Event Notification Report for April 28, 2009
U.S. Nuclear Regulatory Commission
Operations Center
EVENT REPORTS FOR
04/27/2009 - 04/28/2009
EVENT NUMBERS
45026450274502845029
Power Reactor
Event Number: 45026
Facility: NORTH ANNA
Region: 2 State: VA
Unit: [] [2] []
RX Type: [1] W-3-LP,[2] W-3-LP
NRC Notified By: PAGE KEMP
HQ OPS Officer: HOWIE CROUCH
Region: 2 State: VA
Unit: [] [2] []
RX Type: [1] W-3-LP,[2] W-3-LP
NRC Notified By: PAGE KEMP
HQ OPS Officer: HOWIE CROUCH
Notification Date: 04/28/2009
Notification Time: 12:00 [ET]
Event Date: 04/28/2009
Event Time: 09:53 [EDT]
Last Update Date: 04/28/2009
Notification Time: 12:00 [ET]
Event Date: 04/28/2009
Event Time: 09:53 [EDT]
Last Update Date: 04/28/2009
Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(3)(iv)(A) - VALID SPECIF SYS ACTUATION
10 CFR Section:
50.72(b)(3)(iv)(A) - VALID SPECIF SYS ACTUATION
Person (Organization):
MIKE ERNSTES (R2)
JOHN THORP (NRR)
MIKE ERNSTES (R2)
JOHN THORP (NRR)
| Unit | SCRAM Code | RX Crit | Initial PWR | Initial RX Mode | Current PWR | Current RX Mode |
|---|---|---|---|---|---|---|
| 2 | N | Y | 100 | Power Operation | 100 | Power Operation |
EMERGENCY DIESEL AUTO-START DUE TO LOSS OF THE "B" RESERVE STATION SERVICE TRANSFORMER
"At 0953 hours on April 28, 2009, a loss of the 'B' Reserve Station Service Transformer resulted in a Degraded Voltage / Under Voltage automatic start of the Unit 2 'H' Emergency Diesel Generator. The Unit 2 'H' Emergency Diesel Generator is operating and supplying electrical power to the Unit 2 'H' 4160 Volt Bus. The Unit 2 'B' Charging Pump (2-CH-P-1B) and the Unit 2 'B' Component Cooling Water Pump (2-CC-P-1B) also auto started in response to the loss of power. [Procedure] 0-AP-10, 'Loss of Electrical Power', was entered to address the loss of the normal power source for the Unit 2 'H' Emergency Bus.
"Unit 2 is operating at 100% power and an investigation has been initiated to determine the cause of the event and appropriate corrective actions.
"The NRC Resident Inspector has been notified."
There was no impact on Unit 1 operation. There was no loss of significant safety equipment as a result of the transformer loss. The licensee will continue to supply the 'H' 4160V bus with the 'H' Emergency Diesel Generator until compensatory measures are put in place. The loss of the 'H' bus places Unit 2 in a 72-hr shutdown action statement.
"At 0953 hours on April 28, 2009, a loss of the 'B' Reserve Station Service Transformer resulted in a Degraded Voltage / Under Voltage automatic start of the Unit 2 'H' Emergency Diesel Generator. The Unit 2 'H' Emergency Diesel Generator is operating and supplying electrical power to the Unit 2 'H' 4160 Volt Bus. The Unit 2 'B' Charging Pump (2-CH-P-1B) and the Unit 2 'B' Component Cooling Water Pump (2-CC-P-1B) also auto started in response to the loss of power. [Procedure] 0-AP-10, 'Loss of Electrical Power', was entered to address the loss of the normal power source for the Unit 2 'H' Emergency Bus.
"Unit 2 is operating at 100% power and an investigation has been initiated to determine the cause of the event and appropriate corrective actions.
"The NRC Resident Inspector has been notified."
There was no impact on Unit 1 operation. There was no loss of significant safety equipment as a result of the transformer loss. The licensee will continue to supply the 'H' 4160V bus with the 'H' Emergency Diesel Generator until compensatory measures are put in place. The loss of the 'H' bus places Unit 2 in a 72-hr shutdown action statement.
Power Reactor
Event Number: 45027
Facility: WOLF CREEK
Region: 4 State: KS
Unit: [1] [] []
RX Type: [1] W-4-LP
NRC Notified By: BILL STUCKER
HQ OPS Officer: HOWIE CROUCH
Region: 4 State: KS
Unit: [1] [] []
RX Type: [1] W-4-LP
NRC Notified By: BILL STUCKER
HQ OPS Officer: HOWIE CROUCH
Notification Date: 04/28/2009
Notification Time: 17:41 [ET]
Event Date: 04/28/2009
Event Time: 15:27 [CDT]
Last Update Date: 04/28/2009
Notification Time: 17:41 [ET]
Event Date: 04/28/2009
Event Time: 15:27 [CDT]
Last Update Date: 04/28/2009
Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(2)(iv)(B) - RPS ACTUATION - CRITICAL 50.72(b)(3)(iv)(A) - VALID SPECIF SYS ACTUATION
10 CFR Section:
50.72(b)(2)(iv)(B) - RPS ACTUATION - CRITICAL 50.72(b)(3)(iv)(A) - VALID SPECIF SYS ACTUATION
Person (Organization):
WAYNE WALKER (R4DO)
WAYNE WALKER (R4DO)
| 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 |
AUTOMATIC REACTOR TRIP ON LOW STEAM GENERATOR LEVEL DUE TO FEEDWATER VALVE FAILURE
"Wolf Creek experienced a reactor trip at 1527 CDT. The RX trip resulted from low level in 'B' S/G. [The] 'B' Main Feedwater Regulating Valve [FRV] AE FCV-520 closed causing a loss of feedwater to the 'B' S/G. Initial investigation indicates that the FRV controllers lost power due to blown primary and backup fuses associated with its card rack.
"One intermediate range instrument (SE NI-36) stuck at 10E-10 amps requiring manual energization of source range instruments.
"Aux Feedwater and Feedwater Isolation actuations occurred as expected."
All control rods inserted into the core during the trip. There were no relief or safety valves actuated during the transient. The electrical plant is in the normal shutdown alignment with offsite power supply emergency busses. Decay heat is being removed via the steam dumps to condenser. The licensee is investigating the cause of the blown fuses.
The licensee has notified the NRC Resident Inspector.
"Wolf Creek experienced a reactor trip at 1527 CDT. The RX trip resulted from low level in 'B' S/G. [The] 'B' Main Feedwater Regulating Valve [FRV] AE FCV-520 closed causing a loss of feedwater to the 'B' S/G. Initial investigation indicates that the FRV controllers lost power due to blown primary and backup fuses associated with its card rack.
"One intermediate range instrument (SE NI-36) stuck at 10E-10 amps requiring manual energization of source range instruments.
"Aux Feedwater and Feedwater Isolation actuations occurred as expected."
All control rods inserted into the core during the trip. There were no relief or safety valves actuated during the transient. The electrical plant is in the normal shutdown alignment with offsite power supply emergency busses. Decay heat is being removed via the steam dumps to condenser. The licensee is investigating the cause of the blown fuses.
The licensee has notified the NRC Resident Inspector.
Power Reactor
Event Number: 45028
Facility: QUAD CITIES
Region: 3 State: IL
Unit: [1] [2] []
RX Type: [1] GE-3,[2] GE-3
NRC Notified By: CURT STEFFES
HQ OPS Officer: HOWIE CROUCH
Region: 3 State: IL
Unit: [1] [2] []
RX Type: [1] GE-3,[2] GE-3
NRC Notified By: CURT STEFFES
HQ OPS Officer: HOWIE CROUCH
Notification Date: 04/28/2009
Notification Time: 19:34 [ET]
Event Date: 04/28/2009
Event Time: 18:34 [CDT]
Last Update Date: 05/18/2009
Notification Time: 19:34 [ET]
Event Date: 04/28/2009
Event Time: 18:34 [CDT]
Last Update Date: 05/18/2009
Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(3)(xiii) - LOSS COMM/ASMT/RESPONSE
10 CFR Section:
50.72(b)(3)(xiii) - LOSS COMM/ASMT/RESPONSE
Person (Organization):
DAVID HILLS (R3DO)
DAVID HILLS (R3DO)
| Unit | SCRAM Code | RX Crit | Initial PWR | Initial RX Mode | Current PWR | Current RX Mode |
|---|---|---|---|---|---|---|
| 1 | N | N | 0 | Refueling | 0 | Refueling |
| 2 | N | Y | 100 | Power Operation | 100 | Power Operation |
PROCESS COMPUTER REPLACEMENT IMPACTING ERDS AND SPDS
"During the current Unit 1 refueling outage, the plant process computer system will be taken out of service for approximately three weeks to support a planned modification which will install a new (upgraded) plant process computer system. During this timeframe, ERDS [Emergency Response Data System] and SPDS [Safety Parameter Display System] will be unavailable on Unit 1 which will remain in a shutdown condition. The computer outage will also render ERDS and SPDS unavailable on Unit 2 for approximately eight hours. Should the need arise, plant status information will be communicated to the NRC, [State, and local responders] using other available communication systems. The computer outage is expected to commence at 2200 CDT on 4/28/2009, and the process computer will be restored prior to Unit 1 startup currently scheduled for 05/21/2009.
"This event is reportable under 10 CFR 50.72(b)(3)(xiii) as a Major Loss of Assessment Capability."
The licensee will update this report when ERDS and SPDS is restored to Unit 2.
The licensee has notified the NRC Resident Inspector.
* * * UPDATE FROM C. STEFFES TO JOE O'HARA AT 0558 ON 4/29/09 * * *
"Follow-up notification - U2 ERDS and SPDS is restored as of 0315 CDT 4/29/09."
The licensee notified the NRC Resident Inspector.
Notified R3DO(Hills)
* * * UPDATE FROM BRIAN MAGNUSON TO JOE O'HARA AT 1509 EDT ON 5/18/09 * * *
"Follow-up notification - U1 ERDS and SPDS have been restored. Acceptance testing was completed on 5/15/09."
The licensee notified the NRC Resident Inspector.
Notified R3DO(Skokowski)
"During the current Unit 1 refueling outage, the plant process computer system will be taken out of service for approximately three weeks to support a planned modification which will install a new (upgraded) plant process computer system. During this timeframe, ERDS [Emergency Response Data System] and SPDS [Safety Parameter Display System] will be unavailable on Unit 1 which will remain in a shutdown condition. The computer outage will also render ERDS and SPDS unavailable on Unit 2 for approximately eight hours. Should the need arise, plant status information will be communicated to the NRC, [State, and local responders] using other available communication systems. The computer outage is expected to commence at 2200 CDT on 4/28/2009, and the process computer will be restored prior to Unit 1 startup currently scheduled for 05/21/2009.
"This event is reportable under 10 CFR 50.72(b)(3)(xiii) as a Major Loss of Assessment Capability."
The licensee will update this report when ERDS and SPDS is restored to Unit 2.
The licensee has notified the NRC Resident Inspector.
* * * UPDATE FROM C. STEFFES TO JOE O'HARA AT 0558 ON 4/29/09 * * *
"Follow-up notification - U2 ERDS and SPDS is restored as of 0315 CDT 4/29/09."
The licensee notified the NRC Resident Inspector.
Notified R3DO(Hills)
* * * UPDATE FROM BRIAN MAGNUSON TO JOE O'HARA AT 1509 EDT ON 5/18/09 * * *
"Follow-up notification - U1 ERDS and SPDS have been restored. Acceptance testing was completed on 5/15/09."
The licensee notified the NRC Resident Inspector.
Notified R3DO(Skokowski)
Power Reactor
Event Number: 45029
Facility: SEQUOYAH
Region: 2 State: TN
Unit: [1] [] []
RX Type: [1] W-4-LP,[2] W-4-LP
NRC Notified By: JAMES W. FULLER
HQ OPS Officer: JOE O'HARA
Region: 2 State: TN
Unit: [1] [] []
RX Type: [1] W-4-LP,[2] W-4-LP
NRC Notified By: JAMES W. FULLER
HQ OPS Officer: JOE O'HARA
Notification Date: 04/28/2009
Notification Time: 23:50 [ET]
Event Date: 04/28/2009
Event Time: 21:59 [EDT]
Last Update Date: 04/29/2009
Notification Time: 23:50 [ET]
Event Date: 04/28/2009
Event Time: 21:59 [EDT]
Last Update Date: 04/29/2009
Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(2)(iv)(B) - RPS ACTUATION - CRITICAL 50.72(b)(3)(iv)(A) - VALID SPECIF SYS ACTUATION
10 CFR Section:
50.72(b)(2)(iv)(B) - RPS ACTUATION - CRITICAL 50.72(b)(3)(iv)(A) - VALID SPECIF SYS ACTUATION
Person (Organization):
MIKE ERNSTES (R2DO)
MIKE ERNSTES (R2DO)
| Unit | SCRAM Code | RX Crit | Initial PWR | Initial RX Mode | Current PWR | Current RX Mode |
|---|---|---|---|---|---|---|
| 1 | M/R | Y | 18 | Power Operation | 0 | Hot Standby |
MANUAL REACTOR TRIP INITIATED AS A RESULT OF A LOSS OF CONDENSATE FLOW
"At 2159 Sequoyah Unit 1 was manually tripped from 18% reactor power. The manual trip was in response to automatic isolation of all intermediate pressure feedwater heater strings resulting in a loss of all condensate flow. The feedwater heater strings isolated following a manual trip of the main turbine. The main turbine was manually tripped in response to a failed open moisture separator reheater relief valve. At the time of the event Sequoyah Unit 1 was raising power following a refueling outage.
"Following the reactor trip, all safety-related equipment operated as designed. Auxiliary feedwater automatically actuated as expected on loss of the operating main feedwater pump.
"The plant is currently being maintained in Mode 3 at NOT/NOP, approximately 547 F and 2235 psig, with auxiliary feedwater supplying the steam generators and steam dumps to main condenser removing decay heat.
"Maintenance activities have been initiated to repair the MSR relief valve."
Electrical lineup is normal with all safety related equipment operable. Licensee is investigating the cause for the relief valve failure.
The NRC Resident Inspector has been notified.
"At 2159 Sequoyah Unit 1 was manually tripped from 18% reactor power. The manual trip was in response to automatic isolation of all intermediate pressure feedwater heater strings resulting in a loss of all condensate flow. The feedwater heater strings isolated following a manual trip of the main turbine. The main turbine was manually tripped in response to a failed open moisture separator reheater relief valve. At the time of the event Sequoyah Unit 1 was raising power following a refueling outage.
"Following the reactor trip, all safety-related equipment operated as designed. Auxiliary feedwater automatically actuated as expected on loss of the operating main feedwater pump.
"The plant is currently being maintained in Mode 3 at NOT/NOP, approximately 547 F and 2235 psig, with auxiliary feedwater supplying the steam generators and steam dumps to main condenser removing decay heat.
"Maintenance activities have been initiated to repair the MSR relief valve."
Electrical lineup is normal with all safety related equipment operable. Licensee is investigating the cause for the relief valve failure.
The NRC Resident Inspector has been notified.