Event Notification Report for March 13, 2003
U.S. Nuclear Regulatory Commission
Operations Center
EVENT REPORTS FOR
03/12/2003 - 03/13/2003
Power Reactor
Event Number: 39667
Facility: CALLAWAY
Region: 4 State: MO
Unit: [1] [] []
RX Type: [1] W-4-LP
NRC Notified By: BREDEMAN
HQ OPS Officer: CHAUNCEY GOULD
Region: 4 State: MO
Unit: [1] [] []
RX Type: [1] W-4-LP
NRC Notified By: BREDEMAN
HQ OPS Officer: CHAUNCEY GOULD
Notification Date: 03/13/2003
Notification Time: 16:32 [ET]
Event Date: 03/13/2003
Event Time: 08:00 [CST]
Last Update Date: 03/13/2003
Notification Time: 16:32 [ET]
Event Date: 03/13/2003
Event Time: 08:00 [CST]
Last Update Date: 03/13/2003
Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(3)(ii)(B) - UNANALYZED CONDITION
10 CFR Section:
50.72(b)(3)(ii)(B) - UNANALYZED CONDITION
Person (Organization):
MARK SHAFFER (R4)
MARK SHAFFER (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 |
ERROR IN THE CURRENT SAFETY ANALYSIS FOR A STEAM GENERATOR TUBE RUPTURE
"On March 13, 2003 while operating at 100% power, Callaway Plant determined a problem existed in the current safety analysis for a steam generator tube rupture accident accompanied by an overfill condition of the ruptured steam generator. This problem was discovered while reviewing a future plant modification package. The current FSAR analysis concluded that an overfill condition would not occur during a tube rupture accident.
"Through investigations instigated during the modification review process, it was determined that for current plant conditions, an overfill condition could result if an auxiliary feedwater control valve supplying the ruptured steam generator were to fail open. In this case, it was concluded that the ruptured steam generator would overfill and water would be released through the steam generator safety valves, resulting in a radioactive release to the environment greater than allowed by regulatory guidance. Since this accident was not adequately incorporated in the FSAR, specific procedural time limits for operator actions had not been developed. This lack of credited operator time limits coupled with past plant modifications result in a potential to exist for the ruptured steam generator overfill scenario.
"Current Technical Specifications allow a reactor coolant system Dose Equivalent Iodine (DEI) value of 1.0 microcurie per gram. To assure compliance with existing regulatory guidance, plant procedures have been changed to administratively reduce the steady state DEI limit to 0.3 microcurie per gram, a value that has not been exceeded in the last three years. Current steady state DEI concentration in the reactor coolant system is 0.001769 microcurie per gram. The new lower DEI limit will ensure that if an overfill condition were to occur during a steam generator tube rupture, post accident radiological consequences would not exceed the limits imposed by the FSAR and the Standard Review Plan."
The NRC Resident Inspector was notified.
"On March 13, 2003 while operating at 100% power, Callaway Plant determined a problem existed in the current safety analysis for a steam generator tube rupture accident accompanied by an overfill condition of the ruptured steam generator. This problem was discovered while reviewing a future plant modification package. The current FSAR analysis concluded that an overfill condition would not occur during a tube rupture accident.
"Through investigations instigated during the modification review process, it was determined that for current plant conditions, an overfill condition could result if an auxiliary feedwater control valve supplying the ruptured steam generator were to fail open. In this case, it was concluded that the ruptured steam generator would overfill and water would be released through the steam generator safety valves, resulting in a radioactive release to the environment greater than allowed by regulatory guidance. Since this accident was not adequately incorporated in the FSAR, specific procedural time limits for operator actions had not been developed. This lack of credited operator time limits coupled with past plant modifications result in a potential to exist for the ruptured steam generator overfill scenario.
"Current Technical Specifications allow a reactor coolant system Dose Equivalent Iodine (DEI) value of 1.0 microcurie per gram. To assure compliance with existing regulatory guidance, plant procedures have been changed to administratively reduce the steady state DEI limit to 0.3 microcurie per gram, a value that has not been exceeded in the last three years. Current steady state DEI concentration in the reactor coolant system is 0.001769 microcurie per gram. The new lower DEI limit will ensure that if an overfill condition were to occur during a steam generator tube rupture, post accident radiological consequences would not exceed the limits imposed by the FSAR and the Standard Review Plan."
The NRC Resident Inspector was notified.
Power Reactor
Event Number: 39669
Facility: MONTICELLO
Region: 3 State: MN
Unit: [1] [] []
RX Type: [1] GE-3
NRC Notified By: JACK EARCLEY
HQ OPS Officer: RICH LAURA
Region: 3 State: MN
Unit: [1] [] []
RX Type: [1] GE-3
NRC Notified By: JACK EARCLEY
HQ OPS Officer: RICH LAURA
Notification Date: 03/13/2003
Notification Time: 18:35 [ET]
Event Date: 03/13/2003
Event Time: 14:30 [CST]
Last Update Date: 03/13/2003
Notification Time: 18:35 [ET]
Event Date: 03/13/2003
Event Time: 14:30 [CST]
Last Update Date: 03/13/2003
Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(3)(ii)(B) - UNANALYZED CONDITION
10 CFR Section:
50.72(b)(3)(ii)(B) - UNANALYZED CONDITION
Person (Organization):
KENNETH O'BRIEN (R3)
KENNETH O'BRIEN (R3)
| Unit | SCRAM Code | RX Crit | Initial PWR | Initial RX Mode | Current PWR | Current RX Mode |
|---|---|---|---|---|---|---|
| 1 | N | Y | 100 | Power Operation | 100 | Power Operation |
DEGRADED APPENDIX "R" BARRIER AT MONTICELLO
"During inspection of penetration FZ-4900 in the upper 4KV room it was discovered there was a portion of the penetration seal that was degraded which caused the upper and lower 4KV rooms to communicate with each other. Penetration seal FZ-4900 is a gypsum wall board assembly approximately 18 inches wide 4 foot high and 24 foot long. The degraded part of the penetration is approximately 1 inch by 4 inch hole. The upper and lower 4 KV areas are required to be separated in accordance with 10CFR50 Appendix R requirements."
The licensee indicated they established a fire watch as a compensatory measure.
The NRC Resident Inspector was notified.
"During inspection of penetration FZ-4900 in the upper 4KV room it was discovered there was a portion of the penetration seal that was degraded which caused the upper and lower 4KV rooms to communicate with each other. Penetration seal FZ-4900 is a gypsum wall board assembly approximately 18 inches wide 4 foot high and 24 foot long. The degraded part of the penetration is approximately 1 inch by 4 inch hole. The upper and lower 4 KV areas are required to be separated in accordance with 10CFR50 Appendix R requirements."
The licensee indicated they established a fire watch as a compensatory measure.
The NRC Resident Inspector was notified.
Power Reactor
Event Number: 39665
Facility: OYSTER CREEK
Region: 1 State: NJ
Unit: [1] [] []
RX Type: [1] GE-2
NRC Notified By: CIGANIK
HQ OPS Officer: CHAUNCEY GOULD
Region: 1 State: NJ
Unit: [1] [] []
RX Type: [1] GE-2
NRC Notified By: CIGANIK
HQ OPS Officer: CHAUNCEY GOULD
Notification Date: 03/13/2003
Notification Time: 13:38 [ET]
Event Date: 03/13/2003
Event Time: 06:15 [EST]
Last Update Date: 03/13/2003
Notification Time: 13:38 [ET]
Event Date: 03/13/2003
Event Time: 06:15 [EST]
Last Update Date: 03/13/2003
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):
MOHAMED SHANBAKY (R1)
MOHAMED SHANBAKY (R1)
| Unit | SCRAM Code | RX Crit | Initial PWR | Initial RX Mode | Current PWR | Current RX Mode |
|---|---|---|---|---|---|---|
| 1 | N | Y | 100 | Power Operation | 100 | Power Operation |
MAJOR LOSS OF EMERGENCY PREPAREDNESS CAPABILITIES
The licensee was notified by Ocean County that their alarm panel indicated the radio power to the prompt notification sirens had failed with unknown affect on the sirens operation. The Licensee Technicians were dispatched and had to reboot a computer to clear the alarm. Sirens would not operate until the computer was rebooted. Sirens tested satisfactory @ 0930 hours.
This notification was made because it can not be determined if the Ocean County siren activation technicians possess the knowledge and skills to reboot this computer if siren activation would have been required. If it is later determined that the technicians did possess the required knowledge and skills, this notification may be retracted.
The NRC Resident Inspector was notified along with local agencies.
The licensee was notified by Ocean County that their alarm panel indicated the radio power to the prompt notification sirens had failed with unknown affect on the sirens operation. The Licensee Technicians were dispatched and had to reboot a computer to clear the alarm. Sirens would not operate until the computer was rebooted. Sirens tested satisfactory @ 0930 hours.
This notification was made because it can not be determined if the Ocean County siren activation technicians possess the knowledge and skills to reboot this computer if siren activation would have been required. If it is later determined that the technicians did possess the required knowledge and skills, this notification may be retracted.
The NRC Resident Inspector was notified along with local agencies.