05000286/LER-2003-003
| Indian Point Unit 3 | |
| Event date: | |
|---|---|
| Report date: | |
| Reporting criterion: | 10 CFR 50.73(a)(2)(iv)(A), System Actuation |
| Initial Reporting | |
| 2862003003R00 - NRC Website | |
DESCRIPTION OF EVENT
Vote: The Energy Industry Identification System Codes are identified within brackets ().
On June 22, 2003, at approximately 1742 hours0.0202 days <br />0.484 hours <br />0.00288 weeks <br />6.62831e-4 months <br />, following completion of maintenance activities on the 345kv Buchanan distribution system (FK), Indian Point Unit 3 experienced an automatic turbine / reactor trip event (JE). Prior to this event, Indian Point Unit 3 was operating in Mode 1, at 100 percent, steady-state power conditions. Generator output breaker No. 3 (BKR) had been opened earlier at approximately 1654 hours0.0191 days <br />0.459 hours <br />0.00273 weeks <br />6.29347e-4 months <br /> to support maintenance activities on 345kV offsite feeder W-98 (FDR). Indian Point Unit 3 remained in service utilizing generator output breaker No. 1. At approximately 1740 hours0.0201 days <br />0.483 hours <br />0.00288 weeks <br />6.6207e-4 months <br />, upon restoration of feeder W-98, a request was received from the Con Edison District Operator to close generator output breaker No 3. While attempting to close breaker No. 3, a fault occurred which resulted in the tripping of both generator output breakers No. 1 and 3. A main turbine trip occurred as a result of the actuation of the main generator primary (86P) and backup (86BU) lockout relays (86). The main generator trip actuation was caused by the actuation of electrical protective relaying direct trip circuitry from the 345kV (PK) main output breakers (BKR). The cause for this event was attributed to a ground fault on Phase B of generator output breaker No. 3. Control room personnel immediately entered Emergency Operating Procedure (SOP) E-0, "Reactor Trip or Safety Injection" to initiate plant recovery. Subsequently, operators transitioned to ES-0.1, 'Reactor Trip Response,' and then to POP-3.1, 'Plant Shutdown From 45 Percent Power.' The plant was stabilized in the hot shutdown condition and the transient terminated. Subsequently, the plant was brought to cold shutdown conditions at approximately 1304 hours0.0151 days <br />0.362 hours <br />0.00216 weeks <br />4.96172e-4 months <br /> on June 23, 2003. Station offsite power was maintained throughout this event. There was no automatic start of the Emergency Diesel Generators (EK). All control rods (AA) fully inserted. The Auxiliary Feedwater System (BA) pumps automatically started. In accordance with 10 CFR 50.72(b)(2)(iv)(B), the NRC was notified (Event Number 39955) of this event at 2118 hours0.0245 days <br />0.588 hours <br />0.0035 weeks <br />8.05899e-4 months <br /> on June 22, 2003. This event was entered into the Entergy Corrective Action Program under CR-1P3-2003-03809. A post transient evaluation was performed (No. 03-04) on June 23, 2003. This evaluation did not identify any personnel or procedural issues during the event.
Upon completion of plant restoration activities, Indian Point Unit 3 was re- started and synchronized to the grid on June 27, 2003 utilizing Breaker No. 1 only.
Generator output breaker No. 3 is a 345kV, Type 345GA25-30, manufactured by ITS Imperial Corporation. There are six sets of contacts (fixed and moving) mounted inside each tank. The contacts are enclosed in a high gas pressure chamber to quench the arc during contact disengagement between stationary and moving contacts.
NRC FORM SIM (1-2001) NRCFORNI3136AU.S.NUCLEARREGULATORYCOMMISSION 04wil FACILITY NAME (1) LER NUMBER (6) PAGE (3) DOCKETrZ
CAUSE OF EVENT
This event was initiated by the failure of Breaker No. 3, Phase B, and was a result of compromised dielectric. It is believed that moisture and/or a contaminated pull rod assembly compromised the SF6 dielectric gas. The compromised dielectric reduced the breaker's capability to withstand a heavy voltage surge during breaker closure. The closure of breaker No. 3 and subsequent re-closure of Millwood breaker No. 18 generated a larger than usual voltage spike. This resulted in an arc to ground along the bottom of a defective pull rod assembly in lower support insulator No. 3 causing the lower support assembly to fail. The apparent cause for this event is human performance-related, and is attributed to limited in-house knowledge of the maintenance and operational requirements of the ITE Imperial Corporation breaker design. As a result, a heavy reliance on the expertise provided by outside contractors to perform activities such as breaker repair or refurbishment existed. A contributing factor was inadequate supervision and oversight of contractor activities, primarily following recent breaker refurbishment activities.
CORRECTIVE ACTIONS
The following corrective actions have been or will be performed to prevent a recurrence of this event.
1. The immediate corrective actions taken prior to plant restart were to assess the damage to breaker No. 3, re-verify the capability of breaker No. 1 to support 100 percent plant power operation, and restore those Balance of Plant components which did not operate as expected following the reactor trip event. Various equipment repairs were performed prior to the plant's return to service on June 27, 2003. (Completed) 2. The dielectric moisture level in generator output breaker No. 1 was tested and found to be within acceptable limits. (completed) 3. Breaker No. 3 will be replaced with newer designed breaker (GE-Hitachi HVB). (Estimated Completion Date: August 2003) 4. The Training Department will perform a site wide needs analysis to determine the appropriate level of training and qualification needed for personnel to provide supervisory oversight of contractor/vendor personnel performing work on plant equipment. (Estimated Completion Date: December 2003)
EVENT REPORTING
This report is submitted pursuant to 10 CFR 50.73(a)(2)(iv)(A) as condition that resulted in a manual or automatic actuation of the protection system. In accordance with 10 CFR 50.72(b)(2)(iv)(B), notified (Event Number 39955) of this event at 2118 hours0.0245 days <br />0.588 hours <br />0.0035 weeks <br />8.05899e-4 months <br /> on June
PAST SIMILAR EVENTS
an event or reactor the NRC was 22, 2003.
A review of previous Licensee Event Reports generated within the past three years, which involved a reactor trip caused by a Buchanan Switchyard related fault identified two events.
power, an automatic reactor trip occurred. Main generator output breaker No. 3 faulted resulting in breakers Nos. 1, 3, and 6 opening and electrically isolating the plant output. The apparent cause for the phase-to-ground fault in breaker Na. 3 was high resistance at the breaker contact surfaces. The high resistances at the breaker contacts were a result of breaker contact misalignment during previous maintenance performed during the 2001 refueling outage. Subsequent inadequate corrective actions associated with maintenance performed on breaker 3 in December 2002 contributed to the June 22, 2003 failure. In general, a limited knowledge of these breakers and the difficulty in adequate oversight of contractors due to this limited knowledge were identified as factors.
power, an automatic reactor trip occurred as a result of a main turbine trip.
The cause for the main turbine trip was a direct trip from the 345kV Substation.
Degraded electrical cable insulation between several conductors in the direct trip circuitry within underground cable between the plant and the 345kV Substation was the probable cause. The corrective actions for that event did not prevent this event because the cause was different. This event was due to a faulted generator output breaker and was not related to faults in the direct trip cable.
NRC NNW 311004 4001) Fac IUTY NAME (1) PAGE (3) DOCKET (2) LER NUMBER (6) BEQUEN11AL REVISION Indian Point Unit 3 05000-286 A 2003 - 03 -00 50F5 NRCFORM366AU.S.NUCLEARREGULATORYCOMM1SSION (1-nci)
EVENT SAFETY SIGNIFICANCE
During this event, the turbine received a direct trip signal from the Buchanan Switchyard. The turbine trip caused a reactor trip due to signals derived from the turbine auto stop oil pressure switches since the plant was above P-8 (approximately 35t power). There was no loss of offsite power.
There were no safety significant consequences during this event. A loss of external electrical load and/or turbine trip event is described in the Indian Point Unit 3 FSAR Section 14.1.8. As discussed in the MAR, the results of the analyses performed for a total lose of external electrical load without a direct or immediate reactor trip from full power conditions show that the plant design is such that there would be no challenge to the integrity of the reactor coolant system or the main steam system. Pressure relieving devices incorporated in the design of the plant would be adequate to limit the maximum pressures to within the design limits. In addition, the integrity of the core would be maintained by operation of the reactor protection system; i.e., the DNBR would be maintained above the safety analysis limit value. Thus, no core safety limit would be violated. Furthermore, these results, in conjunction with the results for the complete loss of flow event from full power, bound the results for a complete loss of load from Set power without a direct reactor trip on turbine trip. In this event, offsite power was not lost and the reactor did trip immediately. Therefore the Chapter 14 analysis bounds this event.