05000306/LER-2008-002
| Prairie Island Nuclear Generating Plant | |
| Event date: | 10-30-2008 |
|---|---|
| Report date: | 12-19-2008 |
| Reporting criterion: | 10 CFR 50.73(a)(2)(iv)(A), System Actuation |
| 3062008002R00 - NRC Website | |
On October 30, 2008 PINGP was in the process of a Unit 2 reactor startup following a scheduled refueling outage. The reactor was critical and was slightly above the POAH when at 1414 CDT a rod control system' urgent failure alarm was received. At this point, PINGP operators were performing reactor physics testing for control rod bank A, which required both rod control groups 1 and 2 to be inserted. When the urgent failure alarm was received, operators noted that group 1 stopped moving while group 2 continued inward motion. Operators immediately stopped group 2 control rods and based on plant conditions at the time, operations personnel initiated a manual reactor trip of PINGP Unit 2.
The manual trip was initiated since power was decreasing with low power conditions present. All control rods fully inserted as expected and all operator actions were as expected for the reactor trip. The event is being reported as required by 10 CFR 50.73(a)(2)(iv)(A) due to the manual actuation of the reactor protection system.
EVENT ANALYSIS
One of the Reactor Protection System2 design functions is to prevent or suppress conditions that could result in exceeding acceptable fuel damage limits by opening the reactor trip breakers and allowing all control rods to fall inward. This rapid insertion of rods creates negative reactivity which causes a rapid reactivity shutdown. When operators manually tripped the reactor, all control rods fully inserted as expected. Therefore, the reactor protection system operated as designed and there was no loss of safety function per 10 CFR 73(a)(2)(v).
SAFETY SIGNIFICANCE
The reactor protection system is designed so that reactor shutdown with rods is completely independent of the normal rod control functions since the reactor trip breakers completely interrupt power to the rod latching mechanisms regardless of existing control signals. When operators manually initiated the reactor trip, all control rods fully inserted as expected. For these reasons this event was of low safety significance.
CAUSE
It was not known at the time of the trip, but subsequent troubleshooting isolated the cause of the rod control urgent failure alarm and failure of control bank A group 1 rods to insert to a phase C fuse failure in the 21AC power cabinet moveable gripper bus duct disconnect switch. This fuse provides power to the moveable gripper assemblies in power cabinet 21AC. Troubleshooting was unable to determine a fault that would cause the fuse to blow and the apparent cause was determined to be a random fuse failure.
CORRECTIVE ACTION
The blown fuse along with all of the moveable gripper bus duct disconnect-switch fuses for each power cabinet was replaced on Unit 2. This restored the full functionality of Unit 2 control rods.
Troubleshooting and repairs were completed on October 31, 2008 at approximately 0031 CDT. At 1115 CDT, PINGP resumed Unit 2 startup and the reactor was successfully restored to criticality on October 31, 2008, at 1311 CDT.
1 EEIIS Component Identifier: AA 2 EEIIS Component Identifier: JC In order to correct the extent of condition, a preventative maintenance program to replace all of the bus duct fuses in both PINGP Unit 1 and Unit 2 every 10 years will be implemented. In addition, all of the bus duct fuses for Unit 1 will be replaced during the next refueling outage.
PREVIOUS SIMILAR EVENTS
Three reactor trip events were found under LER 1-06-01, LER 2-07-01, and LER 1-08-02. LER 1-06-01 describes a manual reactor trip initiated due to a ground in a condensate pump, LER 2-07-01 was an automatic reactor trip caused by a failed safety injection relay, and LER 1-08-02 was an automatic reactor trip caused by a failed reactor protection controller during testing.
Although similar in that they all describe reactor trips related to equipment problems, they are not significant with regard to the subject event because each equipment problem was significantly different.