05000261/LER-2013-003
| H. B. Robinson Steam Electric Plant, Unit No. 2 | |
| Event date: | |
|---|---|
| Report date: | |
| 2612013003R01 - NRC Website | |
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PLANT IDENTIFICATION
Westinghouse - Pressurized Water Reactor
BACKGROUND
At 1801 hours0.0208 days <br />0.5 hours <br />0.00298 weeks <br />6.852805e-4 months <br /> EST on 11/5/2013, with the unit in Mode I at 19% power and no involvement of out-of-service structures, systems or components, H. B. Robinson Steam Electric Plant (HBRSEP2) experienced an automatic reactor [RCT] trip while operators were transferring loads from the Unit Startup Transformer (SUT) [XFMR] to the Unit Auxiliary Transformer (UAT) [XFMR], in accordance with plant procedures. During the coordinated breaker operation, the reactor tripped when an anomaly occurred that running 'A' Main Feedwater Pump [P] tripped and caused the auto-start of the Auxiliary Feedwater System (AFW), which maintained Steam Generator [SG] water levels within the normal operating band. The 'A' Emergency Diesel Generator (EDG) [DG] auto-started and supplied the E-1 emergency bus [BU] as a result of the undervoltage transient.
The event was reported as a 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> Non-Emergency report per 10 CFR 50.72(b)(2)(iv)(B) due to the valid Reactor Protection System Actuation, and as an 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> Non-Emergency report per 10 CFR 50.72(b)(3)(iv)(A) due to the valid actuation of AFW and EDG auto- start and subsequent starting of required undervoltage loads, save 'A' Service Water (SW) Pump [P].
EVENT DESCRIPTION
At 1801 hours0.0208 days <br />0.5 hours <br />0.00298 weeks <br />6.852805e-4 months <br /> EST on 11/5/2013, with the Unit in Mode I at 19% power, HBRSEP2 experienced an automatic reactor trip while operators were transferring loads from the SUT to the UAT, in accordance with plant procedures. When the breaker (52/7) control switch [33] connecting the UAT to 4KV Bus I [BU] was taken to the 'close' position, indication on the Reactor Turbine Generator Board (RTGB) [MCBD] went from 'Open' to no indication. The 52/7 breaker's 'A' and 'C' phases closed, however 'B' phase operating rod had failed thereby inhibiting closure of the 'B' phase of the breaker. When the 52/7 control switch was returned to center position, breaker 52/12 [BKR], Incoming Line — Startup Transformer to 4KV Bus 2 [BU], tripped open per normal controls.
When breaker 52/12 opened, the 'B' phase current on Reactor Coolant Pumps (RCPs) [P] 'A' and 'C' dropped from approximately 500 amps to 76 amps. As designed, the relays for buses 1 and 2 tripped at a setpoint equal to 75% of nominal bus voltage. Reactor trip signal logic directs a reactor trip when 2 out of 3 4KV bus undervoltage relays trip. As a result of the '13' phase current drop on RCPs 'A' and 'C' and the subsequent trip of their associated undervoltage relays, an automatic reactor trip occurred. The turbine -[TG] automatically tripped due to the reactor trip causing the turbine stop valves [V] to close while the switchyard generator output breakers [BKR] 52/8 and 52/9 remained closed. With the turbine stop valves closed while the switchyard breakers were closed, the 60-second time delay pick-up relays initiated and timed out resulting in Generator Lockout Relay 86BU [RLY] tripping and closing breaker 52/12. Normal voltage was restored to 4KV buses 1 and 2 when breaker 52/12 closed. The voltage drop on the 4KV buses was sufficient to pick up the E-1 Bus undervoltage relays and initiate the blackout sequence on Bus E-l. The 'A' EDG auto-started and the `A' EDG load sequencer [PMC] loaded the required Engineered Safety Features loads onto E-1 with the exception of the 'A' SW Pump, which should have started as part of the sequence.
CAUSAL FACTORS
The root cause of the reactor trip event is advanced aging/fatigue of breaker 52/7 phenolic operating rod materials. Degradation of the rod materials caused failure of the '13' operating rod, which prevented closing of breaker 52/7. The failure of the 'A' SW Pump to start was due to a loose wire termination in the Emergency Control Station. The loose termination was attributed to a historical condition based on maintenance records searches going back more than fifteen years that did not identify any work performed in this particular Emergency Control Station.
CORRECTIVE ACTIONS
Completed:
1. Interim corrective action to address rod failure involved inspection and or replacement of operating rods in all four incoming breakers that are part of the fast bus transfer logic between the UAT and SUT.
2. The 'A' SW pump loose terminal connection in the Emergency Control Station was secured.
Planned:
1. Long term corrective action to prevent recurrence consists of replacing the operating rods for all 4KV incoming breakers with high tensile strength operating rods. Action Request (AR) 642282, Assignment Nos. 05 & 06.
SAFETY ANALYSIS
The risk consequences of this event were minimal based on the successful reactor trip and operators' ability to successfully start SW Pump 'EY [P] after SW Pump 'A' failed to start as expected. SW Pump 'A' remained available by a manual start and the transient was not complicated by additional equipment failures, malfunctions or human errors.
The reactor was at 19% power, and power ascension following completion of the refueling outage was in progress. Therefore, the decay heat load was well below that of a full-power trip after a period of operation and therefore well within the capabilities of the decay heat removal systems. In addition, AFW actuation had been demonstrated successfully in the previous period of plant operation, and no corrective or preventive maintenance had been performed on that system since that time. The Condensate Storage Tank [TK] capacity was adequate to remove the above decay heat load for an extended period before requiring refill. Backup sources of feedwater were available and important accident mitigation equipment remained available throughout this event.
ADDITIONAL INFORMATION
An internal Operating Experience (OE) search for related events at HBRSEP2 was conducted; no similar events were identified. An evaluation of external OE shows that this event would not have been prevented by available OE, therefore this was not considered a missed opportunity to utilize OE.
Energy Industry Identification System (EIIS) codes for systems and components relevant to this event are identified in the text of this document within brackets [].