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Home NRC Library Document Collections Events, Reports Associated with Part 21 Reports 1997

Part 21 Report - 1997-331

 
 


ACCESSION #: 9705300003



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ATTACHMENT "EVENT NOTIFICATION WORKSHEET" OMITTED.



        Attachment to NRC Form 361 - Event Notification Worksheet

  Initial Notification 10CFR Part 21 - BVPS Unit 2 Auxiliary Feedwater

Check Valve Failure



During the March 19, 1997 Beaver Valley Power Station (BVPS) Unit 2 trip

(previously documented in Licensee Event Report 1-97-005-00, dated April

14, 1997), Auxiliary Feedwater (AFW) anomalies were observed.  The AFW

flow through the "B" steam generator was lower (150 vs. 280 GPM) than

expected.  Flow through the "A" and "C" steam generators was as expected.

The performance of all three AFW Pumps was normal for the trip

conditions.  Subsequent inspection of the "B" steam generator check valve

(2FWE-100) revealed that the seat ring had partially moved into the flow

stream, decreasing the available opening for flow to pass through the

valve.  The three Unit 2 AFW check valves were shipped to the vendor's

facility for further examination and analysis.  The resulting

investigation concluded that the thermal gradient conditions created by

flowing cold water through the hot valve created a rapid cooldown of the

seat ring, allowing it to displace.  All three of the subject check

valves were modified to prevent reoccurrence.  The valves were then

shipped back to the site and reinstalled.



The check valves arc normally held shut by steam generator pressure.

Failed check valve 2FWE-100, is located in close proximity to the main

feedwater header, and is at approximately 430 degrees F.  The other two

check valves are below 300 degrees F.  The differences in the

temperatures are attributed to the distance and location of the valves

with respect to the main feedwater header.  During a reactor trip, AFW at

approximately 60 degrees F is injected.  It is estimated that it takes

approximately 5 seconds for the seat to cool down, whereas the massive

valve body stays relatively hot.  It appears that the valve seat loosened

because of cold water passing through the valve.  The massive valve

retained its shape, whereas the seat shrunk.  This relative shrinkage

allowed the seat to displace and move into the flow stream.



An extent of condition evaluation has shown that other Enertech nozzle

check valves of this design in service at Unit 2 are not subject to

thermal gradients of sufficient magnitude to induce the condition

observed for 2FWE-100.  Unit 1 does not have Enertech nozzle check

valves.



A similar failure of AFW check valve 2FWE-100 would have resulted in a

reduction of AFW flow to the "B" steam generator during a postulated

design basis accident.  The reduction in flow caused by the defect would

have resulted in AFW flows less than analyzed for the Unit 2 Accident

Analysis.  Therefore, for the postulated accidents, the ability to

provide adequate AFW cooling would be adversely affected and the system

may not.  have performed its safety function.



An evaluation of this event, completed on April 24, 1997, has determined

that a substantial safety hazard could be created as the result of the

identified valve defect and that it is, therefore, reportable pursuant to

the requirements or 10CFR Part 21.



Component Description:



The component is a nozzle check valve intended for use with water

service.



Supplier:



Enertech

(BW/IP)

2950 Birch Street

Brea, CA 92621



Type:



Enertech "4" Nozzle Check Valve, ANSI Class 600, Type DRV-Z



Valve Body - Dwg, # PD96227, ASME SA105

Seat - Dwg.  # PB96233, ASTM A479 Type 316



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