L-2025-024, And Point Beach, Units 1 and 2 - Submittal of 10 CFR 50.46 Annual Reporting of Changes to, or Errors in Emergency Core Cooling System Models or Applications
| ML25069A525 | |
| Person / Time | |
|---|---|
| Site: | Saint Lucie, Point Beach, Seabrook, Turkey Point  |
| Issue date: | 03/10/2025 |
| From: | Mack K Florida Power & Light Co, Point Beach, NextEra Energy Seabrook |
| To: | Office of Nuclear Reactor Regulation, Document Control Desk |
| References | |
| L-2025-024 | |
| Download: ML25069A525 (1) | |
Text
F=PL..
ATTN: Document Control Desk U. S. Nuclear Regulatory Commission Washington, DC 20555 Re:
Florida Power & Light Company March 10, 2025 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 10 CFR 50.46 Annual Reporting of Changes to, or Errors in Emergency Core Cooling System Models or Applications L-2025-024 10 CFR 50.46 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, 2024 through December 31, 2024.
Evaluations of each reported error have concluded that re-analysis was not required.
This letter contains no new or revised regulatory commitments.
Should you have any questions regarding this report, please contact Maribel Valdez, Fleet Licensing Manager, at (561) 904-5164.
Very truly yours, K~-$?
Director, Licensing and Regulatory Compliance Florida Power & Light Company Florida Power & Light Company 700 Universe Boulevard, Juno Beach, FL 33408
L-2025-024 Page 12 of 2 Attachments (4) cc:
USNRC Regional Administrator, Region I USNRC Regional Administrator, Region II USNRC 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, Point 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
ATTACHMENT 1 Florida Power & Light Company Turkey Point Units 3 and 4
L-2025-024 Page 11 of 2 Table 1:
Turkey Point Unit 3 and 4 Small Break LOCA PCT 2024 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 Prior 1 O CFR 50.46 Changes or Error Corrections - up to Year 2023 (Reference 1) 10 CFR 50.46 Changes or Error Corrections - Year 2024 l Superheated Steam Thermal Conductivity 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 Superheated Steam Thermal Conductivity Net PCT Absolute PCT Effect Effect 0 °F 0 °F 0 °F 0 °F 0 °F 0 °F 1475 °F < 2200 °F An error was identified in WCOBRA/TRAC-TF2 related to an incorrect specific heat value used in the calculation of superheated steam thermal conductivity as part of the post-burst pellet-to-cladding gap heat transfer calculation. The incorrect specific heat value results in the superheated steam thermal conductivity being slightly under-predicted at the burst node after rupture is predicted. This leads to a minimal reduction in the gap conductance and slight over-prediction of the fuel average temperatures at the rod burst location after burst is predicted. The deficiency was qualitatively evaluated, and the nature of the error leads to an estimated peak cladding temperature impact of 0°F.
References:
- 1. Letter from Steve Catron 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-2024-011, March 13, 2024 (ML24073A190).
L-2025-024 Page 12 of 2 Table 2:
Turkey Point Unit 3 and 4 Large Break LOCA PCT 2024 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 Prior 10 CFR 50.46 Changes or Error Corrections -
up to Year 2023 (Reference 1) 10 CFR 50.46 Changes or Error Corrections - Year 2024 Superheated Steam Thermal Conductivity Sum of 1 O 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 Superheated Steam Thermal Conductivity Net PCT Absolute PCT Effect Effect 0 °F 0 °F 0 °F 0 °F 0 °F 0 °F 1981 °F < 2200 °F An error was identified in WCOBRNTRAC-TF2 related to an incorrect specific heat value used in the calculation of superheated steam thermal conductivity as part of the post-burst pellet-to-cladding gap heat transfer calculation. The incorrect specific heat value results in the superheated steam thermal conductivity being slightly under-predicted at the burst node after rupture is predicted. This leads to a minimal reduction in the gap conductance and slight over-prediction of the fuel average temperatures at the rod burst location after burst is predicted. The deficiency was qualitatively evaluated, and the nature of the error leads to an estimated peak cladding temperature impact of 0°F.
References:
- 1. Letter from Steve Catron 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-2024-011, March 13, 2024 (ML24073A190).
ATTACHMENT 2 Florida Power & Light Company St. Lucie Units 1 and 2
L-2025-024 Page J 1 of 4 Table 1:
St. Lucie Unit 1 Small Break LOCA PCT 2024 Annual Report Evaluation Methodology:
Framatome, "PWR Small Break LOCA Evaluation Model, S-RELAP5 Based," EMF-2328(P)(A)
Revision Oas supplemented by ANP-3000(P), Revision 0.
Evaluation Model PCT:
1828°F Prior 10 CFR 50.46 Changes or Error Corrections - up to Year 2023 (Reference 1) 10 CFR 50.46 Changes or Error Corrections - Year 2024 Sum of 1 O 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
References:
Net PCT Absolute PCT Effect Effect
+24 °F 84 °F None None
+24 °F 84 °F 1852 °F < 2200 °F
- 1. Letter from Steve Catron 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-2024-011, March 13, 2024 (ML24073A190).
L-2025-024 Page 12 of 4 Table 2:
St. Lucie Unit 1 Large Break LOCA PCT 2024 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 2023 (Reference 1) 10 CFR 50.46 Changes or Error Corrections - Year 2024 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
References:
Net PCT Absolute PCT Effect Effect
+6 °F 6°F None None
+6 °F 6°F 1794 °F < 2200 °F
- 1. Letter from Steve Catron 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-2024-011, March 13, 2024 (ML24073A190).
L-2025-024 Page 13 of 4 Table 3:
St. Lucie Unit 2 Small Break LOCA PCT 2024 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 1 O CFR 50.46 Changes or Error Corrections - up to Year 2023 (Reference 1) 10 CFR 50.46 Changes or Error Corrections - Year 2024 RCP Homologous Curve (Reference 2)
Sum of 1 O 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
References:
Net PCT Absolute PCT Effect Effect
-279 °F 393 °F 0 °F 0 °F
-279 °F 393 °F 1778 °F < 2200 °F
- 1. Letter from Steve Catron 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-2024-011, March 13, 2024 (ML24073A190).
- 2. Letter from Paul Rasmus to U.S. Nuclear Regulatory Commission, "10 CFR 50.46 - Emergency Core Cooling System SBLOCA 30-Day Report," L-2024-064, April 17, 2024 (ML24109A010).
L-2025-024 Page 14 of 4 Table 4:
St. Lucie Unit 2 Large Break LOCA PCT 2024 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 2023 (Reference 1) 10 CFR 50.46 Changes or Error Corrections - Year 2024 RCP Homologous Curve (Reference 2)
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
References:
Net PCT Absolute PCT Effect Effect 0 °F 0 °F 0 °F 0 °F 0 °F 0°F 1732 °F < 2200 °F
- 1. Letter from Steve Catron 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-2024-011, March 13, 2024 (ML24073A190).
- 2. Letter from Paul Rasmus to U.S. Nuclear Regulatory Commission, "10 CFR 50.46 - Emergency Core Cooling System SBLOCA 30-Day Report," L-2024-064, April 17, 2024 (ML24109A010).
ATTACHMENT 3 NextEra Energy Seabrook, LLC Seabrook Station
L-2025-024 Page 11 of 2 Table 1:
Seabrook Unit 1 Small Break LOCA PCT 2024 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 Prior 10 CFR 50.46 Changes or Error Corrections - up to Year 2023 (Reference 1) 10 CFR 50.46 Changes or Error Corrections - Year 2024 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
References:
Net PCT Absolute PCT Effect Effect 0 °F 0 °F None None 0 °F 0 °F 1373 °F < 2200 °F
- 1. Letter from Steve Catron 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-2024-011, March 13, 2024 (ML24073A190).
L-2025-024 Page 12 of 2 Table 2:
Seabrook Unit 1 Large Break LOCA PCT 2024 Annual Report Evaluation Methodology:
Westinghouse, "Code Qualification Document for Best Estimate LOCA Analysis," WCAP-12945-P-A, March 1998.
Evaluation Model PCT:
1784 °F Prior 10 CFR 50.46 Changes or Error Corrections -
up to Year 2023 (Reference 1) 10 CFR 50.46 Changes or Error Corrections - Year 2024 Volume of the Barrel-Baffle Channel 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 Volume of the Barrel-Baffle Channel Net PCT Effect Absolute PCT Effect 155 °F 155 °F 0 °F 0 °F 155 °F 155 °F 1939 °F < 2200 °F An error was identified during the course of a recent Best Estimate Large Break LOCA analysis in which the volume between the core barrel and the baffle plates within the active fuel length was modeled incorrectly. The corrected values have been evaluated for impact on the current licensing-basis analysis results. The error was evaluated to have a negligible impact on the calculated results, leading to an estimated peak cladding temperature (PCT) impact of 0°F.
References:
- 1. Letter from Steve Catron 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-2024-011, March 13, 2024 (ML24073A190).
ATTACHMENT 4 NextEra Energy Point Beach, LLC Point Beach Units 1 and 2
L-2025-024 Page 11 of 4 Table 1:
Point Beach Unit 1 Small Break LOCA PCT 2024 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 1 O CFR 50.46 Changes or Error Corrections - up to Year 2023 (Reference 1) 10 CFR 50.46 Changes or Error Corrections - Year 2024 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
- 1. Letter from Steve Catron 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-2024-011, March 13, 2024 (ML24073A190).
L-2025-024 Page 12 of 4 Table 2:
Point Beach Unit 1 Large Break LOCA PCT 2024 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 1 O CFR 50.46 Changes or Error Corrections -
up to Year 2023 (Reference 1) 1 O CFR 50.46 Changes or Error Corrections - Year 2024 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 Steve Catron 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-2024-011, March 13, 2024 (ML24073A190).
L-2025-024 Page I 3 of 4 Table 3:
Point Beach Unit 2 Small Break LOCA PCT 2024 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 2023 (Reference 1) 10 CFR 50.46 Changes or Error Corrections - Year 2024 Sum of 1 O 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 Steve Catron 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-2024-011, March 13, 2024 (ML24073A190).
L-2025-024 Page 14 of 4 Table 4:
Point Beach Unit 2 Large Break LOCA PCT 2024 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 2023 (Reference 1) 10 CFR 50.46 Changes or Error Corrections - Year 2024 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 Steve Catron 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-2024-011, March 13, 2024 (ML24073A190).