05000424/LER-2002-003
| Vogtle Electric Generating Plant - Unit 1 | |
| Event date: | |
|---|---|
| Report date: | |
| 4242002003R00 - NRC Website | |
U.S. NUCLEAR REGULATORY COMMISSION FACILITY NAME (1) DOCKET LER NUMBER (6) PAGE (3)
A. REQUIREMENT FOR REPORT
This event is reportable per 10 CFR 50.73 (a)(2)(iv) because unplanned engineered safety feature and unplanned reactor protection system actuations occurred.
B. UNIT STATUS AT TIME OF EVENT
At the time of this event, Unit 1 was in power ascension in Mode 1 (power operations) at 30 percent of rated thermal power. Personnel were making preparations to synchronize the generator to the grid. Other than that described herein, there was no inoperable equipment that contributed to the occurrence of this event.
C. DESCRIPTION OF EVENT
On April 20, 2002, power ascension was in progress following a refueling outage. At approximately 0450 EDT, the operator designated as the steam generator water level controller (SGWLC) was sent from the control room to investigate a main generator alarm. At 0500 EDT, a control room alarm was received due to water level lowering in steam generator (SG) #4. The reactor operator (RO) increased the speed of the A main feedwater pump (MFP). Almost simultaneously, the balance-of-plant operator (BOP) recognized that the A MFP mini-flow valve was open and an operator in the turbine building was directed to shut it. These combined actions caused water levels in all four SGs to rise rapidly. Both the RO and the SGWLC (who had returned to the control room) attempted to control the water levels by lowering the MFP speed and throttling the main feedwater regulating valves. However, at 0509 EDT, SG #4 reached its high level setpoint which led to a turbine trip, a main feedwater isolation, and an auxiliary feedwater (AFW) system actuation. The shift superintendent ordered a manual reactor trip, which occurred at 0510 EDT, and the unit was stabilized in Mode 3 (hot standby).
D. CAUSE OF EVENT
The causes of this event were:
1) The initial lowering of the SG water levels was caused by failure of the operating crew to increase the MFP speed at normal increments to correspond with increased water demand as reactor power was rising.
FACILITY NAME (1) DOCKET LER NUMBER (6) PAGE (3) 2) Inadequate coordination in response to the lowered SG water levels. Specifically, closing the MFP mini-flow valve without fully considering the impact on the plant.
Finally, there were no characteristics of the work location that contributed to the occurrence of these errors by the licensed operators involved.
E. ANALYSIS OF EVENT
The main feedwater system isolated and the auxiliary feedwater system actuated as designed following the receipt of the SG high water level signal. With the main feedwater system isolated and reactor power at 30%, control room personnel acted appropriately to manually trip the reactor and prevent a challenge to the automatic trip actuation circuitry. Based on these considerations, there was no adverse effect on plant safety or on the health and safety of the public as a result of this event.
This event does not represent a safety system functional failure.
F. CORRECTIVE ACTIONS
1) The operating crew has been counseled on the expectations of managing evolutions and performed just-in-time training prior to assuming the next shift.
2) This event will be addressed in licensed operator continuing training, emphasizing error pre- cursors and how proper management of activities could have prevented the event. Also, skill- based tasks performed during unit startup will be reviewed, and appropriate changes made to training program(s) by September 1, 2002.
3) Appropriate procedures will be revised by October 10, 2002, to ensure proper use of the MFP mini-flow valves during unit start-ups.
G. ADDITIONAL INFORMATION
1) Failed Components:
None U.S. NUCLEAR REGULATORY COMMISSION 2) Previous Similar Events:
There have been no previous similar events in the last two years.
3) Energy Industry Identification System Code:
Main Feedwater System — SJ Auxiliary Feedwater System — BA Main Steam System - SB Reactor Coolant System - AB