05000482/LER-2005-001
| Wolf Creek Generating Station | |
| Event date: | |
|---|---|
| Report date: | |
| Reporting criterion: | 10 CFR 50.73(a)(2)(ii)(B), Unanalyzed Condition |
| 4822005001R00 - NRC Website | |
Background:
The centrifugal charging pumps (CCPs) [EIIS Code: P] are used to inject borated water into the reactor to maintain reactor water inventory, maintain shutdown reactivity margin, and maintain a flow of cooling water to the reactor coolant pump (RCP) seals to prevent damage to the seals. A Safety Injection Signal opens the Boron Injection Tank (BIT) inlet and outlet valves for the CCPs to inject borated water into the reactor. A fire in the plant has the potential to cause the BIT inlet valve, EMHV8803A, to remain closed and result in a loss of the capability to maintain reactor water inventory and borate the reactor.
Plant Conditions Prior to the Event:
MODE —1 Power — 100 percent Normal Operating Temperature and Pressure
Event Description:
In early January 2005, it was discovered that corrective actions for LER 2002-004-02 were not fully implemented for auxiliary building corridor at elevation 1974 ft. (Fire Area A-1) by the associated Design Change Package (DCP).
Completion of the fire wrap modification by December 2003 was a commitment made in LER 2002-004-02.
Corrective action program documentation, PIR 2000-2378, had identified several electrical raceways that needed to be wrapped with a fire barrier to protect a safe shutdown success path if a fire occurred in Fire Area A-1. A design modification package, DCP 11038, was generated to implement the modification for the fire wrap barriers. The electrical raceway associated with the Boron Injection Tank (BIT) inlet valve, EMHV8803A, was not included in the modification package. The corrective action document, PIR 2000-2378, was closed incorrectly as a result of the implementation of the design modification.
EMHV8803A is needed in the case of fire in Fire Area A-1 to maintain the water level in the Pressurizer on scale by injecting borated water in excess of reactor water being lost by reactor coolant system (RCS) letdown, if letdown fails to isolate. This lack of fire wrapping could have prevented EMHV8803A from performing its post-fire safe shutdown function.
Basis for Reportability:
Due to intervening combustibles, a fire in Fire Area A-1 has the potential to cause valve EMHV8803A to not function properly and cause a loss of the capability to borate the reactor. Based on this information, WCNOC made an eight hour Emergency Notification System call in accordance with 10 CFR 50.72(b)(3)(ii)(B).
This condition is also reportable pursuant to 10 CFR 50.73(a)(2)(ii)(B) for any event or condition that resulted in the nuclear power plant being in an unanalyzed condition that significantly degraded plant safety.
Root Cause:
The root cause was human error by Design Engineering personnel, Fire Protection personnel, and contracting personnel involved with the design modification process. Greater than 95% of the work for the design modification process was located in the west corridor of Fire Area A-1. The power and control cables for valve EMHV8803A were located in the east/northeast corridor of Fire Area A-1. The personnel involved with the design modification process had tunnel vision that limited their focus to the west corridor modifications.
Additionally, the Design Engineer who closed the corrective action document did not fully utilize procedural guidance to ensure that the design modification process fully corrected the issues identified.
Corrective Actions:
A lesson-learned session for the entire Design Engineering Department will be conducted focusing, in part, on human performance error reduction and engineering rigor. This event will be part of initial and continuing training classes for engineers qualified to review and approve design modification change packages.
Design Engineering will initiate a plant modification change package. This modification will ensure that conduit for valve EMHV8803A will be fire wrapped in Fire Area A-1 of the Auxiliary Building. The implementation of this modification will be completed by January 20, 2006.
Safety Significance:
The potential for cables for valve EMHV8803A to be affected by a fire in Fire Area A-1, such that control of this valve is lost, is low because:
(1) The separation between the trains is approximately 135 feet (Appendix R requires separation of "more than 20 feet with no intervening combustible or fire hazards.
(2) The majority of "intervening combustibles" in this case are IEEE 383 electrical cables; such cables will not self-sustain a fire. Other combustibles in the general area consist only of personal anti-contamination clothing located in the RCA dress-out area. The dress-out area is located about 80 feet from the Train B cables. There is automatic fire suppression in that area.
(3) Valve EMHV8803A cables are in electrical conduit and while the conduit is not a rated fire barrier, it would provide some protection against damage from a fire. In addition, a fire sprinkler line with several sprinkler heads runs less than two feet from this conduit for most of its distance in Fire Area A-1.
(4) Valve EMHV8803A would be needed only in the case of a failure of letdown to isolate.
Therefore, potential safety significance of this condition in all areas (system/component operability, nuclear/radiological safety, health/safety of the public and environmental impact) is considered to be low.
Operating Experience/Previous Events:
An occurrence of a similar event was reported via licensee event report LER 1999-009-00. In LER 1999-009-00, it was determined that there was inadequate separation of cables for valves and level transmitters for the volume control tank. In the event of a fire, a potential existed for gas intrusion into the suction of the centrifugal charging pump. While corrective actions have been taken to address these conditions, an additional corrective action for LER 1999-009-00 was to validate the post fire safe shutdown analysis and to provide necessary correction to the Updated Safety Analysis Report (USAR).
The validation consists of two phases: phase one reverified the design criteria and phase two completes the post-fire safe shutdown analysis review. The conditions identified in LER 2005-001-00 were discovered during phase two and reported in LER 2002-004-02.