ML19260A731
| ML19260A731 | |
| Person / Time | |
|---|---|
| Site: | Point Beach  |
| Issue date: | 11/26/1979 |
| From: | Burstein S WISCONSIN ELECTRIC POWER CO. |
| To: | Harold Denton, Schwencer A Office of Nuclear Reactor Regulation |
| References | |
| NUDOCS 7912030172 | |
| Download: ML19260A731 (2) | |
Text
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O WISCONSIN Electnc eona coura.vr 231 W. MICHIGAN P.O. BOX 2046. MitWAUKEE, WI $3201 November 26, 1979 Mr. Harold R. Denton, Director Office of Nuclear Reactor Regulation U. S. NUCLEAR REGULATORY COMMISSION Washington, D. C.
20555 Attention: Mr. A. Schwencer, Chief Operating Reactor Branch #1 Gentlemen:
DOCKET NOS. 50-266 AND 50-301 REVISED ECCS EVALUATION 18% STEAM GENERATOR TUBE PLUGGING POINTBEACHNUCLEARPLAEU; NITS 1AliD2 On November 23, 1979, Mr. Trammell of your staff telephoned with three questions regarding the revised ECCS reanalysis for up to 18% plugged steam gener.itor tubes. This reanalysis was filed with the NRC by letter dated November 19, 1979. The questions and our responses are provided as follows:
1.
What inlet temperature was used in the analysis?
The inlet temperature used for the analysis was a nominal 544 F.
2.
Why was the analysis performed with a primary pressure of 2280 psia rather than 2220 psia?
The value of 2280 psia for primary system pressure was selected by the vendor for this reanalysis. There was no particular reason for selecting this value rather than 2220 psia. Both values represent the range of pressures around the nominal 2250 psia operating pressure. Selection of 2220 psia would result in some additional conservatism in the reenalysis. However, the sensitivity of this analysis to pressure is only about 2.850F change in peak clad tempera-ture (PCT) per 100 psia from actual calculations in the pressure range being considered.
In this instance, a D
change of 60 psia would thus result in only about a 1.70F change in PCT.
1456 215 7912030/ M
,)_
.w Mr. Harold R. Denton, Director November 26, 1979 3.
Why does this submittal not justify that the break analyzed is the limiting break?
Westinghouse has conducted generic sensitivity studies in the past which show that the limiting break for Westinghouse two-loop plants with 14x14 matrix fuel is a double-ended, cold-leg, guilotine pipe break with a discharge coefficient of 0.4.
The assumptions for this reanalysis do not affect these sensitivity studies and the limiting break size has not changed. Therefore only the analysis of the specific limiting break for Point Beach Nuclear Plant was submitted.
The generic sensitivity study results for two-loop plants with 14x14 fuel were previously povided to you with our letter dated March 20, 1979.
Should you have ar.y additional questions regarding this matter, please let us know.
Very truly yours, C
o M
Sol Burstein Ex utive Vice President 1456 216