05000413/LER-2003-001
| Catawba Nuclear Station, Unit 1 | |
| Event date: | 02-04-2003 |
|---|---|
| Report date: | 03-24-2003 |
| Reporting criterion: | 10 CFR 50.73(a)(2)(iv)(A), System Actuation |
| 4132003001R00 - NRC Website | |
BACKGROUND
This event is being reported under 10 CFR 50.73(a)(2)(iv)(A), any event or condition that resulted in manual or automatic actuation of the RPS including reactor scram and reactor trip, and pressurized water reactor (PWR) auxiliary feedwater system.
Catawba Nuclear Station Unit 1 is a Westinghouse PWR [EIIS: RCT].
The purpose of the Main Feedwater (MFW) System [EIIS: SJ] is to operate in conjunction with the Condensate System [EIIS: KA] to return condensate from the condenser hotwells through the condensate polishing demineralizers and the regenerative feedwater heating cycle to the steam generators [EIIS: SG] while maintaining proper water inventories throughout the cycle. The MFW System consists of two steam _turbine driven feedwater pumps [EIIS: P]; two stages of high pressure feedwater heaters [EIIS: HTR]; and piping, valves [EIIS: V], and instrumentation.
Normally, both MFW pumps are in operation with each pump providing half' the required feedwater flow. The feedwater passes through the high pressure feedwater heaters to a final feedwater header, where final feedwater temperature is equalized. The feedwater is then admitted to the steam generators via four feedwater lines, each of which contains a feedwater control valve and a feedwater flow nozzle. Feedwater flow to the individual steam generators is controlled by a three element feedwater control system using feedwater flow, steam generator water level, and main steam flow as control parameters for the steam generator feedwater control valves. Because of its digital design, Catawba's feedwater control system is referred to as the Digital Feedwater Control System (DFCS).
A safety function of the MFW System is to isolate the steam generators on a feedwater isolation signal. One situation in which MFW isolation is required is to prevent overfilling the steam generators should the normal means of controlling steam generator level malfunction. During an overfill situation, MFW isolation is actuated by a steam generator level high-high signal. A steam generator level high-high signal will also trip the MFW pumps and the main turbine [EIIS: TRB]. Above 69% power, a trip of the main turbine will result in a trip of the reactor.
The Auxiliary Feedwater (AFW) System [EIIS: BA] is the assured source of feedwater to the steam generators during accident conditions. The AFW System consists of two motor driven pumps and one steam turbine driven pump; various safety and non-safety related sources of AFW; and associated piping, valves, and instrumentation. There are various signals which will automatically start the AFW System. In particular, the motor driven AFW pumps are designed to start upon trip of both MFW pumps.
When this event occurred, Unit 1 was operating in Mode 1 at 100% power.
Except as stated in the Event Description, no structures, systems, or components were out of service that had any significant effect on the event.
EVENT DESCRIPTION
(Certain event times are approximate) Date/Time � Event Description February 2, 2003/0220 � Control room received an annunciator indicating trouble with DFCS. It was determined that MFW header pressure transmitter [EIIS: PT] 1CFPT5141 had failed.
February 2-3, 2003 In response to the failure of transmitter 1CFPT5141, plant personnel evaluated available options and developed a plan for replacing the transmitter with the unit on line. The plan was discussed with all affected personnel and site management.
Necessary procedure changes to support the replacement were made, reviewed, and approved. A pre-job brief was held with affected personnel.
February 4, 2003/1002 � DFCS was in automatic control as transmitter 1CFPT5141 was being returned to service following replacement. As the transmitter was valved in, the hydraulic interaction with nearby transmitters 1CFPT5140 and 1CFPT5142 (due to the three transmitters sharing a common impulse line) caused the MFW header pressure indication to DFCS to fluctuate. MFW pump turbine speed increased and the MFW control valves opened in response to the transient. When the difference between transmitters 1CFPT5140 and 1CFPT5142 (the controlling channel) reached 50 psi, DFCS switched from automatic to manual control, as designed.
1005 � Control room operators attempted to control steam generator levels manually.
High-high level on steam generator B resulted in turbine trip, which resulted in reactor trip, since the unit was above 69% power. The
- high-high level on steam generator B also resulted in a trip of both MFW pumps. This started the motor driven AFW pumps, as designed.
1013 � Entered reactor trip response procedure and stabilized unit parameters.
1242 � 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> telephone notification made to NRC per 10 CFR 50.72(b)(2)(iv)(B) (actuation of reactor protection system while reactor is critical).
CAUSAL FACTORS
The root cause of this event was determined to be an inadequate understanding of the DFCS response to a common impulse line hydraulic interaction. The inadequate understanding was present throughout various site groups and management. This inadequate knowledge led to the development of an inadequate procedure and work plan, which upon implementation, resulted in the unanticipated MFW transient. In addition: 1) the level of risk management and oversight for the transmitter replacement activity was insufficient to ensure that the actions of a multi-disciplined team would result in event free operation; and 2) due to insufficient training, control room operator attempts to reestablish control of steam generator water levels were not effective.
CORRECTIVE ACTIONS
Immediate:
1. Operations entered the reactor trip response procedure and stabilized the unit parameters following the reactor trip.
Subsequent:
1. A team was formed to review this event and to determine its cause.
2. Procedures associated with calibration and maintenance of DFCS transmitters were placed on hold pending a-review by Engineering.
Planned:
1. Training will be developed and conducted for appropriate plant personnel concerning potential interactions and consequences of common impulse line hydraulics and the design and response of the DFCS.
2. Procedures associated with calibration and maintenance of transmitters which share common impulse lines will be reviewed and revised as necessary.
3. The risk management and oversight processes used to ensure emergent work activities do not result in plant events will be strengthened to minimize the potential for events of this type.
4. Additional training will be developed and conducted for control room operators concerning response to MFW transients.
The planned corrective actions are being addressed via the Catawba Corrective Action Program. There are no NRC commitments contained in this LER.
SAFETY ANALYSIS
The Solid State Protection System functioned as designed upon receipt of the reactor trip signal. The reactor trip breakers opened within 150 milliseconds of the trip signal as designed. All control and shutdown rods inserted as designed. The rod drop time was within the Technical Specification requirement. Nuclear instrumentation response was normal following the trip.
Reactor coolant temperature control functioned as designed following the trip. In addition, primary system pressure control functioned as expected. No pressurizer power operated relief valves (PORVs) or code safety valves lifted. The pressurizer spray valves and pressurizer heaters controlled primary system pressure as designed. Pressurizer level control was normal following the trip.
Following the MFW isolation signal, all valves associated with the MFW isolation closed within the required time frame. Secondary system control functioned as designed with two exdieptions.
Sections of the MFW System were overpressurized in response to the MFW isolation. The peak pressure encountered on ehe system was 1807 psig. The design pressure on this section ofi the MFW System is 1385 psig; however, the observed peak pressure was less than the hydrostatic test pressure of 1.5 times the design pressure for this section of the system. Following the event, a system walkdown was performed and data trending showed no concern with the structural integrity of the MFW System due to the overpressurization transient. Also, valve 1SB15 (one of the steam dump to condenser valves) opened out of sequence and never reached its full open position as required. An air leak was found on this valve and the leak was repaired. No steam generator PORVs or main steam code safety valves lifted in response to this event. The AFW System response to this event was as designed. The motor driven AFW pumps automatically started on the loss of both MFW pumps. The steam turbine driven AFW pump did not receive a start signal, nor was it required to start in response to this event. AFW flow to all four steam generators was within the acceptable range.
In summary, the post-trip transient response remained within the bounds of the Updated Final Safety Analysis Report.
The core damage significance of this event has been evaluated quantitatively by considering the following:
- A loss of MFW initiating event.
8
- Actual moderator temperature coefficient value at the time of the trip
- Actual plant configuration and maintenance activities at the time of the trip.
- Credit for recent Nuclear Service Water to AFW System modifications to minimize clam intrusion.
The conditional core damage probability for this event is 2.0E-07, which is less than the accident sequence precursor threshold of 1.0E- 06.
.The dominant core damage sequences associated with this event have the significant containment safeguards systems available. These 'include the containment spray and hydrogen mitigation systems.- Furthermore, most have low to moderate reactor coolant system pressures at reactor vessel failure. Sequences of this nature contribute insignificantly to the Large Early Release Frequency (LERF), which is dominated by the interfacing systems loss of coolant accidents and seismic initiating events. Therefore, this event is judged to have no significance with respect to the LERF for Catawba.
While this event resulted in an automatic isolation of MFW, post- trip MFW flow could have been reestablished by the control room operators if desired. Therefore, this event did not meet the initiating event criterion for a reactor trip with loss of the normal heat sink.
This event was of no significance with respect to the health and safety of the public.
ADDITIONAL INFORMATION
Within the last three years, four other reactor trip events occurred from power operation at Catawba as follows:
the turbine trip instrumentation inputs to the Solid State Protection System logic. This resulted from an electrical short within an electrical connector on the normally energized turbine electrical trip solenoid valve. The root cause of the event was a misapplication of the connector insert insulating material.
transient caused by rainwater entering the turbine building and affecting control of one of the MFW pumps. The root cause of this event was determined to be inadequate oversight of a turbine building roof modification.
trip. The root cause of this event was determined to be incomplete troubleshooting analysis associated with the main turbine protection system mechanical trip solenoid valve.
reactor coolant flow when the 2D reactor coolant plump 6900 VAC feeder breaker opened in response to protective relay actuation caused by an electrical fault internal to the pump, motor.
The corrective actions taken in response to these events would not have prevented this latest event from occurring. Therefore, this event was determined to be non-recurring in nature.
Energy Industry Identification System (EIIS) codes are identified in the text as [EIIS: XX]. This event is not considered reportable to the Equipment Performance and Information Exchange (EPIX) program.
This event did not involve a Safety System Functional Failure.
There were no releases of radioactive materials, radiation exposures, or personnel injuries associated with this event.