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> October 20
Event Notification Report for October 20, 2006
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U.S. Nuclear Regulatory Commission
Operations Center
Event Reports For
10/19/2006 - 10/20/2006
** EVENT NUMBERS **
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| Power Reactor |
Event Number: 42919 |
Facility: THREE MILE ISLAND
Region: 1 State: PA
Unit: [1] [ ] [ ]
RX Type: [1] B&W-L-LP,[2] B&W-L-LP
NRC Notified By: CRAIG SMITH
HQ OPS Officer: JOHN MacKINNON
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Notification Date: 10/19/2006
Notification Time: 11:17 [ET]
Event Date: 10/19/2006
Event Time: 09:21 [EDT]
Last Update Date: 10/19/2006
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Emergency Class: NON EMERGENCY
10 CFR Section:
26.73 - FITNESS FOR DUTY
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Person (Organization):
RICHARD BARKLEY (R1)
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| Unit |
SCRAM Code |
RX CRIT |
Initial PWR |
Initial RX Mode |
Current PWR |
Current RX Mode |
| 1 |
N |
Y |
100 |
Power Operation |
100 |
Power Operation |
Event Text
NON-LICENSED SUPERVISOR FAILED FOLLOW-UP TESTING
"A non-licensed employee supervisor had a confirmed positive test for alcohol during a follow up fitness for duty test. The employee's access to the plant has been put on administrative hold, and the individual has been escorted offsite."
The NRC Resident Inspector was notified of this by the licensee.
Contact the Headquarter Operations Officer for further details. |
| Power Reactor |
Event Number: 42920 |
Facility: PALO VERDE
Region: 4 State: AZ
Unit: [ ] [ ] [3]
RX Type: [1] CE,[2] CE,[3] CE
NRC Notified By: DAN HAUTALA
HQ OPS Officer: PETE SNYDER
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Notification Date: 10/19/2006
Notification Time: 17:27 [ET]
Event Date: 10/19/2006
Event Time: 11:47 [MST]
Last Update Date: 10/19/2006
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Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(2)(iv)(B) - RPS ACTUATION - CRITICAL
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Person (Organization):
DALE POWERS (R4)
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| Unit |
SCRAM Code |
RX CRIT |
Initial PWR |
Initial RX Mode |
Current PWR |
Current RX Mode |
| 3 |
M/R |
Y |
100 |
Power Operation |
0 |
Hot Standby |
Event Text
MANUAL REACTOR TRIP FOLLOWING THE TRIP OF SECONDARY CONDENSATE PUMPS
"The following event description is based on information currently available. If through subsequent reviews of this event, additional information is identified that is pertinent to this event or alters the information being provided at this time, a follow-up notification will be made via the ENS or under the reporting requirements of 10CFR50.73.
"On October 19, 2006, at approximately 1147 Mountain Standard Time (MST) Palo Verde Unit 3 plant operators manually tripped the reactor from approximately 100% rated thermal power. The reactor was tripped when lowering hotwell levels caused two condensate pumps to trip. The preliminary cause for the lowering hotwell level was the hotwell draw-off valve spuriously failing open. Unit 3 was at normal temperature and pressure prior to the trip. All CEAs inserted fully into the reactor core. This was an uncomplicated reactor trip. No ESF actuations occurred and none were required. Safety related buses remained energized during and following the reactor trip. The offsite power grid is stable. No significant LCOs have been entered as a result of this event. There was no loss of normal heat removal capabilities, or loss of any safety functions associated with this event. No major equipment was inoperable prior to the event that contributed to the event. The event did not result in any challenges to fission product barriers and there were no adverse safety consequences as a result of this event. The event did not adversely affect the safe operation of the plant or the health and safety of the public.
"The Resident Inspector was informed of the Unit 3 reactor trip and this notification."
The current decay heat removal path is auxiliary feedwater supplying water to the steam generators steaming to the condenser. Emergency Diesel Generators are available. |
| Power Reactor |
Event Number: 42921 |
Facility: RIVER BEND
Region: 4 State: LA
Unit: [1] [ ] [ ]
RX Type: [1] GE-6
NRC Notified By: DAN WILLIAMSON
HQ OPS Officer: PETE SNYDER
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Notification Date: 10/19/2006
Notification Time: 22:35 [ET]
Event Date: 10/19/2006
Event Time: 17:56 [CDT]
Last Update Date: 10/19/2006
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Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(2)(iv)(A) - ECCS INJECTION
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
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Person (Organization):
DALE POWERS (R4)
NILESH CHOKSHI (NRR)
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| Unit |
SCRAM Code |
RX CRIT |
Initial PWR |
Initial RX Mode |
Current PWR |
Current RX Mode |
| 1 |
A/R |
Y |
100 |
Power Operation |
0 |
Hot Shutdown |
Event Text
AUTOMATIC SCRAM FOLLOWING SPONTANEOUS FEEDWATER VALVE CLOSURE
"At 1756 CDT with the plant operating at 100% power, a reactor scram occurred in response to a reactor water level 3 signal from an apparent loss of feedwater. Both feedwater injection lines isolated when isolation valves were inadvertently closed, The cause of the isolation valve closure is under investigation.
"When reactor water level lowered to level 2, high pressure core spray (HPCS) initiated automatically and recovered water level. The reactor core isolation cooling system (RCIC) was tagged out for maintenance at the time of the event.
"Following the scram, main steam isolation valves isolated on low main steam header pressure. As a result, reactor pressure control was being controlled with the safety relief valves. SRV pressure control in turn led to EOP entry conditions on containment pressure and suppression pool level.
"Both feedwater lines were opened, and normal reactor level control was restored. The MSIV's were opened and pressure control was returned to the turbine bypass valves and the main condenser.
"Initial indications are that all plant equipment functioned as designed with the exception of the 'B' feed pump which experienced an apparent seal failure. The plant is stable in Mode 3. All plant conditions are understood.
"This event is being reported in accordance with 10CRF50.72(b)(2) as an RPS actuation and an injection of HPCS into the reactor vessel, and in accordance with 10CRF50.72(b)(3) as a loss of safety function of HPCS, as it was manually disabled during recovery from the event. The HPCS Injection valve was manually overridden closed for 76 minutes. In addition,' containment isolation valves in multiple systems actuated in response to the RPV level 2 signal."
Reactor vessel water level lowered to below level 2. Decay heat is being removed by normal feedwater to the reactor vessel steaming to the main condenser. Offsite power is available and stable. Emergency Diesel Generators are available.
The licensee notified the NRC Resident Inspector. |
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