05000366/LER-2007-005
| Edwin I. Hatch Nuclear Plant Unit 2 | |
| Event date: | 03-15-2007 |
|---|---|
| Report date: | 05-14-2007 |
| Reporting criterion: | 10 CFR 50.73(a)(2)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident |
| 3662007005R00 - NRC Website | |
PLANT AND SYSTEM IDENTIFICATION
General Electric - Boiling Water Reactor Energy Industry Identification System codes appear in the text as (EIIS Code XX).
DESCRIPTION OF EVENT
On March 15, 2007 at approximately 1532 EDT, Unit 2 was in the start-up mode at a reactor pressure of approximately 155 psig and estimated power level of one percent. The low pressure start-up testing of the High Pressure Coolant Injection (HPCI, EIIS Code BJ) system, which is performed at a reactor pressure of less than or equal to 165 psig, was in progress when the system failed to perform as required. The origin of this event is associated with an event that occurred earlier in the refueling outage when a water and oil mixture was found leaking from the Unit 2 HPCI turbine bearings and associated drains. The water intrusion into the HPCI oil system resulted in corrosion of the internals of the EGR, which normally functions to control the position of the HPCI Turbine Stop Valve. When the EGR failed, the Turbine Stop valve did not receive an open signal and thus remained closed resulting in the HPCI start failure. Maintenance replaced the EGR, the oil in the HPCI Turbine, the Reduction Gear Unit, and the Duplex Filters. The low pressure start-up testing of the HPCI system was re performed successfully and the HPCI system was returned to service at 1435 EDT on March 16, 2007.
CAUSE OF EVENT
This event was caused by corrosion of the EGR Actuator internals. The corrosion of the EGR internals is attributed to an event that occurred earlier in the refueling outage. During that event, water was introduced into the HPCI bearing oil system through a path created by a tagout. The root cause associated with the tagout event is that the drafter and the reviewer of the tagout did not adequately address the system or functional impact associated with the components that were tagged or removed from service.
REPORTABILITY ANALYSIS AND SAFETY ASSESSMENT
This event is reportable per 10 CFR 50.73 (a)(2)(v)(D) because an event occurred in which the HPCI system, a single train safety system, was rendered inoperable.
The HPCI system consists of a steam turbine-driven pump and the necessary piping and valves to transfer water from the suppression pool or the condensate storage tank (EIIS Code KA) to the reactor vessel. The system is designed to inject water to the reactor vessel over a range of reactor pressures from approximately 160 psig through full rated pressure. The HPCI system starts and injects automatically whenever low reactor water level or high drywell pressure indicates the possibility of an abnormal loss of coolant inventory. The HPCI system is designed to replace lost reactor coolant inventory in cases where a small line break occurs which does not result in full depressurization of the reactor vessel.
The backup for the HPCI system is the Automatic Depressurization System (ADS) together with two low pressure injection systems: the Low Pressure Coolant Injection (LPCI, EIIS Code BO) system and the Core Spray (CS, EIIS Code BM) system. The CS system is composed of two independent, redundant, 100 percent capacity subsystems. Each subsystem consists of a motor driven pump, its own dedicated spray sparger located above the core, and piping and valves to transfer water from the suppression pool to the sparger. Upon receipt of an initiation signal, the CS pumps in both subsystems start. Once ADS has reduced reactor pressure sufficiently, CS system flow begins.
LPCI is an operating mode of the Residual Heat Removal (EIIS Code BO) system. There are two independent, redundant, 100 percent capacity LPCI subsystems, each consisting of two motor driven pumps and piping and valves to transfer water from the suppression pool to the reactor vessel. Upon receipt of an initiation signal, all four LPCI pumps automatically start. Once ADS has reduced reactor pressure sufficiently, the LPCI flow to the reactor vessel begins.
ADS consists of 7 of the 11 Safety Relief Valves (SRV). It is designed to provide depressurization of the Reactor Coolant System during a small break LOCA if HPCI fails or is unable to maintain required water level in the Reactor Pressure Vessel (RPV). ADS operation reduces the RPV pressure to within the operating pressure range of the low pressure Emergency Core Cooling System (ECCS) subsystems (CS and LPCI), so that these subsystems can provide coolant inventory makeup.
In this event, the EGR would not send a signal to open the HPCI Turbine Stop Valve thus the HPCI system was inoperable. During the time the HPCI system was inoperable, the Reactor Core Isolation Cooling (RCIC, EIIS Code BN) system was available to inject high pressure water into the reactor vessel. Although not an ECCS, the RCIC system is designed, maintained, and tested to the same standards and requirements as the HPCI system and, therefore, should reliably inject water into the reactor vessel when required. If a break exceeded the capacity of the RCIC system (400 gallons per minute (gpm)), and with reactor pressure at approximately 155 psig, either the CS or LPCI systems could have been used to provide water to the reactor core. If the reactor pressure had been higher ADS would be available to depressurize the reactor vessel to the point that either the CS or LPCI systems could have been used. The capacity of one loop of the CS system is equal to that of the HPCI system (4250 gpm each); the capacity of one loop of the LPCI system is approximately three times that of the HPCI system. Therefore, any one of the four loops of the LPCI systems would have provided sufficient injection capacity for a small break LOCA.
Based on this analysis, it is concluded that this event had no adverse impact on nuclear safety. This analysis is applicable to all power levels and operating modes in which a LOCA is postulated to occur.
CORRECTIVE ACTIONS
Maintenance replaced the oil in the HPCI Turbine and reservoir, the Reduction Gear Unit, the Duplex Filters and the EGR was replaced.
Beginning of Shift Training was conducted to educate Operations Personnel on the lessons learned from this event. The training detailed the events leading up to the water intrusion into the HPCI oil system and emphasized that overall system impact must be evaluated anytime components are placed in an off-standard position. This training has been completed.
ADDITIONAL INFORMATION
Other Systems Affected: None Failed Components Information:
Master Parts List Number: 2E41-0002-5 EIIS System Code: BJ Manufacturer: Woodard Reportable to EPIX: Yes Model Number: A9903026 Root Cause Code: A Type: Control Operator, Flow EIIS Component Code: FCO Manufacturer Code: W290 Commitment Information: This report does not create any permanent licensing commitments.
Previous Similar Events: There are no similar events in the last two years in which a single-train safety system was rendered inoperable due to water intrusion.