05000374/LER-2001-003
| Docket Number | |
| Event date: | |
|---|---|
| Report date: | |
| Reporting criterion: | 10 CFR 50.73(a)(2)(iv)(A), System Actuation |
| 3742001003R00 - NRC Website | |
DOCKFT (21 FACILITY NAME (1) LaSalle County Station, Unit 2 YEAR r SEQUENTIAL NUMBER L R NUMBER (61 PAGE (31 2 OF � 3 PLANT AND SYSTEM Identification General Electric - Boiling Water Reactor, 3489 Megawatts Thermal Rated Core Power
A. CONDITION PRIOR TO EVENT
Unit(s): 2 Reactor Mode(s): 1 Mode(s) Name: Run
B. DESCRIPTION OF EVENT
Event Date: 09/03/01 � Event Time: 1728 Power Level(s): 100 On September 3, 2001, Unit 2 ESF Bus 241Y lost power due to actuation of the Division 1 undervoltage (UV)[EB] protective circuit. The feed breaker to the bus opened due to the Bus UV load shedding logic. On loss of power to ESF Bus 241Y, the feedwater level control [JB] circuits lost power. The loss of control power drove the feedwater demand to zero and prevented the operators from taking control of the components. The loss of power caused the "A" Reactor Protection System (RPS)[JE] bus to trip. This results in half of the scram logic being made up and various Primary Containment [JM] logic being initiated.
At 1728 hours0.02 days <br />0.48 hours <br />0.00286 weeks <br />6.57504e-4 months <br /> the operators inserted a manual scram when it became apparent that reactor water level could not be controlled in the normal range with feedwater. All control rods fully inserted. Following the scram, reactor water level continued to decrease to minus 50 inches. High Pressure Core Spray (HPCS)[BJ] and Reactor Core Isolation Cooling (RCIC)[BN] automatically initiated and injected to restore level.
The Anticipated Transient Without a Scram logic tripped both Reactor Recirculation pumps off. The Alternate Rod Injection logic also initiated.
The UV logic started the unit common diesel generator (EDG)[EK] and energized Bus 241Y once it was up to speed. However, the UV signal was still in and prevented the for Division 1). The redundant systems (i.e. RHR Loops B and C, RCIC, HPCS and Low Pressure Coolant Injection) were available and operated as necessary to remove decay heat and control reactor vessel water level and pressure.
All systems operated as designed. The lowest reactor water level reached was minus 55 inches. Operators closed the Main Steam Isolation Valves as level continued to increase above procedural limits. Safety relief valves were manually actuated in accordance with procedure to control reactor water level and remove decay heat.
This event is reportable pursuant to 10CFR50.73(a)(2)(iv)(A) as an event that resulted in manual actuation of the reactor protection system. An Emergency Notification System call was made at 2102 hours0.0243 days <br />0.584 hours <br />0.00348 weeks <br />7.99811e-4 months <br /> on September 3, 2001.
NRC FORM 366A (MM-YYYY) NRCFORM366AU.S. NUCLEAR REGULATORY COMMISSION (MM-YYYY) FACILITY NAME (11 C. C CAUSE OF EVENT DOCKET (21 L R NUMBER (6) 2001 _ 003 _ 00 The root cause of this event was failed fuses in the potential transformer portion of the Division 1 UV protective circuit. This UV protection circuit tripped power to the reactor water level control system, driving the demand signal to zero and preventing operator intervention from controlling level. In the event of a single potential transformer fuse failure, UV protection would not initiate. There are two B phase fuses in parallel, and one fuse each in phases A and C. The failure of both B phase fuses or the failure of both A and C phase fuses will initiate UV protection on bus 241Y. A fuse failure in phase A or C will actuate one channel of UV and degraded voltage relays, and voltage indication in the Main Control Room will not show full voltage on all three phases. The failure of a single B phase fuse will not be detected in the control room or locally at the switchgear. This design deficiency contributed to the event.
D. SAFETY ANALYSIS
This scram was due to loss of feedwater flow at 100% power. This transient is bounded by loss of feedwater event with a single failure. The loss of a single ESF bus caused the failure of some ESF systems (i.e. 2A RHR, LPCS and AC power crosstie to Unit 1 for Division 1) to respond to the event. The redundant systems were available and operated as necessary to remove decay heat and control reactor vessel level and pressure. HPCS injection and RCIC injection were valid responses to low RPV level due to loss of feedwater and loss of AC power. RPV level was restored as expected. RPV pressure was maintained within expected values using RCIC and SRVs.
Decay heat was removed using redundant 2B RHR in suppression pool cooling.
Suppression pool temperature was maintained within procedural limits.
High RPV level was mitigated per procedure by closing the MSIVs. This caused the loss of the condenser as a heat sink. However SRVs and RHR 2B were able to remove the decay heat and control reactor vessel water level. This method is the safety related (UFSAR evaluated) method of responding to this event.
E. CORRECTIVE ACTIONS
Corrective actions include replacing all four Bussmann JCW-lE fuses with GE type EJI fuses which have been found to be more robust.
To minimize the possibility of losing the bus, a modification will install a new PT drawer that will allow a single 'B' phase fuse failure to be detected in the control room (low voltage reading) and locally at the switchgear (UV and degraded voltage relay flags). (ATM # 75014)
F. PREVIOUS OCCURRENCES
A review of Licensee Event Reports over the previous five years found no previous or similar occurrences.
G. COMPONENT FAILURE DATA
Bussmann, 1 amp, JCW-lE fuses