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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-2024-071, Cycle 27 Core Operating Limits Report2024-04-29029 April 2024 Cycle 27 Core Operating Limits Report L-2024-070, Cycle 32 Core Operating Limits Report2024-04-29029 April 2024 Cycle 32 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-2024-054, 2023 Annual Environmental Operating Report2024-04-0909 April 2024 2023 Annual Environmental Operating Report 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-2024-045, Report of 10 CFR 72.48 Plant Changes2024-03-27027 March 2024 Report of 10 CFR 72.48 Plant Changes 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-2024-023, Unusual or Important Environmental Event - Turtle Mortality2024-03-0606 March 2024 Unusual or Important Environmental Event - Turtle Mortality L-2024-015, 2023 Annual Radioactive Effluent Release Report2024-02-29029 February 2024 2023 Annual Radioactive Effluent Release Report L-2024-029, 2023 Annual Operating Report2024-02-28028 February 2024 2023 Annual Operating Report L-2024-026, Revised Reactor Vessel Materials Surveillance Capsule Withdrawal Schedules2024-02-27027 February 2024 Revised Reactor Vessel Materials Surveillance Capsule Withdrawal Schedules 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-2023-102, 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-2023-099, 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-2023-097, 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) 2024-08-07
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Text
NRC.roRM 195 (2 78)
U.S. NUCLEAR REGVLATORV COMMISSION d-3 DOCKET N MUSH NRC DISTRIBuTION roR PAn i'0 DOCKET MATERIAL FILE NUMBER FROM: DATE OF DOCUMENT TO Mr. Victor Stello FPL 06-30-77 Miami, Florida 33101 DATE RECEIVED Robert E. Uhrig 07-05-77 QNOTORIZED PROP INPUT FORM NUMBER OF COPIES RECEIVED E ER 408 C LASS RIG IN Elcopv AL IFIE D S(4 ~ED ENCLOSURE DESCRIPTION Notorized,06-30-77...Amd'lt to Appendi A, Operatin'g Lidense,PPR-67/Change to Tech Specs..
consisting of Page 3/0 1-26 and Page B 3/4 1-4 concerning The maxim(@ alloqable CEA drpp time in specification 3.1.3.$ is changed from 3.3 seconds to 3.0 seconds to be','onsistent with the previous 2 pages. performed. accident ahalysis.....-
4 pages PLANT NAME: ST LUCIE jcm 07-06-77 UNIT,g 1 o
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QQ NOT REMOXB KNOWLEDGZD FOR ACTION/INFORMATION ENVIRHNMENTAL ASSIGNED AD: V% MOORE LTR BRANCH CHIEF:
ECT K%AGER: PROJECT MANAGER:
CENSING ASSISTANT. LICENSING ASSISTANT:
Bo HARLESS INTERNAL DISTR I BUTION STFMS SAFETY PLANT SYSTEMS SITE SAFETY &
HEINEMAN TEDESCO ENVIRON ANALYSIS ROEDER BENAROYA DENTON '& MULLER ENGINEERING IPPOLITO ENVIRO TECH ERNST OPERATING REACTORS BALLARD YOUNG BLOOD BAER VA BUTLER GAMMILL 2 GRIME CHECK SI'7E ANALYSIS VOLLMER AT I BUNCH ALTZMAN J ~ COLLINS R TBERG KREGER EXTERNAL DISTRIBUTION CONTROL NUMBER TIC NSIC REG IV J HMCHETT 16 CYS ACRS SENT CAT GO Y 77187O284
P. O. BOX 013100, MIAMI, FL 33101 FLORIDA POWER 8 LIGHT COMPANY June 30, 1977
~Regulatory Docket Fjje L-77-197 Director 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: St. Lucie Unit 1 Docket No. 50-335 I Proposed Amendment to Facilit 0 eratin License DPR-67 During a recent NSSS vendor review of the St. Lucie Unit 1 Technical Specifications, it was noted that the CEA drop time specification is nonconservative with respect to the accident analysis. The specified drop time is <3.3 seconds, whereas the accident analysis assumes a drop time of 3.0 seconds. The discrepancy has had no safety significance because the measured drop times at St. Lucie Unit 1 have all been well below 3.0 seconds (about 2.4 seconds maximum). However, Technical Specification 3.1.3.4 should be amended to be consistent with the accident analysis.
In accordance with 10 CFR 50.30, Florida Power 6 Light Company submits herewith three (3) signed originals and forty (40) conformed copies of a request to amend Appendix A of Facility Operating License DPR-67. The proposed change is described below and shown on the accompanying Technical Specification pages bearing the date of this letter in the lower right hand corner.
Page 3/4 1-26 and Pa e B 3/4 1-4 The maximum allowable CEA drop time in Specification 3.1.3.4 is changed from 3.3 seconds to 3.0 seconds to be consistent with the previously performed accident analysis.
The proposed amendment has been reviewed by the St. Lucie Facility Review Group (FRG) and the Florida Power 6 Light Nuclear Review Board (CNRB). They have concluded that it Company is an I
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Director of Nuclear Reactor Regulation Page Two administrative change to bring the Technical Specifications into conformance with the accident analysis and does not involve an unreviewed safety question.
Very truly yours, Robert E. Uhrig Vice President REU/MAS/cpc Attachment cc: Mr. Norman C. Moseley, Region XX Robert Lowenstein, Esquire
REACTIVITY CONTROL SYSTEMS CEA DROP TIME LIMITING CONDITION FOR OPERATION 3.1.3.4 The individual full length (shutdown and control) CEA drop time, from a fully withdrawn position, shall be < '3,0 seconds .rom when electrical power is interrupted to the CEA drive mechanism until the CEA reaches its 90 percent insertion position with:
- a. T avg-) 515 F, and
- b. All reactor coolant pumps operating.
APPLICABILITY: HOOE 3.
ACTION:
- a. With the drop time of any full length CEA determined to exceed the above limit, restore the CEA drop time to within the above limit prior to proceeding to MODE 1 or 2.
- b. . With the CEA drop times within limits but determined at less than full reactor coolant flow, operation may proceed provided THERMAL POWER is restricted to less than or equal to the maximum THERMAL POWER level allowable for the reactor coolant pump combination operating at the time of CEA drop time determination.
SURVEILLANCE RE UIREMENTS 4.1.3.4 The CEA drop time of full length CEAs shall be demonstrated through measurement prior to reactor criticality:
a ~ For all CEAs following each removal of the reactor vessel head,
- b. For specifically affected individual CEAs following any main-tenance on or modification to the CEA drive system which could affect the drop time of those specific CEAs, and C. At least once per 18 months.
ST. LUCIE - UNIT 1 3/4 1-26
~ 'r REACTIVITIY CONTROL SYSTEMS BASES 3/4.1.3 MOVABLE CONTROL ASSEMBLIES Continued protective system would not detect the degradation in radial peaking factors and since variations in other system parameters (e.g., pressure and coolant temperature) may not be sufficient to cause trips, it is possible that the reactor could be operating with process vari'ables less conservative than those assumed in generating LCO and LSSS setpoints.
Therefore, the ACTION statement associated with the large misalignment of a CEA requires a prompt and significant reduction in THERMAL POllER prior to attempting realignment of the misaligned CEA.
The ACTION statements applicable to misaligned or inoperable CEAs include requirements to align the OPERABLE CEAs in a given group with the inoperable CEA. Conformance with these alignment requirements bring the core, within a short period of time, to a configuration consistent with that assumed in generating LCO and LSSS setpoints. However, extended operation with CEAs significantly inserted in the core may lead to perturbations in 1) local burnup, 2) peaking factors and 3) available '-
shutdown margin which are more adverse than the conditions assumed to exist in the safety analyses and LCO and LSSS setpoints determination.
Therefore, time limits have been imposed on operation with inoperable CEAs to preclude such adverse conditions from developing.
Operability of the CEA position indicators (Specification 3.1.3.3~
is required to determine CEA positions and thereby ensure compliance with the CEA alignment and insertion limits and ensures proper operation of the rod block circuit. The CEA "Full In" and "Full Out" limits provide an additional independent means for determining the CEA positions when the CEAs are at either their fully inserted or fully withdrawn positions.
Therefore, the ACTION statements applicable to inoperable CEA position indicators permit continued operations when the positions of CEAs with inoperable position indicators can be verified by the "Full In" or "Full Out" limits.
CEA positions and OPERABILITY of the CEA position indicators are required to be verified on a nominal basis of once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> with more frequent verifications required if an automatic monitoring channel is inoperable. These verification frequencies are adequate for assuring that the applicable LCO's are satisfied.
The maximum CEA drop time permitted by Specification 3.1.3.4 is-the assumed CEA drop time of 3.0 seconds used in the accident analyses.
Measurement with Tav ~ 515'F and with all reactor coolant pumps operating ensures that the measured drop times will be representative of insertion times experienced during a reactor trip at operating conditions.
B 3/41-4
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SAFETY EVALUATION Introduction This evaluation supports a proposed change to the maximum allowable CEA drop time specified in Technical Specification 3.1.3.4.
Discussion As discussed in the cover letter, Specification 3.1.3.4 should be changed-to be consistent with the accident analysis. No equip-ment related changes are necessary, so the amendment. will not affect the probability or consequences of equipment malfunctions.
Neither will it affect the probability or consequences of hypo-thetical accidents because we are using the existing accident analysis as the basis for administrative correction of a Technical Specification. Also, there is no decrease in safety margin because the proposed change is in the conservative direction.
Conclusion Based on these considerations, (1) the proposed change does not increase the probability or consequences of accidents or mal-functions of equipment important to safety and does not reduce the margin of safety, therefore, the change does not involve an unreviewed safety question, (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.
STATE OF FLORIDA )
) SS ~
COUNTY OF DADE )
Robert E. Uhrig, being first duly sworn, deposes and says:
That he is a Vice President of Florida Power 6 Light Company, the Licensee 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 Licensee.
Robert E. Uhrig Subscribed and sworn to before me this ctay oa 19~7 NOTARY P BLIC, in and for t e County of Dade, State of Florida IOTARY PUBLIC STATE OP FLORIDA aI LARGE hIY CON~!SSIOII EXPIRES hIAY BQIIDED QIRII BIAYIIARQ BOIIOuYG My .commission expires: AGENCY
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