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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-2024-089, 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-2024-125, 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-2024-114, 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-2024-102, 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-2024-076, 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-2024-060, 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-2024-048, 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-2024-069, 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-2024-047, 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-2024-040, 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-2024-013, Submittal of Periodic Reports2024-03-28028 March 2024 Submittal of Periodic Reports 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-2024-014, 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-2024-008, 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-2024-010, 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-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-2023-166, 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-2023-155, 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-2023-146, 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-2023-110, 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-2023-094, 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-2023-087, 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-2023-074, 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-2023-069, 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-2023-062, 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-2023-049, 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 2024-08-12
[Table view] Category:Report
MONTHYEARL-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-2024-048, 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 ML23333A0142023-11-27027 November 2023 Attachment F - Groundwater Tek Inc - Peer Review Study Final-1 ML23333A0172023-11-27027 November 2023 Attachment I - Chin - the Cooling Canal System at the FPL-Turkey-Point Power Station ML24012A0422023-11-16016 November 2023 FAQ 23-03 Turkey Point IE01 Proposed NRC Response ML24081A1612023-11-16016 November 2023 FAQ 23-03 Turkey Point IE01 Revision 1 ML23265A5512023-09-22022 September 2023 Enclosure 3: WCAP-18830-NP, Turkey Point Fuel Storage Criticality Analysis for 24 Month Cycles 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-2023-049, Correction to U4R33 Steam Generator Tube Inspection Report2023-03-30030 March 2023 Correction to U4R33 Steam Generator Tube Inspection Report 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-2023-010, Supplemental Information Regarding License Amendment Request 274, Reactor Protection System, Engineered Safety Features Actuation System, and Nuclear Instrumentation System Replacement Project - Submittal of RPS / ESFAS / Nis2023-02-10010 February 2023 Supplemental Information Regarding License Amendment Request 274, Reactor Protection System, Engineered Safety Features Actuation System, and Nuclear Instrumentation System Replacement Project - Submittal of RPS / ESFAS / Nis L-2022-168, and Point Beach Units 1 and 2 - 10 CFR 50.46 - Emergency Core Cooling System LBLOCA 30-Day Report2022-10-26026 October 2022 and Point Beach Units 1 and 2 - 10 CFR 50.46 - Emergency Core Cooling System LBLOCA 30-Day Report L-2022-110, License Amendment Request 276, Revise Fire Protection Program in Support of Reactor Coolant Pump Seal Replacement Project (Non-Proprietary)2022-08-26026 August 2022 License Amendment Request 276, Revise Fire Protection Program in Support of Reactor Coolant Pump Seal Replacement Project (Non-Proprietary) L-2022-142, Revised Diversity and Defense-In-Depth Evaluation (D3), Framatome Document No. 51-9324096-0042022-08-19019 August 2022 Revised Diversity and Defense-In-Depth Evaluation (D3), Framatome Document No. 51-9324096-004 L-2022-038, Emergency Response Data System (Eros) Changes2022-05-17017 May 2022 Emergency Response Data System (Eros) Changes L-2022-073, Diversity and Defense-In-Depth Evaluation (D3), Framatome Document No. 51-9324096-0042022-05-0303 May 2022 Diversity and Defense-In-Depth Evaluation (D3), Framatome Document No. 51-9324096-004 L-2022-022, Updated Conditions of Certification Report2022-02-14014 February 2022 Updated Conditions of Certification Report L-2021-182, Summary of Commitment Revisions, Emergency Diesel Generator Fuel Oil Storage Tank Cleaning Inspection Commitment Revisions2021-09-16016 September 2021 Summary of Commitment Revisions, Emergency Diesel Generator Fuel Oil Storage Tank Cleaning Inspection Commitment Revisions L-2021-069, Commitment Revision Summary Report for Reactor Vessel Head Leakage Detection System (Rvhlds)2021-06-24024 June 2021 Commitment Revision Summary Report for Reactor Vessel Head Leakage Detection System (Rvhlds) IR 05000250/20200112020-07-23023 July 2020 NRC Inspection Report 05000250-2020011 and 05000251-2020011 and Investigation Report 2-2019-011; and Apparent Violation Final L-2020-073, Fifth Ten-Year Inservice Inspection Interval Revised Relief Request No. 6 and Supplemental Information for Train B CCW Return Piping2020-04-13013 April 2020 Fifth Ten-Year Inservice Inspection Interval Revised Relief Request No. 6 and Supplemental Information for Train B CCW Return Piping ML20098F3412020-04-0707 April 2020 Exigent License Amendment Request 272, One-Time Extension of TS 6.8.4 Steam Generator Inspection Program - Response to Request for Additional Information L-2019-204, Request for Use and Approval of Vapor Infusion Technology 90-Day Trial - Notification2019-11-25025 November 2019 Request for Use and Approval of Vapor Infusion Technology 90-Day Trial - Notification L-2019-151, 10 CFR 50.46 - Emergency Core Cooling System LBLOCA 30-Day Report2019-08-0606 August 2019 10 CFR 50.46 - Emergency Core Cooling System LBLOCA 30-Day Report L-2019-010, Proposed Alternative for the Use of Encoded Phased Array Ultrasonic Examination Techniques in Lieu of Radiography for Ferritic and Austenitic Welds2019-03-19019 March 2019 Proposed Alternative for the Use of Encoded Phased Array Ultrasonic Examination Techniques in Lieu of Radiography for Ferritic and Austenitic Welds L-2019-054, Baffle-Former Bolts Predictive Evaluations2019-03-13013 March 2019 Baffle-Former Bolts Predictive Evaluations ML19072A1622019-03-0505 March 2019 National Park Service, Southeast Regional Office, Comments Dated March 5, 2019, on Turkey Point Nuclear Generating Units 3 and 4 Preliminary Draft SEIS Regarding Subsequent License Renewal ML18353A8352018-12-31031 December 2018 Biological Assessment for the Turkey Point Units 3 and 4 Proposed Subsequent License Renewal ML18299A1162018-10-15015 October 2018 Structural Integrity Associates Engineering Report No. 1700109.401P, Revision 7 - Redacted, Evaluation of Environmentally-Assisted Fatigue for Turkey Point Units 3 and 4 for Subsequent License Renewal ML18299A1182018-10-12012 October 2018 Structural Integrity Associates Engineering Report No. 0901350.304, Revision 3, Fatigue Crack Growth Evaluation, October 12, 2018 L-2018-174, Structural Integrity Associates Engineering Report No. 0901350.401, Revision 4, Leak-Before-Break Evaluation - Accumulator, Pressurizer Surge, and Residual Heat Removal Lines2018-10-12012 October 2018 Structural Integrity Associates Engineering Report No. 0901350.401, Revision 4, Leak-Before-Break Evaluation - Accumulator, Pressurizer Surge, and Residual Heat Removal Lines L-2018-169, Responses to Requests for Additional Information for Subsequent License Renewal Application No. HC-7-a L-2018-169 Environmental Review2018-10-0505 October 2018 Responses to Requests for Additional Information for Subsequent License Renewal Application No. HC-7-a L-2018-169 Environmental Review L-2018-187, Subsequent License Renewal Application Revision to SLRA Section 3.5.2.2.2.6, Reduction of Strength and Mechanical, Properties of Concrete Due to Irradiation2018-10-0505 October 2018 Subsequent License Renewal Application Revision to SLRA Section 3.5.2.2.2.6, Reduction of Strength and Mechanical, Properties of Concrete Due to Irradiation L-2018-173, Notification of Request for Use and Approval of Polyacrylic Acid Pilot Program2018-09-21021 September 2018 Notification of Request for Use and Approval of Polyacrylic Acid Pilot Program ML18254A3412018-09-11011 September 2018 Fish and Wildlife Service'S List of Migratory Birds Near Turkey Point ML18227B5212018-08-15015 August 2018 Submit Attachment a, Annual Hourly Percent Frequency of Vertical and Horizontal Stability Categories by Wind Direction and Wind Speed ML18227A2902018-08-15015 August 2018 Submit Report Contains Official Summary of Startup Physics Tests, Unit 4 Cycle Iii. the Tests Were Conducted in Accordance with Operating Procedure 0204.5, Startup Sequence After Refueling L-2018-098, Annual Report for the AP1000 Standard Plant Design, 2017 Model Year 10 CFR 50.462018-04-18018 April 2018 Annual Report for the AP1000 Standard Plant Design, 2017 Model Year 10 CFR 50.46 L-2018-054, Attachment B: Process Control Program, 0-HPA-045, Revision 0A, Issued 2/16/172018-02-16016 February 2018 Attachment B: Process Control Program, 0-HPA-045, Revision 0A, Issued 2/16/17 L-2017-148, Special Report - Accident Monitoring Instrumentation2017-08-11011 August 2017 Special Report - Accident Monitoring Instrumentation L-2017-123, Special Report - Standby Steam Generator Feedwater Pumps Inoperable2017-06-29029 June 2017 Special Report - Standby Steam Generator Feedwater Pumps Inoperable L-2017-124, Flooding Focused Evaluation Summary2017-06-29029 June 2017 Flooding Focused Evaluation Summary L-2017-014, Florida Power & Light Company - 10 CPR 50.46 Annual Reporting of Changes to, or Errors in Emergency Core Cooling System Models or Applications for 20162017-04-17017 April 2017 Florida Power & Light Company - 10 CPR 50.46 Annual Reporting of Changes to, or Errors in Emergency Core Cooling System Models or Applications for 2016 L-2016-226, NEI 12-06, Revision 2, Appendix G, G.4.2 Mitigating Strategies Assessment Flex Strategies Report for the New Flood Hazard Information2016-12-20020 December 2016 NEI 12-06, Revision 2, Appendix G, G.4.2 Mitigating Strategies Assessment Flex Strategies Report for the New Flood Hazard Information L-2016-058, Licensee Qualification for Performing Dynamic Rod Worth Measurement Analysis2016-03-23023 March 2016 Licensee Qualification for Performing Dynamic Rod Worth Measurement Analysis ML16013A4722016-01-22022 January 2016 Staff Assessment of Information Provided Pursuant to Title 10 of the Code of Federal Regulations Part 50, Section 50.54(f) Seismic Hazard Reevaluations for Recommendation 2.1 of the Near-Term Task L-2015-313, Special Report - Accident Monitoring Instrumentation2015-12-18018 December 2015 Special Report - Accident Monitoring Instrumentation L-2015-137, Submittal of 10 CFR 50.59(d)(2) Summary Report of Changes, Tests and Experiments2015-04-22022 April 2015 Submittal of 10 CFR 50.59(d)(2) Summary Report of Changes, Tests and Experiments ML15054A0372015-02-23023 February 2015 U.S. Nuclear Regulatory Commission, Record of Review Dispositions to Fire PRA Facts and Observations for Turkey Point Nuclear Generating Stations, Units 3 and 4 ML14338A5552014-12-0404 December 2014 NRC-2013-TN3079-NRC 2014 St. Lucie License Renewal 2024-05-09
[Table view] Category:Technical
MONTHYEARL-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 ML23333A0172023-11-27027 November 2023 Attachment I - Chin - the Cooling Canal System at the FPL-Turkey-Point Power Station ML23333A0142023-11-27027 November 2023 Attachment F - Groundwater Tek Inc - Peer Review Study Final-1 ML23265A5512023-09-22022 September 2023 Enclosure 3: WCAP-18830-NP, Turkey Point Fuel Storage Criticality Analysis for 24 Month Cycles L-2023-049, Correction to U4R33 Steam Generator Tube Inspection Report2023-03-30030 March 2023 Correction to U4R33 Steam Generator Tube Inspection Report L-2023-010, Supplemental Information Regarding License Amendment Request 274, Reactor Protection System, Engineered Safety Features Actuation System, and Nuclear Instrumentation System Replacement Project - Submittal of RPS / ESFAS / Nis2023-02-10010 February 2023 Supplemental Information Regarding License Amendment Request 274, Reactor Protection System, Engineered Safety Features Actuation System, and Nuclear Instrumentation System Replacement Project - Submittal of RPS / ESFAS / Nis L-2022-168, and Point Beach Units 1 and 2 - 10 CFR 50.46 - Emergency Core Cooling System LBLOCA 30-Day Report2022-10-26026 October 2022 and Point Beach Units 1 and 2 - 10 CFR 50.46 - Emergency Core Cooling System LBLOCA 30-Day Report L-2022-110, License Amendment Request 276, Revise Fire Protection Program in Support of Reactor Coolant Pump Seal Replacement Project (Non-Proprietary)2022-08-26026 August 2022 License Amendment Request 276, Revise Fire Protection Program in Support of Reactor Coolant Pump Seal Replacement Project (Non-Proprietary) L-2022-142, Revised Diversity and Defense-In-Depth Evaluation (D3), Framatome Document No. 51-9324096-0042022-08-19019 August 2022 Revised Diversity and Defense-In-Depth Evaluation (D3), Framatome Document No. 51-9324096-004 L-2022-073, Diversity and Defense-In-Depth Evaluation (D3), Framatome Document No. 51-9324096-0042022-05-0303 May 2022 Diversity and Defense-In-Depth Evaluation (D3), Framatome Document No. 51-9324096-004 L-2022-022, Updated Conditions of Certification Report2022-02-14014 February 2022 Updated Conditions of Certification Report L-2021-069, Commitment Revision Summary Report for Reactor Vessel Head Leakage Detection System (Rvhlds)2021-06-24024 June 2021 Commitment Revision Summary Report for Reactor Vessel Head Leakage Detection System (Rvhlds) L-2020-073, Fifth Ten-Year Inservice Inspection Interval Revised Relief Request No. 6 and Supplemental Information for Train B CCW Return Piping2020-04-13013 April 2020 Fifth Ten-Year Inservice Inspection Interval Revised Relief Request No. 6 and Supplemental Information for Train B CCW Return Piping ML20098F3412020-04-0707 April 2020 Exigent License Amendment Request 272, One-Time Extension of TS 6.8.4 Steam Generator Inspection Program - Response to Request for Additional Information L-2019-010, Proposed Alternative for the Use of Encoded Phased Array Ultrasonic Examination Techniques in Lieu of Radiography for Ferritic and Austenitic Welds2019-03-19019 March 2019 Proposed Alternative for the Use of Encoded Phased Array Ultrasonic Examination Techniques in Lieu of Radiography for Ferritic and Austenitic Welds L-2019-054, Baffle-Former Bolts Predictive Evaluations2019-03-13013 March 2019 Baffle-Former Bolts Predictive Evaluations ML19072A1622019-03-0505 March 2019 National Park Service, Southeast Regional Office, Comments Dated March 5, 2019, on Turkey Point Nuclear Generating Units 3 and 4 Preliminary Draft SEIS Regarding Subsequent License Renewal ML18353A8352018-12-31031 December 2018 Biological Assessment for the Turkey Point Units 3 and 4 Proposed Subsequent License Renewal ML18299A1162018-10-15015 October 2018 Structural Integrity Associates Engineering Report No. 1700109.401P, Revision 7 - Redacted, Evaluation of Environmentally-Assisted Fatigue for Turkey Point Units 3 and 4 for Subsequent License Renewal L-2018-174, Structural Integrity Associates Engineering Report No. 0901350.401, Revision 4, Leak-Before-Break Evaluation - Accumulator, Pressurizer Surge, and Residual Heat Removal Lines2018-10-12012 October 2018 Structural Integrity Associates Engineering Report No. 0901350.401, Revision 4, Leak-Before-Break Evaluation - Accumulator, Pressurizer Surge, and Residual Heat Removal Lines ML18299A1182018-10-12012 October 2018 Structural Integrity Associates Engineering Report No. 0901350.304, Revision 3, Fatigue Crack Growth Evaluation, October 12, 2018 L-2018-169, Responses to Requests for Additional Information for Subsequent License Renewal Application No. HC-7-a L-2018-169 Environmental Review2018-10-0505 October 2018 Responses to Requests for Additional Information for Subsequent License Renewal Application No. HC-7-a L-2018-169 Environmental Review L-2018-187, Subsequent License Renewal Application Revision to SLRA Section 3.5.2.2.2.6, Reduction of Strength and Mechanical, Properties of Concrete Due to Irradiation2018-10-0505 October 2018 Subsequent License Renewal Application Revision to SLRA Section 3.5.2.2.2.6, Reduction of Strength and Mechanical, Properties of Concrete Due to Irradiation ML18227B5212018-08-15015 August 2018 Submit Attachment a, Annual Hourly Percent Frequency of Vertical and Horizontal Stability Categories by Wind Direction and Wind Speed ML18227A2902018-08-15015 August 2018 Submit Report Contains Official Summary of Startup Physics Tests, Unit 4 Cycle Iii. the Tests Were Conducted in Accordance with Operating Procedure 0204.5, Startup Sequence After Refueling L-2018-098, Annual Report for the AP1000 Standard Plant Design, 2017 Model Year 10 CFR 50.462018-04-18018 April 2018 Annual Report for the AP1000 Standard Plant Design, 2017 Model Year 10 CFR 50.46 L-2018-054, Attachment B: Process Control Program, 0-HPA-045, Revision 0A, Issued 2/16/172018-02-16016 February 2018 Attachment B: Process Control Program, 0-HPA-045, Revision 0A, Issued 2/16/17 L-2014-085, Seismic Hazard and Screening Report (CEUS Sites), Response to NRC Request for Information Pursuant to 10 CFR 50.54(f) Re Recommendation 2.1 of the Near-Term Task Force Review of Insights from Fukushima Dai-ichi Accident2014-03-27027 March 2014 Seismic Hazard and Screening Report (CEUS Sites), Response to NRC Request for Information Pursuant to 10 CFR 50.54(f) Re Recommendation 2.1 of the Near-Term Task Force Review of Insights from Fukushima Dai-ichi Accident ML14002A1602014-02-0606 February 2014 Interim Staff Evaluation Relating to Overall Integrated Plan in Response to Order EA-12-049 (Mitigation Strategies) ML14029A2522014-01-29029 January 2014 Mega-Tech Services, LLC Technical Evaluation Report Regarding the Overall Integrated Plan for Turkey Point Nuclear Plant, Units 3 and 4, TAC Nos.: MF0982 and MF0983 ML13213A1902013-07-0909 July 2013 Rev. 2 to Seismic Walkdown Report, in Response to the 50.54(f) Information Request Re Fukushima Near-Term Task Force Recommendation 2.3: Seismic for Turkey Point Unit 3. L-2013-198, Rev. 2 to Seismic Walkdown Report, in Response to the 50.54(f) Information Request Re Fukushima Near-Term Task Force Recommendation 2.3: Seismic for Turkey Point Unit 4.2013-07-0909 July 2013 Rev. 2 to Seismic Walkdown Report, in Response to the 50.54(f) Information Request Re Fukushima Near-Term Task Force Recommendation 2.3: Seismic for Turkey Point Unit 4. L-2013-087, Flood Hazard Reevaluation Report in Response to 50.54(f) Information Request Regarding Near-Term Task Force Recommendation 2.1. Part 2 of 22013-03-11011 March 2013 Flood Hazard Reevaluation Report in Response to 50.54(f) Information Request Regarding Near-Term Task Force Recommendation 2.1. Part 2 of 2 ML13095A1962013-03-11011 March 2013 Flood Hazard Reevaluation Report in Response to 50.54(f) Information Request Regarding Near-Term Task Force Recommendation 2.1. Part 1 of 2 L-2012-417, Flooding Walkdown Report FPL061-PR-001, Rev 0, in Response to the 50.54(f) Information Request Regarding Near-Term Task Force Recommendation 2.3: Flooding2012-11-20020 November 2012 Flooding Walkdown Report FPL061-PR-001, Rev 0, in Response to the 50.54(f) Information Request Regarding Near-Term Task Force Recommendation 2.3: Flooding L-2011-190, WCAP-17070-NP, Revision 1, Westinghouse Setpoint Methodology for Protection Systems, Turkey Point, Units 3 & 4 (Power Uprate to 2644 Mwt - Core Power), Attachment 2 to L-2011-1902011-06-30030 June 2011 WCAP-17070-NP, Revision 1, Westinghouse Setpoint Methodology for Protection Systems, Turkey Point, Units 3 & 4 (Power Uprate to 2644 Mwt - Core Power), Attachment 2 to L-2011-190 L-2011-032, WCAP-17094-NP, Rev 3, Turkey Point, Units 3 and 4 New Fuel Storage Rack and Spent Fuel Pool Criticality Analysis, Attachment 2 to L-2011-0322011-02-28028 February 2011 WCAP-17094-NP, Rev 3, Turkey Point, Units 3 and 4 New Fuel Storage Rack and Spent Fuel Pool Criticality Analysis, Attachment 2 to L-2011-032 L-2010-113, Turkey Point, Units 3 and 4 - License Amendment Request for Extended Power Uprate, Attachment 2; Renewed Facility Operating License and Technical Specification Proposed Change Markups2010-12-14014 December 2010 Turkey Point, Units 3 and 4 - License Amendment Request for Extended Power Uprate, Attachment 2; Renewed Facility Operating License and Technical Specification Proposed Change Markups ML1035601802010-12-14014 December 2010 License Amendment Request for Extended Power Uprate, Attachment 04; Appendix C, List of Key Acronyms ML1035601832010-12-14014 December 2010 License Amendment Request for Extended Power Uprate, Attachment 07; Supplemental Environmental Report L-2010-113, Turkey Point, Units 3 and 4 - License Amendment Request for Extended Power Uprate, Attachment 08; Cameron/Caldon Ultrasonics Engineering Reports2010-12-14014 December 2010 Turkey Point, Units 3 and 4 - License Amendment Request for Extended Power Uprate, Attachment 08; Cameron/Caldon Ultrasonics Engineering Reports ML1035601842010-12-14014 December 2010 License Amendment Request for Extended Power Uprate, Attachment 08; Cameron/Caldon Ultrasonics Engineering Reports L-2010-113, Turkey Point, Units 3 and 4 - License Amendment Request for Extended Power Uprate, Attachment 07; Supplemental Environmental Report2010-12-14014 December 2010 Turkey Point, Units 3 and 4 - License Amendment Request for Extended Power Uprate, Attachment 07; Supplemental Environmental Report L-2010-113, Turkey Point, Units 3 and 4 - License Amendment Request for Extended Power Uprate, Attachment 06; Summary of Regulatory Commitments2010-12-14014 December 2010 Turkey Point, Units 3 and 4 - License Amendment Request for Extended Power Uprate, Attachment 06; Summary of Regulatory Commitments ML1035601822010-12-14014 December 2010 License Amendment Request for Extended Power Uprate, Attachment 06; Summary of Regulatory Commitments L-2010-113, Turkey Point, Units 3 and 4 - License Amendment Request for Extended Power Uprate, Attachment 04; Appendix C, List of Key Acronyms2010-12-14014 December 2010 Turkey Point, Units 3 and 4 - License Amendment Request for Extended Power Uprate, Attachment 04; Appendix C, List of Key Acronyms L-2010-113, Turkey Point, Units 3 and 4 - License Amendment Request for Extended Power Uprate, Attachment 04; Appendix B, Additional Codes and Methods2010-12-14014 December 2010 Turkey Point, Units 3 and 4 - License Amendment Request for Extended Power Uprate, Attachment 04; Appendix B, Additional Codes and Methods ML1035601792010-12-14014 December 2010 License Amendment Request for Extended Power Uprate, Attachment 04; Appendix B, Additional Codes and Methods L-2010-113, Turkey Point, Units 3 and 4 - License Amendment Request for Extended Power Uprate, Attachment 04; Appendix a, Safety Evaluation Report Compliance2010-12-14014 December 2010 Turkey Point, Units 3 and 4 - License Amendment Request for Extended Power Uprate, Attachment 04; Appendix a, Safety Evaluation Report Compliance ML1035601782010-12-14014 December 2010 License Amendment Request for Extended Power Uprate, Attachment 04; Appendix a, Safety Evaluation Report Compliance 2024-04-30
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U.S. NUCLEAR REGULATORY COIAMISS ION DOCK 8ER hlRC FORM 195 I2~ 76 I -2 251 NRC DISTRIBUTION soR PART EO DOCKLI P MA I'ERIAL TO: FROM: DATE OF DOCUMENT Florida Power & Light Company 12/9/76 Hr. Victor Stello Hiami, Florida DATE RECEIVED Hr. Robert E. Uhrig 12/9/76
%LETTER DNOTORIZED PROP INPUT FORM NUMSER OF COPIES RECEIVED JHl ORIGINAL gUNC LASSIF IE D Three signed Qcopv 35 copies encl recvd.
DESCRIPTION ENCLOSURE t
L tr. w/a tached erra ta shee t.... re our 6/17/76 and 12/3/76 orders...trans the following: Concerns re-evaluation of ECCS cooling performance calculated in accordance with SOMOT REMOVE an approved Nestinghouse Evaluation Model, with appropriate corrections for upper head water temperature.
ACKNOWLEDGED (3-P) (6O-P) .
PLANT NAME:
Turkey POIInt Units 3 & 4 SAFETY FOR ACTION/INFORMATION 12 10 76 ASSXGNED AD:
Lear RO EC MANA Elliott PROJECT MANAGER'IC L C ASST arrz.s ASST INTERNAL DISTRIBUTION REG FILE SYSTEMS SAFETY PLA1AT SYSTEMS SITE SA E HEINEHAN TEDESCO I &E SCHROEDER N 0 A OELD GOSSICK & STAFF ENGINEERING IPPOLXTO MIPC ERNST CASE KNXGHT HANAUER HARLESS SIHWEIL PAWL CK OPERATING REACTORS STELLO SPANGLER gpss SITE TECH PROJECT MANAGEMENT REACTOR SAFE OPERATING TECH GAMHILL BOYD ROSS EXSENHUT STEPP PE COLLINS NOVAK HULMAN HOUSTON ROSZTOCZY PETERSON CHECK B .E SITE ANALYSXS MELTZ VOLLHER HELTEMES AT & I BUNCH SKOVHOLT SALTZMAN J ~ COLLINS RUTBERG KREGER EXTERNAL DISTRIBUTION, CONTROL NUMBER LPDR ~ Miami, Fla. NAT LAB'EG B 0 K TXC: VOGIE ULR KSON OR NSIC: LA PDR j+
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~~+ FLORIOA POWER & LIGHT COMPANY y'~c @@4 December 9, 1976 L-76-419 Office of Nuclear Reactor Regulation 0 Attention: Mr. Victor Stello, Director Division of Operating Reactors <-'~'~o~
U. S. Nuclear Regulatory Commission +
Washington, D. C. 20555 +e pegutatory Docket I Iie
Dear Mr. Stello:
Re: Turkey Point Units 3 6 4 Docket Nos. 50-250 and 50-251 ECCS Re-evaluation In accordance with the June 17, 1976 and December 3, 1976 Orders for Modification of License issued by the Commission for Turkey Point Unit 4, Florida Power 6 Light Company hereby submits a re-evaluation of ECCS cooling performance calculated in accordance with an approved Westinghouse Evaluation Model, with appropriate corrections for upper head water. temperature..
The re-evaluation also includes the effects of (1) plugged steam generator tubes and (2), modifications to increase the water volume of the Safety Injection accumulators. The re-evaluation was performed using a rated power level of 2192 Mwt instead of the actual Technical Specification rated power level of 2200 Electric Corporation is now revising their calcula-Mwt.'estinghouse tions using 2200 Mwt and, when this is complete, change pages will be forwarded to you to bring the ECCS re-evaluation up to date.
The ECCS re-evaluation assumes -a minimum accumulator water volume of 875 cubic feet, however, the actual m'inimum'water volume for the remainder of core cycle 3 for Unit 4 will be 825 cubic feet. Modifications to increase the Unit 4 accumulator volume are planned for the Spring 1977 refueling outage.
Modifications to increase the Unit 3 accumulator volume will be completed during the Fall 1976 refueling outage which is now in progress.
Core Cycle 3 parameters were used in the ECCS re-evaluat'on.
Westinghouse Electric Corporation then compared Cycle 4 operation with Cycle 3 operation and concluded that Cycle 3 is PEOPLE... SERVING PEOPLE
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Office of Nuclear Reactor Regulation Attention: Mr. Victor Stello, Director Page Two more limiting than Cycle 4 with respect to ECCS performance.
The comparison included consideration of the three Region 3 assemblies which will be reloaded into Unit 4 for Cycle 4 operation.
Proposed Technical Specification amendments incorporating the results of the ECCS re-evaluation will be submitted under separate cover letters.
Very tru ursg Robert, E. Uhrig Vice President REU/MAS/cpc Attachment cc: Mr. Norman C. Moseley Robert Lowenstein, Esquire
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ERRATA SHEET Page 1, paragraph 1, line 7: replace the words "in Section 15.3.1" with the word "herein".
Page 1, paragraph 2, line 1: add the word "would" after the words "Reactor Coolant System".
Page 1, paragraph 2, line 4: the word "injection" should be capitalized.
Page 2, paragraph 3, line 2: The word "serverance" should be spelled "severance".
Page 3, paragraph 1, last line: replace the Table number "15.4-3" with the number "3".
Page 3, paragraph 4, lines 6 and 7: delete the words "the value of the peak linear power density used in this analysis and". This same phrase is used twice.
Page 4, paragraph 2, first subparagraph: the subparagraph heading should be "Figures la through 3d".
Page 5, subparagraph 2, last line: the word "intack" should be spelled "intact".
Page 5, main paragraph 2, line 3: the word "reacotr" should be spelled "reactor".
Page 5, main paragraph 2, line 4: the word "id" should be spelled "is".
Page 6, paragraph 1, line 3: delete the period after the word "saturation".
Page 6, paragraph 1, line 6: the word "whos" should be spelled "whose".
Page 6, paragraph 3, line 1: the word "toal" should be spelled "total", and the number "013" should be ".013".
Page 10, Table 2, second column: the expression "CD+0.6" should be "CD=0.6".
Page 10, Table 2, third and fourth columns: the expressions "CD=0.4" should be completely, enclosed by parentheses.
Page 10, Table 2, near the bottom of the page: the word "cucle" should be spelled "cycle".
Page ll, Table 3, last line: the word "Fastests" should be spelled "Fastest".
e ~ <a MAJOR REACTOR COOLANT SYSTEM PIPE RUPTURES (LOSS OF COOLANT ACCIDENT)
The analysis specified by 10CFR50.46 Cl] Acceptance Criteria for Emer-gency Core Cooling Systems for Light Water Power Reactors", is presented in this section. The results of the loss of coolant accident analvses are shown in Table 2 and show compliance with the Acceptance Criteria.
The analytical techniques used are in compliance with Appendix K of 10CFR50, and are described in Reference [2]. The results for the small
- break loss of coolant accident are presented in Section 15.3.1 and are in conformance with 10CFR50.46 and Appendix K of 1GCRR50.
Should a major break occur, depressurization of the Reactor Coolant System result in a pressure decrease in the pressurizer. Reactor trip signal occurs when the pressurizer low pressure trip setpoint is reached. A Safety injection System signal is actuated when the appropriate set-point is reached. These countermeasures will limit the consequences of the accident in two ways:
Reactor trip and borated water injection complement void formation in causing rapid reduction of power to a residual level corresponding to fission product decay heat.
\
2 Injection of borated water provides heat transfer from the core and prevents excessive clad temperatures.
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At the beginning of the blowdown phase, the entire Reactor Coolant System contains subcooled liauid which transfers heat from the core by forced convection with some fully developed nucleate boiling. After the break develops, the time to departure from nucleate boiling is calculated, consistent with Appendix K of 10CPR50. Thereafter the core heat transfer is based on local conditions with boiling and forced con-vection to steam as the major heat transfer transition mechanisms. During the refill period, rod-to-rod radiation is the only heat transfer mech-anism.
When the Reactor Coolant System pressure falls below 600 psia the ac-cumulators begin to inject borated water. The conservative assumption is made that accumulator water injected bypasses the core and goes out through the break until the termination of bypass, 'his conservatism is again consistent with Appendix K of 10CFR50.
Thermal Anal sis Westinghouse Performance Criteria for Emergency Core Cooling System.
The reactor is designed to withstand thermal effects caused by a loss of coolant accident including the double ended serverance of the largest Reactor Coolant System pipe. The reactor core and internals together with the Emergency Core Cooling System (ECCS) are designed zo that the reactor can be safely shutdown and the essential heat transfer geometry of the core preserved following the accident.
The ECCS, even when operating during the injectio n mode with the most severe single active failure, is designed to meet the Acceptance Criteria December 1976
41 Method of Thermal Analysis The description of the various aspects of the loss of coolant accident analysis is given in Reference [2]. This document describes the major phenomena modeled, the interfaces among the computer codes and features of the codes which maintain compliance with the Acceptance Criteria.
The individual codes are described in detail in References [3] through
[6]. The analyses presented here were performed using the October 1975 version of the Nestinghouse Evaluation Model. This version includes the modifications to the models, referenced above, as specified by the Nuclear Regulatory Commission (NRC) in Reference [7] and complies with Appendix K, of 10CPR50. The October, 1975 Llestinghouse Evaluation Model is documented in References [8, ll and 12]. Containment data used to calculate ECCS backpressure is presented in Table 15.4-3.
Table 1 presents the time sequence of events for both ECCS analyses using the NRC and Westinghouse Parameters.
Results Table 2 presents the peak clad temperatures and hot spot metal reaction for a range of break sizes. This range of break sizes was de-termined to include the limiting case for peak clad temperature from sensitivity studies reported in References 9 and 10.
The Satan VI analysis of the loss of coolant accident is performed at 102 percent of Licensed Application Core Power Level (power level shown in Table 2). The peak linear power, and peaking factor at the license application power level'sed in the'analyses, are also given in Table 2.
Since there is margin between the value of the peak linear power density used in this analysis and the value of the peak linear power density used in this analysis and the value expected in operation, a lower peak clad temperature would be obtained by using the peak linear power density expected during operation.
December 1976
4 A
~4 I 4
Three cases are analyzed with 5% uniform steam generator tube plugging.
An additional case is presented, for the limiting break, with 10%
uniform steam generator tube plugging.
For the results discussed below, the hot spot is defined to be the location of maxiumum peak clad temperature. This location is given in Table 2 for each break size analyzed.
Figures 1 through 16 present the transients for the principal parameters for the break sizes analyzed. The following items are noted:
Figures la The following quantities are presented at the clad burst location and at the hot spot (location of maximum clad temperature) both on the hottest fuel rod (hot rod):
(1) fluid quality (2) mass velocity (3) heat transfer coefficient.
The heat transfer coefficient shown is calculated by the LOCTA IV code.
Figures 4a The system pressure shown is the calculated pressure in the core. The flow rate out the break is plotted as the sum of both ends for the guillotine break cases. This core pressure drop shown is from the lower plenum, near the core, to the upper plenum at the core outlet.
Figures 7a These figures show the hot spot clad temperature transient through 9d and the clad temperature transient at the burst location.
The fluid temperature shown is also for the hot spot and burst location. The core flow (top and bottom) is also shown.
Figures 10a These figures show the core reflood transient.
through 10d December 1976
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Figures lla These figures show the Emergency Core Cooling System through 12d flow for all cases analyzed. .As described earlier, the accumulator delivery during'lowdown is discarded until the end of bypass is calculated.
Accumulator flow, however, is established in refill reflood calculations. The accumulator flow assumed is the sum of that injected in the intack cold legs.
Figures 13 The containment pressure transient is also provided.
a,b,c,d Figures 14 These figures show the core power transient.
a,b,c,d Figures 15 These figures show the break energy released to the a,b,c,d containment during blowdown for the limiting case break.
Figure 16 This fC.gure provides the containment wall condensing heat transfer coefficient for the limiting case break.
In addition to the above, Tables 4 and 5 present the reflood mass and energy release to the containment and the broken loop accumulator mass and energy flowrate to the containment, respectively.
The analysis presented in this section was performed with a reactor vessel upper head temperature equal to the RCS hot leg temperature. The effect of using the hot leg temperature in the reacotr vessel upper head id described in Reference 13. A break spectrum sensitivity study using the hot leg temperature is presented in Reference [14].
The purpose of the Reference 14 sensitivity study is to show that changing the upper head water temperature does not change the limiting break type and location, which is a Double Ended Cold Leg Guillotine, for a three loop plant. The three loop plant configuration used for this sensitivity study is sufficiently similar to the Turkey Point Units 3 and 4 plants to assure that the limiting break is identified.
December 1976
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Particular details of p ant design do not change the ba c effect re-sulting from the higher upper head temperature, that being the flashing of water at the hot leg saturation pressure rather than the cold leg saturation.
pressure. In addition previous sensitivity studies (References 9, 10, and 14) performed for 3 loop plants have been consistent, in verifying this limiting break type and location for Mestinghouse plants whos designs reflect the differences noted between Reference 14 and the Turkey Point 3 6 4 plants.
The clad temperature analysis is based on a total peaking factor of 2.25.
The hot spot metal water reaction reached is 10.732%, which is well below the embrittlement limit of 17 percent, as required by 10CFR50.46.
In addition, the toal core metal-water reaction is less than 013 percent for all breaks as compared with the 1 percent criterion of 10CFR50.46.
The results of several sensitivity studies are reported in Reference 9.
These results are for conditions which are not limiting in nature and hence are reported on a generic basis.
For breaks up to and including the double ended severance of a reactor coolant pipe, the Emergency Core Cooling System will meet the Accep-tance Criteria as presented in 10CFR50.46. That is;
- l. The calculated peak fuel element clad temperature provides margin to the requirement of 2200'F based on Fq value of 2.25.
The amount of fuel element cladding that reacts chemically with water or steam does not exceed 1 percent of the total amount of Zircaloy in the reactor.
3~ The clad temperature transient is terminated at a time when the core geometry is still amenable to cooling. The cladding oxidation limits of 17 percent are not exceeded during or after quenching.
- 4. The core temperature is reduced and decay heat is removed for an extended period of time, as required by the long-lived radio-activity remaining in the core.
December 1976
'a REFERENCES "Acceptance Criteria for Emergency Core Cooling Systems for Light Hater Cooled Nuclear Power Reactors," 10CFR50.46 and Appendix K of 10CFR50. Federal Register, Volume 39 Number 3, January 4, 1974.
Bordelon, F.M., Massie, H.H. and Z ordon T.A., "Westinghouse FCCS Evaluation Model Summary," WCAP-8339, July, 1974 Bordelon, F.M., et al., "SATAN-VI Program: Comprehensive Space-Time Dependent Analysis of Loss of Coolant," HCAP-8302, June, 1974 (Proprietary) and WCAP-8306, June, 1974 (Non-Proprietary).
Bordelon, F. M., et al., "LOCTA-IV Program: Loss of Coolant Tran-sient Analysis," WCAP-8301, June 1974 (Proprietary) and HCAP-8305, June, 1974 (Non-Proprietary).
Kelly, R. D., et al., "Calculational Model for Core Reflooding after a Loss of Coolant Accident (HREFLOOD Code)," WCAP-8170, June 1974 (Proprietary) and HCAP -8171, June 1974 (Non-Proprietary) .
Bordelon, F. M. and Murphy, E. T., "Containment Pressure Analysis Code (COCO)," WCAP-8327, June, 1974 (Proprietary) and HCAP-8326, June, 1974 (Non-Proprietary) .
"Supplement to the Status Report by the Directorate of Licensing in the matter of Westinghouse Flectric Company ECCS Evaluation Model Conformance to 10CFR50 Appendix K." Federal Register, November, 1974.
Bordelon, F.M., et al., "Westinghouse ECCS Evaluation Model-Supplementary Information," HCAP-8471-P-A ,April, 1975, (Proprietary) and HCAP 8472-A, April, 1975 (Non-Proprietary).
Salvatori, R., "Westinghouse ECCS - Plant Sensitivitv Studies,"
HCAP-8340, July, 1974 (Proprietary) and HCAP-8356, July, 1974 (Non-Proprietary).
Buterbaugh, T. L., Julian, H.V. and Tome, A.E., "Westinghouse ECCS-Three Loop Plant (17x17) Sensitivity Studies" WCAP 8572-P July, 1975 (Proprietary) and WCAP 8573, July, 1975 (Non-Proprietary).
December 1976
"Westinghouse ECCS Evaluation 1'fodel October, 1975 Version, WCAP 8622, November 1975 (Proprietary) and WCAP-8623, November 1975, (Non-Proprietary) .
Letter from C. Eicheldinger of Westinghouse Electric Corporation to D. B. Vassallo of the Nuclear Regulartory Commission; letter number NS-CE-924, dated January 23, 1976.
Letter from C. Eicheldinger of Westinghouse Electric Corporation to V. Stello of the Nuclear Regulatory Commission, Letter Number NS-CE-1163 dated August 13, 1976, Julian, H. V., Tabone, C. J., and Thompson, C. N., "Westinghouse ECCS-Three Loop Plant (17x17) Sensitivity Studies, WCAP 8853, September, 1976 (Non-Proprietary) .
December 1976
Ai 4, TABLE 1 LARGE BREA'INE SEOUENCE OF EVENTS
- DECL +DECL ~DECL o'ctcDECLG (CD-1.0) (CD=0.6) (CD=0.4) (CD=0.4)
(Sec) (Sec) (Sec) (Sec)
START 0.0 0.0 0.0 0.0 Reactor Trip Signal 0.573 0.583 0,595 0.595 S. I'ignal 0.44 0.55 0.67 0.67 Acc. Injection 10. 3 12 ' 16.6 16.3 End of Bypass 20.96 23. 81 27. 89 27.66 End of Blowdown 21.18 23.96 28 04
~ 27.84 Bottom of Core Recovery 40.37 42.78 46.94 46.78 Acc. Empty 55.09 57.71 61. 52 61. 19 Pump Injection 25.44 25.55 25.67 25. 67 Contains 5% Stm. Gen. Tube Plugging (Figures lc thru 14c)
>>* Contains 10% Stm. Gen. Tube Plugging (Figures ld through 14d, 15 and 16)
December, 1976
TABLE 2 L+GE BREAK Results >>DECL '<DECL <<DF.CL >>>>DECLG (CD-1.0) (CD+0.6) (CD=0.4 (CD=0.4 Results Peak Clad Temp. 'F 1767 1927 2162 2198 Peak Clad Location Ft. 6.25 6.5 6.0 6.0 Local Zr/H20 Rxn(max)% 2.009 3.627 10.732 12.310 Local Zr/H20 Location Ft. 6,0 6.0 6.0 6' Total Zr/H20 Rxn % L0.3 L0.3 LO ~ 3 L0.3 Hot Rod Burst Time Sec. 45.24 31.0 24.1 22,9 Hog Rod.Burst Location Ft. 6 0
~ 6.0 6.0 6.0 Calculation Core Power'wt 102% of 2192 Peak Linear Power kw/ft 102% of Peaking Factor 2.25 Accumulator Mater Volume (ft3) ~875 ( ar accumulator)
Fuel region + cucle analyzed Cycle Region UNITS 3 and 4
- Contains 5% Stm. Gen. Tube Pilugging
>>>> Contains 10% Stm Gen. Tube Plugging December 1976 10
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TABLE 3 LARGE BREAK CONTAINNFNT DATA (DRY CONTAINMENT)
NET FREE VOLlNi E 1.55xl0 Ft3 INITIAL CONDITIONS Pressure 14.7 psia 900&
Temperature RMST Temperature 39'F Service Hater Temperature 63'F Outside Temperature 39'F SPRAY SYSTEM Number of Pumps Operating Runout Flow Rate 1450 gpm Actuation Time 26 sec SAFEGUARDS FAN COOLERS Number of Fan Coolers Operating Fastests Post Accident Initiation of Fan Coolers 26 sec December 1976
~k LARGE BREAK Table 3 (continued)
CONTAINMENT DATA (Dry Containment)
STRUCTURAL HEAT SINKS Thickness (In) Area (Ft2)
Steel 0.03 31,400 0,063 107,158 0.1 56,371 0.2 57,185 0.24 9931 Steel 0. 2898, 136,000 Concrete 24.0 steel 0. 4896 23,667 0.6396 6537 0.8904 4915 1.256 27802 1.56 5307 2.0 668 2.75 . 1268,7 5.5 1277.4 Steel 9.0 260 '
Stainless Steel 14,392 Concrete Stainless Steel 0.44 768 Stainless Steel 2. 126 3704 Stainless Steel 0.007 102,400 Contrete 24.0 59,132 12 December 1976
TABLE 4 Reflood Mass and Energy Releases for limiting. Case DECLG (DC=0.4) and 5% Stm. Gen. Tube. Plugging TIME TOTAL MASS FLOWRATE TOT~ ENERGY FLOWRATE (Sec) (LBm/Sec) (105 BTU/Sec) 46.943 0.0 0.0
- 49. 068 0.0 0.0 54.575 35. 89 0.4663 64.668 99.81 1.239 77.168 101.9 1.264 91.868 107. 0 1.319 107.468 215.6;,. 1.623 123.568 260. 7 1. 707 157.968 273. 1 1.638 195.568 279.8 1.545 13 December 1976
TABLE 5 Broken Loop Accumulator Flow to Containment For Limiting Case DECLG (CD=0.4) 5% Stm Gen. Tube Plugging Time (Sec) Mass Flowrate ( LBm/Sec) 0.0 0.0 0.02 2723.5 2.00 2276.0 4,00 1994.9 6.00 1793.1 8.00 1645.5 10.0 1526.4 15.0 1302.5 20.0 1137.8 25.0 1034.6 30.0 954.2 35.0 887.0 38.9 842.6 December 1976
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Figure la. Fluid Quality DECLG (Cg = 1.0)
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Figure 1b. Fluid Quality OECLG (CD = 0.6)
I
.7
.6 lpo 5 lpi !02 2 I 03 TIME (SECOHDS)
Figure 1c. Fluid Quality DECLG (CD = 0.4)
l l 0
I.2 I.P
.7
.6 IO 2 IO 2 5 IO 2 IO 2 TIME (SECOIIDS)
Figure 1d. Fluid Quality DECLG (CD = 0.4)
50 0
-50 I
-l00 I
)
ce l50
-200
-250 lo-'O 2 5 lOi TIME (SECONDS) 2 I02 2 Figure 2a. Mass Velocity DECLG (Cp = 1.0)
lpp 50 cn 0 I
-50 CD ED O
g -IOO 2*
- l50
-200 lp-'p 2 5 IO' TIME (SECONDS)
IP 2 IP3 Figure 2b. Mass Velocity DECLG (CD = 0.6)
30 IO I
0
- l0
-20 t
!0' Io'. 5 Ip 2 Io'IME (SECONDS)
Figure 2c. Mass Velocity DECLG (Cp = 0.4)
k 4
'40 30 20 I
C4 I
IO ED ce 0
-IO
-20 IO 2 5 IO 2 5 IO 2 10 2 5 IO
-TIME (SECONDS)
Figure 2d. Mass Velocity -. DECLG (Cg = 0.4)
l03 O
I 2
I l02 I
UJ CD U
LI UJ 5.75'0 ED CD Ip I 5 0 l00 200 300 400- 500 TIME (SECONDS)
Figure 3a. Heat Transfer Coefficient DECLG (CD = 1.0)
CO I
CO
S 0'
lO~
5 O
I 5I 2 CV I
lO~
5 U
g 2 los 5.75'.25'o'00 I
I 200 300 TIME (SECONDS)
C)
Figure 3b. Heat Transfer Coefficient DECLG (Cp = 0.6) CO I
/ CO
IO O
I 2
IO~
I 5
ED 2
" IO'-
I 5
IO' IOO 500 TIME (SECONDS)
Figure 3c. Heat Transfer Coefficient DECLG (Cp = 0.4)
IO'OO 200 TIME (SECONDS) 300 tIOO 500 Figure 3d. Heat Transfer Coefficient -- DECLG (Cg = 0.4)
f0. 749- l3 C)
CV II Cl O
lO O
G9 0
Ch ED UJ C/0 Q)
L UJ CL I- 0) 0 I
CD O (visd) sznssszd
2500 2000
~ 1500 1000 500 0 10 15 20 25 TIME (SECONDS)
Figure 4b. Core Pressure DECLG (Cp = 0.6)
2500 2000 I500 0
0 IO 20 30 40 50 TIME (SECONDS)
Figure 4c. Core Pressure DECLG (Cp = 0.4)
O.,
2500 2000 1500 ul 1000 0
0 10 20 30 40 50 TIME (SECONDS)
Figure 4d. Core Pressure DECLG (Cp = 0.4)
s x i04 6 X lO" e X lO" ED 2X l04 0
-2X 10" 0 lo l5 20 25 TIME (SECONDS)
Figure Sa. Break Flow Rate DECLG (CD = 1.0I
IX IOS 8 X I04 6 x i04 e x i04 2X 10" 0
-2X IO'i 0 lo l5 20 25 TIME (SECONDS)
Figure 5b. Break Flow Rate DECLG (CD = 0.6)
II 0
i.i x ios S X IO" 7X IO 5X IO" hC 3 X i04 I x io" 0
-i x i04 0 'I 0 20 30 50 TIME (SECOHDS)
Figure 5c. Break Flow Rate DECLG (CD = 0.4)
9X IO4 7X IO 5 X I04 hC 3 X I04 I X I04 0
-I X IO 0 IO 20 30 50 TIME (SECONDS)
Figure 5d. Break Flow Rate DECLG (Cg = 0.4)
70 50 25 0
o -25
-50
-70 0 IO I5 25 TIME (SECONDS)
Figure 6a. Core Pressure Drop DECLG (CD = 1.0)
70 50 25 0
o -25
-50
-70 0 10 l5 20 25 TIME (SECONDS) figure 6b. Core Pressure Drop DECLG (CD = 0.6)
I 0'
70 50 25 Cl CO 5
UJ 0
-50
-70 0 IO 20 30 QO 50 TIME (SECONDS)
Figure 6c. Core Pressure Drop DECLG (CD = 0.4)
0" 70 50 25 Cl UJ 0
UJ oC) -25
-50
-70 0 IO 20 30 40 50 TIME (SECONDS)
Figure 6d. Core Pressure Drop DECLG {CD = 0.4)
4 2500 o 2000 C5 CO g I500 6'.
75' i5 g I 000 500 0 100 TIME (SECONDS)
Figure 7a. Peak Clad Temperature DECLG (Cp = 1.0)
2500 o 2000 Ch CO 6.
g l500 25'.75' au I 000 500 0 l00 200 500 TIME (SECONDS)
Figure 7b. Peak CIad Temperature DECLG (Cp = 0.6)
0 2500 o 2000 Ch ED I
g l500 I
5 l 000 Ch 500 0
0 l00 300 %00 TIME (SECONDS)
Figure 7c. Peak Clad Temperature DECLG (Cp = 0.4)
4 0
2500 L 2000 C5 CD I 500 5
IOOO CS 500 IOO 200 400 TIME (SECONDS)
Figure 7d. Peak Clad Temperature DECLG (Cp = 0.4)
l l750 l 500 I250 l000 750 500 250 0 l00 300 TIME (SECONDS)
C)
Figure Sa. Fluid Temperature DECLG (Cp = 1.0) CO I
CO
2000 l750 l500 I250 l000 h
750 500 250 IOO 200 300 400 500 TIME (SECONDS)
Figure Sb. Fluid Temperature DECLG (CD = 0.6)
I750 I 500 I000 I~
750 250 200 '00 TIME (SECONDS)
Figure 8c. Fluid Temperature DECl G (CD = 0.4)
2000 I750 I500 I250 I000 750 500 250 200 %00 TIME (SECONDS)
Figure Sd. Fluid Temperature DECLG (CD = 0.4)
l0.709-33 CD CV II CI O
U O
C5 I
E 0
0 Ch CD UJ UJ K
I P
0 O
CD CD CV LA I
7000 5000 2500 0
TOP ED
-2500 BOTTOH
-5000
-7000 IO I5 20 TIME (SECONDS)
Figure 9b. Core Flow Top and Bottom DECLG (CD = 0.6)
7000 5000 2500 BOTTOM 0
C)
-2500
-7000 l0 20 30 50 TIME (SECONDS)
Figure 9c. Core Flow Top and Bottom DECLG (CD = 0.4)
7000 5000 2500 TOP 0
ED BOTTOM I
M
-2500
-7000 0 IO 20 30 50 TIME (SECONDS)
Figure 9d. Core Flow Top and Bottom DECLG (CD = 0.4)
20.0 2. 00 l7. 5 I.75 OOWNCOHER LEVEL l5.0 l.50 ViN I. 25 l0.0 I .00 I
7.5 75 3 CORE LEVEL 5.0 .50 2.5 .25 40 50 75 I00 l25 l50 l75 200 225 TIME (SECONDS)
Figure 10a. Ref lood Transient DECLG (Cp = 1.0)
Downcomer and Core Water Levels
20.0 2.00
- 17. 5 l.75 OOWNCOHER LEVEL le 0 I.50
$ N I. 25 g Io.o I.00 I 7.5 .75 8 CORE LEVEL 5.0 .50 2.5 .25 50 75 100 125 I 50 I75 225 TIME (SECONDS)
Figure 10b. Ref lood Transient DECLG {Cp = 0.6)
Downcomer and Core Water Levels
h C
I'
20.0 2.00
- 17. 5 1.75 OOWXCOHER LEVEL 15.0 1.50
- 12. 5 I. 25 ViX 10.0 1.00 I
cD 7.5 75 CORE LEVEL 5~0 .50 2.5 .25 50 75 100 125 150 175 200 225 T IME (SECONDS)
Figure 10c. Ref lood Transient DECLG (Cp = 0.4)
Downcomer and Core Water Levels
q~
20.0 2.00 I7.5 I.75 OOWNCOMER LEVEL I 5.0 I. 50 l.25 ~
l l2.5 10.0 l.00 I-Cl 7 5 .75 o CORE LEVEL .50 5.0 2.5 .25 50 75 l00 I25 . I50 I75 200 225 TIME {SECONDS)
Figure 10d. Ref lood Transient DECLG (CD = 0.4)
Downcomer and Core Water Levels
V 5000 co %000 3000 2000 l000 0
0 IO 20 25 TIME (SECONDS)
Figure 11a. Accumulator Flow (Blowdown) DECLG (Cg = 1.0)
r 4 6000 5000
%000 3000 2000 I 000 0
0 IO l5 20 25 TIME (SECONOS)
Figure 11b. Accumulator Flow (Slowdown) DECLG (CD = 0.6)
6000 5000 g %000 CD 3000 8 2000 l000 0
0 IO 20 30 50 TIME (SECONDS)
Figure 11c. Accumulator Flow (Blowdown) DECLG (Cg = 0.4)
P
~ >>
U
6000 5000 4000 ED 3000 I
8 2000 I 000 0
0 lp 20 30 40 50 TIME (SECONDS)
Figure 11d. Accumulator Flow (Blowdown) DECLG (CD = 0.4)
0',
l2 10 0
0 80 I20 I60 200 280 320 TIME (SECONDS)
Figure 12a. Pumped ECCS Flow (Ref lood) DECLG (CO = 1.0)
1 l2 IO 80 I20 I60 200 280 320 360 TIME (SECONDS)
Figure 12b. Pumped ECCS Flow (Ref lood) DECLG (Co = 0.6)
IO 6
80 I20 I60 200 280 320 360 TIME (SECONDS)
Figure 12c. Pumped ECCS Flow (Ref lood) DECLG (Cp = 0.4)
I
l0 80 I 20 I60 200 2%0 280 320 360 TIME (SECONDS)
Figure 12d. Pumped ECCS Flow (Ref lood) DECLG (CD = 0.4)
35 30 25 20 l-IS I
ED IO 80 120 I 60 200 2% 280 320 360 TIME (SECONDS)
Figure 13a. Containment Pressure DECLG (Cp = 1.0j
4 f
35 30 25 I
lxJ I5 I
CO CD
. IO 80 I20 I 60 200 2% 280 320 360 TIME (SECONDS)
CO
'4 Figure 13b. Containment Pressure DECLG (Cp = 0.6) CO I
col CD
'1 P p
30 25 a 20 0 '0 80 I20 I 60 200 TIME (SECONDS) 240 280 320 360 400 Figure 13c. Containment Pressure DECLG (Cp = 0.4)
30 25 C9 2 20 LLI l5 I
IO 5
CD 80 I20 I 60 200 2tIO 280 320 360 TIME (SECONDS)
Figure 13d. Containment Pressure pECLG (Cp = 04)
.2 0
0 IO l5 20 25 TIME (SECONDS)
Figure 14a. Core Power Transient DECLG (Cp = 1.0j
.8
.6
.2 0'
IO l5 20 25 TIME (SECONDS)
Figure 14b. Core Power Transient pECLG (Cp = 0.6)
.8
.2 0
0 IO 20 30 "%0 50 TIME (SECONDS)
Figure 14c. Core Power Transient DECLG (Cp = 0.4)
~ 8 0
20 30 %0 50 TIME (SECONDS)
Figure 14d. Core Power Transient DECLG (Cp = 0.4)
LI.5 X I07 3.5 X 107 2.5 X IO
<<C 1.5 X l07.
~ SX IO'
-.5 X l06 0 IO 20 30 50 TIME (SECONDS)
Figure 15. Break Energy Released to Containment ( CD = 0.4)
q, l0.7%9-53 I 000 900 U
0 I
CV I
U-I 800 I
CO I- 700 UJ U
U UJ 600 UJ U
I- 500 CO C/I 400 UJ Ch ED 300 I
UJ 200 l
l00 l00 200 300 TIME (SECONDS)
Figure 16. Containment Wall Condensing Heat Transfer Coefficient (CD = 0.4)
laMC M