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On September 1, 2017, at approximately 1006 Central Daylight Time (CDT), Browns Ferry Nuclear Plant (BFN) Unit 3 3A Residual Heat Removal (RHR) system pump failed to start during performance of Surveillance 3-SR-3.5.1.6 (RHR I), Quarterly RHR System Rated Flow Test Loop I. The apparent cause was the Electrical Preventive Maintenance Instruction for 4kV Wyle/Siemens Horizontal Vacuum Circuit Breaker (Type-3AF) and Compartment Maintenance was revised to include steps to secure the breaker's mounting hardware which caused internal binding of the indication flag. Binding of the indication flag prevented the closing spring of the breaker from charging and the breaker from closing on demand. As a result, automatic start of the 3A RHR pump was prevented. On September 1, 2017, at approximately 1633 CDT, the 3A RHR Pump was declared operable following lubrication and testing of the breaker's indication flag mounting bolt.

A Past Operability Evaluation concluded that the 3A RHR Pump was inoperable from July 26, 2017 to September 1, 2017, which exceeded the Technical Specification allowed outage time. During this time, the 3B, 3C, and 3D RHR pumps would have started automatically upon receipt of an Emergency Core Cooling System (ECCS) initiation signal or from an Operator manual start demand from the Control Room. Based on results from the Probability Risk Assessment and Engineering inspections, there was no significant risk to the health and safety of the public or plant personnel for this event. The Corrective Action to reduce the probability of similar events occurring in the future will be addressed by revising the Electrical Preventive Maintenance Instruction for 4kV Wyle/Siemens breakers to ensure freedom of movement of the indication flag is present during the breaker inspection.

On January 30, 2017, during routine surveillance testing of the Unit 2 Division 3 Diesel Generator Cooling Water (DGCW) system, the cooling water strainer backwash valve was unable to open. The Division 3 DGCW system was declared inoperable. Upon investigation, operators determined the cause of the valve malfunction was due to stem-disc separation. Division 3 DGCW is a support system for the Division 3 Emergency Diesel Generator and the High Pressure Core Spray (HPCS) system. The required actions of Technical Specifications (TS) 3.7.2 and 3.5.1 were entered on January 30, 2017 when the DGCW and HPCS system, respectively, were determined to be inoperable. TS 3.7.2 Required Action (RA) A.1 requires the supported system to be immediately declared inoperable. TS 3.5.1 RA B.2 requires restoration of the HPCS system to operable within 14 days. TS 3.8.1 was not applicable since a note provides that Division 3 AC electrical power sources are not required to be operable when HPCS is inoperable. The valve was replaced, and the HPCS system was returned to operable on February 2, 2017.

This condition could have prevented the HPCS system, a single train safety system, from performing its design function. This event is reportable in accordance with 10 CFR 50.73(a)(2)(v)(D) as an event or condition that could have prevented fulfillment of the safety function of structures or system that are needed to mitigate the consequences of an accident. There were minimal safety consequences associated with the event since the other emergency safety systems remained operable, and the Division 3 DGCW system remained functional as it retained the ability to provide the required flow through the system. The apparent cause of the stem-disc separation was erosion due to the carbon-steel valve internals in a raw water system environment.

On June 17, 2015, at approximately 1015 Central Daylight Time (CDT), Browns Ferry Nuclear Plant (BFN) Operations personnel attempted to place BFN, Unit 2, Residual Heat Removal (RHR) Loop 1 into Suppression Pool Cooling (SPC). Upon actuating Hand Switch 2-HS-74-5A from the Control Room, Operations personnel observed that 2A RHR pump failed to start, and declared RHR Loop I inoperable.

On June 18, 2015, at 1710 CDT, the pump was started and stopped after completion of troubleshooting, and the RHR Loop I was declared operable.

Troubleshooting discovered a loose terminal wire which intermittently prevented 2A RHR from being manually started from the Control Room. An investigation determined the cause to be human performance errors. RHR Loop 1 was inoperable from March 20, 2015, to June 18, 2015, longer than allowed by BFN, Unit 2, Technical Specifications. This loose terminal wire made the pump vulnerable to failure during a seismic event and, therefore, not in compliance with the design basis for the RHR system. This event did not prevent the 2A RHR Pump from manually starting, from the pump breaker, for SPC in the event of an emergency. Both the SPC manual start and the automatic start response to an Emergency Core Cooling System initiation signal were unaffected by this condition.

Corrective Actions for this event were to discipline the individuals responsible, to tighten the loose fastener, and to revise maintenance instructions to reduce the probability of recurrence.

On February 11, 2015, at 0820 Central Standard Time, Brown's Ferry Nuclear Plant (BFN), Unit 3, declared the High Pressure Coolant Injection (HPCI) and Reactor Core Isolation Cooling (RCIC) systems inoperable due to no suction source aligned. During surveillance testing, the Condensate Storage Tank (CST) emergency discharge isolation valve energized and closed when the breaker was closed, isolating both systems from their suction source. It was subsequently determined that contacts on the local hand switch were stuck closed following performance of a previous maintenance task. Operations personnel re-opened the isolation valve using the hand switch in the Control Room, restoring operability to the HPCI and RCIC systems.

The apparent cause of this event was inadequate design review of a 2010 plant modification which allowed latent design vulnerabilities to be introduced into the plant.

The corrective actions to reduce the probability of a similar event occurring in the future were to remove thermal overload heaters from the affected breakers, preventing valve closure when these breakers are closed; to review a sample of recent engineering change packages for quality of Design Review; to repair a faulty hand switch; and to implement a design change for the CST isolation valves for all three BFN units to prevent spurious operation of the isolation valve when the associated breaker is closed.

On March 29, 2014, Unit 1 was in Mode 1 at 100% power. At 1620 hours CDT, the Division 3 Core Standby Cooling System (CSCS) Pump Room, Switchgear Room, and Battery Room Ventilation failed in such a manner that heat could not be removed from the rooms. Due to the lack of ventilation in the Division 3 switchgear room the High Pressure Core Spray (HPCS) system was declared inoperable and Condition B of Technical Specification (TS) 3.5.1 was entered.

The cause of the event was a failure of the hydramotor pump bearing for the 1VD19Y, Division 3 CSCS Ventilation Return Fan Outlet Damper. Loss of hydraulic pressure in the hydramotor resulted in 1VD19Y failing in the closed position. The corrective action for the event was replacement of the hydramotor for the 1VD19Y damper.

On Thursday, August 22, 2013 at 0755 (EDT) with the reactor critical at 98% core thermal power (CTP), and the mode switch in RUN, the Pilgrim Nuclear Power Station (PNPS) was manually scrammed due to lowering reactor water level resulting from a trip of the reactor feed pumps. The reactor feed pumps tripped due to a loss of power to the pump seal cooling water flow switch relays and resultant automatic actuation of the feed pump trip circuit.

The direct cause of the reactor feed pump trip was an automatic actuation of the feed pump trip circuitry. The feed pump trip circuits actuated as designed in response to a loss of power to the pump seal cooling water flow switch relays. The seal cooling water flow switch relays lost power as a result of a 120V AC breaker trip resulting from a short-to-ground fault in the associated circuits fed by the breaker. Corrective action was taken to repair the ground fault in the associated circuits fed by the 120V AC breaker and to revise reactor feed pump trip circuit design to remove the loss of seal water trip function.

This event posed no threat to public health and safety.

On November 30, 2011, at 1747 hours, with the reactor at 100% core thermal power and steady state conditions, Pilgrim Nuclear Power Station (PNPS) declared the High Pressure Coolant Injection (HPCI) system inoperable due to the HPCI turbine governor control valve (HO-2301-24) failing to open during planned post-maintenance testing. The governor control valve is a hydraulically operated valve and its normal position is closed. The valve has a safety function to open on a demand signal during certain event mitigation scenarios requiring the HPCI system operation.

The governor control valve failed to open during the post-maintenance testing from the Alternate Shutdown Panel (ASP) and subsequently from the Main Control Room. The HPCI system was declared inoperable and the Limiting Condition for Operation (LCO) for Technical Specification (TS) 3.5.C.2 was entered.

This event had no impact on the health_ and/ or safety of the public.

During the 2010 refueling outage, the pneumatic actuators for the four main steam safety relief valves (RV), RV- 2-71-A, B, C & D, were tested and leakage was identified through the shaft to piston thread seal on three of the four RVs. This leakage, when combined with the RV accumulator leakage, caused two of the four RVs to not meet design actuation requirements and therefore be considered inoperable. Technical Specification (TS) 3.6.D requires at least three of the four RVs to be operable for overpressure protection of the Reactor Coolant System and TS 3.5.F requires all four RVs to be operable to support the Automatic Depressurization System (ADS) function of the Core Standby Cooling System. Since inoperability of two valves could constitute an operation or condition prohibited by TS, a cause analysis was performed. On October 25, 2010, based on a review of the analysis, it was determined that there was firm evidence that the condition may have existed for a period of time greater than allowed by the TS. Therefore, this event is reportable in accordance with 10CFR50.73(a)(2)(i)(B) as an operation or condition prohibited by TS. Subsequent material testing of a seal from the same batch lot determined that the apparent cause of the thread seal condition was thermal degradation. The thread seals were replaced and tested on all four RVs prior to start up from the 2010 outage with new seals and will be disassembled for inspection, modified and tested during the next refueling outage.

This event did not affect the automatic function of the RVs to provide overpressure protection and due to redundancy in the RV design and available nitrogen supplies, adequate relief capacity for ADS existed at all times. Therefore, this event did not pose a threat to public health and safety.

On April 16, 2005 at 0955 hours, one drywell pressure transmitter was found isolated (valved out of service) and therefore, inoperable. The transmitter is part of one of two Division 'B' instrument channels that function to initiate safety systems in the event of a high-pressure condition inside the drywell. The transmitter was mistakenly isolated at 0905 hours on April 14, 2005. The Division 'A' pressure transmitters and related instrument channels that provide a similar function were not impacted. The condition existed during reactor power operation for a period longer than allowed by technical specifications.

The root cause of the condition was licensed operator error. The operator mistakenly closed the isolation valve of the pressure transmitter instead of another valve. The error was the result of inadequate use of human performance error prevention tools, i.e. self-checking and a questioning attitude. Corrective actions taken included opening the isolation valve and walk-downs to ensure other instrumentation was correctly configured.

The condition posed no threat to public health and safety.

NRC FORM 368

On September 6, 2003 at 1151 hours, the safety-related train 'A' 480-volt load center and the related motor control centers (MCCs) powered from the load center de-energized. The event resulted in several systems including the high pressure coolant injection and the reactor core isolation cooling systems becoming inoperable. The train '8' 480-volt load center and related MCCs were unaffected.

The affected load center and related MCCs were re-energized by 0022 hours and the affected systems were returned to service by 1000 hours on September 7, 2003.

The direct cause of the event was the unplanned trip of the circuit breaker that powers the load center.

The root cause investigation revealed the circuit breaker tripped due to a malfunction of one of the breaker's three current transformers.

Corrective action included the replacement of the circuit breaker. Corrective action planned includes analysis of the current transformer malfunction and related actions to preclude recurrence.

The event posed no threat to public health and safety.

On July 7, 2002, the High Pressure Coolant Injection (HPCI) system was declared inoperable for 30 minutes.

The cause was a failed fuse that is part of the control power circuitry for the circuit breaker of a normally closed motor-operated valve in the HPCI system injection piping. Inspection of the failed fuse identified a separation that had occurred at an internal solder connection between the fuse end cap and the fusible link. The fusible link was intact. The fuse was replaced and the system was returned to operable status on July 7, 2002.

The event occurred at 100 percent reactor power with the reactor mode selector switch in the RUN position. The reactor vessel pressure was approximately 1035 psig with the water temperature at the saturation temperature for that pressure. This event posed no threat to public health and safety.