05000334/LER-2001-003
| Beaver Valley Power Station Unit No. 1 | |
| Event date: | 11-06-2001 |
|---|---|
| Report date: | 12-21-2001 |
| Reporting criterion: | 10 CFR 50.73(a)(2)(iv), System Actuation |
| 3342001003R00 - NRC Website | |
There were no systems, structures, or components that were inoperable that contributed to the event. One system air compressor was out of service at the time.
DESCRIPTION OF EVENT
At 1416 hours0.0164 days <br />0.393 hours <br />0.00234 weeks <br />5.38788e-4 months <br /> on November 6, 2001, Beaver Valley Power Station Unit No. 1 reactor automatically tripped from 100% power due to low-low level in the C Steam Generator. The Feedwater Level Control Valve on the C Steam Generator unexpectedly failed closed causing the C Steam Generator water level to rapidly drop and reach the low-low water level reactor trip setpoint. Prior to the reactor trip, the Control Board Operator attempted to open the Feedwater Level Control Valve in manual control and observed no increase in the demand signal indicator or feedwater flow. After the reactor trip, Emergency Operating Procedure E-0 for Reactor Trip was performed and the plant was stabilized in Mode 3.
REPORTABILITY
The automatic initiation of a reactor trip from 100% power via the Reactor Protection System on November 6, 2001, was a valid reactor trip and was not part of a pre-planned sequence during testing or reactor operation. Therefore this event is reportable pursuant to 10 CFR 50.72(b)(2)(iv)(B) pursuant to 10 CFR 50.72 (b)(2)(iv)(B) at 1538 hours0.0178 days <br />0.427 hours <br />0.00254 weeks <br />5.85209e-4 months <br /> on November 6, 2001 (ENS No. 38472).
CAUSE OF EVENT
The direct cause of the reactor trip was due to low-low level in the C Steam Generator, which was preceded by the unexpected closure of the C Main Feedwater Regulating Valve. Closure of the feedwater regulating valve resulted in a loss of feedwater flow to the C Steam Generator and the rapid decrease in steam generator water level.
The cause of the unexpected closure of the C Main Feedwater Regulating Valve was the failure of a diode in the 7100 process rack module that controls the valve actuator, causing the output of the Power Supply regulator board to rise from 35 volts to 51.55 volts. This higher voltage resulted in the output of the valve limiter to not rise above 3.8 milliamps. The control range is 4 to 20 milliamps.
Since this limiter is common to both the manual and automatic control circuits and with the output U.S. NUCLEAR REGULATORY COMMISSION r � _, --.NIRC FORM 366A (1 -2001) FACILITY NAME (1) LER NUMBER (6) DOCKET (2) PAGE (3) limited to the 3.8 millamps value, the actuator would not respond to either Manual or Automatic control commands.
There was no evidence that voltage spikes or overheating were involved. In addition, the controller was replaced approximately one and one half years ago; therefore, age was not considered a factor.
Thus, the diode failure was categorized as a random failure.
SAFETY IMPLICATIONS
Following the automatic reactor trip, the control rods fully inserted into the reactor core and the required safety systems operated as designed. Emergency Operating Procedure E-0 for Reactor Trip was performed and the plant was stabilized in Mode 3.
This event was an automatic reactor trip from full power, which is less severe than the design basis event for Loss of External Electrical Load and/or Turbine Trip from full power. The Loss of External Electrical Load and/or Turbine Trip is analyzed in BVPS Unit No. 1 UFSAR Section 14.1.7.
Comparison of the UFSAR Loss of External Electrical Load and/or Turbine Trip with this event indicates that BVPS Unit 1 operated conservatively regarding the UFSAR transient in comparison to the UFSAR analysis assumptions/results. The actual plant response for this event was bounded by the UFSAR analysis for a Loss of External Electrical Load and/or Turbine Trip.
The plant risk associated with this BVPS Unit 1 automatic reactor trip on November 6, 2001 due to low steam generator water level is considered to be low. This is based on the conditional core damage probability for the event when considering the trip occurred during plant operation in Mode 1 at 100% reactor power given a partial loss of main feedwater with one system air compressor out of service.
Based on the above, the safety significance of the automatic reactor trip on November 6, 2001 was small.
CORRECTIVE ACTIONS
1. The deficiency in the Main Feedwater Regulating Valve FC-1FW-498 controller was corrected.
2. A failure analysis will be performed on the failed controller components for FC-1FW-498.
3. Additional enhancements are being investigated for 7100 module components, including power supplies and controllers.
Corrective action completion is being tracked through the corrective action program.
U.S. NUCLEAR REGULATORY COMMISSION -NRC FORM 366A (1-2001) FACILITY NAME (1) LER NUMBER (6) DOCKET (2) PAGE (3) U.S. NUCLEAR REGULATORY COMMISSION
PREVIOUS SIMILAR EVENTS
A review of past Beaver Valley Power Station Units 1 and 2 Licensee Event Reports found two similar events involving a reactor trip associated with steam generator water level within the last five years:
- BVPS Unit 1 LER 98-028, "Automatic Reactor Trip on 'A' Steam Generator Low Level Coincident with Steam Flow/Feed Flow Mismatch Signal from Manual Tripped Transmitter Bistables of F-MS- 475.
- BVPS Unit 1 LER 97-025, "Ground in Feedwater Flow Controller Results in High Steam Generator Level and Subsequent Turbine Trip/Reactor Trip.