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MONTHYEAR1CAN051401, Time-Limited Aging Analysis Regarding Reactor Vessel Internals Loss of Ductility for Arkansas Nuclear One, Unit 1 at 60 Years Arkansas Nuclear One Unit 12014-05-0606 May 2014 Time-Limited Aging Analysis Regarding Reactor Vessel Internals Loss of Ductility for Arkansas Nuclear One, Unit 1 at 60 Years Arkansas Nuclear One Unit 1 Project stage: Request 1CAN071406, Unit 1, Revised Affidavit for the Time-Limited Aging Analysis Regarding Reactor Vessel Internals Loss of Ductility2014-07-30030 July 2014 Unit 1, Revised Affidavit for the Time-Limited Aging Analysis Regarding Reactor Vessel Internals Loss of Ductility Project stage: Request ML14169A5092014-08-0707 August 2014 Request for Withholding Information from Public Disclosure, 3/25/14 Affidavit Executed by G. Elliot, Areva Inc. Regarding Time-Limited Aging Analysis Project stage: Withholding Request Acceptance ML14349A0072014-12-12012 December 2014 Transmittal Email, Proprietary Request for Additional Information Email, Time-Limited Aging Analysis Regarding Reactor Vessel Internals Project stage: RAI ML15028A4932015-01-31031 January 2015 ANP-3281Q1NP, Rev. 0, Response to NRC Request for Additional Information on Time-Limited Aging Analysis for the ANO-1 Reactor Vessel Internals, Attacment 2 Project stage: Response to RAI ML15139A0022015-05-20020 May 2015 Request for Withholding Information from Public Disclosure, 1/25/15 Affidavit Executed by G. Elliot, Areva Inc. ANP-3281Q1P, Revision 0, Response to NRC Request for Additional Information on Time-Limited Aging Analysis Project stage: RAI ML15139A1132015-05-28028 May 2015 Redacted, Approval of Time-Limited Aging Analysis Regarding Reactor Vessel Internals Project stage: Other 2014-08-07
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Category:Letter type:
MONTHYEAR0CAN102401, Response to Request for Additional Information - Arkansas Nuclear One – Units 1 and 2, Proposed Alternatives ANO1-ISI-24-01 and ANO2-ISI-24-01 for Examinations of Steam Generator Pressure-Retaining Welds and Full Penetration Welded Nozzle2024-10-16016 October 2024 Response to Request for Additional Information - Arkansas Nuclear One – Units 1 and 2, Proposed Alternatives ANO1-ISI-24-01 and ANO2-ISI-24-01 for Examinations of Steam Generator Pressure-Retaining Welds and Full Penetration Welded Nozzles 2CAN102401, Cycle 31 Core Operating Limits Report (COLR)2024-10-14014 October 2024 Cycle 31 Core Operating Limits Report (COLR) 1CAN082401, Inservice Inspection Summary Report for the Arkansas Nuclear One, Unit 1 Thirty-First Refueling Outage (1R312024-08-13013 August 2024 Inservice Inspection Summary Report for the Arkansas Nuclear One, Unit 1 Thirty-First Refueling Outage (1R31 2CAN072401, Proposed Alternative for Implementation of Extended Reactor Vessel In-Service Inspection Interval (ANO2-ISI-24-02)2024-07-19019 July 2024 Proposed Alternative for Implementation of Extended Reactor Vessel In-Service Inspection Interval (ANO2-ISI-24-02) 0CAN072401, Annual 10 CFR 50.46 Report for Calendar Year 2023 Emergency Core Cooling System Evaluation Changes2024-07-0808 July 2024 Annual 10 CFR 50.46 Report for Calendar Year 2023 Emergency Core Cooling System Evaluation Changes 1CAN072401, Request for Review and Approval of Changes to the Safety Analysis Report and to a Confirmatory Order Clarifying an Alternate Means of Compliance for Pressurizer Heaters Emergency Power Supply2024-07-0202 July 2024 Request for Review and Approval of Changes to the Safety Analysis Report and to a Confirmatory Order Clarifying an Alternate Means of Compliance for Pressurizer Heaters Emergency Power Supply 0CAN062403, Groundwater Protection Initiative - Voluntary Special Report for Tritium Levels2024-06-25025 June 2024 Groundwater Protection Initiative - Voluntary Special Report for Tritium Levels 0CAN062402, Proposed Alternatives ANO1-ISI-24-01 and ANO2-ISI-24-01 for Examinations of Steam Generator Pressure-Retaining Welds and Full Penetration Welded Nozzles2024-06-0606 June 2024 Proposed Alternatives ANO1-ISI-24-01 and ANO2-ISI-24-01 for Examinations of Steam Generator Pressure-Retaining Welds and Full Penetration Welded Nozzles 0CAN052401, – Units 1 and 2, Submittal of Annual Radiological Environmental Operating Report for 20232024-05-13013 May 2024 – Units 1 and 2, Submittal of Annual Radiological Environmental Operating Report for 2023 1CAN052401, Cycle 32 Core Operating Limits Report2024-05-0404 May 2024 Cycle 32 Core Operating Limits Report 2CAN042403, Response to the Request for Additional Information Regarding ANO-2 License Amendment Request to Revise Technical Specifications to Adopt Risk-Informed Completion Times TSTF-505, Revision 2, Provide Risk-Informed Extended C2024-04-24024 April 2024 Response to the Request for Additional Information Regarding ANO-2 License Amendment Request to Revise Technical Specifications to Adopt Risk-Informed Completion Times TSTF-505, Revision 2, Provide Risk-Informed Extended C 0CAN042402, Annual Occupational Radiation Exposure Report for 20232024-04-23023 April 2024 Annual Occupational Radiation Exposure Report for 2023 0CAN042401, Radioactive Effluent Release Report for the 2023 Calendar Year2024-04-15015 April 2024 Radioactive Effluent Release Report for the 2023 Calendar Year 2CAN042402, Special Report of Non-functional Main Steam Line Radiation Monitor2024-04-11011 April 2024 Special Report of Non-functional Main Steam Line Radiation Monitor 2CAN042401, Request to Revise Typographical Errors in Arkansas Nuclear One, Unit 2 Technical Specifications2024-04-0404 April 2024 Request to Revise Typographical Errors in Arkansas Nuclear One, Unit 2 Technical Specifications 2CAN012401, U.S. Additional Protocol2024-01-17017 January 2024 U.S. Additional Protocol 2CAN012403, Supplemental Information - Adopt Risk-Informed Completion Times TSTF-505, Revision 2, Provide Risk-Informed Extended Completion Times - RITSTF Initiative 42024-01-11011 January 2024 Supplemental Information - Adopt Risk-Informed Completion Times TSTF-505, Revision 2, Provide Risk-Informed Extended Completion Times - RITSTF Initiative 4 0CAN012401, Registration of Cask Use2024-01-10010 January 2024 Registration of Cask Use 1CAN122301, Responses to Request for Additional Information for Request for Alternative for Implementation of Extended Reactor Vessel Inservice Inspection Interval (ANO1-ISl-037)2023-12-14014 December 2023 Responses to Request for Additional Information for Request for Alternative for Implementation of Extended Reactor Vessel Inservice Inspection Interval (ANO1-ISl-037) 0CAN102303, Registration of Cask Use2023-10-24024 October 2023 Registration of Cask Use 0CAN102301, Evacuation Time Estimate (ETE) Study2023-10-0404 October 2023 Evacuation Time Estimate (ETE) Study 1CAN092301, Supplemental Information - Adopt Risk-Informed Completion Times TSTF-505, Revision 2, Provide Risk-Informed Extended Completion Times - RITSTF Initiative 42023-09-21021 September 2023 Supplemental Information - Adopt Risk-Informed Completion Times TSTF-505, Revision 2, Provide Risk-Informed Extended Completion Times - RITSTF Initiative 4 0CAN092302, Supplement to Request for Alternative Frequency to Containment Unbonded Post-Tensioning System Inservice Inspection (ANO-CISI-002)2023-09-14014 September 2023 Supplement to Request for Alternative Frequency to Containment Unbonded Post-Tensioning System Inservice Inspection (ANO-CISI-002) 2CAN092301, Reply to a Notice of Violation2023-09-0808 September 2023 Reply to a Notice of Violation 0CAN092301, Emergency Plan Implementing Procedure Revision2023-09-0505 September 2023 Emergency Plan Implementing Procedure Revision 0CAN082301, Units 1 and 2 - Changes to the Quality Assurance Program Approval Form for Radioactive Material Package No. 03412023-08-17017 August 2023 Units 1 and 2 - Changes to the Quality Assurance Program Approval Form for Radioactive Material Package No. 0341 2CAN082301, Inservice Inspection Summary Report for the Arkansas Nuclear One, Unit 2, Twenty-Ninth Refueling Outage (2R29)2023-08-10010 August 2023 Inservice Inspection Summary Report for the Arkansas Nuclear One, Unit 2, Twenty-Ninth Refueling Outage (2R29) 0CAN072301, Registration of Cask Use2023-07-18018 July 2023 Registration of Cask Use 1CAN062304, Supplement Related to License Amendment Request to Remove Technical Specification Condition Allowing Two Reactor Coolant Pump Operation2023-06-29029 June 2023 Supplement Related to License Amendment Request to Remove Technical Specification Condition Allowing Two Reactor Coolant Pump Operation 0CAN062301, Status of Actions to Return to Compliance2023-06-26026 June 2023 Status of Actions to Return to Compliance 0CAN062302, Submittal of Revision 22 of the ANO Fire Hazards Analysis2023-06-20020 June 2023 Submittal of Revision 22 of the ANO Fire Hazards Analysis 1CAN062301, Request for Alternative for Implementation of Extended Reactor Vessel Inservice Inspection Interval (ANO1-ISI-037)2023-06-0808 June 2023 Request for Alternative for Implementation of Extended Reactor Vessel Inservice Inspection Interval (ANO1-ISI-037) 0CAN052303, Annual 10 CFR 50.46 Report for Calendar Year 20222023-05-24024 May 2023 Annual 10 CFR 50.46 Report for Calendar Year 2022 0CAN052302, Emergency Plan Rev. 482023-05-11011 May 2023 Emergency Plan Rev. 48 0CAN052301, Units 1 and 2 - Annual Radiological Environmental Operating Report for 20222023-05-0909 May 2023 Units 1 and 2 - Annual Radiological Environmental Operating Report for 2022 2CAN052301, Cycle 30 Core Operating Limits Report (COLR)2023-05-0303 May 2023 Cycle 30 Core Operating Limits Report (COLR) 0CAN042302, Annual Occupational Radiation Exposure Report for 20222023-04-27027 April 2023 Annual Occupational Radiation Exposure Report for 2022 0CAN042301, Radioactive Effluent Release Report for 20222023-04-14014 April 2023 Radioactive Effluent Release Report for 2022 2CAN042301, 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 42023-04-0505 April 2023 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 1CAN032301, License Amendment Request to Modify the Arkansas Nuclear One, Unit 1, Technical Specification 3.3.1, Reactor Protection System Instrumentation, Turbine Trip Function on Low Control Oil Pressure2023-03-30030 March 2023 License Amendment Request to Modify the Arkansas Nuclear One, Unit 1, Technical Specification 3.3.1, Reactor Protection System Instrumentation, Turbine Trip Function on Low Control Oil Pressure 2CAN032303, Responses to Request for Additional Information Concerning the Request for Alternative Regarding the Half-Nozzle Repair of Reactor Vessel Closure Head Penetration 462023-03-29029 March 2023 Responses to Request for Additional Information Concerning the Request for Alternative Regarding the Half-Nozzle Repair of Reactor Vessel Closure Head Penetration 46 2CAN032304, Supplement to the Request for Alternative Regarding the Half-Nozzle Repair of Reactor Vessel Closure Head Penetration 462023-03-29029 March 2023 Supplement to the Request for Alternative Regarding the Half-Nozzle Repair of Reactor Vessel Closure Head Penetration 46 2CAN032305, 03 Post Examination Analysis2023-03-23023 March 2023 03 Post Examination Analysis 1CAN032302, Inspection Summary Report for the Thirtieth Refueling Outage (1R30)2023-03-20020 March 2023 Inspection Summary Report for the Thirtieth Refueling Outage (1R30) 1CAN012301, Responses to Request for Additional Information for Request for Relief ANO1-ISI-0352023-01-30030 January 2023 Responses to Request for Additional Information for Request for Relief ANO1-ISI-035 2CAN012303, U.S. Additional Protocol2023-01-23023 January 2023 U.S. Additional Protocol 1CAN122201, 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 42022-12-22022 December 2022 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 0CAN122202, Registration of Cask Use2022-12-21021 December 2022 Registration of Cask Use 0CAN122201, Reply to a Notice of Violation; EA-22-0992022-12-0808 December 2022 Reply to a Notice of Violation; EA-22-099 0CAN112201, Request for Alternative Frequency to Containment Unbonded Post-Tensioning System Inservice Inspection (ANO-CISI-002)2022-11-10010 November 2022 Request for Alternative Frequency to Containment Unbonded Post-Tensioning System Inservice Inspection (ANO-CISI-002) 2024-08-13
[Table view] Category:Report
MONTHYEARML24295A1232024-10-21021 October 2024 Enclosure 3: Relief Request ANO2-RR-24-001, Revision 0 (Non-Proprietary) 1CAN072401, Request for Review and Approval of Changes to the Safety Analysis Report and to a Confirmatory Order Clarifying an Alternate Means of Compliance for Pressurizer Heaters Emergency Power Supply2024-07-0202 July 2024 Request for Review and Approval of Changes to the Safety Analysis Report and to a Confirmatory Order Clarifying an Alternate Means of Compliance for Pressurizer Heaters Emergency Power Supply 1CAN062304, Supplement Related to License Amendment Request to Remove Technical Specification Condition Allowing Two Reactor Coolant Pump Operation2023-06-29029 June 2023 Supplement Related to License Amendment Request to Remove Technical Specification Condition Allowing Two Reactor Coolant Pump Operation 1CAN062302, Enclosure 2: ANO-1 SAR Amendment 31 - Redacted Version2023-06-20020 June 2023 Enclosure 2: ANO-1 SAR Amendment 31 - Redacted Version ML23180A1082023-06-20020 June 2023 ANO Unit 1 SAR Amendment 31, TRM, TS Bases, 10 CFR 50.59 Report, and Commitment Change Summary Report ML23088A2172022-12-31031 December 2022 Relief Request for Half-Nozzle Repair of Reactor Vessel Closure Head Penetration 46 - Technical Report, ANP-4023NP, Revision 0, December 2022 2CAN022202, Requests for Relief from American Society of Mechanical Engineers Section XI Volumetric Examination Requirements - Fourth 10-Year Interval, Second and Third Periods2022-02-24024 February 2022 Requests for Relief from American Society of Mechanical Engineers Section XI Volumetric Examination Requirements - Fourth 10-Year Interval, Second and Third Periods CNRO-2021-00023, Entergy Operations, Inc. - Supplement to CNRO-2021-00002, Basis for Concluding the Terms of Confirmatory Order EA-17-132/EA-17-153 Are Complete, Element L2021-10-0606 October 2021 Entergy Operations, Inc. - Supplement to CNRO-2021-00002, Basis for Concluding the Terms of Confirmatory Order EA-17-132/EA-17-153 Are Complete, Element L 0CAN102102, Units 1 and 210 CFR 50.71(e) Report Revision 20 of the ANO Fire Hazards Analysis2021-10-0606 October 2021 Units 1 and 210 CFR 50.71(e) Report Revision 20 of the ANO Fire Hazards Analysis ML21272A3032021-09-30030 September 2021 Technology Inclusive Content of Application Project (Ticap) for Non-Light Water Reactors Westinghouse Evinci; Micro-Reactor Tabletop Exercise Report ML21237A0512021-08-25025 August 2021 Follow-on Risk Informed Performance Based Implementation Guidance Needed for Advanced Non-Light Water Reactors ML21081A1922021-06-30030 June 2021 Enclosure - USNRC-CNSC Joint Report Concerning X-Energy's Reactor Pressure Vessel Construction Code Assessment 2CAN062103, Request for Alternative ANO2-PT-003 End-of-Interval System Leakage Test for Extended Reactor Coolant Pressure Boundary Piping - Fifth Interval2021-06-29029 June 2021 Request for Alternative ANO2-PT-003 End-of-Interval System Leakage Test for Extended Reactor Coolant Pressure Boundary Piping - Fifth Interval 0CAN052102, Annual 10 CFR 50.46 Report for Calendar Year 2020 Emergency Core Cooling System Evaluation Changes2021-05-10010 May 2021 Annual 10 CFR 50.46 Report for Calendar Year 2020 Emergency Core Cooling System Evaluation Changes ML21272A3382021-04-0101 April 2021 Technology Inclusive Content of Application Project (Ticap) for Non-Light Water Reactors Versatile Test Reactor Ticap Tabletop Exercise Report ML21090A0332021-03-31031 March 2021 Historical Context and Perspective on Allowable Stresses and Design Parameters in ASME Section III, Division 5, Subsection Hb, Subpart B (ANL/AMD-21/1) ML21083A1362021-03-23023 March 2021 Completed Activities ML21083A1372021-03-22022 March 2021 NEIMA Reporting ML21083A1382021-03-22022 March 2021 Rulemaking ML21083A1392021-03-22022 March 2021 Strategy 1 ML21083A1402021-03-22022 March 2021 Strategy 2 ML21083A1412021-03-22022 March 2021 Strategy 3 ML21083A1422021-03-22022 March 2021 Strategy 4 ML21083A1432021-03-22022 March 2021 Strategy 5 ML21083A1442021-03-22022 March 2021 Strategy 6 ML21014A2672021-01-14014 January 2021 Preapplication Engagement to Optimize Application Reviews January 12 Version - Copy 1CAN032001, Supplemental Information Related to License Amendment Request to Revise Loss of Voltage Relay Allowable Values2020-03-19019 March 2020 Supplemental Information Related to License Amendment Request to Revise Loss of Voltage Relay Allowable Values 0CAN121901, Summary of Lost Specimens Investigation Report2019-12-0303 December 2019 Summary of Lost Specimens Investigation Report ML18215A1782018-06-30030 June 2018 WCAP-18169-NP, Rev 1, Arkansas Nuclear One Unit 2 Heatup and Cooldown Limit Curves for Normal Operation. ML17214A0292018-02-12012 February 2018 Staff Assessment of Flooding Focused Evaluation (CAC Nos. MF9809 and MF9810) ML17291A0092017-10-26026 October 2017 Staff Assessment Regarding Program Plan for Aging Management for Reactor Vessel Internals (CAC No. MF8155; EPID L-2016-LRO-0001) ML17236A1792017-08-22022 August 2017 Arkansas, Units 1 and 2, ANO Emergency Plan On-Shift Staffing Analysis Report, Revision 2 0CAN081703, Document 51-9257562-001, Revision 1, Arkansas Nuclear One Hfe - High Frequency Confirmation Report2017-08-16016 August 2017 Document 51-9257562-001, Revision 1, Arkansas Nuclear One Hfe - High Frequency Confirmation Report ML17167A0832017-06-28028 June 2017 Arkansas Nuclear One, Unit 2 - Review of Commitment Submittal for License Renewal Regarding Nickel-Based Alloy Aging Management Program Plan (CAC No. MF8154) 0CAN061701, Transmittal of 10 CFR 50.71(e) Report, Revision 17 of the ANO Fire Hazards Analysis2017-06-0707 June 2017 Transmittal of 10 CFR 50.71(e) Report, Revision 17 of the ANO Fire Hazards Analysis 0CAN051704, Engineering Report No. CALC-ANOC-CS-14-00017, Rev 0, 2017 Focused Evaluation for External Flooding at Arkansas Nuclear One.2017-03-13013 March 2017 Engineering Report No. CALC-ANOC-CS-14-00017, Rev 0, 2017 Focused Evaluation for External Flooding at Arkansas Nuclear One. 2CAN011703, Submittal of Additional Protocol Report2017-01-26026 January 2017 Submittal of Additional Protocol Report ML17024A0362016-12-31031 December 2016 Operating Data Report for 2016 0CAN121602, Mitigating Strategies Assessment (MSA) Report for the New Seismic Hazard Information Per Nuclear Energy Institute (NEI) 12-06, Appendix H, Revision 2, H.4.3 Path 32016-12-30030 December 2016 Mitigating Strategies Assessment (MSA) Report for the New Seismic Hazard Information Per Nuclear Energy Institute (NEI) 12-06, Appendix H, Revision 2, H.4.3 Path 3 ML17003A2902016-12-20020 December 2016 Areva, Inc. - Engineering Information Record - Arkansas Nuclear One HFE-High Frequency Confirmation Report ML16365A0272016-10-31031 October 2016 ANP-3486NP, Revision 0, MRP-227-A Applicant/Licensee Action Item 6 Analysis for Arkansas Nuclear One Unit 1 (ANO-1). ML16293A5842016-09-30030 September 2016 WCAP-18166-NP, Revision 0, Analysis of Capsule 284 from the Entergy Operations, Inc. Arkansas Nuclear One, Unit 2 Reactor Vessel Radiation Surveillance Program. 1CAN091601, Submittal of Initial Examination Completion of Post-Examination Analysis2016-09-0101 September 2016 Submittal of Initial Examination Completion of Post-Examination Analysis ML16202A1672016-07-0505 July 2016 Report 1500227.401, PWR Internals Aging Management Program Plan. ML16147A3242016-05-31031 May 2016 ANP-3417NP, Rev. 1, MRP-227-A Applicant/Licensee Action Item #7 Analysis for Arkansas Nuclear One, Unit 1. ML16004A1792015-12-31031 December 2015 Attachment 2, ANP-3418NP, Revision 0, Arkansas Nuclear One Unit 1 Reactor Vessel Internals License Renewal Scope and MRP-189, Revision 1 Comparison (MRP-227-A Action Item 2) Licensing Report. (Non-Proprietary) ML15278A0242015-09-28028 September 2015 Attachment 2, Areva Document ANP-3417NP, Revision 0, MRP-227-A Applicant / Licensee Action Item No. 7 Analysis for Arkansas Nuclear One, Unit 1 (Non-Proprietary), Attachment 3, Affidavit, and Attachment 4, List of Commitments ML15099A1522015-04-16016 April 2015 Review of Spring 2014 Steam Generator Tube Inspection Report, Inspection During Refueling Outage 2R23 ML15071A0552015-03-31031 March 2015 ANP-3300Q2NP, Revision 0, Response to Request for Additional Information on Reactor Coolant System Pressure/Temperature and Low Temperature Overpressure Protection System Limits to 54 EFPY for Arkansas Nuclear One, Unit 1. ML15086A0242015-03-25025 March 2015 ANP-3300Q3NP, Revision 0 to Response to Request for Additional Information on Reactor Coolant System Pressure/Temperature and Low Temperature Overpressure Protection System Limits to 54 EFPY for Arkansas Nuclear One, Unit 1 (Non-Proprietary 2024-07-02
[Table view] Category:Technical
MONTHYEARML24295A1232024-10-21021 October 2024 Enclosure 3: Relief Request ANO2-RR-24-001, Revision 0 (Non-Proprietary) 1CAN072401, Request for Review and Approval of Changes to the Safety Analysis Report and to a Confirmatory Order Clarifying an Alternate Means of Compliance for Pressurizer Heaters Emergency Power Supply2024-07-0202 July 2024 Request for Review and Approval of Changes to the Safety Analysis Report and to a Confirmatory Order Clarifying an Alternate Means of Compliance for Pressurizer Heaters Emergency Power Supply 1CAN062304, Supplement Related to License Amendment Request to Remove Technical Specification Condition Allowing Two Reactor Coolant Pump Operation2023-06-29029 June 2023 Supplement Related to License Amendment Request to Remove Technical Specification Condition Allowing Two Reactor Coolant Pump Operation ML23180A1082023-06-20020 June 2023 ANO Unit 1 SAR Amendment 31, TRM, TS Bases, 10 CFR 50.59 Report, and Commitment Change Summary Report 1CAN062302, Enclosure 2: ANO-1 SAR Amendment 31 - Redacted Version2023-06-20020 June 2023 Enclosure 2: ANO-1 SAR Amendment 31 - Redacted Version ML23088A2172022-12-31031 December 2022 Relief Request for Half-Nozzle Repair of Reactor Vessel Closure Head Penetration 46 - Technical Report, ANP-4023NP, Revision 0, December 2022 2CAN022202, Requests for Relief from American Society of Mechanical Engineers Section XI Volumetric Examination Requirements - Fourth 10-Year Interval, Second and Third Periods2022-02-24024 February 2022 Requests for Relief from American Society of Mechanical Engineers Section XI Volumetric Examination Requirements - Fourth 10-Year Interval, Second and Third Periods 0CAN102102, Units 1 and 210 CFR 50.71(e) Report Revision 20 of the ANO Fire Hazards Analysis2021-10-0606 October 2021 Units 1 and 210 CFR 50.71(e) Report Revision 20 of the ANO Fire Hazards Analysis ML21272A3032021-09-30030 September 2021 Technology Inclusive Content of Application Project (Ticap) for Non-Light Water Reactors Westinghouse Evinci; Micro-Reactor Tabletop Exercise Report ML21237A0512021-08-25025 August 2021 Follow-on Risk Informed Performance Based Implementation Guidance Needed for Advanced Non-Light Water Reactors ML21081A1922021-06-30030 June 2021 Enclosure - USNRC-CNSC Joint Report Concerning X-Energy's Reactor Pressure Vessel Construction Code Assessment 2CAN062103, Request for Alternative ANO2-PT-003 End-of-Interval System Leakage Test for Extended Reactor Coolant Pressure Boundary Piping - Fifth Interval2021-06-29029 June 2021 Request for Alternative ANO2-PT-003 End-of-Interval System Leakage Test for Extended Reactor Coolant Pressure Boundary Piping - Fifth Interval ML21272A3382021-04-0101 April 2021 Technology Inclusive Content of Application Project (Ticap) for Non-Light Water Reactors Versatile Test Reactor Ticap Tabletop Exercise Report ML21090A0332021-03-31031 March 2021 Historical Context and Perspective on Allowable Stresses and Design Parameters in ASME Section III, Division 5, Subsection Hb, Subpart B (ANL/AMD-21/1) ML18215A1782018-06-30030 June 2018 WCAP-18169-NP, Rev 1, Arkansas Nuclear One Unit 2 Heatup and Cooldown Limit Curves for Normal Operation. ML17236A1792017-08-22022 August 2017 Arkansas, Units 1 and 2, ANO Emergency Plan On-Shift Staffing Analysis Report, Revision 2 0CAN081703, Document 51-9257562-001, Revision 1, Arkansas Nuclear One Hfe - High Frequency Confirmation Report2017-08-16016 August 2017 Document 51-9257562-001, Revision 1, Arkansas Nuclear One Hfe - High Frequency Confirmation Report 0CAN061701, Transmittal of 10 CFR 50.71(e) Report, Revision 17 of the ANO Fire Hazards Analysis2017-06-0707 June 2017 Transmittal of 10 CFR 50.71(e) Report, Revision 17 of the ANO Fire Hazards Analysis 0CAN051704, Engineering Report No. CALC-ANOC-CS-14-00017, Rev 0, 2017 Focused Evaluation for External Flooding at Arkansas Nuclear One.2017-03-13013 March 2017 Engineering Report No. CALC-ANOC-CS-14-00017, Rev 0, 2017 Focused Evaluation for External Flooding at Arkansas Nuclear One. ML16365A0272016-10-31031 October 2016 ANP-3486NP, Revision 0, MRP-227-A Applicant/Licensee Action Item 6 Analysis for Arkansas Nuclear One Unit 1 (ANO-1). ML16293A5842016-09-30030 September 2016 WCAP-18166-NP, Revision 0, Analysis of Capsule 284 from the Entergy Operations, Inc. Arkansas Nuclear One, Unit 2 Reactor Vessel Radiation Surveillance Program. ML16202A1672016-07-0505 July 2016 Report 1500227.401, PWR Internals Aging Management Program Plan. ML16004A1792015-12-31031 December 2015 Attachment 2, ANP-3418NP, Revision 0, Arkansas Nuclear One Unit 1 Reactor Vessel Internals License Renewal Scope and MRP-189, Revision 1 Comparison (MRP-227-A Action Item 2) Licensing Report. (Non-Proprietary) ML15278A0242015-09-28028 September 2015 Attachment 2, Areva Document ANP-3417NP, Revision 0, MRP-227-A Applicant / Licensee Action Item No. 7 Analysis for Arkansas Nuclear One, Unit 1 (Non-Proprietary), Attachment 3, Affidavit, and Attachment 4, List of Commitments ML15071A0552015-03-31031 March 2015 ANP-3300Q2NP, Revision 0, Response to Request for Additional Information on Reactor Coolant System Pressure/Temperature and Low Temperature Overpressure Protection System Limits to 54 EFPY for Arkansas Nuclear One, Unit 1. ML15041A0742015-02-0606 February 2015 ANP-3300Q1NP, Rev. 0, Response to Request for Additional Information on Reactor Coolant System Pressure/Temperature and Low Temperature Overpressure Protection System Limits to 54 EFPY for Arkansas Nuclear One, Unit 1, Attachment 2 to 1CAN0 ML14330A2502014-11-30030 November 2014 Attachment 4 to 1CAN111401, ANP-3300, Revision 1, Pressure-Temperature Limits at 54 Efpy. ML14241A2412014-06-30030 June 2014 ANP-3300, Arkansas Nuclear One (ANO) Unit 1 Pressure-Temperature Limits at 54 EFPY, Attachment 4 ML14139A3812014-05-14014 May 2014 CALC-ANO2-CS-12-00002, Revision 1, Flooding Walkdown Report for Resolution of Fukushima Near Term Task Force Recommendation 2.3, Attachment 2 to 0CAN051402 ML14139A3802014-05-14014 May 2014 CALC-ANO1-CS-12-00003, Revision 1, Flooding Walkdown Report for Resolution of Fukushima Near Term Task Force Recommendation 2.3, Attachment 1 to 0CAN051402 1CAN051401, Time-Limited Aging Analysis Regarding Reactor Vessel Internals Loss of Ductility for Arkansas Nuclear One, Unit 1 at 60 Years Arkansas Nuclear One Unit 12014-05-0606 May 2014 Time-Limited Aging Analysis Regarding Reactor Vessel Internals Loss of Ductility for Arkansas Nuclear One, Unit 1 at 60 Years Arkansas Nuclear One Unit 1 ML14141A5552014-05-0101 May 2014 Attachment 1 to 1CAN051403 PWR Internals Aging Management Program Plan ML14007A4592014-02-25025 February 2014 Interim Staff Evaluation Relating to Overall Integrated Plan in Response to Order EA-12-049 (Mitigation Strategies) ML14045A1562014-02-20020 February 2014 Mega-Tech Services, LLC Technical Evaluation Report Regarding the Overall Integrated Plan for Arkansas Nuclear One, Units 1 and 2, TAC Nos.: MF0942 and MF0943 1CAN091301, Updated Seismic Walkdown Report2013-09-30030 September 2013 Updated Seismic Walkdown Report ML13213A2702013-07-22022 July 2013 Stator Drop Root Cause Evaluation Report CR-ANO-C-2013-00888, Rev. 0, Unit 1 Main Turbine Generator Stator. ML13113A2182013-04-23023 April 2013 Technical Letter Report, PNNL Evaluation and Modeling of Licensee'S Alternative for Volumetric Inspection of Dissimilar Metal Butt Welds at Arkansas Nuclear One ML12334A0092012-11-19019 November 2012 CALC-ANO1-CS-12-00003, Flooding Walkdown Submittal Report for Resolution of Fukushima Near-Term Task Force Recommendation 2.3: Flooding, Attachment 1 to 1CAN111202 ML12334A0072012-11-19019 November 2012 CALC-ANO2-CS-12-00002, Rev. 0, Arkansas Nuclear One Unit 2 Flooding Walkdown Submittal Report for Resolution of Fukushima Near-Term Task Force Recommendation 2.3: Flooding. 1CAN111201, Engineering Report CALC-ANO1-CS-12-00002, Rev. 0, Seismic Walkdown Report for Resolution of Fukushima Near-Term Task Force Recommendation 2.3: Seismic, Pages 332 Through 5602012-11-16016 November 2012 Engineering Report CALC-ANO1-CS-12-00002, Rev. 0, Seismic Walkdown Report for Resolution of Fukushima Near-Term Task Force Recommendation 2.3: Seismic, Pages 332 Through 560 ML12342A2202012-11-16016 November 2012 Engineering Report CALC-ANO1-CS-12-00002, Rev. 0, Seismic Walkdown Report for Resolution of Fukushima Near-Term Task Force Recommendation 2.3: Seismic, Pages 1 Through 331 2CAN111201, Arkansas, Unit 2, Attachment 1 - Engineering Report CALC-ANO2-CS-12-00001, Rev. 0, Seismic Walkdown Report for Resolution of Fukushima Near-Term Task Force Recommendation 2.3: Seismic, Page 364 of 533 Through End2012-11-16016 November 2012 Arkansas, Unit 2, Attachment 1 - Engineering Report CALC-ANO2-CS-12-00001, Rev. 0, Seismic Walkdown Report for Resolution of Fukushima Near-Term Task Force Recommendation 2.3: Seismic, Page 364 of 533 Through End ML12342A0522012-11-16016 November 2012 Arkansas, Unit 2, Attachment 1 - Engineering Report CALC-ANO2-CS-12-00001, Rev. 0, Seismic Walkdown Report for Resolution of Fukushima Near-Term Task Force Recommendation 2.3: Seismic, and Attach. 2, List of Regulatory Commitments, Cover - 1CAN111201, Engineering Report CALC-ANO1-CS-12-00002, Rev. 0, Seismic Walkdown Report for Resolution of Fukushima Near-Term Task Force Recommendation 2.3: Seismic, Pages 1 Through 3312012-11-16016 November 2012 Engineering Report CALC-ANO1-CS-12-00002, Rev. 0, Seismic Walkdown Report for Resolution of Fukushima Near-Term Task Force Recommendation 2.3: Seismic, Pages 1 Through 331 ML1200903102012-01-0909 January 2012 Email Apparent Cause Evaluation Report, Final ACE for Tube to Tube Wear ML0832603222008-11-12012 November 2008 Letter to Elmo E. Collins from FEMA, Region IV, Denton, Texas Dated 11-12-2008 Subj: ANO Radiological EP Final Report for ANO ML0821900132008-08-0707 August 2008 Monthly Operating Reports Second Quarter 2008 2CAN040801, Summary of Design and Analyses of Weld Overlays for Hot Leg Nozzle Dissimilar Metal Welds for Alloy 600 Mitigation at ANO-22008-04-0202 April 2008 Summary of Design and Analyses of Weld Overlays for Hot Leg Nozzle Dissimilar Metal Welds for Alloy 600 Mitigation at ANO-2 ML0713703522007-05-0808 May 2007 Stress Analysis Summary Report, Pressurizer and Reactor Coolant Hot Leg Weld Overlays ML0710002572007-03-26026 March 2007 Attachment 5 - HI-2063601, Holtec Licensing Report for ANO Unit 2 Partial Rerack, (non-propriety) 2024-07-02
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Text
Entergy Operations, Inc.
1448 S.R. 333 Russellville, AR 72802 Tel 479-858-4704 Stephenie L. Pyle Manager, Regulatory Assurance Arkansas Nuclear One 1CAN051401 May 6, 2014 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk 11555 Rockville Pike Rockville, MD 20852
SUBJECT:
Time-Limited Aging Analysis Regarding Reactor Vessel Internals Loss of Ductility for Arkansas Nuclear One, Unit 1 at 60 Years Arkansas Nuclear One - Unit 1 Docket No. 50-313 License No. DPR-51
REFERENCES:
- 1. Entergy letter dated January 31, 2000, License Renewal Application, (1CAN010003) (ML003679667)
- 2. BAW-2248A, Demonstration of the Management of Aging Effects for the Reactor Vessel Internals, April 2000 (ML003708443)
- 3. Entergy letter dated August 24, 2000, License Renewal Application RAIs, (1CAN080003) (ML003746995)
- 4. NRC letter dated April 12, 2001, Arkansas Nuclear One, Unit 1 - License Renewal Safety Evaluation Report, (ML011030091)
- 5. Duke Energy letter dated February 20, 2012, License Renewal Commitment to Submit a Time Limiting Aging Analysis for the Reactor Vessel Internals to the NRC for Review, (ML12053A332)
Dear Sir or Madam:
By letter dated January 31, 2000 (Reference 1), Entergy Operations, Inc. (Entergy) submitted a License Renewal Application (LRA) for Arkansas Nuclear One, Unit 1 (ANO-1). As noted in the LRA, BAW-2248A provides a description of the reactor vessel internals for ANO-1. BAW-2248A was developed on a generic basis for several Babcock & Wilcox units, including ANO-1, to demonstrate that the aging effects for the reactor vessel internals are adequately managed for the period of extended operation.
Attachment 1 to this letter contains proprietary information - Attachment 1 is withheld from public disclosure per 10 CFR 2.390.
1CAN051401 Page 2 of 3 As a result of the NRC review of Reference 2, several Renewal Applicant Action Items were identified. The ANO-1 specific response to Renewal Applicant Action Item #12 states A plant-specific analysis will be performed to demonstrate that under LOCA and seismic loading, the internals have adequate ductility to absorb local strain at the regions of maximum stress intensity and that irradiation accumulated at the expiration of the renewal license will not affect deformation limits. Data will be developed to demonstrate that the internals will meet the deformation limits at the expiration of the renewed license.
Reference 3 transmitted responses to NRC Requests for Additional Information (RAIs) pertaining to the reactor vessel and reactor vessel internals. The response to RAI 4.1-1 states The TLAA reported in BAW-2248A regarding ductility of stainless steel and deformation limits will be evaluated by Entergy Operations when sufficient embrittlement data is collected through the BWOG and EPRI MRP programs. Once the embrittlement data is available, the TLAA reported in BAW-2248A will be updated. The TLAA will be resolved using 10 CFR 54.21(c)(1)(iii), which is consistent with the approach taken by Duke Power for the Oconee Units that was approved by the NRC (NUREG-1723, page 4-24). At present, ANO plans to complete the evaluation prior to the end of the current term of operation.
The NRC Safety Evaluation (Reference 4) stated that the staff found that the applicants responses to the Renewal Applicant Action Items resolve the action items from BAW-2248.
Specifically, for Item #12, the staff stated in part that the analysis will be performed as part of the applicants reactor vessel internals aging management program (RVIAMP).
ANO-1 enters the period of extended operation on May 20, 2014. provides the ANO-1 specific time-limited aging analysis for the loss of ductility of the reactor vessel internals. This analysis demonstrates that the internals will meet the deformation limits at the expiration of the renewal license and fulfills the commitment made in Reference 1. The remaining portion of the RVIAMP will be provided under a separate cover letter.
This analysis is considered to be proprietary to AREVA, Inc. AREVA, Inc. requests that the proprietary information be withheld from public disclosure in accordance with 10 CFR 2.390.
AREVA, Inc. has provided Entergy with authorization to provide the proprietary information. An affidavit by the information owner, AREVA, Inc., supporting the request for non-disclosure is provided in Attachment 2. Therefore, Entergy requests that Attachment 1 of this submittal be withheld from public disclosure in accordance with 10 CFR 2.390. Attachment 3 provides a non-proprietary version of the analysis.
This request and analysis are similar to those provided by Duke Energy via Reference 5.
Attachment 1 to this letter contains proprietary information - Attachment 1 is withheld from public disclosure per 10 CFR 2.390.
1CAN051401 Page 3 of 3 This letter contains no new regulatory commitments.
If you have any questions or require additional information, please contact me.
Sincerely, Original signed by Stephenie L. Pyle SLP/rwc
Attachment:
- 1. Time-Limited Aging Analysis Regarding Reactor Vessel Internal Loss of Ductility for Arkansas Nuclear One Unit 1 at 60 Years - PROPRIETARY
- 2. Affidavit
- 3. Time-Limited Aging Analysis Regarding Reactor Vessel Internal Loss of Ductility for Arkansas Nuclear One Unit 1 at 60 Years - NON-PROPRIETARY cc: Mr. Marc L. Dapas Regional Administrator U. S. Nuclear Regulatory Commission, Region IV 1600 East Lamar Boulevard Arlington, TX 76011-4511 NRC Senior Resident Inspector Arkansas Nuclear One P.O. Box 310 London, AR 72847 U. S. Nuclear Regulatory Commission Attn: Mr. Peter Bamford MS O-8B3 One White Flint North 11555 Rockville Pike Rockville, MD 20852 Attachment 1 to this letter contains proprietary information - Attachment 1 is withheld from public disclosure per 10 CFR 2.390.
Attachment 1 to 1CAN051401 Time-Limited Aging Analysis Regarding Reactor Vessel Internals Loss of Ductility for Arkansas Nuclear One Unit 1 at 60 Years PROPRIETARY to 1CAN051401 Affidavit
Attachment 3 to 1CAN051401 Time-Limited Aging Analysis Regarding Reactor Vessel Internals Loss of Ductility for Arkansas Nuclear One Unit 1 at 60 Years NON-PROPRIETARY
Controlled Document ANP-3281NP Time-Limited Aging Analysis Revision 1 Regarding Reactor Vessel Internals Loss of Ductility for Arkansas Nuclear One Unit 1 at 60 Years Licensing Report March 2014 AREVA Inc.
(c) 2014 AREVA Inc.
Controlled Document Copyright © 2014 AREVA Inc.
All Rights Reserved
Controlled Document AREVA Inc. ANP-3281NP Revision 1 Time-Limited Aging Analysis Regarding Reactor Vessel Internals Loss of Ductility for Arkansas Nuclear One Unit 1 at 60 Years Licensing Report Page i Nature of Changes Section(s) or Item Page(s) Description and Justification Rev. 0 All Initial Issue Rev. 1 All Added assumptions and results sections, renumbered other sections accordingly Nomenclature New acronyms added 1.0 New quotations from and discussions regarding ANO-1 LR documentation 5.0 Added RV internals at the end of the last sentence 7.0 Added References 3, 4, and 6, updated Reference 5
Controlled Document AREVA Inc. ANP-3281NP Revision 1 Time-Limited Aging Analysis Regarding Reactor Vessel Internals Loss of Ductility for Arkansas Nuclear One Unit 1 at 60 Years Licensing Report Page ii Contents Page
1.0 INTRODUCTION
............................................................................................... 1-1
2.0 BACKGROUND
................................................................................................. 2-1 3.0 ASSUMPTIONS ................................................................................................ 3-1 4.0 INPUTS ............................................................................................................. 4-1 5.0 ANALYSIS ......................................................................................................... 5-1 6.0 RESULTS .......................................................................................................... 6-1
7.0 REFERENCES
.................................................................................................. 7-1
Controlled Document AREVA Inc. ANP-3281NP Revision 1 Time-Limited Aging Analysis Regarding Reactor Vessel Internals Loss of Ductility for Arkansas Nuclear One Unit 1 at 60 Years Licensing Report Page iii List of Figures Figure 5-1 Effect of Radiation on Uniform Elongation of Annealed Type 304 Stainless Steel Irradiated at 290°C (554°F) and Tested at Various Temperatures ......................................................................................... 5-2 Figure 5-2 [
] ........................................................................... 5-3 Figure 5-3 Change in Uniform Elongation as a Function of Neutron Dose for Solution-Annealed Type 304, 304L, and 347 Stainless Steels at Elevated Temperatures (270-380°C) ...................................................... 5-4 Figure 5-4 Effect of Strain Rate and Temperature on the Uniform Elongation of Type 304 Stainless Steel ........................................................................ 5-5 Figure 5-5 Effect of Strain Rate and Temperature on the 0.2 Percent Yield Stress of Type 304 Stainless Steel ......................................................... 5-5
Document AREVA Inc. ANP-3281NP Revision 1 Time-Limited Aging Analysis Regarding Reactor Vessel Internals Loss of Ductility for Arkansas Nuclear One Unit 1 at 60 Years Licensing Report Page iv Nomenclature Acronym Definition AMP Aging Management Program AMR Aging Management Review ANO-1 Arkansas Nuclear One Unit 1 ASTM American Society of Testing and Materials B&W Babcock and Wilcox B[&]WOG Babcock and Wilcox Owners Group (now Pressurized Water Reactor Owners Group, or PWROG)
CFR Code of Federal Regulations CLB Current Licensing Basis EFPY Effective Full-Power Years EPRI Electric Power Research Institute LOCA Loss Of Coolant Accident LRA License Renewal Application LWR Light Water Reactor MeV Million Electron Volts NRC Nuclear Regulatory Commission RAl Request for Additional Information RCS Reactor Coolant System RV Reactor Vessel RVIAMP Reactor Vessel Internals Aging Management Programa SC Structures and Components SER Safety Evaluation Report TLAA Time-Limited Aging Analysis UFSAR Updated Final Safety Analysis Report a
RVIAMP is used in the direct quotes from the NRC to indicate Reactor Vessel Internals (RVI) Aging Management Program (AMP). In the AREVA text, other forms such as RV internals or reactor vessel internals are used to avoid confusion with Reactor Vessel Integrity (RVI), as commonly used in other AREVA reports.
Controlled Document AREVA Inc. ANP-3281NP Revision 1 Time-Limited Aging Analysis Regarding Reactor Vessel Internals Loss of Ductility for Arkansas Nuclear One Unit 1 at 60 Years Licensing Report Page 1-1
1.0 INTRODUCTION
Entergy Operations submitted a License Renewal Application (LRA) (Reference 1) in January 2000 for Arkansas Nuclear One Unit 1 (ANO-1), which provided the technical information as required by 10 CFR 54. The application intended to provide sufficient information for the Nuclear Regulatory Commission (NRC) to complete its technical reviews. BAW-2248A (Reference 2) was developed on a generic basis for several Babcock and Wilcox (B&W) units, including ANO-1, to demonstrate that the aging effects for the reactor vessel (RV) internals within the scope of Reference 2 are adequately managed for the period of extended operation.
As detailed in the LRA, the NRC issued several renewal application action items as the result of their review of BAW-2248A. Renewal applicant action item #12 reads as follows:
Plant-specific analysis is required to demonstrate that, under loss-of-coolant-accident (LOCA) and seismic loading, the internals have adequate ductility to absorb local strain at the regions of maximum stress intensity and that irradiation accumulated at the expiration of the renewal license will not adversely affect deformation limits. The RVIAMP must develop data to demonstrate that the internals will meet the deformation limits at that expiration of the renewal license.
Document AREVA Inc. ANP-3281NP Revision 1 Time-Limited Aging Analysis Regarding Reactor Vessel Internals Loss of Ductility for Arkansas Nuclear One Unit 1 at 60 Years Licensing Report Page 1-2 As a result of this requirement, Entergy provided the following response within the LRA (see Section 2.3.1.6 and Table 2.3-5 of Reference 1):
A plant-specific analysis will be performed to demonstrate that under LOCA and seismic loading, the internals have adequate ductility to absorb local strain at the regions of maximum stress intensity and that irradiation accumulated at the expiration of the renewal license will not affect deformation limits. Data will be developed to demonstrate that the internals will meet the deformation limits at the expiration of the renewed license.
This requirement is captured as a time-limited aging analysis (TLAA) in Table 4.1-1 of the LRA (Reference 1) and referenced in the reactor vessel internals aging management program in Appendix B of Reference 1. In response to RAI 4.1-1 provided by the NRC staff on April 25, 2000 regarding the timing and means of how this TLAA will be addressed (Reference 3), ANO-1 provided the following response on August 24, 2000 (Reference 4):
The TLAA reported in BAW-2248A regarding ductility of stainless steel and deformation limits will be evaluated by Entergy Operations when sufficient embrittlement data is collected through the BWOG and EPRI MRP programs.
Once the embrittlement data is available, the TLAA reported in BAW-2248A will be updated. The TLAA will be resolved using 10CFR54.21(c)(1)(iii), which is consistent with the approach taken by Duke Power for the Oconee Units that was approved by the NRC (NUREG-1723, page 4-24). At present, ANO plans to complete the evaluation prior to the end of the current term of operation.
The NRCs safety evaluation report (SER) for the ANO-1 LRA (Reference 5) provided the following in Section 3.3.2.4.3:
Controlled Document AREVA Inc. ANP-3281NP Revision 1 Time-Limited Aging Analysis Regarding Reactor Vessel Internals Loss of Ductility for Arkansas Nuclear One Unit 1 at 60 Years Licensing Report Page 1-3 On the basis of the review described above, the staff finds that the applicant has demonstrated that the effects of aging associated with the reactor vessel internals will be adequately managed so that there is reasonable assurance that the intended function will be maintained with the CLB for the period of extended operation.
The NRCs safety evaluation report (SER) for the ANO-1 LRA (Reference 6) provided the following in Section 4.1.3:
The NRC staff concludes that the applicant has provided a list of acceptable TLAAs as defined in 10 CFR 54.3, and that no 10 CFR 50.12 exemptions have been granted on the basis of a TLAA as defined in 10 CFR 54.3.
This document provides the plant-specific evaluation of ductility for the RV internals at the expiration of the renewed license, which is 54 Effective Full-Power Years (EFPY) using projected fluence values for ANO-1, as required per LRA and SER.
Information considered proprietary to AREVA is marked in square brackets, [ ]
Controlled Document AREVA Inc. ANP-3281NP Revision 1 Time-Limited Aging Analysis Regarding Reactor Vessel Internals Loss of Ductility for Arkansas Nuclear One Unit 1 at 60 Years Licensing Report Page 2-1
2.0 BACKGROUND
In 2010, Appendix E of the 1970 RV internals topical report was updated by AREVA through a contract with the Electric Power Research Institute (EPRI) for 60 years on a generic basis for the B&W units and submitted, for information, to the NRC.
(Reference 7) This update identified the locations of the maximum stress intensity where a loss of ductility because of neutron irradiation would be detrimental (from Appendix E of the 1970 RV internals topical report) as the core barrel flanges.
However, upon more detailed examination of the wording and stress intensity values presented in the 1970 RV internals topical report, the location of highest stress intensity occurs at the core support shield bottom flange.
The excerpts describing such locations from the 1970 RV internals topical report, Section 3.2.3.2 are below:
Controlled Document AREVA Inc. ANP-3281NP Revision 1 Time-Limited Aging Analysis Regarding Reactor Vessel Internals Loss of Ductility for Arkansas Nuclear One Unit 1 at 60 Years Licensing Report Page 2-2
[
] Therefore, the bottom core support shield flange will be evaluated within this document as the region of maximum stress intensity, as required by renewal applicant action item #12 from BAW-2248A.
Controlled Document AREVA Inc. ANP-3281NP Revision 1 Time-Limited Aging Analysis Regarding Reactor Vessel Internals Loss of Ductility for Arkansas Nuclear One Unit 1 at 60 Years Licensing Report Page 3-1 3.0 ASSUMPTIONS There are no assumptions for this document.
Controlled Document AREVA Inc. ANP-3281NP Revision 1 Time-Limited Aging Analysis Regarding Reactor Vessel Internals Loss of Ductility for Arkansas Nuclear One Unit 1 at 60 Years Licensing Report Page 4-1 4.0 INPUTS This section identifies and provides inputs to the ANO-1-specific evaluation of ductility for the RV internals at 54 EFPY. The first required input is a projected fluence value specifically generated for ANO-1 at 54 EFPY. The methodology used to determine the neutron fluence was based on AREVAs NRC approved fluence analysis methodology, described in topical report BAW-2241P-A (References 8, 9, and 10). The fluence methodology is consistent with the guidance of Regulatory Guide 1.190, Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence.
As described in Section 2.0 of this document, the bottom of the core support shield (i.e.,
at the lower flange) is the location of interest and the projected 54 EFPY fluence for this location at ANO-1 is [ ] Using the light water reactor (LWR) conversion factor of 15 dpa = 1 x 1022 n/cm2 (E>1.0 MeV) (Reference 11), this converts to [ ] In addition, it is noted that the fluence at the top of the core support shield (i.e., at the upper flange) would be less than at the bottom of the core support shield because of the increased distance from the core.
The second required input is the material used for the manufacturing of the core support shield top and bottom flanges. As detailed in Reference 2, the core support shield flanges are fabricated from American Society of Testing and Materials (ASTM) A 473-63 Type 304 austenitic stainless steel.
Controlled Document AREVA Inc. ANP-3281NP Revision 1 Time-Limited Aging Analysis Regarding Reactor Vessel Internals Loss of Ductility for Arkansas Nuclear One Unit 1 at 60 Years Licensing Report Page 5-1 5.0 ANALYSIS The projected 54 EFPY fluence is [ ] for the core support shield lower flange, which has been shown to be the region of maximum stress intensity for the RV internals; the fluence at the top of the core support shield would be less than this value because of the increased distance from the core.
Figure 5-1 (Figure E-3 of the 1970 RV internals topical report) depicts that for a fluence of [
] Note that Figure 5-2 (Figure 3-12 in Reference 7) provides recent irradiated Type 304 test data to compare to the curves in Figure 5-1. The test data validates the conservatism of the curves in Figure 5-1. This slight decrease in uniform elongation at this level of fluence is confirmed in Figure 5-3 (Figure 13(c) of Reference 11).
In addition, the uniform elongation of unirradiated solution annealed Type 304 stainless steel at 600°F is seen to only decrease slightly with increasing strain rate as shown in Figure 5-4 (Figure 5 of Reference 12). However, even at the highest tested strain rates, at 600°F, the uniform elongation is above the 20 percent uniform elongation of irradiated material credited for 40 years in Appendix E of the 1970 RV internals topical report and the 8.6 percent allowable strain specified in Appendix A of the 1970 RV internals topical report. It is also observed that yield strength increases with increasing strain rate at 600°F as shown in Figure 5-5 (Figure 3 of Reference 12). In addition to having sufficient ductility at 60 years relative to the allowables of the 1970 RV internals topical report, the upper and lower core support shield flanges will have greater resistance to plastic deformation at increased strain rates.
Controlled Document AREVA Inc. ANP-3281NP Revision 1 Time-Limited Aging Analysis Regarding Reactor Vessel Internals Loss of Ductility for Arkansas Nuclear One Unit 1 at 60 Years Licensing Report Page 5-2 Therefore, the conclusions from Appendix E to the 1970 RV internals topical report concerning the acceptable ductility and deformation limits for a 40-year lifetime remain valid for a 60-year lifetime for the ANO-1 RV internals.
Figure 5-1 Effect of Radiation on Uniform Elongation of Annealed Type 304 Stainless Steel Irradiated at 290°C (554°F) and Tested at Various Temperatures Note: This is Figure E-3 in Appendix E of the 1970 RV internals topical report.
Controlled Document AREVA Inc. ANP-3281NP Revision 1 Time-Limited Aging Analysis Regarding Reactor Vessel Internals Loss of Ductility for Arkansas Nuclear One Unit 1 at 60 Years Licensing Report Page 5-3 Figure 5-2 [
]
Controlled Document AREVA Inc. ANP-3281NP Revision 1 Time-Limited Aging Analysis Regarding Reactor Vessel Internals Loss of Ductility for Arkansas Nuclear One Unit 1 at 60 Years Licensing Report Page 5-4 Figure 5-3 Change in Uniform Elongation as a Function of Neutron Dose for Solution-Annealed Type 304, 304L, and 347 Stainless Steels at Elevated Temperatures (270-380°C)
Note: This is Figure 13(c) in Reference 11.
Controlled Document AREVA Inc. ANP-3281NP Revision 1 Time-Limited Aging Analysis Regarding Reactor Vessel Internals Loss of Ductility for Arkansas Nuclear One Unit 1 at 60 Years Licensing Report Page 5-5 Figure 5-4 Effect of Strain Rate and Temperature on the Uniform Elongation of Type 304 Stainless Steel Note: This is Figure 5 in Reference 12.
Figure 5-5 Effect of Strain Rate and Temperature on the 0.2 Percent Yield Stress of Type 304 Stainless Steel Note: This is Figure 3 in Reference 12.
Controlled Document AREVA Inc. ANP-3281NP Revision 1 Time-Limited Aging Analysis Regarding Reactor Vessel Internals Loss of Ductility for Arkansas Nuclear One Unit 1 at 60 Years Licensing Report Page 6-1 6.0 RESULTS By NRC submittal, Entergy Operations committed to providing a plant-specific analysis to demonstrate that under LOCA and seismic loading, the internals have adequate ductility to absorb local strain at the regions of maximum stress intensity and that irradiation accumulated at the expiration of the renewed license will not affect deformation limits for ANO-1 prior to the end of the current term of operation. A projected fluence value at 54 EFPY was developed for ANO-1. Based on material data included in a 1970 RV internals topical report and newer data, the conclusions from Appendix E to the 1970 RV internals topical report concerning the acceptable ductility and deformation limits for a 40-year lifetime remain valid for a 60-year lifetime for the ANO-1 RV internals.
Controlled Document AREVA Inc. ANP-3281NP Revision 1 Time-Limited Aging Analysis Regarding Reactor Vessel Internals Loss of Ductility for Arkansas Nuclear One Unit 1 at 60 Years Licensing Report Page 7-1
7.0 REFERENCES
- 1. Arkansas Nuclear One - Unit 1 Docket No. 313, License No. DPR-51, License Renewal Application, January 31, 2000. NRC Accession Number ML003679667.
- 2. BAW-2248A, Demonstration of the Management of Aging Effects for the Reactor Vessel Internals, April 2000. NRC Accession Number ML003708443.
- 3. Request for Additional Information for the Review of the Arkansas Nuclear One, Unit 1, License Renewal Application, April 25, 2000, NRC Accession Number ML003707431.
- 4. Arkansas Nuclear One - Unit 1, Docket No. 50-313, License No. DPR-51, License Renewal Application RAIs (TAC No. MA8064), August 24, 2000, NRC Accession Number ML003746995.
- 5. NUREG-1743, Safety Evaluation Report Related to License Renewal of Arkansas Nuclear One, Unit 1, Chapter 3, May 2001, NRC Accession Number ML011640177.
- 6. NUREG-1743, Safety Evaluation Report Related to License Renewal of Arkansas Nuclear One, Unit 1, Chapter 4-End, May 2001, NRC Accession Number ML011640217.
- 7. Information in Support of the EPRI Materials Reliability Program (MRP):
Pressurized Water Reactor Internals Inspection and Evaluation Guidelines (MRP-227-Rev. 0) Review, October 29, 2010, NRC Accession Number ML103090248.
- 8. Letter from James F. Malley to NRC Document Control Desk, Submittal of BAW-2241P, Revision 2, Fluence and Uncertainty Methodologies, June 2, 2003, NRC Accession Number ML031550365.
Controlled Document AREVA Inc. ANP-3281NP Revision 1 Time-Limited Aging Analysis Regarding Reactor Vessel Internals Loss of Ductility for Arkansas Nuclear One Unit 1 at 60 Years Licensing Report Page 7-2
- 9. BAW-2241NP-A, Revision 2, Fluence and Uncertainty Methodologies, April 30, 2006, NRC Accession Number ML073310660.
- 10. Letter from Ronnie L. Gardner to NRC Document Control Desk, Publication of Revision 1 of Appendix G to BAW-2241(P), Revision 2, Fluence and Uncertainty Methodologies, November 20, 2007, NRC Accession Number ML073310655.
- 11. NUREG/CR-7027, Degradation of LWR Core Internal Materials Due to Neutron Irradiation, December 2010, NRC Accession Number ML102790482.
- 12. High Strain Rate Tensile Properties of AISI Type 304 Stainless Steel, J.
M. Steichen, Journal of Engineering Materials and Technology, July 1973.