05000325/LER-2011-002
| Brunswick Steam Electric Plant (Bsep), Unit 1 | |
| Event date: | 10-13-2011 |
|---|---|
| Report date: | 12-08-2011 |
| Reporting criterion: | 10 CFR 50.73(a)(2)(ii)(B), Unanalyzed Condition |
| Initial Reporting | |
| ENS 47341 | 10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition |
| 3252011002R00 - NRC Website | |
Energy Industry Identification System (EIIS) codes are identified in the text as [XX].
Introduction Initial Conditions At the time of the event, both Unit 1 and Unit 2 were in Mode 1, operating at approximately 100 percent of rated thermal power (RTP).
Reportability Criteria This event is being reported in accordance with 10 CFR 50.73(a)(2)(ii)(B), as an event or condition that resulted in the nuclear power plant being in an unanalyzed condition that significantly degraded plant safety. The NRC was initially notified of this event on October 13, 2011 (i.e., Event Number 47341).
Event Description
On October 13, 2011, in preparation for converting from 10 CFR 50, Appendix R, to National Fire Protection Association (NFPA) Standard 805, a review of the Brunswick Steam Electric Plant (BSEP) safe shutdown analysis identified conditions that may not ensure a protected train of equipment remains available under certain postulated fire scenarios. The analysis determined that a postulated fire in specific fire areas could cause spurious actuation of critical components, potentially resulting in loss of equipment required for safe shutdown. A fire in one of the specified fire areas could potentially adversely affect the following: Suppression Pool level instrument 2-CAC-LT-2602, Residual Heat Removal (RHR) [BO] net positive suction head (i.e., drywell containment overpressure), Reactor Core Isolation Cooling (RCIC) [BN], Emergency Bus E-1, and Emergency Bus E-3 [EK]. The review identified an enhancement for manual operation of the RCIC barometric condenser vacuum pump. Though RCIC is required for certain alternative safe shutdown scenarios, operation of the barometric condenser vacuum pump is considered a good operating practice and does not affect RCIC operability.
Per 10 CFR 50, Appendix R, fire damage limits requires that one train of equipment and components, needed to achieve and maintain safe shutdown during a single fire event, be functional and free of fire damage. This includes electrical raceway, instrumentation, controls emergency electrical power and other auxiliary equipment required for the components to perform their function. Contrary to this requirement, BSEP's current review of the safe shutdown analysis determined that there were certain electrical raceways, required for safe shutdown, that may not remain free of fire damage during a fire event. The damaged electrical cables can be classified as affecting three general equipment functions. These are:
1) Emergency Bus load shedding, 2) containment pressure to maintain RHR pump net positive suction head (NPSH), and 3) Suppression Pool level indication.
Emergency Bus Load Shedding Safe shutdown analysis revalidation found that control power cables to conventional service water pumps Event Description (continued) (i.e., 1B, 2A, and 2C) [BI], may be damaged during a fire in the Turbine Building. Cables routed inside the Turbine Building were not included in the original analysis. A fire event inside the Turbine Building was not postulated since the assumption was made that fire damage in this area would have no shutdown impact.
However, control power for conventional service water pumps are routed through the Turbine Building and fire damage to these cables may not allow the emergency diesel generator output breaker to close and energize Emergency Bus E-1 or E-3.
In addition, control power and 4KV power cables for the 2A Control Rod Drive (CRD) [AA] pump are also exposed to fire during an event in the Unit 1 Reactor Building. This may cause the pumps' 4KV [EB] feeder breaker to remain closed during an automatic load shed of Emergency Bus E-1 or E-3. Fire damaged cable in the Unit 1 Reactor Building was identified during the original safe shutdown analysis since the Reactor Building is an Appendix R fire area. It appears that cables routed for the 2A CRD pump were not identified in that evaluation.
Containment Pressure The safe shutdown analysis revalidation found several cases where a fire inside the Unit 2 Reactor Building may damage cables associated with Containment Atmospheric Control (CAC) [BB] valves. Spurious operation of these valves would allow a decrease in primary containment pressure. The revalidation also determined that continued operation of the 1B and 2B Reactor Building Closed Cooling Water (RBCCW) [CC] pumps during a fire event may lower primary containment pressure. Appendix R assumptions include loss of containment cooling which results in a requirement to cool down rapidly to limit drywell temperature. High cool-down rates, combined with the additional decay heat from extended power uprate, results in elevated suppression pool temperatures that, in turn, reduce the available NPSH for the RHR and Core Spray [BG] pumps. This effect can be offset by allowing containment and suppression pool airspace pressure to rise as drywell temperature increases. The Appendix R fire protection review at BSEP did not consider containment pressure maintenance when implementing the extended power uprate. As a result, the Appendix R program and associated safe shutdown procedures were not evaluated or revised to incorporate this new requirement.
Suppression Pool Level Safe shutdown analysis revalidation determined that control cables for solenoid valves associated with suppression pool level instrument 2-CAC-LT-2602 may be damaged during a fire in the Unit 2 Reactor Building. Cable damage may cause the valves to close losing level indication in the Control Room. Fire damaged cable in the Unit 2 Reactor Building, an Appendix R fire area, was identified during the original safe shutdown analysis. It appears that cables routed for equipment associated with the CAC level transmitter were not initially identified.
Event Cause No root cause was determined due to it being a historical condition. The respective calculations were originally prepared in the 1990 to 1993 time frame. As such, a conclusive root cause cannot be established.
Safety Assessment The safety significance of this event is minimal. Fire watches were established for the affected portions of fire areas RB1-1, RB2-1, TB1, CB-2, CB-13, and CB-23. Additionally, fire detection and suppression equipment in the affected areas were fully functional at the time of discovery. The findings identified here are based on fire scenarios that have not actually occurred.
Corrective Actions
The following actions have been completed.
- Establish a once per hour fire watch in the affected portions of fire areas RB1-1, RB2-1, TB1, CB-2a(b), and CB-13a(b). In addition, credit is taken for fire area CB-23 (i.e., Control Building 49 ft. elevation, back-panel areas) since this area is continuously occupied.
- Revise alternative safe shutdown procedures OASSD-02, 1ASSD-04, 1 ASSD-06, 2ASSD-04, 2ASSD-05, and 2ASSD-06 to include compensatory actions to address fire damaged cables and equipment operation to ensure safe shutdown.
The following corrective actions will be taken.
- Complete the new safe shutdown analysis, calculation BNP-E-9.010. This action is expected to be complete by March 10, 2012.
- Complete the transition to National Fire Protection Association (NEPA) Standard 805. This action is expected to be complete by December 31, 2014.
Previous Similar Events
A review of LERs and corrective action program condition reports for the past three years did not identify any similar previous occurrences.
Commitments No regulatory commitments are contained in this report.