Event Notification Report for June 20, 2011
U.S. Nuclear Regulatory Commission
Operations Center
EVENT REPORTS FOR
06/19/2011 - 06/20/2011
Power Reactor
Event Number: 46973
Facility: LIMERICK
Region: 1 State: PA
Unit: [1] [2] []
RX Type: [1] GE-4,[2] GE-4
NRC Notified By: BRANDON SHULTZ
HQ OPS Officer: STEVE SANDIN
Region: 1 State: PA
Unit: [1] [2] []
RX Type: [1] GE-4,[2] GE-4
NRC Notified By: BRANDON SHULTZ
HQ OPS Officer: STEVE SANDIN
Notification Date: 06/20/2011
Notification Time: 16:59 [ET]
Event Date: 06/20/2011
Event Time: 11:56 [EDT]
Last Update Date: 06/20/2011
Notification Time: 16:59 [ET]
Event Date: 06/20/2011
Event Time: 11:56 [EDT]
Last Update Date: 06/20/2011
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):
CHRISTOPHER CAHILL (R1DO)
CHRISTOPHER CAHILL (R1DO)
| Unit | SCRAM Code | RX Crit | Initial PWR | Initial RX Mode | Current PWR | Current RX Mode |
|---|---|---|---|---|---|---|
| 1 | N | Y | 100 | Power Operation | 100 | Power Operation |
| 2 | N | Y | 100 | Power Operation | 100 | Power Operation |
TECHNICAL SUPPORT CENTER CHARCOAL FAILED TESTING FOR METHYL IODINE
"During 24 month Technical Support Center (TSC) Ventilation System Charcoal Analysis testing, the charcoal failed due to Methyl Iodine penetration of 3.1%, which is above the procedural limit of 1%.
"The TSC non-emergency ventilation system remains functional. Charcoal replacement is planned.
"If an emergency is declared requiring TSC activation during this period, the TSC will be staffed and activated using existing Emergency Planning procedures and checklists. If radiological conditions require TSC facility evacuation prior to completion of charcoal replacement; the Station Emergency Director will evacuate and relocate the TSC staff in accordance with applicable site procedures."
The licensee informed the NRC Resident Inspector.
"During 24 month Technical Support Center (TSC) Ventilation System Charcoal Analysis testing, the charcoal failed due to Methyl Iodine penetration of 3.1%, which is above the procedural limit of 1%.
"The TSC non-emergency ventilation system remains functional. Charcoal replacement is planned.
"If an emergency is declared requiring TSC activation during this period, the TSC will be staffed and activated using existing Emergency Planning procedures and checklists. If radiological conditions require TSC facility evacuation prior to completion of charcoal replacement; the Station Emergency Director will evacuate and relocate the TSC staff in accordance with applicable site procedures."
The licensee informed the NRC Resident Inspector.
Power Reactor
Event Number: 46974
Facility: PERRY
Region: 3 State: OH
Unit: [1] [] []
RX Type: [1] GE-6
NRC Notified By: JEFF TUFTS
HQ OPS Officer: PETE SNYDER
Region: 3 State: OH
Unit: [1] [] []
RX Type: [1] GE-6
NRC Notified By: JEFF TUFTS
HQ OPS Officer: PETE SNYDER
Notification Date: 06/20/2011
Notification Time: 17:10 [ET]
Event Date: 06/20/2011
Event Time: 09:45 [EDT]
Last Update Date: 06/20/2011
Notification Time: 17:10 [ET]
Event Date: 06/20/2011
Event Time: 09:45 [EDT]
Last Update Date: 06/20/2011
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):
ERIC DUNCAN (R3DO)
ERIC DUNCAN (R3DO)
| Unit | SCRAM Code | RX Crit | Initial PWR | Initial RX Mode | Current PWR | Current RX Mode |
|---|---|---|---|---|---|---|
| 1 | N | Y | 100 | Power Operation | 100 | Power Operation |
UNPLANNED POWER OUTAGES AFFECT EMERGENCY RESPONSE SYSTEMS
"On June 20, 2011, at approximately 0945 hours, electrical power was lost to non-essential 120 volt busses V-1-F and V-2-F. The busses supply electrical power to the plant integrated computer system (ICS). As a result of the power loss, the Safety Parameter Display System (SPDS), the Emergency Response Data System (ERDS), and the automatic mode calculation of the Computer Aided Dose Assessment Program (CADAP) were unavailable.
"Preliminary investigation indicates the power loss was caused by receipt of a spurious high temperature alarm (greater than 95 degrees F) in the Technical Support Center (TSC) facility where the electrical busses are located. The alarm signal is believed to be invalid because the high temperature alarm came in for two minutes and reset, and the ambient temperature in the TSC remained steady at 65 degrees F. Electrical busses V-1-F and V-2-F were re-energized and restored to service at 1505 hours.
"The ICS, SPDS, ERDS, and CADAP systems were fully restored at 1652 hours on June 20, 2011. In the event of an emergency while these systems were unavailable, contingency plans were in place to transmit plant parameter data and perform the dose assessment function.
"This event is being reported in accordance with 10 CFR 50.72(b)(3)(xiii), as a condition that results in a major loss of emergency assessment and communications capability. The NRC Resident Inspector has been notified."
"On June 20, 2011, at approximately 0945 hours, electrical power was lost to non-essential 120 volt busses V-1-F and V-2-F. The busses supply electrical power to the plant integrated computer system (ICS). As a result of the power loss, the Safety Parameter Display System (SPDS), the Emergency Response Data System (ERDS), and the automatic mode calculation of the Computer Aided Dose Assessment Program (CADAP) were unavailable.
"Preliminary investigation indicates the power loss was caused by receipt of a spurious high temperature alarm (greater than 95 degrees F) in the Technical Support Center (TSC) facility where the electrical busses are located. The alarm signal is believed to be invalid because the high temperature alarm came in for two minutes and reset, and the ambient temperature in the TSC remained steady at 65 degrees F. Electrical busses V-1-F and V-2-F were re-energized and restored to service at 1505 hours.
"The ICS, SPDS, ERDS, and CADAP systems were fully restored at 1652 hours on June 20, 2011. In the event of an emergency while these systems were unavailable, contingency plans were in place to transmit plant parameter data and perform the dose assessment function.
"This event is being reported in accordance with 10 CFR 50.72(b)(3)(xiii), as a condition that results in a major loss of emergency assessment and communications capability. The NRC Resident Inspector has been notified."
Power Reactor
Event Number: 46971
Facility: MILLSTONE
Region: 1 State: CT
Unit: [] [2] []
RX Type: [1] GE-3,[2] CE,[3] W-4-LP
NRC Notified By: FRED PERKINS
HQ OPS Officer: STEVE SANDIN
Region: 1 State: CT
Unit: [] [2] []
RX Type: [1] GE-3,[2] CE,[3] W-4-LP
NRC Notified By: FRED PERKINS
HQ OPS Officer: STEVE SANDIN
Notification Date: 06/20/2011
Notification Time: 13:01 [ET]
Event Date: 06/20/2011
Event Time: 11:52 [EDT]
Last Update Date: 06/20/2011
Notification Time: 13:01 [ET]
Event Date: 06/20/2011
Event Time: 11:52 [EDT]
Last Update Date: 06/20/2011
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):
CHRISTOPHER CAHILL (R1DO)
CHRISTOPHER CAHILL (R1DO)
| Unit | SCRAM Code | RX Crit | Initial PWR | Initial RX Mode | Current PWR | Current RX Mode |
|---|---|---|---|---|---|---|
| 2 | A/R | Y | 60 | Power Operation | 0 | Hot Standby |
UNIT 2 REACTOR TRIP FOLLOWING THE LOSS OF THE "B" MAIN FEEDWATER (MFW) PUMP
At 1152 EDT on 6/20/11 while operating at 60% power, the "B" MFW Pump tripped for reasons unknown. There was no maintenance or I&C work on-going at the time involving this pump. Operators initiated a manual reactor trip, however, [they] are not certain whether the automatic reactor trip setpoint of 49.5% Steam Generator Water Level Narrow Range (SGWL NR) was reached first. SGWL decreased to the Auxiliary Feedwater (AFW) setpoint of 26.8% NR causing the initiation of both motor-driven AFW Pumps. All Control Rods fully inserted.
Unit 2 is currently stable in Mode 3, Hot Standby, removing decay heat via the Main Steam line to the Condenser. Operators secured AFW and will initiate feed to the Steam Generators using the "A" MFW Pump. Unit 2 is in a normal post-trip electrical lineup with all sources of offsite power available. The licensee has the cause of the "B" MFW Pump trip under investigation.
The licensee informed both state/local (Waterford Dispatch) and the NRC Resident Inspector
At 1152 EDT on 6/20/11 while operating at 60% power, the "B" MFW Pump tripped for reasons unknown. There was no maintenance or I&C work on-going at the time involving this pump. Operators initiated a manual reactor trip, however, [they] are not certain whether the automatic reactor trip setpoint of 49.5% Steam Generator Water Level Narrow Range (SGWL NR) was reached first. SGWL decreased to the Auxiliary Feedwater (AFW) setpoint of 26.8% NR causing the initiation of both motor-driven AFW Pumps. All Control Rods fully inserted.
Unit 2 is currently stable in Mode 3, Hot Standby, removing decay heat via the Main Steam line to the Condenser. Operators secured AFW and will initiate feed to the Steam Generators using the "A" MFW Pump. Unit 2 is in a normal post-trip electrical lineup with all sources of offsite power available. The licensee has the cause of the "B" MFW Pump trip under investigation.
The licensee informed both state/local (Waterford Dispatch) and the NRC Resident Inspector