Event Notification Report for April 19, 2003
U.S. Nuclear Regulatory Commission
Operations Center
EVENT REPORTS FOR
04/18/2003 - 04/19/2003
Power Reactor
Event Number: 39772
Facility: CALVERT CLIFFS
Region: 1 State: MD
Unit: [] [2] []
RX Type: [1] CE,[2] CE
NRC Notified By: JAMES GROOMS
HQ OPS Officer: HOWIE CROUCH
Region: 1 State: MD
Unit: [] [2] []
RX Type: [1] CE,[2] CE
NRC Notified By: JAMES GROOMS
HQ OPS Officer: HOWIE CROUCH
Notification Date: 04/19/2003
Notification Time: 19:33 [ET]
Event Date: 04/19/2003
Event Time: 15:30 [EDT]
Last Update Date: 04/19/2003
Notification Time: 19:33 [ET]
Event Date: 04/19/2003
Event Time: 15:30 [EDT]
Last Update Date: 04/19/2003
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):
DAVID SILK (R1)
DAVID SILK (R1)
| Unit | SCRAM Code | RX Crit | Initial PWR | Initial RX Mode | Current PWR | Current RX Mode |
|---|---|---|---|---|---|---|
| 2 | N | N | 0 | Hot Standby | 0 | Hot Standby |
REACTOR PROTECTION SYSTEM ACTUATION
The following information was received from the licensee via facsimile:
"While performing STP [Surveillance Test Procedure] M-200-2, the RPS [Reactor Protection System] S/G [Steam Generator] Low Press[ure] Trip Units were bypassed using the S/G Low Press[ure] Bypass key. The bypass was automatically removed early before the S/G Low Press[ure] trips were clear[ed]. This resulted in the RPS calling for the TCBs [Trip Circuit Breakers] to open. The CEAs [control rods] were already in the core. This is an 8 hr. report[able] per [10 CFR] 50.72(b)(3)(iv)(A). S/G Press[ures] were [approximately] 700 psi."
The surveillance test procedure was to allow time response testing on the trip circuit breakers by the Electrical department (which requires the TCBs to be closed). The plant was also performing a plant heatup in preparation for plant startup. During normal operations, the S/G Low Pressure Trip setpoint is 700 psi. The surveillance test procedure allowed the S/G Low Pressure Trip to be bypassed via the bypass key. If the S/G Low Pressure Trip is keyed to bypass and the S/G pressure increases to greater than 785 psi, then the 700 psi trip setpoint is automatically re-instated regardless of bypass key position. The licensee is investigating the reason that the S/G Low Pressure Trip was automatically re-instated with S/G pressure at approximately 700 psi.
The NRC Resident Inspector has been notified.
The following information was received from the licensee via facsimile:
"While performing STP [Surveillance Test Procedure] M-200-2, the RPS [Reactor Protection System] S/G [Steam Generator] Low Press[ure] Trip Units were bypassed using the S/G Low Press[ure] Bypass key. The bypass was automatically removed early before the S/G Low Press[ure] trips were clear[ed]. This resulted in the RPS calling for the TCBs [Trip Circuit Breakers] to open. The CEAs [control rods] were already in the core. This is an 8 hr. report[able] per [10 CFR] 50.72(b)(3)(iv)(A). S/G Press[ures] were [approximately] 700 psi."
The surveillance test procedure was to allow time response testing on the trip circuit breakers by the Electrical department (which requires the TCBs to be closed). The plant was also performing a plant heatup in preparation for plant startup. During normal operations, the S/G Low Pressure Trip setpoint is 700 psi. The surveillance test procedure allowed the S/G Low Pressure Trip to be bypassed via the bypass key. If the S/G Low Pressure Trip is keyed to bypass and the S/G pressure increases to greater than 785 psi, then the 700 psi trip setpoint is automatically re-instated regardless of bypass key position. The licensee is investigating the reason that the S/G Low Pressure Trip was automatically re-instated with S/G pressure at approximately 700 psi.
The NRC Resident Inspector has been notified.
Power Reactor
Event Number: 39773
Facility: NORTH ANNA
Region: 2 State: VA
Unit: [1] [] []
RX Type: [1] W-3-LP,[2] W-3-LP
NRC Notified By: BRAD BROWN
HQ OPS Officer: HOWIE CROUCH
Region: 2 State: VA
Unit: [1] [] []
RX Type: [1] W-3-LP,[2] W-3-LP
NRC Notified By: BRAD BROWN
HQ OPS Officer: HOWIE CROUCH
Notification Date: 04/19/2003
Notification Time: 23:23 [ET]
Event Date: 04/19/2003
Event Time: 20:08 [EDT]
Last Update Date: 04/19/2003
Notification Time: 23:23 [ET]
Event Date: 04/19/2003
Event Time: 20:08 [EDT]
Last Update Date: 04/19/2003
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):
THOMAS DECKER (R2)
THOMAS DECKER (R2)
| Unit | SCRAM Code | RX Crit | Initial PWR | Initial RX Mode | Current PWR | Current RX Mode |
|---|---|---|---|---|---|---|
| 1 | M/R | Y | 74 | Power Operation | 0 | Hot Standby |
MANUAL REACTOR TRIP WITH A SUBSEQUENT AUXILIARY FEEDWATER ACTUATION
The following information was received from the licensee via facsimile:
"A 4 hr. report due to a manual reactor trip (RPS actuation) due to an oil leak on the main turbine.
"An 8 hr. report due to Aux[iliary] Feedwater System initiation due to the reactor trip (expected response)."
The manual reactor trip was due to a lube oil leak through the diaphragm on the EHC (Electro-Hydraulic Control) interface valve. The leak caused the interface valve to close therefore isolating EHC hydraulic fluid to the turbine control system. During the trip, all rods inserted into the core. As a result of the transient, a pressurizer power-operated relief valve lifted and indications show that it subsequently re-seated. There are no indications that the blow-out plug on the pressurizer relief tank was damaged. Pressurizer relief tank pressure has been reduced to normal. Additionally, a secondary tube side relief lifted and was re-seated after the second running feedwater pump was secured. Decay heat is being removed via the steam dump to condenser system operating in automatic. The electrical grid is stable and supplying plant loads. There was no affect on the other operating unit.
As expected, the Auxiliary Feedwater System initiated. The system has been returned to normal, standby configuration and steam generator level is being controlled by the main feedwater system operating in manual control.
The licensee has notified the NRC Resident Inspector.
The following information was received from the licensee via facsimile:
"A 4 hr. report due to a manual reactor trip (RPS actuation) due to an oil leak on the main turbine.
"An 8 hr. report due to Aux[iliary] Feedwater System initiation due to the reactor trip (expected response)."
The manual reactor trip was due to a lube oil leak through the diaphragm on the EHC (Electro-Hydraulic Control) interface valve. The leak caused the interface valve to close therefore isolating EHC hydraulic fluid to the turbine control system. During the trip, all rods inserted into the core. As a result of the transient, a pressurizer power-operated relief valve lifted and indications show that it subsequently re-seated. There are no indications that the blow-out plug on the pressurizer relief tank was damaged. Pressurizer relief tank pressure has been reduced to normal. Additionally, a secondary tube side relief lifted and was re-seated after the second running feedwater pump was secured. Decay heat is being removed via the steam dump to condenser system operating in automatic. The electrical grid is stable and supplying plant loads. There was no affect on the other operating unit.
As expected, the Auxiliary Feedwater System initiated. The system has been returned to normal, standby configuration and steam generator level is being controlled by the main feedwater system operating in manual control.
The licensee has notified the NRC Resident Inspector.