05000446/LER-2002-001
| Event date: | |
|---|---|
| Report date: | |
| Reporting criterion: | 10 CFR 50.73(a)(2)(iv)(A), System Actuation |
| 4462002001R00 - NRC Website | |
I. DESCRIPTION OF THE REPORTABLE EVENT
A. REPORTABLE EVENT CLASSIFICATION
Any event or condition that resulted in manual or automatic actuation of the Reactor Protection System (RPS) including reactor trip or reactor scram.
B. PLANT OPERATING CONDITIONS PRIOR TO THE EVENT
On June 6, 2002, CPSES Unit 2 was in Mode 1, Power Operation, at 100 percent power.
C. STATUS OF STRUCTURES, SYSTEMS, OR COMPONENTS THAT WERE
INOPERABLE AT THE START OF THE EVENT AND THAT CONTRIBUTED TO
THE EVENT
There were no structures, systems, or components that were inoperable at the start of the event that contributed to the event.
D. NARRATIVE SUMMARY OF THE EVENT, INCLUDING DATES AND
APPROXIMATE TIMES
On June 6, 2002, CPSES Unit 2 was in Mode 1, Power Operation, operating at 100 percent power. At 1858 hours0.0215 days <br />0.516 hours <br />0.00307 weeks <br />7.06969e-4 months <br />, Operators (utility, licensed) in the Unit 2 Control Room received a "Generator Electronic Protection Failure Alarm." Maintenance personnel (utility, non- licensed) investigated and found faults indicated in Electronic Generator Protection cabinet 2- JC52 (HIS: (CAB)), which contains the Channel 1 Primary Water Pump (EIIS:(P)) 2-01 Shaft Vibration Detector (EIIS:(DET)). The Primary Water Pump Shaft Vibration trip uses a 2 out of 2 channel logic with a trip setpoint of 6.3 volts. Voltage readings were taken on both vibration detector channels (EITS:(CHA)), and the voltages measured on both channels were normal (3.6 volts and 3.5 volts), indicating that there was not an actual problem with pump vibration.
At 1910 hours0.0221 days <br />0.531 hours <br />0.00316 weeks <br />7.26755e-4 months <br />, the Maintenance personnel returned to cabinet 2-JC52 and found that the conditions had changed. Channel 1 at that time was showing additional alarms and the voltage measured was zero. Maintenance and Operations personnel were evaluating the next course of action when, at 1924 hours0.0223 days <br />0.534 hours <br />0.00318 weeks <br />7.32082e-4 months <br />, a "Turbine Generator Channel 1 Primary Water Pump Vibration High" signal was received. This resulted in a Unit 2 Turbine Generator (EllS:(TG)) trip which caused a trip of the Unit 2 Main Turbine (EIIS:(TRB)). The Unit 2 reactor (EIIS:(RCT)) automatically tripped on a "Turbine Trip >50% Power" signal. All control rods (ELLS: (AA)) fully inserted, and all Auxiliary Feedwater (EIIS:(BA)) pumps started automatically on Steam Generator (EIIS:(SG)) Lo-Lo level.
E. THE METHOD OF DISCOVERY OF EACH COMPONENT OR SYSTEM FAILURE,
OR PROCEDURAL OR PERSONNEL ERROR
Operators (utility, licensed) in the Unit 2 Control Room received a "Turbine Trip >50% Power" alarm.
II. COMPONENT OR SYSTEM FAILURES
A. FAILURE MODE, MECHANISM, AND EFFECTS OF EACH FAILED COMPONENT
A circuit breaker (EIIS:(BKR)) in Vibration Expansion Measuring cabinet 2-JMO3 (EIS:
(CAB)) failed open during this event. The failure mode, mechanism, and effects of the circuit breaker failure have not yet been determined. The breaker failure is being analyzed at a testing laboratory.
B. CAUSE OF EACH COMPONENT OR SYSTEM FAILURE
The cause of the circuit breaker failure has not yet been determined. The cause of the breaker failure is being determined through failure analysis techniques at a testing laboratory.
C. SYSTEMS OR SECONDARY FUNCTIONS THAT WERE AFFECTED BY FAILURE
OF COMPONENTS WITH MULTIPLE FUNCTIONS
Not applicable — No failures of components with multiple functions have been identified.
D. FAILED COMPONENT INFORMATION
Siemens Power Group Model Number 340/032 Circuit Breaker
III. ANALYSIS OF THE EVENT
A. SAFETY SYSTEM RESPONSES THAT OCCURRED
The Reactor Protection System and The Auxiliary Feedwater System actuated during the event.
The Unit 2 reactor automatically tripped on a "Turbine Trip >50% Power" signal, and all three Auxiliary Feedwater pumps automatically started on "Steam Generator Lo-Lo water level" signals.
B. DURATION OF SAFETY SYSTEM TRAIN INOPERABILITY
Not applicable - no safety system train was deemed inoperable.
C. SAFETY CONSEQUENCES AND IMPLICATIONS OF THE EVENT
This event is specifically bounded by the Final Safety Analysis Report (FSAR) accident analysis of the turbine trip presented in Section 15.2.3 of the CPSES FSAR. The analysis uses conservative assumptions to demonstrate the capability of pressure relieving devices and to demonstrate core protection margins. The event of June 6, 2002, occurred at 100 percent reactor power, and all systems and components functioned as designed. There were no safety system functional failures associated with this event.
Based on the above, it is concluded that the event of June 6, 2002, did not adversely affect the safe operation of CPSES Unit 2 or the health and safety of the public.
IV. CAUSE OF THE EVENT
TXU Energy believes that a circuit breaker (EIIS:(BKR)) in Vibration Expansion Measuring cabinet 2-JMO3 (EIIS: (CAB)) failed open which caused a spurious "Turbine Generator Channel 1 Primary Water Pump Vibration High" signal. Post trip troubleshooting using a replacement breaker revealed that when the channel 1 vibration detector circuit breaker is opened a voltage spike enters the channel Vibration Detector circuit.
V. CORRECTIVE ACTIONS
The failed breaker has been replaced. The cause of the breaker failure is being determined through failure analysis techniques at a testing laboratory. Maintenance personnel received instruction related to blocking the Primary Water Pump vibration detector trip prior to working on the circuit, and Operations personnel were provided information on the trip via a lessons learned.
Engineering will evaluate other turbine/generator protection cabinets to determine if similar circuit designs exist. Digital upgrades planned for this equipment eliminate this circuitry thereby minimizing the probability of a single component failure generating a Turbine Generator trip signal. The Turbine Generator vendor (Siemens) has provided potential equipment modifications which could be implemented prior to the digital upgrade.
VI. PREVIOUS SIMILAR EVENTS
There have been other previous events which resulted in a turbine trip followed by an automatic reactor trip. However, the causes of those events were sufficiently different such that the corrective actions taken for the previous events would not have prevented this event.