Event Notification Report for March 15, 2008
U.S. Nuclear Regulatory Commission
Operations Center
EVENT REPORTS FOR
03/14/2008 - 03/15/2008
Power Reactor
Event Number: 44066
Facility: FT CALHOUN
Region: 4 State: NE
Unit: [1] [] []
RX Type: [1] CE
NRC Notified By: KEVIN R. BOSTON
HQ OPS Officer: PETE SNYDER
Region: 4 State: NE
Unit: [1] [] []
RX Type: [1] CE
NRC Notified By: KEVIN R. BOSTON
HQ OPS Officer: PETE SNYDER
Notification Date: 03/15/2008
Notification Time: 11:19 [ET]
Event Date: 03/15/2008
Event Time: 09:33 [EDT]
Last Update Date: 03/17/2008
Notification Time: 11:19 [ET]
Event Date: 03/15/2008
Event Time: 09:33 [EDT]
Last Update Date: 03/17/2008
Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(2)(iv)(B) - RPS ACTUATION - CRITICAL
10 CFR Section:
50.72(b)(2)(iv)(B) - RPS ACTUATION - CRITICAL
Person (Organization):
TROY PRUETT (R4)
TROY PRUETT (R4)
| Unit | SCRAM Code | RX Crit | Initial PWR | Initial RX Mode | Current PWR | Current RX Mode |
|---|---|---|---|---|---|---|
| 1 | A/R | Y | 85 | Power Operation | 0 | Hot Standby |
REACTOR TRIP DUE TO TURBINE TRIP
"At 09:33 EDT the plant experienced a turbine trip which resulted in a reactor trip from 85%. Plant power was less than 100% due to a [turbine] control valve oscillating problem identified on 3/13/08.
"Post-trip [complications]: one of the four 4160 V busses, Bus 1A1, did not fast transfer from 22 KV to the 161 KV system as expected. This buss loss caused Reactor Coolant Pump RC-3A [to] trip off. Seals on RC-3A appear to be failed. The other 3 RC pumps remained in service for forced circulation through the Steam Generators
"Steam Generator levels are being maintained by Main Feedwater Pump FW-4C. Steaming is through the steam dump and bypass valves to the condensers.
"No primary or secondary relief valves lifted during this event.
"Standard post-trip actions were taken per EOP-00 , followed by a transition to EOP-2. Loss of Off-Site Power. During EOP-00 the plant computer indicated 8 rods not inserted into the reactor core. Secondary and primary rod indication showed all control rods inserted. In response to the difference of rod position indication, emergency boration was implemented. Emergency boration was secured after shutdown margin was verified.
"4160 V buss 1A was reenergized on the 161kv system at 10:31 EDT.
"A transition to plant procedure OP-3A, plant shutdown was made at 10:58 EDT.
"Reactor Coolant Pump RC-3A seal evaluation is in progress. The plant will remain in Mode 3."
The licensee notified the NRC Resident Inspector and plans a press release.
While RCP RC-3A seals indicate the same pressure across them there were no accompanying indications of containment sump water level increasing or high radiation levels in containment.
Decay heat is being removed by normal feedwater feeding the steam generators steaming to the condenser steam dumps. Nuclear Instruments indicated plant shutdown throughout the trip. Safety buses are powered by offsite power and emergency diesel generators are in standby.
* * * UPDATE FROM ERICK MATZKE TO PETE SNYDER AT 1033 ON 3/17/08 * * *
"None of the reactor coolant pump seals at [Fort Calhoun Station] (FCS) were failed. There were abnormal indications on the A reactor coolant pump seal. The A reactor coolant pump seal is now working normally."
Notified R4DO (Whitten).
"At 09:33 EDT the plant experienced a turbine trip which resulted in a reactor trip from 85%. Plant power was less than 100% due to a [turbine] control valve oscillating problem identified on 3/13/08.
"Post-trip [complications]: one of the four 4160 V busses, Bus 1A1, did not fast transfer from 22 KV to the 161 KV system as expected. This buss loss caused Reactor Coolant Pump RC-3A [to] trip off. Seals on RC-3A appear to be failed. The other 3 RC pumps remained in service for forced circulation through the Steam Generators
"Steam Generator levels are being maintained by Main Feedwater Pump FW-4C. Steaming is through the steam dump and bypass valves to the condensers.
"No primary or secondary relief valves lifted during this event.
"Standard post-trip actions were taken per EOP-00 , followed by a transition to EOP-2. Loss of Off-Site Power. During EOP-00 the plant computer indicated 8 rods not inserted into the reactor core. Secondary and primary rod indication showed all control rods inserted. In response to the difference of rod position indication, emergency boration was implemented. Emergency boration was secured after shutdown margin was verified.
"4160 V buss 1A was reenergized on the 161kv system at 10:31 EDT.
"A transition to plant procedure OP-3A, plant shutdown was made at 10:58 EDT.
"Reactor Coolant Pump RC-3A seal evaluation is in progress. The plant will remain in Mode 3."
The licensee notified the NRC Resident Inspector and plans a press release.
While RCP RC-3A seals indicate the same pressure across them there were no accompanying indications of containment sump water level increasing or high radiation levels in containment.
Decay heat is being removed by normal feedwater feeding the steam generators steaming to the condenser steam dumps. Nuclear Instruments indicated plant shutdown throughout the trip. Safety buses are powered by offsite power and emergency diesel generators are in standby.
* * * UPDATE FROM ERICK MATZKE TO PETE SNYDER AT 1033 ON 3/17/08 * * *
"None of the reactor coolant pump seals at [Fort Calhoun Station] (FCS) were failed. There were abnormal indications on the A reactor coolant pump seal. The A reactor coolant pump seal is now working normally."
Notified R4DO (Whitten).
Power Reactor
Event Number: 44213
Facility: BRUNSWICK
Region: 2 State: NC
Unit: [1] [] []
RX Type: [1] GE-4,[2] GE-4
NRC Notified By: LEE GRZECK
HQ OPS Officer: JEFF ROTTON
Region: 2 State: NC
Unit: [1] [] []
RX Type: [1] GE-4,[2] GE-4
NRC Notified By: LEE GRZECK
HQ OPS Officer: JEFF ROTTON
Notification Date: 05/14/2008
Notification Time: 13:16 [ET]
Event Date: 03/15/2008
Event Time: 16:35 [EDT]
Last Update Date: 05/14/2008
Notification Time: 13:16 [ET]
Event Date: 03/15/2008
Event Time: 16:35 [EDT]
Last Update Date: 05/14/2008
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):
MARVIN SYKES (R2)
MARVIN SYKES (R2)
| Unit | SCRAM Code | RX Crit | Initial PWR | Initial RX Mode | Current PWR | Current RX Mode |
|---|---|---|---|---|---|---|
| 1 | N | N | 0 | Cold Shutdown | 0 | Cold Shutdown |
INVALID PRIMARY CONTAINMENT ISOLATION SYSTEM (PCIS) ACTUATION
"This 60-day telephone notification is being made under reporting requirements specified by 10 CFR 50.73(a)(2)(iv)(A) and 10 CFR 50.73(a)(1).
"On March 15, 2008, while resetting the main turbine in preparation for performing 1MST-RPS25R, 'RPS Turbine Stop Valve and Control Valve Closure Chan Cal,' a Group 1 isolation was received. Main Steam Lines (MSLs) A and B were isolated for local leak rate testing, however, the MSLs C and D were not. The inboard and outboard Main Steam Isolation Valves (MSIVs) for the C and D MSL closed as a result of this event.
"The isolation occurred during work associated with replacement of the main generator output breakers 22A and 22B. In order for the breakers to be removed, wiring that provides breaker position information was disconnected from the turbine control system (EHC). This resulted in a 'breaker closed' input to EHC. When the main turbine was reset, several logic interactions occurred resulting in the Group 1 isolation. No actual plant condition existed which required a Group 1 isolation and, with the Unit in Mode 4, Group 1 isolation is not required to be operable. Therefore, this actuation is considered to be invalid.
"The affected MSIVs and PCIS instrumentation operated properly. This was a full Group 1 isolation with only the MSIVs associated with the C and D MSLs closing since the A ad B lines were isolated when the actuation occurred. The invalid PCIS actuation was entered into the corrective action program (i.e., NCR 270475).
"The safety significance of this event was minimal. Plant equipment performed as expected. Group 1 isolation was not required to be operable in Mode 4.
"The Group 1 isolation was reset and a jumper was installed to prevent a subsequent invalid Group 1 isolation. Procedures will be developed or revised to disable the Group 1 Isolation during modes in which it is not required and to reset the main turbine during outages.
"The NRC Resident Inspector has been notified at this report."
"This 60-day telephone notification is being made under reporting requirements specified by 10 CFR 50.73(a)(2)(iv)(A) and 10 CFR 50.73(a)(1).
"On March 15, 2008, while resetting the main turbine in preparation for performing 1MST-RPS25R, 'RPS Turbine Stop Valve and Control Valve Closure Chan Cal,' a Group 1 isolation was received. Main Steam Lines (MSLs) A and B were isolated for local leak rate testing, however, the MSLs C and D were not. The inboard and outboard Main Steam Isolation Valves (MSIVs) for the C and D MSL closed as a result of this event.
"The isolation occurred during work associated with replacement of the main generator output breakers 22A and 22B. In order for the breakers to be removed, wiring that provides breaker position information was disconnected from the turbine control system (EHC). This resulted in a 'breaker closed' input to EHC. When the main turbine was reset, several logic interactions occurred resulting in the Group 1 isolation. No actual plant condition existed which required a Group 1 isolation and, with the Unit in Mode 4, Group 1 isolation is not required to be operable. Therefore, this actuation is considered to be invalid.
"The affected MSIVs and PCIS instrumentation operated properly. This was a full Group 1 isolation with only the MSIVs associated with the C and D MSLs closing since the A ad B lines were isolated when the actuation occurred. The invalid PCIS actuation was entered into the corrective action program (i.e., NCR 270475).
"The safety significance of this event was minimal. Plant equipment performed as expected. Group 1 isolation was not required to be operable in Mode 4.
"The Group 1 isolation was reset and a jumper was installed to prevent a subsequent invalid Group 1 isolation. Procedures will be developed or revised to disable the Group 1 Isolation during modes in which it is not required and to reset the main turbine during outages.
"The NRC Resident Inspector has been notified at this report."