ML20247N124
| ML20247N124 | |
| Person / Time | |
|---|---|
| Site: | Point Beach  |
| Issue date: | 07/27/1989 |
| From: | Fay C WISCONSIN ELECTRIC POWER CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| CON-NRC-89-088, CON-NRC-89-88 GL-88-14, IEB-80-04, IEB-80-4, VPNPD-89-417, NUDOCS 8908020284 | |
| Download: ML20247N124 (3) | |
Text
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POWER COMPANY h
231 W Mehigon, Po. Box 2046. Mitwoukee, WI S3201 (414) 221-2345 4
VPNPD 417
.NRC 088 July 27, 1989 Document Control Desk U.
S.. NUCLEAR REGULATORY COMMISSION Mail Station Pl-137 Washington, D.C.
20555 Gentlemen:
DOCKET NOS. 50-266 and 50-301 SUPPLEMENT TO BULLETIN 80-04 AND TO GENERIC LETTER 88-14 RESPONSES POINT BEACH NUCLEAR PLANT, UNITS 1 AND 2 In a letter dated February 20, 1989, Wisconsin Electric presented information responding to the concerns of Generic Letter 88-14, Instrument Air Problems Affecting Safety-Related Equipment.
In that letter, the fail-safe design of the auxiliary feedwater system for a loss of instrument air was considered potentially unacceptable for a Main Steam Line Break inside containment.
The discharge valves for the electric motor-driven auxiliary feedwater pumps at the Point Beach Nuclear Plant (PBNP) fail open on'a loss of instrument air.
In the February submittal, we committed to study and determine a course of action for this issue prior to July 31, 1989.
The main steara line break accident inside containment is the accident of primary concern, since the addition of feed water to a' faulted steam generator would continue to blow down to containment as steam.
This would result in a potential containment over-pressure condition.
Wisconsin Electric requested and received the results of a reference 2-Loop Plant Containment Pressure Analyses for Steam Line Break from the Point Beach NSSS vendor (Westinghouse Electric Jorp.) to evaluate the potential for overpressurization of containment.
Detailed evaluations of the inputs and results of the reference
, analyses were performed.
We determined that the reference analyses differed from PBNP in two key areas: 1) modelling of containment structural heat sink and 2) the assumption for auxiliary feedwater flow rates..The PBNP containment is not insulated, which results in significantly more energy removal than the insulated containment assumed in the reference plant analysis.
The maximum PBNP auxiliary feedwater flow rate could be as high as 1060 gpm to a faulted steam generator compared to approximately 600 gpm in the reference plant analysis.
The PBNP AFW flow rate would be limited to about 676 gpm if the AFW pump discharge valve did not fail open.
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U.S. NRC Document Control Desk j
July 27, 1989 Page 2 The maximum peak pressure calculated for the reference plant was 59.8 psig.
The counterbalancing effects of higher auxiliary feedwater input and higher containment structural heat removal were quantified for PBNP.
The maximum peak pressure for PBNP was determined to be approximately 61 psig, slightly in excess of the PBNP containment design pressure of 60 psig.
This evaluation assumes-the concurrent failure of a containment spray pump and the
_AFW pump discharge valve.
This scenario conservatively assumes the maximum. steam generator water inventory, off-site power available, and a core decay heat rate of 120% of the ANS-1971 standard.
The results of our evaluation show that if the concurrent failure of-the AFW pump discharge valve and a containment spray pump is assumed, containment design pressure, 60 psig, may be exceeded.
The resulting pressure, approximately 61 psig, is well within the structural design margin of the containments, which were verified during a pre-operational structural integrity test at 115% of design pressure (69 psig).
If only one of these failures occurs, contain-ment pressure would remain below the decign pressure of 60 psig.
Due to the sustantial amount of consert ism in these analyses, the very low probability of failure in both of these two components, the risks associated with modifying the AFW system to limit flow, and the minimal radiological consequences of a main steam line break, Wisconsin Electric believes that further action with respect to this issue is not required.
We will, however, revise the PBNP FSAR to remove inaccuracies in section 14.2.5, Rupture of a Steam Pipe, regarding auxiliary feedwater flow rate and energy addition, which were based on the original WE IE Bulletin 80-04 response.
In addition to the above concerns, in our February 20, 1989 response to' Generic Letter 88-14 we committed to provide you with the results of loss of instrument air tests, conducted during the Unit 1 spring 1989 outage, on the Point Beach Unit 1 containment purge and exhaust valves.
The tests were conducted to determine if leakage existed past the air accumulator check valves.
The accumulators supply valve seal pressure during loss of instrument air pressure.
During the tests, we found that three out of four check valves leaked.
These check valves were replaced.
The tests also revealed other system air leaks, associated with the solenoid operated vent, which could potentially affect the integrity of the purge supply and exhaust valve seals.
We completed temporary modifications to isolate these leakage problems associated with the vent valves and restored the air supply to the purge supply and exhaust valve seals to full operability.
Permanent corrective action consisting of replacing the defective components associated with the leaking vent
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U.S.
NRC Document Control Desk July 27, 1989 Page 3 valves will be taken during the 1990 Unit 1 outage.
These repairs could not be completed during the 1989 outage due to the unavailability-J of replacement parts.
If you have any additional questions, please contact us.
Very truly yours, Bu (g' C. W. ' Fay Vice President Nuclear Power Copies to NRC Regional Administrator, Region III NRC Resident Inspector Subscribe and sworn to before me this 27 day of
\\,1 1989.
co Notary Pablic, State 6f-Wisconsin My Commission expires 5-27-9d.
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