Event Notification Report for March 26, 2011
U.S. Nuclear Regulatory Commission
Operations Center
EVENT REPORTS FOR
03/25/2011 - 03/26/2011
Power Reactor
Event Number: 46701
Facility: DIABLO CANYON
Region: 4 State: CA
Unit: [] [2] []
RX Type: [1] W-4-LP,[2] W-4-LP
NRC Notified By: DOUGLAS DYE
HQ OPS Officer: HOWIE CROUCH
Region: 4 State: CA
Unit: [] [2] []
RX Type: [1] W-4-LP,[2] W-4-LP
NRC Notified By: DOUGLAS DYE
HQ OPS Officer: HOWIE CROUCH
Notification Date: 03/26/2011
Notification Time: 21:10 [ET]
Event Date: 03/26/2011
Event Time: 14:49 [PDT]
Last Update Date: 03/26/2011
Notification Time: 21:10 [ET]
Event Date: 03/26/2011
Event Time: 14:49 [PDT]
Last Update Date: 03/26/2011
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):
GREG PICK (R4DO)
GREG PICK (R4DO)
| Unit | SCRAM Code | RX Crit | Initial PWR | Initial RX Mode | Current PWR | Current RX Mode |
|---|---|---|---|---|---|---|
| 2 | M/R | Y | 88 | Power Operation | 0 | Hot Standby |
MANUAL REACTOR TRIP WHEN A LOSS OF THE 2-1 MAIN FEEDWATER PUMP OCCURRED
"This notification provides the 4-hour non-emergency event report for the manual reactor trip of Diablo Canyon Power Plant Unit 2 in accordance with 10 CFR 50.72(b)(2)(iv)(B) 'RPS Actuation (scram)'. Additionally, this notification provides the 8-hour non-emergency event report of the automatic actuation of the auxiliary feedwater system as a result of the reactor trip in accordance with 10 CFR 50.72(b)(3)(iv)(A) 'Specified System Actuation'.
"On March 26, 2011 at 1449 PDT operators at Diablo Canyon Power Plant Unit 2 manually initiated a reactor trip in response to loss of main feedwater pump 2-1. The main feedwater pump is believed to have tripped due to non-radioactive water spray on its control console. The water spray was caused by leakage from the flange of the relief valve on the feedwater heater 2-1A condenser dump valve line. Emergency plan activation was not required.
The unit is stable in mode 3 (Hot Standby) with offsite power being supplied to all buses via the 230 kV startup circuit. Diesel generators 2-1 and 2-2 remain OPERABLE in standby. Diesel generator 2-3 remains unavailable due to pre-planned maintenance. All rods fully inserted on the reactor trip. The reactor is being cooled by the auxiliary feedwater system with the condenser in service. All systems performed as designed with no unexpected pressure or level transients. ECCS actuation was not required. Automatic main feedwater isolation, auxiliary feedwater actuation, and steam generator blowdown isolation occurred as expected.
"Unit 1 was unaffected by this event and remains at 100% power."
The licensee has notified the NRC Resident Inspector, San Luis Obispo County, and the State of California.
"This notification provides the 4-hour non-emergency event report for the manual reactor trip of Diablo Canyon Power Plant Unit 2 in accordance with 10 CFR 50.72(b)(2)(iv)(B) 'RPS Actuation (scram)'. Additionally, this notification provides the 8-hour non-emergency event report of the automatic actuation of the auxiliary feedwater system as a result of the reactor trip in accordance with 10 CFR 50.72(b)(3)(iv)(A) 'Specified System Actuation'.
"On March 26, 2011 at 1449 PDT operators at Diablo Canyon Power Plant Unit 2 manually initiated a reactor trip in response to loss of main feedwater pump 2-1. The main feedwater pump is believed to have tripped due to non-radioactive water spray on its control console. The water spray was caused by leakage from the flange of the relief valve on the feedwater heater 2-1A condenser dump valve line. Emergency plan activation was not required.
The unit is stable in mode 3 (Hot Standby) with offsite power being supplied to all buses via the 230 kV startup circuit. Diesel generators 2-1 and 2-2 remain OPERABLE in standby. Diesel generator 2-3 remains unavailable due to pre-planned maintenance. All rods fully inserted on the reactor trip. The reactor is being cooled by the auxiliary feedwater system with the condenser in service. All systems performed as designed with no unexpected pressure or level transients. ECCS actuation was not required. Automatic main feedwater isolation, auxiliary feedwater actuation, and steam generator blowdown isolation occurred as expected.
"Unit 1 was unaffected by this event and remains at 100% power."
The licensee has notified the NRC Resident Inspector, San Luis Obispo County, and the State of California.
Power Reactor
Event Number: 46881
Facility: BROWNS FERRY
Region: 2 State: AL
Unit: [] [2] []
RX Type: [1] GE-4,[2] GE-4,[3] GE-4
NRC Notified By: RICKY GIVENS
HQ OPS Officer: HOWIE CROUCH
Region: 2 State: AL
Unit: [] [2] []
RX Type: [1] GE-4,[2] GE-4,[3] GE-4
NRC Notified By: RICKY GIVENS
HQ OPS Officer: HOWIE CROUCH
Notification Date: 05/24/2011
Notification Time: 18:09 [ET]
Event Date: 03/26/2011
Event Time: 20:05 [CDT]
Last Update Date: 05/24/2011
Notification Time: 18:09 [ET]
Event Date: 03/26/2011
Event Time: 20:05 [CDT]
Last Update Date: 05/24/2011
Emergency Class: NON EMERGENCY
10 CFR Section:
50.73(a)(1) - INVALID SPECIF SYSTEM ACTUATION
10 CFR Section:
50.73(a)(1) - INVALID SPECIF SYSTEM ACTUATION
Person (Organization):
KATHLEEN O'DONOHUE (R2DO)
KATHLEEN O'DONOHUE (R2DO)
| Unit | SCRAM Code | RX Crit | Initial PWR | Initial RX Mode | Current PWR | Current RX Mode |
|---|---|---|---|---|---|---|
| 2 | N | N | 0 | Cold Shutdown | 0 | Cold Shutdown |
60-DAY TELEPHONIC NOTIFICATION OF INVALID PRIMARY CONTAINMENT ISOLATION SIGNAL DUE TO MAINTENANCE
"This 60-day telephone notification is being made per the reporting requirements specified by 10 CFR 50.73(a)(2)(iv) and 10 CFR 50.73(a)(1) to describe an invalid actuation of multiple secondary containment isolation signals affecting more than one system.
"On March 26, 2011, at 2005 hours Central Daylight Time (CDT), following planned maintenance activities, an unplanned actuation of secondary containment isolation signals affecting containment isolation valves in more than one system occurred and resulted in a partial Primary Containment Isolation System (PCIS) (Group 6, Secondary Containment) isolation and receipt of a Unit 2 Reactor Building Ventilation Abnormal Alarm. Unit 2 was in Mode 4, in a refueling outage, and at 0 percent power (0 MWT). Units 1 and 3 were both in Mode 1 and operating at approximately 100 percent power.
"Plant Conditions, which initiate PCIS (Group 6, Secondary Containment) actuations, are Low Reactor Vessel Water Level, High Drywell Pressure, High Reactor Zone Exhaust Radiation, or High Refueling Floor Exhaust Radiation. At the time of the event, these conditions did not exist; therefore, the partial actuations were invalid.
"For this occurrence, equipment normally affected by a complete Group 6 isolation responded as follows. Trains 'B' and 'C' of Standby Gas Treatment (SGT) System started while Train 'A' did not. Train 'B' of the Control Room Emergency Ventilation System (CREVS) started while Train 'A' did not. Secondary containment (Unit 1, 2, and 3 Reactor and Refuel Zones) normal ventilation isolated - fans stopped and dampers closed. Suppression Chamber Exhaust Inboard Isolation Valve 2-FSV-64-32 and Drywell Exhaust Inboard Isolation Valve 2-FCV-64-29 did not close. Because only a partial (B part of the logic) PCIS isolation was relayed, only the affected Group 6 equipment received the isolation demand. All equipment that received the isolation demand responded and performed as designed.
"Control Room personnel commenced actions of the applicable Alarm Response Procedure and the Group 6 Ventilation System Isolation Abnormal Operations Procedure. Train 'A' of the SGT System was later started.
"This event was entered in the Corrective Action Program as Service Request (SR) 344609 and Problem Evaluation Report (PER) 344680.
"There were no safety consequences or impact on the health and safety of the public as a result of these events.
"The NRC Resident Inspector has been notified of this event."
Although Group 6 PCIS testing occurred earlier that day, the licensee could not determine the signal initiator and considered it spurious.
"This 60-day telephone notification is being made per the reporting requirements specified by 10 CFR 50.73(a)(2)(iv) and 10 CFR 50.73(a)(1) to describe an invalid actuation of multiple secondary containment isolation signals affecting more than one system.
"On March 26, 2011, at 2005 hours Central Daylight Time (CDT), following planned maintenance activities, an unplanned actuation of secondary containment isolation signals affecting containment isolation valves in more than one system occurred and resulted in a partial Primary Containment Isolation System (PCIS) (Group 6, Secondary Containment) isolation and receipt of a Unit 2 Reactor Building Ventilation Abnormal Alarm. Unit 2 was in Mode 4, in a refueling outage, and at 0 percent power (0 MWT). Units 1 and 3 were both in Mode 1 and operating at approximately 100 percent power.
"Plant Conditions, which initiate PCIS (Group 6, Secondary Containment) actuations, are Low Reactor Vessel Water Level, High Drywell Pressure, High Reactor Zone Exhaust Radiation, or High Refueling Floor Exhaust Radiation. At the time of the event, these conditions did not exist; therefore, the partial actuations were invalid.
"For this occurrence, equipment normally affected by a complete Group 6 isolation responded as follows. Trains 'B' and 'C' of Standby Gas Treatment (SGT) System started while Train 'A' did not. Train 'B' of the Control Room Emergency Ventilation System (CREVS) started while Train 'A' did not. Secondary containment (Unit 1, 2, and 3 Reactor and Refuel Zones) normal ventilation isolated - fans stopped and dampers closed. Suppression Chamber Exhaust Inboard Isolation Valve 2-FSV-64-32 and Drywell Exhaust Inboard Isolation Valve 2-FCV-64-29 did not close. Because only a partial (B part of the logic) PCIS isolation was relayed, only the affected Group 6 equipment received the isolation demand. All equipment that received the isolation demand responded and performed as designed.
"Control Room personnel commenced actions of the applicable Alarm Response Procedure and the Group 6 Ventilation System Isolation Abnormal Operations Procedure. Train 'A' of the SGT System was later started.
"This event was entered in the Corrective Action Program as Service Request (SR) 344609 and Problem Evaluation Report (PER) 344680.
"There were no safety consequences or impact on the health and safety of the public as a result of these events.
"The NRC Resident Inspector has been notified of this event."
Although Group 6 PCIS testing occurred earlier that day, the licensee could not determine the signal initiator and considered it spurious.