|Power Reactor||Event Number: 47590|
|Facility: WOLF CREEK
Region: 4 State: KS
Unit:  [ ] [ ]
RX Type:  W-4-LP
NRC Notified By: TERRY DAMASHEK
HQ OPS Officer: HOWIE CROUCH
|Notification Date: 01/13/2012
Notification Time: 16:05 [ET]
Event Date: 01/13/2012
Event Time: 14:03 [CST]
Last Update Date: 01/13/2012
|Emergency Class: UNUSUAL EVENT
10 CFR Section:
50.72(a) (1) (i) – EMERGENCY DECLARED
50.72(b)(2)(iv)(B) – RPS ACTUATION – CRITICAL
50.72(b)(3)(iv)(A) – VALID SPECIF SYS ACTUATION
50.72(b)(3)(v)(A) – POT UNABLE TO SAFE SD
DALE POWERS (R4DO)
ART HOWELL (R4RA)
ERIC LEEDS (NRR)
JANE MARSHALL (IRD)
JOHN THORP (NRR)
FRED HILL (DHS)
KEVIN BISCOE (FEMA)
|Unit||SCRAM Code||RX CRIT||Initial PWR||Initial RX Mode||Current PWR||Current RX Mode|
|1||A/R||Y||100||Power Operation||0||Hot Standby|
|NOTIFICATION OF UNUSUAL EVENT AND REACTOR TRIP DUE TO LOSS OF OFFSITE POWER”At 1403 CST, Wolf Creek experienced a reactor trip due to loss of power in the switchyard. At 1415 CST, Wolf Creek declared a Notification of Unusual Event (NOUE) when it was determined that the switchyard would not be restored within 15 minutes.”All systems functioned as expected in response to this event and both Emergency Diesel Generators started and energized the safety-related buses.
“The plant is currently stable in Mode 3 and investigation into the cause for loss of power in the switchyard is underway.”
During the trip, all rods inserted into the core. No primary relief valves lifted as a result of the transient. Decay heat is being removed via the atmospheric steam dumps with auxiliary feedwater supplying the steam generators. The plant is stable at NOP/NOT. No safety significant equipment is reported out of service.
The licensee has notified state and local governments and the NRC Resident Inspector.
* * * UPDATE FROM DAVE DEES TO VINCE KLCO AT 1851 EST ON 1/13/12 * * *
At 1709 CST, the licensee exited the NOUE when power was restored to the east bus from offsite. Additionally, the licensee is reporting a loss of safe shutdown capability in accordance with 10CFR50.72(b)(3)(v)(A) due to the initial loss of offsite power.
The licensee has notified state and local governments, the NRC Resident Inspector, and will be issuing a press release on the event.
Notified R4DO (Powers), IRD (Marshall), NRR (Cheok), FEMA (Burckart) and DHS (Hill).
This is a highly unusual event, especially since there was no severe weather in the area at that time. The good news is that both Emergency Diesel Generators Worked and Offsite Power was Restored approximately 3 hours after the LOOP. The bad news is that we don’t know what caused the disruption of Switchyard Power.
Was there a CME in progress? Was the SCADA (Supervisory Control and Data Acquisition System – that controls the Switchyard Breakers) hacked?
The Wolf Creek Switchyard is typical of many nuclear stations in that multiple offsite power lines from different circuits feed the plant’s onsite power systems. (See Attached Drawing 2_5 Electrical.pdf)
According to the Wolf Creek FSAR (Final Safety Analysis Report, which is their license) Section 18.104.22.168.1 the Offsite Power System is highly reliable and meets the following (so what went wrong):
22.214.171.124.1 Safety Design Bases
SAFETY DESIGN BASIS ONE – Electrical power from the power grid to the plant site is supplied by two physically independent circuits designed and located so as to minimize the likelihood of simultaneous failure.
SAFETY DESIGN BASIS TWO – Each of these independent circuits has the capability to safely shut down the unit. The first preferred circuit, which is connected to the startup transformer, has the capacity to supply the startup and all the auxiliary loads (both group 1 and group 2 simultaneously) of the unit.
SAFETY DESIGN BASIS THREE – The second preferred power circuit, which supplies power to the ESF transformer, has the capacity to supply all the safety-related loads of the unit.
SAFETY DESIGN BASIS FOUR – The loss of the nuclear unit or the most critical unit on the grid will not result in the loss of offsite power to the Class IE busses.
According to the Wolf Creek FSAR Section 126.96.36.199, Transmission Network:
There are three 345-kV lines connecting the Wolf Creek 345-kV Substation to the area transmission system. The three lines are as follows:
a. Wolf Creek-LaCygne 345-kV Line:
58 miles long, connecting to the LaCygne Steam Electric
Station which has three additional 345-kV lines.
b. Wolf Creek-Rose Hill 345-kV Line:
98 miles long, connecting to the Rose Hill Substation
southeast of Wichita. Rose Hill Substation has two
additional 345-kV connections.
c. Wolf Creek-Benton 345-kV Line:
90 miles long, connecting to the Benton Substation
northeast of Wichita. Benton Substation has two
additional 345-kV lines, one of which is to the Wichita
345-kV Substation, near the Gordon Evans Steam Electric
The above 345-kV lines do not share common rights-of-way, do not have any crossovers, and are not in close proximity of one another, except in the immediate vicinity of the switchyard. In those areas, the design is such as to maintain clearances as required by the National Electric Safety Code (4).
Further, it describes that No One Line Failure Should Remove ALL OFFSITE POWER. Why did Wolf Creek Lose ALL OFFSITE POWER??
If one of the three 345-kV lines faulted, the breakers located at Wolf Creek Substation (Figure 8.2-5) would trip, deenergizing the line.
Any one of the two remaining incoming 345-kV Transmission lines at Wolf Creek Substation can carry the total ESF load required for safe shutdown by controlled switching of the Wolf Creek substation breakers, providing a separate transmission line feeding each ESF transformer.
Comment: This means that there should have been no Loss of Offsite Power due to any one or two of the three power supplies. In my opinion, this narrows the cause down to AT LEAST Breakers Breakers 345-70 and 345-40 (See Attached Drawing 2_5 Electrical.pdf). Possibly Breakers 345-50 and 345-60 also opened. The Specific Reactor Trip Signal would be the Reactor Coolant Pump Undervoltage Trip. From FSAR Chapter 7 (See Attached Table 7.2-1 Sheet 2 Item 12.).
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