05000416/LER-2008-004
| Docket Numbersequential Revmonth Day Year Year Month Day Yearnumber No. N/A N/A | |
| Event date: | 10-23-2008 |
|---|---|
| Report date: | 12-15-2008 |
| Reporting criterion: | 10 CFR 50.73(a)(2)(iv)(A), System Actuation |
| 4162008004R00 - NRC Website | |
On October 23, 2008 at 0721 Grand Gulf Nuclear Station experienced and, automatic reactor SCRAM due to decreasing reactor coolant level. Upon decreasing to Level 3 (+11.4 inches) a Reactor Protection System (RPS) [JC] SCRAM signal was auto initiated. The initiation of RPS resulting in an automatic reactor SCRAM is reportable pursuant to 10 CFR 50.73(a)(2)(iv)(A).
B. INITIAL CONDITIONS
At the time of the event, the reactor was in OPERATIONAL MODE 1 with reactor power at approximately 10 percent. There were no additional inoperable structures, systems, or components at the start of the event that contributed to the event.
C. DESCRIPTION OF OCCURRENCE
On October 23, 2008 at 0721 Grand Gulf Nuclear Station experienced an automatic reactor SCRAM due to decreasing reactor coolant level caused by loss of "A" RFPT train. At the initiation of the SCRAM the unit was operating in Mode 1 at approximately 10 percent power. Prior to the SCRAM Operations had performed a Reactor Feed Pump Turbine (RFPT) [SJ] "B" Overspeed Trip Test with the RFPT uncoupled from the pump. The system was ready for restoration therefore the control room reactor operator (RO) dispatched a non licensed operator (NLO) into the plant to restore the system. The NLO proceeded to manipulate controls on what was thought to be the "B" RFPT train however the NLO manipulated the steam supply drains on the operating "A" RFPT train. These steps did not initiate the loss of the "A" RFPT. As the NLO proceeded to the next step it was noticed that there was no light indication for the next valve to be manipulated. Having performed the valve line up on the "A" RFPT train the day before the NLO now realized it was the "A" train rather than the "B" train due to the fact that the NLO remembered the lack of light indication. At the same time the NLO realized the mistake the main control room called the NLO on the plant public address system.
The NLO concluded that the page concerned the mistake and attempted to correct the error. In an attempt to undo the error the NLO closed the High Pressure Steam Supply Valves to the "A" RFPT. As a result of the isolation of the steam inlet valves the "A" RFPT lost steam pressure. Shortly thereafter the Reactor Feed Pump discharge pressure dropped below reactor pressure, which resulted in a loss of feedwater flow to the reactor vessel. This resulted in a reactor SCRAM upon reaching Level 3 (+11.4 inches).
The reactor SCRAM came in as designed at Level 3 (+11.4 inches). Operators manually initiated Reactor Core Isolation Cooling (RCIC) to restore and maintain reactor vessel level. The lowest indicated reactor vessel level was -39 inches Wide Range. All withdrawn control rods fully inserted and no emergency core cooling system (ECCS) initiations were received. No Safety Relief Valves lifted as a result of this event and all other equipment operated as expected.
NRC FORM 366A U.S. NUCLEAR REGULATORY COMMISSION (9-2007) I0 REV , D. CAUSE of OCCURRENCE The cause of the event was that the NLO did not use the self checking standard of TOUCH-READ-READ which involves touching the component intended to be manipulated, reading the tag, and reading the procedural step to verify the correct component was about to be manipulated. The NLO also did not use the required circle and slash method of place keeping.
The root cause of this event was determined to be a lack of commitment to Human Performance Program implementation.
E. CORRECTIVE ACTIONS
Immediate Corrective Actions — Reactor water level was restored and the plant placed in a stable condition.
The individual involved was removed from shift and a human performance error review was conducted.
Additionally, temporary covers were placed over the switches on the 1H22-P175 Panel for the RFPT "A" and "B" controls.
Long Term Corrective Actions - Condition Report CR-GGN-2008-06195 was written and will address any additional actions.
F. SAFETY ASSESSMENT
Immediate actions performed by the Operations staff were adequate and appropriate in placing and maintaining the reactor in a safe shutdown condition. The lowest reactor level indicated was -39 inches Wide Range. This is above the initiation setpoint for ECCS systems. RCIC was manually initiated to restore and maintain reactor vessel level.
The Group 2 Residual Heat Removal (RHR) to Radwaste and Group 3 RHR Shutdown Cooling automatic isolations were received however no valves isolated because they were in their normally closed position prior to the event. No Safety Relief Valves lifted as a result of this event and all other systems performed as required.
This event did not prevent the fulfillment of a safety function therefore there were no safety system functional failures. The health and safety of the public was not compromised by this event.
G. ADDITIONAL INFORMATION
Previous Similar Events — Pursuant to 10CFR50.73(b)(5) this issue is considered an infrequent event. There has not been any occurrence of the same underlying concern in the past two years at Grand Gulf Nuclear Station.