05000339/LER-2003-001
| Document Number | |
| Event date: | 03-31-2003 |
|---|---|
| Report date: | 04-17-2003 |
| 3392003001R00 - NRC Website | |
FACILITY NAME (1) NORTH ANNA POWER STATION UNIT 2
DOCKET
05000 - 339 1 YEAR I SEQUENTIAL � REVISION 2003 -001 - � 00 LER NUMBER 6) 1.0 DESCRIPTION OF THE EVENT On March 31, 2003, at 1259 hours0.0146 days <br />0.35 hours <br />0.00208 weeks <br />4.790495e-4 months <br /> with Unit 2 operating at 100 percent power an automatic reactor trip occurred. The initiating signal was the "C" steam generator (SG) (EIIS System AB, Component SG) low level coincident with a steam flow greater than feed flow mismatch caused by closure of the "C" main feed regulating valve (MFRV) (EIIS System SJ, Component FCV). This resulted in a reactor and turbine trip.
At 1259 hours0.0146 days <br />0.35 hours <br />0.00208 weeks <br />4.790495e-4 months <br /> the Unit 2 operator at the controls received a process rack power supply failure alarm followed by a "C" steam generator steam flow-feed flow mismatch alarm.
The operator noted an absence of indicating lights for the "C" feed reg. valve controller and that the valve indicated closed. The reactor automatically tripped from a steam flow- feed flow mismatch with low steam generator level initiating signal within seconds of the receipt of the alarms. Control Room personnel responded to the event in accordance with emergency procedure E-0, Reactor Trip or Safety Injection. Control Room personnel stabilized the plant using ES-0.1 Reactor Trip recovery. Initially, Reactor Coolant System (RCS) (EIIS System AB) pressure and temperature decreased to approximately 1983 psig and 544 degrees Fahrenheit. Subsequently, RCS pressure and temperature returned to their normal programmed values.
Following the reactor trip the Reactor Protection System (RPS) and all Engineered Safety Feature (ESF) (EIIS System JE) equipment responded as designed including proper operation of AMSAC, and the Auxiliary Feedwater System (AFW) (EIIS System BA). No other major equipment issues were noted.
At 1540 hours0.0178 days <br />0.428 hours <br />0.00255 weeks <br />5.8597e-4 months <br /> a 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> Non-Emergency Report was made to the NRC in accordance with 10 CFR 50.72 (b)(2)(iv)(B) for an event causing actuation of the Reactor Protection System when the reactor is critical. An 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> Non-Emergency Report was also made to the NRC in accordance with 10 CFR 50.72 (b)(3)(iv)(A) for an event causing actuation of the Auxiliary Feedwater System.
2.0 SIGNIFICANT SAFETY CONSEQUENCES AND IMPLICATIONS This event posed no significant safety implications because the reactor protection system and ESF systems functioned as designed following the reactor trip. Therefore, the health and safety of the public were not affected by this event.
This event is reportable pursuant to 10 CFR 50.73 (a)(2)(iv) for a condition that resulted in an automatic actuation of any engineered safety feature including the reactor protection system.
FACILITY NAME (1) NORTH ANNA POWER STATION UNIT 2 LER NUMBER (6)
DOCKET
05000 - 339 3.0 CAUSE Cause of the automatic reactor trip was the "C" SG low level coincident with a steam flow greater than feed flow mismatch. The initiating signal was caused by closure of the "C" MFRV. Closure of the "C" MFRV was the result of a failed driver card in the SG water level control system for "C" SG. The driver card failed (i.e., de-energized) as a result of a blown fuse.
4.0 IMMEDIATE CORRECTIVE ACTION(S) Control Room personnel responded to the event in accordance with emergency procedure E-0, Reactor Trip or Safety Injection. Control Room personnel stabilized the plant using ES-0.1 Reactor Trip Recovery. All safety systems responded appropriately. The unit was stabilized at no-load conditions, the MFW System was placed in service to all three S/Gs and the AFW System secured and returned to normal AUTO/Standby alignment.
Subsequently, Control Room personnel transitioned to 2-0P-1.5 in preparation for unit re- start.
5.0 ADDITIONAL CORRECTIVE ACTIONS The "C" MFRV failed driver card was replaced with a refurbished card and a successful functional test was performed. The driver cards for the Unit 2 "A" and "B" MFRVs were inspected and the fuses were replaced. Functional tests were performed successfully. Unit 2 entered Mode 1 at 0256 hours0.00296 days <br />0.0711 hours <br />4.232804e-4 weeks <br />9.7408e-5 months <br /> on April 1, 2003. Unit 2 achieved 100 percent power at 0122 hours0.00141 days <br />0.0339 hours <br />2.017196e-4 weeks <br />4.6421e-5 months <br /> on April 2, 2003.
With Unit 1 defueled for a scheduled refueling outage the driver cards for the "A", "B", and "C" MFRVs and the by-pass valves were pulled, the fuses inspected and replaced as necessary.
6.0 ACTIONS TO PREVENT RECURRENCE Unit 2 last experienced an automatic reactor trip from a MFRV driver card failure in 1991.
Corrective actions implemented included developing preventive maintenance (PM) MFRV driver cards were last replaced in 1999. The next scheduled driver card replacement was due during the 2004 refueling outage. Following the 1999 driver card replacement, industry operating experience identified the need to inspect and replace fuses as part of the 7300 processor card refurbishment. Subsequently, the PM was revised to include this action in the next schedule PM. The current controls ensure driver card refurbishment's included proper inspections and necessary repairs.
FACILITY NAME (1) NORTH ANNA POWER STATION UNIT 2
DOCKET
05000 - 339 LER NUMBER (6) I NUMBER � NUMBER A root cause evaluation is being performed regarding the automatic reactor trip. Corrective actions will be performed as necessary following completion of the evaluation.
7.0 SIMILAR EVENTS LER N2-91-009-00 dated 10/10/91, documents an automatic reactor trip from "B" steam generator low level coincident with a steam flow greater than feed flow mismatch caused by closure of the "B" main feed regulating valve.
8.0 ADDITIONAL INFORMATION At the time of this event Unit 1 was defueled as part of a scheduled refueling outage.
Component failure information:
Description: � Driver Card Mark No. 02-FW-FCY-2498 Manufacturer: Westinghouse Model No.: � 2837A16G03 Our Instrument and Calibration group is refurbishing these driver cards on site.