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Category:Letter type:L
MONTHYEARL-2024-132, 2024 Population Update Analysis2024-08-13013 August 2024 2024 Population Update Analysis L-2024-129, Relief Request (RR) 14. Limited Coverage Exams Due to Impractical Inservice Inspection Requirements - Fourth Ten-Year Inservice Inspection Program Interval2024-08-0707 August 2024 Relief Request (RR) 14. Limited Coverage Exams Due to Impractical Inservice Inspection Requirements - Fourth Ten-Year Inservice Inspection Program Interval L-2024-121, Subsequent License Renewal Commitment 30 Revision2024-07-30030 July 2024 Subsequent License Renewal Commitment 30 Revision L-2024-123, Submittal of In-Service Inspection Program Owners Activity Report (OAR-1)2024-07-29029 July 2024 Submittal of In-Service Inspection Program Owners Activity Report (OAR-1) 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-2024-110, Environmental Protection Plan Report, Unusual or Important Environmental Event - Manatee in Intake2024-07-10010 July 2024 Environmental Protection Plan Report, Unusual or Important Environmental Event - Manatee in Intake 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-2024-109, Schedule for Subsequent License Renewal Environmental Review2024-07-0303 July 2024 Schedule for Subsequent License Renewal Environmental Review L-2024-104, Response to Request for Additional Information, St. Luce Relief Request (RR) 7, Proposed Alternative in Accordance with 10 CFR 50.55a(z)(1 Extension of Inspection Interval for Reactor Pressure Vessel Welds from 102024-06-26026 June 2024 Response to Request for Additional Information, St. Luce Relief Request (RR) 7, Proposed Alternative in Accordance with 10 CFR 50.55a(z)(1 Extension of Inspection Interval for Reactor Pressure Vessel Welds from 10 L-2024-097, Technical Specification Special Report2024-06-20020 June 2024 Technical Specification Special Report L-2024-102, Official Service List Update2024-06-19019 June 2024 Official Service List Update L-2024-090, Revised Steam Generator Tube Inspection Reports2024-06-0404 June 2024 Revised Steam Generator Tube Inspection Reports L-2024-075, Notification of Improved Standard Technical Specifications (ITS) Implementation2024-05-13013 May 2024 Notification of Improved Standard Technical Specifications (ITS) Implementation L-2024-053, License Amendment Request L-2024-053, Updated Spent Fuel Pool Criticality Analysis2024-04-30030 April 2024 License Amendment Request L-2024-053, Updated Spent Fuel Pool Criticality Analysis L-2024-070, Cycle 32 Core Operating Limits Report2024-04-29029 April 2024 Cycle 32 Core Operating Limits Report L-2024-071, Cycle 27 Core Operating Limits Report2024-04-29029 April 2024 Cycle 27 Core Operating Limits Report L-2024-056, Annual Radiological Environmental Operating Report for Calendar Year 20232024-04-17017 April 2024 Annual Radiological Environmental Operating Report for Calendar Year 2023 L-2024-064, Florida Power & Light Company - 10 CFR 50.46 - Emergency Core Cooling System SBLOCA 30-Day Report2024-04-17017 April 2024 Florida Power & Light Company - 10 CFR 50.46 - Emergency Core Cooling System SBLOCA 30-Day Report 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-2024-004, Relief Request (RR) 7, Proposed Alternative in Accordance with 10 CFR 50.55a(z)(1) Extension of Inspection Interval for Reactor Pressure Vessel Welds from 10 to 20 Years2024-01-18018 January 2024 Relief Request (RR) 7, Proposed Alternative in Accordance with 10 CFR 50.55a(z)(1) Extension of Inspection Interval for Reactor Pressure Vessel Welds from 10 to 20 Years L-2024-002, Withdrawal of Proposed Alternative to American Society of Mechanical Engineers (ASME) Operation and Maintenance (OM) Code for the Auxiliary Feedwater (AFW) 2C Pump2024-01-0808 January 2024 Withdrawal of Proposed Alternative to American Society of Mechanical Engineers (ASME) Operation and Maintenance (OM) Code for the Auxiliary Feedwater (AFW) 2C Pump 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-2023-179, Unusual or Important Environmental Event - Turtle Mortality2023-12-14014 December 2023 Unusual or Important Environmental Event - Turtle Mortality L-2023-168, License Amendment Request Supplement to Revision 2 for the Technical Specifications Conversion to NUREG-1432 Revision 52023-12-12012 December 2023 License Amendment Request Supplement to Revision 2 for the Technical Specifications Conversion to NUREG-1432 Revision 5 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-2023-162, Response to 50.69 2nd Round of Rals2023-11-21021 November 2023 Response to 50.69 2nd Round of Rals L-2023-131, Subsequent License Renewal Application - Second Annual Update2023-09-28028 September 2023 Subsequent License Renewal Application - Second Annual Update L-2023-136, Supplement to License Amendment Request to Adopt 10 CFR 50.69, Risk-Informed Categorization and Treatment of Structures, Systems and Components for Nuclear Power Reactors2023-09-26026 September 2023 Supplement to License Amendment Request to Adopt 10 CFR 50.69, Risk-Informed Categorization and Treatment of Structures, Systems and Components for Nuclear Power Reactors L-2023-122, Corrections to the 2022 Annual Radiological Environmental Operating Report2023-09-20020 September 2023 Corrections to the 2022 Annual Radiological Environmental Operating Report L-2023-127, Correction to the 2022 Annual Radioactive Effluent Release Report2023-09-18018 September 2023 Correction to the 2022 Annual Radioactive Effluent Release Report L-2023-113, Correction to the 2020 Annual Radiological Environmental Operating Report2023-09-14014 September 2023 Correction to the 2020 Annual Radiological Environmental Operating Report L-2023-118, Response to Request for Additional Information Regarding License Amendment Request to Adopt 10 CFR 50.69, Risk-Informed Categorization and Treatment of Structures, Systems and Components for Nuclear Power Reactors2023-09-11011 September 2023 Response to Request for Additional Information Regarding License Amendment Request to Adopt 10 CFR 50.69, Risk-Informed Categorization and Treatment of Structures, Systems and Components for Nuclear Power Reactors L-2023-108, Report of 10 CFR 50.59 Plant Changes2023-09-11011 September 2023 Report of 10 CFR 50.59 Plant Changes L-2023-107, Technical Specification Bases Control Program Periodic Report of Bases Changes TS 6.8.4.j.42023-09-0606 September 2023 Technical Specification Bases Control Program Periodic Report of Bases Changes TS 6.8.4.j.4 L-2023-112, Corrections to the 2021 Annual Radioactive Effluent Release Report2023-09-0606 September 2023 Corrections to the 2021 Annual Radioactive Effluent Release Report 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-2023-105, Preparation and Scheduling of Operator Licensing Examinations2023-08-0303 August 2023 Preparation and Scheduling of Operator Licensing Examinations L-2023-099, Pump Relief Request 10 (PR-10), One-Time Request for an Alternative to the American Society of Mechanical Engineers (ASME) Operation and Maintenance (OM) Code for the Auxiliary Feedwater (AFW) 2C Pump2023-07-26026 July 2023 Pump Relief Request 10 (PR-10), One-Time Request for an Alternative to the American Society of Mechanical Engineers (ASME) Operation and Maintenance (OM) Code for the Auxiliary Feedwater (AFW) 2C Pump L-2023-102, Relief Request PSL2-15-RR-01, Proposed Alternative to ASME Section XI Code Examination Requirements for Reactor Vessel Bottom Area and Piping in Covered Trenches2023-07-26026 July 2023 Relief Request PSL2-15-RR-01, Proposed Alternative to ASME Section XI Code Examination Requirements for Reactor Vessel Bottom Area and Piping in Covered Trenches L-2023-097, Subsequent License Renewal Application Revision 1 - Supplement 62023-07-13013 July 2023 Subsequent License Renewal Application Revision 1 - Supplement 6 L-2023-076, In-Service Inspection Program Owner'S Activity Report (OAR-1)2023-07-11011 July 2023 In-Service Inspection Program Owner'S Activity Report (OAR-1) 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-2023-082, Subsequent License Renewal Application Revision 1, Supplement 52023-06-14014 June 2023 Subsequent License Renewal Application Revision 1, Supplement 5 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-2023-071, 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-2023-059, Subsequent License Renewal Application - Aging Management Requests for Additional Information (RAI) Set 4 Supplemental Response2023-04-21021 April 2023 Subsequent License Renewal Application - Aging Management Requests for Additional Information (RAI) Set 4 Supplemental Response L-2023-055, 2022 Annual Environmental Operating Report2023-04-12012 April 2023 2022 Annual Environmental Operating Report L-2023-041, Annual Radiological Environmental Operating Report for Calendar Year 20222023-04-0404 April 2023 Annual Radiological Environmental Operating Report for Calendar Year 2022 L-2023-051, Report of 10 CFR 50.59 Plant Changes2023-04-0404 April 2023 Report of 10 CFR 50.59 Plant Changes 2024-08-07
[Table view] Category:Report
MONTHYEARL-2024-123, Submittal of In-Service Inspection Program Owners Activity Report (OAR-1)2024-07-29029 July 2024 Submittal of In-Service Inspection Program Owners Activity Report (OAR-1) L-2023-131, Subsequent License Renewal Application - Second Annual Update2023-09-28028 September 2023 Subsequent License Renewal Application - Second Annual Update L-2023-076, In-Service Inspection Program Owner'S Activity Report (OAR-1)2023-07-11011 July 2023 In-Service Inspection Program Owner'S Activity Report (OAR-1) 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-2023-009, Owner'S Activity Report2023-01-31031 January 2023 Owner'S Activity Report ML22227A0532022-08-15015 August 2022 Biological Opinion for the Continued Operation of St. Lucie Nuclear Power Plant ML22124A0112022-04-30030 April 2022 Scoping Summary Report - Final L-2022-046, Subsequent License Renewal Application Revision - Documents WCAP-18623-P/NP Revision 1 Submittal2022-04-13013 April 2022 Subsequent License Renewal Application Revision - Documents WCAP-18623-P/NP Revision 1 Submittal L-2022-015, Relief Request Number 20 - Request for an Alternative to the Requirements of the ASME Code for Examination of Reactor Vessel Closure Head Control Element Drive Mechanism (CEDM) Housing 27 Canopy Seal Weld - RAI2022-01-14014 January 2022 Relief Request Number 20 - Request for an Alternative to the Requirements of the ASME Code for Examination of Reactor Vessel Closure Head Control Element Drive Mechanism (CEDM) Housing 27 Canopy Seal Weld - RAI L-2022-011, Relief Request Number 20 - Request for an Alternative to the Requirements of the ASME Code for Examination of Reactor Vessel Closure Head Control Element Drive Mechanism (CEDM) Housing 27 Canopy Seal Weld2022-01-12012 January 2022 Relief Request Number 20 - Request for an Alternative to the Requirements of the ASME Code for Examination of Reactor Vessel Closure Head Control Element Drive Mechanism (CEDM) Housing 27 Canopy Seal Weld ML22010A0942022-01-0404 January 2022 Trp 29 St. Lucie SLRA - Tank Breakout L-2021-178, Report of 10 CFR 50.59 Plant Changes2021-11-0808 November 2021 Report of 10 CFR 50.59 Plant Changes L-2021-142, Westinghouse Report LTR-REA-21-1-NP, Revision 1, St. Lucie Nuclear Plant, Units 1 and 2, Subsequent License Renewal: Unit 1 Reactor Vessel, Vessel Support, and Bioshield Concrete Exposure Data, May 26, 20212021-08-0303 August 2021 Westinghouse Report LTR-REA-21-1-NP, Revision 1, St. Lucie Nuclear Plant, Units 1 and 2, Subsequent License Renewal: Unit 1 Reactor Vessel, Vessel Support, and Bioshield Concrete Exposure Data, May 26, 2021 ML19252A4002019-09-0909 September 2019 FPL to NRC, Notification of Smalltooth Sawfish Capture at St. Lucie 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 ML18096B3952018-04-0606 April 2018 Exhibit III Estimate of Construction Costs and Exhibit IV Technical Qualifications of Contractors ML18088B1952018-03-29029 March 2018 Hutchinson Island Plant Units 1 and 2 - Chapter 9, Auxiliary Systems and Chapter 10, Steam and Power Conversion System ML18088B1942018-03-29029 March 2018 Hutchinson Island Plant Units 1 and 2 - Chapter 11, Radioactive Waste Management System, Chapter 12, Radiation Protection, and Chapter 13, Conduct of Operations ML18088A0942018-03-29029 March 2018 Unit II Plants ECCS Performance Results L-2017-173, Environmental Protection Plan Report, Unusual or Important Environmental Event - Turtle Mortality - 09/11/2017 Event2017-09-28028 September 2017 Environmental Protection Plan Report, Unusual or Important Environmental Event - Turtle Mortality - 09/11/2017 Event L-2018-081, Kld Engineering, Pc - 2017 Population Update Analysis2017-09-20020 September 2017 Kld Engineering, Pc - 2017 Population Update Analysis L-2017-117, Submittal of SL2-23 Outage, Owner'S Activity Report, Form OAR-12017-06-20020 June 2017 Submittal of SL2-23 Outage, Owner'S Activity Report, Form OAR-1 L-2018-015, Plan of Study 316(b) Implementation2017-04-28028 April 2017 Plan of Study 316(b) Implementation L-2017-015, PWROG-15105-NP PA-MSC-1288 PWR Rv Internals Cold-Work Assessment, Materials Committee.2016-04-30030 April 2016 PWROG-15105-NP PA-MSC-1288 PWR Rv Internals Cold-Work Assessment, Materials Committee. ML16084A6162016-03-24024 March 2016 Submittal of Biological Opinion for the Continued Operation of St. Lucie Nuclear Power Plant, Units 1 and 2 in St. Lucie County, Florida ML16063A0072016-02-26026 February 2016 Participation in Additional Work Under the Support for Applicant Action Items 1, 2, and 7 from the Final Safety Evaluation on MRP-227, Revision 0 PA-MSC-0983 R2 Cafeteria Task 8 and Acceptance Criteria for Measurement Of.. ML15352A0532016-01-0707 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 Revaluations for Recommendation 2.1 of the Near-Term Task Force Review of Insights L-2015-297, Status of Required Actions for EA-12-049 Order Modifying Licenses with Regard to Requirements for Mitigation Strategies for Beyond-Design-Basis External Events, and Submittal of Site FLEX Final Integrated Plan2015-12-10010 December 2015 Status of Required Actions for EA-12-049 Order Modifying Licenses with Regard to Requirements for Mitigation Strategies for Beyond-Design-Basis External Events, and Submittal of Site FLEX Final Integrated Plan L-2015-300, ANP-3352NP, Revision 1, Transition License Amendment Request, Technical Report.2015-11-30030 November 2015 ANP-3352NP, Revision 1, Transition License Amendment Request, Technical Report. ML15314A1602015-10-29029 October 2015 St. Lucie, Units 1 and 2 - License Renewal Commitment, Submittal of Pressurizer Surge Line Welds Inspection Program L-2015-221, Report of 10 CFR 50.59 Plant Changes2015-10-16016 October 2015 Report of 10 CFR 50.59 Plant Changes ML15240A1542015-09-0808 September 2015 Staff Observations of Sump Strainer Head Loss Testing at Alden Laboratory for Generic Safety Issue 191 L-2015-206, ANP-3428NP, Revision 0, St. Lucie Unit 2 Fuel Transition: Response to SNPB-RAI-1, Attachment 4 to L-2015-2062015-07-31031 July 2015 ANP-3428NP, Revision 0, St. Lucie Unit 2 Fuel Transition: Response to SNPB-RAI-1, Attachment 4 to L-2015-206 L-2015-177, Fuel Transition Small Break LOCA Summary Report, ANP-3345NP, Revision 12015-06-30030 June 2015 Fuel Transition Small Break LOCA Summary Report, ANP-3345NP, Revision 1 L-2015-143, Status of Required Actions for EA-12-049 Order Modifying Licenses with Regard to Requirements for Mitigation Strategies for Beyond-Design-Basis External Events2015-05-14014 May 2015 Status of Required Actions for EA-12-049 Order Modifying Licenses with Regard to Requirements for Mitigation Strategies for Beyond-Design-Basis External Events L-2015-272, 1301103.401, Revision 0, Flaw Tolerance Evaluation of St. Lucie Surge Line Welds Using ASME Code Section XI, Appendix L, May 20152015-05-0808 May 2015 1301103.401, Revision 0, Flaw Tolerance Evaluation of St. Lucie Surge Line Welds Using ASME Code Section XI, Appendix L, May 2015 L-2016-052, TN-5696-00-02, Revision 0, Technical Note Assessment of Laboratory PWSCC Crack Growth Rate Data Compiled for Alloys 690, 52, and 152 with Regard to Factors of Improvement (Foi) Versus Alloys 600 and 182.2015-03-31031 March 2015 TN-5696-00-02, Revision 0, Technical Note Assessment of Laboratory PWSCC Crack Growth Rate Data Compiled for Alloys 690, 52, and 152 with Regard to Factors of Improvement (Foi) Versus Alloys 600 and 182. L-2015-091, ANP-3396NP, Revision 0, Fuel Transition Supplemental Information to Support the LAR2015-03-31031 March 2015 ANP-3396NP, Revision 0, Fuel Transition Supplemental Information to Support the LAR L-2015-093, Acceptance Criteria for Emergency Core Cooling Systems for Light Water Nuclear Power Reactors 10 CFR 50.46 Annual Report2015-03-24024 March 2015 Acceptance Criteria for Emergency Core Cooling Systems for Light Water Nuclear Power Reactors 10 CFR 50.46 Annual Report L-2015-048, St. Lucie, Units 1 and 2 - Submittal of Revision 0 to FPL-072-PR-002, Flooding Hazards Reevaluation Report2015-03-10010 March 2015 St. Lucie, Units 1 and 2 - Submittal of Revision 0 to FPL-072-PR-002, Flooding Hazards Reevaluation Report ML15083A2642015-03-10010 March 2015 St. Lucie, Units 1 and 2 - Submittal of Revision 0 to FPL-072-PR-002, Flooding Hazards Reevaluation Report L-2015-048, to FPL-072-PR-002, Flooding Hazards Reevaluation Report for St. Lucie Nuclear Power Plant Units 1 & 2. Cover Page to Page 1092015-02-0606 February 2015 to FPL-072-PR-002, Flooding Hazards Reevaluation Report for St. Lucie Nuclear Power Plant Units 1 & 2. Cover Page to Page 109 L-2015-048, to FPL-072-PR-002, Flooding Hazards Reevaluation Report for St. Lucie Nuclear Power Plant Units 1 & 2. Table 2-1 Through Figure 4-92015-02-0606 February 2015 to FPL-072-PR-002, Flooding Hazards Reevaluation Report for St. Lucie Nuclear Power Plant Units 1 & 2. Table 2-1 Through Figure 4-9 ML15083A2662015-02-0606 February 2015 to FPL-072-PR-002, Flooding Hazards Reevaluation Report for St. Lucie Nuclear Power Plant Units 1 & 2. Figure 4-10 Through the End L-2015-048, to FPL-072-PR-002, Flooding Hazards Reevaluation Report for St. Lucie Nuclear Power Plant Units 1 & 2. Figure 4-10 Through the End2015-02-0606 February 2015 to FPL-072-PR-002, Flooding Hazards Reevaluation Report for St. Lucie Nuclear Power Plant Units 1 & 2. Figure 4-10 Through the End ML15083A2652015-02-0606 February 2015 to FPL-072-PR-002, Flooding Hazards Reevaluation Report for St. Lucie Nuclear Power Plant Units 1 & 2. Table 2-1 Through Figure 4-9 L-2014-366, ANP-3352NP, Revision 0, St. Luice, Unit 2, Fuel Transition License Amendment Request, Technical Report2014-12-31031 December 2014 ANP-3352NP, Revision 0, St. Luice, Unit 2, Fuel Transition License Amendment Request, Technical Report ML14338A5542014-12-0404 December 2014 NRC-2013- TN2986-NRC 2014 St. Lucie ML14338A5552014-12-0404 December 2014 NRC-2013-TN3079-NRC 2014 St. Lucie License Renewal L-2014-125, Report of 10 CFR 50.59 Plant Changes2014-05-0606 May 2014 Report of 10 CFR 50.59 Plant Changes 2024-07-29
[Table view] Category:Miscellaneous
MONTHYEARL-2024-123, Submittal of In-Service Inspection Program Owners Activity Report (OAR-1)2024-07-29029 July 2024 Submittal of In-Service Inspection Program Owners Activity Report (OAR-1) L-2023-131, Subsequent License Renewal Application - Second Annual Update2023-09-28028 September 2023 Subsequent License Renewal Application - Second Annual Update L-2023-076, In-Service Inspection Program Owner'S Activity Report (OAR-1)2023-07-11011 July 2023 In-Service Inspection Program Owner'S Activity Report (OAR-1) 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-2023-009, Owner'S Activity Report2023-01-31031 January 2023 Owner'S Activity Report ML22227A0532022-08-15015 August 2022 Biological Opinion for the Continued Operation of St. Lucie Nuclear Power Plant ML22124A0112022-04-30030 April 2022 Scoping Summary Report - Final ML22010A0942022-01-0404 January 2022 Trp 29 St. Lucie SLRA - Tank Breakout L-2021-178, Report of 10 CFR 50.59 Plant Changes2021-11-0808 November 2021 Report of 10 CFR 50.59 Plant Changes ML19252A4002019-09-0909 September 2019 FPL to NRC, Notification of Smalltooth Sawfish Capture at St. Lucie L-2017-173, Environmental Protection Plan Report, Unusual or Important Environmental Event - Turtle Mortality - 09/11/2017 Event2017-09-28028 September 2017 Environmental Protection Plan Report, Unusual or Important Environmental Event - Turtle Mortality - 09/11/2017 Event L-2017-117, Submittal of SL2-23 Outage, Owner'S Activity Report, Form OAR-12017-06-20020 June 2017 Submittal of SL2-23 Outage, Owner'S Activity Report, Form OAR-1 ML15352A0532016-01-0707 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 Revaluations for Recommendation 2.1 of the Near-Term Task Force Review of Insights L-2015-297, Status of Required Actions for EA-12-049 Order Modifying Licenses with Regard to Requirements for Mitigation Strategies for Beyond-Design-Basis External Events, and Submittal of Site FLEX Final Integrated Plan2015-12-10010 December 2015 Status of Required Actions for EA-12-049 Order Modifying Licenses with Regard to Requirements for Mitigation Strategies for Beyond-Design-Basis External Events, and Submittal of Site FLEX Final Integrated Plan ML15314A1602015-10-29029 October 2015 St. Lucie, Units 1 and 2 - License Renewal Commitment, Submittal of Pressurizer Surge Line Welds Inspection Program L-2015-221, Report of 10 CFR 50.59 Plant Changes2015-10-16016 October 2015 Report of 10 CFR 50.59 Plant Changes ML15240A1542015-09-0808 September 2015 Staff Observations of Sump Strainer Head Loss Testing at Alden Laboratory for Generic Safety Issue 191 L-2015-143, Status of Required Actions for EA-12-049 Order Modifying Licenses with Regard to Requirements for Mitigation Strategies for Beyond-Design-Basis External Events2015-05-14014 May 2015 Status of Required Actions for EA-12-049 Order Modifying Licenses with Regard to Requirements for Mitigation Strategies for Beyond-Design-Basis External Events ML15083A2642015-03-10010 March 2015 St. Lucie, Units 1 and 2 - Submittal of Revision 0 to FPL-072-PR-002, Flooding Hazards Reevaluation Report L-2015-048, St. Lucie, Units 1 and 2 - Submittal of Revision 0 to FPL-072-PR-002, Flooding Hazards Reevaluation Report2015-03-10010 March 2015 St. Lucie, Units 1 and 2 - Submittal of Revision 0 to FPL-072-PR-002, Flooding Hazards Reevaluation Report ML14338A5552014-12-0404 December 2014 NRC-2013-TN3079-NRC 2014 St. Lucie License Renewal ML14338A5542014-12-0404 December 2014 NRC-2013- TN2986-NRC 2014 St. Lucie L-2014-125, Report of 10 CFR 50.59 Plant Changes2014-05-0606 May 2014 Report of 10 CFR 50.59 Plant Changes ML13360A2022013-12-12012 December 2013 EPA Echo Report Martin County, Fl L-2013-193, Transition to 10 CFR 50.48(c) - NFPA 805 Performance-Based Standard for Fire Protection for Light Water Reactor Generating Plants (2001 Edition) Acceptance Review Clarification Response2013-06-14014 June 2013 Transition to 10 CFR 50.48(c) - NFPA 805 Performance-Based Standard for Fire Protection for Light Water Reactor Generating Plants (2001 Edition) Acceptance Review Clarification Response L-2013-175, Report of 10 CFR 50.59 Plant Changes2013-05-23023 May 2013 Report of 10 CFR 50.59 Plant Changes L-2012-437, Standardization of Senior Reactor Operator Fuel Handling Duties2012-12-27027 December 2012 Standardization of Senior Reactor Operator Fuel Handling Duties L-2012-105, Response to Nuclear Performance and Code Review Branch Request for Additional Information Identified During an Audit of Analyses Supporting the Extended Power Uprate License Amendment Request2012-03-15015 March 2012 Response to Nuclear Performance and Code Review Branch Request for Additional Information Identified During an Audit of Analyses Supporting the Extended Power Uprate License Amendment Request ML1026405722010-09-14014 September 2010 E-mail St. Lucie Capture Summary Update for January Through August 2010 ML1019304172010-05-0606 May 2010 Tritium Database Report ML1012704392010-05-0505 May 2010 Y020100187 - List of Historical Leaks and Spills at U.S. Commercial Nuclear Power Plants L-2010-059, Florida Dept of Environmental Protection - Wastewater Application Form 2CS - Permit to Discharge Process Wastewater from New or Existing Industrial Wastewater Facilities to Surface Water2010-03-18018 March 2010 Florida Dept of Environmental Protection - Wastewater Application Form 2CS - Permit to Discharge Process Wastewater from New or Existing Industrial Wastewater Facilities to Surface Water ML1008304422010-03-18018 March 2010 Florida Dept of Environmental Protection - Wastewater Application Form 2CS - Permit to Discharge Process Wastewater from New or Existing Industrial Wastewater Facilities to Surface Water L-2009-101, Report of 10 CFR 50.59 Plant Changes, for the Period of May 28, 2007 Through November 21, 20082009-04-22022 April 2009 Report of 10 CFR 50.59 Plant Changes, for the Period of May 28, 2007 Through November 21, 2008 L-2008-261, Technical Specification Special Report Inoperable Containment Sump Wide Range Level Channel B2008-12-16016 December 2008 Technical Specification Special Report Inoperable Containment Sump Wide Range Level Channel B ML0807010172008-03-10010 March 2008 EPRI Reports (St. Lucie Pressurizer Nozzle DM Weld Examination Project Internal Office Report) L-2008-030, Supplemental Response to NRC Generic Letter 2004-02, Potential Impact of Debris Blockage on Emergency Recirculation During Design Basis Accidents at Pressurized-Water Reactors2008-02-27027 February 2008 Supplemental Response to NRC Generic Letter 2004-02, Potential Impact of Debris Blockage on Emergency Recirculation During Design Basis Accidents at Pressurized-Water Reactors NOC-AE-07002231, Request for Relaxation of Requirements from Revision 1 of Order EA-03-009 Establishing Interim Inspection Requirements for Reactor Pressure Vessel Head Penetrations (Relief Request RR-ENG-2-46)2007-11-0707 November 2007 Request for Relaxation of Requirements from Revision 1 of Order EA-03-009 Establishing Interim Inspection Requirements for Reactor Pressure Vessel Head Penetrations (Relief Request RR-ENG-2-46) L-2007-167, Report of 10 CFR 50.59 Plant Changes During the Period of December 19, 2005 Through May 27, 20072007-10-29029 October 2007 Report of 10 CFR 50.59 Plant Changes During the Period of December 19, 2005 Through May 27, 2007 L-2007-168, Technical Specifications Bases Control Program Periodic Report of Bases Changes TS 6.8.4.j.42007-10-29029 October 2007 Technical Specifications Bases Control Program Periodic Report of Bases Changes TS 6.8.4.j.4 L-2007-137, Technical Specification Special Report, Inoperable Containment Sump Wide Range Level Channel B2007-08-31031 August 2007 Technical Specification Special Report, Inoperable Containment Sump Wide Range Level Channel B L-2007-029, Semi-Annual Fitness-For-Duty Program Performance Data for the Six-Month Period Ending December 31, 20062007-02-26026 February 2007 Semi-Annual Fitness-For-Duty Program Performance Data for the Six-Month Period Ending December 31, 2006 L-2007-020, Annual Summary of Commitment Changes Implemented Without Prior NRC Notification for Calendar Year 20062007-02-0606 February 2007 Annual Summary of Commitment Changes Implemented Without Prior NRC Notification for Calendar Year 2006 L-2006-078, Proposed License Amendment, Containment Spray Nozzle Surveillance Change2006-10-19019 October 2006 Proposed License Amendment, Containment Spray Nozzle Surveillance Change L-2006-217, Day Post Outage Steam Generator Report Technical Specification 4.4.5.5.c2006-09-20020 September 2006 Day Post Outage Steam Generator Report Technical Specification 4.4.5.5.c L-2006-152, NRC Order EA-03-009 - Reactor Vessel Head and Vessel Head Penetration Nozzle Inspection Results SL2-162006-08-10010 August 2006 NRC Order EA-03-009 - Reactor Vessel Head and Vessel Head Penetration Nozzle Inspection Results SL2-16 L-2006-180, Sation; St. Lucie, Units 1 and 2 and Turkey Point, Units 3 and 4 - Groundwater Questionnaire2006-07-31031 July 2006 Sation; St. Lucie, Units 1 and 2 and Turkey Point, Units 3 and 4 - Groundwater Questionnaire L-2006-167, Environmental Protection Plan Report Responses to FDEP Comments on FPL Proposal for Information Collection Clean Water Act Section 316(b)2006-07-0606 July 2006 Environmental Protection Plan Report Responses to FDEP Comments on FPL Proposal for Information Collection Clean Water Act Section 316(b) L-2006-139, Report of 10 CFR 50.59 Plant Changes2006-06-16016 June 2006 Report of 10 CFR 50.59 Plant Changes L-2006-063, Technical Specification Special Report, Radwaste Building Exhaust System, Plant Ventilation Stack Particulate Iodine and Noble Gas (Sping) Radiation - Out of Service2006-02-22022 February 2006 Technical Specification Special Report, Radwaste Building Exhaust System, Plant Ventilation Stack Particulate Iodine and Noble Gas (Sping) Radiation - Out of Service 2024-07-29
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w May 6, 2014 L-2014-125 10 CFR 50.59(d)U. S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, D. C. 20555 Re: St. Lucie Unit 1 Docket No. 50-335 Report of 10 CFR 50.59 Plant Changes Pursuant to 10 CFR 50.59(d)(2), the enclosed report contains a brief description of any changes, tests, and experiments, including a summary of the 50.59 evaluation of each which were made on Unit 1 during the period of April 22, 2012 through November 9, 2013. This submittal correlates with the information included in Amendment 26 of the Updated Final Safety Analysis Report to be submitted under separate cover.Please contact us should there by any questions regarding this information.
Sincerely, Eric S. Katzman Licensing Manager St. Lucie Plant ESK/tlt Enclosure 1:jZ7E4-Florida Power & Light Company 6501 S. Ocean Drive, Jensen Beach, FL 34957 St. Lucie Unit 1 L-2014-125 Docket No. 50-335 Enclosure ST. LUCIE UNIT 1 DOCKET NUMBER 50-335 CHANGES, TESTS AND EXPERIMENTS MADE AS ALLOWED BY 10 CFR 50.59 FOR THE PERIOD OF APRIL 22, 2012 THROUGH NOVEMBER 9, 2013 (13 PAGES INCLUDING COVER)I St. Lucie Unit I L-2014-125 Docket No. 50-335 Enclosure INTRODUCTION This report is submitted in accordance with 10 CFR 50.59 (d)(2),which requires that: i) changes in the facility as described in the SAR;ii) changes in procedures as described in the SAR; and iii) tests and experiments not described in the SAR that are conducted without prior Commission approval be reported to the Commission in accordance with 10 CFR 50.90 and 50.4. This report is intended to meet these requirements for the period of April 22, 2012 through November 9, 2013.This report is typically divided into three (3) sections.
First, changes to the facility as described in the Updated Final Safety Analysis Report (UFSAR) performed by a Permanent Modification.
Second, changes to the facility/procedures as described in the UFSAR, or tests/experiments not described in the UFSAR, which are not performed by a Permanent Modification.
And third, a summary of any Fuel Reload 10 CFR 50.59 evaluation.
Sections 1, 2 and 3 summarize specific 10 CFR 50.59 evaluations that evaluated the specific change(s).
Each of these 10 CFR 50.59 evaluations concluded that the change does not require a change to the plant technical specifications, and prior NRC approval is not required.2 St. Lucie Unit 1 Docket No. 50-335 L-2014-125 Enclosure TABLE OF CONTENTS SECTION 1 PERMANENT MODIFICATIONS PAGE EC 277049, REVISION 0 EC 275043, REVISION 0 SECTION 2 PERMANENT REMOVAL OF ST. LUCIE UNIT 1 RCP 1A1 WHIP (CABLE) RESTRAINTS PERMANENT REMOVAL OF ST. LUCIE UNIT 1 RCP 1A2 WHIP (CABLE) RESTRAINTS 10 CFR 50.59 EVALUATIONS 5 8 11 NONE SECTION 3 FUEL RELOAD EVALUATIONS ST. LUCIE UNIT 1 CYCLE 25 RELOAD EC 277290, REVISION 0 13 3 St. Lucie Unit I Docket No. 50-335 L-2014-125 Enclosure SECTION 1 PLANT CHANGE / MODIFICATIONS 4
St. Lucie Unit I L-2014-125 Docket No. 50-335 Enclosure EC 277049, REVISION 0 PERMANENT REMOVAL OF ST. LUCIE UNIT I RCP IAI WHIP (CABLE) RESTRAINTS
SUMMARY
- To allow for future maintenance of the reactor coolant pump the RCP 1Al, whip (cable)restraints shall be permanently removed. In addition to removing an interference that impedes the disassembly of the pump it will also reduce the radiological dose required to reinstall the 4-inch cables.The St. Lucie Unit I Construction Permit was issued on July 1, 1970 and an Operating Licensing was licensed in March 1976. Prior to 1986, General Design Criterion (GDC) 4,"Environmental and Missile Design Bases," required that systems and components important to safety be appropriately protected against dynamic effects, including the effects of missiles, pipe whipping, and discharge fluids, which may result in equipment failures.
In accordance with NRC Branch Technical Position ASB 3-1, plants for which construction permits were tendered before July 1, 1993, and operating licenses were issued after July 1, 1975, should follow the guidance of Appendix B of ASB 3-1 (letter by A Giambusso, December 1972, General Information Required for Consideration of the Effects of a Piping System Break Outside Containment")
and also provide moderate energy piping failure analyses in accordance with Branch Technical Position ASB 3-1. Accordingly, the original St. Lucie Unit I design bases considered all dynamic effects (missile generation, pipe whipping, pipe break reaction forces, jet impingement forces, compartment, sub-compartment and cavity pressurizations and decompression waves with the ruptured pipe) and all environmental effects (pressure, temperature, humidity, and flooding) resulting from arbitrary intermediate pipe ruptures.Circa- 1986, GDC 4 was revised to read: "Environmental and dynamic effects design bases. Structures, systems, and components important to safety shall be designed to accommodate the effects of and to be compatible with the environmental conditions associated with normal operation, maintenance, testing, and postulated accidents, including loss-of-coolant accident.
These structures, systems, and components shall be appropriately protected against dynamic effects, including the effects of missiles, pipe whipping, and discharging fluids, that may result from equipment failures and from events and conditions outside the nuclear power unit.However, dynamic effects associated with postulated pipe ruptures in nuclear power units may be excluded from the design basis when analyses reviewed and approved by the Commission demonstrate that the probability of fluid system piping rupture is extremely low under conditions consistent with the design basis for the piping." Consistent with the revision to GDC 4, on June 19, 1987, the NRC issued Generic Letter 87-11, Relaxation in Arbitrary Intermediate Pipe Rupture Requirements which finalized a revision to Branch Technical Position (BTP) MEB 3-1 of Standard Review Plan (SRP)5 St. Lucie Unit 1 L-2014-125 Docket No. 50-335 Enclosure Section 3.6.2 in NUREG-0800.
The revisions to BTP MEB 3-1 and SRP 3.6.2 eliminated all dynamic effects and all environmental effects resulting from arbitrary intermediate pipe ruptures.
This action allows the elimination of pipe whip restraints and jet impingement shields placed to mitigate the effects of arbitrary intermediate pipe ruptures, and other related changes.On October 30, 1990, the NRC accepted Topical Report CEN-367, "Leak-Before-Break Evaluation of Primary Loop Piping in Combustion Engineering Designed Nuclear Steam Supply Systems, which was submitted for staff review by Combustion Engineering Owners Group (CEOG) letter dated November 20, 1987. FPL was a participating CEOG member and St. Lucie Units 1 and 2 were included in the bounding analyses submitted.
By letter to the NRC dated August 26, 1992, FPL proposed to eliminate the dynamic effects associated with high energy pipe rupture in the reactor coolant system piping from the licensing and design bases of St. Lucie Units 1 and 2 by the application of leak-before-break (LBB) technology.
This change to the licensing and design bases is permitted by revised GDC-4 of Appendix A to 10 CFR 50.By NRC letter dated March 5, 1993, the staff concluded that since the St. Lucie Units are bounded by the CEOG analyses and the leakage detection systems are capable of detecting the specified leakage rate, the dynamic effects associated with postulated pipe breaks in the primary coolant system piping can be excluded from the licensing and design bases of the St.Lucie Units.The Unit 1 UFSAR was updated to incorporate the effects of the staff's approval.
Section 3.1.4 was revised to read: Due to the application of leak before break methodology to the RCS hot and cold leg piping, the dynamic effects of a loss of coolant accident do not have to be considered.
A technical evaluation was performed to demonstrate that the probability or likelihood of large pipe breaks occurring in the primary coolant loops is sufficiently low that they need not be a design basis.UFSAR Section 3.6.2 was revised to include the following acceptance criteria: It should be noted that circumferential (guillotine) and longitudinal (slot) breaks in RCS hot and cold leg piping are no longer considered a design basis for GDC 4 (NRC acceptance letter for leak-before-break).
The primary loop piping is not susceptible to failure from the effects of corrosion, water hammer, fatigue, brittle fracture or indirect causes such as missiles or failure of nearby components.
As a result, the mechanical/structural loadings associated with the dynamic effects of a large hot or cold leg break need not be considered.
The proposed permanent removal of the upper two 4-inch reactor coolant pump RCP IAI whip (cable) restraints meets the acceptance criteria found in Generic Letter 87-11. Also as documented in NRC letter dated March 5, 1993, the NRC staff has concluded that since the St. Lucie Units are bounded by the CEOG analyses and the leakage detection systems are 6 St. Lucie Unit I L-2014-125 Docket No. 50-335 Enclosure capable of detecting the specified leakage rate, the dynamic effects associated with postulated pipe breaks in the primary coolant system piping can be excluded from the licensing and design bases of the St. Lucie Units. The permanent removal of the upper RCP lAl whip restraints meets the acceptance criteria of BTP 3-1, as contained in SRP Section 3.6.2, in that absent the whip restraints the primary coolant system piping continues to meet the applicable ASME Code design requirements.
There are no Technical Specifications that address the restraints.
However, as discussed in the Safety Evaluation by the NRC on Leak-Before-Break (LBB) Technology, the acceptance of LBB is based on a leakage detection system consistent with Regulatory Guide 1.45,"Reactor Coolant Pressure Boundary Leakage Detection Systems." Technical Specification Section 3/4.4.6.1 addresses the RCS leakage detection system and no changes to this section or any other section is required as a result of the removal of the cable restraints.
A License Amendment Request is not required.
The RCS whip restraints were installed during initial construction of St. Lucie Unit 1 to mitigate the dynamic effects associate with postulated RCS hot and cold leg piping breaks. In 1993 the dynamic effects associated with postulated pipe ruptures in RCS hot and cold leg piping were excluded from the design and licensing bases of St. Lucie Unit 1, as described in UFSAR Section 3.6. Because the proposed change does not require a change to the technical specifications and does not meet any of the criteria in 10 CFR 50.59(c)(2), the change can be made without obtaining a license amendment pursuant to 10 CFR 50.90.7 St. Lucie Unit 1 L-2014-125 Docket No. 50-335 Enclosure EC 275043, REVISION 0 PERMANENT REMOVAL OF ST. LUCIE UNIT I RCP 1A2 WHIP (CABLE) RESTRAINTS
SUMMARY
- To allow for future maintenance of the reactor coolant pump the RCP 1A2 whip (cable)restraints shall be permanently removed. In addition to removing an interference that impedes the disassembly of the pump it will also reduce the radiological dose required to reinstall the 4-inch cables.The St. Lucie Unit 1 Construction Permit was issued on July 1, 1970 and an Operating Licensing was licensed in March 1976. Prior to 1986, General Design Criterion (GDC) 4, "Environmental and Missile Design Bases," required that systems and components important to safety be appropriately protected against dynamic effects, including the effects of missiles, pipe whipping, and discharge fluids, which may result in equipment failures.
In accordance with NRC Branch Technical Position ASB 3-1, plants for which construction permits were tendered before July 1, 1993, and operating licenses were issued after July 1, 1975, should follow the guidance of Appendix B of ASB 3-1 (letter by A Giambusso, December 1972, General Information Required for Consideration of the Effects of a Piping System Break Outside Containment")
and also provide moderate energy piping failure analyses in accordance with Branch Technical Position ASB 3-1. Accordingly, the original St. Lucie Unit I design bases considered all dynamic effects (missile generation, pipe whipping, pipe break reaction forces, jet impingement forces, compartment, sub-compartment and cavity pressurizations and decompression waves with the ruptured pipe) and all environmental effects (pressure, temperature, humidity, and flooding)resulting from arbitrary intermediate pipe ruptures.Circa- 1986, GDC 4 was revised to read: "Environmental and dynamic effects design bases". Structures, systems, and components important to safety shall be designed to accommodate the effects of and to be compatible with the environmental conditions associated with normal operation, maintenance, testing, and postulated accidents, including loss-of-coolant accident.
These structures, systems, and components shall be appropriately protected against dynamic effects, including the effects of missiles, pipe whipping, and discharging fluids, that may result from equipment failures and from events and conditions outside the nuclear power unit. However, dynamic effects associated with postulated pipe ruptures in nuclear power units may be excluded from the design basis when analyses reviewed and approved by the Commission demonstrate that the probability of fluid system piping rupture is extremely low under conditions consistent with the design basis for the piping." Consistent with the revision to GDC 4, on June 19, 1987, the NRC issued Generic Letter 87-11, Relaxation in Arbitrary Intermediate Pipe Rupture Requirements which finalized a revision to Branch Technical Position (BTP) MEB 3-1 of Standard Review Plan (SRP) Section 3.6.2 in 8 St. Lucie Unit I L-2014-125 Docket No. 50-335 Enclosure NUREG-0800.
The revisions to BTP MEB 3-1 and SRP 3.6.2 eliminated all dynamic effects and all environmental effects resulting from arbitrary intermediate pipe ruptures.
This action allows the elimination of pipe whip restraints and jet impingement shields placed to mitigate the effects of arbitrary intermediate pipe ruptures, and other related changes.On October 30, 1990, the NRC accepted Topical Report CEN-367, "Leak-Before-Break Evaluation of Primary Loop Piping in Combustion Engineering Designed Nuclear Steam Supply Systems, which was submitted for staff review by Combustion Engineering Owners Group (CEOG) letter dated November 20, 1987. FPL was a participating CEOG member and St. Lucie Units 1 and 2 were included in the bounding analyses submitted.
By letter to the NRC dated August 26, 1992, FPL proposed to eliminate the dynamic effects associated with high energy pipe rupture in the reactor coolant system piping from the licensing and design bases of St. Lucie Units I and 2 by the application of leak-before-break (LBB)technology.
This change to the licensing and design bases is permitted by revised GDC-4 of Appendix A to 10 CFR 50.By NRC letter dated March 5, 1993, the staff concluded that since the St. Lucie Units are bounded by the CEOG analyses and the leakage detection systems are capable of detecting the specified leakage rate, the dynamic effects associated with postulated pipe breaks in the primary coolant system piping can be excluded from the licensing and design bases of the St. Lucie Units. Furthermore by NRC letter dated July 9, 2012, NRC staff concludes that Leak Before Break remains valid for SL- 1 under EPU conditions.
The Unit 1 UFSAR was updated to incorporate the effects of the staff's approval.
Section 3.1.4 was revised to read: Due to the application of leak before break methodology to the RCS hot and cold leg piping, the dynamic effects of a loss of coolant accident do not have to be considered.
A technical evaluation was performed to demonstrate that the probability or likelihood of large pipe breaks occurring in the primary coolant loops is sufficiently low that they need not be a design basis.UFSAR Section 3.6.2 was revised to include the following acceptance criteria: It should be noted that circumferential (guillotine) and longitudinal (slot) breaks in RCS hot and cold leg piping are no longer considered a design basis for GDC 4 (NRC acceptance letter for leak-before-break).
The primary loop piping is not susceptible to failure from the effects of corrosion, water hammer, fatigue, brittle fracture or indirect causes such as missiles or failure of nearby components.
As a result, the mechanical/structural loadings associated with the dynamic effects of a large hot or cold leg break need not be considered.
The proposed permanent removal of the upper two 4-inch reactor coolant pump RCP 1A2 whip (cable) restraints meets the acceptance criteria found in Generic Letter 87-11. Also as documented in NRC letter dated March 5, 1993, the NRC staff has concluded that since the St.Lucie Units are bounded by the CEOG analyses and the leakage detection systems are capable of detecting the specified leakage rate, the dynamic effects associated with postulated pipe breaks in 9 St. Lucie Unit I L-2014-125 Docket No. 50-335 Enclosure the primary coolant system piping can be excluded from the licensing and design bases of the St.Lucie Units. The EPU leak-before-break evaluation is based on evaluation CEN-367-A.
The primary loop piping normal operating, SSE and pressure loads due to the EPU conditions were used in the EPU evaluation.
The results of the evaluation demonstrated that leak-before-break recommended margins for the primary loop piping continue to be satisfied for the EPU conditions.
Calculation CN-MRCDA-09-68 provides an analysis of the RCP and surrounding components to demonstrate the RCP HELB cable restraints for pumps lAl, 1A2, 1B1 and 1B2 can be permanently removed. In addition, calculation CN-MRCDA-09-06 documents that sufficient leak-before-break margin on crack stability is maintained on the main coolant loop hot and cold leg pipes under EPU conditions.
NRC letter dated July 9, 2012, concludes that Leak Before Break remains valid for SL-1 under EPU conditions.
The permanent removal of the upper RCP 1A2 whip restraints meets the acceptance criteria of BTP 3-1, as contained in SRP Section 3.6.2, in that absent the whip restraints the primary coolant system piping continues to meet the applicable ASME Code design requirements.
There are no Technical Specifications that address the restraints.
However, as discussed in the Safety Evaluation by the NRC on Leak-Before-Break (LBB) Technology, the acceptance of LBB is based on a leakage detection system consistent with Regulatory Guide 1.45, "Reactor Coolant Pressure Boundary Leakage Detection Systems." Technical Specification Section 3/4.4.6.1 addresses the RCS leakage detection system and no changes to this section or any other section is required as a result of the removal of the cable restraints.
A License Amendment Request is not required.
The RCS whip restraints were installed during initial construction of St. Lucie Unit I to mitigate the dynamic effects associated with postulated RCS hot and cold leg piping breaks. In 1993 the dynanmic effects associated with postulated pipe ruptures in RCS hot and cold leg piping were excluded from the design and licensing bases of St.Lucie Unit 1, as described in UFSAR Section 3.6. In 2012, it was concluded that the leak before break analysis remains valid for EPU conditions.
Because the proposed change does not require a change to the technical specifications and does not meet any of the criteria in 10 CFR 50.59(c)(2), the change can be made without obtaining a license amendment pursuant to 10 CFR 50.90.10 St. Lucie Unit I Docket No. 50-335 L-2014-125 Enclosure SECTION 2 50.59 EVALUATIONS For the time period of this report, there were no changes to the facility (outside of the plant design modifications discussed in Section 1) as described in the Updated Final Safety Analysis Report (UFSAR) performed by a 10 CFR 50.59 Evaluation.
11 St. Lucie Unit I Docket No. 50-335 L-2014-125 Enclosure SECTION 3 CORE RELOAD EVALUATION 12 St. Lucie Unit I L-2014-125 Docket No. 50-335 Enclosure EC 277290, REVISION 0 ST. LUCIE UNIT I CYCLE 25 RELOAD
SUMMARY
The l0CFR50.59 Applicability/Screening performed for the St. Lucie Unit 1 Cycle 25 Core Reload concluded, based upon the changes provided in the reload EC [i.e., the new fuel batch Region GG (88 fresh assemblies), implementation of a Zircaloy-4 MONOBLOC comer guide tube design, reduction in AFW flow and increased AFW temperature in accident analyses, SIT line MOVs position to be less than 100% open (>90%), SG blowdown flow isolation in 30 minutes instead of 20 minutes, and tripping of all four RCPs instead of two at 30 minutes], some items could not screen out and had to be evaluated further in this document.
The items to be covered in this 10CFR50.59 evaluation are the analytical parameter value changes to the AFW pressure/flow conditions, increased AFW temperature, and SG blowdown isolation timing in the accident analyses, which constitute a change that adversely affects the UFSAR described functions of the systems relied upon to mitigate the consequences of the loss of feedwater (LOFW) and feedwater (FW) line break events.The actual AFW operating point has been observed to be at pressures up to 1030 psia, whereas the accident analyses had only considered conditions up to 1000 psia. Also, pump heat added to the fluid had not been considered in the analyses.
As a result, the input parameters for the AFW pumps was revised to add an operating point at 1030 psia, with a minimum degraded AFW flow rate of 276 gpm, and a maximum uncertainty adjusted AFW fluid temperature of ll 1.5°F (previous value being used was 104'F). At the same time, the analyses were run with the time assumed for the operator to isolate the SG blowdown flow to be 30 minutes instead of the previously assumed timing in the Chapter 10 LOFW analysis of 20 minutes. These items were considered adverse, as these changes could adversely affect the AFW system capability to remove heat from the primary system via the steam generators to cool down the RCS during post-accident conditions, and because of the delay in SG blowdown isolation which could affect SG inventory and challenge the heat removal to maintain RCS subcooling.
These parameters required the re-analyses to be performed for three UFSAR events: the loss of normal feedwater (UFSAR Section 15.2.8), the loss of normal feedwater concurrent with Auxiliary Feedwater pipe break (UFSAR Chapter 10), and the feedwater line break (UFSAR Section 10.5.3). The new results of these three analyses provided in Reference I and discussed in the reload EC show that the analyses meet the same acceptance criteria as before and the consequences of these analyses remain unchanged from those presented in the UFSAR. These analyses are being incorporated into the UFSAR with a change package provided in Attachment 2 of the EC.The discussions within this EC, along with the 10 CFR 50.59 Screening and Evaluation justify that the design and operation of the Cycle 25 reload core will meet the 10 CFR 50.59 (c)(2)criteria.
The core reload activities can be implemented with no changes to the St. Lucie Unit 1 Technical Specifications.
The safety analyses results are within the acceptance limits provided by the USNRC regulatory criteria and within the criteria provided by 10 CFR 50.59. Therefore, prior NRC approval is not required for implementation of this EC-DCP for operation in all Modes.13