ML20006E237

LER 90-001-00:on 900123,reactor Trip Occurred on Steam Feedwater Flow Mismatch.Caused by Failed Circuit Driver Card on Feedwater Regulating Valve.Feedwater Regulating Valve Driver Card replaced.W/900208 Ltr

ML20006E237
Person / Time
Site:	North Anna
Issue date:	02/08/1990
From:	Kane G VIRGINIA POWER (VIRGINIA ELECTRIC & POWER CO.)
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
LER-90-001, LER-90-1, N-90-001, N-90-1, NUDOCS 9002220365
Download: ML20006E237 (7)

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. 1 10 CFR 50.73 i VIRGINl A ELECT MlC AND POWER COMP ANY NORTN ANN A P0t ER $T ATION P.C.90K 402 MINE R AL, VIRGINI A 23117 February 8, 1990 l U. S. Nuclear Regulatory Commission Serial No. N 90-001 )

Attention: Document Control Desk NAPS /CSW:csw ,

Washington, D.C. 20555 Docket No. 50 338 i

License No. NPF4 i

Dear Sirs:

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The Virginia Electric and Power Company hereby submits the following Licensee Event Report applicable to Nonh Anna Unit 1.

Report No. LER 90-01-00 This Repon has been reviewed by the Station Nuclear Safety and Operating Committee and will be forwarded to Safety Evaluation and Control for their review. ,

Very Truly Yours, l G. i. Kane Station Manager 1-Enclosum:

l cc: U.S. Nuclear Regulatory Commission 101 Marietta Street, N.W.

Suite 2900 Atlanta, Georgia 30323 Mr. J. L. Caldwell NRC Senior Resident Inspector North Anna Power Station

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The closure was caused by a failed printed circuit driver card in the valve controller. After event investigation and corrective action, Unit 1 was returned to critical on January 24, 1990 at 0241 hours0.00279 days <br />0.0669 hours <br />3.984788e-4 weeks <br />9.17005e-5 months <br />.

This event constitutes an automatic actuation of the Reactor Protection System and is reportable pursuant to 10 CFR 50.73 (a) (2) (iv) .

No significant safety consequences resulted from the reactor trip becaune plant safety systems functioned as designed. The Reactor Coolant System parameters stabilized at their normal post trip values. There was no release of radioactive materials due to the trip.

The health and safety of the public were not affected at any time  !

during this event.

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) 0 0l2 op 0 l6 l rsssu,, , . m .m. a c,n ama m, 1.0 Description of the Event At 1522 hours0.0176 days <br />0.423 hours <br />0.00252 weeks <br />5.79121e-4 months <br /> on January 23, 1990, Unit 1 experienced an automatic trip from 100 percent power. The initiating signal for the reactor trip was a low level in the "C" Steam Generator (S/G) (EIIS System Identifier AB, Component Identifier SG) with a steam flow greater than feedwater flow i mismatch. The mismatch resulted from closure of the 'C' Main Feedwater Regulating Valve . The closure was caused by a f ailed Westinghouse 7300 printed circuit driver card in the valve's electronic control system. This event constitutes an automatic actuation of the Reactor Protection a System and is reportable pursuant to 10 CFR 50.73 (a) (2) (iv) . A four hour report was made to the NRC at 1725 hours0.02 days <br />0.479 hours <br />0.00285 weeks <br />6.563625e-4 months <br /> on January 23, 1990, in accordance with 10 CFR 50.72 (b) (2) (ii) .

Control Room Operators responded to the reactor trip in accordance with Emergency Operating Procedure EP-0, " Reactor i Trip or Safety Injection". The plant reponded as expected ,

with Pressurizer pressure decreasing to 1950 psig, pressurizer level decreasing to 21% and Reactor Coolant System (RCS) temperature decreasing to 54 3. 5* F. Following evaluation of the RCS parameters and indications, Control Room personnel transitioned from EP-0 to ES-0.1 " Reactor Trip Response".

Subsequent to the reactor trip, the "C" phase of the electrical generator output breaker (G-12) control logic indicated closed due to a dirty contact, although the breaker had actually opened. As a result, the switchyard breakers G102 and G175H opened as designed to isolate the turbine generator from the electrical grid. Station service electrical buses automatically transferred to the reserve station service transformers. At 1525 hours0.0177 days <br />0.424 hours <br />0.00252 weeks <br />5.802625e-4 months <br />, Control Room Operators entered Abnormal procedure AP-10.1, " Loss of Electrical Power" to identify the electrical malfunction and verify electrical system lineups.

While monitoring intermediate range neutron flux, Control Room Operators observed that intermediate range detector N-35 was undercompensated, preventing automatic reinstatement of the source range detectors. At 1538 hours0.0178 days <br />0.427 hours <br />0.00254 weeks <br />5.85209e-4 months <br />, the operators manually reinstated the source range detectors in accordance with ES-0.1.

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Plant Equipment responded as expected with the ,

following exceptions:

• TheC' Main Feedwater Regulating Valve, 1-FW-FCV-1498 (EIIS System Identifier SJ, Concoonent Identifier FCV,

• Vendor Identifier C635, Model Nder D100-12), failed ,

closed due to a failed printed circuit driver card.

• The Intermediate Range Nuclear Instrument (N-35) (EIIS  ;

System Identifier IG, Component Identifier DET, Vendor Identifier W120) was undercompensated.

• The Condenser Steam Dump 'B' (1-MS-TCV-1408B) (EIIS System Identifier SB, Component Identifier TCV Vendor Identifier C635 Model Number 8-RA36RG) indicated mid position after closure.

• The 1 Reheat Right Stop Valve (EIIS System Identifier SB, Component Identifier ISV, Vendor Identifier C635, Model Number D100-160-3) had a broken indicator arm.

• The 'C' Phase Generator Output Breaker (G-12) (EIIS System Identifier 41, Component Identifier BKR, Vendor Identifier B455) failed to indicate open.

• The tube side relief valves on the 2A and 4B feedwater heaters (1-RV-SV-112A and 114B) (EIIS System Identifier SJ, Component Identifier RV, Vendor Identifier C710) opened.

The suction relief valve (1-FW-RV-102C) on the 'C' Main Feedwater Pump (EIIS System Identifier SJ, Component Identifier RV, Vendor Identifier S012, Model Number 451132-B) opened.

After event investigation and corrective action, Unit I was returned to critical on January 24, 1990 at 0241 hours0.00279 days <br />0.0669 hours <br />3.984788e-4 weeks <br />9.17005e-5 months <br />.

2.0 Significant Safety Consequences and Implications No significant safety consequences resulted from t.he reactor trip because plant safety systems functioned as I designed. The Reactor Coolant System parameters stabilized at their normal post trip values. There was no release of radioactive materials due to the trip.

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3.0 Cause of the Event The cause of this event was failure of the printed circuit driver card on the 'C' Main Feedwater Regulating valve which caused valve closure. Preliminary investigation has indicated that the driver card failed due to age. VEPCO .

is conducting an engineering evaluation and root cause investigation to discern any contributing factors to the  ;~

failure and determine if other driver cards of similar vintage have an increased failure risk.

Upon completion of the engineering and root cause ,

evaluation, necessary corrective actions will be ,

implemented.

4.0 Immediate Corrective Action As an immediate corrective action, Emergency Operating Procedure EP-0, " Reactor Trip or Safety Injection", was entered and the plant stabili..ed in Hot Standby.

5.0 Additional Corrective Actions The following corrective actions were taken to correct the hardware problems that occurred during this event:

• The 'C' Main Feedwater Regulating Valve, 1-FW-FCV-1498 failed driver card was replaced. A functional test was performed on the similar driver cards for the 'A' and

'B' Main Feedwater Regulating Valves. The' cards and associated drivers were verified to be operating properly.

• The Intermediate Range Nuclear Instrument (N-35) compensation voltage was adjusted in accordance with Westinghouse methodology.

• The Condenser Steam Dump 'B' (1-MS-TCV-1408B) limit switch was adjusted.

• The Broken indicator arm on 1 Reheat Right Stop Valve was repaired.

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• The Generator Output Breaker (G-12) phase "C"  ;

indication contact was cleaned. As a preventive i measure, the Generator Output Breaker phase "A" and "B" l indication contacts were also cleaned.

• The Tube side relief valves on the 2A and 4B feedwater  !

heaters (1-RV-SV-112A and 114P) reseated, l I

• The Suction relief valve (1-FW-RV-102C) on 'C' Main l Feedwater Pump resented. l 6.0 Actions to Prevent Recurrence Engineering evaluations will be conducted on the the following malfunctions:

• Intermediate Range Monitor N-35 undercompensation

• Failed driver card on 'C' Main Feedwater Regulating Valve, 1-FW-FCV-1498 .

• Lifting of the tube side relief valves on the 2A and 4B teedwater heaters (1-RV-SV-112A and 114B) .

• Lifting of the suction relief valve (1-FW-RV-102 C ) on the 'C' Main Feedwater Pump.

Corrective actions will be taken as required as a result of these evaluations.

7.0 Similar Events  !

Previous reactor trips due to steam flow greater than feedwater flow mismatch coincident with a low steam generator level occurred on Unit 1: May 20, 1986 (LER-N1 008), August 6, 1988 (LER N1-88-020) , February 25, 1989 (LER N1-89-005) and on Unit 2: March 13, 1984 (LER N2-84-001).

June 25, 1984, (LER N2-84-005) and June 29, 1986 (LER N2 009)

The Reactor Trip reported in LER N1-89-005 was also caused by the failure of the 'C' Main Feedwater Regulating Valve. However, the cause of the regulating valve failure for that event was a broken air line to the valve, wac r =4 i.4ei

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