05000461/LER-2009-005
| Docket Numbersequential Revmonth Day Year Year Month Day Year 05000Number No. | |
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| 4612009005R01 - NRC Website | |
PLANT AND SYSTEM IDENTIFICATION
General Electric — Boiling Water Reactor, 3473 Megawatts Thermal Rated Core Power Energy Industry Identification System (EllS) codes are identified in the text as [XX].
EVENT IDENTIFICATION
Manual Scram on High Water Level Due to Reactor Recirc Pump Trip
A. CONDITION PRIOR TO EVENT
Unit: 1 Event Date: 10/15/09� Event Time: 0538 hours0.00623 days <br />0.149 hours <br />8.895503e-4 weeks <br />2.04709e-4 months <br /> CST Reactor Mode: 1 Mode Name: Power Operation� Power Level: 96.6 percent
B. DESCRIPTION OF EVENT
On 10/15/09 at 0537 hours0.00622 days <br />0.149 hours <br />8.878968e-4 weeks <br />2.043285e-4 months <br />, operators in the Main Control Room (MCR) received an alarm [ALM] indicating the 'B' Reactor Recirculation (RR) [AD] Pump [P] had unexpectedly tripped from fast speed to off. The 'A' Reactor Recirculation Pump remained in fast speed. The trip of 'B' Reactor Recirculation Pump caused Reactor Pressure Vessel (RPV) water level to increase, and the high RPV water level alarm annunciated.
At 0538 hours0.00623 days <br />0.149 hours <br />8.895503e-4 weeks <br />2.04709e-4 months <br />, with RPV water level increasing to the predetermined level of 48 inches, operators responded as expected by placing the Reactor Mode Switch [HS] into the ShutdoWn position to initiate a manual reactor scram prior to the high RPV water Level 8 reactor scram setpoint of 52.0 inches. All control rods fully inserted as a result of the manual reactor scram. Immediately after the scram RPV water level decreased below the low RPV water Level 3 setpoint as expected and operators responded by entering Emergency Operating Procedure (EOP) 1, RPV Level Control.
As expected, the low RPV water Level 3 trip caused primary containment isolation valves [ISV] in Group 2 (Residual Heat Removal (RHR) [BO]), Group 3 (RHR), and Group 20 (miscellaneous systems) to receive signals to shut; operators verified that the containment isolation valves properly responded to the Level 3 trip.
At 0556, plant conditions were stable with RPV water level above the level 3 trip setpoint and the reactor scram was reset.
At 0558 hours0.00646 days <br />0.155 hours <br />9.22619e-4 weeks <br />2.12319e-4 months <br />, operators were unable to fully shut Main Steam [SB] to Moisture Separator Reheater inlet valves 1B21F302A and 1B21F500A resulting in reactor vessel cooldown rates challenging the prescribed pressure band. As a result, operators manually shut the Main Steam Isolation Valves to maintain reactor vessel cooldown rates within limits. Main Steam Line drains were used to control reactor pressure. Issue Report 979911 was initiated to investigate and correct the failure of these valves to shut.
At 0717 hours0.0083 days <br />0.199 hours <br />0.00119 weeks <br />2.728185e-4 months <br />, operators started the 'B' Residual Heat Removal system in suppression pool cooling mode to support running the Reactor Core Isolation Cooling (RCIC) system [BN] in the pressure control mode. At 0725 hours0.00839 days <br />0.201 hours <br />0.0012 weeks <br />2.758625e-4 months <br />, operators started the RCIC system in the tank to tank mode to supplement reactor pressure control.
At about 0927 hours0.0107 days <br />0.258 hours <br />0.00153 weeks <br />3.527235e-4 months <br />, operators exited EOP-1.
Initial investigation at the 3B and 4B feed breakers [BKR] for the 'B' RR pump motor [MO] revealed that the ground relays [64] had actuated on both breakers. Further troubleshooting and megger testing of the motor determined the cause of 'B' Reactor Recirculation pump trip was a phase to ground fault inside the pump motor. The motor has been sent to a vendor for failure analysis.
Upon trip of the 'B' RR pump, an immediate operator action is to shut the RR pump discharge valve for the idle reactor recirculation pump to prevent reverse rotation of the pump. During this event, operators attempted to close the 'B' RR pump discharge valve, 1B33F067B, but it would not fully close. Issue Report 979732 was initiated to investigate and correct the failure of valve 1B33F067B to fully close.
No other inoperable equipment or components directly affected this event.
Issue Report 979700 was initiated to investigate this event and initiate corrective actions.
The NRC Operations Center was notified at 0806 hours0.00933 days <br />0.224 hours <br />0.00133 weeks <br />3.06683e-4 months <br /> about this reactor scram via Event Notification 45433.
C. CAUSE OF EVENT
The cause of this event is attributed to the failure of the B' RR pump motor. The pump motor was shipped to a vendor for failure analysis. The vendor concluded that the failure in the RR pump motor was a B-Phase turn-to-turn fault located at a connection end coil knuckle. The failure was limited to a single coil end turn knuckle of the winding and was located in a position that is not susceptible to air flow or foreign material. The affected area was cut out and the remaining windings passed all electrical testing. The apparent cause of the failure was a random insulation breakdown of the original winding that resulted in a turn-to-turn failure.
Clinton Power Station is designed to stay on line with a single recirculation pump trip. A review of the response of the Feedwater Level Control System (FWLCS) [JB] to the 'B' RR Pump trip on 10/15/09 identified that it was very similar to that of a previous event on 2/10/08 when the 'B' RR Pump tripped to off and an unexpected reactor scram occurred on high RPV water Level 8.
The cause of the 2/10/08 Reactor Scram identified that FWLCS tuning was inadequate to ensure that Reactor water level does not approach the Level 8 Scram setpoint when a single RR pump trips off.
Specifically, the FWLCS flow controllers were not properly tuned to provide adequate margin to the high water Level 8 scram setpoints. Consequently, the feedwater demand signal did not decrease fast enough to reduce feedwater pump flow after the RR pump trip. In response to the cause an action plan was developed to correct the plant response to a single RR pump trip including dynamic tuning of the FWLCS.
At the time of this 10/15/09 event, all of the actions were not completed from the 2/10/08 event but were completed during startup from refueling outage Cl R12 in February 2010.
D. SAFETY CONSEQUENCES
This event is reportable under the provisions of 10 CFR 50.73 (a) (2) (iv) (A) as an event that resulted in a manual reactor scram while the reactor was critical. No significant safety consequences resulted from this event because required safety systems were available and functioned as designed within safety limits. The reactor was shut down safely and maintained in a safe shut down condition.
This reactor scram event was compared to similar previous events and the plant response and behavior was almost identical to the previous events. The reactor scram was compared to Updated Safety Analysis Report (USAR) Section 15.3, Decrease in Reactor Coolant System Flow Rate. The fission product barriers (fuel clad, reactor, pressure boundary, containment) were not challenged during this event. No Safety Relief Valves lifted during this event and pressure control maintained by Main Steam Line Drains and the RCIC system. The Motor Driven Reactor Feed Pump was used to maintain RPV water level.
This event report does not identify any safety system functional failures.
E. CORRECTIVE ACTIONS
A spare RR pump motor was installed in place of the 'B' RR pump motor that failed. The 'B' RR pump motor that failed has been repaired.
Final tuning of the FWLCS was completed prior to and during refueling outage Cl R12.
F. PREVIOUS OCCURRENCES
unexpected trip of the 'B' RR pump. This event is discussed further in the CAUSE OF EVENT section of this report.
G. COMPONENT FAILURE DATA
The Reactor Recirculation Pump B motor was manufactured by General Electric, and is 6300 horsepower, 6600 volts, 484 amps, 3-phase, non-safety, Critical Class 1, continuous run motor, model 264X805.