Event Notification Report for May 15, 1999
U.S. Nuclear Regulatory Commission
Operations Center
EVENT REPORTS FOR
05/14/1999 - 05/15/1999
Power Reactor
Event Number: 35719
Facility: DRESDEN
Region: 3 State: IL
Unit: [] [2] [3]
RX Type: [1] GE-1,[2] GE-3,[3] GE-3
NRC Notified By: PETE KARABA
HQ OPS Officer: LEIGH TROCINE
Region: 3 State: IL
Unit: [] [2] [3]
RX Type: [1] GE-1,[2] GE-3,[3] GE-3
NRC Notified By: PETE KARABA
HQ OPS Officer: LEIGH TROCINE
Notification Date: 05/15/1999
Notification Time: 09:47 [ET]
Event Date: 05/15/1999
Event Time: 04:56 [CDT]
Last Update Date: 05/15/1999
Notification Time: 09:47 [ET]
Event Date: 05/15/1999
Event Time: 04:56 [CDT]
Last Update Date: 05/15/1999
Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(2)(iii)(D) - ACCIDENT MITIGATION
10 CFR Section:
50.72(b)(2)(iii)(D) - ACCIDENT MITIGATION
Person (Organization):
ROGER LANKSBURY (R3)
ROGER LANKSBURY (R3)
| Unit | SCRAM Code | RX Crit | Initial PWR | Initial RX Mode | Current PWR | Current RX Mode |
|---|---|---|---|---|---|---|
| 2 | N | Y | 100 | Power Operation | 100 | Power Operation |
| 3 | N | Y | 100 | Power Operation | 100 | Power Operation |
FAILURE OF A CONTROL ROOM HVAC REFRIGERATION AND CONDENSING UNIT TO RESTART DURING SURVEILLANCE TESTING
The following text is a portion of a facsimile received from the licensee:
"At 0456 hours on May 15, 1999, the 'B' Control Room HVAC Refrigeration and Condensing Unit (RCU) would not restart during surveillance testing. The 'B' RCU is a single train system and, therefore, is believed to be reportable per ComEd Reportability Manual SAF 1.17. The 'B' RCU is required to operate during design basis accidents to remove the heat from the Main Control Room. The Air Filtration Unit remains operable. This places both units in a 30-day Tech. Spec. LCO."
The licensee plans to notify the NRC resident inspector.
The following text is a portion of a facsimile received from the licensee:
"At 0456 hours on May 15, 1999, the 'B' Control Room HVAC Refrigeration and Condensing Unit (RCU) would not restart during surveillance testing. The 'B' RCU is a single train system and, therefore, is believed to be reportable per ComEd Reportability Manual SAF 1.17. The 'B' RCU is required to operate during design basis accidents to remove the heat from the Main Control Room. The Air Filtration Unit remains operable. This places both units in a 30-day Tech. Spec. LCO."
The licensee plans to notify the NRC resident inspector.
Power Reactor
Event Number: 35720
Facility: BROWNS FERRY
Region: 2 State: AL
Unit: [] [2] []
RX Type: [1] GE-4,[2] GE-4,[3] GE-4
NRC Notified By: RODNEY NACOSTE
HQ OPS Officer: JOHN MacKINNON
Region: 2 State: AL
Unit: [] [2] []
RX Type: [1] GE-4,[2] GE-4,[3] GE-4
NRC Notified By: RODNEY NACOSTE
HQ OPS Officer: JOHN MacKINNON
Notification Date: 05/15/1999
Notification Time: 18:25 [ET]
Event Date: 05/15/1999
Event Time: 14:56 [CDT]
Last Update Date: 05/20/1999
Notification Time: 18:25 [ET]
Event Date: 05/15/1999
Event Time: 14:56 [CDT]
Last Update Date: 05/20/1999
Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(2)(ii) - ESF ACTUATION
10 CFR Section:
50.72(b)(2)(ii) - ESF ACTUATION
Person (Organization):
RUDOLPH BERNHARD (R2)
RUDOLPH BERNHARD (R2)
| Unit | SCRAM Code | RX Crit | Initial PWR | Initial RX Mode | Current PWR | Current RX Mode |
|---|---|---|---|---|---|---|
| 2 | A/R | Y | 100 | Power Operation | 0 | Hot Shutdown |
TURBINE TRIP/REACTOR SCRAM DURING MAIN TURBINE MECHANICAL OVERSPEED TESTING
While the licensee was resetting the mechanical overspeed trip during main turbine mechanical overspeed testing, the main turbine tripped which generated an automatic reactor scram. All control rods fully inserted into the core. The resulting scram caused reactor vessel water level to lower to below the primary containment isolation system group 2 (drywell), group 3 (reactor water cleanup), group 6 (ventilation), and group 8 (traversing incore probes) isolation setpoint resulting in those isolations along with control room emergency ventilation system and the standby gas treatment system initiations. Reactor feedwater is maintaining reactor vessel water level, and the turbine bypass valves are maintaining reactor vessel pressure. All emergency core cooling systems and the emergency diesel generators are fully operable if needed. None of the safety-relief valves opened, and the offsite electrical grid is stable.
The licensee is investigating what caused the turbine trip/reactor scram. A Licensee Event Report will be filed with the NRC within 30 days concerning this event.
The NRC Resident Inspector was notified of this event by the licensee.
* * * UPDATE AT 1207 ON 05/20/99 BY JOE BENNETT TO JOLLIFFE * * *
During an investigation of the above event, the licensee determined that five main steam relief valves had opened automatically due to high reactor pressure and immediately closed completely following the reactor scram. The licensee plans to include this additional information in their Licensee Event Report on this event.
The licensee plans to inform the NRC Resident Inspector of this update. The NRC Operations Officer notified the R2DO Ann Boland.
While the licensee was resetting the mechanical overspeed trip during main turbine mechanical overspeed testing, the main turbine tripped which generated an automatic reactor scram. All control rods fully inserted into the core. The resulting scram caused reactor vessel water level to lower to below the primary containment isolation system group 2 (drywell), group 3 (reactor water cleanup), group 6 (ventilation), and group 8 (traversing incore probes) isolation setpoint resulting in those isolations along with control room emergency ventilation system and the standby gas treatment system initiations. Reactor feedwater is maintaining reactor vessel water level, and the turbine bypass valves are maintaining reactor vessel pressure. All emergency core cooling systems and the emergency diesel generators are fully operable if needed. None of the safety-relief valves opened, and the offsite electrical grid is stable.
The licensee is investigating what caused the turbine trip/reactor scram. A Licensee Event Report will be filed with the NRC within 30 days concerning this event.
The NRC Resident Inspector was notified of this event by the licensee.
* * * UPDATE AT 1207 ON 05/20/99 BY JOE BENNETT TO JOLLIFFE * * *
During an investigation of the above event, the licensee determined that five main steam relief valves had opened automatically due to high reactor pressure and immediately closed completely following the reactor scram. The licensee plans to include this additional information in their Licensee Event Report on this event.
The licensee plans to inform the NRC Resident Inspector of this update. The NRC Operations Officer notified the R2DO Ann Boland.
Power Reactor
Event Number: 35721
Facility: SAN ONOFRE
Region: 4 State: CA
Unit: [] [] [3]
RX Type: [1] W-3-LP,[2] CE,[3] CE
NRC Notified By: MATT THURBURN
HQ OPS Officer: LEIGH TROCINE
Region: 4 State: CA
Unit: [] [] [3]
RX Type: [1] W-3-LP,[2] CE,[3] CE
NRC Notified By: MATT THURBURN
HQ OPS Officer: LEIGH TROCINE
Notification Date: 05/16/1999
Notification Time: 04:08 [ET]
Event Date: 05/15/1999
Event Time: 23:58 [PDT]
Last Update Date: 05/16/1999
Notification Time: 04:08 [ET]
Event Date: 05/15/1999
Event Time: 23:58 [PDT]
Last Update Date: 05/16/1999
Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(2)(ii) - RPS ACTUATION
10 CFR Section:
50.72(b)(2)(ii) - RPS ACTUATION
Person (Organization):
CHARLES CAIN (R4)
CHARLES CAIN (R4)
| Unit | SCRAM Code | RX Crit | Initial PWR | Initial RX Mode | Current PWR | Current RX Mode |
|---|---|---|---|---|---|---|
| 3 | M/R | Y | 24 | Power Operation | 0 | Hot Standby |
MANUAL REACTOR TRIP DUE TO FAILURE OF A MAIN FEEDWATER REGULATING VALVE AND SUBSEQUENT INCREASING STEAM GENERATOR WATER LEVELS (Refer to event #35713 for a similar event that occurred on 05/13/99.)
While operators were making preparations to place the turbine on line, feedwater regulation valve #3FV1111 for the #1 steam generator (E089) failed open. Operators attempted to manually close the valve, but the valve did not respond. As steam generator water levels trended up, a pre-trip alarm on high steam generator level was received, and operators manually tripped the reactor from 24% power at 2358 PDT on 05/15/99. (Steam generator water levels did not reach the trip setpoint.) All rods fully inserted, and all systems functioned as required. There were no engineered safety feature or emergency core cooling system actuations, and none were required. The main steam isolation valves remained open, none of the relief valves opened, and there was no loss of offsite power.
Reactor coolant system (RCS) cooled down farther than normal following the manual trip due to the overfeeding of the steam generators. RCS temperature normally stabilizes at approximately 545°F. However, during this transient, the RCS cooled down to approximately 530°F before operators tripped the main feedwater pump and manually initiated auxiliary feedwater.
The reactor is currently stable in Mode 3 (Hot Standby). The reactor coolant pumps, normal charging and letdown, and pressurizer heaters and sprays are being utilized for primary system transport, level, and pressure control. All containment parameters are normal. Auxiliary feedwater is supplying water to the steam generators, and secondary steam is being dumped to the main condenser. Offsite power is available. The emergency diesel generators and emergency core cooling systems are operable and available if needed.
The cause of the feedwater regulation valve failure is under investigation. The licensee stated that there were no maintenance activities ongoing at the time of the event.
The licensee notified the NRC resident inspector.
While operators were making preparations to place the turbine on line, feedwater regulation valve #3FV1111 for the #1 steam generator (E089) failed open. Operators attempted to manually close the valve, but the valve did not respond. As steam generator water levels trended up, a pre-trip alarm on high steam generator level was received, and operators manually tripped the reactor from 24% power at 2358 PDT on 05/15/99. (Steam generator water levels did not reach the trip setpoint.) All rods fully inserted, and all systems functioned as required. There were no engineered safety feature or emergency core cooling system actuations, and none were required. The main steam isolation valves remained open, none of the relief valves opened, and there was no loss of offsite power.
Reactor coolant system (RCS) cooled down farther than normal following the manual trip due to the overfeeding of the steam generators. RCS temperature normally stabilizes at approximately 545°F. However, during this transient, the RCS cooled down to approximately 530°F before operators tripped the main feedwater pump and manually initiated auxiliary feedwater.
The reactor is currently stable in Mode 3 (Hot Standby). The reactor coolant pumps, normal charging and letdown, and pressurizer heaters and sprays are being utilized for primary system transport, level, and pressure control. All containment parameters are normal. Auxiliary feedwater is supplying water to the steam generators, and secondary steam is being dumped to the main condenser. Offsite power is available. The emergency diesel generators and emergency core cooling systems are operable and available if needed.
The cause of the feedwater regulation valve failure is under investigation. The licensee stated that there were no maintenance activities ongoing at the time of the event.
The licensee notified the NRC resident inspector.