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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:License-Application for Facility Operating License (Amend/Renewal) DKT 50
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 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-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-2023-086, 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. ML23095A0072023-04-0404 April 2023 License Amendment Request Revision 2 for the Technical Specifications Conversion to NUREG-1431 Revision 5 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-2023-004, Supplemental Information Regarding License Amendment Request 274, Reactor Protection System, Engineered Safety Features Actuation System, and Nuclear Instrumentation System Replacement Project - Submittal of Huma Factors Results2023-01-17017 January 2023 Supplemental Information Regarding License Amendment Request 274, Reactor Protection System, Engineered Safety Features Actuation System, and Nuclear Instrumentation System Replacement Project - Submittal of Huma Factors Results L-2022-160, Station,, Point Beach Units 1 and 2, License Amendment Request for Common Emergency Plan Consistent with NUREG-0654, Revision 22022-10-0404 October 2022 Station,, Point Beach Units 1 and 2, License Amendment Request for Common Emergency Plan Consistent with NUREG-0654, Revision 2 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) ML22243A1622022-08-26026 August 2022 Submittal of License Amendment Request 276, Revise Fire Protection Program in Support of Reactor Coolant Pump Seal Replacement Project ML21265A3722021-09-23023 September 2021 Improved Technical Specifications Conversion, Volume 1, Revision 0, Application of Selection Criteria to the Turkey Point Nuclear Generating Station Unit 3 and Unit 4 Technical Specifications L-2021-158, Contents of the Turkey Point Nuclear Plant, Units 3 and 4 Improved Technical Specifications (Its) Submittal2021-09-22022 September 2021 Contents of the Turkey Point Nuclear Plant, Units 3 and 4 Improved Technical Specifications (Its) Submittal ML21265A3822021-09-22022 September 2021 Improved Technical Specifications Conversion, Volume 11, Revision 0, Section 3.6, Containment Systems ML21265A3802021-09-22022 September 2021 Improved Technical Specifications Conversion, Volume 9, Revision 0, Section 3.4, Reactor Coolant Systems (RCS) ML21265A3772021-09-22022 September 2021 Improved Technical Specifications Conversion, Volume 6, Revision 0, Section 3.1, Reactivity Control Systems ML21265A3812021-09-22022 September 2021 Improved Technical Specifications Conversion, Volume 9, Revision 0, Section 3.5, Emergency Core Cooling Systems (ECCS) ML21265A3872021-09-22022 September 2021 Improved Technical Specifications Conversion, Volume 16, Revision 0, Section 5.0, Administrative Controls ML21265A3792021-09-22022 September 2021 Improved Technical Specifications Conversion, Volume 8, Revision 0, Section 3.3, Instrumentation ML21265A3852021-09-22022 September 2021 Improved Technical Specifications Conversion, Volume 14, Revision 0, Section 3.9, Refueling Operations ML21265A3842021-09-22022 September 2021 Improved Technical Specifications Conversion, Volume 13, Revision 0, Section 3.8, Electrical Power Systems ML21265A3832021-09-22022 September 2021 Improved Technical Specifications Conversion, Volume 12, Revision 0, Section 3.7, Plant Systems ML21265A3862021-09-22022 September 2021 Improved Technical Specifications Conversion, Volume 15, Revision 0, Section 4.0, Design Features ML21265A3782021-09-22022 September 2021 Improved Technical Specifications Conversion, Volume 7, Revision 0, Section 3.2, Power Distribution Limits ML21265A3752021-09-22022 September 2021 Improved Technical Specifications Conversion, Volume 4, Revision 0, Chapter 2.0, Safety Limits (Sls) ML21265A3762021-09-22022 September 2021 Improved Technical Specifications Conversion, Volume 5, Revision 0, Section 3.0, LCO and SR Applicability ML21265A3742021-09-22022 September 2021 Improved Technical Specifications Conversion, Volume 3, Revision 0, Chapter 1.0, Use and Application ML21265A3732021-09-22022 September 2021 Improved Technical Specifications Conversion, Volume 2, Revision 0, Determination of No Significant Hazards Considerations Generic Changes L-2021-071, License Amendment Request 273, Update Listing of Approved LOCA Methodologies to Adopt Full Spectrum Tm LOCA Methodology2021-04-15015 April 2021 License Amendment Request 273, Update Listing of Approved LOCA Methodologies to Adopt Full Spectrum Tm LOCA Methodology L-2020-157, Supplement to License Amendment Request 264, Adopt Emergency Action Level (EAL) Scheme Described in NEI 99-01, Revision 6, Development of Emergency Action Levels for Non-Passive Reactors2020-11-0505 November 2020 Supplement to License Amendment Request 264, Adopt Emergency Action Level (EAL) Scheme Described in NEI 99-01, Revision 6, Development of Emergency Action Levels for Non-Passive Reactors L-2020-053, Exigent License Amendment Request 272, One-Time Extension of TS 6.8.4 Steam Generator Inspection Program2020-04-0404 April 2020 Exigent License Amendment Request 272, One-Time Extension of TS 6.8.4 Steam Generator Inspection Program L-2020-003, License Amendment Request 268, Request to Extend Containment Leakage Rate Test Frequency2020-01-27027 January 2020 License Amendment Request 268, Request to Extend Containment Leakage Rate Test Frequency L-2019-203, License Amendment Request 264, Adopt Emergency Action Level (EAL) Scheme Described in NEI 99-01, Revision 6, Development of Emergency Action Levels for Non-Passive Reactors.2019-12-0606 December 2019 License Amendment Request 264, Adopt Emergency Action Level (EAL) Scheme Described in NEI 99-01, Revision 6, Development of Emergency Action Levels for Non-Passive Reactors. L-2019-192, License Amendment Request 270, Modify Containment Atmosphere Radioactivity Monitoring, Containment Ventilation Isolation and RCS Leakage Detection System Requirements2019-11-0404 November 2019 License Amendment Request 270, Modify Containment Atmosphere Radioactivity Monitoring, Containment Ventilation Isolation and RCS Leakage Detection System Requirements L-2019-005, License Amendment Request 266, Clarify Requirements When One Unit Is Outside the Applicability of Certain Technical Specifications2019-02-14014 February 2019 License Amendment Request 266, Clarify Requirements When One Unit Is Outside the Applicability of Certain Technical Specifications L-2018-219, Supplement to License Amendment Request 265, Revise NFPA 805 License Condition for Reactor Coolant Pump Seals2018-12-0303 December 2018 Supplement to License Amendment Request 265, Revise NFPA 805 License Condition for Reactor Coolant Pump Seals L-2018-203, Supplement to License Amendment Request 257, Modify Emergency Diesel Generator Partial-Load Rejection Surveillance Requirement2018-11-20020 November 2018 Supplement to License Amendment Request 257, Modify Emergency Diesel Generator Partial-Load Rejection Surveillance Requirement L-2018-170, License Amendment Request 265, Revise NFPA 805 License Condition for Reactor Coolant Pump Seals2018-10-17017 October 2018 License Amendment Request 265, Revise NFPA 805 License Condition for Reactor Coolant Pump Seals L-2018-176, Subsequent License Renewal Application, Responses to the August 2018 NRC On-Site Regulatory Audit Follow-Up Items2018-10-17017 October 2018 Subsequent License Renewal Application, Responses to the August 2018 NRC On-Site Regulatory Audit Follow-Up Items L-2018-175, Subsequent License Renewal Application, Safety Review Requests for Additional Information (RAI) Set 5 Responses2018-10-17017 October 2018 Subsequent License Renewal Application, Safety Review Requests for Additional Information (RAI) Set 5 Responses L-2018-177, Subsequent License Renewal Application Safety Review Requests for Confirmation of Information (Rci) Responses2018-10-0909 October 2018 Subsequent License Renewal Application Safety Review Requests for Confirmation of Information (Rci) Responses 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 ML18130A4662018-06-12012 June 2018 Amendments Regarding Technical Specifications Pertaining to IST Program and ISI Program Requirements and Surveillance Frequency Control Program Applicability L-2018-049, Application to Add Limiting Condition for Operation (LCO) 3.0.6 to the Technical Specifications2018-05-29029 May 2018 Application to Add Limiting Condition for Operation (LCO) 3.0.6 to the Technical Specifications L-2018-044, License Amendment Request 257, Modify Emergency Diesel Generator Partial-Load Rejection Surveillance Requirement2018-05-14014 May 2018 License Amendment Request 257, Modify Emergency Diesel Generator Partial-Load Rejection Surveillance Requirement L-2018-065, License Amendment Request 258, Revise Reactor Core Safety Limit to Reflect WCAP-17642-PA, Revision 12018-05-0303 May 2018 License Amendment Request 258, Revise Reactor Core Safety Limit to Reflect WCAP-17642-PA, Revision 1 L-2018-082, Turkey Point, Units 3 and 4 - Enclosure 2 to L-2018-082, Applications for Withholding Proprietary Information from Public Disclosure Per 10 CFR 2.390 (Public Version)2018-04-10010 April 2018 Turkey Point, Units 3 and 4 - Enclosure 2 to L-2018-082, Applications for Withholding Proprietary Information from Public Disclosure Per 10 CFR 2.390 (Public Version) L-2018-082, Enclosure 4 to L-2018-082 - Non-Proprietary Reference Documents and Redacted Versions of Proprietary Documents (Public Version)2018-04-10010 April 2018 Enclosure 4 to L-2018-082 - Non-Proprietary Reference Documents and Redacted Versions of Proprietary Documents (Public Version) ML18113A1422018-04-10010 April 2018 Enclosure 2 to L-2018-082, Applications for Withholding Proprietary Information from Public Disclosure Per 10 CFR 2.390 (Public Version) ML18113A1412018-04-10010 April 2018 Enclosure 1 to L-2018-082, Subsequent License Renewal Application, Enclosure Summary 2024-04-30
[Table view] |
Text
NRC FORM 195 U.S. NUCLCAR REGULATORY COMh<ISSION DOCKET NUMBER V
{2.76) I 50-251 FILE NUMBER NRC DISTRIBUTION FOR PART 60 DOCI<ET MATERIAL TO: FROM: OATE OF OOCUMENT Mr. Victor Stello Florida Power & Light Company 12/9/76 Miami, Florida OATE RECEIVED Mr. Robert E. Uhrig 12/9/76 RLETTE R RNOTOR IZE 0 PROP INPUT FORM NUMBER OF COPIES RECEIVED
/@ORIGINAL QUN C LASS I F I E O three signed DCOP Y 40 copies encl recvd.
OESCQIPTION ~ ENCLOSU RE Ltr. notorized 12/9/76.... trans the following: Amdt. to ol/change to Appendix A tech specs
....submitted as a result of a re-evaluation of ECCS cooling performance.
(S-P) 'ACKNOWLEDGED PLANT NAME: DP NOT REMOVE .
Turkey Point Unit 04 SAFETY FOR ACTION/INFORMATION 12 10 76 ASSXGNED PD:
N Lear PR ECT MAN E Elliott PROJECT lfANAGER'IC L C ASST Parrish ASST lfcGough INTERNAL D I ST R I BUT ION REG F SYSTEMS SAFETY PLANT SYSTEMS S TE SAFE NRC PDR HEXNEMAN TEDESCO I&E SCHROEDER OELD GOSSXCK & STAFF ENGINEERXNG XPPOLXTO MXPC ERNST CASE KNIGHT HANAUER SIHMEIL OPERATING REACTORS SPANGLER HARLESS PAWL CK STELLO PROJECT MANAGEMENT REACTOR SAFE OPERATING TECH GAlfMILL BOYD ROSS EISENHUT STEPP PE COLLINS NOVAK HOUSTON ROSZTOCZY PETERSON CHECK BUTLE SITE ANALYSIS.
MELTZ VOLLMER HELTEMES AT&I BUNCH SKOVHOLT SALTZlfAN J COLLINS RUTBERG KREGER EXTERNAL DISTRIBUTION CONTROL NUiiilBE R LPDR ~ Miami Fla. NAT LAB B 00 TIC: REG V,XE ULRXKSON OR NSIC: LA PDR ASLB: CONSULTANTS:
ACRS CYS S&bBKH&/ E T ~
NRC FORM 196 (2 76)
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4< P. O. BOX 013100, MIAMI, FL 33101
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~+ gc g FLORIDA POWER & LIGHT COMPANY December 9, 1976 L-76-418 Office of Nuclear Reactor Regulation Attention: Mr. Victor Stello, Director Division of Operating Reactors U. S. Nuclear Regulatory Commission Washington, D. C. 20555
Dear Mr. Stello:
Re: Turkey Point Unit 4 Docket No. 50-251 Proposed Amendment to Facilit 0 crating License DPR-41 i I In accordance with 10 CFR,50.30, Florida Power 6 Light Company submits herewith three >>(3) signed orig'inals and forty (40).
copies of a request to amend Appendix A of Facility, Operating License DPR-41.
i, This proposal is being submitted as a result of a re-evaluation of HCCS coolingperformance calculated in accordance wi'th an approved Westinghouse Evaluation Model. The proposed change described below and shown on the accompanying Technical
's Specification pages bearing the date of this letter in the lower right hand corner.
Page 3.2-3 is designated applicable to Unit 3 only.
New page 3.2-3a is designated applicable to Unit 4 only.
The new page contains a revision to Specification 3.2.6.a such that the limit on the Heat. Flux Hot Channel Factor (F<) for Unit 4 is reduced from 2.32 to 2.24.
Pa es B3.2-4 and 83.2-6 Pages B3.2-4 and B3.2-6 are designated applicable to Unit 3 only.
~Pa es B3.2-4a and B3.2-6a New pages B3.2-4a and B3.2-6a are designated applicable to Unit 4 only. The new 'pages present the basis for the revised Unit 4 limit on F<.
PEOPLE.;. SERVING PEOPLE
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Office of Nuclear Reactor Regulation Attention: Mr. Victor Stello, Director Page Two Page 3.4-1 is designated applicable to Unit 3'nly.
New page 3.4-1a is designated applicable to Unit 4 only. The accumulator water volume in Specification 3.4.1.3 is revised from 825-841 ft to 875-891 ft Although the above amendments are being proposed now, they will not be applicable until after the Spring 1977 Unit 4 refueling outage. NRC approval of these proposed amendments is requested during the spring refueling outage, however, the amendments are not necessary for the conduct of the refueling, or return to op-eration following the refueling. For the remainder of the current core Cycle 3, the accumulator minimum. water volume will remain:at 825 cubic feet. During the refueling outage, the accumulators will be modified to increase the minimium water volume to 875 cubic feet in conformance with the ECCS re-evaluation. Therefore, for the re-mainder of Cycle 3, the maximum allowable nuclear peaking factor(Fg) will be limited to 2.08 December in'accordance with the Order for Modifi-cation of License dated 3, 1976.,issued by the Commission for Turkey Point Unit 4.
The proposed amendment has been reviewed by the Turkey Point Plant Nuclear Safety Committee and the Florida Power & Light Company'Nu-clear Review Board. They have concluded that safety question. written it safety does not. involve evaluation is at-an unreviewed A tached.
Very tr y ours,'
~
Robert E. Uhrig ~
Vice President REU/MAS/cpc/ls Attachments cc: Mr. Norman C. Moseley Robert Lowenstein, Esquire
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STATE OF FLORIDA )'"
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COUNTY OF DADE )
being first duly sworn, deposes and says:
'E Executive Vice President ~
of Florida Power &
Light Company, the ><<nsee herein; That he has executed the foregoing document; that the state-ments made in this said document are true and correct to the best of his knowledge, information, and belief, and that, he is authorized to execute the document, on behalf of said >>ce>>ee-E. A. Adomat Subscribed and sworn to before me this ctay of , F9~7 L
NOTAR PUBLIC, in and for the County of Dade, State of Florida QQ74Ry pI IoI Ir <47E OF FLORIDA 4> i >>rF, 1979 c:L MY COMMISSION EXP IRKS!4N. 26, UNDERWRILEi5 BONDED TIIRl! GENERAL INSURANCE r,"~'Q 7a, '
My commission expires:
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Unit 3 reactivity insertion upon ejection greater than 0.3% A k/k at rated power. Inoperable rod worth shall be determined within 4 weeks.
- b. A control rod shall'be considered inoperable if (a) the rod cannot be moved by the CRDM, or (b) the rod is misaligned 'from its bank by more than 15 inches, or (c) the rod drop time is not met.
- c. If a control rod cannot be moved by the drive mechanism, shutdown margin shall be increased by boron addition to compensate for the withdrawn worth of the inoperable rod.
- 5. CONTROL ROD POSITION INDICATION If either the power range channel deviation alarm or the rod deviation monitor alarm are not operable rod positions shall be logged once per shift and after a load change greater than 10% of rated power. If both alarms are inoperable for two hours or more, the nuclear overpower tri'p shall be reset to 93% of rated power.
- 6. POSER DISTRIBUTION LIMITS
- a. At all times except during low power physics tests, the hot channel factors defined in the basis must meet the following limits:
Fq(Z) < (2 '2/P) x K(Z) for P ~ ~ 5 Fq(Z) < (4 ~ 64) x K(Z) for P < ~ 5 FNA < 1 55 [1 + 0.2 (1-P) ]
where P is the fraction of design power at which the core is operating. K(Z) is the function given in Figure 3.2-3 and Z is the core height location of Fq.
Following. initial loading before the reactor is
~
~
b.
operated above 75% of rated power and at regular effective full rated power monthly intervals thereafter, power distribution maps, using the movable detector system shall be made, to conform that the hot channel factor. limits of the specifica-tion are satisfied. For the purpose of this comparison, Unit 3 3~ 2 3
Si Unit 4 reactivity insertion upon ejection greater than 0.3% A k/k at rated power. Inoperable rod worth shall be determined within 4 weeks.
- b. A control rod shall be considered inoperable if (a) the rod cannot be moved by the CRDM, or (b) the rod's misaligned from its bank by more than 15 inches, or (c) the rod drop time is not met.
- c. If a control rod cannot be moved by the drive mechanism, shutdown margin shall be increased by boron addition to compensate for the withdrawn worth of the inoperable rod.
- 5. CONTROL ROD POSITION INDICATION If either the power range channel deviation alarm or the rod deviation monitor alarm are not operable rod positions shall be logged once per shift and after a load change greater than 10% of rated power. If both alarms are inoperable for two hours or more, the nuclear overpower trip shall be reset to 93% of rated power-
- 6. POWER DISTRIBUTION LIMITS
- a. At all times except during low power physics tests, the hot channel factors defined in the'asis must meet the following limits:
Fq(Z) < (2.24/P) x K(Z) for P > 5 ~
Fq(Z) < (4.64) x K(Z) for P < .5
< 1'55 [1 + 0.2 (1-P) ]
FAH where P is the fraction of design power at which the core is operating. K(Z) is the function given in Figure 3.2-3 and Z is the core height location of l
Fq.
- b. Following, initial loading before the reactor is operated above 75% of rated power and at regular effective full rated power monthly intervals thereafter, power distribution maps, using the movable detector system shall be made, to conform that the hot channel factor limits of the specifica<<
tion are satisfied. For the purpose of this comparison, 30 2 3 a 12/9/76 i$
Unit 4
I
~ ~ ~
Unit 3 An upper bound envelope of, 2.32 times the normalized peaking factor axial dependence of Figure 3.2>>3 has been determined (from extensive analyses at.
design power considering all operating maneuvers) to be consistent with the technical specifications on power distribution control as given in Section 3.2. The, results of the loss of coolant accident analyses based on this upper bound envelope indicate a peak clad temperature of 2150'F at design power; corresponding to. a 50'F margin to the 2200 F FAG limit.
When an F measurement is taken, both experimental error and manufacturing tolerance must be allowed for. Five percent is the appropriate experimen'tal uncertainty allowance for a full core map taken with the movable incore detector flux mapping system and three percent is the appropriate allowance for manufacturing tolerance.
N In the, specified limit of F , there is an 8 percent allowance for uncertain-ties which means that normal operation of the core is expected to result. in F (1.55/1.08. The logic behind the 1'arger uncertainty in this case is that AH-(a) normal perturbations in the radial power shape (e.g., rod misalign-N ment) affect P H, in most cases without necessarily affecting F ,(b) the has a direct influence on F through moveme'nt of rods, and can limit AH'perator it to the desired value, he has no direct control over F<H N and (c) an error in the predictions for radial power shape, which may be detected during startup physics tests can be compensated for in P by tighter axial control, but compensation for F is less readily available. When a measurement of F
AH is taken, experimental error must be allowed for and 4% is the appro-priate allowance for a full core map taken with the movable incore detector flux mapping system.
Measurements of the hot channel factors are required as part of start-up physics tests, at least once each full rated power month or operation~ and whenever abnormal power distribution conditions require a reduction of core power to a level based on measured hot channel factors. The incore map taken following initial goading provides confirmation of the basic nuclear Unit 3 a3.2-4 12/9/76
Unit 4 An upper bound envelope of 2. 24 times the normalized peaking factor axial dependence of Figure 3.2-3 has been determined to be consistent with the technical specifications on power distrib'ution control as given in Section 3.2.
When an F measurement is taken, both experimental error and manufacturing tolerance must be allowed for. Five percent is the appropriate experimental uncertainty allowance for a full core map taken with the movable incore detector flux mapping syst'm and three percent is the appropriate allowance for manufacturing tolerance.
In thh specified limit of N F>H, there is an 8 percent allowance for uncertain-ties which means that normal operation of the core is expected to result in F <1.55/1.08. The logic behind the larger uncertainty in this case is that (a) normal perturbations in the radial power shape (e.g., rod misalign-ment) affect F , in most cases without necessarily affecting F ,(b) the has a direct influence on F through movement of rods, and can limit q'perator it to the desired value, he has no direct control over F>H N and (c) an error in the predictions for radial power shape, which may be detected during startup physics tests can be compensated for in F by tighter axial control, but compensation for F is less readily available. When a measurement of F
AH is taken, experimental error must be allowed for and 4% is the appro-priate allowance for a full core map taken with the movable incore detector flux mapping system.
1feasurements of the hot channel factors are required as part of start-up physics tests, at least once each full rated power month of operation, and whenever abnormal power distribution conditions require a reduction of core power to a level based on measured hot channel factors. The incore map taken following initial Roading provides confirmation of the basic nuclear Unit 4 B3.2-4 a
~ ~
Unit 3 Flux Difference (4$ ) and a reference value which corresponds to the full design power equilibrium value of Axial Offset (Axial Offset 44/fractional power) . The ref erence value of flux difference varies with power leve1 and burnup but expressed as axial offset it varies only with burnup.
The technical specifications on power distribution- contro1 assure that th' F upper bound envelope of 2.32 times Figure 3.2-3 is not exceeded and xenon q
distributions are not developed which at. a later time, would cause greater local power peaking even though the flux difference is then within the 1imits specified by the procedure.
The target (or reference) value of flux difference is determined as follows.
At any time that equilibrium xenon conditions have been established, the in-dicated flux difference is noted with part length rods withdrawn from the core and with the full length rod control rod bank more than 190 steps withdrawn (i.e., normal rated power operating position appropriate for the time in life.
Control rods are usually withdrawn farther as burnup proceeds).- This value, divided by the fraction of design power at which the core was operating is the design power value of the target flux difference. Values f'r all other core power levels are obtained by multiplying tne design power value by the fractional .power. Since the indicated equilibrium value was noted, no allowances for excore detector error are necessary and indicated deviation of
+5% 4I are permitted from the indicated reference value. During periods where extensive load following is required, it may be impractical to establish the required core conditions for measuring the target flux difference every rated power month. For this reason, methods are permitted by Item 6c of Section 3,2 fo'r updating the target flux differences. Figure B3.2>>1 shows a typical construction of the target flux difference band.. at BOL and Figure B3.2-2 shows the typical variation of the full power value with burnup.
Strict:control of the flux difference (and rod*position) ih not as necessary during part power operation. This is because xenon distribution contro1 at part power is not as significant as the control at full power and allowance t
has been made in predicting the heat flux peaking factors for less strict co"..
trol at part power. Strict control of the flux difference i.s not possible during certain physics tests or during the required, periodic excore calibra-3 Unit 3 B3.2-6
,,12/9/76
~ ~
Unit 4, Flux Difference (hP) and a reference value which corresponds to the full design power equilibrium value of Axial Offset (Axial Offset ~ ~4/fractional power). The reference value of flux difference varies with power level and burnup but expressed as axial offset it varies only with burnup.
The technical specifications on power distribution contro1 assure 'that the F upper bound env'elope of 2. 24 times Figure 3.2-3 is not exceeded and xenon distributions are not developed which at a later time, would cause greater local power peaking even though the flux difference is then within the limits specified by the procedure.
The target (or reference) value of flux difference is determined as follows.
At any time that equilibrium xenon conditions have been established, the in-V dicated flux difference is noted with part length rods witndrawn from the core and with the full length rod control rod bank more than 190 steps withdrawn (i.e., normal rated power operating position appropriate for the time in life.
Control rods are usually withdrawn farth r as burnup proceeds).'his value, divided by the fraction of design power at wnich the core was operating is the design power value of the target flux difference. Values for all other core power level@ are obtained by multiplying tne design power value by the fractional power. Since the indicated equilibrium valu was noted, no allowances for excore detector error are necessary and indicated deviation of
+5% GI are permitted from the indicated reference value. During periods where extensive load following is required, it may be impractical to establish the required core conditions for measuring the target flux difference every rated power month. For this reason, methods are permitted by Item 6c of Section 3.2 for updating the target flux differences. Figure B3.2-1 shows a typical construction of the target flux difference band at BOL and Figure B3.2-2 shows the typical variation of the full power value with burnup.
Stiict control of the flux difference (and rod position) is not as necessary II during part power operation. This is because xenon distribution control at part power is not as significant as the control at full power and allowance has been made in predicting the heat flux peaking factors for less strict co..
trol at part power. Strict control of the flux difference is not possible during certain physics'tests or during the required, periodic excore calibra-Unit 4 B3.2-6 a 12/9/76
UNIT 3 3.4 ENGINEERED SAFETY FEATURES Applies to the operating status of the Engineered Safety Features.
~Ob ective: To define those limiting conditions for operation that are necessary: (1) to remove decay heat from the core in emergency or normal shutdown situations, (2) to re-move heat from containment in normal operating and emergency situations, and (3) to remove airborne iodine from the containment atmosphere in the event of a Maximun Hypothetical Accident.
- a. The reactor shall not be 1
made critical, except for low power physics tests, unless the following conditions are met:
- 1. The refueling water tank shall contain not less than 320,000 gal. of water with a boron con-II centratlon of at least. 1950 ppm
- 2. The boron injection tank shall contain not less than 900 gal. of a 20,000 to 22,500 ppm boron solution. The solution in the tank, and in isolated portions. of the inlet and outlet piping, shall be maintained at a temperature of at least 145F. TMO channels of heat tracing shall be operable f'r the flow path.
- 3. Each accumulator* shall be pressurized to at least 600 psig and contain 825-841 ft3 of water with a boron concentration of at least 1950 ppm, and shall not be isolated.
- 4. FOUR safety injection pumps shall be operable.
3.4-1 UNIT 3
'i2/9/76
I~
UNIT 4 3.4 ENGINEERED SAFETY FEATURES Features.
~Ob ective: To define those limiting conditions for operation that are necessary: (1) to remove decay heat from the core in emergency or normal shutdown situations, (2) to re-move heat from containment in normal operating and emergency situations, and (3) to remove airborne iodine from the containment atmosphere in the event. of a Maximum Hypothetical Accident.
- a. The reactor shall not be made critical, except for low power physics tests, unless the following conditions are met:
- 1. The refueling water tank shall contain not less than 320,000 gal. of water with a boron con-centration of at least 1950 ppm.
- 2. The boron injection tank shall contain not less than 900 gal. of a 20,000 to 22,500 ppm boron, solution. The solution in the tank, and in isolated portions of the inlet and outlet piping, shall be maintained at a temperature of at least 145F. TWO channels of heat tracing shall be operable for the flow path.
- 3. Each accumulator shall be pressurized to at least 600 psig and contain 875-891 ft of water with a boron concentration of at least 1950 ppm, and shall not be isolated.
FOUR safety injection pumps shal3. be operable.
UNIT 4 3e4-1 a
. 12/9/76
SAFETY EVALUATION I. Introduction This safety evaluation supports the following proposed changes to the Unit 4 Technical Specifications:
(1) The maximum allowable nuclear peaking factor (F )
is decreased from 2.32 to 2.24.
(2) The limits on Safety Injection accumulator water volume are increased from 825-841 ft to 875-891 ft II. Discussion A. Core C cle 4 A re-evaluation of ECCS cooling performance calculated in accordance with an approved Westinghouse Evaluation Model has been performed. The re-evaluation shows that for breaks up to and including the double ended severence of a reactor coolant pipe, the ECCS will meet the Acceptance Criteria presented in 10 CFR 50.46.
The detailed re-evaluation is contained in FPL letter L-76-419 of December 9 , 1976, and shows that, at a core power level of 102% of 2192 Mwt and a minimum accumulator water volume of 875 ft, per accumulator, the maximum allowable nuclear peaking factor is 2.25.
However, since the Technical Specifications allow a maximum core power level of 2200 Mwt, the re-evaluation is being revised using the higher power level. The revised calculation is expected to yield a maximum Fq of 2.24.
B.
The following analysis shows that, by operating with F<
<2.08 for the remainder of Cycle 3, we can provide conservative assurance of safe operation.
The Westinghouse ECCS re-evaluation assumed:
- 1. 10% steam generator tube plugging ft2.25
- 2. F =
- 3. 895 accumulator minimum water volume
- 4. 2192 Mwt core power level However, for the remainder of Cycle 3, the actual parameters will be:
- l. 7% steam generator tube plugging
=
- 2. F ft2.08
- 3. 895 accumulator minimum water volume
- 4. 2200 Mwt core power level
SAFETY EVALUATION (Continued)
Therefore, the following adjustments are needed to adapt the ECCS re-evaluation to the remainder of Cycle 3:
(1) Tube Plu ing Ad'ustment From the Westinghouse ECCS re-evaluation, Peak Clad Temperature (PCT) is:
(a) PCT =
2192 Mwt 2198'F for 10% plugging, 875 ft (b) PCT =
Mwt 2162'F for 5% plugging, 875 ft , 2192 As shown, the PCT varies by 36'F as tube plugging increases from 5% to 10%. Therefore, for 7% tube pluggingF PCT = 2198 F [ (. 60) (36'F) ]
2198DF 21.6DF 2176.4DF (2) F Adjustment PCT is further reduced by lowering F from 2.25 to 2e08 (BFq = e17) ~
using the relationship 10 F By it can be shown that . 17F q 01F
(
10 F
.01F ) = 170'F, i.e., lowering F f=om 2.25 to 2.08 results in a reduction of 190'F in PCT. Therefore, for Fq=2.08, PCT = 2176.4'F 170'F 2006.4DF
SAFETY EVALUATION (Continued)
(3) Accumulator Wa ter Volume Ad'ustment Westinghouse ECCS analyses show that a decrease in accumulator water volume from 875 ft to 825 ft Therefore, corresponds to a for 825 ft3, 36'F increase in PCT.
PCT = 2006.4 F + 36 F 2042.4oF (4) Core Thermal Power Adjustment Increasing the core power from 2192 Mwt to 2200 Mwt will increase the PCT as follows:
(a) hFq = ) (2.25) = .008 2192 (b) (. 008Pq) . 01F 01Fq 8 P Therefore, for 2200 Mwt, PCT = 2042.4 F + 8 F 2050.4 F Thus, operation with F <2.08 for the remainder of Cycle 3 will result in a PCT of 2050.4'F, which is well below the ECCS acceptance criteria of 2200'F.
III.Con'clusions s
Based on these considerations, (1) the proposed change does not increase the probability or consequences of accidents or malfunctions of equipment important to safety and does not reduce the margin of safety as defined in the basis for any technical specification, therefore, the change does not involve a significant hazards consideration, (2) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, and (3) such activities will be conducted in compliance with the Commission's regulations and the issuance of this amendment will not be inimical to the common defense and
'security or to the health and safety of the public.
'