|Donald C. Cook Nuclear Plant Unit 2|
|Reporting criterion:||10 CFR 50.73(a)(2)(iv)(A), System Actuation|
|ENS 49220||10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation, 10 CFR 50.72(b)(3)(iv)(A), System Actuation|
|3162013001R00 - NRC Website|
Conditions Prior to Event The Unit 2 reactor [RCT] was operating at 100 percent power.
Description of Event
On July 28, 2013, Cook Nuclear Plant (CNP) Unit 2 reactor was operating at 100 percent power.
At 0931, air operated valve 2-MRV-411, West Bypass Steam to Right Moisture Heating Coils Shutoff Valve MMO-431 Bypass Valve [SB][V], failed to the closed position from the normally full open position. Closure of the valve initiated a secondary plant transient, causing 2-HE-4A, Condensate Heater 4A [SD][HX], to reach a low-low level condition, which resulted in a trip of 2-PP-22N, North Heater Drain Pump [SN][P] at 0936. Operations reduced turbine load by 25 MW in response to the North Heater Drain Pump trip. At 1010 an Auxiliary Equipment Operator reported that valve 2-MRV-411 was locally observed in the closed position.
Controller demand for 2-MRV-411 was observed full open and in manual control.
At 1012, heater level fluctuations occurring in 2-HE-4B, Condensate Heater 4B [SD][HX], resulted in a heater low-low level condition and tripped 2-PP-22S, South Heater Drain Pump [SN][P]. Operations initiated an additional turbine load reduction of 25 MW following the trip of South Heater Drain Pump. At 1017, 2-PP-1W, West Main Feedwater Pump [SJ][P], automatically tripped on low suction pressure.
All three Auxiliary Feedwater (AFW) pumps [BA][P] were started in accordance with the Loss of One Main Feedwater Pump procedure. The Unit Supervisor directed a manual reactor trip if steam generator levels lowered to 23% (automatic reactor trip setpoint is 22%). At 1018 Steam Generator #4 [SG] level lowered to 23% and Operators manually tripped the reactor.
All safety systems responded normally following the reactor trip with the exception of valve 2-FMO-221, Turbine Driven Auxiliary Feedwater Pump PP-4 Discharge to Steam Generator 2-OME-3-2 Control Valve [BA][V], which did not position as required, following the manual reactor trip. The affected discharge valve was normally open as expected during the AFW pump start and then closed instead of throttling to a preset intermediate position upon high flow rate. The Turbine Driven Auxiliary Feedwater Pump (TDAFP) [BA][P] was declared inoperable. Feedwater flow to 2-OME-3-2, Steam Generator #2 [SG], was maintained by 2-PP-3E, East Motor Driven Auxiliary Feedwater Pump [BA][P].
In accordance with 10 CFR 50.72(b)(2)(iv)(B), Event Notification 49220 was submitted on July 28, 2013, at 1343 to report the actuation of the Reactor Protection System. The specified system actuation of the AFW System was also reported in accordance with 10 CFR 50.72(b)(3)(iv)(A).
Cause of Event
The initiating event was a failure of air operated valve 2-MRV-411 failing to the closed position due to fretting of the control air signal air line by a tubing support bracket.
The Root Cause of the event was determined to be a lack of procedural guidance and adequate controls for the Condensate Heater Condensate Bypass Control Valve, 2-CRV-224 [SD][V], automatic control functions.
Analysis of Event
The event was analyzed regarding the contributing factors leading up to the manual reactor trip on lowering steam generator levels caused by the low suction pressure automatic trip of the West Main Feedwater pump.
Investigation following the event concluded that 2-CRV-224 [SD][V] did not open to help prevent a low feedwater suction pressure from tripping the main feedwater pump. The associated controller [SD][PMC] was found at a setpoint lower than the intended design, which prevented 2-CRV-224 from opening.
Controller upgrades occurred in 1995 and 2003 that changed the methodology for setpoint inputs. The root cause was stated as a lack of procedural guidance and controls for the valve controller functions following controller upgrades.
A failure of 2-LPD-320N, North Heater Drain Pump PP-22N Discharge Check Valve [SN][CKV], was discovered following the manual reactor trip. The check valve was found in a partially open position which allowed a diversion of a portion of condensate flow away from the main feedwater pump suction.
The TDAFP discharge valve that did not automatically position as expected following the manual reactor trip was found with limit switches adjusted incorrectly. The incorrect limit switch settings did not affect the main feedwater pump suction during the event or impact feedwater flow to #3 steam generator. The limit switches were adjusted and tested satisfactory.
A risk impact review for this event indicates the trip was uncomplicated. All control rods [JC][JD] inserted, the main turbine-generator [TB][TG] tripped, offsite power transferred and remained energized by the station Reserve Auxiliary Transformers [EB][XFMR] as designed. No safety injection or engineered safety feature actuations occurred or were warranted beyond those expected for a nominal reactor trip. Main feedwater function was recoverable if it had been needed. The control room operators did not enter any additional emergency operating procedures after the trip, except those optimally expected. The operators noted that 2-FMO-221 (TDAFP Discharge to #3 SG flow control valve) ran fully closed rather than stopping at its intermediate flow retention setting. Discussion with the control room operators afterward and review of the Reactor Trip Report indicates there was sufficient flow from the motor driven AFW pumps such that the closed AFW valve had no significant impact on plant trip response. The operators were not required to manually reopen valve 2-FMO-221 immediately following the trip. The operators subsequently opened 2-FMO-221 to re-align the TDAFP to an available status and the valve did respond as expected to manual operation. The operators could have opened the valve during the event for additional AFW flow to #3 SG if needed.
Overall, the trip was an uncomplicated event with a malfunction of one AFW valve. The crew had observed and was aware of the AFW valve condition, and could have manually used the valve if needed. For these reasons, this trip did not pose any significant risk.
Completed Corrective Actions
Plant procedures were revised to provide guidance to establish and maintain condensate bypass valve 2-CRV-224 controller setpoint to modulate beginning at 240 psig.
North Heater Drain Pump discharge check valve 2-LPD-320N was replaced.
Failed control air tubing supply line to 2-MRV-411 was replaced.
The motor operated valve actuator limit switch settings for 2-FMO-221 were adjusted and tested.
Planned Corrective Actions
No additional corrective actions are planned.
Previous Similar Events
LERs for CNP Unit 1 and Unit 2 were reviewed for the previous five years and found no similar events.