05000456/LER-2008-002
| Braidwood Station | |
| Event date: | 12-17-2008 |
|---|---|
| Report date: | 02-17-2009 |
| Reporting criterion: | 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications |
| 4562008002R00 - NRC Website | |
A. Plant Operating Conditions Before The Event:
Event Date:�December 17, 2008 Event Time: 13:48 Unit: 1 MODE: 1� Reactor Power: 99.9 percent Unit 1 Reactor Coolant System (RC) [AB]: Normal operating temperature and pressure
B. Description of Event:
There were no structures, systems or components inoperable at the beginning of the event that contributed to the severity of the event.
On December 17, 2008, Operations was performing a pressurization of a portion of the primary sampling system (PS) [KN] piping, as part of an American Society of Mechanical Engineers (ASME) surveillance. During the surveillance, excessive leakage was identified, and troubleshooting determined the leakage was due to leak-by of containment isolation valve 1PS228B. At 13:12 hours, Operations declared 1PS228B inoperable and entered Technical Specification (TS) 3.6.3, Containment Isolation Valves, Condition A. Condition A requires the penetration flowpath be isolated within four hours by closing and de-activating at least one valve. Operations determined that the redundant containment isolation valve, 1PS229B, would be closed and deactivated. At 13:48 hours, valve 1PS229B was de-energized by fuse block removal, and TS 3.6.3 Condition A was considered to be met.
During subsequent review, it was determined that the December 17, 2008 pressure test performed on the PS system piping had air flow directed toward containment and not away from containment. This is not the flow direction necessary to evaluate whether there is excessive leakage on the 1PS228B valve.
On December 19, 2008, a local leak rate test (LLRT) on the affected containment penetration was successfully performed. Based on the results of this LLRT, the determination was made that 1PS228B was never inoperable.
At 09:30 hours, during performance of this LLRT, 1PS229B was reenergized to allow for 1PS229B manipulation.
When the fuse block for 1PS229B was reinstalled, the valve indicated Open. A review of 1PS229B identified the valve to be a fail as-is valve upon loss of power. Therefore, while the fuse block for 1PS229B was removed to comply with TS 3.6.3 Condition A, with the valve open, the valve could not close on a containment isolation signal, and was determined to be inoperable with the fuse block removed.
On December 19, 2008, at 09:30 hours, 1PS229B was restored to operable status when the fuse block was reinstalled.
TS 3.6.3 Condition A applied to 1PS229B, which required the valve be isolated within four hours. Valve 1PS229B was failed open for approximately 44 hours5.092593e-4 days <br />0.0122 hours <br />7.275132e-5 weeks <br />1.6742e-5 months <br /> without being isolated. Therefore, this event is reportable under 10 CFR 50.73(a)(2)(i)(B), any operation or condition which was prohibited by the Technical Specifications.
C. Cause of Event
Upon declaring 1PS228B inoperable, Operations determined that 1PS229B should be closed and deactivated to comply with TS 3.6.3 Condition A. The Operations supervisor identified the fuse block that needed to be removed to deactivate 1PS229B, and directed a nuclear station operator (NSO) to remove the fuse block. The NSO did not close 1PS229B prior to removing the fuse block and incorrectly believed the valve would fail to the closed position when the fuse block was removed.
�NRC FORM 366A (9-2007) PRINTED ON RECYCLED PAPER All solenoid operated containment isolation valves at Braidwood Station fail to the closed position with the exception of two valves associated with each hydrogen monitor. Each hydrogen monitor has one containment isolation valve that fails open and one containment isolation valve that fails as-is. 1PS229B is a fail as-is valve.
The investigation of this event determined that the procedure used for entering and verifying the TS action statement requirements are met for inoperable containment isolation valves, is inadequate. The procedure documents the applicable TS action statement(s) entered, any loss of safety functions, affected support systems, and follow-up surveillances to ensure the TS action statement is met. In this event, the procedure documents that the 1PS229B is deactivated, but did not verify the final valve position. The procedure did not provide adequate guidance to ensure the components are placed in the proper position and subsequently verified in the position once the component is deactivated.
Therefore, the root cause of this event was determined to be inadequate procedure guidance to ensure the component was placed in the proper position and subsequently verified in the position once the component was deactivated to meet TS requirements.
D. Safety Consequences:
There were no safety consequences impacting the plant or public safety as a result of this event. During this event, isolation valve 1PS228B was closed while 1PS229B was deactivated in the open position. Based on the successful LLRT testing of 1PS228B, the 1PS228B valve was determined to be operable throughout the entire time the 1PS229B was open.
If a loss of coolant accident (LOCA) event occurred while the 1PS229B was deactivated open, the 1PS228B would have received a close signal for the valve and would have maintained containment integrity. Since 1PS228B is a fail open valve on a loss of power, the containment leakage is bounded by the Large Early Release Frequency (LERF) analysis for the unit, as described below.
The PS system is used to monitor containment hydrogen concentration during a beyond design basis event. The PS system takes a sample from inside of containment, flows through the hydrogen analyzer, and discharges back into containment. The sample piping is composed of one-half inch inside diameter stainless steel.
The plant specific probabilistic risk assessment screens all containment penetrations as being a non-significant source of leakage if the penetration is two inches in diameter or less. Assuming a complete loss of pressure boundary integrity simultaneously for all four affected penetrations per unit, would not result in a flow area equivalent to a two inch diameter penetration. Due to the size of the containment penetration, the resulting leakage during a LOCA event would not be sufficient to impact the LERF for the unit. The amount of leakage through the PS system piping is negligible compared to the TS leakage limit acceptance criteria for the containment. Therefore, the failure to isolate the penetration does not result in a significant risk increase.
This is not a safety system functional failure.
E. Corrective Actions:
The corrective action to prevent recurrence is to revise the operating procedure to add specific controls to prompt the user to identify the proper isolation valve, the failure mode of the valve, if appropriate, and verifications that the actions taken properly isolated the penetration.
Additional corrective actions include review of other procedures for single point vulnerabilities and revise those procedures as necessary.
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C. Previous Occurrences:
There have been no similar Licensee Event Report events at Braidwood Station in the last three years.
G. Component Failure Data:
Manufacturer Nomenclature Model Mfg. Part Number N/A N/A N/A N/A NRC FORM 366A (9-2007) PRINTED ON RECYCLED PAPER