05000482/LER-2010-005
| Docket Number Sequential Revmonth Day Year Year Month Day Yearnumber Na 05000 | |
| Event date: | 03-02-2010 |
|---|---|
| Report date: | 05-03-2010 |
| Reporting criterion: | 10 CFR 50.73(a)(2)(iv), System Actuation 10 CFR 50.73(a)(2)(iv)(A), System Actuation |
| 4822010005R00 - NRC Website | |
On 3/2/2010 at 1458 CST, Wolf Creek Generating Station (WCGS) experienced an automatic reactor trip due to Steam Generator (SG) Water Level — Low Low actuation signal [EllS Code: JB]. The SG water level — Low Low actuation signal was caused by the loss of the speed sensor to the train "A" main feedwater (MFW) pump control circuitry [EllS Code: SJ]. The loss of the speed sensor was caused by the loss of 120 VAC power to the circuit.
The loss of power to the speed sensor occurred when inverter PNO9 [El IS Code: SJ—INVT] was being transferred to its alternate power supply to facilitate a minor maintenance activity to replace a light bulb. The power supply was being transferred to the alternate supply using procedure SYS PN-200, "Energizing and Deenergizing Inverters PNO9 or PN10." Inverter PNO9 did not transfer from the normal to alternate power supply due to the sticking of the reed relay on the static transfer switch circuit board.
The unit received a feedwater isolation and auxiliary feedwater actuation (both motor-driven and turbine-driven) as a result of the SG water level — Low Low actuation signal. All control rods inserted fully and the Reactor Trip System (RTS) and the Engineered Safety Feature Systems [EllS Code: JE] performed as expected. Decay heat removal following the reactor trip was via the atmospheric relief valves and auxiliary feedwater flow to the steam generators. At the time of the reactor trip the steam dump valves were not available as the loss of inverter PNO9 caused the loss of the C-9 interlock blocking steam dump operation.
The steam dump valves became available at 1554 CST, on 3/2/2010, when inverter PNO9 was reenergized from its alternate power supply.
All systems functioned as designed with the exception of the instrumentation powered by inverter PN09. PNO9 does not supply any safety-related equipment. At the time of the reactor trip the train "A" diesel generator [EllS Code: EK] and the train "A" Class 1 E air conditioning unit [EllS Code: VI] were out of service for maintenance.
BASIS FOR REPORTABILITY
The reactor trip and subsequent actuation of Engineered Safety Feature Actuation System instrumentation described in this event is reportable per 10 CFR 50.73(a)(2)(iv)(A), which requires reporting of "Any event or condition that resulted in manual or automatic actuation of any of the systems listed in paragraph (a)(2)(iv)(B) of this section." Paragraph (B)(1) of 10 CFR 50.73(a)(2)(iv) includes "Reactor protection system (RPS) including: reactor scram or reactor trip." Paragraph (B)(6) of 10 CFR 50.73(a)(2)(iv) includes "PWR auxiliary or emergency feedwater.
During the performance of procedure SYS PN-200, inverter PNO9 did not transfer from the normal to alternate power supply due to the sticking of the reed relay on the static transfer switch circuit board.
A contributing cause was the failure to perform timely preventive maintenance (PM) on components exceeding their recommended life expectancy. Life Cycle Management and Preventive Maintenance Optimization plans determined the schedule for implementing PMs for the inverters. The inverters were on a staggered schedule for PM parts replacement during the refueling outages. PM activities on PN010 were completed in Refueling Outage 17 in Fall 2009. PM activities for PN009 were scheduled for Refueling Outage 18 in Spring 2011.
CORRECTIVE ACTIONS
The static switch control circuit board was replaced along with a number of other circuit boards and fuses. After replacement, the power supplies were transferred successfully several times.
Preventive maintenance has been established to replace the circuit boards in accordance with the manufacturer recommendations.
SAFETY SIGNIFICANCE
The safety significance of this event is low. This event is analyzed as reported in WCGS Updated Safety Analysis Report (USAR) Section 15.2.7, "Loss of Normal Feedwater Flow." Results of the analysis show that a loss of normal feedwater does not adversely affect the core, the reactor coolant system, or the steam system, since the auxiliary feedwater capacity is such that reactor coolant water is not relieved from the pressurizer relief or safety valves.
There were no adverse effects on the health and safety of the public.
OPERATING EXPERIENCE/PREVIOUS EVENTS
in response to failures of the primary and secondary fuses for the Westinghouse 7300 control card frame that contained the associated control cards for the MFRV.
MFRV separating from the valve stem.