ML20084U839

AO 50-265/74-30:on 741230,excessive Leakage Found from Feedwater Check Valves 2-220-58B & 62B.Caused by Seals Allowing Flow Between Seat Ring & Valve Body.Stainless Steel Rings Replaced W/Silicon Rubber O Rings

ML20084U839
Person / Time
Site:	Quad Cities
Issue date:	03/10/1975
From:	COMMONWEALTH EDISON CO.
To:	US ATOMIC ENERGY COMMISSION (AEC)
References
AO-50-265-74-30, NUDOCS 8306290307
Download: ML20084U839 (2)

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nv REPORT NUMBER: A0-50-265/74-30

_ REPORT DATE: March 10, 1975 OCCURRENCE DATE: December 30, 1974 FACILITY: Quad-Cities NucIcar Power Station Cordova, Illinois 61242 IDENTIFICATION OF OCCURRENCE:

Excessive leakage from feedwater check valves 2-220-58B and 2-220-628 CONDITIONS PRIOR TO OCCURRENCE:

Unit 2 in cold shutdown cendition for refueling outage.

DESCRIPTION OF OCCURRENCE:

Feedwater check valves 2-220-588 and 628 failed to pass the local leak rate test limitation of 18.36 scfh at 48 psig as allowed by Technical Specifica-tion 4.7.A.2.i.(2)(b). The test conditions and results were as follows:

TEST TEST VALVES LEAKAGE VALVE PRESSURE FOR PRESSURIZING (SCFli)

(PSIC) 2-220-58B 48 2-220-115B 2,016 2-220-116B 2-220-62B 48 2-220-86B 2,647 2-220-87B Following modification of the seal rings and remachining of the valve seating surfaces, the following leak rates were obtained:

• VALVE t.EAKAGE (SCFH) TEST DATE 2-220-58B 10.3 January 18, 1975 2-220-628 1.357 February 18, 1975 DESIGNATION OF APPARENT CAUSE OF OCCURRENCE:

Ecutonent Failure - The excessive leakage of the "B" loop feedsater check valves was found to be the result of seals allowing flow between the seat ring and the valve body.

8306290307 750310 PDR ADDCK 05000265 S PM ,

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o ANALYSIS OF OCCURRENCE:

j If an accident were to occur, the 2-220-588 valve would determine the maxi-4 mum amount of leakage from the reactor, if an unlikely situation is assumed

where the feedwater system experiences a rupture between the 2-220-58B and 2-220-62B valves, the '58B check valve would leak excessively if the reactor 4

were at the leak rate test pressure of 48 psig. Since the reactor would be at a pressure on the order of 950 psig, these valves would be driven into

their seats allowing for a better seal and thereby reducing leakage. If 1 this break occurred, it could happen either inside the drywell or the out-j board main steam isolation valve room. The main. con

:nts of this leakage j would be feedwater, steam from the reactor steam space, and fission product gases. The water could leak only until the vessel level was lowered to the i feedwater sparger level. Any steam that leaked would condense and this water contained inside the plant by the floor drain system. Any gases that

escaped would be handled by the Standby Gas Treatment System. At no time

j. would the fuel be in danger of being uncovered due to this situation. i j When deinerted, a drywell entry could be made and the 1-220-57B valve l

could be closed preventing any further leakage. This type of worst case l i

analysis puts the plant in an inoperable status, but no radioactivity would be ' released to the environs and the health and safety of the public would l not be affected.

CORRECTIVE ACTION

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The corrective action has been the installation of a modification (M-4-2 l 15) which replaced the stainless steel seal rings with Viton silicon rubber

'V" rings. The valve seating surfaces were also remachined.

FAILURE DATA:

} This is the first time these valves have been tested since the pre-operational tests; however, due to leakage previously experienced at Dresden Station, the modification was initiated to replace the stainless steel seat seal rings with rubber "0" rings. This modification is designed to reduce check valve l eakage.

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