05000270/LER-2006-001
| Oconee Nuclear Station | |
| Event date: | |
|---|---|
| Report date: | |
| Reporting criterion: | 10 CFR 50.73(a)(2)(iv)(A), System Actuation |
| 2702006001R00 - NRC Website | |
EVALUATION:
BACKGROUND
This event is reportable per 10CFR 50.73(a)(2)(iv)(A) as an event which resulted in a Reactor Trip at Oconee Nuclear Station.
As described in Section 7.2.2 of the Updated Final Safety Analysis Report (UFSAR), the Reactor Protective System (RPS) [EIIS:JC] has trip when two channels actuate. Each RPS channel laas multiple functions which will cause the channel to trip when the applicable setpoint is reached. One such function is Flux/Flow/Imbalance.
.Flux is neutron flux, which indicates Reactor power. Flow is ,Reactor Coolant System (RCS) [EIIS:AB] flow. Imbalance is reactor 'power imbalance, defined as power in the top half of the core minus the power in the bottom half of the core, expressed as a percentage of full power. When the neutron flux signal exceeds a variable setpoint calculated from measured reactor coolant 'flow and reactor power imbalance, the associated protective channel trips.
The Integrated Control System (ICS) [EIIS:JA] provides automatic control to maintain the unit at the selected power level. In response to certain plant events, the ICS system will attempt to runback (automatically reduce power) to a predetermined setpoint.
A Reactor Coolant Pump (RCP) [EIIS:P] trip is one of the events which initiates an ICS runback. However, this runback is not expected to occur fast enough to avoid a reactor trip if the' RCP trip occurs at 100% power.
The Anticipated Transient Without Scram (ATWAS) [EIIS:JC] Mitigation System Actuation Circuitry (AMSAC) and the Diverse Scram System (DSS) provide non-safety back-up signals to mitigate an ATWAS event. DSS will trip the reactor in the event that the RPS fails to successfully trip the reactor and the transient results in a very high RCS pressure.
Prior to this event Unit 2 was operating at 100% power in Mode 1.
Operations began activities to remove the Inadequate Core Cooling Monitor (ICCM) [EIIS:XI] train 2B from service for planned calibrations. � The ICCM 2B train affects the operability of a number of safety systems, so the 2B trains of those systems were also logged out of service. However, both the Operations Control Room Senior Reactor Operator and the Instrument and Electrical (I&E) technicians involved in the planned work confirm that only the ATWAS systems (AMSAC and DSS) had physically been disabled (bypassed) at the time of the trip. As a result, at the time of the trip, no other safety systems or components were out of service that would have contributed to this event.
EVENT DESCRIPTION
At approximately 12:45 on April 12, 2006, a team of I&E technicians began work to inspect and test an AC watt transducer [EIIS:TD] for RCP 2B2 as part of a maintenance activity. They found no abnormal readings. By 13:35 hours they had completed their data collection task and were preparing to disconnect their test equipment.
At 13:35:51 RCP 2B2 tripped, which caused the ICS to initiate a runback. However, this runback is not expected to occur fast enough to avoid a reactor trip if the RCP trip occurs at 100% power.
At 13:35:57 RPS channels A and D tripped on Flux/Flow/Imbalance, which tripped the reactor.
Post trip response was normal. All control rod drive [EIIS:AA] breakers tripped and control rods dropped into the core. The turbine tripped as expected due to the reactor trip. Unit auxiliary power automatically transferred from the Normal to Start- up source as expected. Main Feedwater [EIIS:SJ] remained in service, with flow demand automatically reduced by the ICS. RCS temperature, RCS pressure, RCS inventory, Main Steam (MS) [EIIS:SB] Pressure, and Steam Generator (SG) [EIIS:SG] inventory remained within expected limits. No actuations or actuation demands occurred related to Emergency Feedwater [EIIS:BA] or Engineered Safeguards [EIIS:JE] (i.e. Emergency Core Cooling [EIIS:BG and BP], Containment Isolation [EIIS:NH], Containment Spray/Cooling [EIIS:BE and BK], and Emergency Power [EIIS:EK]). Since the RPS operated properly to trip the unit, the fact that AMSAC and DSS were out of service did not contribute to the event.
One control rod (Rod 7 Group 3) position indication showed the rod to be at 20% withdrawn while a second control rod (Rod 7 Gr6Up 2) position indication showed an out-limit light. The operators opened valves connecting the Borated Water Storage Tank to the suction of the operating High Pressure Injection (HPI) [EIIS:CB and BG] pump to increase the boron concentration in the RCS to ensure adequate shutdown margin. Further evaluation of other indications, including individual rod in-limit and group in-limit lights and core response, confirmed that both rods were fully inserted.
One Main Steam Relief Valve [EIIS:RV] (MSRV) (2MS-11, on the "B" header) did not reseat at 995 psig as expected. MS header pressure was lowered to 970 psig (89.8% of set pressure), at which point the valve reseated. The relief valve operation was reviewed by Valve Engineering personnel. Although the recorded reseat value (89.8%) was slightly lower than expected, it was not outside of the potential relief valve drift (3%) and blowdown (10%). Therefore, Valve Engineering concluded that the slight recorded variation of the reseat value (0.2% from the procedure value of 90%), was not indicative of a valve problem.
The post-trip investigation validated the plant response described above and confirmed the cause of the trip.
CAUSAL FACTORS
The Reactor Trip was a direct result of the trip of the 2B2 Reactor Coolant Pump. A root cause team was created to determine the cause of the RCP trip.
The Reactor Coolant Pump is powered from a three phase 6900 volt source. The pump protective relaying uses current transformer (CT) [ICT] circuits which allow control and monitoring devices to operate at reduced voltages with currents proportional to the current in the corresponding 6900 volt phase. One set, using 2 phases, of these CT circuits includes watt transducers which provide computer indication of the power used by the pump. This CT circuit also contains an electronic device which provides several protective relaying functions, including a phase balance current relay.
In this CT circuit only two phases, X and Z, are monitored and the third phase is calculated from the two monitored phases. The phase balance current relay function compares the measured and calculated currents. If an imbalance is detected between any two phases, real or calculated, the relay will trip the pump.
The work activity in progress involved connection of test instruments to one of the power transducers providing computer indication. The root cause team found that, while the test equipment was in place, there was a loss of isolation between the power transducer being tested and the operating CT circuit. The test setup involves isolation of the CT circuit by inserting a test plug into a knife switch. It is indeterminate if the loss of isolation occurred due to an interaction between the test plug and the knife switch or due to a problem with the knife switch itself.
The loss of isolation caused a portion of the current normally in the neutral circuit of the CT circuit to interact with the current injected by the test equipment and pass through ground as a return path, reducing the current in the neutral circuit. When the current in the neutral leg dropped, the protective relay detected the current unbalance and provided a trip signal to the pump breaker.
CORRECTIVE ACTIONS
Immediate:
1. Operators took appropriate actions to bring the unit to stable hot shutdown (Mode 3).
Subsequent:
1. The instrument procedure in use during this activity was placed on technical hold (which prevents its use) to address risk issues associated with performing this procedure on vital equipment at power.
Planned:
1. Additional troubleshooting will be performed during the next Unit 2 refueling outage to investigate potential failure modes involving the knife switch/test plug interactions.
This corrective action is not considered to be an NRC Commitment item. There are no NRC Commitment items contained in this LER.
SAFETY ANALYSIS
This event did not include a Safety System Functional Failure.
A reactor trip is an anticipated transient and is considered the safe end state following many plant transients and most accident scenarios addressed in Chapter 15 of the UFSAR.
As described in Section 7.2.2 of the UFSAR, the flux/flow/imbalance parameter is one of the inputs which initiates a reactor trip.
During this event, the unit tripped in response to actual plant parameters. All required systems and equipment operated as designed to mitigate the consequences of the RCP trip and stabilize the unit in Mode 3.
The minor discrepancies observed (two control rod position indications conflicting with the rod fully inserted indications, and the slight shift in reseat pressure of 2MS-11) had no adverse impact on the event.
Core Damage Significance (CDS) Impact The CDS of this event has been evaluated quantitatively by considering the following:
- A reactor trip initiating event (reactor trip with no complications)
- Actual plant configuration and maintenance activities at the time of the trip
- Current Probabilistic Risk Assessment (PRA) model accounting for any necessary updates since the last revision The event was an uncomplicated reactor scram with no appreciable impact on any safety systems.
Large Early Release Frequency (LERF) Significance Impact The LERF significance of this event has been evaluated quantitatively using the same considerations as for the CDS Impact.
The event has an insignificant impact on the core damage risk.
Therefore, there was no actual impact on the health and safety of the public due to this event.
ADDITIONAL INFORMATION
A review of Unit trip events over the last five years indicates that there have been no similar reactor trip events at Oconee within that time period.
There were no releases of radioactive materials, radiation exposures or personnel injuries associated with this event.
The trip of the reactor constitutes a Maintenance Rule functional failure and is considered reportable under the Equipment Performance and Information Exchange (EPIX) program.