05000293/LER-2009-002
| Pilgrim Nuclear Power Station | |
| Event date: | 2-2-2009 |
|---|---|
| Report date: | 02-19-2010 |
| Reporting criterion: | 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications 10 CFR 50.73(a)(2)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident |
| 2932009002R01 - NRC Website | |
BACKGROUND
The Secondary Containment System, in conjunction with other engineered safeguards and nuclear safety systems, limits radioactive material release during normal plant operations to within 10 CFR 20 limits and limits the release to the environs of radioactive materials so that the offsite dose from a postulated DBA will be below the guideline values of 10CFR100. The secondary containment is designed to minimize any ground level release of radioactive materials that might result from a serious accident. The Reactor Building provides secondary containment during reactor operation, when the drywell is sealed and in service; the Reactor Building provides primary containment during periods when the reactor is shutdown, the drywell is open, and activities are ongoing that require secondary containment to be operable. Because the secondary containment'is an integral part of the complete containment system, secondary containment is required at all times that primary containment is required as well as during movement of recently irradiated fuel and during operations with the potential to drain the reactor vessel (OPDRVs). There are two principal accidents for which credit is taken for secondary containment operability. These are a loss of coolant accident (LOCA) although not specifically evaluated for alternate source term methodology and a fuel handling accident involving recently irradiated fuel.
The secondary containment performs no active function in response to each of these limiting events; however, its leak tightness is required to ensure that the release of radioactive materials from primary containment is restricted to those leakage paths and associated leakage rates assumed in the accident analysis that fission products entrapped within the secondary containment structure will be treated by the Standby Gas Treatment System (SGTS) prior to discharge to the environment. An operable secondary containment provides a control volume into which fission products that bypass or leak from primary containment, or are released from the reactor coolant pressure boundary components located in secondary containment can be diluted and processed prior to release to the environment. For the secondary containment to be considered operable, it must have adequate leak tightness to ensure that the required vacuum can be established and maintained. The Reactor Building ventilation system always maintains flow from areas of least potential contamination to areas of highest potential contamination. The Reactor Building is maintained at all times at a small negative pressure with respect to its surroundings to ensure any contamination will be contained with its boundaries.
All normally open drains which are open both to the secondary containment and the outside atmosphere are provided with water seals to maintain secondary containment integrity. This is exemplified by the four 14 inch dewatering lines for the Reactor Building auxiliary bay floor sumps. These lines penetrate the secondary containment boundary, two below each of the two sumps, and terminate in a pair of troughs within the torus compartment about 6 inches above the trough floor. The two 4 foot cubic shaped troughs maintain secondary containment integrity by providing water seals for each of the four lines. High and low trough water levels are alarmed in the control room. On low water level, the operators are directed by procedure to refill the troughs via the Condensate Transfer System, to maintain containment integrity. The troughs provide pipe break flood protection for the auxiliary bay RBCCW/ TBCCW equipment. Pipe break events causing flooding in the auxiliary bay would drain into the torus troughs and then overflow onto the torus room floor.
The water level in the trough is intended to ensure auxiliary bay drain pipe submergence and therefore a water seal between the reactor building (secondary containment) and the auxiliary bay which would allow either normal reactor building ventilation for routine operations or the standby gas treatment system to establish the 0.25 inches of negative water pressure required by Technical Specifications.
Technical Specification (TS) Limiting Condition for Operation (LCO) 3.7.C.1 requires, in part, that whenever the reactor is critical, secondary containment integrity must be maintained.
EVENT DESCRIPTION
On December 22, 2009, at approximately 0845 hours0.00978 days <br />0.235 hours <br />0.0014 weeks <br />3.215225e-4 months <br /> a design engineer performing a walk down of the torus room notified the main control room (MCR) that the torus trough (bay #15) for auxiliary bay `A' was dry. The engineer also indicated that the torus trough (bay #13) for auxiliary bay 'B' appeared to be at a lower level than normally observed. The Shift Manager immediately took actions to verify the engineer's observations.
Main Control Room (MCR) alarm, C904L-A7, Torus Trough Hi/Lo, which was not in alarm status, was verified to be enabled. The Operations Field Shift Supervisor was dispatched to the torus room to verify the engineer's observations.
Secondary containment integrity is ensured by maintaining a controlled water level above the drain pipe openings in each trough such that a water seal ensures the ability to establish and maintain 0.25 inches of negative water pressure within secondary containment. Active LCO, LCO ACT-1-09-0219, was entered at 0845 hours0.00978 days <br />0.235 hours <br />0.0014 weeks <br />3.215225e-4 months <br /> because secondary containment integrity could not be ensured with one trough dry. The trough was filled and the LCO was exited one hour later at 0945 hours0.0109 days <br />0.263 hours <br />0.00156 weeks <br />3.595725e-4 months <br />.
Additionally, an 8-hour 50.72 notification was made to the USNRC.
CAUSE
The apparent cause of this event is due to a small leak of water from the torus bay 'A' trough. This leakage caused the trough level to slowly lower over time which ultimately challenged the failed low level alarm level switch, LS-9038B. Plans for repair of the torus trough are ongoing and are being tracked in PNPS's Corrective Action Program via CR-PNP-2009-5309.
The second apparent cause of this event was the failure of level switch LS-9038B which provides the torus trough low level alarm signal. The level switch actuating plate was found misaligned which had required compensation with the adjustment screws during functional calibrations and eventually over time, failed to provide the trough low water level condition to the control room. This alarm is designed to alert the main control room of a high or low level in the torus trough condition and to initiate appropriate corrective actions.
The lack of a specific trough level acceptance criteria in the operator rounds contributed to this event. While operators performed the weekly torus room check/ tour, there were no inspections of the troughs or criteria for acceptable level bands.
CORRECTIVE ACTION
Immediate corrective actions taken were to refill the trough to the correct level, repair the level switch which was found to be defective, ensured other the trough had adequate water level and its level switches were working properly, and enhanced the weekly tour requirement of the Torus Compartment performed by plant operations.
Corrective actions planned include leak repair of the torus trough, level switch surveillance test enhancements, and level switch preventative maintenance (PM) basis document revision.
These above actions are being tracked in the Pilgrim Station Corrective Action Program (CR-PNP-2009- 5295 and CR-PNP-2009-5309).
SAFETY CONSEQUENCES
The event posed no threat to public health and safety.
The plant was operating at 100% power prior to and during the time period when the torus trough was found without a water seal and required repair. All other secondary containment sub-systems were operable during this time period.
Secondary containment integrity is ensured by maintaining a given water level in each of the two torus troughs.
The lack of a water seal in a torus trough creates a scenario in which the auxiliary bay atmosphere would communicate directly with the secondary containment atmosphere. This would cause the effective volume upon which the standby gas system would attempt to maintain at 0.25 inches of water negative pressure to be larger.
This presents the potential for secondary containment air pressure to approach, equal or be greater than the air pressure in ambient building or atmospheric pressures. If this occurred and remained undetected the design principle of leakage into secondary containment, filtration by standby gas and release from the main stack would be reduced or neutralized. Ultimately the potential for an unmonitored ground level release would increase.
Technical Specification 3.7.0 requires that the secondary containment be operable in the RUN mode. The Secondary Containment integrity definition was not satisfied during this time period.
Technical Specification definition for Secondary Containment Integrity means that the reactor building is intact and the following conditions are met:
1. At least one door in each access opening is closed, 2. The standby gas treatment system is operable, 3. All automatic ventilation system isolation valves are operable or secured in the isolated position.
Since the Secondary Containment System was able to be restored to an operable status following the re filling of the torus troughs to a proper level, there was no long term safety.significance associated with this event.
REPORTABILITY
This report is being submitted in accordance with 10 CFR 50.73(a)(2)(i)(B) for operation or condition prohibited by Technical Specifications and 10 CFR 50.73(a)(2)(v)(D) for an event or condition that could have prevented fulfillment of a safety function needed to mitigate the consequences of an accident.
SIMILARITY TO PREVIOUS EVENTS
A review was conducted of Pilgrim Station LERs since 1974. There were no LERs related to the torus troughs with the failure to maintain secondary containment.
The review identified Secondary Containment events that occurred in 1985 and 2008.
- LER 1985-18 addressed an event where a Secondary Containment damper (AO-N-90) would not fully close.
- LER 2008-001 addressed on-line testing of the Reactor Building Isolation Control System (RBICS) ventilation dampers. This testing identified that in the closed position, damper AO-N-78 did not fully close. The damper was reported to have a one-half inch gap opening across two of the four damper blades and did not meet Technical Specification requirements for full damper closure.
These events were reported as events where Technical Specifications were not satisfied.
ENERGY INDUSTRY IDENTIFICATION SYSTEM (EllS) CODES The El IS codes for this report are as follows:
SYSTEMS � CODES Containment Leak System � BD