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| * This letter contains no new or revised regulatory commitments. | | * This letter contains no new or revised regulatory commitments. |
| Florida Power & Light Company 700 Universe Boulevard, Juno Beach, FL 33408 L-2019-058 Page 2 of2 Should you have any questions regarding this report, please contact Mr. Steve Catron, Fleet Licensing Manager, at (561) 304-6206. | | Florida Power & Light Company 700 Universe Boulevard, Juno Beach, FL 33408 L-2019-058 Page 2 of2 Should you have any questions regarding this report, please contact Mr. Steve Catron, Fleet Licensing Manager, at (561) 304-6206. |
| Very truly yours, /v~71-P--William Parks Director, Nuclear Licensing and Regulatory Compliance Florida Power & Light Company Attachments (3) cc: USNRC Regional Administrator, Region I USNRC Regional Administrator, Region II USNRC Regional Administrator, Region III USNRC Project Manager, Point Beach Nuclear Plant USNRC Project Manager, Turkey Point Nuclear Plant USNRC Project Manager, Seabrook Station USNRC Senior Resident Inspector, Point Beach Nuclear Plant USNRC Senior Resident Inspector, Turkey Point Nuclear Plant USNRC Senior Resident Inspector, Seabrook Station | | Very truly yours, /v~71-P--William Parks Director, Nuclear Licensing and Regulatory Compliance Florida Power & Light Company Attachments (3) cc: USNRC Regional Administrator, Region I USNRC Regional Administrator, Region II USNRC Regional Administrator, Region III USNRC Project Manager, Point Beach Nuclear Plant USNRC Project Manager, Turkey Point Nuclear Plant USNRC Project Manager, Seabrook Station USNRC Senior Resident Inspector, Point Beach Nuclear Plant USNRC Senior Resident Inspector, Turkey Point Nuclear Plant USNRC Senior Resident Inspector, Seabrook Station |
| (' ATTACHMENT 1 NextEra Energy Point Beach Units 1 and 2 L-2019-058 Attachment 1 Page 1 ofl Point Beach Units 1 and 2 Large Break LOCA PCT 30-Day Report Evaluation Methodology: | | (' ATTACHMENT 1 NextEra Energy Point Beach Units 1 and 2 L-2019-058 Attachment 1 Page 1 ofl Point Beach Units 1 and 2 Large Break LOCA PCT 30-Day Report Evaluation Methodology: |
| Westinghouse, "Realistic Large-Break LOCA Evaluation Methodology Using the Automated Statistical Treatment of Uncertainty Method (ASTRUM)," WCAP-16009-P-A, January 2005. Westingti.ouse, "Application of Best Estimate Large Break LOGA Methodology to Westinghouse PWRs with Upper Plenum Injection," WCAP-14449-P-A Revision 1, October 1999. Evaluation Model PCT: 1975°F/1810°F Prior 10 CFR 50.46 Changes or Error Corrections-up to Year 2018(1) Prior 10 CFR 5 0 .46 Changes or Error Gorrections | | Westinghouse, "Realistic Large-Break LOCA Evaluation Methodology Using the Automated Statistical Treatment of Uncertainty Method (ASTRUM)," WCAP-16009-P-A, January 2005. Westingti.ouse, "Application of Best Estimate Large Break LOGA Methodology to Westinghouse PWRs with Upper Plenum Injection," WCAP-14449-P-A Revision 1, October 1999. Evaluation Model PCT: 1975°F/1810°F Prior 10 CFR 50.46 Changes or Error Corrections-up to Year 2018(1) Prior 10 CFR 5 0 .46 Changes or Error Gorrections |
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| : 1. L-2019-057, W. Parks (NextEra Energy) to U.S. Nuclear Regulatory Commission, "10CFR50.46 Annual Reporting of Changes to, or Errors in Emergency Core Cooling System Models or Applications," March 19, 2019. | | : 1. L-2019-057, W. Parks (NextEra Energy) to U.S. Nuclear Regulatory Commission, "10CFR50.46 Annual Reporting of Changes to, or Errors in Emergency Core Cooling System Models or Applications," March 19, 2019. |
| ATTACHMENT 2 Florida Power & Light Company Turkey Point Units 3 and 4 L-2019-058 Attachment 2 Page 1 of1 Turkey Point Units 3 & 4 Large Break LOCA PCT 30-Day Report . ***""' Evaluation Methodology: | | ATTACHMENT 2 Florida Power & Light Company Turkey Point Units 3 and 4 L-2019-058 Attachment 2 Page 1 of1 Turkey Point Units 3 & 4 Large Break LOCA PCT 30-Day Report . ***""' Evaluation Methodology: |
| Westinghouse, "Realistic Large-Break LOCA Evaluation Methodology Using the Automated Statistical Treatment Of Uncertainty Method (ASTRUM)," WCAP-16009-P-A, Revision 0, January 2005 Evaluation Model PCT: 2152 °F (Reference | | Westinghouse, "Realistic Large-Break LOCA Evaluation Methodology Using the Automated Statistical Treatment Of Uncertainty Method (ASTRUM)," WCAP-16009-P-A, Revision 0, January 2005 Evaluation Model PCT: 2152 °F (Reference |
| : 1) Prior 10 CPR 50.46 Changes or Error Corrections | | : 1) Prior 10 CPR 50.46 Changes or Error Corrections |
| -up to year 2018 (Reference | | -up to year 2018 (Reference |
| : 2) Prior 10 CPR 50.46 Changes or Errors Corrections-year 2019 New 10 CPR 50.46 Changes or Errors Corrections-year 2019 I Error in Vapor Temperature Resetting Logic Sum of 10 CPR 50.46 Changes or Errors Corrections The sum of the PCT from the most recent analysis using an acceptable evaluation model and the estimates of PCT impact for changes and errors identified since this analysis Error in Vapor Temperature Resetting Logic Net PCT . Absolute PCT Effect Effect -28 °P 80 °P None 0 op 0 op -28 °P 80 °P 2124 °F < 2200 °F When the vapor temperature is greater than the wall temperature, and several other conditions are met, the vapor temperature is reset to the saturation temperature for heat transfer calculations. | | : 2) Prior 10 CPR 50.46 Changes or Errors Corrections-year 2019 New 10 CPR 50.46 Changes or Errors Corrections-year 2019 I Error in Vapor Temperature Resetting Logic Sum of 10 CPR 50.46 Changes or Errors Corrections The sum of the PCT from the most recent analysis using an acceptable evaluation model and the estimates of PCT impact for changes and errors identified since this analysis Error in Vapor Temperature Resetting Logic Net PCT . Absolute PCT Effect Effect -28 °P 80 °P None 0 op 0 op -28 °P 80 °P 2124 °F < 2200 °F When the vapor temperature is greater than the wall temperature, and several other conditions are met, the vapor temperature is reset to the saturation temperature for heat transfer calculations. |
| It is discovered that the vapor temperature resetting logic results in an inconsistency between tqe conduction solution and the hydraulic solution, such that energy is not conserved between the two solutions. | | It is discovered that the vapor temperature resetting logic results in an inconsistency between tqe conduction solution and the hydraulic solution, such that energy is not conserved between the two solutions. |
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| : 1. L-2012-019, M. Kiley (NextEra Energy) to U.S. Nuclear Regulatory Commission, "Response to NRC Reactor Systems Branch Request for Additional Information Regarding Extended Power Uprate License Amendment Request No. 205 and Thermal Conductivity Degradation," January 16, 2012. 1. L-2019-057, W. Parks (NextEra Energy) to U.S. Nuclear Regulatory Commission, "I0CPR50.46 Annual Reporting of Changes to, or Errors in Emergency Core Cooling System Models or Applications," March 19, 2019. | | : 1. L-2012-019, M. Kiley (NextEra Energy) to U.S. Nuclear Regulatory Commission, "Response to NRC Reactor Systems Branch Request for Additional Information Regarding Extended Power Uprate License Amendment Request No. 205 and Thermal Conductivity Degradation," January 16, 2012. 1. L-2019-057, W. Parks (NextEra Energy) to U.S. Nuclear Regulatory Commission, "I0CPR50.46 Annual Reporting of Changes to, or Errors in Emergency Core Cooling System Models or Applications," March 19, 2019. |
| ATTACHMENT 3 N extEra Energy Seabrook Station L-2019-058 Attachment 3 Page 1 of1 Seabrook Unit 1 Large Break LOCA PCT 30-Day Report Evaluation Methodology: | | ATTACHMENT 3 N extEra Energy Seabrook Station L-2019-058 Attachment 3 Page 1 of1 Seabrook Unit 1 Large Break LOCA PCT 30-Day Report Evaluation Methodology: |
| Westinghouse, "Code Qualification Document for Best Estimate LOCA Analysis," WCAP-12945-P-A, March 1998 Evaluation Model PCT: 1784 °F (Reference | | Westinghouse, "Code Qualification Document for Best Estimate LOCA Analysis," WCAP-12945-P-A, March 1998 Evaluation Model PCT: 1784 °F (Reference |
| : 1) Prior 10 CPR 50.46 Changes or Error Corrections | | : 1) Prior 10 CPR 50.46 Changes or Error Corrections |
| -up to year 2018 (Reference | | -up to year 2018 (Reference |
| : 2) Prior 10 CPR 5 0 .46 Changes or Errors Corrections | | : 2) Prior 10 CPR 5 0 .46 Changes or Errors Corrections |
| -year 2019 New 10 CPR 50.46 Changes or Errors Corrections-year 2019 J Error in Vapor Temperature Resetting Logic Sum of 10 CPR 50.46 Changes or Errors Corrections The sum of the PCT from the most recent analysis using an acceptable evaluation model and the estimates of PCT impact for changes and errors identified since this a,nalysis Error in Vapor Temperature Resetting Logic Net PCT Absolute PCT Effect Effect 155 °P 155 °P None 0 op 0 op 155 °P 155 °P 1939 °F < 2200 °F When the vapor temperature is greater than the wall temperature, and several other conditions are met, the vapor temperature is reset to the saturation temperature for heat transfer calculations. | | -year 2019 New 10 CPR 50.46 Changes or Errors Corrections-year 2019 J Error in Vapor Temperature Resetting Logic Sum of 10 CPR 50.46 Changes or Errors Corrections The sum of the PCT from the most recent analysis using an acceptable evaluation model and the estimates of PCT impact for changes and errors identified since this a,nalysis Error in Vapor Temperature Resetting Logic Net PCT Absolute PCT Effect Effect 155 °P 155 °P None 0 op 0 op 155 °P 155 °P 1939 °F < 2200 °F When the vapor temperature is greater than the wall temperature, and several other conditions are met, the vapor temperature is reset to the saturation temperature for heat transfer calculations. |
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Category:Letter type:L
MONTHYEARL-2024-001, Relief Request CISl-03-01 for Relief Concerning Containment Unbonded Post-Tensioning System Inservice Inspection Requirements2024-01-26026 January 2024 Relief Request CISl-03-01 for Relief Concerning Containment Unbonded Post-Tensioning System Inservice Inspection Requirements L-2024-010, Point Units 3 and 4, Seabrook, Duane Arnold, and Point Beach Units 1 and 2, Nuclear Property Insurance - 10 CFR 50.54(w)(3)2024-01-25025 January 2024 Point Units 3 and 4, Seabrook, Duane Arnold, and Point Beach Units 1 and 2, Nuclear Property Insurance - 10 CFR 50.54(w)(3) L-2024-007, Inservice Inspection Program Owner'S Activity Report (OAR-1)2024-01-18018 January 2024 Inservice Inspection Program Owner'S Activity Report (OAR-1) L-2024-003, NextEra Energy Seabrook, LLC - 10 CFR 50.46 - Emergency Core Cooling System LBLOCA 30-Day Report2024-01-11011 January 2024 NextEra Energy Seabrook, LLC - 10 CFR 50.46 - Emergency Core Cooling System LBLOCA 30-Day Report L-2023-173, Quality Assurance Topical Report (FPL-1) Revision 30 Update2023-12-15015 December 2023 Quality Assurance Topical Report (FPL-1) Revision 30 Update L-2023-180, Submittal of Changes to the Technical Specification Bases2023-12-13013 December 2023 Submittal of Changes to the Technical Specification Bases L-2023-174, Subsequent License Renewal Application - Third Annual Update2023-12-13013 December 2023 Subsequent License Renewal Application - Third Annual Update L-2023-166, Turkey Points Units 3 and 4, Correction to the 2022 Annual Radioactive Effluent Release Report2023-12-0606 December 2023 Turkey Points Units 3 and 4, Correction to the 2022 Annual Radioactive Effluent Release Report L-2023-172, Supplement to Exemption Request Regarding Enhanced Weapons. Firearms Background Checks. and Security Event Notifications Final Rule2023-11-29029 November 2023 Supplement to Exemption Request Regarding Enhanced Weapons. Firearms Background Checks. and Security Event Notifications Final Rule L-2023-177, Supplement to Seabrook Exemption Request Regarding Enhanced Weapons, Firearms Background Checks, and Security Event Notifications Final Rule2023-11-29029 November 2023 Supplement to Seabrook Exemption Request Regarding Enhanced Weapons, Firearms Background Checks, and Security Event Notifications Final Rule L-2023-176, Supplement to Exemption Request Regarding Enhanced Weapons, Firearms Background Checks, and Security Event Notifications Final Rule2023-11-29029 November 2023 Supplement to Exemption Request Regarding Enhanced Weapons, Firearms Background Checks, and Security Event Notifications Final Rule L-2023-155, Supplement to Response to Request for Additional Information, Revised NextEra Common Emergency Plan, and Revised Site-Specific Emergency Plan Annexes Regarding License Amendment Request for Common Emergency Plan Consistent with NUREG-06542023-11-28028 November 2023 Supplement to Response to Request for Additional Information, Revised NextEra Common Emergency Plan, and Revised Site-Specific Emergency Plan Annexes Regarding License Amendment Request for Common Emergency Plan Consistent with NUREG-0654, L-2023-160, Part 73 Exemption Request Regarding Enhanced Weapons, Firearms, Background Checks, and Security Event Notifications Final Rule2023-11-16016 November 2023 Part 73 Exemption Request Regarding Enhanced Weapons, Firearms, Background Checks, and Security Event Notifications Final Rule L-2023-159, Part 3 Exemption Request Regarding Enhanced Weapons, Firearms Background Checks and Security Event Notifications Final Rule2023-11-16016 November 2023 Part 3 Exemption Request Regarding Enhanced Weapons, Firearms Background Checks and Security Event Notifications Final Rule L-2023-146, Part 73 Exemption Request Regarding Enhanced Weapons, Firearms Background Checks, and Security Event Notifications Final Rule2023-11-16016 November 2023 Part 73 Exemption Request Regarding Enhanced Weapons, Firearms Background Checks, and Security Event Notifications Final Rule L-2023-078, License Amendment Request 278, Incorporate Advanced Fuel Products, Extend Surveillance Intervals and 10 CFR 50.46 Exemption Request to Facilitate Transition to 24-Month Fuel Cycles2023-11-15015 November 2023 License Amendment Request 278, Incorporate Advanced Fuel Products, Extend Surveillance Intervals and 10 CFR 50.46 Exemption Request to Facilitate Transition to 24-Month Fuel Cycles L-2023-077, License Amendment Request 277 Updated Spent Fuel Pool Criticality Analysis2023-10-11011 October 2023 License Amendment Request 277 Updated Spent Fuel Pool Criticality Analysis L-2023-128, License Amendment Request to Revise TS 5.5.17, Pre-Stressed Concrete Containment Tendon Surveillance Program2023-09-19019 September 2023 License Amendment Request to Revise TS 5.5.17, Pre-Stressed Concrete Containment Tendon Surveillance Program L-2023-110, Response to Requests for Additional Information Regarding License Amendment Request No. 276, Revise Fire Protection Program in Support of Reactor Coolant Pump Seal Replacement Project2023-08-25025 August 2023 Response to Requests for Additional Information Regarding License Amendment Request No. 276, Revise Fire Protection Program in Support of Reactor Coolant Pump Seal Replacement Project L-2023-115, Inservice Inspection Program Owner'S Activity Report (OAR-1)2023-08-21021 August 2023 Inservice Inspection Program Owner'S Activity Report (OAR-1) L-2023-114, Proposed Turkey Point Units 6 and 7; Seabrook Station; Point Beach Units 1 and 2 - Official Service List Update2023-08-17017 August 2023 Proposed Turkey Point Units 6 and 7; Seabrook Station; Point Beach Units 1 and 2 - Official Service List Update L-2023-098, and Point Beach Units 1 and 2 - Response to Request for Additional Information Regarding License Amendment Request for Common Emergency Plan Consistent with NUREG-0654, Revision 22023-08-0707 August 2023 and Point Beach Units 1 and 2 - Response to Request for Additional Information Regarding License Amendment Request for Common Emergency Plan Consistent with NUREG-0654, Revision 2 L-2023-104, Preparation and Scheduling of Operator Licensing Examinations2023-08-0303 August 2023 Preparation and Scheduling of Operator Licensing Examinations L-2023-103, Inservice Inspection Examination Report2023-08-0303 August 2023 Inservice Inspection Examination Report L-2023-094, Response to Requests for Additional Information Regarding License Amendment Request No. 276, Revise Fire Protection Program in Support of Reactor Coolant Pump Seal Replacement Project2023-07-27027 July 2023 Response to Requests for Additional Information Regarding License Amendment Request No. 276, Revise Fire Protection Program in Support of Reactor Coolant Pump Seal Replacement Project L-2023-089, Refueling Outage Owner'S Activity Report (OAR-1) Unit 2 for Inservice Inspections2023-07-24024 July 2023 Refueling Outage Owner'S Activity Report (OAR-1) Unit 2 for Inservice Inspections L-2023-087, Florida Power & Light/Nextera Energy, Results of the Safety Culture Program Effectiveness Review, March 20, 2023 (ADAMS Accession No. ML22340A452)2023-06-29029 June 2023 Florida Power & Light/Nextera Energy, Results of the Safety Culture Program Effectiveness Review, March 20, 2023 (ADAMS Accession No. ML22340A452) L-2023-086, Request Temporary Suspension of Turkey Point License Amendment Request 274, Reactor Protection System, Engineered Safety Features Actuation System, and Nuclear Instrumentation.2023-06-28028 June 2023 Request Temporary Suspension of Turkey Point License Amendment Request 274, Reactor Protection System, Engineered Safety Features Actuation System, and Nuclear Instrumentation. L-2023-088, 10 CFR 50.55a Requests, Relief Requests I6-RR-1, I6-RR-2, and I6-RR-3 Sixth Ten-Year Inservice Inspection Program Interval2023-06-27027 June 2023 10 CFR 50.55a Requests, Relief Requests I6-RR-1, I6-RR-2, and I6-RR-3 Sixth Ten-Year Inservice Inspection Program Interval L-2023-075, Response to Request for Additional Information (RAI) Regarding Exemption Request, License Amendment Request and Revised Response in Support of a Risk-Informed Resolution of Generic Letter 2004-022023-06-0909 June 2023 Response to Request for Additional Information (RAI) Regarding Exemption Request, License Amendment Request and Revised Response in Support of a Risk-Informed Resolution of Generic Letter 2004-02 L-2023-074, Addendum to 2021 Decommissioning Funding Status Reports / Independent Spent Fuel Storage Installation Ctsfsi) Financial Assurance Update2023-06-0202 June 2023 Addendum to 2021 Decommissioning Funding Status Reports / Independent Spent Fuel Storage Installation Ctsfsi) Financial Assurance Update L-2023-073, Subsequent License Renewal Application, Second Annual Update Request for Additional Information Set 1 Response2023-06-0101 June 2023 Subsequent License Renewal Application, Second Annual Update Request for Additional Information Set 1 Response L-2023-069, Response to Requests for Additional Information Regarding License Amendment Request No. 276, Revise Fire Protection Program in Support of Reactor Coolant Pump Seal Replacement Project2023-05-31031 May 2023 Response to Requests for Additional Information Regarding License Amendment Request No. 276, Revise Fire Protection Program in Support of Reactor Coolant Pump Seal Replacement Project L-2023-072, Preparation and Scheduling of Operator Licensing Examinations2023-05-22022 May 2023 Preparation and Scheduling of Operator Licensing Examinations L-2023-071, NextEra Energy Quality Assurance Topical Report (FPL-1) Revision 29 and Florida Power and Light Company Quality Assurance Program Description for 10 CFR Part 52 Licenses (FPL-2) Revision 11, Annual Submittal2023-05-22022 May 2023 NextEra Energy Quality Assurance Topical Report (FPL-1) Revision 29 and Florida Power and Light Company Quality Assurance Program Description for 10 CFR Part 52 Licenses (FPL-2) Revision 11, Annual Submittal L-2023-061, 2022 Annual Radiological Environmental Operating Report2023-05-12012 May 2023 2022 Annual Radiological Environmental Operating Report L-2023-064, License Amendment Request 23-01, Revision 1, Update the Period of Applicability (Poa) for the Pressure-Temperature Limits (PTL) and Low Temperature Overpressure Protection (L Top) Curves2023-05-11011 May 2023 License Amendment Request 23-01, Revision 1, Update the Period of Applicability (Poa) for the Pressure-Temperature Limits (PTL) and Low Temperature Overpressure Protection (L Top) Curves L-2023-062, Cycle 33 Core Operating Limits Report2023-04-27027 April 2023 Cycle 33 Core Operating Limits Report L-2023-060, Radiological Emergency Plan, Revision 752023-04-26026 April 2023 Radiological Emergency Plan, Revision 75 L-2023-054, Submittal of Periodic Reports2023-04-12012 April 2023 Submittal of Periodic Reports L-2023-058, 2022 Annual Monitoring Report2023-04-10010 April 2023 2022 Annual Monitoring Report L-2023-053, Core Operating Limits Report for Cycle 232023-04-0505 April 2023 Core Operating Limits Report for Cycle 23 L-2023-049, Correction to U4R33 Steam Generator Tube Inspection Report2023-03-30030 March 2023 Correction to U4R33 Steam Generator Tube Inspection Report L-2023-021, Units, 1 and 2, Turkey Point, Units 3 and 4, Seabrook Station and Point Beach, Units 1 and 2 - Decommissioning Funding Status Reports / Independent Spent Fuel Storage Installation (ISFSI) Financial Assurance Update2023-03-28028 March 2023 Units, 1 and 2, Turkey Point, Units 3 and 4, Seabrook Station and Point Beach, Units 1 and 2 - Decommissioning Funding Status Reports / Independent Spent Fuel Storage Installation (ISFSI) Financial Assurance Update L-2023-028, and Point Beach Units 1 and 2, 10 CFR 50.46 Annual Reporting of Changes to, or Errors in Emergency Core Cooling System Models or Applications2023-03-27027 March 2023 and Point Beach Units 1 and 2, 10 CFR 50.46 Annual Reporting of Changes to, or Errors in Emergency Core Cooling System Models or Applications L-2023-048, Core Operating Limits Report (COLR) Cycle 40 (U2C40)2023-03-18018 March 2023 Core Operating Limits Report (COLR) Cycle 40 (U2C40) L-2023-012, License Amendment Request 23-01, Remove Period of Applicability (Poa) from Pressure Temperature Limits (PTL) and Low Temperature Overpressure Protection (L Top) Curves2023-03-15015 March 2023 License Amendment Request 23-01, Remove Period of Applicability (Poa) from Pressure Temperature Limits (PTL) and Low Temperature Overpressure Protection (L Top) Curves L-2023-025, Fleet Relief Request (Frr) 23-01, Proposed Alternative to ASME Section XI Authorizing Implementation of ASME Code Case N-752-12023-03-15015 March 2023 Fleet Relief Request (Frr) 23-01, Proposed Alternative to ASME Section XI Authorizing Implementation of ASME Code Case N-752-1 L-2023-029, and Point Beach Units 1 and 2 Nuclear Property Insurance - 10 CFR 50.54(w)(3)2023-03-10010 March 2023 and Point Beach Units 1 and 2 Nuclear Property Insurance - 10 CFR 50.54(w)(3) L-2023-023, Revised Steam Generator Tube Inspection Report2023-03-0303 March 2023 Revised Steam Generator Tube Inspection Report 2024-01-26
[Table view] Category:Licensee 30-Day Written Event Report
MONTHYEARL-2019-058, CFR 50.46 - Emergency Core Cooling System LBLOCA 30-Day Report2019-03-25025 March 2019 CFR 50.46 - Emergency Core Cooling System LBLOCA 30-Day Report L-2018-063, CFR 50.46 Annual Reporting and 30-day Notification of Changes or Errors in Emergency Core Cooling System Models or Applications2018-03-26026 March 2018 CFR 50.46 Annual Reporting and 30-day Notification of Changes or Errors in Emergency Core Cooling System Models or Applications L-2017-102, Submittal of 10 CFR 50.46 - Emergency Core Cooling 30-Day Report2017-06-0202 June 2017 Submittal of 10 CFR 50.46 - Emergency Core Cooling 30-Day Report L-2017-069, Special Report - Containment Tendon Corrosion Protection Medium Volume Reduction2017-04-17017 April 2017 Special Report - Containment Tendon Corrosion Protection Medium Volume Reduction NRC-2017-0012, Submitting 10 CFR 26.719(c) Report on Inaccurate Reporting of Laboratory Test Result2017-02-24024 February 2017 Submitting 10 CFR 26.719(c) Report on Inaccurate Reporting of Laboratory Test Result L-2016-124, 10 CFR 50.46 30-Day Report for Turkey Point, Seabrook, and Point Beach2016-06-17017 June 2016 10 CFR 50.46 30-Day Report for Turkey Point, Seabrook, and Point Beach NRC 2016-0015, Large Break Loss-of-Coolant Accident Margin Summary Sheet - 30-Day Report2016-04-0606 April 2016 Large Break Loss-of-Coolant Accident Margin Summary Sheet - 30-Day Report L-2016-065, Acceptance Criteria for Emergency Core Cooling Systems in Light Water Nuclear Power Reactors, - 30 Day Special Report2016-04-0505 April 2016 Acceptance Criteria for Emergency Core Cooling Systems in Light Water Nuclear Power Reactors, - 30 Day Special Report NRC 2014-0086, 10 CFR 50.46 30-Day Report2014-12-11011 December 2014 10 CFR 50.46 30-Day Report L-2014-359, 10 CFR 50.46 Acceptance Criteria for Emergency Core Cooling Systems in Light Water Nuclear Power Reactors, 30 Day Special Report2014-12-0303 December 2014 10 CFR 50.46 Acceptance Criteria for Emergency Core Cooling Systems in Light Water Nuclear Power Reactors, 30 Day Special Report NRC 2014-0029, CFR 50.46 Annual Report/30-Day Report2014-04-18018 April 2014 CFR 50.46 Annual Report/30-Day Report SBK-L-14073, Best Estimate Large Break Loss of Coolant Accident 10 CFR 50.46 30-Day Report2014-04-16016 April 2014 Best Estimate Large Break Loss of Coolant Accident 10 CFR 50.46 30-Day Report L-2014-077, CFR 50.46, Acceptance Criteria for Emergency Core Cooling Systems in Light Water Nuclear Power Reactors - 30 Day Special Report2014-03-24024 March 2014 CFR 50.46, Acceptance Criteria for Emergency Core Cooling Systems in Light Water Nuclear Power Reactors - 30 Day Special Report L-2014-065, Special Report - Accident Monitoring Instrumentation2014-02-25025 February 2014 Special Report - Accident Monitoring Instrumentation L-2014-037, Emergency Core Cooling Systems in Light Water Nuclear Power Reactors - 30 Day Special Report2014-02-18018 February 2014 Emergency Core Cooling Systems in Light Water Nuclear Power Reactors - 30 Day Special Report NRC 2014-0012, Large Break Loss-of-Coolant Accident Margin Summary Sheet - 30-Day Report2014-02-13013 February 2014 Large Break Loss-of-Coolant Accident Margin Summary Sheet - 30-Day Report NRC 2013-0081, Large Break Loss-of Coolant Accident Margin Summary Sheet- 30-Day Report2013-08-23023 August 2013 Large Break Loss-of Coolant Accident Margin Summary Sheet- 30-Day Report L-2013-184, Special Report - Accident Monitoring Instrumentation2013-05-30030 May 2013 Special Report - Accident Monitoring Instrumentation L-2013-117, Special Report - Accident Monitoring Instrumentation2013-03-29029 March 2013 Special Report - Accident Monitoring Instrumentation SBK-L-12264, Nuclear Fuel Pellet Thermal Conductivity Degradation Impact on Current Seabrook Be LBLOCA Analysis Using the 1996 Cqd Methodology 10 CFR 50.46 30-Day Report2012-12-21021 December 2012 Nuclear Fuel Pellet Thermal Conductivity Degradation Impact on Current Seabrook Be LBLOCA Analysis Using the 1996 Cqd Methodology 10 CFR 50.46 30-Day Report NRC 2012-0038, ECCS 30-Day Report for the Thermal Conductivity Degradation Impact on Large Break Loss of Coolant Accident Analyses with Astrum2012-05-30030 May 2012 ECCS 30-Day Report for the Thermal Conductivity Degradation Impact on Large Break Loss of Coolant Accident Analyses with Astrum SBK-L-11141, Submittal of Special Report Regarding Containment Enclosure Building Structural Integrity2011-07-11011 July 2011 Submittal of Special Report Regarding Containment Enclosure Building Structural Integrity L-2010-235, LER, Turkey Point, Units 3 and 4, Special Report - Accident Monitoring Instrumentation Inoperable2010-10-13013 October 2010 LER, Turkey Point, Units 3 and 4, Special Report - Accident Monitoring Instrumentation Inoperable L-2010-114, Special Report Re Inoperable Main Steam Lines High Range-Noble Gas Effluent Monitor Greater than 7 Days2010-06-14014 June 2010 Special Report Re Inoperable Main Steam Lines High Range-Noble Gas Effluent Monitor Greater than 7 Days L-2010-100, Accident Monitoring Instrumentation Special Report2010-05-14014 May 2010 Accident Monitoring Instrumentation Special Report NRC-2007-0055, Error Identified in ECCS Evaluation Model 30-Day Report Required by 10 CFR 50.462007-07-0303 July 2007 Error Identified in ECCS Evaluation Model 30-Day Report Required by 10 CFR 50.46 L-2005-096, CFR 50.46, Acceptance Criteria for Emergency Core Cooling Systems in Light Water Nuclear Power Reactors - Annual Report and 30-Day Report2005-04-27027 April 2005 CFR 50.46, Acceptance Criteria for Emergency Core Cooling Systems in Light Water Nuclear Power Reactors - Annual Report and 30-Day Report ML0508303282004-05-0606 May 2004 Event Notification 40728 for Point Beach NRC 2003-0042, Reporting of Fire Barriers Degraded for More than Seven Days2003-05-0909 May 2003 Reporting of Fire Barriers Degraded for More than Seven Days NRC 2002-0095, Reporting of Fire Barriers Degraded for More than Seven Days2002-10-22022 October 2002 Reporting of Fire Barriers Degraded for More than Seven Days NRC 2002-0083, Reporting of Fire Barriers Degraded for More than Seven Days2002-09-26026 September 2002 Reporting of Fire Barriers Degraded for More than Seven Days ML0307702772002-08-0707 August 2002 2-page Abnormal Occurrence Writeup for 11/29/2001 Event Potential Loss of All Auxiliary Feedwater at Point Beach Plant ML18227A5071977-07-29029 July 1977 Transmittal of Licensee Event Report 251-77-4, Steam Generator Tube Plugs ML18227A5081977-05-13013 May 1977 Transmittal of Licensee Event Report 251-77-2 Boric Acid Transfer Pumps 2019-03-25
[Table view] |
Text
/ A TIN: Document Control Desk U. S. Nuclear Regulatory Commission Washington, DC 20555 MAR 2 5 2019 Re: N extEra Energy Point Beach, ILC Point Beach Units 1 and 2, Docket Nos. 50-266, 50-301 Florida Power & Light Company Turkey Point Units 3 and 4, Docket Nos. 50-250, 50-251 N extEra Energy Seabrook, LLC Seabrook Station, Docket No. ,50-443 L-2019-058
- 10 CFR 50.46 10 CFR 50.46 -Emergency Core Cooling System LBLOCA 30-Day Report Pursuant to 10 CFR 50.46(a)(3)(ii), this letter contains the 30-day report for the Point Beach Nuclear Plant, Units 1 and 2, Turkey Point Nuclear Plant, Units 3 and 4, and* Seabrook Station for the emergency core cooling system analysis performed by Westinghouse Electric Company, LLC, in the respective attachments to this letter. One error was identified by Westinghouse that affects the Large-Break (LB) LOCA models. The LBLOCA error is related to an error in the vapor temperature resetting logic in the Best-Estimate LOCA evaluation model. The error has a negligible impact on the results of LBLOCA analyses, leading to an estimated peak cladding temperature (PCT) impact of 0°F. As the reported error is of a 0°F impact, the PCT continues to remain within the limits. However, as the cumulative PCT change already exceeds 50 °F for the LBLOCA analysis, a 30 day 1 OCFR50.46 report must be issued. Evaluations of each reported error have concluded that re-analysis was not required.
- This letter contains no new or revised regulatory commitments.
Florida Power & Light Company 700 Universe Boulevard, Juno Beach, FL 33408 L-2019-058 Page 2 of2 Should you have any questions regarding this report, please contact Mr. Steve Catron, Fleet Licensing Manager, at (561) 304-6206.
Very truly yours, /v~71-P--William Parks Director, Nuclear Licensing and Regulatory Compliance Florida Power & Light Company Attachments (3) cc: USNRC Regional Administrator, Region I USNRC Regional Administrator, Region II USNRC Regional Administrator, Region III USNRC Project Manager, Point Beach Nuclear Plant USNRC Project Manager, Turkey Point Nuclear Plant USNRC Project Manager, Seabrook Station USNRC Senior Resident Inspector, Point Beach Nuclear Plant USNRC Senior Resident Inspector, Turkey Point Nuclear Plant USNRC Senior Resident Inspector, Seabrook Station
(' ATTACHMENT 1 NextEra Energy Point Beach Units 1 and 2 L-2019-058 Attachment 1 Page 1 ofl Point Beach Units 1 and 2 Large Break LOCA PCT 30-Day Report Evaluation Methodology:
Westinghouse, "Realistic Large-Break LOCA Evaluation Methodology Using the Automated Statistical Treatment of Uncertainty Method (ASTRUM)," WCAP-16009-P-A, January 2005. Westingti.ouse, "Application of Best Estimate Large Break LOGA Methodology to Westinghouse PWRs with Upper Plenum Injection," WCAP-14449-P-A Revision 1, October 1999. Evaluation Model PCT: 1975°F/1810°F Prior 10 CFR 50.46 Changes or Error Corrections-up to Year 2018(1) Prior 10 CFR 5 0 .46 Changes or Error Gorrections
-Year 2019 New 10 CFR 50.46 Changes or Error Corrections-Year 2019 / Error in Vapor Temperature Resetting Logic Sum of 10 CFR 50.46 Changes or Error Corrections The sum of the PCT from the most recent analysis using an acceptable evaluation model and the estimates of PCT* impact for changes and errors identified since this analysis Error in Vapor Temperature Resetting Logic Net PCT Effect Absolute PCT Effect Unit 1/Unit 2 Unit 1/Unit 2 +210°F/+248°F 210°F/340°F None None 0°F/0°F 0°F/0°F +210°F/+248°F 210°F/340°F ) 2185°F/2058°F
< 2200 °F When the vapor temperature is greater than the wall temperature, and several other conditions are met, the vapor.temperature is reset to the saturation temperature for heat transfer calculations.
It is discovered that the vapor temperature resetting logic results in an inconsistency between the conduction solution and the hydraulic solution, such that energy is not conserved between the two solutions.
The correction of this error has a negligible impact on the results of LBLOCA analyses.
The PCT.impact of correcting this error is estimated to be 0°F.
Reference:
- 1. L-2019-057, W. Parks (NextEra Energy) to U.S. Nuclear Regulatory Commission, "10CFR50.46 Annual Reporting of Changes to, or Errors in Emergency Core Cooling System Models or Applications," March 19, 2019.
ATTACHMENT 2 Florida Power & Light Company Turkey Point Units 3 and 4 L-2019-058 Attachment 2 Page 1 of1 Turkey Point Units 3 & 4 Large Break LOCA PCT 30-Day Report . ***""' Evaluation Methodology:
Westinghouse, "Realistic Large-Break LOCA Evaluation Methodology Using the Automated Statistical Treatment Of Uncertainty Method (ASTRUM)," WCAP-16009-P-A, Revision 0, January 2005 Evaluation Model PCT: 2152 °F (Reference
- 1) Prior 10 CPR 50.46 Changes or Error Corrections
-up to year 2018 (Reference
- 2) Prior 10 CPR 50.46 Changes or Errors Corrections-year 2019 New 10 CPR 50.46 Changes or Errors Corrections-year 2019 I Error in Vapor Temperature Resetting Logic Sum of 10 CPR 50.46 Changes or Errors Corrections The sum of the PCT from the most recent analysis using an acceptable evaluation model and the estimates of PCT impact for changes and errors identified since this analysis Error in Vapor Temperature Resetting Logic Net PCT . Absolute PCT Effect Effect -28 °P 80 °P None 0 op 0 op -28 °P 80 °P 2124 °F < 2200 °F When the vapor temperature is greater than the wall temperature, and several other conditions are met, the vapor temperature is reset to the saturation temperature for heat transfer calculations.
It is discovered that the vapor temperature resetting logic results in an inconsistency between tqe conduction solution and the hydraulic solution, such that energy is not conserved between the two solutions.
The correction of this . error has a negligible impact on the results of LBLOCA analyses.
The PCT impact of correcting this error is estimated to be 0°P.
References:
- 1. L-2012-019, M. Kiley (NextEra Energy) to U.S. Nuclear Regulatory Commission, "Response to NRC Reactor Systems Branch Request for Additional Information Regarding Extended Power Uprate License Amendment Request No. 205 and Thermal Conductivity Degradation," January 16, 2012. 1. L-2019-057, W. Parks (NextEra Energy) to U.S. Nuclear Regulatory Commission, "I0CPR50.46 Annual Reporting of Changes to, or Errors in Emergency Core Cooling System Models or Applications," March 19, 2019.
ATTACHMENT 3 N extEra Energy Seabrook Station L-2019-058 Attachment 3 Page 1 of1 Seabrook Unit 1 Large Break LOCA PCT 30-Day Report Evaluation Methodology:
Westinghouse, "Code Qualification Document for Best Estimate LOCA Analysis," WCAP-12945-P-A, March 1998 Evaluation Model PCT: 1784 °F (Reference
- 1) Prior 10 CPR 50.46 Changes or Error Corrections
-up to year 2018 (Reference
- 2) Prior 10 CPR 5 0 .46 Changes or Errors Corrections
-year 2019 New 10 CPR 50.46 Changes or Errors Corrections-year 2019 J Error in Vapor Temperature Resetting Logic Sum of 10 CPR 50.46 Changes or Errors Corrections The sum of the PCT from the most recent analysis using an acceptable evaluation model and the estimates of PCT impact for changes and errors identified since this a,nalysis Error in Vapor Temperature Resetting Logic Net PCT Absolute PCT Effect Effect 155 °P 155 °P None 0 op 0 op 155 °P 155 °P 1939 °F < 2200 °F When the vapor temperature is greater than the wall temperature, and several other conditions are met, the vapor temperature is reset to the saturation temperature for heat transfer calculations.
It is discovered that the vapor temperature resetting logic results in an inconsistency between the conduction solution and the hydraulic solution, such that energy is not conserved between the two solutions.
The correction of this error has a negligible impact on the results ofLBLOCA analyses.
The PCT impact of correcting this error is estimated to be 0°P.
References:
- 1. NYN-04016, M. Warner (NextEra Energy) to U.S. Nuclear Regulatory Commission, "License Amendment Request 04-03, Application for Stretch Power Uprate," March 17, 2004. 2. L-2019-057, W. Parks (NextEra Energy) to U.S. Nuclear Regulatory Commission, "10CPR50.46 Annual Reporting of Changes to, or Errors in Emergency Core Cooling System Models or Applications," March 19, 2019.