05000281/LER-2003-001
| Event date: | 01-25-2003 |
|---|---|
| Report date: | 03-26-2003 |
| Reporting criterion: | 10 CFR 50.73(a)(2)(iv)(A), System Actuation |
| 2812003001R00 - NRC Website | |
FACILITY NAME (1)
DOCKET
05000 - 280 05000 - 281 LER NUMBER (6) PAGE (3) 2 � OF 4 1.0 DESCRIPTION OF THE EVENT The Surry Main Generator Protection System provides fault protection for the main generator leads and main transformer by interrupting electrical flow if an electrical fault exists. A current transformer (CT) provides secondary current to a differential relay that provides this fault protection.
On January 25, 2003 at 1414 hours0.0164 days <br />0.393 hours <br />0.00234 weeks <br />5.38027e-4 months <br />, with Surry Power Station Unit 2 at 100% reactor power, an automatic reactor trip was generated due to a main turbine generator trip. The main generator tripped due to a Main Transformer and Generator Leads differential lockout. All three auxiliary feedwater pumps automatically initiated as designed on low low steam generator level following the trip. All control rod bottom lights were lit, however, Individual Rod Position Indications (IRPIs) for three control rods indicated between 11 and 20 steps following the trip. In accordance with emergency operating procedures, emergency boration was initiated at 1434 hours0.0166 days <br />0.398 hours <br />0.00237 weeks <br />5.45637e-4 months <br /> and secured at 1440 hours0.0167 days <br />0.4 hours <br />0.00238 weeks <br />5.4792e-4 months <br />, followed by normal boration to ensure adequate shutdown margin. The three IRPIs that initially did not indicate zero position drifted to less than 10 steps by 1501 hours0.0174 days <br />0.417 hours <br />0.00248 weeks <br />5.711305e-4 months <br />. Boron concentration shutdown margin for Unit 2 was determined to be satisfactory at 1645 hours0.019 days <br />0.457 hours <br />0.00272 weeks <br />6.259225e-4 months <br />.
A load shed feature is provided to reduce electrical loading in the event two units were simultaneously loaded on the Reserve Station Service Transformers (RSSTs). This feature ensures that the voltages on the emergency busses will be within acceptable electrical loads were provided by the RSSTs. When Unit 2 tripped, loads automatically transferred to the RSSTs and load shedding was initiated on Unit 1. As a result, the Unit 1 'B' Main Feed Pump tripped, and since the Unit 1 'A' Main Feed Pump had previously been shut down, a start signal was initiated to both Unit 1 Motor Driven Auxiliary Feedwater Pumps (MDAFWPs) at approximately 1414 hours0.0164 days <br />0.393 hours <br />0.00234 weeks <br />5.38027e-4 months <br />. The Unit 1 'A' Main Feed Pump, was restarted at 1623 hours0.0188 days <br />0.451 hours <br />0.00268 weeks <br />6.175515e-4 months <br />, and the Motor Driven Auxiliary Feedwater Pumps were secured at 1653 hours0.0191 days <br />0.459 hours <br />0.00273 weeks <br />6.289665e-4 months <br />.
At 1725 hours0.02 days <br />0.479 hours <br />0.00285 weeks <br />6.563625e-4 months <br />, a four-hour and an eight-hour non-emergency report was made to the NRC as required by 10 CFR 50.72(b)(2)(iv)(B) and 10 CFR 50.72(b)(3)(iv)(A), respectively. The automatic actuation of the Unit 2 reactor protection system and the initiation of Unit 2 Auxiliary Feedwater (AFW) system are reportable in accordance with 10 CFR 50.73(a)(2)(iv)(A). The automatic actuation of the Unit 1 AFW system is also reportable in accordance with 10 CFR 50.73(a)(2)(iv)(A).
2.0 SIGNIFICANT SAFETY CONSEQUENCES AND IMPLICATIONS This event resulted in no significant safety consequences or implications. Emergency systems functioned as required for the Unit 2 trip. All three AFW pumps automatically initiated as designed on low low steam generator level following the trip. The operating 4,.. , ■ i , 9 N.
FACILITY NAME (1)
DOCKET
05000 - 280 05000 - 281 LER NUMBER 6) PAGE (3) 1 crew noted the three IRPI indications and in accordance with emergency procedures, initiated boration to ensure adequate shutdown margin. All electrical busses transferred properly following the trip and all emergency diesel generators were operable. The Reactor Coolant System (RCS) [EllS-AB] cooled to a minimum average temperature (Tave) of approximately 540 degrees Fahrenheit (F) and then stabilized to the no load Tave value of 547 degrees F. No indications of primary to secondary leakage existed.
Emergency systems functioned as required for the Unit 1 load shed. The actuation initiated flow from the two motor driven auxiliary feed pumps. In addition, the cross- connect from Unit 2 AFW system remained operable. Therefore, the health and safety of the public were not affected.
3.0 CAUSE The Unit 2 automatic reactor trip was caused by a main generator trip. The main generator tripped due to a Main Transformer and Generator Leads differential lockout. It was observed by walkdown that a conduit fitting had separated from a junction box containing the wiring for the differential lockout CT secondary leads. A 'B' phase secondary lead became shorted when the associated conduit disconnected from the junction box [EDS- TB, JBX] insulated bushing and the conduit became supported by the CT leads. The preliminary cause was the detachment of the insulated bushing locking collar allowing the insulated bushing and conduit connection to loosen, and ultimately disconnect. The inherent vibrations from the main turbine generator existing over a long period of time contributed to the failure.
The cause of the Unit 1 automatic start of the two MDAFWPs was the load shed feature designed to reduce electrical loading on the RSSTs. At the time of the Unit 2 reactor trip, Unit 1 was off-line. Load shedding tripped the operating Unit 1 Main Feed Pump and initiated a start signal for the Unit 1 MDAFWPs.
4.0 IMMEDIATE CORRECTIVE ACTION(S) The CT secondary leads were repaired and tested satisfactorily. The insulating bushing and conduit were repaired. Other Unit 2 insulating bushings were examined and other conduits containing protective relay wiring were identified as having degraded bushings.
The insulating bushings and wiring were repaired and tested satisfactorily.
5.0 ADDITIONAL CORRECTIVE ACTIONS Similar protective relay conduits and insulated bushings on Unit 1 were inspected and one insulated bushing was still attached to the conduit and was providing adequate support and protection for the internal CT wires. The insulating bushing was repaired.
FACILITY NAME (1)
DOCKET
05000 - 280 05000 - 281 LER NUMBER (6) 6.0 ACTIONS TO PREVENT RECURRENCE A Root Cause Evaluation (RCE) was initiated to determine the cause of the reactor trip due to generator differential. Conclusions from the RCE will be evaluated and the approved recommendations from the RCE necessary to prevent recurrence will be implemented through the corrective action program.
7.0 SIMILAR EVENTS None.
8.0 MANUFACTURER/MODEL NUMBER The insulated bushing was a Seimens Westinghouse Manufacturing part number 57D2226G01.
9.0 ADDITIONAL INFORMATION None.