05000366/LER-2006-002
| Edwin I. Hatch Nuclear Plant - Unit 2 | |
| Event date: | |
|---|---|
| Report date: | |
| Reporting criterion: | 10 CFR 50.73(a)(2)(iv)(A), System Actuation |
| 3662006002R00 - 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
At 0016 EDT, on 4/5/06, a reactor scram occurred on Turbine Control Valve (TCV) (EIIS Code TA) fast closure. Reactor water level initially decreased due to void collapse and then increased to 60 inches. Reactor pressure spiked to approximately 1125 psig, which resulted in an additional trip signal due to high reactor pressure. All control rods fully inserted (EIIS Code JD). Eight of the eleven Safety Relief Valves (SRVs) (EIIS Code SB) opened as designed to mitigate the pressure transient. The remaining three SRVs did not open and should not have opened since the reactor pressure did not reach their lift set points. Reactor water level (RWL) increased to a maximum value of approximately 60 inches above instrument zero. This swell in RWL was due to a combination of the SRVs lifting and both recirculation pumps (EIIS Code AD) tripping automatically on End of Cycle-Recirculation Pump Trip (EOC-RPT) per design. Reactor water level subsequently decreased when the reactor feed pumps (RFP) (EIIS Code SJ) automatically tripped on high reactor water level per design. The 'A' RFP was restarted and subsequently tripped per design due to low vacuum in the main condenser (EIIS Code SG). It was once again restarted but subsequently tripped again on low condenser vacuum when the reactor feed pump turbine speed exceeded 3500 rpm. (The feed pump turbine low vacuum trip may be reset, and the feed pumps re-started with the low vacuum condition in effect, provided the turbine speed does not exceed 3500 rpm). As reactor water level (RWL) continued to decrease, Operations manually initiated the Reactor Core Isolation Cooling (RCIC) system (EIIS Code BN) which slowed the rate of RWL decrease. At this point, Operations also initiated a pressure reduction by reducing the turbine pressure set.
The 'A' RFP was again re-started but could not inject due to the main condenser vacuum limiting injection pressure of the RFP to a value below the actual reactor pressure. Operations manually initiated the High Pressure Coolant Injection (HPC() system (EIIS Code BJ) at a RWL of approximately seven inches above instrument zero.
RWL increased and HPCI was secured at approximately 28 inches above instrument zero. RWL continued to increase due to the RCIC injection and the swell effect of decay heat heating the cooler water injected by HPCI and RCIC. As a result, the 'A' RFP subsequently tripped on high RWL, and was subsequently re-started.
Operations continued to decrease reactor pressure until pressure reached approximately 475 psig. At that point, the 'A' RFP was secured and RWL was controlled by the condensate booster pump (EIIS Code SD).
Additional Details:
Main condenser vacuum decreased because the turbine steam sealing pressure was in manual control. The regulator for the Steam Seal Feed Valve Controller was replaced during the 2005 spring refueling outage.
Later in 2005, additional problems were identified indicating problems with the steam seal feed valve. These problems were documented in the corrective action program. The Steam Seal System was placed in alternate seal steam pressure control per plant procedures. During normal operation manual control maintained main condenser vacuum. During the forced shut-down the valve was repaired.
In addition, the reactor coolant system (RCS) temperature in the bottom head region decreased by more than the 100 degrees Fahrenheit in one hour limit provided in the Technical Specifications. An evaluation of this condition was performed and the RCS and the reactor vessel were found acceptable for continued operation.
CAUSE OF EVENT
The direct cause of this event was a main turbine trip from a sensed power load unbalance (PLU) due to the performance of procedure 57CP-CAL-010-2, "Esterline Angus Megavar & KV Recorder," in Mode 1 (Run).
The power-load unbalance caused the TCV fast closure. I&C was calibrating recorder 2S32-R017, KVR Generator & Main Transformer Protection Recorder, at the time of the trip. Review of plant and vendor drawings determined that shorting the CTs to ground during the performance of 57CP-CAL-010-2 was the cause of the turbine trip. Shorting the CTs to ground in 2H11-P658 caused the current to go to zero. The PLU module in the Mark I turbine control system looks at the current in each generator output phase and determines the electrical power based on the percent current. The PLU then compares the electrical power to the mechanical power and trips the turbine if a greater than 40 percent unbalance is detected. With the CTs grounded, the PLU percent current was zero and mechanical power was 100 percent. Therefore, the turbine tripped per design even though no actual power unbalance condition existed. This procedure was found to be inadequate in that it allowed this calibration to be performed with the main turbine on-line and at greater than 40 percent rated thermal power.
REPORTABILITY ANALYSIS AND SAFETY ASSESSMENT
This event is reportable per 10 CFR 50.73 (a)(2)(iv)(A) because of unplanned actuations of Engineered Safety Feature systems. First, the reactor protection system actuated on turbine control valve fast closure when the main turbine tripped following a sensed power load unbalance. Second, HPCI and RCIC were manually initiated since the RFP could not inject due to the main condenser vacuum limiting RFP operation to less than or equal to 3500 rpm.
Fast closure of the turbine control valves is initiated whenever a power load unbalance is sensed. The turbine control valves close as rapidly as possible to prevent overspeed of the turbine-generator rotor. Valve closing causes a sudden reduction in steam flow that, in turn, results in a reactor vessel pressure increase. If the pressure increases to the pressure relief setpoints, some or all of the safety/relief valves will briefly discharge steam to the suppression pool (MIS Code BL).
Reactor Pressure did not increase to the level of the Safety Relief Valve's mechanical lift setpoint of 1150 psig.
However, SRVs B, D, F, G, A, C, K, and M lifted due to reactor pressure reaching their electrical lift setpoints.
Reactor pressure did not reach the electrical setpoint for the other three SRVs, E, H, and L. The solenoid operated electrical lift feature for the SRVs is intended as a non-safety related, back-up system to the safety related mechanical lift system.
In this event, the reactor recirculation pumps tripped on End-of-Cycle Recirculation Pump Trip, initiated by the turbine control valve fast closure. Reactor scram and recirculation pump trip initiation by turbine control valve fast closure prevents the core from exceeding thermal hydraulic safety limits following a main generator or main turbine trip. A reactor scram is initiated on turbine control valve fast closure. The scram, along with the reactor recirculation pump trip system, ensures that the minimum critical power ratio safety limit is not exceeded.
The main turbine tripped as designed in response to the sensed power load unbalance. The turbine trip actuated the reactor protection system and scrammed the reactor. Vessel water level was maintained well above the top of the active fuel throughout the transient and never decreased to the reactor scram actuation setpoint. The water level decrease was terminated prior to reaching the automatic initiation set point for HPCI and RCIC when Operations personnel manually initiated these systems. Therefore, there were no automatic safety system actuations on low water level.
Typically, the bottom head region of the pressure vessel experiences rapid cooling following a scram coincident with a trip of the reactor recirculation pumps. This cooling is the result of the loss of effective water mixing due to the trip of the recirculation pumps and increased cold water flow from the control rod drive (HIS Code AA) system following a scram. In this event, the temperature in the vessel bottom head region decreased by 129 degrees Fahrenheit in one hour. However, a bounding analysis indicated cool down up to 165°F in one hour will not place unacceptable stress on components of the reactor coolant system.
Based upon the preceding analysis, it is concluded this event had no adverse impact on nuclear safety. The analysis is applicable to all power levels.
CORRECTIVE ACTIONS
Maintenance procedure 57CP-CAL-010-2, "Esterline Angus Megavar and KV Recorder," has been revised to prevent performance while the reactor is operating.
2N33-F001, steam seal feed valve has been repaired.
ADDITIONAL INFORMATION
Other Systems Affected: No systems other than those already mentioned in this report were affected by this event.
A review for impact on Unit 1 was performed. Procedure 57CP-CAL-010-1 was reviewed by system engineering and it was determined that due to the Mark VI turbine controls modification the same vulnerability to cause a turbine control valve fast closure and resulting reactor scram does not exist.
This event was due to an inadequate procedure. There were no equipment failures that caused this event.
Failed Components Information: None Commitment Information: This report does not create any permanent licensing commitments.
Previous Similar Events in the last two years in which the reactor scrammed automatically while critical were reported in the following Licensee Event Reports:
1-2005-002, dated 12-21-2005 Previous Similar Events in the last two years in which an ECCS system has actuated either automatically of manually:
1-2004-004, dated 5-5-2004