05000263/LER-2001-011
| Monticello Nuclear Generating Plant | |
| Event date: | 10-23-2001 |
|---|---|
| Report date: | 12-21-2001 |
| Reporting criterion: | 10 CFR 50.73(a)(2)(iv)(A), System Actuation |
| 2632001001R00 - NRC Website | |
Description On October 23, 2001, while operating in coastdown to refueling at 98% power, a reactor scram occurred at 1712. The reactor scram signal was initiated by a Main Steams Isolation Valve2 (MSIV) closure, resulting from a Primary Containment Isolation System (PCIS)3 Group 1 Isolation signal.
In response to the scram and MSIV closure, a typical prompt reactor vessel level drop occurred due to the collapse of voids. When level reached +9 inches, Group 2 and 3 Containment Isolations occurred.
All safety systems required to respond to the scram functioned properly. The Safety Relief Valves4 (SRV) cycled six times to control reactor pressure in the first 24 minutes following the scram.
The scram recovery was affected by the Feedwater5 Regulating Valves6 (FRV) locking in an open position and not responding to water level rising after the initial void collapses. The reactor feedwater pumps' tripped on high reactor level less than one minute after the scram occurred. By procedure, an operator would have been required to trip one feed pump as level returned to normal operating levels. Since both feed pumps tripped, the procedure required an operator to manually restart a reactor feed pump. This occurred at approximately 1737, corresponding to the end of the last SRV cycle.
1 EllS System Code: SB 2 EIIS Component Code: ISV 3 EIIS System Code: JM 4 EllS Component Code: RV 5 EllS System Code: SJ 6 EIIS Component Code: LCV EIIS Component Code: P
Cause
The cause of the event was personnel error. A non-licensed utility Radiation Protection Specialist (RPS) inadvertently contacted a structural support brace of sensitive instrument rack C-126 on the east side of the 935 foot elevation of the reactor building.
A Level 1 Condition Report and associated investigation were conducted. The review found that the RPS had determined a contaminated area existed that required posting behind instrument rack C-126. In accordance with Radiation Protection Procedures, the RPS proceeded to setup a contaminated area boundary in a narrow passage area behind the instrument panel.
Four stanchions were to be utilized to provide the placement of radiation barrier rope to define the contaminated boundary. These stanchions are odd-shaped and heavy, with a base approximately one foot square and a three-foot vertical pipe.
The RPS chose to move the four stanchions, aware of the nearby instrument panel's sensitivity to physical shock. While moving the fourth stanchion into place, the RPS lost his balance and made physical contact with a support for the instrument panel. The physical contact (bump) caused a PCIS Group 1 isolation and resultant reactor scram.
The RPS was qualified to perform the task and was under no time pressure to complete the task. He was fully aware of the sensitivity of the instrument to physical contact and moved the stanchions cautiously. A fitness for duty test was performed with negative results.
Two causal factors to the event were identified. The first was inadequate work practices. The work practice used by the RPS to transport the stanchions was inadequate. This was due to overconfidence by the RPS in his ability to move the barriers without affecting the critical instrumentation.
The second factor was a work environment susceptible to incidental contact. There was no physical barrier in the vicinity of the specific brace or any other support that would prevent activation of the equipment as the result of incidental contact. Posted warnings in the area were determined to be inadequate.
With respect to the FRV lock up, subsequent investigation identified a deficiency in the software for the Digital8 Feedwater Control System9 (DFCS) as the cause of the FRV lockup.
The deficiency was corrected prior to restart.
8 EIIS Component Code: DCC 9 EIIS System Code: JB
Event Analysis
Analysis of Reportability The event is reportable under 10CFR 50.73(a)(2)(iv)(A), as an automatic activation of the Reactor Protection System and reactor scram, and as Group 1, 2 and 3 Containment Isolations.
The FRV lock up was determined not to be reportable.
The event does not constitute a safety system functional failure.
Safety Significance
The safety significance of the event is considered to be low, based on the operating crew successfully completing the abnormal operating procedure for a reactor scram, including manual run back of the recirculation pumps and restarting a reactor feed pump. Therefore, the health and safety of the public was not affected by the event.
A risk assessment performed by the Monticello Probabilistic Risk Assessment (PRA) Group concluded that this event is of low safety significance. This conclusion was based on the following considerations:
1'. The Monticello PRA model includes an initiating event for MSIV closure. The event that initiated this unplanned shutdown (bumping an instrument that caused a Group I isolation, MSIV closure, and reactor scram) would be grouped into the MSIV closure initiating event.
Recent operating history, including this event, is well within the bounds of the initiating event frequency used in the PRA model.
2. A review of the event also reveals that there were no equipment failures that resulted in the unavailability of systems modeled in the Monticello PRA. The feedwater pumps tripped, and one was restarted, just as is assumed in the PRA model for MSIV closure events. Therefore the event had negligible risk impact with regard to equipment reliability.
3. A quantitative assessment of the event was performed to confirm that the event had little impact on safety. This assessment was performed by calculating the probability of core damage, given that the event occurred. The result of this assessment is that the event was found to be of low safety significance.
Corrective Actions
Cause 1: Inadequate work practices. The worker was counseled by Site management, including the Site Vice President, regarding the need to perform work activities in a deliberate controlled manner with due consideration of the risk of actions taken.
Cause 2: Work Environment. An action was assigned to provide a physical barrier and signs requiring Shift Supervisor authorization to enter the work area behind the C-126 instrument rack. The C-126 instrument rack is unique in that normal access behind the rack is necessary for routine work activities.
Failed Component Identification Advanced Process Automation and Control System (APACS), manufactured by Moore Products Company for the DFCS.
Previous Similar Events
None within approximately ten years.
Past Corrective Action Analysis A reactor scram related to physical agitation of equipment has not occurred in over ten years.
Corrective actions resulting from four scrams from 1987 to 1990 were effective for a significant time period. The circumstances of this event were different from the previous scrams in that the worker was aware of the sensitivity of the critical instruments, and was utilizing what he considered appropriate care for the work being performed.
The work environment issue for this specific instrument panel has not been addressed to the extent it will be now. Although the area had warning signs for sensitivity, the signage will be improved, and the area will have more restrictive controls for the performance of work.
An action will be pursued to evaluate the effectiveness of previous actions to prevent instrument perturbations.