Semantic search

On December 9, 2017 at 1347 CDT the Main Control Room received annunciators that indicated a trip of a 4160V 1A1 Breaker, the 480V transformer 1A and Al feed breaker. The loss of Division 1 480V power caused the instrument air (IA) containment isolation valves to fail close as designed. The loss of IA affected various containment loads, including the scram pilot air header and containment isolation valves. Another consequence of this event was that secondary containment differential pressure became positive due to fuel building ventilation dampers failing closed by design due to the loss of power. Operations entered Emergency Operating Procedure (EOP) -08, Secondary Containment Control, and Technical Specification (TS) Limiting Condition for Operation (LCO), 3.6.4.1 Action A.1. Division 2 Standby Gas Treatment System was activated at 1350 and restored secondary containment differential pressure within allowable TS values at 1351. The TS LCO and EOP were exited when allowable TS values were restored. Due to the loss of IA, a manual reactor scram was inserted at 1353 when two control rods began drifting in as expected.

A phase to ground fault was identified on 480V transformer 1A (1AP11E). On December 14, the 480V transformer was replaced and the plant returned to Mode 1 operations on December 15. The condition described in this report was determined to be reportable under 10 CFR50.73(a)(2)(iv)(A), 10 CFR 50.73(a)(2)(v)(C) and 10 CFR 50.73 (a)(2)(ii)(B). The cause of the transformer failure is currently under investigation and will be provided in a supplemental report. This event is classified as an unplanned scram with complications due to the loss of the Division 1 480V power.

On June 2, 2017, plant personnel were performing welding activities on the 'B' Reactor Water Cleanup System (RT) pump while the plant was in operating Mode 4. The Operations Work Control Supervisor (WCS) had given maintenance personnel authorization to prop both 'B' RT pump room doors open so that welding cables could extend through both doors in support of maintenance activities. The doors are part of the plant secondary containment boundary. Authorization granted by the WCS was executed without utilizing the plant barrier impairment process (PBI) per plant procedures. Prior to the plant's transition to operating Mode 2, Operations personnel made a plant announcement requiring the establishment of primary and secondary containment. The plant then transitioned to operating Mode 2 at 0241. However, both RT Pump 'B' doors remained propped open during the plant mode change.

This condition was discovered by shift personnel at 0300 (CDT) and secondary containment was declared inoperable.

The loss of secondary containment was caused by personnel not following the FBI per plant procedure. Secondary containment was subsequently restored twenty four minutes after discovery of the open 'B' RT doors. Corrective actions taken and planned include implementing management action response checklist (MARC) principles for the responsible supervisor and completing a read and sign for planners and schedulers associated with the PBI process.

On December 5, 2016 at 2011 hours during a surveillance test of the Residual Heat Removal (RHR) "C" pump, a Westinghouse DHP Breaker for the RHR "C" pump failed to close. This occurrence resulted in a failure of RHR "C" pump to start. The pump was declared INOPERABLE and Technical Specification (TS) Limiting Condition of Operation (LCO) 3.5.1, "Emergency Core Cooling System -Operating," Required Action A.1 was entered to restore the pump to OPERABLE status in seven days. Performance records indicate the breaker was installed on March 11, 2016 in the RHR "C" pump cubicle and successfully operated on March 11, June 9, and September 6, 2016, prior to failure.

An investigation determined that the breaker latch check switch contacts were not closed and that the latch check switch setting was out of procedural limits. The most likely time that the switch would have become out of adjustment would have been during transport of the breaker to the RHR "C" pump breaker cubicle on March 11, 2016. The apparent cause of the failure was an inadequate verification of the latch check switch setting performed prior to installation of the breaker in the cubicle. Corrective measures have been initiated that include revising plant procedures to record latch check switch adjustment value. The condition described in this report is reportable under 10CFR50.73(a)(2)(i)(B) as an "operation or condition which was prohibited by the plant's Technical Specifications.

On 12/8/13 at 2026 hours with the plant in Mode 1 at 97.3 percent reactor power, operators received multiple alarms due to the trip of 4160 volt 1A1 breaker which resulted in a loss of power to two Division 1 480 volt unit substations.

Operators were immediately dispatched and found a 4160/480 volt stepdown transformer failed. Many Division I components lost power. The loss of power caused an instrument air (IA) containment isolation. The loss of IA affected various containment loads, including the scram pilot air header, the main steam isolation valves and the reactor water cleanup system. At 2036 hours, the scram pilot air header low pressure alarm was received, and in response to an anticipated automatic reactor scram, operators immediately initiated a manual reactor scram. All control rods fully inserted into the core. Reactor pressure vessel water level dropped to the low reactor water level 3 setpoint (normal result of a scram from high power) and operators entered the Reactor Pressure Vessel Control Emergency Operating Procedure. The most probable cause of the transformer failure was a tum to turn failure of the high side windings due to insulation breakdown over time, prior to its expected end of life. An installed spare was connected to replace the failed Division 1 transformer.

On March 24, 2016, it was determined that placing both Reactor Water Cleanup System (RT) Leak Detection System (LD) bypass switches in the Bypass position per plant procedure when the RT Filter/Demineralizer (F/D) was placed in service following backwash and pre-coat operations on- January 25, 2016 was a reportable condition. Both divisions of the RT LD were bypassed for seven minutes. Backwashing and pre-coating a RT F/D is a normal system operation and not considered maintenance.

A review of the Updated Safety Analysis Report (USAR) determined that the associated isolation functions are credited to mitigate the consequences of an RT pipe break accident described in USAR Chapter 6. Therefore, placing both divisions of RT LD in Bypass constituted a condition that could have prevented the fulfillment of the safety function of a system that is needed to mitigate the consequences of an accident. The direction for bypassing the RT LD system had been included in procedures since 1989 but did not constitute a reportable event until the issuance of NUREG-1022, Rev. 3 in 2013. The failure to report this condition was caused by not revising plant procedures when the Exelon fleet reporting requirements were revised to align with NUREG-1022, Rev. 3.

During a Clinton Power Station (CPS) review of an event at the Browns Ferry Station (ENS 47374), CPS determined that the original plant wiring design for the station battery ammeter circuits contains a shunt in the current flow from each direct current (DC) battery. Bolted onto the shunt bar are two Institute of Electrical and Electronics Engineers (IEEE) IEEE-383 qualified leads to an ammeter in the main control room (MCR). The small difference in voltage between the two taps on the shunt is enough to deflect the current meter in the MCR when current flows from the battery through the shunt. The ammeter wiring attached to the shunt does not have fuses, and if one of the ammeter wires shorts to ground at the same time another DC wire from the opposite polarity on the same battery also shorts to ground, a ground loop through the unfused ammeter cable could occur. With enough current going through the cable, the potential exists that the overloaded ammeter wiring could damage safe shutdown wiring in direct physical contact with the cable resulting in a loss of the associated safe shutdown function/capability.

The cause of this deficiency is the original design criteria not specify protection for shunt fed ammeter circuits. A modification is planned to correct the deficiency. Compensatory measures have been established until the modification has been installed.

On December 1, 2011, with the reactor shut down for refueling, an as-found Local Leak Rate Test (LLRT) was performed for Reactor Core Isolation Cooling turbine exhaust check valve 1E51-F040. The station primary containment leakage rate testing program which implements the requirements of 10 CFR 50, Appendix J, requires an as-found test to be performed prior to a scheduled maintenance activity to open and inspect the valve. The recorded leakage for the as-found LLRT was within acceptable limits. A step in the test procedure establishing the test lineup required the valve to be stroked open and shut by normal means.

A subsequent evaluation concluded that the pre-test stroking was unacceptable preconditioning; therefore, the as-found LLRT was invalid and considered to be a missed surveillance and a condition prohibited by Technical Specification 3.6.1.3. The test procedure incorrectly required the preconditioning and caused this event. Corrective actions include revising the test procedure and reviewing other similar check valve test procedures to determine the extent of this condition. An analysis to determine the actual impact of the valve stroke concluded that the preconditioning, although unnecessary to establish the test conditions, would have had little actual influence on the as-found test results.

On 5/23/11, Operators in the Main Control Room (MCR) noticed a change in noise levels originating from ventilation equipment outside the MCR. Investigation revealed the increased noise was from the MCR (VC) B Return Fan.

Vibration readings on the fan assembly were in the alert range, but not in the shutdown range.

Vibration readings were taken over an 8-hour period after the increase in vibration was noted. During that time, vibration levels remained constant. All other MCR Ventilation parameters for the VC B train were normal and unchanged. After this run, VC B train was placed in standby and VC A train was placed in service. An operability evaluation was completed for VC B which concluded VC B was operable.

On 6/7/11, an inspection of the fan was performed. The inspection identified a crack in the fan hub. A new fan was installed and the system was restored to an operable status on 6/10/11.

The fan hub was sent offsite for failure analysis, which concluded that the crack was the result of end-of-life fatigue, caused by low stress, high cycle loading. Analysis concluded that the hub assembly could not support the ability of the fan to perform its specified safety function over the designed mission time of 30 days, and thus the fan was inoperable.

The VC A system was operable during this period; therefore, no loss of safety function occurred during the period of inoperability.

On February 3, 2010, after entering Mode 2 (Startup) following refueling outage C1R12, it was discovered that a primary containment local leak rate test (LLRT) performed on feedwater check valve, 1B21-F032B, exceeded its acceptance criteria. Technical Specification (TS) Surveillance Requirement (SR) 3.6.1.3.11 requires that the combined leakage rate for both primary containment feedwater penetrations to be less than or equal to 2 gallons per minute (gpm) for the worst of the isolation valves. The measured leakage rate for 1B21-F032B was reported to be 2.5 gpm. This leakage rate is greater than that assumed in the plant safety analysis.

The unit was placed in Mode 4 (Cold Shutdown) to repair the valve and to re-perform the LLRT. After lubricating the valve actuator and stroking the check valve, the LLRT was performed to confirm the valve could meet its safety function for the Feedwater Leakage Control System (FWLCS). Following the satisfactory completion of the LLRT, plant startup was re-commenced on February 4, 2010.

The cause of the 1B21-F032B.check valve to fail its leak rate test was age-related degradation of the lubrication causing increased friction in the actuator.

Corrective action for this event includes establishing preventive maintenance activities to lubricate and overhaul the actuators.