Event Notification Report for October 07, 2010

FEEDWATER ISOLATION INADVERTENTLY DISABLED

"At about 1315 hours EDT on October 7, 2010, with the unit in Mode 3 and operators performing recovery actions following a reactor trip that had occurred at 0013 hours (see EN#46313), it was discovered that actions that had been performed to restore the main feedwater system had inadvertently resulted in disabling the feedwater isolation function. The feedwater isolation function, as described in Technical Specifications Section 3.3.2, Table 3.3.2 1, Function 5, requires that the feedwater isolation function be operable in Modes 1, 2, and 3, when the feedwater system is not isolated by the main feedwater isolation valves, main feedwater regulating valves, and bypass valves or by a closed manual valve. At approximately 1018 hours, during actions being taken to restore operation of the main feedwater system, the feedwater isolation key switches for the three steam generators, A, B, and C, were placed in the override/reset position. Although it was not realized at that time, this action was contrary to the Technical Specifications Section 3.3.2 operability requirements for the feedwater isolation function. This inoperability of the feedwater isolation function would have prevented the automatic feedwater isolation function described in the basis of Technical Specifications Section 3.3.2, which states that the primary function of the feedwater isolation signal is to stop excessive flow of feedwater into the steam generators. It also states that this function is necessary to mitigate the effects of overfeeding the steam generators, which could result in overcooling of the primary system. This function is actuated by a safety injection signal. There is no Technical Specifications allowed condition for both trains of the feedwater isolation function to be inoperable. Therefore, Technical Specifications Limiting Condition for Operation (LCO) 3.0.3 was applicable from the time the feedwater isolation switches were placed in the override/reset position, until the feedwater isolation function operability was restored at approximately 1329 hours. The LCO 3.0.3 completion time to be in Mode 4 within 13 hours was not exceeded. This condition is reportable in accordance with 10 CFR 50.72(b)(3)(v)(D), any event or condition that at the time of discovery could have prevented the fulfillment of a safety function."

The licensee notified the NRC Resident Inspector.

DEGRADED FIRE BARRIER

"During a walkdown for the Fire Penetration Seal Project a degraded fire barrier was identified in the wall between the Unit 1 and Unit 2 Normal 480V Switchgear Rooms. The wall is listed as an Appendix R wall between Fire Area (FA) 37 and FA 38. The wall separates redundant safe shutdown cables.

"There is a 2 inch gap where the top of the wall meets the ceiling. The gap is filled with combustible foam - some loose and some stationary. This has been identified as a missing fire barrier such that the required degree of separation for redundant safe shutdown trains is lacking.

"A fire watch has been established as a compensatory measure and will remain in place until the fire barrier is repaired.

"The discovery of this non-compliance is being reported as an unanalyzed condition as defined by 10CFR50.72(b)(3)(ii)(B).

"The licensee has notified the NRC Resident Inspector of this event."

* * * RETRACTION ON 11/15/2010 AT 1328 EST FROM STEPHEN SEILHYMER TO MARK ABRAMOVITZ * * *

"An eight hour report per 10 CFR 50.72(b)(3)(ii)(B) was conservatively reported on October 7, 2010 for a degraded fire barrier between Fire Area (FA) 37 and FA 38. Subsequent engineering analysis determined that the degraded fire barrier maintained the required degree of separation for redundant safe shutdown trains and plant safety was not significantly degraded. The 10 CFR 50.72(b)(3)(ii)(B) report is retracted.

"The NRC Resident [Inspector] has been informed."

Notified the R3DO (Cameron).

LOSS OF 4160 VOLT SAFETY BUS

"Oyster Creek experienced a loss of 4160 volt 'D' safety bus. Emergency diesel generator number 2 fast started and re-powered the 'D' bus, as expected. All other safety systems responded as expected. Investigation into the loss of 'D' 4160 volt safety bus is in progress."

The licensee was performing grid undervoltage testing on the 'D' 4160 at the time of the event. The licensee has staffed the Outage Control Center. There is no affect on the other 4160 volt bus. Offsite power is normal. Plant risk was 'yellow' prior to and following the event as a result of Reactor Building Closed Cooling Water Pump 1-2 out of service for planned maintenance.

The licensee notified the NRC Resident Inspector.

AUTOMATIC REACTOR TRIP DUE TO A ONE LOOP REACTOR COOLANT LOW FLOW

"At 0013 hours EDT on October 7, 2010, with the unit in Mode 1 at 100% power, an automatic reactor trip occurred. The reactor trip signal was based on the One Loop Low Flow reactor protection function. Reactor Coolant Pump (RCP) 'C' tripped during the event.

"The Auxiliary Feedwater System automatically actuated, as expected due to low steam generator water level, and provided feedwater to the steam generators. Steam generator and pressurizer Power Operated Relief Valves (PORVs) and Main Steam Safety Valves did not open during the event. All control rods indicated fully inserted following the reactor trip.

"The normal post-trip electrical lineup is providing power to the plant and the electrical system is stable. Decay heat is being removed via the condenser. The reactor is currently stable in Mode 3.

"The Main Turbine Lube Oil Deluge System actuated without a fire during the event. A fire hose station pipe ruptured after the deluge system actuation. The fire hose station was isolated.

"In addition, the 'B' Main Feedwater (MFW) Pump tripped during the event and the 'A' Main Feedwater pump subsequently tripped due to high steam generator level in the 'C' Steam Generator at about 0023 hours.

"There is currently indication of RCP 'C' Number 2 seal leakage of approximately 2.5 gpm.

"At approximately 0405 EDT the Auxiliary Feedwater (AFW) system actuated due to a trip of MFW Pump 'A' while attempting to start the pump in accordance with procedure GP-004, Post Trip Stabilization. The AFW system actuation signal caused motor-driven AFW Pump 'B' to start. The motor-driven AFW Pump 'A' was already in operation due to the post-trip condition.

"The cause of the MFW Pump 'A' trip is under investigation. This AFW system actuation is being reported in accordance with 10 CFR 50.72(b)(3)(iv)(A).

"The cause of the reactor trip is under investigation."

The licensee notified the NRC Resident Inspector.