05000282/LER-2001-004
| Event date: | |
|---|---|
| Report date: | |
| Reporting criterion: | 10 CFR 50.73(a)(2)(v)(A), Loss of Safety Function - Shutdown the Reactor |
| 2822001004R00 - NRC Website | |
EVENT DESCRIPTION
On August 1, 2001, Prairie Island Nuclear Generating Plant Unit 1 (PINGP-1) was operating at 100% power. Atmospheric dew point temperatures at the time were greater than 72 degrees F. At 0703 CST, an automatic reactor trip occurred due to negative flux rate.
Immediately prior to the trip, an Apprentice Plant Attendant (APA) on routine rounds detected water from the drip tray beneath the 11 Control Rod Drive Room Unit Cooler overflowing onto the 12AC Control Rod Drive Power Cabinet' and surrounding floor. The reactor trip occurred as the APA was about to notify the Assistant Plant Equipment Operator. A review of plant data indicates the Control Bank C2 control rods energized by the 12AC power cabinet had dropped into the core, resulting in a negative flux rate reactor trip. Instrument and Control (I&C) technicians who responded to troubleshoot the malfunction detected moisture inside the 12AC cabinet. l&C personnel later demonstrated through bench testing of a spare that the stationary gripper varistors of the type installed in the cabinet indicated a significant reduction in phase-to-neutral and phase-to-phase resistance when exposed to a water mist.
Inspection of the 11 Control Rod Drive Room Unit Cooler determined that the drip tray was full and the 1 inch drain line was partially obstructed. The blockage was cleared by plant workers.
Following the reactor trip, six (6) of twenty-nine (29) control rods initially failed to provide rod bottom light indications (instead indicating 15 — 20 steps). This was within the tolerance of the position indicator but did not cause the rod bottom bistables for these lights to actuate.
The operating crew responded by borating in accordance with plant emergency operating procedures. The Auxiliary Feedwater pumps automatically started in response to indicated low steam generator levels. This is an expected response following a reactor trip from 100 percent power. The remainder of the transient recovery proceeded normally.
CAUSE OF THE EVENT
The cause of the event was the lack of preventative maintenance on the 11 Control Rod Drive Room Unit Cooler drip tray drain2. High ambient humidity at the time of the event resulted in a condensation rate on the cooling coils that exceeded the drainage rate through the partially blocked drain line. Contributing factors include the location of the unit cooler which is directly above the 12AC Power Cabinet. Also, there was a similar event occurrence on Unit 2 in 1982. During that event, water seeped into the 21BD Control Rod Drive Power Supply Cabinet and blew a fuse. The Corrective Action Program was not utilized to fully document and evaluate this condition. Instead, informal work processes were relied on.
1 EIIS System Code: AA; Component Code: VRC 2 EIIS System Code: VF; Component Code: DRN The cause of the erroneous control rod position indication following the reactor trip was a calibration bias (equating to approximately 8 steps) that was inadvertently introduced by a recent calibration procedure change. The change failed to anticipate differences between the test equipment used to develop the procedure and actual circuit output. The circumstances that led to the introduction of the bias are still under investigation.
ANALYSIS OF THE EVENT
The event is reportable under 10CFR50.73(a)(2)(v)(A) as an automatic actuation of the reactor protection system in response to the automatic reactor trip. The health and safety of the public were unaffected since the plant systems, with the exception of control rod position indication, responded to the trip as designed.
Impact of Event on Performance Indicators This event affects the number of unplanned scrams per 7,000 critical hours. However, the unit has had three or fewer scrams within the last 7,000 critical hours.
Risk Significance This event involved an unplanned reactor trip without impact to systems required for normal post- trip decay heat removal. The additional boron added to the reactor coolant system did not present a significant challenge to successful operator response nor to the ability of mitigating systems to perform as required. This type of event was anticipated in the design of the plant, and plant procedures and training adequately cover the operator actions that are necessary. As such, the risk significance of this event is very low.
CORRECTIVE ACTIONS
Immediate 1. The blockage in the unit cooler drip tray was cleared by plant workers.
2. The control rod position indications were recalibrated using the previously accepted methodology.
Planned 1. A preventative maintenance program will be established to periodically inspect and clean the drip tray drains on the Unit 1 and Unit 2 Control Rod Drive Unit Coolers.
2. The status of all plant Unit coolers and air handlers near sensitive plant equipment will be evaluated. Preventative maintenance programs, the installation of physical barriers or splash guards, and the removal or relocation of equipment will be initiated as needed.
3. The cause of the calibration bias that led to the inaccurate control rod position indications is currently being investigated. Corrective actions to prevent recurrence will be initiated in accordance with the Prairie Island Corrective Action Program.
These corrective actions have been assigned a high priority within the Prairie Island Corrective Actions Program.
FAILED COMPONENT IDENTIFICATION
None.
PREVIOUS SIMILAR EVENTS
A similar event occurred in May 1982 on Unit 2 due to water intrusion from a drip pan onto the 12BD Control Rod Drive Power Supply Cabinet. This resulted in a blown fuse.