05000334/LER-1981-032-01, /01T-0:on 810502,vibrations in Feed Line a & High Feed Flow W/Rising Level in Steam Generator a Found During Moisture Carryover Test.Caused by Malfunction of Feed Regulating Valve a Due to Disconnected Actuator

/01T-0:on 810502,vibrations in Feed Line a & High Feed Flow W/Rising Level in Steam Generator a Found During Moisture Carryover Test.Caused by Malfunction of Feed Regulating Valve a Due to Disconnected Actuator

ML19347F647
Person / Time
Site:	Beaver Valley
Issue date:	05/15/1981
From:	Lacey W DUQUESNE LIGHT CO.
To:	NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
Shared Package
ML19347F646	List: 05000334/LER-1981-032, Forwards LER 81-032/01T-0 05000334/LER-1981-032-01, /01T-0:on 810502,vibrations in Feed Line a & High Feed Flow W/Rising Level in Steam Generator a Found During Moisture Carryover Test.Caused by Malfunction of Feed Regulating Valve a Due to Disconnected Actuator
References
LER-81-032-01T, LER-81-32-1T, NUDOCS 8105220304
Download: ML19347F647 (4)

LER-1981-032, /01T-0:on 810502,vibrations in Feed Line a & High Feed Flow W/Rising Level in Steam Generator a Found During Moisture Carryover Test.Caused by Malfunction of Feed Regulating Valve a Due to Disconnected Actuator

Event date:
Report date:
3341981032R01 - NRC Website

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4 EVENTDESCRIFT1CN AND PRC9AstE CONSEcuENCES h ITTU l With the reactor at 100% power and'a moisture carrvover test in progress, hich feed l vibrations in l

g l flow and rising level in the "A" steam generator were accompanied by 1

A load reduction was commenced orior to reactor trio at 83%

1o141 I the "A" feed line.

I l o i s i I poder due to hich-high level in the "A" steari generator which occurred af ter the Af ter the trip. excess steam pressure was relieved to the f

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Q l atmosphere, resulting in a calculated release concentration less than I

~ l o # a j l 10 CFR 20 limits for Na-24, the moisture carrvover tracer.

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{ g l The feedwater transient was due to the malfun. ion of the "A" feed regulating valve f

,,,,,l due to a disconnected actuator to positioner feed back linkage and a fractured valvel The valve stem was replaced, the valve was rebuilt, and actions recommended j

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Attachment To LER 81-32/OlT.

Beaver Valley Power Stat' on i

Duquesne Light Company Docket No. 50-334 At about 0041 hours4.74537e-4 days <br />0.0114 hours <br />6.779101e-5 weeks <br />1.56005e-5 months <br /> on May 2,1981, the Reactor Operator (RO) began to reduce Tavg from 576F to 571F as per procedure BVT 1.2 - 2.21.1, " Steam Generator Moistur.e Carryover Test".

After completing the reduction in Tavg, turbine. load was being increased to bring the reactor back up to an indicated 100% power for the test.

At approximately 0157 hours0.00182 days <br />0.0436 hours <br />2.595899e-4 weeks <br />5.97385e-5 months <br />, the Plant Operator (PO) acknowledged a feed flow /

stem flow mismatch alarm. He noted that the feed flow on the "A" steam generator had gone off scale high on [FR-MS-478] and that-the "A" steam generator had a rapidly increasing level. At about this time, the plant experienced a feedwater vibration. The PO took manual control of the "A" main feed regulating valve [FCV-FW-478] and attempted to reduce feed flow to restore proper level in the steam generator.

A power reduction was commenced at 2% per minute while attempting to stabilize "A" steam generato'r level, in an effort to transfer control from the main feed regulating valve to the "A" feed regulating bypvass valve [FCV-FW-479].

Approximately 1 or 2 minutes after the first vibration, the plant experienced a second feed water vibration while the PO was having problems controlling flow and level'to the "A" steam generator.

Operators were sent throughout the plant to determine what, if any, were the effects of the second feed water vibration. A third feedwater vibration was subsequently experienced and line' movement was observed by operators in the feed regulating valve room. This vibration was followed. shortly by a reactor trip, at 0208 hours0.00241 days <br />0.0578 hours <br />3.439153e-4 weeks <br />7.9144e-5 months <br />, on high-high level in the "A" steam generator. At the time of the trip, reactor power'had been reduced to'approximately 83%.

After the reactor trip, 'the Reactor Operator (RO) followed Emergency Operating Procedure E-5, " Reactor Trip".

The primary plant behavior was normal and normal pressurizer level and pressure was restored.

The auxiliary feed water pumps were used to restore normal steam generator level.

A main feed pump was then started in an attempt to feed the steam generators via the bypass feed control valves. After the auxiliary feed pumps were shut down, decreasing "A" and "C" steam generator levels and zero feed flow were observed.

Normal levels were restored using an auxiliary fead water pump.

Meter and Control Repairmen (MCRs) were sent to the f'eed bypass valves to determine the cause and repair the valves if possible. They discovered that the l

control air supply lines to the "A" and "C" valves [FCV-FW-479] 2nd [FCV-FW-499]

i had been broken off. The air lines were repaired but the "A" valve still would not work.due to s failed actuator pisten seal.

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4 Attichment To LER 81-32/01T Beaver Valley Power Station Duquesne Light Company Docket No. 50-334 Following the reactor trip, steam. pressure in the generators increased to an indicated value of approximately'1080 psig as observed on [PR-MS-474]. At this time, it was suspected that the atmospheric dump or safety valves opened.

This pressure increase above normal no-load pressure (1005 psig) was due to-the slow' response of the condenser steam dump system. About 20 minutes after the reactor trip, steam generator pressure again increased to approximately 1080 psig, possibly reopening the atmospheric dump / safety valves. While the l

pressure was increasing the third time, the PO switched to the Steam Pressure control mode. The dump system responded properly to' control steam generator pressure in this mode. An investigation revealed that the slow condenser steam-dump response was due to a failed summator in the condenser dump control loop.

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At 0330 hours0.00382 days <br />0.0917 hours <br />5.456349e-4 weeks <br />1.25565e-4 months <br />, The Shift Supervisor initiated the Emergency Preparedness Plan j

(EPP) and declared that an " Unusual Event" had occurred, under the understanding that a steam generator atmospheric dump / safety valve had opened, causing an unplanned release to the atmosphere of Na-24, a radioactive tracer used in the moisture carryover test. "The release concentration was calculated to be 1.0 x 10-13 pCi/cc and the EPP was terminated at 0420 hours0.00486 days <br />0.117 hours <br />6.944444e-4 weeks <br />1.5981e-4 months <br />. The health end safety of the general public was not jeopardized since the calculated release was less than.001% of 10 CFR 20 limits for Na-24.

- 2 No problems were experienced with primary plant systems during this event and i

there were no releases of radioactivity from the primary plant.

Damages believed to result from the feed water vibrations experienced were i

confined primarily to the "A" feed line. The only related damage to the "B"

feed line is the failure of flow transmitter [FT-FW-486]. The "C" feed line damage was confined to the severed instrument air line on the bypass valve, j

[FCV-FW-499].

"A" feed line damages were confined to snubber and pipe support damage inside containment, except for the broken instrument air line to the bypass valve, [FCV-FW-479]. The snubber damage involved a broken extension rod and cracked bushing. Remaining support damage was limited to loose bolts and shims shaken from a pipe whip restraint. All were repaired.

i The malfunction of tha "A" main feed regulating valve [FCV-FW-478] was the cause of the event. This malfunction'was caused by a disconnected feed back linkage to the valve positioner, which resulted in a full open positioner demand when a load increase initiated, and a fractured valve stem, which resulted in valve plug motion due.to hydraulic forces rather than actuator force..

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Attachment Ta LER 81-32/OlT Beaver Valley Power Station Duquesne Light Company Docket No. 50-334 Analysis of the valve ste'm by Westinghouse Electric Corporation concluded.that the valve stem was fractured for some time prior to the incident and that the stem fracture was due to fatigue loading - this analysis was conducted on a crack found approximately one inch from the point of fracture. It is believed that in the time period prior to the-incident the fractured valve stem was held together by the stem locking anti-rotation device installed on the valve stem at the fracture point.

Further analysis by ' Westinghouse and the valve vendor, Copes Vulcan, revealed that no similar failures had occurred previously in these valves as long as the valve trim was manufactured by the valve vendor.

It was noted that similar failure had occurred with valves whose trim was manufactured by someone other than the vendor.

It was the opinion of,,the valve vendor that stem failure was enhanced by increased lateral movement due to the fact that the stem locking anti-rotation device was not properly bolted to the valve actuator yoke. This device is intended for damping stem lateral moviment near the yoke.

Analysis by Westinghouse determined that no unreviewed safety question was involved since only a calculated value of 5000 lbf was exerted on the feed piping based on an 80% flow change in 1.5 seconds.

All three main feed regulating valves were disassembled, inspected, and rebuilt. The "A" and "C" line valve stems were replaced (the "C" valve stem

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was found to have a hairline crack). The stem locking anti-rotation devices were properly installed as instructed by the yalve vendor and all valve actuator to positioner linkage connections were sealed with Loc-tite.

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