L-2003-227, LBLOCA Evalution Model 30-Day 10 CFR 50.46 Report

LBLOCA Evalution Model 30-Day 10 CFR 50.46 Report

ML032580420
Person / Time
Site:	Saint Lucie
Issue date:	09/10/2003
From:	Jefferson W Florida Power & Light Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
L-2003-227
Download: ML032580420 (3)

Text

Florida Power & Light Company, 6501 S.Ocean Drive, Jensen Beach, FL 34951 0

FPL September 10, 2003 L-2003-227 10 CFR 50.46 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555 Re: St. Lucie Unit 2 Docket No. 50-389 LBLOCA Evaluation Model 30-Day 10 CFR 50.46 Report Westinghouse Electric A) is the current fuel vendor for St. Lucie Unit 2, and performs the calculations to demonstrate that the Unit 2 emergency core cooling system (ECCS) performance conforms to 10 CFR 50.46. W employs an acceptable evaluation model consistent with 10 CFR 50, Appendix K. Model changes/errors in the large break loss-of-coolant accident (LBLOCA) analysis has resulted in a significant change to the calculated peak cladding temperature (PCT), and is hereby reported pursuant to 10 CFR 50.46(a)(3)(ii). The small break loss-of-coolant accident (SBLOCA) analysis PCT remains unchanged from that reported in FPL letter L-2002-196 dated October 15, 2002.

Please contact George Madden at 772-467-7155 if you have any questions regarding this matter.

Vice President St. Lucie Plant WJ/spt Attachment N ()(:)

an FPL Group company

St. Lucie Unit 2 Docket No. 50-389 L-2003-227 Attachment Page 1 St. Lucie Unit 2 10 CFR 50.46 LBLOCA 30-Day Report Westinghouse Electric W) is the current fuel vendor for St. Lucie Unit 2, and performs the calculations to demonstrate that the Unit 2 emergency core cooling system (ECCS) performance conforms to 10 CFR 50.46. W employs an acceptable evaluation model consistent with 10 CFR 50, Appendix K. Model changes/errors in the large break loss-of-coolant accident (LBLOCA) analysis have resulted in a significant change to the calculated peak cladding temperature (PCT), and is hereby reported pursuant to 10 CFR 50.46(a)(3)(ii). The small break loss-of-coolant accident (SBLOCA) analysis PCT remains unchanged from that reported in Reference 1.

Nature of the Model Changes and Corrective Action Error in the Locked-Rotor K-Factor Value Description of Deviation The St. Lucie Unit 2 ECCS performance analyses PCTs applicable to the current operating cycle (Cycle 14) were previously reported in References I and 2.

A review of the LBLOCA analysis revealed that the reactor coolant pump locked rotor k-factor used In the analysis was incorrect. The locked-rotor k-factor using the as-built reactor coolant pump test data was found to be approximately 25% greater than the value currently used. The k-factor value was corrected and the impact of the k-factor error on the limiting break was estimated using the updated evaluation model, 1999 EM (Reference 3). This evaluation model (1999 EM) was approved by the NRC in Reference 4 for application to CE PWRs.

Impact of the Code Error The impact of the model change (1999 EM) on the LBLOCA analysis is estimated to be a reduction in the PCT of 140 0F. The impact of the k-factor error on the LBLOCA PCT is estimated to be an increase of I 10F.

The cumulative change of the PCT changes becomes 2660F as provided in the table below. The final LBLOCA PCT becomes 21360F.

References

1. FPL Letter L-2002-196, St. Lucie Unit 2, Docket No. 50-389, Proposed License Amendment - Reduce the Minimum Reactor Coolant System Flow, dated October 15,2002.

St. Lucie Unit 2 Docket No. 50-389 L-2003-227 Attachment Page 2 St. Lucie Unit 2 10 CFR 50.46 LBLOCA 30-Day Report

2. FPL Letter L-2003-078, St. Lucie Units I 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, dated March 26, 2003.

3. CENPD-132, Supplement 4-P-A, Calculative Methods for the CE Nuclear Power Large Break LOCA Evaluation Model, dated March 2001.

4. NRC Letter, S. A. Richards (NRC) to P. W. Richardson (Westinghouse), Safety Evaluation of Topical Report CENPD-132, Supplement 4, Revision 1, Calculative Methods for the CE Nuclear Power Large Break LOCA Evaluation Model, dated December 15,2000.

Unit 2 LBLOCA Summary - PCT Current evaluation model calculated LBLOCA PCT 21500F Estimated impact of STRIKIN-I errors (previously reported in L-2003-078) 1 0F Estimated impact of RCS flow reduction from 363,000 gpm to 355,000 gpm 150 F (previously reported in L-2002-196)

Estimated impact due to model change from 1985 EM to 1999 EM -140 0 F Estimated impact of locked-rotor k-factor error 110 0 F Cumulative Change 2660 F Total PCT Change -140F Final LBLOCA PCT 2136F