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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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Florida Power & Light Company, 6501 S. Ocean Drive, Jensen Beach, FL 34957 November 30, 2009 F=PL L-2009-277 10 CFR 50.46 U. S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555 Re: St. Lucie Unit 1 Docket No. 50-335 Acceptance Criteria for Emergency Core Cooling Systems for Light Water Nuclear Power Reactors 10 CFR 50.46 Change Report Florida Power and Light (FPL) was informed of inaccuracies in the large and small break loss of coolant accident (LOCA) analyses. The subsequent calculation that assessed the effects of these issues, along with other previous changes and errors in the accident analysis, resulted in a new peak cladding temperature (PCT) value within the regulatory limit of 2200 'F provided in 10 CFR 50.46. However, the accumulated effect on PCT was greater than 50 'F.
According to 10 CFR 50.46, when a cumulative PCT change in errors or changes to the analysis of 50 'F occurs, a 30 day report shall be sent to the NRC with a proposed schedule for providing reanalysis or taking action as may be needed to show compliance with 50.46 requirements. This information is provided in the attachment to this letter.
Please contact Ken Frehafer at (772) 467-7748 should you have any questions regarding this submittal.
Sincerely, Eric S. Katzman Licensing Manager St. Lucie Plant ESK/KWF Attachment Aoc an FPL Group company
St. Lucie Unit I L-2009-277 Docket No. 50-335 Attachment 10 CFR 50.46 Change Report Page 1 of 5 St. Lucie Unit 1 10 CFR 50.46 30-Day Change Report Emergency core cooling system (ECCS) analyses for St. Lucie Unit 1 is performed by AREVA NP Inc. The following 30-Day report pertaining to the application of the AREVA NP Inc. large break loss of coolant accident (LBLOCA) evaluation model (SEM/PWR-98) and small break loss of coolant accident (SBLOCA) evaluation model (ANF-RELAP) to St. Lucie Unit 1, is provided pursuant to 10 CFR 50.46(a)(3)(ii). A summary of the calculated peak cladding temperature (PCT) changes for LBLOCA and SBLOCA is provided in Tables I and 2, respectively.
1.0 Changes to LBLOCA 1.1 Pellet thermal conductivity degradation with burnup effect on RODEX2 An issue is identified with the RODEX2 code wherein burnup dependent thermal conductivity is not accounted for. The code may under-predict fuel pellet temperatures at burnups beyond approximately 20 GWd/MTU and, therefore, may under-predict the stored energy initial condition for LOCA analyses. The impact on PCT for St. Lucie Unit 1 is estimated to be +200 F.
1.2 Previous LBLOCA PCT changes are documented in References 3.1 and 3.2. Table 1 summarizes the estimated impact of the changes/errors on the St. Lucie Unit 1 LBLOCA PCT. The limiting LBLOCA PCT with the estimated effect of all the changes/errors is 2079' F. With the impact of all changes/errors, St. Lucie Unit 1 LBLOCA PCT of 2079' F continues to comply with the 10 CFR 50.46 acceptance criterion of < 2200' F.
Therefore, no reanalysis is required.
LBLOCA is currently being re-analyzed as part of the ongoing Extended Power Uprate (EPU) project, and is tentatively scheduled to be completed in the first half of Year 2010.
2.0 Changes to SBLOCA
- 2. .a Error in the radiation heat transfer model The radiation to fluid heat transfer model in ANF-RELAP is identified to have used flawed figure as the basis for determining coefficients for the correlation of emissivity of water vapor. The result is that the radiation to fluid correlation under predicts the radiative heat transfer. This issue was caused by flawed data used within the industry community. The impact on PCT for St. Lucie Unit I is estimated to be -64' F.
St. Lucie Unit I L-2009-277 Docket No. 50-335 Attachment 10 CFR 50.46 Change Report Page 2 of 5 2.1.b Legacy error in the RELAP5 series point kinetics model Previously Idaho National Laboratory (INL) announced an error in the coding of the point kinetics model in RELAP5 series of codes. Recently, INL announced that the previous error corrections were incorrect and that the recommended convergence criteria supplied with those corrections should be retained. ANF-RELAP used in SBLOCA analysis was not modified since it used strict convergence criteria and, as noted by INL, the index corrections were of secondary importance. The estimated impact on PCT for St. Lucie Unit I is +80 F. (This is a retraction of previously reported estimated impact of
-8o F.)
2.1 .c Legacy error in the RELAP5 series heat conduction model INL announced that the heat conduction solution in RELAP5 series of codes is incorrectly programmed. The error is associated with using the incorrect heat capacity when evaluating the right boundary mesh point. Instead of using the last (adjacent) mesh interval heat capacity, the code incorrectly uses the next to last mesh interval heat capacity. The impact is minimized with an increased number of mesh points. SBLOCA methodology guidelines further minimize the effect by requiring close mesh spacing at the left and right boundaries. The impact on PCT for St. Lucie Unit 1 is estimated to be 00 F.
2.1 .d Pellet thermal conductivity degradation with burnup effect on RODEX2 An issue is identified with RODEX2 code wherein burnup dependent thermal conductivity is not accounted for. The code may under-predict fuel pellet temperatures at burnups beyond approximately 20 GWd/MTU and, therefore, may under-predict the stored energy initial condition for LOCA analyses. SBLOCA analysis is insensitive to initial stored energy because sufficient excess cooling capacity exists during the blowdown phase of the transient to effectively remove any excess initial stored energy prior to the extended heatup period when PCT occurs. The estimated impact on PCT of this error for St. Lucie Unit 1 is 00 F.
2.2 Previous SBLOCA PCT changes are documented in References 3.1 and 3.2. Table 2 summarizes the estimated impact of the changes/errors on the St. Lucie Unit I SBLOCA PCT. The limiting SBLOCA PCT with the estimated effect of all the changes/errors is 17020 F. With the impact of all changes/errors, St. Lucie Unit 1 SBLOCA PCT of 17020 F continues to comply with the 10 CFR 50.46 acceptance criterion of < 22000 F.
Therefore, no reanalysis is required.
SBLOCA is currently being re-analyzed as part of the ongoing Extended Power Uprate (EPU) project, and is tentatively scheduled to be completed in the first half of Year 2010.
St. Lucie Unit I L-2009-277 Docket No. 50-335 Attachment 10 CFR 50.46 Change Report Page 3 of 5 3.0 References 3.1 FPL Letter L-2009-061, Eric Katzman to U.S. Nuclear Regulatory Commission Document Control Desk, "St. Lucie Units 1 and 2, Docket Nos. 50-335 and 50-389, Acceptance Criteria for Emergency Core Cooling Systems for Light Water Nuclear Power Reactors 10 CFR 50.46 Annual Report," March 11, 2009.
3.2 FPL Letter L-2008-254, Eric Katzman to U.S. Nuclear Regulatory Commission Document Control Desk, "St. Lucie Unit I Docket No. 50-335 Acceptance Criteria for Emergency Core Cooling Systems for Light Water Nuclear Power Reactors 10 CFR 50.46 Change Report," December 4, 2008.
St. Lucie Unit 1 L-2009-277 Docket No. 50-335 Attachment 10 CFR 50.46 Chanae Renort Page 4 of 5 Table 1: St. Lucie Unit 1 LBLOCA PCT Margin Summary Sheet Day Report Evaluation Model: EMF-2087(P)(A), Revision 0 Evaluation Model PCT: 2005 OF Absolute Net PCT Effect PCT Effect A. Prior 10 CFR Corrections - 50.46 PreviousChanges Years or Error (L-2009-061, r+ APcT 54rF 56F Reference 3.1) ___
B. Prior 10 CFR 50.46 Changes or Error APCT 0°F 0°F Corrections - Year 2009 ' _
C. Current 10 CFR 50.46 Changes
- 1. Effect of pellet thermal conductivity degradation with bumup not included in APCT + 20°F 20°F RODEX2 D. Absolute Sum of 10 CFR 50.46 Changes APCT 76 0F The sum of the PCTfrom the most recent analysis using an acceptable evaluation model and the estimates of PCT impactfor 2079OF < 2200°F changes and errors identified since this analysis
St. Lucie Unit 1 L-2009-277 Docket No. 50-33 5 Attachment 10 CFR 50.46 Change Report Page 5 of 5 Table 2: St. Lucie Unit 1 SBLOCA PCT Margin Summary Sheet Day Report Evaluation Model: XN-NF-82-49(P)(A), Revision 1, Supplement 1 Evaluation Model PCT: 1765 OF Absolute Net PCT Effect PCT Effect Prior 10 CFR 50.46 Changes or Error A. Corrections - Previous Years (L-2009-061, APCT - 70F I1T Reference 3.1) ',
Prior 10 CFR 50.46 Changes or Error Corrections - Year 2009 PC F ,
C. Current 10 CFR 50.46 Changes
- 1. Error in the radiation heat transfer model APCT - 64 0 F 64 0F
- 2. Legacy error in the RELAP5 series point kinetics model
- 3. Legacy error in the RELAP5 series heat APCT ', 0F ', 0F conduction model C ,
- 4. Effect of pellet thermal conductivity degradation with burnup not included in 0 APCT 0°F " F RODEX2 ',
D. Absolute Sum of 10 CFR 50.46 Changes APCT 83 0 F The sum of the PCTfrom the most recent analysis using an acceptable evaluation model and the estimates of PCT impactfor 1702°F < 2200°F changes and errorsidentified since this analysis