05000220/LER-2005-003
| Docket Number Sequential Revmonth Day Year Year Month Day Yearnumber No. 05000 | |
| Event date: | 08-18-2005 |
|---|---|
| Report date: | 10-12-2005 |
| Reporting criterion: | 10 CFR 50.73(a)(2)(iv)(A), System Actuation |
| Initial Reporting | |
| 2202005003R00 - NRC Website | |
I. Description of Event
On August 18, 2005, at approximately 0950 hours0.011 days <br />0.264 hours <br />0.00157 weeks <br />3.61475e-4 months <br />, with Nine Mile Point Unit 1 (NMP1) operating at approximately 100 percent power, an automatic reactor scram was experienced. The scram occurred when 4160 VAC Power Board 11 (PB 11) was de-energized due to inadvertent protective over-current relay actuations. PB-11 provides power primarily to non safety related station power production equipment.The loss of PB-11 resulted in loss of power to one of the two sources of power to the Reactor Trip Bus (RTB). Concurrently with the loss of PB-11, planned surveillance testing was in progress that introduced a half-scram signal on the opposite (Channel 12) side of the RTB. With a half-scram signal on the Channel 12 side and loss of power on the Channel 11 side of the RTB, a full automatic reactor scram signal was generated. Had the surveillance testing on the Channel 12 side not been in progress, the loss of PB-11 and the downstream loads would have resulted in a significant power reduction or plant shutdown due to the loss of key components, including two strings of feedwater heating, a condensate pump, two reactor recirculation pumps, a service water pump, three drywell cooling fans, a condenser circulating water pump, and numerous other loads.
Following the automatic scram, all control rods fully inserted and the reactor was shutdown.
Reactor pressure did not rise above its initial value during the event. The turbine bypass valves were used to control reactor pressure during scram recovery and plant cool down. The cooldown rate was maintained less than 75 degrees Fahrenheit per hour.
Reactor water level lowered as expected following the scram to approximately 27 inches. The Low Reactor Level scram/High Pressure Coolant Injection (HPCI) initiation signals (53 inches) were received as expected. At the time of the scram, feedwater pump 11 was operating in manual and feedwater pump 12 was in standby. Upon HPCI initiation on the turbine trip signal, feedwater pump 12 started and injected. Feedwater pump 13 was secured and the HPCI logic was reset per procedure. Level continued to increase, reaching the High Reactor Level trip (-93 inches), at which time feedwater pump 12 tripped as designed. Maximum reactor water level reached was 104 inches, partially covering the Emergency Cooling System steam line nozzles but remaining below the Main Steam Line nozzles. The Emergency Cooling System was not required to operate for this event.
House loads on PB-12 transferred to offsite power as expected. PB-11 did not fast transfer per design as there was a lockout signal present to the offsite feeder breaker. There were no challenges to operators with respect to any electrical system beyond the initiating event.
II. Cause of Event
PB-11 was de-energized due to inadvertent protective over-current relay actuations that occurred when a 4.16 kV circuit breaker was being rolled into position within its PB-11 cubicle. Though workers did not recall creating any abnormal mechanical shock or jarring of the cubicle during this activity, analysis of the relays' physical and electrical characteristics, their control circuits, and surrounding electrical equipment, did not reveal any other credible mode of actuation. Actuation of multiple relays mounted on the cubicle door due to mechanical jarring was demonstrated using a mockup of a 4.16 kV power board and circuit breaker.
The underlying cause of the event is failure to adequately identify, assess, and mitigate potential risks of conducting work. The potential plant impacts that could be caused by inadvertent actuation or damage to surrounding equipment due to planned work were not identified or effectively managed.
III. Analysis of Event
This event is reportable in accordance with 10 CFR 50.73(a)(2)(iv)(A) as an event or condition that resulted in manual or automatic actuation of the Reactor Protection System (RPS). The automatic initiation of the HPCI system is also reportable under 10 CFR 50.73(a)(2)(iv)(A).
There were no actual safety consequences associated with this event. All available plant safety systems operated as designed and the operators effectively stabilized reactor parameters. Water level was maintained well above the top of active fuel throughout the event and was restored to its desired value without the need for emergency core cooling system actuation. The conditional core damage probability for the event was calculated as 2.8E-6.
Based on the above, the event did not pose a threat to the health and safety of the public or plant personnel.
IV. Corrective Actions
A. Action Taken to Return Affected Systems to Pre-Event Normal Status Plant shutdown was accomplished without assistance from PB-11. The over-current (51B) relays were tested and found to be within the design specification requirements. The power board was successfully re-energized from offsite power and loads were sequenced back on. No further issues were found with PB-11.
B. Action Taken or Planned to Prevent Recurrence NOTE: There are no NRC regulatory commitments in this Licensee Event Report.
Corrective actions have been developed such that, when completed, the work control process will yield consistent and effective risk assessments of work activities that will prevent the same type of inadvertent relay actuation and other similar events (i.e., events caused by failure to identify and effectively manage potential risks associated with surrounding equipment). These actions include:
- Identifying other relays on switchgear/power boards and other trip sensitive equipment that, if inadvertently impacted, could cause undesirable equipment and plant impacts. Based on potential risk impacts, compensatory measures will be established to prevent significant adverse plant impacts.
- Improving risk assessment effectiveness by work planners, supervisors, and operations staff through enhanced administrative controls, improved tools, training, and qualification, and by re-affirming the lead role of the Operations department in the risk assessment process.
V. Additional Information
A. Failed Components:
None
B. Previous similar events:
None C. Identification of components referred to in this Licensee Event Report:
Components IEEE 805 System ID IEEE 803.1 Function Reactor Protection System JC N/A High Pressure Coolant Injection System BJ N/A Feedwater System SJ N/A Emergency Cooling System BL N/A Main/Reheat Steam System SB N/A Control Rods M ROD Turbine Bypass Valves SB PCV Power Board EA SWGR Pump SJ, BJ P Breaker EA 52 Relay EA 51