05000334/LER-1986-001, :on 860210,120-volt Ac Bus III Deenergized, Resulting in Reactor/Turbine Trips & Feedwater Isolation. Caused by Failed Input Fuse.On 860211,reactor Tripped During Replacement of intermediate-range Detector
| ML20141N690 | |
| Person / Time | |
|---|---|
| Site: | Beaver Valley |
| Issue date: | 03/10/1986 |
| From: | Lacey W DUQUESNE LIGHT CO. |
| To: | NRC OFFICE OF ADMINISTRATION (ADM) |
| References | |
| LER-86-001, LER-86-1, ND1SS1:2728, NUDOCS 8603170525 | |
| Download: ML20141N690 (6) | |
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On 2/10/86, at 1926 hours0.0223 days <br />0.535 hours <br />0.00318 weeks <br />7.32843e-4 months <br />, 120 VAC Vital Bus III de-energized. The loss of the vital bus caused the 'A' Steam Generator controller to overfeed, resulting in a High-High Steam Generator (75%) turbine trip, reactor trip, and feedwater isolation at 1927 hours0.0223 days <br />0.535 hours <br />0.00319 weeks <br />7.332235e-4 months <br />. The plant was stabilized in Hot Standby by 1950 hours0.0226 days <br />0.542 hours <br />0.00322 weeks <br />7.41975e-4 months <br />. The cause of the vital bus loss was a failed input fuse to the inverter. The vendor believes the failure to have been caused by a misfiring silicon rectifier card exposed to high temperaturn. Following the trip, Source Range Detector N-31 failed and was declared inoperable. During replacement, the Intermediate Range Detector N-35 was placed on clearance.
While removing the instrument fuses, the operators failed to bypass the High Flux trip signal and the reactor tripped at 1727 hours0.02 days <br />0.48 hours <br />0.00286 weeks <br />6.571235e-4 months <br /> on 2/11/86. There were no safety implications from either trip since the reactor protection system functioned as designed.
Procedures will be changed to permit a more rapid change to vital bus alternate supply, and caution tags will be postod on flux detectors to remind operators of the trip logic. The reactor returned to criticality at 1430 hours0.0166 days <br />0.397 hours <br />0.00236 weeks <br />5.44115e-4 months <br /> on 2/12/86.
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-- 0lq1 0 l0 0l2 or 0 l4 rm n-w w =ae s mew nn on 2/10/86, the unit was operating at 100% reactor power. At 1926 hours0.0223 days <br />0.535 hours <br />0.00318 weeks <br />7.32843e-4 months <br />, numerous alarms annunciated, including abnormal steam generator level, steam flow and feedwater alarms, along with the Number 3 120 VAC Vital Bus Inverter trouble alarm. The operators quickly determined that the problem was a failure of the Number 3 Vital Bus. The steam generator level controllers were in the automatic mode before the loss of the vital bus, which powers the level signals that input to the Steam Generator Level Control System (SGWLC) for all three steam generators. The loss of the bus caused the 'A' Steam Generator controller to open its feedwater regulating valve fully, in response to the failure of the level signal. The 'B' Steam Generator controller failed to manual control due to a loss of power to its automatic controller, and its feedwater regulating valve remained in the same position. The
'C' Steam Generator remained in automatic while its manual control failed, as it was powered by Vital Bus Number 3.
An investigation will be conducted during an upcoming refueling outage to determine whether the steam generator level controllers are functioning correctly. The plant operator, in accordance with a special operating order, immediately took manual control of the SGWLC and attempted to close the valves. However, before his action could take full effect and overcome the large open demand on the 'A' Steam Generator feedwater valve, the 'A' Steam Generator level reached the High-High setpoint of 75%
narrow range level, causing a turbine trip. The signal also generated a feedwater isolation which tripped the main feedpumps and closed all required feedwater valves. Since reactor power was above the P-9 permissive setpoint of 10%, the turbine caused a reactor trip at 1927 hours0.0223 days <br />0.535 hours <br />0.00319 weeks <br />7.332235e-4 months <br />. There were no safety implications resulting from the trip because the Reactor Protection System functioned properly to shut down the reactor, while the operators quickly took action to stabilize the plant in Hot Standby. They accomplished this task by performing the immediate manual actions for a reactor trip, and following procedures E-0, " Reactor Trip," and ES 0.1, " Response to Reactor Trip."
Abnormal Operating Procedure (AOP) 33, " Loss of Vital Bus III," was also followed. The bus was transferred to an alternate power supply at 1930 hours0.0223 days <br />0.536 hours <br />0.00319 weeks <br />7.34365e-4 months <br />.
As the recovery progressed, operators securing the turbine plant noticed the relief valves lifting on the feedwater heaters. This relief was caused by the feedwater isolation and condensate recirculation, which led to a decrease in condensate cooling flow to the feedwater heaters.
In addition, the vital bus loss caused a letdown isolation, while the post trip cooldcun resulted in increased charging flow. Consequently, the Volume Control Tank (VCT) level dropped, and the suction of the charging pumps switched to the Refueling Water Storage Tank (RWST). Approximately one minute of erratic charging flow, pressure, and seal injection flow occurred at this point.
It is judged entrained gas was rdsponsible for thene problems. At 1946 hours0.0225 days <br />0.541 hours <br />0.00322 weeks <br />7.40453e-4 months <br />, the charging suction was returned to the VCT and the RWST was isolated.
The plant was stabilized in Hot Standby at 1950 hours0.0226 days <br />0.542 hours <br />0.00322 weeks <br />7.41975e-4 months <br />. Also at 1950 hours0.0226 days <br />0.542 hours <br />0.00322 weeks <br />7.41975e-4 months <br />, the NRC was notified of the event in accordance with the provisions of 10 CFR 50.72 (b)(2)(ii).
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ac r assm on Several equipment failures were discovered following the trip. The first involved the Neutron Flux Source Range Detector N-31, menufactured by Westinghouse. At 1952 hours0.0226 days <br />0.542 hours <br />0.00323 weeks <br />7.42736e-4 months <br />, this detector started giving erratic indications and low power supply voltage readings. A0P-10. " Malfunction of Nuclear 1
Instrumentation," was followed by the operators and the detectar fuses were pulled at 2005 hours0.0232 days <br />0.557 hours <br />0.00332 weeks <br />7.629025e-4 months <br />. The detector was declared inoperable at this time.
Troubleshooting by the Station Instrument and Control Section determined that the failure was caused by a faulty connector on the high voltage power supply. The detector was replaced and a surveillance test performed successfully on 2/12/86.
In order to replace the Source Range Detector, it was necessary to de-energize the Intermediate Range Neutron Flux Detector N-35, which is located above N-31 in a common detector well. While removing the instrument power fuses for N-35 on 2/11/86 control cabinet mounted in the Control Room, the operators failed to bypass the high flux reactor trip signal. The removal of the fuses satisfied the 1/2 logic for an Intermediate Range High Flux trip. The bistables changed state and initiated a reactor trip at 1727 hours0.02 days <br />0.48 hours <br />0.00286 weeks <br />6.571235e-4 months <br /> on 2/11/86. The NRC was notified of the trip in accordance with 10 CFR 50.72(b)(ii) at 1750 hours0.0203 days <br />0.486 hours <br />0.00289 weeks <br />6.65875e-4 months <br />. The plant was in Mode 3 at the time of the trip
'i with the control rods inserted; therefore, the trip caused only the shutdown banks to drop. There were no safety implications to the public because no actual high neutron flux was present and the plant remained in a stable condition. To prevent a recurrence of this event, permanent caution signs will be installed on the Control Room detector cabinets to alert the operators to the reactor protection implication of removing the Instrument and Control power fuses. In addition, the event is being discussed at Operations retraining modules.
Another equipment failure resulting from the trip of 2/10/86 involved the 'A' l
Quench Spray Pump Chemical Addition Tank temperature indicator. This indicator was noticed to be failed at zero, when Vital Bus No. 3 was swapped to its alternate supply at 1938 hours0.0224 days <br />0.538 hours <br />0.0032 weeks <br />7.37409e-4 months <br />. Troubleshooting of this instrument was performed, and the problem was determined to be a failed power amplifier. The amplifier was repaired and the transmitter loop recalibrated. A cailorine detector that tripped on the vital bus failure was reset satisfactorily.
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The cause of the vital bus failure was a blown input fuse to the Number 3 120 VAC inverter (manufactured by Cyberex Inc.).
It has been determined by Station Instrument and Control that the inverter fuse failed due to misfiring of the Silicon Control Rectifiers. According to the vendor, a number of circuit cards had been exposed to high temperatures. Mounted on the circuit cards are temperature sensors designed to melt and indicate at 150*F.
The sensor on the gate drive control module card (Cyberex 9549-51) was found to be pyg,0..
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,ancr assamm indicating and some solder joints were dull. The sensors on several other cards were darkened. According to the vendor, the inverter cabinet should be maintained below 55'C (131*F) to minimize any high temperature effects on the circuit boards. The ambient temperature in the switchgear area was 72*F; therefore, it is not considered to have been a factor in the problem at the time of the trip. Lessons learned from this event indicate that in order to prevent the vital bus loss from leading to a control problem, a more rapid transfer of vital bus power to the auxiliary supply is necessary. The Abnomal Operating Procedures are being revised to reflect these lessons.
The reactor was returned to criticality at 1430 hours0.0166 days <br />0.397 hours <br />0.00236 weeks <br />5.44115e-4 months <br /> on 2/12/86.
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'Af Telephorie (412) 393-6000 Nuclear Group P.O. Box 4 Shippingport, PA 150774004 March 10, 1986 NDISSl:2728 Beaver Valley Power Station, Unit No. 1 Docket No. 50-334, License No. DPR-66 LER 86-001-00 United States Nuclear Regulatory Commission Document Control Desk Washington, DC 20555 Gentlemen:
In accordance with Appendix A, Beaver Valley Technical Specifications, The following Licensee Event Report is submitted:
LER 86-001-00,10 CFR 50.73.a.2.iv, " Automatic Actuation of Reactor Protection System (RPS) and Engineered Safety Feature (ESF)."
Very truly yours, k
Wm.
. Lacey Plant Manager Attachment i
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.e March 10, 1986 NDISS1:2728
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cc: Dr. Thomas E. Murley Regional Administrator-
' United States Nuclear Regulatory Commission Region 1 King of Prussia, PA 19406 C. A. Roteck, Ohio Edison
.Mr. Peter Tam, BVPS Licensing Project Manager United States Nuclear Regulatory Commission Washington, DC 20555 W. Troskoski, Nuclear Regulatory Commision, BVPS Site Inspector Mr. Alex Timme, CAPCO Nuclear Projects Coordinator, Toledo Edison INPO Records Center Suite 1500 1100 Circle 75 Parkway Atlanta, GA 30339 G. E. Muckle, Factory Mutual Engineering, Pittsburgh Mr. J. A. Triggiani, Operating Plant Projects Manager Mid Atlantic Area Westinghouse Electric Corporation Nuclear Services Integration Division Box 2728 Pittsburgh, PA 15230 American Nuclear Insurers e/o Dottie Sherman, ANI Library The Exchange Suite 245 270 Farmington Avenue Farmington, CT 06032 f
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