Event Notification Report for October 27, 2006
U.S. Nuclear Regulatory Commission
Operations Center
EVENT REPORTS FOR
10/26/2006 - 10/27/2006
!!!!! THIS EVENT HAS BEEN RETRACTED !!!!!
!!!!! THIS EVENT HAS BEEN RETRACTED !!!!!
Power Reactor
Event Number: 42939
Facility: COOPER
Region: 4 State: NE
Unit: [1] [] []
RX Type: [1] GE-4
NRC Notified By: STEVE WHEELER
HQ OPS Officer: BILL HUFFMAN
Region: 4 State: NE
Unit: [1] [] []
RX Type: [1] GE-4
NRC Notified By: STEVE WHEELER
HQ OPS Officer: BILL HUFFMAN
Notification Date: 10/27/2006
Notification Time: 10:02 [ET]
Event Date: 10/27/2006
Event Time: 03:37 [CDT]
Last Update Date: 11/17/2006
Notification Time: 10:02 [ET]
Event Date: 10/27/2006
Event Time: 03:37 [CDT]
Last Update Date: 11/17/2006
Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(3)(ii)(A) - DEGRADED CONDITION
10 CFR Section:
50.72(b)(3)(ii)(A) - DEGRADED CONDITION
Person (Organization):
JEFFREY CLARK (R4)
JEFFREY CLARK (R4)
| Unit | SCRAM Code | RX Crit | Initial PWR | Initial RX Mode | Current PWR | Current RX Mode |
|---|---|---|---|---|---|---|
| 1 | N | N | 0 | Refueling | 0 | Refueling |
CONTAINMENT PATHWAY DECLARED INOPERABLE FOLLOWING LOCAL LEAK RATE TESTING
"This condition is being reported in accordance with the requirements of 10CFR50.72(b)(3)(ii) as an event or condition that results in: (A) The condition of the nuclear power plant, including its principal safety barriers, being seriously degraded.
"During local leak rate testing (LLRT) of the 'B' reactor feed line to the reactor, both primary containment boundary check valves (RF-CV-13CV and RF-CV-14CV) failed LLRT. Efforts to quantify the leak rate were unsuccessful. The maximum leakage indicated on the test equipment used is 424 standard cubic feet per hour (SCFH). The allowable leakage (La) for Primary Containment is 317.41 SCFH. The affected primary containment pathway was declared inoperable. CNS is in MODE 5 for refueling with vessel level greater than 21 feet above the flange and the spent fuel pool gates removed. No fuel movements or operations with the potential to drain the reactor vessel were in progress. Primary containment is not required per CNS Technical Specifications in this mode of operation.
"The NRC Senior Resident Inspector has been notified."
* * * RETRACTION FROM FLEMING TO HUFFMAN AT 1500 EST ON 11/17/06 * * *
"This notification is being made to retract Event Notification #42939 which reported a Primary Containment isolation pathway being declared inoperable following Local Leak Rate Testing (LLRT); specifically, the 'B' reactor feed line containment boundary inboard and outboard check valves. CNS has determined that the LLRT test method established test conditions that were not representative of the conditions under which the check valves would be required to perform their safety function. Inspection, in-situ testing and evaluation demonstrated the reactor feed line check valves would have performed their isolation functions during credible accident conditions. Thus, the condition was not a serious degradation of a principal safety barrier.
"During the normal plant shutdown to refueling outage RE23 with gradually decreasing feed flow, the check valve discs had not completely settled into the valve seats. Due to the slow pressurization rate of the LLRT working medium (air), sufficient differential pressure across the valve could not be developed which resulted in failing the test. Inspection of the check valves in their as-found condition revealed that a slight contact between the disc and the top of seat left a crescent shaped gap which, at the maximum (the bottom), was no more than 1/2 inch. An in-situ test measured the force needed to settle the disc into the seat and found it equated to a flow related pressure drop of less than 1 psid. A review of accident sequences concluded that post-accident pressurization of the feed lines would have closed the check valves for all credible accident scenarios. Check valve disc to pivot arm spacing adjustments were made to ensure sufficient disc float to reseat and fully close the valve from any open position. Subsequently, LLRT results were satisfactory."
The licensee notified the NRC Resident Inspector. R4DO (Johnson) notified.
"This condition is being reported in accordance with the requirements of 10CFR50.72(b)(3)(ii) as an event or condition that results in: (A) The condition of the nuclear power plant, including its principal safety barriers, being seriously degraded.
"During local leak rate testing (LLRT) of the 'B' reactor feed line to the reactor, both primary containment boundary check valves (RF-CV-13CV and RF-CV-14CV) failed LLRT. Efforts to quantify the leak rate were unsuccessful. The maximum leakage indicated on the test equipment used is 424 standard cubic feet per hour (SCFH). The allowable leakage (La) for Primary Containment is 317.41 SCFH. The affected primary containment pathway was declared inoperable. CNS is in MODE 5 for refueling with vessel level greater than 21 feet above the flange and the spent fuel pool gates removed. No fuel movements or operations with the potential to drain the reactor vessel were in progress. Primary containment is not required per CNS Technical Specifications in this mode of operation.
"The NRC Senior Resident Inspector has been notified."
* * * RETRACTION FROM FLEMING TO HUFFMAN AT 1500 EST ON 11/17/06 * * *
"This notification is being made to retract Event Notification #42939 which reported a Primary Containment isolation pathway being declared inoperable following Local Leak Rate Testing (LLRT); specifically, the 'B' reactor feed line containment boundary inboard and outboard check valves. CNS has determined that the LLRT test method established test conditions that were not representative of the conditions under which the check valves would be required to perform their safety function. Inspection, in-situ testing and evaluation demonstrated the reactor feed line check valves would have performed their isolation functions during credible accident conditions. Thus, the condition was not a serious degradation of a principal safety barrier.
"During the normal plant shutdown to refueling outage RE23 with gradually decreasing feed flow, the check valve discs had not completely settled into the valve seats. Due to the slow pressurization rate of the LLRT working medium (air), sufficient differential pressure across the valve could not be developed which resulted in failing the test. Inspection of the check valves in their as-found condition revealed that a slight contact between the disc and the top of seat left a crescent shaped gap which, at the maximum (the bottom), was no more than 1/2 inch. An in-situ test measured the force needed to settle the disc into the seat and found it equated to a flow related pressure drop of less than 1 psid. A review of accident sequences concluded that post-accident pressurization of the feed lines would have closed the check valves for all credible accident scenarios. Check valve disc to pivot arm spacing adjustments were made to ensure sufficient disc float to reseat and fully close the valve from any open position. Subsequently, LLRT results were satisfactory."
The licensee notified the NRC Resident Inspector. R4DO (Johnson) notified.
Power Reactor
Event Number: 42940
Facility: GRAND GULF
Region: 4 State: MS
Unit: [1] [] []
RX Type: [1] GE-6
NRC Notified By: KARL EHRHARDT
HQ OPS Officer: BILL GOTT
Region: 4 State: MS
Unit: [1] [] []
RX Type: [1] GE-6
NRC Notified By: KARL EHRHARDT
HQ OPS Officer: BILL GOTT
Notification Date: 10/27/2006
Notification Time: 12:03 [ET]
Event Date: 10/27/2006
Event Time: 09:15 [CDT]
Last Update Date: 10/27/2006
Notification Time: 12:03 [ET]
Event Date: 10/27/2006
Event Time: 09:15 [CDT]
Last Update Date: 10/27/2006
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):
JEFFREY CLARK (R4)
JEFFREY CLARK (R4)
| Unit | SCRAM Code | RX Crit | Initial PWR | Initial RX Mode | Current PWR | Current RX Mode |
|---|---|---|---|---|---|---|
| 1 | N | Y | 100 | Power Operation | 100 | Power Operation |
EMERGENCY OPERATIONS FACILITY NON OPERATIONAL DUE TO PLANNED MAINTENANCE
"The Emergency Operations Facility (EOF) became non operational on October 27, 2006, at 0915 due to planned maintenance on the electrical power distribution for that facility. This call is required due to not being able to promptly restore the facility in the event of an accident as noted in NUREG 1022 Section 3.2.13. The maintenance should be completed by October 28, 2006.
"The backup EOF will be used during this maintenance."
The licensee notified the NRC Resident Inspector.
"The Emergency Operations Facility (EOF) became non operational on October 27, 2006, at 0915 due to planned maintenance on the electrical power distribution for that facility. This call is required due to not being able to promptly restore the facility in the event of an accident as noted in NUREG 1022 Section 3.2.13. The maintenance should be completed by October 28, 2006.
"The backup EOF will be used during this maintenance."
The licensee notified the NRC Resident Inspector.
Power Reactor
Event Number: 42937
Facility: COMANCHE PEAK
Region: 4 State: TX
Unit: [] [2] []
RX Type: [1] W-4-LP,[2] W-4-LP
NRC Notified By: GARY CASTERSON
HQ OPS Officer: JASON KOZAL
Region: 4 State: TX
Unit: [] [2] []
RX Type: [1] W-4-LP,[2] W-4-LP
NRC Notified By: GARY CASTERSON
HQ OPS Officer: JASON KOZAL
Notification Date: 10/27/2006
Notification Time: 05:06 [ET]
Event Date: 10/27/2006
Event Time: 03:08 [CDT]
Last Update Date: 10/27/2006
Notification Time: 05:06 [ET]
Event Date: 10/27/2006
Event Time: 03:08 [CDT]
Last Update Date: 10/27/2006
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):
JEFFREY CLARK (R4)
JEFFREY CLARK (R4)
| Unit | SCRAM Code | RX Crit | Initial PWR | Initial RX Mode | Current PWR | Current RX Mode |
|---|---|---|---|---|---|---|
| 2 | M/R | Y | 28 | Power Operation | 0 | Hot Standby |
MANUAL REACTOR SCRAM DUE TO LOSS OF BOTH MAIN FEED PUMPS
The licensee was performing load reject testing on the Unit 2 turbine generator when oscillations were observed on the plant steam dump system. These oscillations caused a high steam generator water level that caused both Main Feed Pumps (MFP) to trip. Upon receiving this indication the licensee manually tripped the reactor. All Auxiliary Feed Water (AFW) pumps started as expected.
All rods inserted as expected. Decay heat is being removed using AFW supplying both steam generators and steaming to the steam dumps. The plant is in a normal shutdown electrical lineup.
The licensee notified the NRC Resident Inspector.
The licensee was performing load reject testing on the Unit 2 turbine generator when oscillations were observed on the plant steam dump system. These oscillations caused a high steam generator water level that caused both Main Feed Pumps (MFP) to trip. Upon receiving this indication the licensee manually tripped the reactor. All Auxiliary Feed Water (AFW) pumps started as expected.
All rods inserted as expected. Decay heat is being removed using AFW supplying both steam generators and steaming to the steam dumps. The plant is in a normal shutdown electrical lineup.
The licensee notified the NRC Resident Inspector.