05000366/LER-2009-004
| Docket Number | |
| Event date: | 06-23-2009 |
|---|---|
| Report date: | 08-10-2009 |
| Reporting criterion: | 10 CFR 50.73(a)(2)(iv)(A), System Actuation |
| 3662009004R00 - NRC Website | |
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PLANT AND SYSTEM IDENTIFICATION
General Electric - Boiling Water Reactor Energy Industry Identification System codes appear in the text as (EllS Code XX).
DESCRIPTION OF EVENT
On June 23, 2009 at 03:51 EDT, Unit 2 was in mode 1 with an approximate reactor power of 1710 CMWTh. At this time an automatic reactor scram from a turbine (EllS Code TA) trip due to high reactor water level occurred. Prior to the event the reactor was increasing in power during startup from a recent outage. At approximately 03:40 EDT, the 2C32-K648 median controller (EllS Code SJ) failed and no longer tracked actual reactor water level. This controller provides input for the indicated level on the plasma display (EllS Code SJ) and the 2C32-R600 feedwater master controller (EDS Code SJ). Reactor water level began to decrease approaching 30 inches as shown by Safety Parameter Display System (SPDS, EDS Code IU) data however the plasma panel did not indicate this change in level. Reactor water level then began to increase. The High Pressure Coolant Injection (HPCI, EDS Code BJ) and Reactor Core Isolation Cooling (RCIC, El IS Code BN) High Reactor Water Level Trip alarms were received. Approximately 30 seconds after receiving the alarm the turbine tripped on the turbine control valve (EllS Code TA) fast closure trip signal due to high reactor water level. All control rods fully inserted (EllS Code JD) and Reactor Feed pumps (RFP, EllS Code SJ) tripped. Reactor water level initially decreased to approximately negative 25 inches, due to void collapse. Primary Containment isolation Valve Group 2 (EllS Code JM) isolation setpoint was reached and the Group 2 valves isolated. Both the 'A' and 'B' Reactor Feed Pumps initially tripped, and the 'A' feed pump was restarted and used to restore and maintain reactor water level. Reactor pressure reached an upper value of approximately 964 psig. No Safety Relief Valves (SRVs, EllS Code SB) opened, nor were they required to open, based on the maximum pressure reached.
Reactor water level increased to a maximum value of approximately 60 inches above instrument zero but was restored to normal range.
CAUSE OF EVENT
The cause of this event was the failure of an internal power supply electrolytic capacitor on the power supply board which caused a failure of the DC power supply for the Yokogawa level controller 2C32- K648.
During recovery efforts it was determined that the 2C32-K648 controller was not responding to reactor water level increases and was displaying the error code P.error. Per the Yokogawa vendor manual, this is indicative of an internal power supply failure. The P.error code was intermittently displayed during the recovery process.
The controller was removed from service and transported to the Maintenance lab for analysis. Power was applied to the controller, and the P.error code was again displayed. This error was intermittent during the analysis period. Internal inspection of the power supply identified a failed electrolytic capacitor.
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This event is reportable per 10 CFR 50.73(a)(2)(iv)(A) because unplanned actuations of a safety feature system listed in 10 CFR 50.73 occurred. In this instance, a reactor protection system (RPS) actuation resulted in a reactor scram. The reactor protection system (ENS Code JC) actuated on turbine control valve fast closure when a high reactor water level was sensed.
Fast closure of the turbine control valves is initiated when a reactor high water level condition exists. The turbine control valves close as rapidly as possible to prevent water intrusion into the main steam lines.
Valve closing causes a sudden reduction in steam flow that, in turn, results in a reactor vessel pressure increase. If the pressure increases to the pressure relief setpoints, some or all of the SRVs will briefly discharge steam to the suppression pool (EllS Code BL). In this event the reactor pressure did not reach corresponding setpoint of the SRVs.
Reactor scram and recirculation pump (EllS Code AD) trip initiation by turbine control valve fast closure prevents the core from exceeding thermal hydraulic safety limits following a main generator or main turbine trip. A reactor scram is initiated on turbine control valve fast closure. The scram, along with the reactor recirculation pump trip system, ensures that the minimum critical power ratio safety limit is not exceeded.
The recirculation pump trip system, upon sensing a turbine control valve fast closure, trips the reactor recirculation pumps resulting in a decrease in core flow. The rapid core flow reduction increases void content and reduces reactivity in conjunction with the reactor scram to reduce the severity of the transients caused by the turbine trip.
In this event, the main turbine tripped as designed in response to the sensed power load unbalance.
The turbine trip actuated the reactor protection system and scrammed the reactor. Vessel water level was maintained well above the top of the active fuel throughout the transient. The water level decrease was terminated prior to reaching the automatic initiation set point for HPCI and RCIC.
Operations restarted the 'A' feed pump immediately following the reactor scram and subsequent feed pump trip. This pump was used to restore and maintain reactor water level. Therefore no safety system actuations on low water level, including emergency core cooling systems, were automatically initiated nor were any required.
Based on this analysis, it is concluded that this event had no adverse impact on nuclear safety.
CORRECTIVE ACTIONS
The 2C32-K648 controller was replaced.
Repetitive Tasks have been created to replace the power supply for this and similar controllers at a prescribed interval.
A new subsection was added to maintenance procedures 57CP-CAL-226-1 and 57CP-CAL-226-2 to provide instructions for replacement of the Yokogawa power supplies
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Previous Similar Events:
There are no similar events within the past two years in which a power supply card failure occurred on a reactor water level control instrument.
� PRINTED ON RECYCLED PAPERNRC FORM 366A (9-2007)