ML20084K305
| ML20084K305 | |
| Person / Time | |
|---|---|
| Site: | Quad Cities  |
| Issue date: | 07/18/1974 |
| From: | COMMONWEALTH EDISON CO. |
| To: | US ATOMIC ENERGY COMMISSION (AEC) |
| Shared Package | |
| ML20084K292 | List: |
| References | |
| AO-50-254-74-8E, NUDOCS 8305190267 | |
| Download: ML20084K305 (2) | |
Text
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REPORT NUMBER: A0 254/74,8e REPORT DATE: July 18, 1974 OCCURRENCE DATE: April 25, 1974 FAClllTY: Quad-Cities Nuclear Power Station Cordova, Illinois 61242 IDENTIFICATION OF OCCURRENCE:
Feedwater check valves 1-220-58A, 1-220-588, 1-220-62A, and 1-220-628 excessive leakage.
CONDITIONS PRIOR TO OCCURRENCE:
Reactor in cold shutdown condition for refueling outage, reactor feed piping isolated and drained.
DESCRIPTION OF OCCURRENCE:
While performing local leak rate tests on primary isolation valves it was deter-mined that the 1-220-58A,1-220-588,1-220-62A, and 1-220-62B all leaked or were suspected of leaking in excess of the 18.36 SCFH limit allowed by Technical Speci-l fication 4.7.A.2.1.(2)(b).
The test conditions and results were as follows:
TEST TEST VALVES VALVE PRESSURE (PSIG)
FOR PRESSURIZING LEAKAGE (SCFH) 1 -220-58A 48 1-220-Il5A 646.5 l-220-Il6A i
1-220-58B 48 1-220-1158 888.2
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l-220-1168 1 -220-62B 48 1-220-868 1332.0 1-220-878 1 -220-62A 48 l -220-86A Couldn't test because 1 -220-87A using clean-up system but expected leakage.
Repairs were initiated and a subsequent leak test on each valve indicated satis-factory results as follows:
VALVE LEAKAGE (SCFH) 1-220-58A 1.49 l-220-58B 2.99 1-220-628 11.10 1
1 -220-62A 0.0 8305190267 y % $4 PDR ADOCK PDR i
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p A0 254/74-Be
. U July 18,-1974 DESIGNATION OF APPARENT CAUSE OF OCCURRENCE:
The apparent cause of excessive leakage through the check valves is leakage between the seat ring and the valve body.
ANALYSIS OF OCCURRENCE:
If 'n accident occurred, the 1-220-58A and 1-220-58B valves would determine the max nam amount of leakage from the reactor, if an unlikely situation is assumed where the feedwater system experiences ruptures between both the 1-220-5BA and 1-220-62A valves and the 1-220-588 and 1-220-628 valves the '58 checks would leak 1534.7 SCFH if the reactor was at 48 psi. Since.the reactor would be at a pres-sure on the order of 950 psi these valves would be driven into.their seats allow-Ing for a better seal and reducing leakage. Af ter the reactor pressure was re-duced below 300 psi the leakage wouM' approach the 1534.7 SCFH figure more closely.
If this break occurred it could happen either inside the drywell or inside the out-board main steam isolation valve room. The main contents of.this leakage 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 feedwater sparger level. At this point feedwater leakage would stop. Any steam that leaked would condense because the local temperatures would be below boiling conditions. This water or condensate could be handled and contained inside the plant by the floor drain system. Any gases that escaped would be handled by the Stand-By Gas Treat-ment system. At no time would the fuel be in danger of being uncovered due to -
this situation. When deinerted, a drywell entry could be made and the 1-220-57A and 1-220-57B valves could be closed to stop any leakage. This type of worst-case analysis puts the plant in an inoperable status but the leakage would have
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m inimal co nsequences from the standpoint of public health and safety.
CORRECTIVE ACTION:
The two main corrective actions were the installation of a modification (M-4 74-15) which replaces.the steel Real ring with the Viton rubber "0" ring and the relapping of the valve' seating surfaces.
This same modification will be installed in Unit Two if these valves are found to leak excessively when tested.
FAILURE DATA:
This is the first time these valves have been tested since the pre-operational tests.
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