05000413/LER-2003-005
| Docket Number | |
| Event date: | |
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| Report date: | |
| Reporting criterion: | 10 CFR 50.73(a)(2)(iv)(A), System Actuation |
| 4132003005R00 - NRC Website | |
BACKGROUND
This event is being reported under 10 CFR 50.73(a)(2)(iv)(A), any event or condition that resulted in manual or automatic actuation of any Engineered Safety Feature including the Reactor Protection System.
Catawba Nuclear Station Unit 1 is a Westinghouse Pressurized Water Reactor [EIIS: ROT]. Unit I was operating in Mode 1, "Power Operation" at 95% power immediately prior to this event. Reactor power was being maintained at 95% as a precautionary measure against possible spurious temperature spikes while loop lA reactor coolant system temperature bistables were placed in the tripped condition. The bistables were tripped due to a failure of the hot leg temperature detectors.
The Reactor Protection System [EIIS: JC] is designed to open the reactor trip breakers to shutdown the reactor whenever a plant operating condition exceeds a specific setpoint. The OTdT reactor trip setpoint is designed to protect the reactor from departure from nucleate boiling (DNBR) conditions by calculating a trip setpoint based on pressurizer pressure, reactor coolant system temperature, and axial power distribution. When two of the four OTdT channels actuate, a reactor trip signal is generated.
With the exception of the loop lA OTdT channel being in the tripped condition, there were no other systems, structures, or components out of service that had any significant effect on the event.
EVENT DESCRIPTION
(Dates and times are approximate) Date/Time � Event Description 08/06/03 � Loop lA OTdT channel placed in tripped condition due to failure of the hot leg temperature detectors. The detectors were inadvertently damaged during surveillance testing.
DOCKET (2)
NTIAL
08/29/03 -0222 08/29/03 -0222 Main condenser dump valves operated as designed to reduce the reactor coolant system temperature to the no load value of 557 degrees F.
During the transfer of the main condenser dump valve controls (EITS: 01] to the steam pressure control mode, the valves opened for approximately 1 minute causing a plant cooldown to 548 degrees F.
The plant cooldown to 548 degrees F caused the pressurizer level to decrease to approximately 16%. Letdown isolation occurred when the pressurizer level decreased below the isolation setpoint of 17%.
During the letdown isolation, one of the letdown isolation valves, 1NV1A, did not close.
The redundant series valves, 1NV10A and 1NV2A, closed and terminated letdown flow automatically.
Pressurizer level restored above 17%.
Condenser dump valves placed in the steam pressure control mode and operated as designed to control reactor coolant system temperature.
Excess letdown placed in service.
All other plant control systems operated as designed. Plant conditions stabilized at no load conditions.
Required 4-hour notification to the NRC completed.
Pressurizer pressure channel 2 repaired.
Valve 1NV1A repaired and normal letdown placed in service. Excess letdown isolated.
08/29/03 -0223 08/29/03 -0247 08/29/03 -0527 08/29/03 -1904 08/30/03 -0104 NRC FORM 36.6A (1 -Mt) 09/02/03 � Plant cooldown to Mode 5 and reactor coolant loop IA hot leg temperature detectors replaced.
09/09/03-0612 � Unit 1 placed on line.
09/10/03 -1228 � Unit 1 returned to 100% power.
CAUSAL FACTORS
The cause of the reactor trip was failure of the pressurizer pressure loop power supply card.
The cause of the condenser dump valves opening was the failure of the control room operator to adjust the controller correctly prior to transferring to the steam pressure control mode.
Engineering concluded that letdown isolation valve 1NV1A did not receive an automatic close signal because pressurizer level channel 2 remained slightly above the letdown isolation setpoint.
Pressurizer level channel 1 decreased slightly below the isolation setpoint thereby generating a close signal for valves 1NV10A and 1NV2A and abating the decrease in pressurizer level. During troubleshooting of 1NV1A controls, a relay was discovered failed that prevented the operators from manually closing 1NV1A. The relay was repaired prior to plant startup.
CORRECTIVE ACTIONS
Immediate:
1. Plant conditions were stabilized at no-load conditions.
Subsequent:
1. Pressurizer pressure channel 2 was repaired and returned to operable status.
2. Reactor coolant system loop lA hot leg temperature detectors were repaired and returned to operable status.
7 3. Letdown isolation valve 1NV1A relay was repaired.
4. Operator training was completed on transferring the condenser dump valves to the steam pressure control mode.
Planned:
1. Thermography measurements of the reactor protection cabinets will be scheduled on a six month frequency to detect suspect cards prior to failure. This frequency will be evaluated and adjusted based on results. Thermography readings were previously scheduled on a two-year frequency.
The planned corrective actions, as well as any future corrective actions, will be addressed through the Catawba Corrective Action Program. There are no NRC commitments contained in this LER.
SAFETY ANALYSIS
The Solid State Protection System functioned as designed upon receipt of the reactor trip signal. The reactor trip breakers opened within 150 milliseconds of the reactor trip signal as designed. All control rods and shutdown rods inserted as designed.
The rod drop time was within the Technical Specification requirement. Nuclear instrumentation response was normal following the trip.
Reactor coolant system pressure control functioned as expected. No pressurizer power operated relief valves or code safety valves lifted. The pressurizer spray valves and pressurizer heaters controlled primary system pressure as designed. Pressurizer heaters were unavailable for a brief period of time when pressurizer level decreased below 17%. During the event, reactor coolant system pressure ranged from 2096 psig to 2232 psig.
Pressurizer level control functioned as expected with the exception of the failure of letdown isolation valve 1NV1A to close. This failure did not impact plant operations due to the operation of valves 1NV10A and 1NV2A. Pressurizer level control was maintained by normal charging, normal letdown initially, and excess letdown.
Pressurizer level was maintained between 16% and 46%.
7 Reactor coolant temperature control was by the condenser dump valves. No steam generator power operated relief valves or code safety valves lifted. The plant cooldown rate was less than the Technical Specification limit of 100 degrees F per hour. The plant cooldown rate was less than 36 degrees F per hour.
Steam generator level control functioned as expected with makeup by the motor driven auxiliary feedwater pumps. The main feedwater isolation system operated as designed. Steam generator levels were maintained between 23% and 40%.
In summary, the post-trip transient response remained within the bounds of the Updated Final Safety Analysis Report.
The core damage significance of this event has been evaluated quantitatively by considering the following:
Reactor trip initiating event Actual plant configuration and maintenance activities at the time of the trip Credit for recent Nuclear Service Water to Auxiliary Feedwater System modifications to minimize clam intrusion The conditional core damage probability for the event being evaluated is estimated to be 8.3E-08, which is less than the accident sequence precursor threshold of 1.0E-06.
The dominant core damage sequences associated with this event involve loss of secondary side heat removal and failure of feed and bleed cooling. In contrast, the dominant Large Early Release Frequency (LERF) sequences for Catawba involve Inter-system Loss of Coolant Accidents (ISLOCA) and seismic initiated sequences. In addition, the reliability of the important containment safeguards systems (containment spray and hydrogen mitigation) was not impacted by the reactor trip. Therefore, this event is judged not to be significant with respect to the LERF.
This event was of no significance with respect to the health and safety of the public.
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ADDITIONAL INFORMATION
Within the last three years, three other reactor trip events occurred from power operation at Catawba as follows:
trip. The root cause of this event was determined to be an incomplete troubleshooting analysis associated with the main turbine protection system mechanical trip solenoid valve.
reactor coolant flow when the 2D reactor coolant pump 6900 VAC feeder breaker opened in response to protective relay actuation caused by an electrical fault internal to the pump motor.
turbine trip. The turbine trip was due to a steam generator high level. The root cause of this event was determined to be an inadequate understanding of the digital feedwater control system response to a common impulse line hydraulic interaction.
The corrective actions taken in response to these events would not have prevented this latest event from occurring. Therefore, this event was determined to be non-recurring in nature.
Energy Industry Identification System (EIIS) codes are identified in the text as (EIIS: XX). This event is reportable to the Equipment Performance and Information Exchange (EPIX) program.
This event did not involve a Safety System Functional Failure.
There were no releases of radioactive materials, radiation exposure, or personnel injuries associated with this event.
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