L-2003-227, 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
0 FPL Florida Power & Light Company, 6501 S. Ocean Drive, Jensen Beach, FL 34951 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 1400F. 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) 10F Estimated impact of RCS flow reduction from 363,000 gpm to 355,000 gpm 150F (previously reported in L-2002-196)
Estimated impact due to model change from 1985 EM to 1999 EM
-140 0F Estimated impact of locked-rotor k-factor error 110 0F Cumulative Change 2660F Total PCT Change