Event Notification Report for October 24, 2000
U.S. Nuclear Regulatory Commission
Operations Center
EVENT REPORTS FOR
10/23/2000 - 10/24/2000
General Information or Other
Event Number: 37459
Rep Org: GEORGIA RADIOACTIVE MATERIAL PGM
Licensee: UNKNOWN
Region: 2
City: EAST POINT State: GA
County:
License #:
Agreement: Y
Docket:
NRC Notified By: TOM HILL
HQ OPS Officer: JOHN MacKINNON
Licensee: UNKNOWN
Region: 2
City: EAST POINT State: GA
County:
License #:
Agreement: Y
Docket:
NRC Notified By: TOM HILL
HQ OPS Officer: JOHN MacKINNON
Notification Date: 10/26/2000
Notification Time: 16:03 [ET]
Event Date: 10/24/2000
Event Time: 12:00 [EDT]
Last Update Date: 10/26/2000
Notification Time: 16:03 [ET]
Event Date: 10/24/2000
Event Time: 12:00 [EDT]
Last Update Date: 10/26/2000
Emergency Class: NON EMERGENCY
10 CFR Section:
10 CFR Section:
Person (Organization):
ANNE BOLAND (R2)
JOSIE PICCONE (NMSS)
ANNE BOLAND (R2)
JOSIE PICCONE (NMSS)
STRONTIUM-90 SOURCE FOUND IN SCRAP METAL.
Environmental Radiation Program (ERP) personnel were notified by South Carolina DHEC of a shipment of scrap metal returning to Newell Recycling; East Point, Georgia; from SMI Steel in Caycese, South Carolina. October 25, 2000 ERP staff surveyed the exterior of the truck and located the area with elevated radiation levels in the front left corner of the trailer. Personnel from Newell and the ERP began unloading scrap until a Sr-90 source was located lying on the floor of the trailer. The source was removed, bagged up and isolated approximately 30 yards away from the trailer in another trailer that could be locked and secured as a temporary storage area until remedial actions can be taken by Newell. Source markings were as follows 3F1L, 100mCi, SR-90, 3M1389, 7-84. Note: Based on the information on the source, the Registry of Radioactive Sealed Sources and Devices registry number for this source is NR-460-S-921-S.
Initial radiation measurements on the outside of the trailer in the area of the front corner were as high as 300 micro-R/hour. Levels in the immediate area of the source inside the trailer were greater than 3R/hour (beta/gamma) unshielded and in contact with the source. After isolating the source, the remaining scrap metal was surveyed by ERP personnel using Ludlum Micro-R Meter and Eberline ESP-2 with HP-260 probe. No radiation levels above background were noted. General radiation levels inside the trailer ranged from 5 to 15 micro-R/hour while background radiation levels outside of the trailer were 15 to 20 micro-R/hour. 26 smears were taken on various pieces of scrap removed from the trailer and the floor in the vicinity of where the source was found. No removable or fixed contamination was noted once the source had been removed and isolated
State of Georgia event number is GA-2000-20I.
Environmental Radiation Program (ERP) personnel were notified by South Carolina DHEC of a shipment of scrap metal returning to Newell Recycling; East Point, Georgia; from SMI Steel in Caycese, South Carolina. October 25, 2000 ERP staff surveyed the exterior of the truck and located the area with elevated radiation levels in the front left corner of the trailer. Personnel from Newell and the ERP began unloading scrap until a Sr-90 source was located lying on the floor of the trailer. The source was removed, bagged up and isolated approximately 30 yards away from the trailer in another trailer that could be locked and secured as a temporary storage area until remedial actions can be taken by Newell. Source markings were as follows 3F1L, 100mCi, SR-90, 3M1389, 7-84. Note: Based on the information on the source, the Registry of Radioactive Sealed Sources and Devices registry number for this source is NR-460-S-921-S.
Initial radiation measurements on the outside of the trailer in the area of the front corner were as high as 300 micro-R/hour. Levels in the immediate area of the source inside the trailer were greater than 3R/hour (beta/gamma) unshielded and in contact with the source. After isolating the source, the remaining scrap metal was surveyed by ERP personnel using Ludlum Micro-R Meter and Eberline ESP-2 with HP-260 probe. No radiation levels above background were noted. General radiation levels inside the trailer ranged from 5 to 15 micro-R/hour while background radiation levels outside of the trailer were 15 to 20 micro-R/hour. 26 smears were taken on various pieces of scrap removed from the trailer and the floor in the vicinity of where the source was found. No removable or fixed contamination was noted once the source had been removed and isolated
State of Georgia event number is GA-2000-20I.
!!!!! THIS EVENT HAS BEEN RETRACTED !!!!!
!!!!! THIS EVENT HAS BEEN RETRACTED !!!!!
Power Reactor
Event Number: 37452
Facility: SOUTH TEXAS
Region: 4 State: TX
Unit: [1] [] []
RX Type: [1] W-4-LP,[2] W-4-LP
NRC Notified By: BILL MOOKHOEK
HQ OPS Officer: FANGIE JONES
Region: 4 State: TX
Unit: [1] [] []
RX Type: [1] W-4-LP,[2] W-4-LP
NRC Notified By: BILL MOOKHOEK
HQ OPS Officer: FANGIE JONES
Notification Date: 10/24/2000
Notification Time: 09:35 [ET]
Event Date: 10/24/2000
Event Time: 04:58 [CDT]
Last Update Date: 10/25/2000
Notification Time: 09:35 [ET]
Event Date: 10/24/2000
Event Time: 04:58 [CDT]
Last Update Date: 10/25/2000
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):
GREG PICK (R4)
GREG PICK (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 |
ESF ACTUATION RESULTING IN A SINGLE VALVE ISOLATION
"On October 24, 2000, at 0458 hours, while Unit 1 was operating at 100% power, a spurious actuation of the Train C Phase B Containment Isolation Signal occurred. This actuation resulted in the closure of a single Engineered Safety Features Containment Isolation Valve servicing the cooling supply to the Reactor Coolant Pumps. Cooling to the Reactor Coolant Pumps was not interrupted since there are two parallel paths in this penetration to prevent inadvertent isolation of the cooling supply. No other components changed position and the Unit continued to operate at 100% power. The cause of the actuation is not yet known, but the actuation only effected valve 1-CC-MOV-0404. Trouble shooting is in process to determine the cause of the event, No other components are controlled by the Train C Phase B actuation logic."
The licensee notified the NRC Resident Inspector.
* * * UPDATE ON 10/25/00 @ 1557 BY MOOKHOEK TO GOULD * * * RETRACTION
South Texas Project notified the NRC of an Engineered Safety Feature actuation at 0835 CDT on October 24, 2000 (Event #37452) as a result of a spurious actuation of the Train C Phase B Containment Isolation Signal. Review of the guidance in NUREG 1022 Revision 1 on ESF actuations indicates that the actuation of single components would not be reportable. The referenced notification reported the actuation of a single component. 1-CC-MOV-0404. Cooling to the Reactor Coolant Pumps was not interrupted by the closure of 1-CC-MOV-0404 since there are two parallel paths in this penetration to prevent inadvertent isolation of the cooling supply. Since this actuation did not stop the flow of cooling water through the containment penetration and only a single component was actuated, South Texas Project has determined that this event is not reportable and retracts the notification.
The NRC Resident has been informed.
REG 4 RDO(Pick) was notified.
"On October 24, 2000, at 0458 hours, while Unit 1 was operating at 100% power, a spurious actuation of the Train C Phase B Containment Isolation Signal occurred. This actuation resulted in the closure of a single Engineered Safety Features Containment Isolation Valve servicing the cooling supply to the Reactor Coolant Pumps. Cooling to the Reactor Coolant Pumps was not interrupted since there are two parallel paths in this penetration to prevent inadvertent isolation of the cooling supply. No other components changed position and the Unit continued to operate at 100% power. The cause of the actuation is not yet known, but the actuation only effected valve 1-CC-MOV-0404. Trouble shooting is in process to determine the cause of the event, No other components are controlled by the Train C Phase B actuation logic."
The licensee notified the NRC Resident Inspector.
* * * UPDATE ON 10/25/00 @ 1557 BY MOOKHOEK TO GOULD * * * RETRACTION
South Texas Project notified the NRC of an Engineered Safety Feature actuation at 0835 CDT on October 24, 2000 (Event #37452) as a result of a spurious actuation of the Train C Phase B Containment Isolation Signal. Review of the guidance in NUREG 1022 Revision 1 on ESF actuations indicates that the actuation of single components would not be reportable. The referenced notification reported the actuation of a single component. 1-CC-MOV-0404. Cooling to the Reactor Coolant Pumps was not interrupted by the closure of 1-CC-MOV-0404 since there are two parallel paths in this penetration to prevent inadvertent isolation of the cooling supply. Since this actuation did not stop the flow of cooling water through the containment penetration and only a single component was actuated, South Texas Project has determined that this event is not reportable and retracts the notification.
The NRC Resident has been informed.
REG 4 RDO(Pick) was notified.
Power Reactor
Event Number: 37453
Facility: SURRY
Region: 2 State: VA
Unit: [1] [] []
RX Type: [1] W-3-LP,[2] W-3-LP
NRC Notified By: LISA HILBERT-SEMMES
HQ OPS Officer: FANGIE JONES
Region: 2 State: VA
Unit: [1] [] []
RX Type: [1] W-3-LP,[2] W-3-LP
NRC Notified By: LISA HILBERT-SEMMES
HQ OPS Officer: FANGIE JONES
Notification Date: 10/24/2000
Notification Time: 10:30 [ET]
Event Date: 10/24/2000
Event Time: 07:41 [EDT]
Last Update Date: 10/24/2000
Notification Time: 10:30 [ET]
Event Date: 10/24/2000
Event Time: 07:41 [EDT]
Last Update Date: 10/24/2000
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):
ANNE BOLAND (R2)
ANNE BOLAND (R2)
| Unit | SCRAM Code | RX Crit | Initial PWR | Initial RX Mode | Current PWR | Current RX Mode |
|---|---|---|---|---|---|---|
| 1 | A/R | Y | 100 | Power Operation | 0 | Hot Shutdown |
AUTOMATIC REACTOR TRIP DUE TO WORK DONE ON WRONG UNIT'S EHC
"At 0741 hours, with Surry Power Station Unit 1 at 100% reactor power, a low low S/G level condition was received and an automatic reactor trip occurred. lmmediately prior to the trip, the Main Steam Dump Valves were observed to have received an open demand, and a loss of turbine load was exhibited. The loss of turbine load was caused by human error due to work on the wrong unit. It was determined that work was being conducted on the Unit 1 EHC System. The work was intended to be one on the Unit 2 EHC System.
"All control rods inserted into the core as designed. IRPI B8 rod bottom light was not received due to a burned out light bulb. The light bulb was replaced and indicated properly. The IRPI indication for this rod trended with all other IRPI indications, and indicated <20 steps withdrawn. The shutdown margin for Unit 1 was determined to be satisfactory. Auxiliary feedwater automatically initiated as designed on low low steam generator level following the trip. Primary RCS temperature decreased to approximately 543 degrees following the reactor trip.
"No primary safety or power operated relief valves were actuated during he event. As a result of the loss of turbine load, the secondary power operated relief valves actuated during the transient. No indication of primary to secondary leakage exists, therefore no adverse radiological consequences resulted from this event.
"All electrical busses transferred properly following the trip and all emergency diesel generators are operable. There were no radiation releases due to this event, nor were there any personnel injuries or contamination events. The cause of the event is being investigated.
"Unit 1 is currently at Hot Shutdown with RCS temperature being maintained at approximately 547 degrees.
"Unit 2 was not affected by this event and remains stable at CSD with RCS temperature at approximately 100°F and the RCS depressurized to atmospheric pressure with containment Type A test conditions being established.
"This event is being reported in accordance with 10CFR50.72(b)(2)(ii)."
The licensee notified the NRC Resident Inspector.
"At 0741 hours, with Surry Power Station Unit 1 at 100% reactor power, a low low S/G level condition was received and an automatic reactor trip occurred. lmmediately prior to the trip, the Main Steam Dump Valves were observed to have received an open demand, and a loss of turbine load was exhibited. The loss of turbine load was caused by human error due to work on the wrong unit. It was determined that work was being conducted on the Unit 1 EHC System. The work was intended to be one on the Unit 2 EHC System.
"All control rods inserted into the core as designed. IRPI B8 rod bottom light was not received due to a burned out light bulb. The light bulb was replaced and indicated properly. The IRPI indication for this rod trended with all other IRPI indications, and indicated <20 steps withdrawn. The shutdown margin for Unit 1 was determined to be satisfactory. Auxiliary feedwater automatically initiated as designed on low low steam generator level following the trip. Primary RCS temperature decreased to approximately 543 degrees following the reactor trip.
"No primary safety or power operated relief valves were actuated during he event. As a result of the loss of turbine load, the secondary power operated relief valves actuated during the transient. No indication of primary to secondary leakage exists, therefore no adverse radiological consequences resulted from this event.
"All electrical busses transferred properly following the trip and all emergency diesel generators are operable. There were no radiation releases due to this event, nor were there any personnel injuries or contamination events. The cause of the event is being investigated.
"Unit 1 is currently at Hot Shutdown with RCS temperature being maintained at approximately 547 degrees.
"Unit 2 was not affected by this event and remains stable at CSD with RCS temperature at approximately 100°F and the RCS depressurized to atmospheric pressure with containment Type A test conditions being established.
"This event is being reported in accordance with 10CFR50.72(b)(2)(ii)."
The licensee notified the NRC Resident Inspector.