05000400/LER-2010-004
| Docket Numbersequential Revmonth Day Year Year Month Day Yearnumber No. N/A 05000 | |
| Event date: | 11-05-2010 |
|---|---|
| Report date: | 1-4-2011 |
| Reporting criterion: | 10 CFR 50.73(a)(2)(iv)(A), System Actuation |
| 4002010004R00 - NRC Website | |
Energy Industry Identification System (El IS) codes are identified in the text within brackets [].
I. DESCRIPTION OF EVENT
On November 5, 2010 at 0104 the station entered Abnormal Operating Procedure AOP-025 due to loss of 1B-SB safety bus [EB].
The plant was in a refueling outage (Mode 5) with the Reactor Coolant System (RCS) [AB] at approximately 80 psig, core exit thermocouples approximately 129 degrees F, and pressurizer [PZR] water level at 100%, when the B Emergency Diesel Generator (EDG) [DG] automatically started. Loads were properly started by the B sequencer [JE]. Prior to the event, the B Residual Heat Removal pump (RHR) [BP] was in service providing core cooling and the B safety train was the protected train in accordance with key safety function status sheet configuration 15A-5A "Mode 5 RCS loops not filled, not vented". This temporary interruption in power to the bus also resulted in the loss of the B RHR pump and associated shutdown cooling for approximately 3 minutes. The B RHR pump was restarted at 0107 in accordance with AOP-025 section 3.2 step 13. Core exit thermocouples showed a pre-event temperature of 129 degrees F and a maximum temperature of 130 degrees F while B RHR was secured. At the time of the event work order 1843093-02 was in progress to perform post maintenance testing of generator 86G1B lock out relay. The technicians were taking a voltage measurement on TBE-87 inside the generator relay panel which is in the circuit for CR3/1748 which opens breaker 125 on a sensed loss of offsite power. The meter being used for this voltage check had one lead connected to a ground point, and the other lead was connected to TBE-87. The inadvertent contact between TBE-87 and TBE-86 resulted in the loss of the 1B-SB safety bus [EB].
II. CAUSE OF THE EVENT
Root Cause
The root cause of the event is the lack of guidance in PLP-400, Post Maintenance Testing, for establishing PMT instructions for complex relay replacements.
Primary Contributing Cause The use of a standard alligator clip lead allowed contact between terminal TBE-87 and terminal TBE-86 in the generator relay panel.
III. SAFETY SIGNIFICANCE
Actual Safety Consequences:
There were no safety significant consequences as a result of this event. Automatic starting and loading of an EDG and temporary loss of shutdown cooling are analyzed for the Harris Plant (HNP) and are described in the FSAR. The plant is designed for both of these events and it responded as designed for given conditions. All available equipment operated per design. Decay heat removal was lost for approximately three minutes with an approximate one degree F rise in core exit temperature being observed during the time the B RHR pump was not running. The Main Control Room staff took prompt action to maintain key plant parameters within normal control bands for the existing plant conditions without challenge to plant or personnel safety.
Potential Safety Consequences:
The potential safety consequences under alternate conditions are bounded by plant design. The, most credible alternate condition would be the failure of the B RHR pump to start. In this condition, the Main Control Room staff would start the A RHR pump per plant procedures. Therefore, no significant safety consequences exist under alternate scenarios that would place the plant in a condition beyond its design bases.
This report is submitted pursuant to 10 CFR 50.73(a)(2)(iv)(A) for the automatic actuation of B EDG.
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IV. CORRECTIVE ACTIONS
Immediate Corrective Actions
The plant entered AOP-025, Loss of One Emergency AC Bus (6.9KV) or One Emergency DC Bus (125V). After approximately three minutes the B Residual Heat Removal pump was started per AOP-025 step 13. The Emergency Action Level (EAL) network for loss of the 1B-SB 6.9kV Bus was reviewed, no classification was required.
Corrective Actions to prevent recurrence Revise plant procedure PLP-400 to establish guidance for PMT development when replacing complex relays such as 86 UV or 86 lock out relays.
Planned Actions
1.Review this event with applicable personnel.
2. Implement the use of improved electrical test leads at HNP.
V. PREVIOUS SIMILAR EVENTS
Crystal River Unit 3 - Reactor Trip Caused by Group 7 Rods Dropping into the Core The cause of the dropped rods event in August, 2009 was the failure of the programmer. The cause of that failure is a gross over-current of the output driver during implementation of PM-126. The over-current condition was triggered by inadvertent contact of an inadequately protected (fused) test jumper to an incorrect/unintended point. The jumper made inadvertent contact with a positive voltage/current source. Although not a Harris Nuclear Plant event, this occurred in the Progress Energy fleet and actions were in place to apply the lessons learned regarding use of specialized test leads to all fleet nuclear plants. This action did not prevent the Harris event because the actions to procure and use standard leads were planned but not yet implemented.