Part 21 Report - 1997-331

ACCESSION #: 9705300003 NOTE: This text document was processed from a scanned version or an electronic submittal and has been processed as received. Some tables, figures, strikeouts, redlines, and enclosures may not have been included with this submittal, or have been omitted due to ASCII text conversion limitations. In order to view this document in its entirety, you may wish to use the NUDOCS microfiche in addition to the electronic text. 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 *** END OF DOCUMENT ***

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