Event Notification Report for June 08, 2000
U.S. Nuclear Regulatory Commission
Operations Center
EVENT REPORTS FOR
06/07/2000 - 06/08/2000
Power Reactor
Event Number: 37068
Facility: COOK
Region: 3 State: MI
Unit: [1] [2] []
RX Type: [1] W-4-LP,[2] W-4-LP
NRC Notified By: DEPUYDT
HQ OPS Officer: CHAUNCEY GOULD
Region: 3 State: MI
Unit: [1] [2] []
RX Type: [1] W-4-LP,[2] W-4-LP
NRC Notified By: DEPUYDT
HQ OPS Officer: CHAUNCEY GOULD
Notification Date: 06/08/2000
Notification Time: 13:11 [ET]
Event Date: 06/08/2000
Event Time: 09:31 [EDT]
Last Update Date: 06/08/2000
Notification Time: 13:11 [ET]
Event Date: 06/08/2000
Event Time: 09:31 [EDT]
Last Update Date: 06/08/2000
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):
JOHN JACOBSON (R3)
ED GOODWIN (NRR)
MICHAEL CASE (EDO)
JOHN JACOBSON (R3)
ED GOODWIN (NRR)
MICHAEL CASE (EDO)
| Unit | SCRAM Code | RX Crit | Initial PWR | Initial RX Mode | Current PWR | Current RX Mode |
|---|---|---|---|---|---|---|
| 1 | N | N | 0 | Refueling | 0 | Refueling |
| 2 | N | N | 0 | Hot Shutdown | 0 | Hot Shutdown |
THE PLANT HAD AUTOSTART OF THE EMERGENCY DIESELS ON BOTH UNITS
On June 8, 2000, at 0931 hours, power was lost to the CD bus on each unit during switching operations involving the #4 and #5 transformers in the main switchyard. During the switching operation a loss of offsite power occurred, affecting one train of 4KV electrical power in each unit and causing the auto start of the CD EDGs on each unit. The EDGs started and loaded as expected.
Unit 1 is currently defueled. When power was lost, the Spent Fuel Pool Cooling pump that was in service tripped. It was placed back in service 0953 hours. The spent fuel pool temperature rose 1 degree from 84 to 85 during this time. Temperature is now stable at 85 degrees.
Unit 2 is currently in Mode 4. When power was lost to the CD bus, Reactor Coolant pumps (RCPs) #22 and #23 tripped. Additionally, the 4 KV bus under frequency signal caused the #21 and #24 RCPs to trip as designed. As a result, Unit 2 was in the natural circulation mode using the steam generators as a heat sink. RCS temperature at the time of loss of power was 335 degrees and cooled to approximately 315 degrees during the time the unit was in natural circulation.
Offsite power was restored through TR5 at 1019 hours. #24 RCP was restarted at 1047 hours, restoring forced circulation, #23 RCP was restarted at 1159 hours. RCPs #21and #22 are still currently secured. The unit is stable and equipment is being restored.
An Event Response Team has been convened to assess the event, and develop recovery plans.
The NRC Resident Inspector was notified.
On June 8, 2000, at 0931 hours, power was lost to the CD bus on each unit during switching operations involving the #4 and #5 transformers in the main switchyard. During the switching operation a loss of offsite power occurred, affecting one train of 4KV electrical power in each unit and causing the auto start of the CD EDGs on each unit. The EDGs started and loaded as expected.
Unit 1 is currently defueled. When power was lost, the Spent Fuel Pool Cooling pump that was in service tripped. It was placed back in service 0953 hours. The spent fuel pool temperature rose 1 degree from 84 to 85 during this time. Temperature is now stable at 85 degrees.
Unit 2 is currently in Mode 4. When power was lost to the CD bus, Reactor Coolant pumps (RCPs) #22 and #23 tripped. Additionally, the 4 KV bus under frequency signal caused the #21 and #24 RCPs to trip as designed. As a result, Unit 2 was in the natural circulation mode using the steam generators as a heat sink. RCS temperature at the time of loss of power was 335 degrees and cooled to approximately 315 degrees during the time the unit was in natural circulation.
Offsite power was restored through TR5 at 1019 hours. #24 RCP was restarted at 1047 hours, restoring forced circulation, #23 RCP was restarted at 1159 hours. RCPs #21and #22 are still currently secured. The unit is stable and equipment is being restored.
An Event Response Team has been convened to assess the event, and develop recovery plans.
The NRC Resident Inspector was notified.
Power Reactor
Event Number: 37069
Facility: INDIAN POINT
Region: 1 State: NY
Unit: [] [3] []
RX Type: [2] W-4-LP,[3] W-4-LP
NRC Notified By: RUSSEL LONG
HQ OPS Officer: DOUG WEAVER
Region: 1 State: NY
Unit: [] [3] []
RX Type: [2] W-4-LP,[3] W-4-LP
NRC Notified By: RUSSEL LONG
HQ OPS Officer: DOUG WEAVER
Notification Date: 06/08/2000
Notification Time: 15:26 [ET]
Event Date: 06/08/2000
Event Time: 15:15 [EDT]
Last Update Date: 06/08/2000
Notification Time: 15:26 [ET]
Event Date: 06/08/2000
Event Time: 15:15 [EDT]
Last Update Date: 06/08/2000
Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(1)(ii)(B) - OUTSIDE DESIGN BASIS
10 CFR Section:
50.72(b)(1)(ii)(B) - OUTSIDE DESIGN BASIS
Person (Organization):
TOM MOSLAK (R1)
TOM MOSLAK (R1)
| Unit | SCRAM Code | RX Crit | Initial PWR | Initial RX Mode | Current PWR | Current RX Mode |
|---|---|---|---|---|---|---|
| 3 | N | Y | 100 | Power Operation | 100 | Power Operation |
OUTSIDE DESIGN BASIS - CABLE SEPARATION ISSUE
At approximately 1515 hours on June 8, 2000, operations determined that plant cable separation criteria had potentially been violated which would place the plant outside the design basis. Two wires were routed from a test jack in the control rod cabinets located in the cable spreading room to a terminal block in Rack C3 in the control room. The routing of the wires takes them into cable trays in the cable spreading room that are carrying control cables for channels 1, 2 and 3. This is in violation of the electrical separation criteria defined in the FSAR. Corrective action is being taken to remove the wires from the cable trays. A preliminary evaluation indicates the wires were used during testing of the control rod drive mechanism during initial plant startup. They are no longer in use. This event was identified during an extent of condition walkdown for the separation criteria violation identified in incident log 37043, dated May 26, 2000.
The licensee notified the NRC resident inspector.
At approximately 1515 hours on June 8, 2000, operations determined that plant cable separation criteria had potentially been violated which would place the plant outside the design basis. Two wires were routed from a test jack in the control rod cabinets located in the cable spreading room to a terminal block in Rack C3 in the control room. The routing of the wires takes them into cable trays in the cable spreading room that are carrying control cables for channels 1, 2 and 3. This is in violation of the electrical separation criteria defined in the FSAR. Corrective action is being taken to remove the wires from the cable trays. A preliminary evaluation indicates the wires were used during testing of the control rod drive mechanism during initial plant startup. They are no longer in use. This event was identified during an extent of condition walkdown for the separation criteria violation identified in incident log 37043, dated May 26, 2000.
The licensee notified the NRC resident inspector.
Power Reactor
Event Number: 37070
Facility: SOUTH TEXAS
Region: 4 State: TX
Unit: [1] [] []
RX Type: [1] W-4-LP,[2] W-4-LP
NRC Notified By: KATHLEEN WORK
HQ OPS Officer: DOUG WEAVER
Region: 4 State: TX
Unit: [1] [] []
RX Type: [1] W-4-LP,[2] W-4-LP
NRC Notified By: KATHLEEN WORK
HQ OPS Officer: DOUG WEAVER
Notification Date: 06/08/2000
Notification Time: 15:59 [ET]
Event Date: 06/08/2000
Event Time: 07:45 [CDT]
Last Update Date: 06/08/2000
Notification Time: 15:59 [ET]
Event Date: 06/08/2000
Event Time: 07:45 [CDT]
Last Update Date: 06/08/2000
Emergency Class: NON EMERGENCY
10 CFR Section:
10 CFR Section:
Person (Organization):
JOHN PELLET (R4)
JOHN PELLET (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 |
24 HOUR REPORT - HIGH BATTERY ELECTROLYTE LEVEL
On May 15, 2000, a surveillance test was performed on the E1A11 battery. The electrolyte level in cell #18 was reported as + ¼ inch above the high level line on the cell which exceeds the Tech spec limit of < ¼ inch, as stated in Tech Spec Table 4.8-2, Battery Surveillance Requirements for Category B. This condition was not recognized during Maintenance review of the data. Tech Spec Surveillance 4.8.2.1.b.1 which allows 7 days to restore electrolyte level within the Category B limits or declare the battery inoperable was exceeded and the battery was considered inoperable. The LCO for an inoperable battery, which is 2 hours to restore to operable or be in hot standby within 6 hours, was not met. This was a violation of Tech Spec 3.8.2.1.a. The condition was corrected on May 30, 2000. There is no safety significance to this event. The electrolyte level in cell 18 at no time compromised the ability of the E1A11 battery to perform its design function.
24-hour notification is being made pursuant to Operating License NPF-76, paragraph 2.G.
The licensee notified the NRC resident inspector.
On May 15, 2000, a surveillance test was performed on the E1A11 battery. The electrolyte level in cell #18 was reported as + ¼ inch above the high level line on the cell which exceeds the Tech spec limit of < ¼ inch, as stated in Tech Spec Table 4.8-2, Battery Surveillance Requirements for Category B. This condition was not recognized during Maintenance review of the data. Tech Spec Surveillance 4.8.2.1.b.1 which allows 7 days to restore electrolyte level within the Category B limits or declare the battery inoperable was exceeded and the battery was considered inoperable. The LCO for an inoperable battery, which is 2 hours to restore to operable or be in hot standby within 6 hours, was not met. This was a violation of Tech Spec 3.8.2.1.a. The condition was corrected on May 30, 2000. There is no safety significance to this event. The electrolyte level in cell 18 at no time compromised the ability of the E1A11 battery to perform its design function.
24-hour notification is being made pursuant to Operating License NPF-76, paragraph 2.G.
The licensee notified the NRC resident inspector.