05000327/LER-2010-003
| Docket Numbersequential ,Revmonth Day Year ' Year Month Day Yearnumber No 0' 1 | |
| Event date: | 12-20-2010 |
|---|---|
| Report date: | 02-17-2011 |
| Reporting criterion: | 10 CFR 50.73(a)(2)(v), Loss of Safety Function |
| 3272010003R00 - NRC Website | |
L�PLANT CONDITION(S) Sequoyah Nuclear Plant (SQN), Unit 1 was operating in Mode 1 at approximately 100 percent power when the reactor was manually tripped.
H. DESCRIPTION OF EVENT
A. Event:
On December 20, 2010, at 0050 Eastern Standard Time (EST), SQN, Unit 1 reactor was manually tripped as a result of a fire associated with the main generator [EMS Code EL]. At approximately 0045 EST site personnel reported smoke in and around the main generator of the Turbine Building. Operations personnel witnessed a fire in the neutral bus duct [EMS Code BDUC] housing of the main generator and initiated an emergency alarm. Operations personnel entered into the abnormal operating procedure for plant fires. The fire brigade and the Incident Commander responded to the fire alarm and validated a fire present inside a vent/inspection opening of the neutral bus duct housing. After validation of the fire, at 0050 EST operations personnel manually tripped the reactor and turbine to support the fire brigade actions to arrest the fire. The unit was stabilized in Mode 3. The fire brigade reported the fire to be extinguished at around 0100 EST.
B. Inoperable Structures, Components, or Systems that Contributed to the Event:
None.
C. Dates and Approximate Times of Major Occurrences:
Date Description December 20, 2010 Approximate Time 0045 EST Smoke was reported in the turbine building and site personnel witnessed a fire in neutral bus duct housing of the main generator.
0050 EST Fire Brigade and the Incident Commander witnessed flames in an opening of the main generator neutral bus duct housing.
reactor and turbine, and stabilized the unit in Mode 3.
0055 EST Operations personnel took actions to vent hydrogen from the main generator.
0100 EST Fire Brigade Leader reports to the Operations Shift Manager that the fire is extinguished.
D. Other Systems or Secondary Functions Affected:
Following the reactor trip, the condensate and feedwater system [EllS Code SN] experienced a feedwater heater drain tank pump [EllS Code P] trip and feedwater heater [EllS Code HX] string isolation. The Number 7 Heater Drain Tank Pump 1B Motor tripped because of an apparent low level in the Number 7 Heater Drain Tank.
Subsequently, the low pressure "C" feedwater heater string (i.e., stages 5, 6, and 7) isolated. Maintenance was performed on this equipment to determine and correct identified issues. No equipment issues were noted. It is suspected based on operating experience of this equipment that the drain tank did experience a low level condition resulting in the pump trip with subsequent heater string isolation.
E. Method of Discovery:
Site personnel reported smoke in and around the SQN, Unit 1 main generator of the turbine building. Operations personnel located the source of the smoke and initiated an emergency alarm.
F. Operator Actions:
Based on the valid information of a fire located in the neutral bus duct housing of the main generator, the SQN, Unit 1 Unit Supervisor in accordance with an operating procedure directed a manual reactor trip.
G. Safety System Responses:
The safety systems performed as designed for the reactor trip. Auxiliary feedwater (AFW) system [EllS Code BA] automatically initiated following the reactor trip. Operators reduced the flow of AFW following the reactor trip to mitigate a decrease in reactor coolant system (RCS) [EllS Code AB] average temperature. All safety systems remained within Technical Specifications (TS) and Updated Final Safety Analysis Report (UFSAR) limits.
CAUSE OF THE EVENT
A. Immediate Cause:
The immediate cause of the unit trip was the initiation of a manual reactor trip signal.
B. Root Cause:
The root cause was identified as a cyclic vibration induced failure of the 'A' phase neutral bushing porcelain, that allowed a hydrogen leak that ignited.
C. Contributing Factor:
The SQN, Unit 1 generator ran from 2004 to the fall of 2010 with a rotor that showed evidence of a shorted turn in the rotor. The shorted turn in the rotor increased vibration, which could have been a contributor to the degradation of the bushing porcelain. This rotor with suspected shorted turn was replaced during the fall 2010 refueling outage, which significantly reduced the vibration levels.
IV. ANALYSIS OF THE EVENT
This event is most similar to and bounded by the analyzed event of loss of external electrical load and/or turbine trip, with the plant safety systems operating as designed during and following the manual reactor trip. Prior to the event, SQN, Unit 1 was operating in Mode 1 at approximately 100 percent power. The reactor coolant system (RCS) pressure was approximately 2235 pounds per square inch gauge (psig) with an average temperature near the program value of 578 degrees Fahrenheit (F) prior to the plant transient. Following the reactor trip, the loss of nuclear heat generation and the introduction of cold AFW resulted in a decrease in RCS average temperature to approximately 545 degrees F and RCS pressure declined to 2030 psig. Operators at approximately 5 minutes following the reactor trip, reduced the AFW flow to approximately 200 gallons per minute (gpm) to each steam generator (SG) [EllS Code SG] to mitigate the RCS temperature and pressure decrease.
Steam pressure was nearly 867 psig before the reactor trip and increased to approximately 1023 psig when the turbine stop valves closed. Steam dumps to the condenser [EllS Code COND] operated as expected and remained available. The introduction of AFW decreased steam pressure to approximately 990 psig within 30 minutes and once AFW flow was matched to decay heat generation, steam pressure returned to the no-load value of approximately 1005 psig. RCS temperature was restored to its no-load value of 547 degrees F as the secondary side pressure recovered. Emergency boration was not required based on the shutdown margin requirements. Additionally, AFW system actuated as designed on the SG low-low level signal, recovering the SG water levels following the reactor trip. The plant responded as expected for the conditions of the trip, with exception to the secondary plant equipment described above. No TS limits were exceeded and the UFSAR analysis of this event remained bounding.
V. ASSESSMENT OF SAFETY CONSEQUENCES
Based on the above "Analysis of The Event," this event did not adversely affect the health and safety of plant personnel or the general public.
VI.�CORRECTIVE ACTIONS A. Immediate Corrective Actions:
The immediate corrective actions included replacement of the failed neutral bushing.
B. Corrective Actions to Prevent 'Recurrence:
A preventive maintenance instruction is being developed to define a service life for refurbishment/replacement of SQN, Unit 1 and Unit 2 main and neutral bushings.
Additionally, the SQN, Unit 2 bushings are being visually inspected during the next scheduled refueling outage.
VII. ADDITIONAL INFORMATION
A. Failed Components:
The failed component was a main generator bushing from Westinghouse Electrical Corporation, style number 5621D28G01, insulation class 25kV.
B. Previous LERs on Similar Events:
A review of previous reportable events for the past three years did not identify any previous similar events.
C. Additional Information:
None.
D. Safety System Functional Failure:
This event did not result in a safety system functional failure in accordance with 10 CFR 50.73(a)(2)(v).
E. Unplanned Scram with Complications:
This event did not result in an unplanned scram with complications.
VIII. COMMITMENTS
None.