ML17221A505
| ML17221A505 | |
| Person / Time | |
|---|---|
| Site: | Saint Lucie  |
| Issue date: | 11/12/1987 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML17221A504 | List: |
| References | |
| NUDOCS 8711200151 | |
| Download: ML17221A505 (3) | |
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UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO LARGE BREAK LOCA ECCS REANALYSIS FLORIDA POWER 8t LIGHT COMPANY, ET AL.
ST.
LUCIE PLANT, UNIT NO.
2 DOCKET NO. 50-389 INTRODUCTION Florida Power and Light Company (FPEL) letter L-87-327 from C. 0. Woody (FP8L) to the NRC, dated August 25, 1987, provided the results of the St. Lucie Plant, Unit No.
2 large break loss of coolant accident -(LOCA) reanalysis.
This reanalysis was performed by Combustion Engineering (CE) to reflect the use of a recent NRC-approved large break LOCA evaluation model.
The recent NRC-approved model was the subject of a letter sent on July 3, 1986 from the NRC to CE entitled, "Safety Evaluation of Combustion Engineering ECCS Large Break Evaluation Model and Acceptance for Referencing of Related Licensing Topical Reports."
The reanalysis also utilized a few numerical changes to the model input assumptions.
The previous model and input assumptions were reviewed by the staff and sent to the licensee by letter dated April 17, 1986.
EVALUATION All of the LOCA ECCS performance analyses were made using the NRC-approved CE large break LOCA ECCS evaluation model, as performed. by Combustion Engineering.
Blowdown, refill/reflood hydraulics, and hot rod temperature calculations were performed assuming fuel parameters that bound the current fuel cycle and the expected conditions for future cycles of St. Lucie 2.
The 11miting axial power shape was used to reanalyze the large break LOCA.
This was the same limiting axial powe~ shape used by the licensee in the previous analysis.
A break spectrum analysis was performed to identify the limiting break.
The double-ended guillotine break at pump discharge with a discharge coefficient of 0.6 (0.6 DEG/PD) produced the highest peak clad temperature of 2107'F, a
peak local oxidation of 7.62'A, and the highest core-wide oxidation of less than 0.70%, thereby meeting the 10 CFR 50.46 acceptance criteria lim'its of 2200'F, 171',
and 1%, respectively.
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The analysis accounts for steam generator U-tube plugging of 1430 average length tubes per steam generator.
This represents approximately 17% of the steam generator tubes.
Less than 3/ of the tubes per generator are currently plugged.
The previous analysis assumed 1250 tubes plugged per generator.
The analysis assumed a nuclear flux peaking augmentation factor of 1.00, versus the previously used value of 1.01.
The staff previously evaluated this augmentation factor as part of a technical specification change request by the licensee on November 7, 1986.
The staff's safety evaluation of Harch 5,
- 1987, in support of Amendment No.
17, found the augmentation factor of 1.00 to be acceptable.
The analysis assumed a safety injection tank (SIT) pressure of 200 psig, versus 570 psig used in the previous analysis.
The current technical specifications require the SIT pressure to be at least 570 psig.
The use of a 200 psig SIT pressure versus the actual 570 psig SIT pressure in the large break LOCA analysis does not produce significant. differences in the analysis results.
Thus, the SIT pressure used in the analysis is acceptable.
- However, if the licensee eventually desires to lower the SIT pressure Technical Specifi-cation limit in the future, the change in SIT pressure will have a significant effect on other accident analyses that must be reanalyzed, notably the small break LOCA analysis.
The analysis assumed an initial containment tempera+ure of 90'F, versus the previously used value of 60'F.
This value is more representative of the actual containment temperature; its use is acceptable.
The results o, the new analysis performed in conformance to 10 CFR 50.46 show that St. Lucie, Unit2 may be opera+ed at a core power level of 2700 Nlt and a
peak linear heat generator rate (PLHGR) of 13.0 Kw/ft.
These are equal to the existing limits for St. Lucie, Unit 2.
CONCLUSION The large break LOCA ECCS reanalysis presented for St. Lucie, Unit 2 supports operation at a core power level of 2700 Wt and a
PLHGR of 13.0 Kw/ft.
The reanalysis assumed steam generator tube plugging of up to 1430 average length tubes per steam generator.
The reanalysis also assumed other numerical changes in input assumptions, namely an augmentation factor of 1.00, an SIT pressure of 200 psig, and an initial containment temperature of 90'F.
Based upon the above, the staff concludes that the reanalysis demonstrates that St. Lucie, Unit 2 is in compliance with the criteria contained in 10 CFR 50.46.
Princi al Contributor:
E. Tourigny Dated:
November 12I 1987