ML20084U473

AO 50-265/75-6:on 750125,feedwater Check Valves 2-220-58A & 62A Failed to Pass Local Leak Rate Test Tech Spec Limit. Caused by Seals Allowing Flow Between Seat Ring & Valve Body.Stainless Steel Seat Seal Rings Replaced

ML20084U473
Person / Time
Site:	Quad Cities
Issue date:	03/17/1975
From:	Kalivianakis N COMMONWEALTH EDISON CO.
To:
Shared Package
ML20084U443	List: ML20084U440 ML20084U458 ML20084U473
References
AO-50-265-75-6, NUDOCS 8306290062
Download: ML20084U473 (2)

Text

.

C, O REPORT NUMBER: A0-50-265/75-6 RE' PORT DATE: March 17, 1975 OCCURRENCE DATE: January 25, 1975 FACILITY: Quad-Cities Nuclear Power Station Cordova, IL 61242 IDENTIFICATION OF OCCURRENCE:

Excessive leakage from feedwater check valves 2-220-58A and 62A.

CONDITIONS PRIOR TO OCCURRENCE:

Unit 2 was in the cold shutdown condition for the current refueling outage.

DESCRIPTION OF OCCURRENCE:

Feedwater check valves 2-220-58A and 62A failed to pass the local leak rate test limi-tation of 18.36 SCFH at 48 PSIG as allowed by Technical Specification 4.7.A.2.1.(2)(b) .

The test conditions and results were as follows:

TEST TEST VALVES vat.VE PRESSURE (PSIG) FOR PRESSURIZING LEAKAGE (SCFH) 2-220-58A 48 2-220-Il5A 97 2-220 - 116.^.

2-220-62A 48 2-220-86A 472 2-220-87A Following modification of the seal rings and remachining of the valve seating surfaces, the following satisfactory leak rates were obtained:

, VALVE LEAKAGE (SCFH) TEST DATES 2-220-58A 8.795 February 8, 1975 2-220-62A 0 March 17, 1975 DESIGNATION OF APPARENT CAUSE OF OCCURRENCE:

Enutement Failurn - The excessive leakage of the "A" loop feedwater check valves was found to be the result of seals allowing flow between the seat ring and the valve body.

ANALYSIS OF OCCURRENCE:

If an accident were to occur, the 1-220-58A valve would determine'the maximum emount of leakage from the reactor. If an unlikely situation is assumed where the feedwater system experiences rupture between the 1-220-58A and 1-220-62A valves, check valve 58A would leak excessively if the reactor 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 this break occurred, it could happen either inside the drywell or the cutboard main steam isolation valve room. The ma!n contents of this leakage would be_feedwater, steam from the reactor steam space, and fission product gases. The water could Icak only until the vessel level was lowered to the feedwater sparger 8306290062 750306 PDR ADOCK 05000265

.m PDR _ _ _ _ _ . _ _ _ _ . _ _ _

o _

o level. Any steam that leaked would condense, and this water or condensate would be contained inside the plant by the floor dra,in system. Any gases that escaped would be. handled by the Standby Gas Treatment System. At no time would the fuel be in danger of bcIng uncovered due to this situation. When deinerted, a drywell entry could be made and the 1-220-57A valve could be closed preventing any further leakage. This type of worst case 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 not be affected.

CORRECTIVE ACTION:

The corrective action has been the installation of a modification (M-4-2-74-15) which replaced the stainless steci seal rings with Viton silicon rubber "0" rings.

The volve seating surfaces were also remachined.

FAILURE DATA:

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

G S

l l

i

_ , _ _ , , -- - . - _ - - - .