05000397/LER-2017-007

LER-2017-007, Valve Closure Results in Momentary Increase in Secondary Containment Pressure

Columbia Generating Station
Event date:	10-03-2017
Report date:	11-30-2017
Reporting criterion:	10 CFR 50.73(a)(2)(v)(C), Loss of Safety Function - Release of Radioactive Material 10 CFR 50.73(a)(2)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident
Initial Reporting
ENS 52999	10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident, 10 CFR 50.72(b)(3)(v)(C), Loss of Safety Function - Release of Radioactive Material
3972017007R00 - NRC Website

LER 17-007-00 for Columbia Generating Station Regarding Valve Closure Results in Momentary Increase in Secondary Containment Pressure

ML17334B440
Person / Time
Site:	Columbia
Issue date:	11/30/2017
From:	Javorik A L Energy Northwest
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
GO2-17-187 LER 17-007-00
Download: ML17334B440 (4)

comments regarding burden estimate to the Information Services Branch (T-2 F43), U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, or by e-mail to used to impose an information collection does not display a currently valid OMB control number, the NRC may not conduct or sponsor, and a person is not required to respond to, the information Columbia Generating Station 05000- 397 17 - 007 - 00

Plant Conditions

At the time of the event, the plant was operating in Mode 1 at 100% power. The division 1 emergency diesel generator [EK] [DG] (DG1) was out of service for maintenance but did not contribute to this event. There were no other structures, systems, or components that were inoperable at the start of the event that contributed to the event.

Event Description

On October 3, 2017 at 0800 PDT, Operations personnel received an annunciator alarm for high Secondary Containment [NH] differential pressure when the Reactor Building [NG] exhaust inboard isolation valve [V] failed shut. For a time period of approximately two minutes, Secondary Containment differential pressure increased above -0.25 inches of water gauge (inwg), resulting in Technical Specification (TS) 3.6.4.1 not being met. The pressure rise was due to unexpected isolation of the Reactor Building inboard exhaust isolation valve during electrical switchgear inspections.

This event is reportable as an event that could have prevented fulfillment of safety functions needed to control the release of radiation, and mitigate the consequences of an accident per 10 CFR 50.73(a)(2)(v)(C) and 10 CFR 50.73(a)(2)(v)(D).

This condition was reported under 10 CFR 50.72(b)(3)(v)(C) and 10 CFR 50.72(b)(3)(v)(D) via Event Notification #52999 for an event or condition that could have prevented fulfillment of a safety function needed to control the release of radioactive material and accident mitigation.

Immediate Corrective Actions

Control Room operators reopened the Reactor Building exhaust inboard isolation valve and pressure was restored within TS limits.

Assessment of Safety Consequences

This event resulted in an unplanned entry into TS 3.6.4.1.A, in which Secondary Containment pressure was greater than -0.25 inwg for approximately two minutes. The peak pressure during this event was +3.7 inwg. While the actual pressure was beyond the range allowed by Technical Specifications, the purpose of maintaining a slight negative pressure is to assist in drawdown of secondary containment to support accident response of the safety related SGT system. Existing engineering analysis demonstrates the drawdown credited in Columbia's accident analysis could have been attained using either of the two available trains of the SGT system. There was no actual safety consequence associated with this event since opening the Reactor Building exhaust inboard isolation valve restored Secondary Containment, and there was no loss of safety function or potential for radiological release.

Cause of Event

The apparent cause of the event is a surface degradation on the lower stab of an electrical disconnect causing a momentary high resistance when the cubicle door is opened. This event occurred during performance of thermography in the cubicle.

Further Corrective actions Cleaning and inspection of the disconnect switches including the stab to bus connection on the electrical disconnects and correction of any deficiency found will be performed in the next available outage. As an interim corrective action, thermography on the electrical panels associated with the degraded electrical disconnects will be discontinued until after the cleaning and inspection of the electrical disconnect is completed.

comments regarding burden estimate to the Information Services Branch (T-2 F43), U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, or by e-mail to used to impose an information collection does not display a currently valid OMB control number, the NRC may not conduct or sponsor, and a person is not required to respond to, the information collection.

Columbia Generating Station 05000- 397 17 - 007 - 00 Similar Events A loss of the ability to maintain Secondary Containment pressure greater than required by Technical Specifications has occurred at Columbia Generating Station six times in the past two years. One event was due to inclement weather, one event was due degraded lubrication on a circuit breaker [52] fan [FAN],one was a faulty control switch on a fan, one was due to human performance errors during rescheduled work, and two events were due to latent errors from initial construction of the plant.

Energy Industry Identification System (EllS) Information codes from IEEE Standards 805-1984 and 803-1983 are represented in brackets as [XX] and [XXX] throughout the body of the narrative.