ML24073A190
| ML24073A190 | |
| Person / Time | |
|---|---|
| Site: | Saint Lucie, Point Beach, Seabrook, Turkey Point  |
| Issue date: | 03/13/2024 |
| From: | Catron S Florida Power & Light Co, Point Beach, NextEra Energy Seabrook |
| To: | Office of Nuclear Reactor Regulation, Document Control Desk |
| References | |
| L-2024-011 | |
| Download: ML24073A190 (1) | |
Text
F=PL.
March 13, 2024 ATIN: Document Control Desk U. S. Nuclear Regulatory Commission Washington, DC 20555 Re:
Florida Power & Light Company Turkey Point Units 3 and 4, Docket Nos. 50-250, 50-251 Florida Power & Light Company St. Lucie Units 1 and 2, Docket Nos. 50-335, 50-389 NextEra Energy Seabrook, LLC Seabrook Station, Docket No. 50-443 NextEra Energy Point Beach, LLC Point Beach Units 1 and 2, Docket Nos. 50-266, 50-301 L-2024-011 10 CFR 50.46 10 CFR 50.46 Annual Reporting of Changes to, or Errors in Emergency Core Cooling System Models or Applications Pursuant to 10 CFR 50.46(a)(3)(ii), the nature of any change to or error discovered in the evaluation models for emergency core cooling systems (ECCS), or in the application of such models, that affect the fuel cladding temperature calculations for Turkey Point Nuclear Plant, Units 3 and 4; and St. Lucie Nuclear Plant, Units 1 and 2; Seabrook Station; and Point Beach Nuclear Plant, Units 1 and 2 are reported in the attachments to this letter by Florida Power & Light Company (FPL), on behalf of itself and its affiliates, NextEra Energy Seabrook, LLC and NextEra Energy Point Beach, LLC. The data interval for this report is from January 1, 2023 through December 31, 2023.
Evaluations of each reported error have concluded that re-analysis was not required.
This letter contains no new or revised regulatory commitments.
Florida Power & Light Company 700 Universe Boulevard, Juno Beach, FL 33408
L-2024-011 Page 12 of 2 Should you have any questions regarding this report, please contact Mr. Kenneth Mack, Fleet Licensing Manager, at (561) 904-3635.
Very truly yours, Steve Catr n Licensing d Regulatory Compliance Director - Nuclear Fleet Florida Power & Light Company Attachments (4) cc:
USNRC Regional Administrator, Region I USN RC Regional Administrator, Region II USN RC Regional Administrator, Region Ill USNRC Project Manager, Seabrook Station USNRC Project Manager, St. Lucie Nuclear Plant USNRC Project Manager, Turkey Point Nuclear Plant USNRC Project Manager, PoinJ Beach Nuclear Plant USNRC Senior Resident Inspector, Seabrook Station USNRC Senior Resident Inspector, St. Lucie Nuclear Plant USNRC Senior Resident Inspector, Turkey Point Nuclear Plant USNRC Senior Resident Inspector, Point Beach Nuclear Plant Florida Power & Light Company 700 Universe Boulevard, Juno Beach, FL 33408
ATTACHMENT 1 Florida Power & Light Company Turkey Point Units 3 and 4
L-2024-011 Page 11 of 2 Table 1:
Turkey Point Unit 3 and 4 Small Break LOCA PCT 2023 Annual Report Evaluation Methodology:
Westinghouse, "Engineering Summary Report of the Turkey Point Units 3 and 4 Loss-of-Coolant Accident (LOCA) Analysis with the FULL SPECTRUM LOCA (FSLOCA) Methodology,"
WCAP-18597-P, Revision 0, November 2020.
Evaluation Model PCT:
1475 °F (Reference 1)
Prior 10 CFR 50.46 Changes or Error Corrections -
up to 12/31/2022 10 CFR 50.46 Changes or Errors Corrections - Year 2023 Sum of 10 CFR 50.46 Changes or Errors Corrections The sum of the PCT from the most recent analysis using an acceptable evaluation model and the estimates of PCT impact for changes and errors identified since this analysis Summary of 2023 Changes and Errors:
Net PCT Absolute PCT Effect Effect N/A1 N/A 0 °F 0 °F 0 °F 0 °F 1475 °F < 2200 °F Vapor/Continuous Liquid lnterfacial Drag Coefficient In Churn-Turbulent Flow Regime:
Two deficiencies were identified in the calculation of the churn-turbulent vapor/continuous liquid interfacial drag coefficient calculation within WCOBRA/TRAC-TF2 due to the potential calculation of a negative critical liquid fraction. The negative liquid crystal fraction results in an over-prediction of the vapor/continuous liquid interfacial area and a negative vapor/continuous liquid interfacial drag coefficient which leads to a significantly large interfacial drag coefficient.
These closely-related group of deficiencies was qualitatively evaluated, and the nature of the deficiencies leads to an estimated peak cladding temperature impact of O °F.
Reference:
- 1. Letter from U.S. Nuclear Regulatory Commission to B. Coffey, "Turkey Point Nuclear Generating Units Nos. 3 and 4 - Issuance of Amendments Nos. 296 and 289 Regarding Implementation of Full Spectrum Loss-of-Coolant Accident (FSLOCA) Methodology (EPID L-2021-LLA-0070)," May 24, 2022 (ML22028A066).
1 FSLOCA is a new methodology; there is no information to report from the previous year.
L-2024-011 Page 12 of 2 Table 2:
Turkey Point Unit 3 and 4 Large Break LOCA PCT 2023 Annual Report Evaluation Methodology:
Westinghouse, "Engineering Summary Report of the Turkey Point Units 3 and 4 Loss-of-Coolant Accident (LOCA) Analysis with the FULL SPECTRUM LOCA (FSLOCA) Methodology,"
WCAP-18597-P, Revision 0, November 2020.
Evaluation Model PCT:
1981 °F (Reference 1)
Prior 10 CFR 50.46 Changes or Error Corrections --
up to 12/31/2022 10 CFR 50.46 Changes or Errors Corrections -- Year 2023 Sum of 10 CFR 50.46 Changes or Errors Corrections The sum of the PCT from the most recent analysis using an acceptable evaluation model and the estimates of PCT impact for changes and errors identified since this analysis Summary of 2023 Changes and Errors:
Net PCT Absolute PCT Effect Effect N/A1 N/A 0 °F 0 °F 0 °F 0 °F 1981 °F < 2200 °F Vapor/Continuous Liquid lnterfacial Drag Coefficient In Churn-Turbulent Flow Regime:
Two deficiencies were identified in the calculation of the churn-turbulent vapor/continuous liquid interfacial drag coefficient calculation within WCOBRA/TRAC-TF2 due to the potential calculation of a negative critical liquid fraction. The negative liquid crystal fraction results in an over-prediction of the vapor/continuous liquid interfacial area and a negative vapor/continuous liquid interfacial drag coefficient which leads to a significantly large interfacial drag coefficient.
These closely-related group of deficiencies was qualitatively evaluated, and the nature of the deficiencies leads to an estimated peak cladding temperature impact of O °F.
Reference:
- 1. Letter from U.S. Nuclear Regulatory Commission to 8. Coffey, "Turkey Point Nuclear Generating Units Nos. 3 and 4 - Issuance of Amendments Nos. 296 and 289 Regarding Implementation of Full Spectrum Loss-of-Coolant Accident (FSLOCA) Methodology (EPID L-2021-LLA-0070)," May 24, 2022 (ML22028A066).
1 FSLOCA is a new methodology; there is no information to report from the previous year.
ATTACHMENT 2 Florida Power & Light Company St. Lucie Units 1 and 2
L-2024-011 Page 11 of 4 Table 1:
St. Lucie Unit 1 Small Break LOCA PCT 2023 Annual Report Evaluation Methodology:
Framatome, "PWR Small Break LOCA Evaluation Model, S-RELAP5 Based," EMF-2328(P)(A)
Revision O as supplemented by ANP-3000(P), Revision 0.
Evaluation Model PCT:
1828°F Prior 10 CFR 50.46 Changes or Error Corrections - up to Year 2022 (Reference 1) 10 CFR 50.46 Changes or Error Corrections - Year 2023 Sum of 10 CFR 50.46 Changes or Error Corrections The sum of the PCT from the most recent analysis using an acceptable evaluation model and the estimates of PCT impact for changes and errors identified since this analysis
Reference:
Net PCT Absolute PCT Effect Effect
+24 °F 84 °F None None
+24 °F 84 °F 1852 °F < 2200 °F
- 1. Letter from D. Strand to U.S. Nuclear Regulatory Commission, "10 CFR 50.46 Annual Reporting of Changes to, or Errors in Emergency Core Cooling System Models or Applications," L-2023-028, March 27, 2023 (ML23086C017).
L-2024-011 Page 12 of 4 Table 2:
St. Lucie Unit 1 Large Break LOCA PCT 2023 Annual Report Evaluation Methodology:
Framatome, "Realistic Large Break LOCA Methodology for Pressurized Water Reactors," EMF-2103(P)(A) Revision Oas supplemented by ANP-2903(P), Revision 1.
Evaluation Model PCT:
1788°F Prior 10 CFR 50.46 Changes or Error Corrections -
up to Year 2022 (Reference 1) 10 CFR 50.46 Changes or Error Corrections - Year 2023 Sum of 10 CFR 50.46 Changes or Error Corrections The sum of the PCT from the most recent analysis using an acceptable evaluation model and the estimates of PCT impact for changes and errors identified since this analvsis
Reference:
Net PCT Effect Absolute PCT Effect
+6 °F 6°F None None
+6 °F 6°F 1794 °F < 2200 °F
- 1. Letter from D. Strand to U.S. Nuclear Regulatory Commission, "10 CFR 50.46 Annual Reporting of Changes to, or Errors in Emergency Core Cooling System Models or Applications," L-2023-028, March 27, 2023 (ML23086C017).
L-2024-011 Page I 3 of 4 Table 3:
St. Lucie Unit 2 Small Break LOCA PCT 2023 Annual Report Evaluation Methodology:
Framatome, "PWR Small Break LOCA Evaluation Model, S-RELAP5 Based," EMF-2328(P)(A)
Revision.a.
Evaluation Model PCT:
2057°F Prior 10 CFR 50.46 Changes or Error Corrections - up to Year 2022 (Reference 1) 10 CFR 50.46 Changes or Error Corrections - Year 2023 Sum of 10 CFR 50.46 Changes or Error Corrections The sum of the PCT from the most recent analysis using an acceptable evaluation model and the estimates of PCT impact for changes and errors identified since this analysis
Reference:
Net PCT Effect Absolute PCT Effect
-279°F 393 °F None None
-279°F 393 °F 1778 °F < 2200 °F
- 1. Letter from D. Strand to U.S. Nuclear Regulatory Commission, "10 CFR 50.46 Annual Reporting of Changes to, or Errors in Emergency Core Cooling System Models or Applications," L-2023-028, March 27, 2023 (ML23086C017).
L-2024-011 Page 14 of 4 Table 4:
St. Lucie Unit 2 Large Break LOCA PCT 2023 Annual Report Evaluation Methodology:
Framatome, "Realistic Large Break LOCA Methodology for Pressurized Water Reactors," EMF-2103(P)(A) Revision 0.
Evaluation Model PCT:
1732°F Prior 10 CFR 50.46 Changes or Error Corrections - up to Year 2022 (Reference 1) 10 CFR 50.46 Changes or Error Corrections - Year 2023 Sum of 10 CFR 50.46 Changes or Error Corrections The sum of the PCT from the most recent analysis using an acceptable evaluation model and the estimates of PCT impact for changes and errors identified since this analysis
Reference:
Net PCT Effect Absolute PCT Effect 0 °F 0 °F None None 0 °F 0 °F 1732 °F < 2200 °F
- 1. Letter from D. Strand to U.S. Nuclear Regulatory Commission, "10 CFR 50.46 Annual Reporting of Changes to, or Errors in Emergency Core Cooling System Models or Applications," L-2023-028, March 27, 2023 (ML23086C017).
ATTACHMENT 3 NextEra Energy Seabrook, LLC Seabrook Station
L-2024-011 Page 11 of 2 Table 1:
Seabrook Unit 1 Small Break LOCA PCT 2023 Annual Report Evaluation Methodology:
Westinghouse, "Westinghouse Small Break ECCS Evaluation Model Using the NOTRUMP Code," WCAP-10054-P-A, August 1985 and Addendum 2, Revision 1, July 1997.
Evaluation Model PCT:
1373 °F (Reference 1)
Prior 10 CFR 50.46 Changes or Error Corrections -
up to 12/31/2022 (Reference 2) 10 CFR 50.46 Changes or Errors Corrections - year 2023 Sum of 10 CFR 50.46 Changes or Errors Corrections The sum of the PCT from the most recent analysis using an acceptable evaluation model and the estimates of PCT impact for changes and errors identified since this analysis
References:
Net PCT Absolute PCT Effect Effect 0 °F 0 °F None None 0 °F 0 °F 1373 °F < 2200 °F
- 1. Letter from M. Warner to U.S. Nuclear Regulatory Commission, "License Amendment Request 04-03, Application for Stretch Power Uprate," NYN-04016, March 17, 2004.
- 2. Letter from D. Strand to U.S. Nuclear Regulatory Commission, "10 CFR 50.46 Annual Reporting of Changes to, or Errors in Emergency Core Cooling System Models or Applications/' L-2023-028, March 27, 2023 (ML23086C017).
L-2024-011 Page 12 of 2 Table 2:
Seabrook Unit 1 Large Break LOCA PCT 2023 Annual Report Evaluation Methodology:
Westinghouse, "Code Qualification Document for Best Estimate LOCA Analysis," WCAP-12945-P-A, March 1998.
Evaluation Model PCT:
1784 °F (Reference 1)
Prior 10 CFR 50.46 Changes or Error Corrections -
up to 12/31/2022 (Reference 2) 10 CFR 50.46 Changes or Errors Corrections - year 2022 Error in Flow Area and Volume of Thimble Components (Reference 3)
Sum of 10 CFR 50.46 Changes or Errors Corrections The sum of the PCT from the most recent analysis using an acceptable evaluation model and the estimates of PCT impact for changes and errors identified since this analysis
References:
Net PCT Absolute PCT Effect Effect 155 °F 155 °F None None 0 °F 0 °F 155 °F 155 °F 1939 °F < 2200 °F
- 1. Letter from M. Warner to U.S. Nuclear Regulatory Commission, "License Amendment Request 04-03, Application for Stretch Power Uprate," NYN-04016, March 17, 2004.
- 2. Letter from D. Strand to U.S. Nuclear Regulatory Commission, "1 O CFR 50.46 Annual Reporting of Changes to, or Errors in Emergency Core Cooling System Models or Applications," L-2023-028, March 27, 2023 (ML23086C017).
- 3. Letter from D. Strand to U.S. Nuclear Regulatory Commission, "1 O CFR 50.46 - Emergency Core Cooling System LBLOCA 30-Day Report," L-2024-003, January 11, 2024 (ML24011A130).
ATTACHMENT 4 NextEra Energy Point Beach, LLC Point Beach Units 1 and 2
L-2024-011 Page 11 of 4 Table 1:
Point Beach Unit 1 Small Break LOCA PCT 2023 Annual Report Evaluation Methodology:
Westinghouse, "Westinghouse Small Break ECCS Evaluation Model Using the NOTRUMP Code,"
WCAP-10054-P-A, August 1985 and Addendum 2, Revision 1, July 1997.
Evaluation Model PCT:
1049°F Prior 10 CFR 50.46 Changes or Error Corrections - up to Year 2022 (Reference 1) 10 CFR 50.46 Changes or Error Corrections - Year 2023 Sum of 10 CFR 50.46 Changes or Error Corrections The sum of the PCT from the most recent analysis using an acceptable evaluation model and the estimates of PCT impact for changes and errors identified since this analysis
Reference:
Net PCT Effect Absolute PCT Effect 0°F 0°F None None 0°F 0°F 1049°F < 2200°F
- 1. Letter from D. Strand to U.S. Nuclear Regulatory Commission, 1110 CFR 50.46 Annual Reporting of Changes to, or Errors in Emergency Core Cooling System Models or Applications," L-2023-028, March 27, 2023 (ML23086C017).
L-2024-011 Page 12 of 4 Table 2:
Point Beach Unit 1 Large Break LOCA PCT 2023 Annual Report Evaluation Methodology:
Westinghouse, "Realistic Large-Break LOCA Evaluation Methodology Using the Automated Statistical Treatment of Uncertainty Method (ASTRUM)," WCAP-16009-P-A, January 2005.
Westinghouse, "Application of Best Estimate Large Break LOCA Methodology to Westinghouse PWRs with Upper Plenum Injection," WCAP-14449-P-A Revision 1, October 1999.
Evaluation Model PCT:
1975°F Prior 10 CFR 50.46 Changes or Error Corrections -
up to Year 2022 (Reference 1) 10 CFR 50.46 Changes or Error Corrections - Year 2023 Sum of 10 CFR 50.46 Changes or Error Corrections The sum of the PCT from the most recent analysis using an acceptable evaluation model and the estimates of PCT impact for changes and errors identified since this analysis
Reference:
Net PCT Effect Absolute PCT Effect
+210°F 210°F None None
+210°F 210°F 2185°F < 2200°F
- 1. Letter from D. Strand to U.S. Nuclear Regulatory Commission, "10 CFR 50.46 Annual Reporting of Changes to, or Errors in Emergency Core Cooling System Models or Applications," L-2023-028, March 27, 2023 (ML23086C017).
L-2024-011 Page I 3 of 4 Table 3:
Point Beach Unit 2 Small Break LOCA PCT 2023 Annual Report Evaluation Methodology:
Westinghouse, "Westinghouse Small Break ECCS Evaluation Model Using the NOTRUMP Code,"
WCAP-10054-P-A, August 1985 and Addendum 2, Revision 1, July 1997.
Evaluation Model PCT:
1103°F Prior 10 CFR 50.46 Changes or Error Corrections - up to Year 2022 (Reference 1) 10 CFR 50.46 Changes or Error Corrections - Year 2023 Sum of 10 CFR 50.46 Changes or Error Corrections The sum of the PCT from the most recent analysis using an acceptable evaluation model and the estimates of PCT impact for changes and errors identified since this analysis
Reference:
Net PCT Effect Absolute PCT Effect 0°F 0°F None None 0°F 0°F 1103°F < 2200°F
- 1. Letter from D. Strand to U.S. Nuclear Regulatory Commission, "10 CFR 50.46 Annual Reporting of Changes to, or Errors in Emergency Core Cooling System Models or Applications," L-2023-028, March 27, 2023 (ML23086C017).
L-2024-011 Page 14 of 4 Table 4:
Point Beach Unit 2 Large Break LOCA PCT 2023 Annual Report Evaluation Methodology:
Westinghouse, "Realistic Large-Break LOCA Evaluation Methodology Using the Automated Statistical Treatment of Uncertainty Method (ASTRUM)," WCAP-16009-P-A, January 2005.
Westinghouse, "Application of Best Estimate Large Break LOCA Methodology to Westinghouse PWRs with Upper Plenum Injection," WCAP-14449-P-A Revision 1, October 1999.
Evaluation Model PCT:
1810°F Prior 10 CFR 50.46 Changes or Error Corrections -
up to Year 2022 (Reference 1) 10 CFR 50.46 Changes or Error Corrections - Year 2023 Sum of 10 CFR 50.46 Changes or Error Corrections The sum of the PCT from the most recent analysis using an acceptable evaluation model and the estimates of PCT impact for changes and errors identified since this analysis
Reference:
Net PCT Effect Absolute PCT Effect
+248°F 340°F None None
+248°F 340°F 2058°F < 2200°F
- 1. Letter from D. Strand to U.S. Nuclear Regulatory Commission, "1 O CFR 50.46 Annual Reporting of Changes to, or Errors in Emergency Core Cooling System Models or Applications," L-2023-028, March 27, 2023 (ML23086C017).