05000325/LER-2009-002
| Brunswick Steam Electric Plant (Bsep), Unit 1 | |
| Event date: | 07-08-2009 |
|---|---|
| Report date: | 09-08-2009 |
| Reporting criterion: | 10 CFR 50.73(a)(2)(iv)(B), System Actuation 10 CFR 50.73(a)(2)(iv)(A), System Actuation |
| 3252009002R00 - NRC Website | |
Energy Industry Identification System (EIIS) codes are identified in the text as [XX].
Introduction Initial Conditions At the time of the event, Units 1 and 2 were in Mode 1, operating at approximately 100 percent of Rated Thermal Power (RTP). The Unit 1 B loop of the Low Pressure Coolant Injection (LPCI) system [BO] was under clearance for planned maintenance. No other Unit 1 or Unit 2 major equipment was inoperable.
Reportability Criteria This event resulted in the automatic actuation of Emergency Diesel Generator 2 (EDG 2) [EK] and various Primary Containment Isolation System (PCIS) [JM] isolations. As such, this event is being reported in accordance with 10 CFR 50.73(a)(2)(iv)(A) as an event that resulted in valid actuation of systems listed in 10 CFR 50.73(a)(2)(iv)(B). The NRC was initially notified of this event on July 8, 2009 (i.e., Event Number 45190). Due to the shared configuration of the onsite AC Electrical Distribution System [EB], this event is applicable to both Units 1 and 2.
Event Description
On July 7, 2009, technicians began a preventive maintenance activity, in accordance with plant procedure OPIC-CNV023, "Calibration of Westinghouse & Scientific Columbus Teleductors." The activity was to calibrate six emergency bus E2 voltage transducers. The activity was expected to be a two day evolution.
On July 8, 2009, at 1013 hours0.0117 days <br />0.281 hours <br />0.00167 weeks <br />3.854465e-4 months <br /> Eastern Daylight Time (EDT), during calibration of the 1-E2-AG6-VTR transducer (i.e., the last of the transducers to be calibrated), electrical power was lost to the 4160V emergency bus E2 [EB]. EDG 2 automatically started and re-energized the E2 bus.
The loss of power to emergency bus E2 resulted in isolation signals to Unit 1 PCIS Group 2 (i.e., Drywell Equipment and Floor Drain, Residual Heat Removal (RHR) Discharge to Radwaste, RHR Process Sample, and Traversing Incore Probe), Group 3 (i.e., Reactor Water Cleanup), Group 6 (i.e., Containment Atmosphere Control/Dilution, Containment Atmosphere Monitoring, and Post Accident Sampling Systems), and Group 10 (i.e., Drywell Pneumatics). Appropriate primary containment isolation valves closed.
Other Unit 1 actuations included the Reactor Building Ventilation System [VA] isolation (i.e., Secondary Containment isolation), automatic start of both trains of the Standby Gas Treatment System [BH] and automatic start of both trains of the Control Room Emergency Ventilation System [VI]. The affected equipment responded as designed.
Per design, no Unit 2 isolations or actuations occurred.
Event Description (continued) It has been determined that the loss of power to emergency bus E2 occurred as a result of a blown fuse in the C phase of emergency bus E2, which in turn caused a loss of power to the emergency bus E2 and the E2 master/slave breaker to trip. In preparation for calibrating the 1-E2-AG6-VTR transducer, OPIC-CNV023 called for the opening of knife switch TS-1, however, the technicians incorrectly opened knife switch TS-2.
As a result, arcing occurred when test equipment was connected to an energized circuit. This caused the blown fuse in the C phase of emergency bus E2.
After completing replacement of the blown fuse, emergency bus E2 was re-aligned to its normal offsite power source at 2307 hours0.0267 days <br />0.641 hours <br />0.00381 weeks <br />8.778135e-4 months <br /> EDT on July 8, 2009.
Event Cause The root cause of this event is inadequacies associated with procedure OPIC-CNV023 and the associated work order used to perform the preventive maintenance (PM) task.
The investigation identified the following weaknesses in procedure OPIC-CNV023: (1) the procedure contained less than adequate information for craft personnel to adequately assess risk significance of the task, (2) the component tag number labeling in the procedure does not match the tag number labeling on the cabinet in the field, (3) multiple-action steps are not written in the order of task performance, and (4) steps that must be performed have been omitted, (i.e., removing and installing test switch covers). Additionally, the associated PM model work order is written in a manner that creates a pattern matching bias. For instance, all but one instrument on the PM model work order uses the same test switch. As such, the procedure and associated work order were inadequate to ensure successful completion of the task.
The investigation of this event also identified other human performance weaknesses that took place during the calibration effort. The technicians performing this calibration had very limited experience in calibration of instruments on the emergency busses. It was determined that before work began, the supervisor decided not to perform a full pre-job briefing per plant procedure OAI-122, "Pre-Job Briefings & Post Job Critiques," but rather allowed the technicians to perform a Simple Task Brief. A full pre-job briefing would have been more appropriate and could have identified some of the work package deficiencies as well as provided an opportunity to re-enforce expectations concerning the use of human performance tools.
Although the event occurred as a result of procedure and work order inadequacies, a contributing cause has been attributed to supervisory oversight.
Safety Assessment The safety significance of this event is considered minimal. EDG 2 started and loaded per design. All other automatic actuations functioned properly upon the interruption of power to emergency bus E2.
Corrective Actions
The following corrective actions to prevent recurrence have been identified.
- Procedure OPIC-CNV023 will be revised to address inadequacies identified during the investigation of this event. This revision will be completed by December 2, 2009.
- The PM model work order will be revised to address pattern matching bias concerns. This revision will be completed by December 2, 2009.
In addition, the following corrective actions have been identified.
- Human performance has been recognized as an area requiring improvement at BSEP. As a result, a Human Performance Improvement Plan has been developed. This plan strives for event-free performance through reducing the frequency of errors and strengthening defenses, thereby optimizing the performance of individuals, leaders, and the organization. The Human Performance Improvement Plan will improve individual behaviors by communicating established standards and expectations, monitoring and coaching performance to the established standards and expectations, then making adjustments as they are identified for continuous improvement. The Human Performance Improvement Plan is specifically focused on addressing weaknesses in (1) site leadership's practices with respect to recognizing and reinforcing human performance behaviors and standards, (2) risk identification and establishment of mitigation barriers, (3) procedure use and adherence, and (4) work control and implementation of human performance barriers.
Implementation of the plan is being tracked by Nuclear Condition Report (NCR) 340674.
Previous Similar Events
A review of LERs and corrective action program condition reports for the past three years identified the following similar events.
- LER 2-2006-002 dated January 10, 2007, "Manual Scram Due to Conductivity Increase," documents a conductivity excursion which lead to a manual scram. The root cause of the event was determined to be failure to have procedural guidance to inspect the condenser water boxes for missing tube plugs following a Loss of Offsite Power (LOOP) event. The corrective actions associated with improvements. As such, they could not have reasonably been expected to prevent the condition reported in this LER.
- LER 2-2006-003, dated February 23, 2007, "Automatic Reactor Scram due to Trip from Neutron Monitoring System," documents a reactor scram generated by the Neutron Monitoring System Oscillation Power Range Monitors. The root cause of the event was inadequate incorporation of Operating Experience into plant procedures and training. The corrective actions to prevent Previous Similar Events (continued) recurrence included development of a formal process to evaluate correspondence from non-typical sources, such as Boiling Water Reactor Owners Group committees, and improvement of procedures associated with single loop operation. As such, they could not have reasonably been expected to prevent the condition reported in this LER.
- LER 2-2007-001, dated May 22, 2007, "Operation Prohibited by Technical Specification 3.3.1.2, Source Range Monitor Instrumentation," documents an event where a control rod was withdrawn a single notch in a fueled quadrant of the core where there was not an operable source range monitor.
The root cause of this event was inadequate procedures. The corrective action to prevent recurrence implemented enhancements to the control rod drive operating procedures. As such, they could not have reasonably been expected to prevent the condition reported in this LER.
- LER 1-2009-001, dated March 22, 2009, " Loss of Control Room Air Conditioning and Emergency Ventilation System," documents a loss of Control Room Emergency Ventilation and Control Room Air Conditioning due to loss of control air. The root cause of this event was determined to be the failure to incorporate system operating parameters for temperature into the associated operating procedures. The corrective action to prevent recurrence implemented enhancements to appropriate operating and annunciator procedures. As such, they could not have reasonably been expected to prevent the condition reported in this LER.
Commitments No regulatory commitments are contained in this report.