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EPID:L-2022-LLA-0161, Application to Remove Technical Specification Condition Allowing Two Reactor Coolant Pump Operation (Approved, Closed) |
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MONTHYEAR1CAN102202, Application to Remove Technical Specification Condition Allowing Two Reactor Coolant Pump Operation2022-10-31031 October 2022 Application to Remove Technical Specification Condition Allowing Two Reactor Coolant Pump Operation Project stage: Request ML22341A1202022-12-0707 December 2022 NRR E-mail Capture - ANO Unit 1 - Acceptance of Requested Licensing Action LAR to to Remove Technical Specification Condition Allowing Two Reactor Coolant Pump Operation (L-2022-LLA-0161) Project stage: Acceptance Review ML23020A9362023-01-27027 January 2023 Audit Plan to Support Review of License Amendment Request to Remove Technical Specification Condition Allowing Two Reactor Coolant Pump Operation Project stage: Other ML23188A1732023-07-12012 July 2023 Request for Withholding Information from Public Disclosure Project stage: Withholding Request Acceptance ML23208A2132023-08-0303 August 2023 Regulatory Audit Summary Concerning Review of License Amendment Request to Remove Technical Specification Condition Allowing Two Reactor Coolant Pump Operation Project stage: Approval ML23275A2072023-11-28028 November 2023 Issuance of Amendment No. 280 Removal of Technical Specification Condition Allowing Two Reactor Coolant Pump Operation Project stage: Approval 2023-01-27
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Category:Letter type:
MONTHYEAR2CAN012401, 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) 2CAN112302, Submittal of Amendment 31 to Safety Analysis Report2023-11-16016 November 2023 Submittal of Amendment 31 to Safety Analysis Report 0CAN102303, Registration of Cask Use2023-10-24024 October 2023 Registration of Cask Use 0CAN102302, (ANO) Emergency Plan Revision 49 and Emergency Plan On-Shift Staffing Analysis Revision 32023-10-11011 October 2023 (ANO) Emergency Plan Revision 49 and Emergency Plan On-Shift Staffing Analysis Revision 3 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 2CAN012302, Relief Request ANO2-RR-23-001, Half-Nozzle Repair of Reactor Vessel Closure Head Penetration 462023-01-20020 January 2023 Relief Request ANO2-RR-23-001, Half-Nozzle Repair of Reactor Vessel Closure Head Penetration 46 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) 1CAN102202, Application to Remove Technical Specification Condition Allowing Two Reactor Coolant Pump Operation2022-10-31031 October 2022 Application to Remove Technical Specification Condition Allowing Two Reactor Coolant Pump Operation 1CAN102203, Cycle 31 Core Operating Limits Report2022-10-24024 October 2022 Cycle 31 Core Operating Limits Report 1CAN102201, Supplement to Proposed Technical Specifications 3.4.12 and 3.4.13 Revised Dose Calculations2022-10-13013 October 2022 Supplement to Proposed Technical Specifications 3.4.12 and 3.4.13 Revised Dose Calculations 0CAN092201, Supplement to License Amendment Request for Approval of Changes to the Emergency Plan Staffing Requirements2022-09-29029 September 2022 Supplement to License Amendment Request for Approval of Changes to the Emergency Plan Staffing Requirements 1CAN092201, Supplement to Request for Relief from American Society of Mechanical Engineers Section XI Volumetric Examination Requirements - Fifth 10-Year Interval, First Period2022-09-0808 September 2022 Supplement to Request for Relief from American Society of Mechanical Engineers Section XI Volumetric Examination Requirements - Fifth 10-Year Interval, First Period 0CAN082201, License Amendment Request for Approval of Changes to the Emergency Plan Staffing Requirements2022-08-30030 August 2022 License Amendment Request for Approval of Changes to the Emergency Plan Staffing Requirements 1CAN082201, Request for Relief from American Society of Mechanical Engineers Section XI Volumetric Examination Requirements - Fifth 10-Year Interval, First Period2022-08-24024 August 2022 Request for Relief from American Society of Mechanical Engineers Section XI Volumetric Examination Requirements - Fifth 10-Year Interval, First Period 2CAN072201, Response to Request for Additional Information Requests for Relief from American Society of Mechanical Engineers Section XI Volumetric Examination Requirements - Fourth 10-Year Interval, Second and Third Periods2022-07-20020 July 2022 Response to Request for Additional Information Requests for Relief from American Society of Mechanical Engineers Section XI Volumetric Examination Requirements - Fourth 10-Year Interval, Second and Third Periods 0CAN072201, Registration of Cask Use2022-07-14014 July 2022 Registration of Cask Use 0CAN062202, Registration of Cask Use2022-06-0909 June 2022 Registration of Cask Use 1CAN062201, Response to the Request for Additional Information Proposed Technical Specifications 3.4.12 and 3.4.13 Revised Dose Calculations2022-06-0202 June 2022 Response to the Request for Additional Information Proposed Technical Specifications 3.4.12 and 3.4.13 Revised Dose Calculations 1CAN052201, Request for Relief Related to American Society of Mechanical Engineers (ASME) Code Case N-729-6 Augmented Examination Requirements ANO1-ISI-0352022-05-31031 May 2022 Request for Relief Related to American Society of Mechanical Engineers (ASME) Code Case N-729-6 Augmented Examination Requirements ANO1-ISI-035 2024-01-17
[Table view] Category:Report
MONTHYEAR1CAN062304, 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 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 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 ML21083A1422021-03-22022 March 2021 Strategy 4 ML21083A1402021-03-22022 March 2021 Strategy 2 ML21083A1412021-03-22022 March 2021 Strategy 3 ML21083A1442021-03-22022 March 2021 Strategy 6 ML21083A1382021-03-22022 March 2021 Rulemaking ML21083A1432021-03-22022 March 2021 Strategy 5 ML21083A1372021-03-22022 March 2021 NEIMA Reporting ML21083A1392021-03-22022 March 2021 Strategy 1 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 ML15043A1032015-02-10010 February 2015 Areva Document ANP-3383NP, Response to Request for Additional Information for the Reactor Pressure Vessel Internals Aging Management Program Plan 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 2023-06-29
[Table view] Category:Technical
MONTHYEAR1CAN062304, 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. 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 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 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 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 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) ML0622204422006-08-10010 August 2006 Holtec Report HI-2022867 Spent Fuel Pool Racks Modifications with Poison Material Inserts in ANO Unit 1 for Entergy. ML0522303712005-08-0808 August 2005 Attachment 2, HI-2043262NP, Rev 0 - Part 50 Criticality Analysis of the MPC-32 for ANO Unit 2, Holtec Project No: 1104, Report Class Safety Related 2023-06-29
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Text
NOTICE: Enclosure 1 to this letter contains Proprietary Information to be withheld from public disclosure per 10 CFR 2.390. Upon separation from Enclosure 1, this letter is DECONTROLLED.
Phil Couture Senior Manager Fleet Regulatory Assurance - Licensing Tel 601-368-5102 1CAN062304 10 CFR 50.90 June 29, 2023 ATTN: Document Control Desk U.S. Nuclear Regulatory Commission Washington, DC 20555
Subject:
Supplement Related to License Amendment Request to Remove Technical Specification Condition Allowing Two Reactor Coolant Pump Operation Arkansas Nuclear One, Unit 1 NRC Docket No. 50-313 Renewed Facility Operating License No. DPR-51
Reference:
Entergy letter to the NRC, "Application to Remove Technical Specification Condition Allowing Two Reactor Coolant Pump Operation," (1CAN102202),
ML22304A669, dated October 31, 2022.
By letter dated October 31, 2022 (Reference), Entergy Operations, Inc. (Entergy) requested U.S. Nuclear Regulatory Commission (NRC) approval of a proposed change to the Arkansas Nuclear One, Unit 1 (ANO-1) Technical Specifications (TS) that would remove the condition that allows for two reactor coolant pump (RCP) operation while at power with one RCP in operation in each Reactor Coolant System (RCS) loop. The change removes TS 3.4.4 "RCS Loops -
MODES 1 and 2" Condition A, "One RCP not in operation in each loop," Required Action A.1 to "Restore one non-operating RCP to operation," and its Completion Time of "18 hours2.083333e-4 days <br />0.005 hours <br />2.97619e-5 weeks <br />6.849e-6 months <br />." This change makes ANO-1 TS 3.4.4 consistent with the NRC Standard Technical Specifications for Babcock and Wilcox (B&W) Plants (NUREG-1430 Revision 5).
To ensure that an adjustment to the Reactor Protection System (RPS) setpoint for reactor power with one RCP in service per loop was not required, Framatome developed the proprietary calculation FS1-0061634, "ANO-1 TS 3.4.4 LAR Support - RPS Pump-to-Power Monitor Setpoint." Since this document was proprietary, it was not submitted with the license amendment request (Reference). The NRC staff has requested that the calculations used to justify the License Amendment Request (LAR) and retain the 55% Rated Thermal Power (RTP) with 1 RCP in service per loop allowable value be provided in a supplement to the Reference letter.
Entergy Operations, Inc., 1340 Echelon Parkway, Jackson, MS 39213
NOTICE: Enclosure 1 to this letter contains Proprietary Information to be withheld from public disclosure per 10 CFR 2.390. Upon separation from Enclosure 1, this letter is DECONTROLLED.
1CAN062304 Page 2 of 3 Also, during a review of the Reference letter, Entergy discovered that an incorrect value for the maximum hypothetical pump-to-power monitor (PPM) trip setpoint was listed. On Enclosure 1, Page 9 of 16, in the 1st paragraph, the following statement was made:
" the ANO-1 PPM (1/1) trip setpoint could be as high as 62.3% RTP and still provide the required steady-state and transient protection."
This should state:
" the ANO-1 PPM (1/1) trip setpoint could be as high as 62.1% RTP and still provide the required steady-state and transient protection."
This value can be found calculated in Enclosure 1, Section 6.0, in the calculation preceding the last paragraph on the page. The last paragraph correctly states the ANO-1 TS 3.4.4 pump-to-power monitor trip with one RCP in service per loop could be as high as 62.1% and still provide the required steady-state and transient protection. Replacing the incorrect value for PPM provided in the Reference letter does not change Entergy's conclusions that the proposed change meets applicable requirements and does not adversely affect nuclear safety. It has no impact on the no significant hazards consideration provided in the original amendment request (Reference).
The enclosed requested supporting Framatome calculations have no impact on the no significant hazards consideration provided in the original amendment request (Reference).
Some information provided in Enclosure 1 is considered proprietary to Framatome who requests that it to be withheld from public disclosure in accordance with 10 CFR 2.390 of the Commissions regulations. The proprietary information is identified by text enclosed within bolded brackets [Example]. The non-proprietary version is provided in Enclosure 2.
This information is supported by an affidavit, signed by Morris Byram, Manager, Licensing and Regulatory Affairs for Framatome Inc. (2101 Horn Rapids Road, Richland, WA 99354), the owner of the information. The affidavit sets forth the basis by which the information may be withheld from public disclosure by the Commission and addresses with specificity the considerations listed in paragraph (b)(4) of 10 CFR 2.390 of the Commission's regulations. The affidavit is included as Enclosure 3.
There are no new regulatory commitments established in this submittal.
In accordance with 10 CFR 50.91, Entergy is notifying the State of Arkansas of Entergy's supplemental information by transmitting a copy of this letter and enclosure to the designated State Official.
If there are any questions or if additional information is needed, please contact Riley Keele, Manager, Regulatory Assurance, Arkansas Nuclear One, at 479-858-7826.
NOTICE: Enclosure 1 to this letter contains Proprietary Information to be withheld from public disclosure per 10 CFR 2.390. Upon separation from Enclosure 1, this letter is DECONTROLLED.
1CAN062304 Page 3 of 3 I declare under penalty of perjury that the foregoing is true and correct. Executed on June 29, 2023.
Respectfully, Philip Digitally sig ned by Philip Co uture Couture Dat e: 2023 .06.29 14:47:1 7 -05'00' Phil Couture PC/mar
Enclosures:
- 1. Technical Report-ANO-1 TS 3.4.4 LAR Support- RPS Pump-to-Power Monitor Setpoint Summary ANP-4048P, Revision 1 (PROPRIETARY)
- 2. Technical Report-ANO-1 TS 3.4.4 LAR Support- RPS Pump-to-Power Monitor Setpoint Summary ANP-4048NP, Revision 1 (NON-PROPRIETARY)
- 3. Affidavit cc: NRC Region IV Regional Administrator NRC Senior Resident Inspector - Arkansas Nuclear One NRC Project Manager - Arkansas Nuclear One Designated Arkansas State Official
ENCLOSURE 2 1CAN062304 Technical Report - ANO-1 TS 3.4.4 LAR Support - RPS Pump-to-Power Monitor Setpoint Summary ANP-4048NP, Revision 1 (18 Pages)
(NON-PROPRIETARY)
ANO-1 TS 3.4.4 LAR Support - RPS ANP-4048NP Revision 1 Pump-to-Power Monitor Setpoint Summary Technical Report June 2023 (c) 2023 Framatome Inc.
0414-12-F04 (Rev. 004, 04/27/2020)
ANP-4048NP Revision 1 Copyright © 2023 Framatome Inc.
All Rights Reserved 0414-12-F04 (Rev. 004, 04/27/2020)
Framatome Inc. ANP-4048NP Revision 1 ANO-1 TS 3.4.4 LAR Support - RPS Pump-to-Power Monitor Setpoint Summary Technical Report Page i Nature of Changes Section(s) or Item Page(s) Description and Justification 1 Section 3.0 Corrected instances of 3-to-1 PCD to 3-to-2 PCD and Table 4-1 2 Section 4.0 Added and corrected modifiers for >, <, and = for PPM and Table 4-1 trips 3 Section 6.0 Removed DNBR specific results
Framatome Inc. ANP-4048NP Revision 1 ANO-1 TS 3.4.4 LAR Support - RPS Pump-to-Power Monitor Setpoint Summary Technical Report Page ii Contents Page
1.0 INTRODUCTION
............................................................................................... 1-1 2.0 ANALYTICAL METHODOLOGY ....................................................................... 2-1 3.0 HISTORICAL BASIS ......................................................................................... 3-1 4.0 TRANSIENT PROTECTION .............................................................................. 4-1 5.0 STEADY-STATE PROTECTION ....................................................................... 5-1 5.1 Maximum Real Power - RPS Flux/Flow Setpoint .................................... 5-1 5.2 Maximum Real Power - RPS PPM (1/1) Setpoint ................................... 5-2 6.0 2 RCP (1/1) STEADY-STATE DNB ANALYSES ............................................... 6-1 7.0
SUMMARY
AND CONCLUSION ....................................................................... 7-1
8.0 REFERENCES
.................................................................................................. 8-1
Framatome Inc. ANP-4048NP Revision 1 ANO-1 TS 3.4.4 LAR Support - RPS Pump-to-Power Monitor Setpoint Summary Technical Report Page iii List of Tables Table 4-1: ANO-1 LOCF Events and RPS Trip Functions ........................................... 4-2
Framatome Inc. ANP-4048NP Revision 1 ANO-1 TS 3.4.4 LAR Support - RPS Pump-to-Power Monitor Setpoint Summary Technical Report Page iv Nomenclature Acronym Definition ANO-1 Arkansas Nuclear One Unit 1 AV Allowable Value B&W Babcock and Wilcox CHF Critical Heat Flux COLR Core Operating Limits Report DNB Departure from Nucleate Boiling DNBR Departure from Nucleate Boiling Ratio FP Full Power LAR License Amendment Request LCO Limiting Condition for Operation LOCF Loss of Coolant Flow NRC Nuclear Regulatory Commission PCD Pump Coastdown PPM Pump-to-Power, Power-to-Pump, or Power/Pump Monitor P-T Pressure Temperature RCP Reactor Coolant Pump RCS Reactor Coolant System RPS Reactor Protection System RSS Root Square Sum RTP Rated Thermal Power SCD Statistical Core Design SDL Statistical Design Limit SQRT Square Root TDL Thermal Design Limit TS Technical Specifications
Framatome Inc. ANP-4048NP Revision 1 ANO-1 TS 3.4.4 LAR Support - RPS Pump-to-Power Monitor Setpoint Summary Technical Report Page 1-1
1.0 INTRODUCTION
Entergy is submitting a license amendment request (LAR) for Arkansas Nuclear One Unit 1 (ANO-1) Technical Specification (TS) 3.4.4, RCS Loops, Modes 1 and 2 (Reference 1). ANO-1 TS 3.4.4 Condition A currently allows for one reactor coolant pump (RCP) in operation per loop for 18 hours2.083333e-4 days <br />0.005 hours <br />2.97619e-5 weeks <br />6.849e-6 months <br /> before requiring the plant to be in Mode 3 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. To be in line with NUREG 1430 (Reference 2) as well as the TS for the other operating Babcock & Wilcox (B&W) designed plants, the LAR will remove TS 3.4.4 Condition A. The remaining condition will then require the plant to enter Mode 3 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> if the limiting condition for operation (LCO) is not met.
This document provides analytical support that may be used to justify maintaining the reactor protection system (RPS) RCP-to-Power trip allowable value (AV) at 55 %RTP for 1/1 operation as defined in TS Table 3.3.1-1 (Reference 1).
Framatome Inc. ANP-4048NP Revision 1 ANO-1 TS 3.4.4 LAR Support - RPS Pump-to-Power Monitor Setpoint Summary Technical Report Page 2-1 2.0 ANALYTICAL METHODOLOGY The NRC-approved methods used to establish RPS setpoints and perform departure from nucleate boiling (DNB) calculations for the ANO-1 plant are described in Sections 7 and 6 of Reference 3, respectively.
The B&W plant methodology for DNB response analysis uses the Statistical Core Design (SCD) methodology (Reference 4) to statistically treat uncertainties. This results in a plant specific Statistical Design Limit (SDL). The SDL is higher than the applicable critical heat flux (CHF) correlation because it accounts for uncertainties. Additional margin can be added by arbitrarily increasing the SDL. The increased value is the Thermal Design Limit (TDL) which the DNB analyses then target. This methodology, as applied to B&W fuel reload analyses, is further described in Section 6.2.9 of Reference 3.
The steady-state and transient core thermal-hydraulic DNB response is analyzed with the LYNXT code (Reference 5). The LYNXT code has been approved by the NRC for use in core thermal-hydraulic analyses for the B&W-designed operating plants. The methodology used for the DNB analyses is described further in Sections 6.6 and 6.7 of Reference 3.
Framatome Inc. ANP-4048NP Revision 1 ANO-1 TS 3.4.4 LAR Support - RPS Pump-to-Power Monitor Setpoint Summary Technical Report Page 3-1 3.0 HISTORICAL BASIS The RPS power-to-pump monitor (PPM) overpower trip setpoint for 2 RCP (1/1) operation was originally designed to allow the plant to perform a controlled runback or shutdown following the loss of 1 or 2 RCPs without tripping the plant. This was based on the original Technical Specifications when the flux/flow trip was the primary trip and the power/pump monitor trip was secondary. However, with the transition of the B&W plants to the Standard Technical Specifications (Reference 2), the power/pump monitor became the primary trip and flux/flow the secondary trip. The development of the PPM 2-RCP (1/1) setpoint methodology established the following equations:
The methodology identified that the 4-to-2 and 3-to-2 pump coastdown (PCD) events, which are loss of coolant flow (LOCF) transients evolving to 2-RCP flow conditions, needed to be justified for the PPM (1/1) overpower portion of the setpoint. Both accident scenarios were already analyzed using the flux/flow setpoint, but it was also required to demonstrate that the PPM (1/1) trip provided protection for these events as well.
Framatome Inc. ANP-4048NP Revision 1 ANO-1 TS 3.4.4 LAR Support - RPS Pump-to-Power Monitor Setpoint Summary Technical Report Page 3-2 After determining the maximum allowable power, the PPM (1/1) TS setpoint (SP14) was then established as Historical calculations for several B&W plants demonstrated that the PPM (1/1) TS setpoint could be established at values greater than 65 %FP. However, the final PPM (1/1) AV was ultimately set at a conservative value of 55 %RTP for the Standard TS format for all B&W-designed plants. This approach precluded the need to further demonstrate that the PPM (1/1) trip protects the 4-to-2 PCD and 3-to-2 PCD events on a plant- or cycle-specific basis.
When considering a scenario where the plant may be operating with 2 RCPs during TS 3.4.4 LCO conditions, both the PPM (1/1) trip and the flux/flow trip provide steady-state DNB overpower protection. The following sections discuss the transient and steady-state protection specifically for the ANO-1 plant.
Framatome Inc. ANP-4048NP Revision 1 ANO-1 TS 3.4.4 LAR Support - RPS Pump-to-Power Monitor Setpoint Summary Technical Report Page 4-1 4.0 TRANSIENT PROTECTION As modeled in the accident analyses, the PPM trip is a digital type trip. If the accident analysis assumes a loss of primary flow from a reactor coolant pump trip, the trip condition is based on the initial core power. If the core power is greater than the allowed final pump status power level, a trip condition exists and a reactor trip is initiated. For example, a 4 PCD transient results in a final RCP configuration of 0/0, but the PPM will trip the plant for all idle loop conditions if the power level is greater than 0 %FP.
Consequently, the 4 PCD transient analysis models the PPM trip signal at time zero with the control rods beginning to drop after the PPM trip delay time.
The ANO-1 plant has non-redundant power-pump monitors. A single failure of a monitor would affect all four RPS channels. The primary trip for the 2 PCD event from 4 RCP operation at 100 %FP is an immediate trip on PPM for flux greater than 0 %RTP (0/2, 2/0) OR on PPM for flux greater than 55 %RTP (1/1). In accident analysis space, however, the single failure requirement results in no PPM trip and the accident is protected by the flux/flow trip. Because of the delayed trip on flux/flow, the 4-to-2 PCD event produces the lowest DNBR and sets the basis for the RPS flux/flow setpoint at ANO-1.
Framatome Inc. ANP-4048NP Revision 1 ANO-1 TS 3.4.4 LAR Support - RPS Pump-to-Power Monitor Setpoint Summary Technical Report Page 4-2 Table 4-1 summarizes the ANO-1 LOCF accidents that must be considered and the RPS setpoints that protect each scenario. The table shows that all LOCF events protected by the primary PPM trip will trip immediately regardless of whether the PPM trip is > 0 %RTP or > 55 %RTP since the initial power for 4-pump or 3-pump normal operating conditions is greater than 55 %RTP. For a scenario where the plant is operating at a power level less than or equal to 55 %FP, a 4-to-2 PCD or 3-to-2 PCD event would not initiate a plant trip. This would allow the operator to begin a controlled cooldown to Mode 3 to be in compliance with LCO requirements outlined in the proposed TS 3.4.4 LAR.
Table 4-1: ANO-1 LOCF Events and RPS Trip Functions Potential Plant Initial RPS Trip LOCF Initial Final RCP Potential Plant RPS Trips w/
RCP Used in Event Power Configuration RPS Trips Single Failure Operation Analysis of a Pump Monitor 4-to-3 4 RCP 100 %FP 2/1, 1/2 flux/flow 1.07 flux/flow 1.07 flux/flow PCD 4-to-2 4 RCP 100 %FP 2/0 or 0/2 PPM > 0 %RTP flux/flow 1.07 flux/flow PCD 1/1 PPM flux > 55 %RTP 2/0, 0/2, or 1/1 flux/flow 1.07 4-to-0 4 RCP 100 %FP 0/0 PPM flux > 0 %RTP PPM flux > 0 %RTP PPM PCD flux/flow 1.07 flux/flow 1.07 3-to-2 3 RCP 75 %FP 2/0 or 0/2 PPM flux > 0 %RTP flux/flow 1.07 flux/flow PCD 1/1 PPM flux > 55 %RTP 2/0, 0/2, or 1/1 flux/flow 1.07 Locked 4 RCP 100 %FP 2/1, 1/2 flux/flow 1.07 flux/flow 1.07 flux/flow Rotor
Framatome Inc. ANP-4048NP Revision 1 ANO-1 TS 3.4.4 LAR Support - RPS Pump-to-Power Monitor Setpoint Summary Technical Report Page 5-1 5.0 STEADY-STATE PROTECTION As previously described, the PPM trip does not place any requirement on the allowed power level for 4- or 3-RCP operation, which allows the RPS high flux and flux/flow trips to maintain the required protection. For a scenario where 2-RCP operation may occur for a short period of time, as defined by TS 3.4.4 LCO, both the RPS flux/flow and PPM (1/1) setpoints provide overpower protection.
5.1 Maximum Real Power - RPS Flux/Flow Setpoint The maximum allowable power levels associated with the RPS power/imbalance/flow (P/I/F) trip are calculated as part of the P/I/F Protective Limits provided in the core operating limits report (COLR) each cycle. The maximum power levels are determined using the equation for partial-pump scaling of the protective limits as provided in Reference (3, Section 7.3.3.2).
[
] The current ANO-1 P/I/F maximum allowable power level for 2-RCP operation is then calculated as:
Framatome Inc. ANP-4048NP Revision 1 ANO-1 TS 3.4.4 LAR Support - RPS Pump-to-Power Monitor Setpoint Summary Technical Report Page 5-2 The combination of power and flow in the steady-state DNB analysis bounds the ratio of power to flow that would be calculated for higher best-estimate 2-RCP flows. For example, [ ] produces a real power of [ ] and a corresponding power/flow ratio of [ ] which is bounded by the more conservative power/flow ratio of [ ]
5.2 Maximum Real Power - RPS PPM (1/1) Setpoint The maximum real power associated with the PPM setpoint for 2-RCP (1/1) operation can be calculated similar to the RPS high flux trip (nuclear overpower) as described in Reference (3, Section 7.4.3):
Framatome Inc. ANP-4048NP Revision 1 ANO-1 TS 3.4.4 LAR Support - RPS Pump-to-Power Monitor Setpoint Summary Technical Report Page 5-3
[
]
The ANO-1 PPM maximum real power for 2-RCP (1/1) operation can then be calculated as:
The power of 62.833 %FP represents the maximum real power that could be achieved under 2-RCP (1/1) flow conditions if the RPS PPM (1/1) AV setpoint of 55 %RTP is reached.
Framatome Inc. ANP-4048NP Revision 1 ANO-1 TS 3.4.4 LAR Support - RPS Pump-to-Power Monitor Setpoint Summary Technical Report Page 6-1 6.0 2 RCP (1/1) STEADY-STATE DNB ANALYSES The steady-state DNB analyses protect overpower conditions based on the steady-state minimum thermal design flow for different RCP combinations. ANO-1 steady-state DNB analyses for 2-RCP minimum DNB flow conditions have previously considered a maximum real power level of 64.2 %FP for the current Mk-B-HTP fuel design. This power and flow combination bounds the real power and flow combinations that could exist if either the flux/flow or PPM (1/1) trip setpoints were reached during the TS 3.4.4 LCO time frame.
To provide additional support for the TS PPM (1/1) AV setpoint of 55 %RTP, a 2-RCP steady-state DNB analysis was performed with the LYNXT code (Reference 5) at a real core power level of 70 %FP. The resulting minimum DNBR was calculated to be greater than the applicable DNBR limit for the Mk-BHTP fuel as established using the SCD methodology (Reference 4). The acceptable DNB result for this case demonstrates the steady-state DNB protection as well as the transient DNB protection for the 4-to-2 PCD and 3-to-2 PCD events as discussed in Section 3.0. The maximum allowable power can then be used with Equation 3-3 to show the margin available in the current TS PPM (1/1) AV setpoint.
This result shows that the ANO-1 TS PPM (1/1) AV could be as high as 62.1 %RTP and still provide the required steady-state and transient protection.
Framatome Inc. ANP-4048NP Revision 1 ANO-1 TS 3.4.4 LAR Support - RPS Pump-to-Power Monitor Setpoint Summary Technical Report Page 7-1 7.0
SUMMARY
AND CONCLUSION Entergy is submitting a LAR to remove TS 3.4.4 Condition A, which will then require the plant to shut down within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> if only 2 RCPs are operating (1/1). The change to TS 3.4.4 will also allow removal of the 2-RCP Safety Limits, LCO limits, and RPS P/I/F setpoints and protective limits from the COLR. To support these changes, Entergy requested that an evaluation be performed to develop the basis for the current RPS RCP-to-power AV setpoint of 55 %RTP for 1/1 operation and demonstrate that it remains applicable considering the proposed LAR.
The review of the historical basis of the PPM trip showed that the original ANO-1 Technical Specifications were based on the RPS flux/flow setpoint as the primary protection and the PPM setpoint as the secondary protection for the LOCF events. The goal was to allow continued 2-RCP (1/1) operation following a 4-to-2 PCD or a 3-to-2 PCD if the flux/flow setpoint was not reached. However, with the transition to the NUREG-1430 Standard TS, the PPM setpoint became the primary protection and the flux/flow setpoint became the secondary. Consequently, the TS PPM (1/1) AV for all the B&W plants was conservatively set at 55 %RTP. This precluded the need for future plant-specific or cycle-specific PPM setpoint calculations.
To provide additional support for the current TS PPM (1/1) AV of 55 %RTP, a 2-RCP steady-state DNB analysis was performed at a real core power level of 70 %FP. The acceptable DNB result for this case demonstrates both steady-state overpower protection as well as transient protection for the 4-to-2 PCD and 3-to-2 PCD LOCF events. Additionally, a maximum allowable power of 70 %FP can support a PPM (1/1)
AV setpoint as high as 62.1 %RTP, which is conservatively bounded by the current PPM (1/1) AV of 55 %RTP by more than 7 %FP margin. Consequently, it is concluded that the ANO-1 PPM (1/1) AV has sufficient basis and analytical support to remain at 55 %RTP following the proposed TS 3.4.4 LAR changes that will further restrict 2-RCP operation and remove associated 2-RCP limits and setpoints from the COLR.
Framatome Inc. ANP-4048NP Revision 1 ANO-1 TS 3.4.4 LAR Support - RPS Pump-to-Power Monitor Setpoint Summary Technical Report Page 8-1
8.0 REFERENCES
- 1. Arkansas Nuclear One, Unit 1, Technical Specifications and Bases, Docket No 50-313, Amendment 274, Bases Revision 76
- 2. NUREG-1430, Revision 5, Standard Technical Specifications for Babcock
& Wilcox Plants
- 3. 43-10179PA-009, Safety Criteria and Methodology for Acceptable Cycle Reload Analyses, BAW-10179P-A, Rev. 9, November 2017
- 4. 43-10187PA-00, Statistical Core Design for B&W Designed 177 FA Plants, BAW-10187P-A, March 1994
- 5. 43-10156A-01, LYNXT - Core Transient Thermal-Hydraulic Program,,
BAW-10156-A, Rev. 1, March 1991
ENCLOSURE 3 1CAN062304 Affidavit (3 Pages)
AFFIDAVIT
- 1. My name is Morris Byram. I am Manager, Licensing & Regulatory Affairs for Framatome Inc. (Framatome) and as such I am authorized to execute this Affidavit.
- 2. I am familiar with the criteria applied by Framatome to determine whether certain Framatome information is proprietary. I am familiar with the policies established by Framatome to ensure the proper application of these criteria.
- 3. I am familiar with the Framatome information contained in Framatome technical report ANP-4048P, entitled ANO-1 TS 3.4.4 LAR Support - RPS Pump-to-Power Monitor Setpoint Summary, and referred to herein as Document. Information contained in this Document has been classified by Framatome as proprietary in accordance with the policies established by Framatome for the control and protection of proprietary and confidential information.
- 4. This Document contains information of a proprietary and confidential nature and is of the type customarily held in confidence by Framatome and not made available to the public. Based on my experience, I am aware that other companies regard information of the kind contained in this Document as proprietary and confidential.
- 5. This Document has been made available to the U.S. Nuclear Regulatory Commission in confidence with the request that the information contained in this Document be withheld from public disclosure. The request for withholding of proprietary information is made in accordance with 10 CFR 2.390. The information for which withholding from disclosure is requested qualifies under 10 CFR 2.390(a)(4) Trade secrets and commercial or financial information.
- 6. The following criteria are customarily applied by Framatome to determine whether information should be classified as proprietary:
(a) The information reveals details of Framatomes research and development plans and programs or their results.
(b) Use of the information by a competitor would permit the competitor to significantly reduce its expenditures, in time or resources, to design, produce, or market a similar product or service.
(c) The information includes test data or analytical techniques concerning a process, methodology, or component, the application of which results in a competitive advantage for Framatome.
(d) The information reveals certain distinguishing aspects of a process, methodology, or component, the exclusive use of which provides a competitive advantage for Framatome in product optimization or marketability.
(e) The information is vital to a competitive advantage held by Framatome, would be helpful to competitors to Framatome, and would likely cause substantial harm to the competitive position of Framatome.
The information in this Document is considered proprietary for the reasons set forth in paragraph 6(c) and 6(d) above.
- 7. In accordance with Framatomes policies governing the protection and control of information, proprietary information contained in this Document has been made available, on a limited basis, to others outside Framatome only as required and under suitable agreement providing for nondisclosure and limited use of the information.
- 8. Framatome policy requires that proprietary information be kept in a secured file or area and distributed on a need-to-know basis.
- 9. The foregoing statements are true and correct to the best of my knowledge, information, and belief.
I declare under penalty of perjury that the foregoing is true and correct.
Executed on: (6/20/2023)
Digitally signed by BYRAM BYRAM Morris Morris Date: 2023.06.20 14:01:25 -07'00' (NAME) morris.byram@framatome.com