05000282/LER-2008-002
| Prairie Island Nuclear Generating Plant | |
| Event date: | |
|---|---|
| Report date: | |
| Reporting criterion: | 10 CFR 50.73(a)(2)(iv)(A), System Actuation 10 CFR 50.73(a)(2)(v), Loss of Safety Function |
| 2822008002R00 - NRC Website | |
On July 31, 2008, 0817 CDT, Prairie Island Nuclear Generating Plant (PINGP), Unit 1 was operating at 100 percent power when the reactor tripped on an over-temperature delta T (OT delta T) reactor trip signal from the reactor protection systeml.
At the time of the event, instrumentation and control personnel were performing the quarterly analog reactor protection functional test on the yellow channel when the red channel OT delta T bistable was actuated. Subsequent troubleshooting and root cause investigation determined that the red channel bistable actuation was caused by the failure of a Foxboro H-line (model 62H-2E-O) F delta Q controller2 in the OT delta T circuit. The controller output failed high causing the OT delta T setpoint to drop below the actual delta T parameter thus causing a red channel reactor trip signal. The red channel OT delta T reactor trip signal combined with the yellow channel OT delta T bistables being in test (trip) as directed by the surveillance procedure completed the 2 out of 4 coincidence logic required to initiate a reactor trip. During the performance of yellow channel OT delta T analog testing, the red channel OT delta T setpoint was not expected to be challenged nor was a reactor trip expected at any point.
All automatic actions for a reactor trip occurred as required with the following exceptions:
Subsequent to the trip, the Unit 1 turbine-driven auxiliary feedwater pump (11 TDAFWP) auto started as designed, but tripped 42 seconds later on low discharge pressure. And a Unit 1 Turbine 2 Reheat Stop Valve indicated intermediate vice closed. However, physical inspection verified that this valve was indeed closed and that the intermediate indication was caused due to a failed switch rod (linkage) that actuates a proximity switch to indicate valve position. Operator response and recovery actions for the reactor trip were completed as expected.
EVENT ANALYSIS
A reactor trip is required to be reported per 10 CFR 50.73(a)(2)(iv)(A). The reactor trip by itself did not result in a condition that could have prevented the fulfillment of a safety function per 10 CFR 50.73(a)(2)(v). Issues associated with the 11 TDAFWP are addressed in LER 1-08-03. The erroneous indication for the Unit 1 Turbine 2 Reheat Stop Valve is not directly related to the reactor trip and was repaired under the site's corrective action program on 08/02/2008.
1 EIIS System Code: JC 2 EIIS Component Identifier: IMOD U.S. NUCLEAR REGULATORY COMMISSIONNRC FORM 366A LICENSEE EVENT REPORT (LER)(9-2007)
CONTINUATION SHEET
SAFETY SIGNIFICANCE
The OT delta T trip along with the overpower delta T trip is designed to keep the departure from nuclear boiling ratio (DNBR) greater than the limit for slow reactivity additions. This event was due to an equipment failure and not related to a reactivity addition. With the exception of the 11 TDAFWP trip and a Unit 1 Turbine 2 Reheat Stop Valve position indication, all systems performed as expected to the reactor trip signal and operators responded and recovered as expected. Thus, this event did not affect the health and safety of the public and the safety significance of this event is considered minimal.
CAUSE
The equipment root cause for the failure of the F delta Q controller is attributed to the random failure of varactor diode (CR1) located inside the controller. Although this controller was refurbished in 1985, only the capacitors were routinely replaced as part of refurbishments. Therefore, CR1 was not replaced as part of the 1985 refurbishment.
The organizational cause was found to be the inadequate prioritization by the site in the creation of a preventive maintenance strategy for the analog components within the reactor protection and control system.
CORRECTIVE ACTION
Immediate corrective action:
1. Replaced the failed F delta Q proportional controller.
Planned corrective actions include:
1. Replacement or refurbishment of all F delta Q proportional controllers.
2. Implement an improved preventive maintenance strategy for the Foxboro H-Line components of the reactor protection and control system.
3. Implement a Life Cycle Management Plan for the reactor protection and control system. This will ensure timely preventative replacement of the Foxboro H-Line components.
Although both of these events describe reactor trips due to equipment related issues, the MG-6 style relay failure was due to high contact resistance while the ground caused by degraded motor insulation was an age related failure. Both of the previous LERs include preventive maintenance in the corrective actions. LER 1-06-01 included an action to institute a large motor program that would not prevent the event of this LER. LER 2-07-01 included an action to implement preventive maintenance strategy for all critical equipment. This action was completed in February of 2008, but some improvements were made apparent by the event of this LER (see Corrective Action discussion, above).
3 EIIS System Code: BQ 4 EIIS System Code: SD