05000366/LER-2007-008
| Edwin I. Hatch Nuclear Plant - Unit 2 | |
| Event date: | |
|---|---|
| Report date: | |
| Reporting criterion: | 10 CFR 50.73(a)(2)(iv)(A), System Actuation |
| 3662007008R00 - 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
This event is reportable, per 10 CFR 50.73 (a)(2)(iv)(A), because an event occurred which resulted in an automatic scram and automatic closure of primary containment isolation system valves.
On August 7, 2007 at 15:06 EDT, Unit 2 was in Mode 1 at 2804 CMWT, 100 percent power. An inadvertent actuation of contacts on an over-current relay protecting one of the three phases for the normal supply breaker of the 2D 4160V switchgear occurred. This resulted in a partial loss of the Condensate System (HIS Code SD) causing a reduction in Feedwater (EIIS Code SJ) flow. The reduction in Feedwater flow caused a decrease in reactor water level. The Recirculation System (PCIS, EIIS Code AD) runback setpoint was reached and the runback initiated. With the loss of two of the three condensate pumps the Recirculation System runback was not able to prevent the continued decrease in reactor water level. As water level continued to decrease, the Reactor Protection System (RPS, EIIS Code JC) reached the reactor low water setpoint and initiated an automatic scram signal. The Group 2 Primary Containment Isolation System (PCIS, EIIS Code JM) actuation setpoint was reached and the isolation signal initiated. Following the reactor scram, water level was recovered automatically with the normal condensate and feedwater system. No automatic initiation setpoints were reached for the Emergency Core Cooling Systems and the operators had no need to manually actuate those systems.
CAUSE OF EVENT
The root cause of this event was determined to be ineffective execution of a screening procedure written to determine scram/transient potential of I&C activities. The screening procedure was executed for the calibration of the overcurrent relay and errantly determined that there was no reactor trip potential when performing the procedure on-line and did not include a precaution for installation of the relay cover.
SAFETY ASSESSMENT
Following the automatic scram on low reactor water level, reactor vessel water level continued to decrease due to void collapse. Level reached a minimum of about thirty two inches below instrument zero (about 126 inches above the top of the active fuel). The decrease in water level resulted in a Group 2 PCIS isolation on low water level and thus closure of the Group 2 Primary Containment Isolation Valves per design. The RPS and PCIS are Engineered Safety Feature systems.
The operating Reactor Feedwater Pumps automatically restored water to its designed setpoint. Operations personnel verified correct system response and restored the isolation valves and the RPS to their normal configuration.
All systems functioned as expected and per their design given the water level transient. Water level was maintained well above the top of the active fuel throughout the transient and was restored to its desired value without the need for emergency core cooling system actuation. Therefore, it is concluded the event had no adverse impact on nuclear safety. This analysis is applicable to all power levels.
CORRECTIVE ACTIONS
Procedure 57CP-CAL-108-2, "Westinghouse CO Over Current Relay," has been revised to delete the statement indicating "no trip potential" and precautions for installation of relay covers have been added.
Procedure AG-MNT-03-0606, Procedure Review for Trip Potentials, will be revised to include guidance to assume components affected (or potentially affected) by the activity are actuated during the activity for the determination of a trip and transient potential. Implementation of this corrective action will be tracked under the corrective action program.
All I&C procedures previously screened will be re-evaluated per the revised guidance. Implementation of this corrective action will be tracked under the corrective action program.
ADDITIONAL INFORMATION
Other Systems Affected: None Failed Components Information: None Commitment Information: This report does not create any permanent licensing commitments.
Previous Similar Events:
automatic reactor scram. The root cause of that event identified an error in the calibration procedure which allowed work to be performed on-line that would directly cause an automatic scram if performed on-line.
Additional corrective actions for that event required a review of calibration procedures to determine if an automatic scram would be a potential effect of performing the procedure on-line. This review failed to achieve the desired result of identifying procedures which could result in a unit scram if performed on-line.
Therefore the corrective action for the event reported in LER 2-2006-002 was not effective in preventing the current event.