural phenomena such as earthquakes. Subsequently, the USNRC issued a 50.54(f) letter that requests information to ensure that all U.S. nuclear power plants address these recommendations. This letter requests that licensees and holders of construction permits under 10 CFR Part 50 reevaluate the seismic hazards at their sites against present-day USNRC requirements and guidance.
In response to the 50.54(f) letter, site-specific seismic hazard estimates have been developed for nuclear plants in the central and eastern United States (CEUS) using the guidance in Electric Power Research Institute (EPRI) 1025287 (EPRI, 2013a) and in a 2013 letter from the Nuclear Energy Institute (NEI, 2013). These hazards form the basis for determining whether further seismic evaluation may be needed on a plant-by-plant basis.
The USNRC has requested that interim actions (that is, actions that can be implemented before more extensive seismic evaluations could be completed) be taken for plants whose ground motion response spectrum (GMRS) exceeds the design basis (USNRC, 2012; USNRC, 2014).
In response to this request, the U.S. nuclear industry proposed an Expedited Seismic Evaluation Process (ESEP) as an effective interim action. Guidance for conducting such an evaluation was developed by EPRI (EPRI, 2013b), and the process and guidance were endorsed by the USNRC (USNRC, 2013). The expedited evaluation is being carried out for any site with a GMRS that exceeds the safe shutdown earthquake (SSE) in the spectral frequency range from 1 to 10 Hz. As an input to the consideration of whether additional interim actions may be warranted, EPRI has estimated, for the fleet of nuclear power plants operating in the CEUS, the seismic core-damage frequencies (SCDFs) based on the newly completed site-specific seismic hazards.
1
2 OBJECTIVES Because it does not explicitly account for the capability of a nuclear power plant to maintain a safe condition during an earthquake, the GMRS calculated from the new seismic hazard characterization provides an incomplete perspective regarding overall seismic safety. The objective of this study is to provide an initial assessment of the safety implications of the new seismic hazard estimates across the CEUS fleet of operating plants. This assessment involves comparing SCDF estimates reflecting the new seismic hazard estimates for the fleet of operating plants to SCDF estimates previously developed by the USNRC in its 2010 Safety / Risk Assessment for GI-199 (USNRC, 2010). To perform this assessment, point estimates of the SCDF have been developed using (1) the methods defined by the USNRC in the 2010 Safety / Risk Assessment for GI-199, (2) the plant-level fragilities determined by the USNRC in the GI-199 Assessment, and (3) new site-specific seismic hazard estimates. The resulting SCDF estimates are compared with the baseline SCDFs developed by the USNRC in 2010 using the 2008 U.S. Geological Survey (USGS) and 1994 Lawrence Livermore National Laboratory (LLNL) seismic hazard curves. These are, respectively, the most recent seismic hazard assessment available at the time of the 2010 study, and the hazard assessment used by the USNRC in its review of seismic evaluations submitted as part of the Individual Plant Examination of External Events (IPEEE) in the 1990s.
3 ESTIMATING SEISMIC CORE DAMAGE FREQUENCY As described in Section 1, new probabilistic seismic hazard analyses (PSHA) have been completed for all U.S. nuclear power plant sites located in the CEUS. The potential safety and risk implications of these new seismic hazard estimates can most comprehensively be assessed with a modern Seismic Probabilistic Risk Assessment (SPRA) in accordance with the PRA Standard (ASME, 2013), but these modern SPRAs are not yet available for most plants. In 2010, the USNRC used a simplified approach to estimate the SCDF for all of the CEUS plants as part of the GI-199 program. The GI-199 program was associated with the changing understanding of seismic hazards in much of the United States and the implications of that understanding for nuclear plant safety. The USNRC simplified seismic risk estimation approach involved estimating the plant seismic fragility (i.e., conditional probability of plant damage at a given seismic hazard input level) from the results of the earlier IPEEE submittals, and convolving that plant fragility estimate with the new seismic hazard to obtain an SCDF estimate.
EPRI is conducting a similar assessment of SCDFs for the fleet of CEUS plants using the same IPEEE-derived plant-level fragilities combined with the new site-specific seismic hazard curves.
4 SITE-SPECIFIC SEISMIC HAZARDS The first major step in responding to Enclosure 1 of the 50.54(f) letter (USNRC, 2012) is to calculate seismic hazards at existing plant sites following the USNRC endorsed guidance in the Screening, Prioritization and Implementation Details (SPID) (EPRI, 2013a). These seismic hazards incorporate PSHA methods using the recently developed CEUS Seismic Source Characterization (CEUS-SSC) for Nuclear Facilities (CEUS-SSC, 2012), together with an updated ground-motion model (GMM) for the CEUS (EPRI, 2013c), and site-specific site amplification calculations. CEUS plants will submit these site-specific seismic hazards by March 31, 2014, in accordance with NEIs letter dated 9 April 2013 (NEI, 2013). These newly developed seismic hazard characterizations were used for the subject fleet SCDF calculations.
2
5 PLANT-LEVEL FRAGILITIES Plant-level fragility curves for each GI-199 plant were developed by the USNRC as part of the 2010 Safety / Risk Assessment based on information provided in the IPEEE submittals.
Appendix C of the USNRC GI-199 report (USNRC, 2010) defines three methods for estimating a plant-level fragility from information reported in the IPEEE submittals. The methods are briefly summarized in Table C.1 of the USNRC report (2010), which is reproduced below as Table 1.
About one-third of the plants in the CEUS performed an SPRA as part of their IPEEE program.
Many of the plants that performed SPRAs provided plant-level fragility information in their IPEEE submittals (Method 1a below), and the remaining plants that performed SPRAs provided SCDF estimates based on a variety of seismic hazard curves (EPRI 1989, LLNL1994, or site-specific curves developed specifically for the IPEEE program). For these remaining plants, plant-level fragility values were approximated in the USNRC GI-199 assessment by estimating and matching the reported SCDFs and using engineering judgment (Methods 1b and 1c below).
In cases where reasonable engineering judgments could not be readily made, the USNRC performed sensitivity studies to estimate the potential plant level fragilities. This resulted in more than one potential plant level fragility for a number of specific plant sites / units.
Two-thirds of the plants conducted a seismic margins analysis (SMA) as part of their IPEEE program. For these plants, the USNRC estimated the plant-level fragility based on the reported plant-level high confidence of a low probability of failure (HCLPF) value and an estimate of the composite variability, c (Methods 2, 3a, and 3b below). The USNRC used a c of 0.4 to develop the plant-level fragilities for the SMA plants.
Table 1 -Summary of USNRC GI-199 Methods for Estimating Plant Damage State Fragilities Bases for Establishing Plant-Level Fragility Curves Parameters From IPEEE Information Basis Source Parameters*
C50 and C determined by probability plot of the reported plant-level 1a SPRA fragility curve C50 found by matching the computed SCDF to the SCDF stated in the IPEEE for the specified hazard curve (EPRI, LLNL, or plant-1b SPRA specific).
Assumed C = 0.4.
C50 and C determined by matching computed SCDFs to IPEEE 1c SPRA SCDFs for a pair of hazard curves.
SMA C50 found by using the stated HCLPF 2
(HCLPF < RLE) Assumed C = 0.4.
C50 found by using the stated HCLPF/RLE SMA 3a Assumed C = 0.4 (HCLPF = RLE)
Note: The RLE is a lower bound on the actual HCLPF.
C50 found by using the stated HCLPF/RLE/SSE SMA Assumed C = 0.4 3b (HCLPF = RLE = SSE) Note: The SSE is a lower bound on the actual HCLPF; applies to reduced scope SMA plants.
th
- C50 is the median (50 percentile) plant-level acceleration capacity and c is the composite variability 3
These plant-level fragility values developed by the USNRC (USNRC, 2010) were used directly for the SCDF calculations in this current EPRI fleet risk assessment, which allows for a direct comparison of the SCDF estimates using the newly developed seismic hazards and the USNRCs SCDF estimates in 2010 using the 2008 USGS and 1994 LLNL seismic hazards. For convenience, the plant-level fragilities from the GI-199 Safety / Risk Assessment are reproduced in Table 2 below. As noted above, some SPRA plants have more than one plant-level fragility estimate (sensitivity studies were conducted in the NRC GI-199 Safety / Risk Assessment for those plants where adequate information was not submitted as part of the IPEEE process).
The columns to the right side of Table 2 summarize the sixty-one CEUS sites for which new site-specific seismic hazards have been calculated. For purposes of the SCDF calculations in this study, the following decisions are made relative to calculating a single SCDF for each of these sixty-one sites:
For sites with multiple units, the plant-level fragility that results in the highest SCDF estimate is conservatively selected (most sites with multiple units have the same plant-level fragilities defined due to similarity, but several sites had submitted different plant-level fragilities as part of their IPEEE efforts).
For sites where the USNRC defined multiple plant-level fragilities (due to uncertainty in the correct spectral ratios from the IPEEE submittals), the plant-level fragility that results in the highest resulting SCDF value is conservatively selected.
These sixty-one plant level bounding fragilities are documented on the right half of Table 2.
4
Table 2 - Plant-Level Fragilities from USNRC Safety / Risk Assessment for GI-199 (USNRC, 2010)
Bounding Case
- Plant-Level Fragility from Appendix C of 2010 USNRC GI-199 Safety Report - Safety/Risk Assessment Plant Level Fragility from Appendix C of 2010 CEUS Site with New Appendix C Plant Data USNRC Safety/Risk Assessment Hazard Estimates Point PGA Fragility ** Spectral Ratios PGA Fragility ** Spectral Ratios C50 (g) C 10 Hz 5 Hz 1 Hz C50 (g) C 10 Hz 5 Hz 1 Hz Arkansas Nuclear One 1 0.76 0.4 1.87 2.12 0.96 Arkansas Nuclear 0.76 0.4 1.87 2.12 0.96 Arkansas Nuclear One 2 0.76 0.4 1.87 2.12 0.96 Beaver Valley 1 0.36 0.26 1.71 1.54 0.68 Beaver Valley 0.36 0.26 1.71 1.54 0.68 Beaver Valley 2 0.53 0.34 1.71 1.54 0.68 Braidwood 1 0.76 0.4 1.87 2.12 0.96 Braidwood 0.76 0.4 1.87 2.12 0.96 Braidwood 2 0.76 0.4 1.87 2.12 0.96 Browns Ferry 1 0.76 0.4 1.87 2.12 0.96 Browns Ferry 2 0.66 0.4 1.87 2.12 0.96 Browns Ferry 0.66 0.4 1.87 2.12 0.96 Browns Ferry 3 0.66 0.4 1.87 2.12 0.96 Brunswick 1 0.76 0.4 1.85 2.12 1.32 Brunswick 0.76 0.4 1.85 2.12 1.32 Brunswick 2 0.76 0.4 1.85 2.12 1.32 Byron 1 0.76 0.4 1.87 2.12 0.96 Byron 0.76 0.4 1.87 2.12 0.96 Byron 2 0.76 0.4 1.87 2.12 0.96 Callaway 0.76 0.4 1.85 2.12 1.32 Callaway 0.76 0.4 1.85 2.12 1.32 Calvert Cliffs 1 0.62 0.4 1.38 1.72 0.6 Calvert Cliffs 0.58 0.4 1.38 1.72 0.6 Calvert Cliffs 2 0.58 0.4 1.38 1.72 0.6 Catawba 1 0.44 0.63 1.87 2.12 0.96 Catawba 0.44 0.63 1.87 2.12 0.96 Catawba 2 0.44 0.63 1.87 2.12 0.96 Clinton (0098) 0.76 0.4 1.85 2.12 1.32 Clinton 0.76 0.4 1.67 1.81 0.59 Clinton(UHS) 0.76 0.4 1.67 1.81 0.59 Comanche Peak 1 0.30 0.4 2.26 2.56 1.28 Comanche Peak 0.3 0.4 2.26 2.56 1.28 Comanche Peak 2 0.30 0.4 2.26 2.56 1.28 Cooper 0.76 0.4 1.85 2.12 1.32 Cooper 0.76 0.4 1.85 2.12 1.32 Crystal River 3 0.25 0.4 1.22 1.51 1.58 Crystal River 0.25 0.4 1.22 1.51 1.58 D.C. Cook 1 0.48 0.27 2.27 2.13 0.65 D.C. Cook 0.48 0.27 2.27 2.13 0.65 D.C. Cook 2 0.48 0.27 2.27 2.13 0.65 Davis-Besse 0.66 0.4 1.87 2.12 0.96 Davis-Besse 0.66 0.4 1.87 2.12 0.96 Dresden 2 0.51 0.4 1.87 2.12 0.96 Dresden 0.51 0.4 1.87 2.12 0.96 Dresden 3 0.51 0.4 1.87 2.12 0.96 Duane Arnold 0.30 0.4 1.85 2.68 1.07 Duane Arnold 0.3 0.4 1.85 2.68 1.07 5
Bounding Case
- Plant-Level Fragility from Appendix C of 2010 USNRC GI-199 Safety Report - Safety/Risk Assessment Plant Level Fragility from Appendix C of 2010 CEUS Site with New Appendix C Plant Data USNRC Safety/Risk Assessment Hazard Estimates Point PGA Fragility ** Spectral Ratios PGA Fragility ** Spectral Ratios C50 (g) C 10 Hz 5 Hz 1 Hz C50 (g) C 10 Hz 5 Hz 1 Hz Farley 1 (1st spectral ratios) 0.25 0.4 1.87 2.12 0.96 Farley 1 (2nd spectral ratios) 0.25 0.4 1.85 2.12 1.32 Farley 0.25 0.4 1.87 2.12 0.96 Farley 2 (1st spectral ratios) 0.25 0.4 1.87 2.12 0.96 Farley 2 (2nd spectral ratios) 0.25 0.4 1.85 2.12 1.32 Fermi 2 0.76 0.4 1.87 2.12 0.96 Fermi 0.76 0.4 1.87 2.12 0.96 FitzPatrick 0.56 0.4 1.87 2.12 0.96 FitzPatrick 0.56 0.4 1.87 2.12 0.96 Fort Calhoun 0.63 0.4 1.85 2.12 1.32 Fort Calhoun 0.63 0.4 1.85 2.12 1.32 Ginna 0.51 0.4 2.14 2.42 1.36 Ginna 0.51 0.4 2.14 2.42 1.36 Grand Gulf 1 0.38 0.4 1.92 2.65 1.33 Grand Gulf 0.38 0.4 1.92 2.65 1.33 Harris 1 0.74 0.4 1.87 2.12 0.96 Harris 0.74 0.4 1.87 2.12 0.96 Hatch 1 0.76 0.4 1.85 2.12 1.32 Hatch 0.76 0.4 1.85 2.12 1.32 Hatch 2 0.76 0.4 1.85 2.12 1.32 Hope Creek 1 1.66 0.7 1.97 2.27 0.98 Hope Creek 1.66 0.7 1.97 2.27 0.98 Indian Point 2 0.68 0.4 1.62 1.23 0.41 Indian Point 0.34 0.34 1.56 1.61 0.81 Indian Point 3 0.34 0.34 1.56 1.61 0.81 Kewaunee 0.41 0.22 1.8 1.79 0.4 Kewaunee 0.41 0.22 1.8 1.79 0.4 La Salle 1 (0098) 1.32 0.4 1.85 2.12 1.32 La Salle 1 (SSE) 1.32 0.4 1.85 2.62 1.31 La Salle 1 (UHS) 1.32 0.4 1.67 1.83 0.923 La Salle 1.32 0.4 1.67 1.83 0.923 La Salle 2 (0098) 1.32 0.4 1.85 2.12 1.32 La Salle 2 (SSE) 1.32 0.4 1.85 2.62 1.31 La Salle 2 (UHS) 1.32 0.4 1.67 1.83 0.923 Limerick 1 0.38 0.4 2.59 2.47 1.18 Limerick 0.38 0.4 2.59 2.47 1.18 Limerick 2 0.38 0.4 2.59 2.47 1.18 McGuire 1 0.45 0.74 1.88 2.35 1.19 McGuire 0.45 0.74 1.88 2.35 1.19 McGuire 2 0.45 0.74 1.88 2.35 1.19 Millstone 2 0.63 0.4 1.87 2.12 0.96 Millstone 0.54 0.4 2.27 2.27 1.26 Millstone 3 0.54 0.4 2.27 2.27 1.26 Monticello 0.30 0.4 2.29 2.69 1.12 Monticello 0.3 0.4 2.29 2.69 1.12 Nine Mile Point 1 0.68 0.4 1.87 2.12 0.96 Nine Mile Point 0.58 0.4 1.87 2.12 0.96 Nine Mile Point 2 0.58 0.4 1.87 2.12 0.96 6
Bounding Case
- Plant-Level Fragility from Appendix C of 2010 USNRC GI-199 Safety Report - Safety/Risk Assessment Plant Level Fragility from Appendix C of 2010 CEUS Site with New Appendix C Plant Data USNRC Safety/Risk Assessment Hazard Estimates Point PGA Fragility ** Spectral Ratios PGA Fragility ** Spectral Ratios C50 (g) C 10 Hz 5 Hz 1 Hz C50 (g) C 10 Hz 5 Hz 1 Hz North Anna 1 (1st spectral 0.41 0.4 1.87 2.12 0.96 ratios)
North Anna 1 (2nd spectral 0.41 0.4 1.85 2.12 1.32 ratios)
North Anna 0.41 0.4 1.85 2.12 1.32 North Anna 2 (1st spectral 0.41 0.4 1.87 2.12 0.96 ratios)
North Anna 2 (2nd spectral 0.41 0.4 1.85 2.12 1.32 ratios)
Oconee 1 0.62 0.32 1.66 1.32 0.35 Oconee 2 0.62 0.32 1.66 1.32 0.35 Oconee 0.62 0.32 1.66 1.32 0.35 Oconee 3 0.62 0.32 1.66 1.32 0.35 Oyster Creek 0.57 0.36 2 1.78 0.796 Oyster Creek 0.57 0.36 2 1.78 0.796 Palisades 0.49 0.35 2.13 2.44 0.74 Palisades 0.49 0.35 2.13 2.44 0.74 Peach Bottom 2 0.51 0.4 1.87 2.12 0.96 Peach Bottom 0.51 0.4 1.87 2.12 0.96 Peach Bottom 3 0.51 0.4 1.87 2.12 0.96 Perry 1 0.76 0.4 1.87 2.12 0.96 Perry 0.76 0.4 1.87 2.12 0.96 Pilgrim 1 0.49 0.27 1.55 1.66 0.5 Pilgrim 0.49 0.27 1.55 1.66 0.5 Point Beach 1 0.45 0.45 1.78 1.75 0.675 Point Beach 0.45 0.45 1.78 1.75 0.675 Point Beach 2 0.45 0.45 1.78 1.75 0.675 Prairie Island 1 0.71 0.4 1.85 2.12 1.32 Prairie Island 0.71 0.4 1.85 2.12 1.32 Prairie Island 2 0.71 0.4 1.85 2.12 1.32 Quad Cities 1 0.23 0.4 1.87 2.12 0.96 Quad Cities 0.23 0.4 1.87 2.12 0.96 Quad Cities 2 0.23 0.4 1.87 2.12 0.96 River Bend 1 0.25 0.4 2.35 2.75 1.41 River Bend 0.25 0.4 2.35 2.75 1.41 Robinson 2 0.71 0.4 1.85 2.12 1.32 Robinson 0.71 0.4 1.85 2.12 1.32 Saint Lucie 1 (s4) 0.25 0.4 1.18 1.5 0.8 Saint Lucie 1 (s5) 0.25 0.4 1.18 1.5 0.8 Saint Lucie 0.25 0.4 1.18 1.5 0.8 Saint Lucie 2 (s4) 0.25 0.4 1.18 1.5 0.8 Saint Lucie 2 (s5) 0.25 0.4 1.18 1.5 0.8 Salem 1 1.31 0.84 1.97 2.27 0.68 Salem 1.31 0.84 1.97 2.27 0.68 Salem 2 1.31 0.84 1.97 2.27 0.68 Seabrook 1 0.90 0.52 2.223 2.42 1.36 Seabrook 0.9 0.52 2.223 2.42 1.36 Sequoyah 1 0.68 0.4 1.87 2.12 0.96 Sequoyah 0.68 0.4 1.87 2.12 0.96 7
Bounding Case
- Plant-Level Fragility from Appendix C of 2010 USNRC GI-199 Safety Report - Safety/Risk Assessment Plant Level Fragility from Appendix C of 2010 CEUS Site with New Appendix C Plant Data USNRC Safety/Risk Assessment Hazard Estimates Point PGA Fragility ** Spectral Ratios PGA Fragility ** Spectral Ratios C50 (g) C 10 Hz 5 Hz 1 Hz C50 (g) C 10 Hz 5 Hz 1 Hz Sequoyah 2 0.68 0.4 1.87 2.12 0.96 South Texas 1 0.38 0.59 2.47 2.97 1.53 South Texas 0.38 0.59 2.47 2.97 1.53 South Texas 2 0.38 0.59 2.47 2.97 1.53 Summer 0.56 0.4 1.87 2.12 0.96 Summer 0.56 0.4 1.87 2.12 0.96 Surry 1 0.74 0.66 2.08 1.95 0.97 Surry 0.74 0.66 2.08 1.95 0.97 Surry 2 0.74 0.66 2.08 1.95 0.97 Susquehanna 1 0.53 0.4 1.87 2.12 0.96 Susquehanna 0.53 0.4 1.87 2.12 0.96 Susquehanna 2 0.53 0.4 1.87 2.12 0.96 Three Mile Island 1 0.29 0.28 2.73 2.6 1.127 Three Mile Island 0.29 0.28 2.73 2.6 1.127 Turkey Point 3 0.38 0.4 1.26 1.58 0.85 Turkey Point 0.38 0.4 1.26 1.58 0.85 Turkey Point 4 0.38 0.4 1.26 1.58 0.85 Vermont Yankee 0.63 0.4 1.87 2.12 0.96 Vermont Yankee 0.63 0.4 1.87 2.12 0.96 Vogtle 1 0.76 0.4 1.85 2.12 1.32 Vogtle 0.76 0.4 1.85 2.12 1.32 Vogtle 2 0.76 0.4 1.85 2.12 1.32 Waterford 3 0.25 0.4 1.72 2.4 1.19 Waterford 0.25 0.4 1.72 2.4 1.19 Watts Bar 1 (rock) 0.76 0.4 1.87 2.12 0.96 Watts Bar 0.76 0.4 1.87 2.12 0.96 Watts Bar 1 (soil) 0.76 0.4 1.85 2.12 1.32 Wolf Creek 1 0.51 0.4 1.83 2.25 0.32 Wolf Creek 0.51 0.4 1.83 2.25 0.32
- Plant level fragility that results in the maximum SCDF for the site when combined with the newly developed site-specific seismic hazard (2013/2014) th
- C50 is the median (50 percentile) plant-level acceleration capacity and c is the composite variability 8
6 QUANTIFICATION APPROACH The USNRC used approximate methods to estimate the SCDF for each operating nuclear plant as part of their 2010 study to assess the safety implications of changing seismic hazards as part of GI-199. These approximate SCDF estimates were developed using a method that involved integrating the mean seismic hazard curve and an approximation of the mean plant-level fragility curve for each plant. This approximate method was first developed by Kennedy (Kennedy, 1999) and is discussed in Section 10-B.9 of the ASME/ANS RA-Sa-2009 Standard (ASME, 2009), as well as Appendix D of the SPID (EPRI, 2013a). This same approach is judged to be the most appropriate method to assess this latest set of new site-specific seismic hazard estimates developed in accordance with the USNRCs 50.54(f) letter.
In the NRC Safety/Risk Assessment of GI-199, SCDF estimates were computed at four spectral frequencies: 10 Hz, 5 Hz, 1 Hz and the peak ground acceleration (PGA). The terminology defined within the GI-199 Safety/Risk Assessment included the concept of a derived SCDF estimate which consisted of an estimate of the seismic core-damage frequency that was developed from these four spectral SCDF estimates:
SCDFpga = SCDF estimate obtained by using the PGA-based seismic hazard and plant-level fragility curves SCDF10 = SCDF estimate obtained by using the 10 Hz seismic hazard and plant-level fragility curves SCDF5 = SCDF estimate obtained by using the 5 Hz seismic hazard and plant-level fragility curves SCDF1 = SCDF estimate obtained by using the 1 Hz seismic hazard and plant-level fragility curves The seismic core damage frequency for a plant can most accurately be generated by incorporating each individual seismic fragility function into the complete plant logic model and convolving with the hazard to develop the SCDF. However, since the plant logic model was not typically included as part of the IPEEE submittal, this approximate approach is the best alternative to estimating these SCDFs. Past SPRAs have demonstrated that the actual plant risk is a function of the seismic response at a variety of spectral frequencies. The plant risk is very site specific and is a function of:
Failure modes governing the lower capacity structures, systems and components Soil frequencies for those structures founded on soil columns Structure fundamental frequencies Equipment fundamental frequencies The frequency ranges that drive the plant seismic risk are typically very broad, including contributions from 1 Hz to PGA. One of the methods to account for the spectral frequency contribution to the SCDF used in the GI-199 Safety / Risk Assessment considered each of the four frequencies (1, 5, 10 Hz and PGA) to contribute equally to the overall SCDF. The resulting derived SCDF estimate associated with this spectral weighting is shown mathematically in the equation below:
9
This averaging of the four frequencies approach is judged to be appropriate for this study as past SPRAs have demonstrated that typically there are risk contributions from all these frequencies due to the variety of equipment, systems and structures that end up contributing to the risk. In addition, EPRI has conducted some limited additional sensitivity studies related to this frequency weighting (expanding the number of frequencies from 4 to 6 and also considering an alternate approach in the GI-199 Safety / Risk Assessment referred to as the IPEEE weighted average SCDF approach) and the overall results and conclusions are relatively insensitive to the approach taken. EPRI does not recommend using any very conservative approaches to estimate the SCDF such as use of the maximum SCDFs calculated at any one frequency. This type of bounding approach is overly conservative and judged to not provide realistic risk estimates consistent with SCDFs calculated in actual SPRAs.
7 SCDF RESULTS To provide an initial assessment of the safety implications of the new seismic hazard estimates across the fleet of CEUS operating plants, point estimates of the mean SCDF are developed using the new site-specific seismic hazard curves. These are compared with the baseline SCDFs developed by the USNRC in 2010 using the 2008 USGS and 1994 LLNL seismic hazard curves.
Figure 1 provides a comparison of the cumulative fleet SCDF distribution calculated using the new site-specific seismic hazards, the 1994 LLNL seismic hazards, and the 2008 USGS seismic hazards. The SCDF values computed using the new hazard range from approximately 4E-7/year to 6E-5/year. The comparison shows that the overall distribution of SCDFs for the fleet has not changed significantly due to the new site-specific seismic hazards.
10
Figure 1 - Comparison of CEUS NPP Site Cumulative Distribution of Seismic CDFs 8 CONCLUSIONS In 2010, the USNRC conducted a Safety / Risk Assessment for the GI-199 program and developed simplified methods to calculate a point estimate of the SCDF. The USNRC developed an estimate of the seismic hazard at that time using the 2008 USGS seismic source to develop a new rock hazard, and EPRI site amplification factors. This 2008 hazard, along with the previously developed 1994 LLNL hazard, was then used to estimate the SCDFs for the fleet of U.S. plants using the plant-level fragilities estimated from each plants IPEEE submittals.
The USNRC concluded in 2010 that the overall SCDF estimates are indicative of performance consistent with the Commissions Safety Goal Policy Statement because they are within the subsidiary objective of 1E-4/year. The specific USNRC statement from the GI-199 Safety / Risk Assessment (USNRC, 2010) was:
Overall seismic core damage risk estimates are consistent with the Commissions Safety Goal Policy Statement because they are within the subsidiary objective of 10-4/year for core damage frequency. The GI-199 Safety / Risk Assessment, based in part on information from the U.S. Nuclear Regulatory Commissions (NRCs) Individual Plant Examination of External Events (IPEEE) program, indicates that no concern exists regarding adequate protection and that the current seismic design of operating reactors 11
provides a safety margin to withstand potential earthquakes exceeding the original design basis.
New seismic hazard analyses have been completed for all sixty-one CEUS nuclear power plant sites. EPRI calculated the approximate SCDFs for each of these sites using methods that the USNRC used to assess changing seismic hazard in the past. As can be seen from Figure 1 above, the overall distribution of SCDFs for the fleet indicates that the impact of the updated seismic hazard has been to reduce risk for most plants relative to estimates obtained using either the 2008 USGS or the 1994 LLNL hazard assessments.
The range of SCDFs still falls between 1E-7/year and 1E-4/year.
For individual plants, some plant SCDF estimates have increased, but the vast majority have decreased somewhat.
In the case of the sites for which increases were seen, none of the SCDF values approaches 1E-4/year.
Comparisons of the SCDF estimates developed in 2010 by the USNRC to the SCDF estimates developed by EPRI for the new site-specific seismic hazards show that there clearly has not been an overall increase in seismic risk for the fleet of U.S. plants. In addition, all sixty-one of the CEUS sites have SCDF estimates below the 1E-4/year threshold considered in the USNRC 2010 Safety / Risk Assessment. Thus it can be concluded that the current seismic design of operating reactors continues to provide a safety margin to withstand potential earthquakes exceeding the seismic design basis, as was concluded in the USNRC 2010 Safety / Risk Assessment.
12
9 REFERENCES ASME (2009). Standard for Level 1/Large Early Release Frequency Probabilistic Risk Assessment for Nuclear Power Plant Applications. American Society of Mechanical Engineers and American Nuclear Society Standard ASME/ANS RA-Sb-2009 (Addenda to ASME/ANS RA-S-2008).
ASME (2013). Standard for Level 1/Large Early Release Frequency Probabilistic Risk Assessment for Nuclear Power Plant Applications. American Society of Mechanical Engineers and American Nuclear Society Standard ASME/ANS RA-Sb-2013 (Addenda to ASME/ANS RA-S-2008), September 2013.
CEUS-SSC (2012). Central and Eastern United States Seismic Source Characterization for Nuclear Facilities, U.S. Nuclear Regulatory Commission Report, NUREG-2115; Electric Power Research Institute Report 1021097, 6 Volumes; DOE Report# DOE/NE-0140.
EPRI (2013a). Seismic Evaluation Guidance: Screening, Prioritization and Implementation Details (SPID) for the Resolution of Fukushima Near-Term Task Force Recommendation 2.1: Seismic, Electric Power Research Institute Report 1025287, February 2013.
EPRI (2013b). Seismic Evaluation Guidance: Augmented Approach for the Resolution of Fukushima Near-Term Task Force Recommendation 2.1 - Seismic, Electric Power Research Institute Report 3002000704, May 2013.
EPRI (2013c). EPRI (2004, 2006) Ground-Motion Model (GMM) Review Project, 2 volumes, Electric Power Research Institute Report 3002000717, June, 2013.
NEI (2013). Proposed Path Forward for NTTF Recommendation 2.1: Seismic Reevaluations, A. Pietrangelo Letter to D. Skeen, Nuclear Energy Institute, 9 April 2013.
Kennedy, R.P. (1999). Overview of Methods for Seismic PRA and Margins Including Recent Innovations, Proceedings of the Organization for the Economic Cooperation and Development/Nuclear Energy Agency Workshop on Seismic Risk, Tokyo, Japan, 10 - 12 August 1999.
USNRC (2007). A Performance-Based Approach to Define the Site-Specific Earthquake Ground Motion, U.S. Nuclear Regulatory Commission Reg. Guide 1.208, U.S. Nuclear Regulatory Commission, Washington, DC.
USNRC (2010). Implications of Updated Probabilistic Seismic Hazard Estimates In Central And Eastern United States On Existing Plants Generic Issue 199 (GI-199), Safety Risk Assessment, U.S. Nuclear Regulatory Commission, Washington, DC, Aug.
USNRC (2012). 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, E. Leeds and M. Johnson Letter to All Power Reactor Licensees et al., U.S. Nuclear Regulatory Commission, Washington, DC, 12 March.
13
USNRC (2013). Electric Power Research Institute Final Draft Report XXXXXX, Seismic Evaluation Guidance: Augmented Approach for the Resolution of Fukushima Near-Term Task Force Recommendation 2.1: Seismic, as an Acceptable Alternative to the March 12, 2012, Information Request for Seismic Reevaluations, Eric Leeds Letter to Joseph Pollock (NEI), 7 May 2013.
USNRC (2014). Supplemental Information Related to Request for Information Pursuant to Title 10 of the Code of Federal Regulations 50.54(f) Regarding Seismic Hazard Reevaluations for Recommendation 2.1 of the Near-Term Task Force Review of Insights from the Fukushima Dai-ichi Accident, E. Leeds Letter to All Power Reactor Licensees et al., U.S. Nuclear Regulatory Commission, Washington, DC 20 February 2014.
14