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Category:Letter type:L
MONTHYEARL-2024-122, Core Operating Limits Report2024-08-12012 August 2024 Core Operating Limits Report L-2024-106, Fifth and Sixth 10-Year Inservice Testing Interval Relief Request No. VR-022024-08-12012 August 2024 Fifth and Sixth 10-Year Inservice Testing Interval Relief Request No. VR-02 L-2024-137, Second Supplement to NextEra Energy Seabrook, LLC, Letter L-2024-108, “Response to Request for Additional Information (RAI) Regarding One Time Extension to Technical Specification 3.8.1.1.a, Allowed Outage Time with One Independent .2024-08-12012 August 2024 Second Supplement to NextEra Energy Seabrook, LLC, Letter L-2024-108, “Response to Request for Additional Information (RAI) Regarding One Time Extension to Technical Specification 3.8.1.1.a, Allowed Outage Time with One Independent . L-2024-131, Response to Request for Additional Information Regarding License Amendment Request 300, Modify Containment Average Air Temperature Requirements2024-08-0909 August 2024 Response to Request for Additional Information Regarding License Amendment Request 300, Modify Containment Average Air Temperature Requirements L-2024-127, Supplement to Letter L-2024-108, “Response to Request for Additional Information (RAI) Regarding One Time Extension to Technical Specification 3.8.1.1.a, Allowed Outage Time with One Independent2024-08-0505 August 2024 Supplement to Letter L-2024-108, “Response to Request for Additional Information (RAI) Regarding One Time Extension to Technical Specification 3.8.1.1.a, Allowed Outage Time with One Independent L-2024-089, Notification of Deviation from Pressurized Water Owners Group (PWROG) Report WCAP- 17 451-P. Revision 1. Reactor Internals Guide Tube Wear - Westinghouse Domestic Fleet Operational Projections2024-07-25025 July 2024 Notification of Deviation from Pressurized Water Owners Group (PWROG) Report WCAP- 17 451-P. Revision 1. Reactor Internals Guide Tube Wear - Westinghouse Domestic Fleet Operational Projections L-2024-113, License Amendment Request 294, Application to Revise Technical Specifications to Adopt TSTF- 577, Revised Frequencies for Steam Generator Tube Inspections2024-07-24024 July 2024 License Amendment Request 294, Application to Revise Technical Specifications to Adopt TSTF- 577, Revised Frequencies for Steam Generator Tube Inspections L-2024-125, Notice of Intent to Provide Supplemental Information to License Amendment Request to Adopt Common Emergency Plan with Site-Specific Annexes2024-07-24024 July 2024 Notice of Intent to Provide Supplemental Information to License Amendment Request to Adopt Common Emergency Plan with Site-Specific Annexes L-2024-108, Response to Request for Additional Information (RAI) Regarding One Time Extension to Technical Specification 3.8.1.1.a, Allowed Outage Time with One Independent Circuit Between the Offsite.2024-07-16016 July 2024 Response to Request for Additional Information (RAI) Regarding One Time Extension to Technical Specification 3.8.1.1.a, Allowed Outage Time with One Independent Circuit Between the Offsite. L-2024-116, Preparation and Scheduling of Operator Licensing Examinations2024-07-11011 July 2024 Preparation and Scheduling of Operator Licensing Examinations L-2024-114, Quality Assurance Topical Report (FPL-1 Revision 31 Annual Submittal2024-07-10010 July 2024 Quality Assurance Topical Report (FPL-1 Revision 31 Annual Submittal L-2024-105, License Amendment Request 300, Modify Containment Average Air Temperature Requirements2024-06-26026 June 2024 License Amendment Request 300, Modify Containment Average Air Temperature Requirements L-2024-107, Schedule for Subsequent License Renewal Environmental Review2024-06-25025 June 2024 Schedule for Subsequent License Renewal Environmental Review L-2024-102, Official Service List Update2024-06-19019 June 2024 Official Service List Update L-2024-100, Withdrawal of License Amendment Request 274, Reactor Protection System, Engineered Safety Features Actuation System, and Nuclear Instrumentation System Replacement Project2024-06-19019 June 2024 Withdrawal of License Amendment Request 274, Reactor Protection System, Engineered Safety Features Actuation System, and Nuclear Instrumentation System Replacement Project L-2024-098, Preparation and Scheduling of Operator Licensing Examinations2024-06-12012 June 2024 Preparation and Scheduling of Operator Licensing Examinations L-2024-093, Steam Generator Divider Plate Assemblies Bounding Analysis Evaluation for Aging Management Commitment 14 Revision2024-06-10010 June 2024 Steam Generator Divider Plate Assemblies Bounding Analysis Evaluation for Aging Management Commitment 14 Revision L-2024-084, Relief Request 4A-01, Rev 1 - Revision to Relief Request for Examination of Control Rod Drive Mechanism (Rod) Housing H-4 Canopy Seal Weld2024-05-30030 May 2024 Relief Request 4A-01, Rev 1 - Revision to Relief Request for Examination of Control Rod Drive Mechanism (Rod) Housing H-4 Canopy Seal Weld L-2024-076, Reply to Notice of Violation; NOV 05000250, 05000251/2024010-052024-05-29029 May 2024 Reply to Notice of Violation; NOV 05000250, 05000251/2024010-05 L-2024-082, 2023 Annual Radiological Environmental Operating Report2024-05-15015 May 2024 2023 Annual Radiological Environmental Operating Report L-2024-061, NextEra Energy Seabrook, LLC, License Amendment Request - One Time Extension to Technical Specification 3.8.1.1.a, Allowed Outage Time with One Independent Circuit Between the Offsite Transmission Network and the Onsite Class 1E Distribut2024-05-10010 May 2024 NextEra Energy Seabrook, LLC, License Amendment Request - One Time Extension to Technical Specification 3.8.1.1.a, Allowed Outage Time with One Independent Circuit Between the Offsite Transmission Network and the Onsite Class 1E Distributio L-2024-060, 10 CFR 50.59(d)(2) Evaluation and 10 CFR 50.71(e)(2) Technical Specification Bases Summaries Report2024-05-0909 May 2024 10 CFR 50.59(d)(2) Evaluation and 10 CFR 50.71(e)(2) Technical Specification Bases Summaries Report L-2024-073, Cycle 34 Core Operating Limits Report2024-05-0101 May 2024 Cycle 34 Core Operating Limits Report L-2024-072, Cycle 33 Core Operating Limits Report2024-05-0101 May 2024 Cycle 33 Core Operating Limits Report L-2024-078, 2023 Annual Radioactive Effluent Release Report2024-05-0101 May 2024 2023 Annual Radioactive Effluent Release Report L-2024-077, 2023 Annual Radiological Environmental Operating Report2024-04-30030 April 2024 2023 Annual Radiological Environmental Operating Report L-2024-048, Divider Plate Assemblies Bounding Analysis Evaluation for Subsequent License Renewal Commitment Revision2024-04-30030 April 2024 Divider Plate Assemblies Bounding Analysis Evaluation for Subsequent License Renewal Commitment Revision L-2024-067, Annual Monitoring Report2024-04-26026 April 2024 Annual Monitoring Report L-2024-069, Radiological Emergency Plan Revision 762024-04-22022 April 2024 Radiological Emergency Plan Revision 76 L-2024-066, Sixth 10-Year Inservice Testing Interval Relief Request No. PR-022024-04-17017 April 2024 Sixth 10-Year Inservice Testing Interval Relief Request No. PR-02 L-2024-013, Submittal of Periodic Reports2024-03-28028 March 2024 Submittal of Periodic Reports L-2024-040, Notification of Deviation from Pressurized Water Owners Group (PWROG) Report WCAP-17451-P, Revision 1, Reactor Internals Guide Tube Wear - Westinghouse Domestic Fleet Operational Projections2024-03-28028 March 2024 Notification of Deviation from Pressurized Water Owners Group (PWROG) Report WCAP-17451-P, Revision 1, Reactor Internals Guide Tube Wear - Westinghouse Domestic Fleet Operational Projections L-2024-047, Proposed Use of a Subsequent ASME Code Edition and Addenda2024-03-28028 March 2024 Proposed Use of a Subsequent ASME Code Edition and Addenda L-2024-030, Supplement to Relief Request CISl-03-01 for Relief Concerning Containment Unbonded Post-Tensioning System Inservice Inspection Requirements2024-03-27027 March 2024 Supplement to Relief Request CISl-03-01 for Relief Concerning Containment Unbonded Post-Tensioning System Inservice Inspection Requirements L-2024-043, Revised Reactor Vessel Materials Surveillance Capsule Withdrawal Schedules2024-03-25025 March 2024 Revised Reactor Vessel Materials Surveillance Capsule Withdrawal Schedules L-2024-044, Revised Steam Generator Tube Inspection Reports2024-03-19019 March 2024 Revised Steam Generator Tube Inspection Reports L-2024-011, 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 Applications2024-03-13013 March 2024 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-2024-038, to Seabrook Emergency License Amendment Request - One Time Extension to Technical Specifications (TS) 3/4.8.1 Action a.3 Allowed Outage Time for an Inoperable Offsite Source2024-03-0808 March 2024 to Seabrook Emergency License Amendment Request - One Time Extension to Technical Specifications (TS) 3/4.8.1 Action a.3 Allowed Outage Time for an Inoperable Offsite Source L-2024-037, to Seabrook Emergency License Amendment Request - One Time Extension to Technical Specifications (TS) 3/4.8.1 Action a.3 Allowed Outage Time for an Inoperable Offsite Source2024-03-0606 March 2024 to Seabrook Emergency License Amendment Request - One Time Extension to Technical Specifications (TS) 3/4.8.1 Action a.3 Allowed Outage Time for an Inoperable Offsite Source L-2024-035, Supplement to Seabrook Emergency License Amendment Request - One Time Extension to Technical Specifications (TS) 3/4.8.1 Action a.3 Allowed Outage Time for an Inoperable Offsite Source2024-03-0505 March 2024 Supplement to Seabrook Emergency License Amendment Request - One Time Extension to Technical Specifications (TS) 3/4.8.1 Action a.3 Allowed Outage Time for an Inoperable Offsite Source L-2024-032, Emergency License Amendment Request- One Time Extension to Technical Specifications (TS) 3/4.8.1 Action a.3 Allowed Outage Time for an Inoperable Offsite Source2024-03-0404 March 2024 Emergency License Amendment Request- One Time Extension to Technical Specifications (TS) 3/4.8.1 Action a.3 Allowed Outage Time for an Inoperable Offsite Source L-2024-033, Request for Enforcement Discretion - Technical Specification (TS) 3/4.8.1.1 Ac. Sources Required Action Completion Time to Replace ED-X-3-B2024-03-0404 March 2024 Request for Enforcement Discretion - Technical Specification (TS) 3/4.8.1.1 Ac. Sources Required Action Completion Time to Replace ED-X-3-B L-2024-014, Turkey Points, Units 3 and 4 - 2023 Annual Radioactive Effluent Release Report2024-02-29029 February 2024 Turkey Points, Units 3 and 4 - 2023 Annual Radioactive Effluent Release Report L-2024-025, Notification of Improved Standard Technical Specifications (ITS) Implementation2024-02-22022 February 2024 Notification of Improved Standard Technical Specifications (ITS) Implementation L-2024-020, Refueling Outage Owners Activity Report (OAR-1) Unit 1 for Inservice Inspections2024-02-22022 February 2024 Refueling Outage Owners Activity Report (OAR-1) Unit 1 for Inservice Inspections L-2024-016, Radiological Emergency Plan (Ssrep), Revision 822024-02-13013 February 2024 Radiological Emergency Plan (Ssrep), Revision 82 L-2024-008, Supplement to 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 Cycles2024-02-0909 February 2024 Supplement to 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-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) 2024-08-09
[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
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Text
L-2019-058 MAR 2 52019
- 10 CFR 50.46 ATIN: Document Control Desk U. S. Nuclear Regulatory Commission Washington, DC 20555 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 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 10CFR50.46 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 Net PCT Effect Absolute PCT Effect Unit 1/Unit 2 Unit 1/Unit 2 Prior 10 CFR 50.46 Changes or Error Corrections- up to
+210°F/+248°F 210°F/340°F Year 2018( 1)
Prior 10 CFR 50 .46 Changes or Error Gorrections - Year None None 2019 New 10 CFR 50.46 Changes or Error Corrections- Year 2019
/ Error in Vapor Temperature Resetting Logic 0°F/0°F 0°F/0°F Sum of 10 CFR 50.46 Changes or Error Corrections +210°F/+248°F 210°F/340°F The sum of the PCTfrom the most recent analysis using an )
acceptable evaluation model and the estimates ofPCT* 2185°F/2058°F < 2200 °F impact for changes and errors identified since this analysis Error in Vapor Temperature Resetting Logic 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)
Net PCT . Absolute PCT Effect Effect Prior 10 CPR 50.46 Changes or Error Corrections - up to
-28 °P 80 °P year 2018 (Reference 2)
Prior 10 CPR 50.46 Changes or Errors Corrections-year None 2019 New 10 CPR 50.46 Changes or Errors Corrections-year 2019 I Error in Vapor Temperature Resetting Logic 0 op 0 op Sum of 10 CPR 50.46 Changes or Errors Corrections -28 °P 80 °P The sum of the PCT from the most recent analysis using an acceptable evaluation model and the estimates of PCT impact 2124 °F < 2200 °F for changes and errors identified since this analysis Error in Vapor Temperature Resetting Logic 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)
Net PCT Absolute PCT Effect Effect Prior 10 CPR 50.46 Changes or Error Corrections - up to 155 °P 155 °P year 2018 (Reference 2)
Prior 10 CPR 50 .46 Changes or Errors Corrections - year None 2019 New 10 CPR 50.46 Changes or Errors Corrections-year 2019 J Error in Vapor Temperature Resetting Logic 0 op 0 op Sum of 10 CPR 50.46 Changes or Errors Corrections 155 °P 155 °P The sum of the PCTfrom the most recent analysis using an acceptable evaluation model and the estimates ofPCT impact 1939 °F < 2200 °F for changes and errors identified since this a,nalysis Error in Vapor Temperature Resetting Logic 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.