Event Notification Report for August 19, 2006
U.S. Nuclear Regulatory Commission
Operations Center
EVENT REPORTS FOR
08/18/2006 - 08/19/2006
EVENT NUMBERS
42787427884279042878
Power Reactor
Event Number: 42787
Facility: BROWNS FERRY
Region: 2 State: AL
Unit: [] [] [3]
RX Type: [1] GE-4,[2] GE-4,[3] GE-4
NRC Notified By: MATHEW RASMUSSEN
HQ OPS Officer: MARK ABRAMOVITZ
Region: 2 State: AL
Unit: [] [] [3]
RX Type: [1] GE-4,[2] GE-4,[3] GE-4
NRC Notified By: MATHEW RASMUSSEN
HQ OPS Officer: MARK ABRAMOVITZ
Notification Date: 08/19/2006
Notification Time: 15:15 [ET]
Event Date: 08/19/2006
Event Time: 11:05 [CDT]
Last Update Date: 08/19/2006
Notification Time: 15:15 [ET]
Event Date: 08/19/2006
Event Time: 11:05 [CDT]
Last Update Date: 08/19/2006
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):
MARK LESSER (R2)
MARK LESSER (R2)
| Unit | SCRAM Code | RX Crit | Initial PWR | Initial RX Mode | Current PWR | Current RX Mode |
|---|---|---|---|---|---|---|
| 3 | M/R | Y | 100 | Power Operation | 0 | Hot Shutdown |
MANUAL REACTOR SCRAM Of UNIT 3 DUE TO THE LOSS OF BOTH REACTOR RECIRCULATION PUMPS
"At 1105 CST both Reactor Recirculation pumps tripped on Unit 3. In accordance with procedure 3-AOI-68-1, a manual scram was initiated. Reactor water level lowered to Reactor Pressure Vessel Level 3, resulting in the automatic actuation of the Primary Containment Isolation System as expected: Group 2 (RHR Shutdown Cooling), Group 3 (Reactor Water (Clean Up), Group 6 (Ventilation), and Group 8 (Traversing Incore Probe) along with the automatic start of Control Room Emergency Ventilation and all 3 trains of the Standby Gas Treatment System. Reactor water level was recovered to normal levels with the reactor feedwater system.
"During this time Unit 2 was at 100% power and was unaffected by the event.
"This event is reportable as a 4-hour and 8-hour notification along with a 60-day written report in accordance with 10 CFR 50.72(b)(2)(iv)(B), 10 CFR 50.72(b)(3)(iv)(A) and 10 CFR 50.73(a)(2)(iv)(A)."
All control rods fully inserted on the scram. Decay heat is being removed with normal feedwater and the turbine bypass valves. No relief valves lifted during this event. Electrical power to the plant is aligned for the normal shutdown lineup. The cause of this event is under investigation.
The licensee notified the NRC Resident Inspector.
"At 1105 CST both Reactor Recirculation pumps tripped on Unit 3. In accordance with procedure 3-AOI-68-1, a manual scram was initiated. Reactor water level lowered to Reactor Pressure Vessel Level 3, resulting in the automatic actuation of the Primary Containment Isolation System as expected: Group 2 (RHR Shutdown Cooling), Group 3 (Reactor Water (Clean Up), Group 6 (Ventilation), and Group 8 (Traversing Incore Probe) along with the automatic start of Control Room Emergency Ventilation and all 3 trains of the Standby Gas Treatment System. Reactor water level was recovered to normal levels with the reactor feedwater system.
"During this time Unit 2 was at 100% power and was unaffected by the event.
"This event is reportable as a 4-hour and 8-hour notification along with a 60-day written report in accordance with 10 CFR 50.72(b)(2)(iv)(B), 10 CFR 50.72(b)(3)(iv)(A) and 10 CFR 50.73(a)(2)(iv)(A)."
All control rods fully inserted on the scram. Decay heat is being removed with normal feedwater and the turbine bypass valves. No relief valves lifted during this event. Electrical power to the plant is aligned for the normal shutdown lineup. The cause of this event is under investigation.
The licensee notified the NRC Resident Inspector.
Power Reactor
Event Number: 42788
Facility: MONTICELLO
Region: 3 State: MN
Unit: [1] [] []
RX Type: [1] GE-3
NRC Notified By: BART BLAKESLEY
HQ OPS Officer: MARK ABRAMOVITZ
Region: 3 State: MN
Unit: [1] [] []
RX Type: [1] GE-3
NRC Notified By: BART BLAKESLEY
HQ OPS Officer: MARK ABRAMOVITZ
Notification Date: 08/19/2006
Notification Time: 23:10 [ET]
Event Date: 08/19/2006
Event Time: 18:49 [CDT]
Last Update Date: 09/22/2006
Notification Time: 23:10 [ET]
Event Date: 08/19/2006
Event Time: 18:49 [CDT]
Last Update Date: 09/22/2006
Emergency Class: NON EMERGENCY
10 CFR Section:
50.73(a)(1) - INVALID SPECIF SYSTEM ACTUATION
10 CFR Section:
50.73(a)(1) - INVALID SPECIF SYSTEM ACTUATION
Person (Organization):
DAVID HILLS (R3)
DAVID HILLS (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 |
RADIATION MONITOR ELECTRICAL SPIKE CAUSES SECONDARY CONTAINMENT ISOLATION
"The 'B' fuel pool radiation monitor spiked high causing an actuation of the secondary containment relays. 'A' SBGTS [Standby Gas Treatment System] train started, reactor building, turbine building, and rad waste building ventilation isolated. The Drywell Continuous Air Monitor (DW CAM) and O2 analyzer containment valves isolated. The control room ventilation system transferred to the high radiation mode.
"At the time of the occurrence the 'A' Fuel Pool radiation monitor was reading normal (0.5 mr/hr). The 'B' Fuel Pool radiation monitor was reading 80 mr/hr. The 'B Fuel Pool radiation monitor, soon after lowered to 30-40 mr/hr and then back to normal (10 mr/hr). The trip setpoint is 50 mr/hr. The trip was reset. The secondary containment isolation and all ventilation trips were restored to normal. The DW CAM and control room ventilation were restored to normal. A local survey was performed on the refuel floor, readings obtained were < 5 mr/hr at the detector and < 2 mr/hr in surrounding areas. The 'B' fuel pool radiation monitor high level trips were placed in the trip/bypass position. The 'B' fuel pool radiation monitor was placed in downscale trip condition to comply with Technical Specification Table 3.2.4. A team is being established to investigate the cause of the trip."
The licensee notified the NRC Resident Inspector.
* * * Update on 09/22/06 at 1456 ET from Dave Burnett to MacKinnon * * *
"This report is being reclassified from a 50.72 (b)(3)(iv)(A) [valid system actuation] to 50.73(a)(2)(iv)(A) [invalid system actuation].
"Based on further investigation, Monticello has determined that the actuation signal was an invalid spurious signal since other radiation monitors in the area did not indicate any change in value. The cause of the spurious signal was determined to be inducted noise through an unshielded cable which resulted in an increased output current to the instrument. Because the actuation signal was not valid, Monticello is reclassifying the initial event report as an unplanned system actuation under 50.73(a)(2)(iv)(A). This report will be made in lieu of reporting the event as an LER.
"In accordance with 50.73(a)(2)(iv)(A):
"This report is not considered an LER and the report is being made under 50.73(a)(2)(iv)(A). The original event report (ENS #42788) detailed the system affected, whether the actuation was complete or partial, and whether each affected system started and functioned successfully.
"The licensee has notified the NRC Resident Inspector.'
NRC R3DO (Julio Lara) notified.
"The 'B' fuel pool radiation monitor spiked high causing an actuation of the secondary containment relays. 'A' SBGTS [Standby Gas Treatment System] train started, reactor building, turbine building, and rad waste building ventilation isolated. The Drywell Continuous Air Monitor (DW CAM) and O2 analyzer containment valves isolated. The control room ventilation system transferred to the high radiation mode.
"At the time of the occurrence the 'A' Fuel Pool radiation monitor was reading normal (0.5 mr/hr). The 'B' Fuel Pool radiation monitor was reading 80 mr/hr. The 'B Fuel Pool radiation monitor, soon after lowered to 30-40 mr/hr and then back to normal (10 mr/hr). The trip setpoint is 50 mr/hr. The trip was reset. The secondary containment isolation and all ventilation trips were restored to normal. The DW CAM and control room ventilation were restored to normal. A local survey was performed on the refuel floor, readings obtained were < 5 mr/hr at the detector and < 2 mr/hr in surrounding areas. The 'B' fuel pool radiation monitor high level trips were placed in the trip/bypass position. The 'B' fuel pool radiation monitor was placed in downscale trip condition to comply with Technical Specification Table 3.2.4. A team is being established to investigate the cause of the trip."
The licensee notified the NRC Resident Inspector.
* * * Update on 09/22/06 at 1456 ET from Dave Burnett to MacKinnon * * *
"This report is being reclassified from a 50.72 (b)(3)(iv)(A) [valid system actuation] to 50.73(a)(2)(iv)(A) [invalid system actuation].
"Based on further investigation, Monticello has determined that the actuation signal was an invalid spurious signal since other radiation monitors in the area did not indicate any change in value. The cause of the spurious signal was determined to be inducted noise through an unshielded cable which resulted in an increased output current to the instrument. Because the actuation signal was not valid, Monticello is reclassifying the initial event report as an unplanned system actuation under 50.73(a)(2)(iv)(A). This report will be made in lieu of reporting the event as an LER.
"In accordance with 50.73(a)(2)(iv)(A):
"This report is not considered an LER and the report is being made under 50.73(a)(2)(iv)(A). The original event report (ENS #42788) detailed the system affected, whether the actuation was complete or partial, and whether each affected system started and functioned successfully.
"The licensee has notified the NRC Resident Inspector.'
NRC R3DO (Julio Lara) notified.
Other Nuclear Material
Event Number: 42790
Rep Org: PENNONI ASSOCIATES
Licensee: PENNONI ASSOCIATES
Region: 1
City: BETHLEHEM State: PA
County:
License #: 37-1763702
Agreement: N
Docket:
NRC Notified By: CHARLES SNYDER
HQ OPS Officer: JOE O'HARA
Licensee: PENNONI ASSOCIATES
Region: 1
City: BETHLEHEM State: PA
County:
License #: 37-1763702
Agreement: N
Docket:
NRC Notified By: CHARLES SNYDER
HQ OPS Officer: JOE O'HARA
Notification Date: 08/21/2006
Notification Time: 09:47 [ET]
Event Date: 08/19/2006
Event Time: 23:00 [EDT]
Last Update Date: 08/23/2006
Notification Time: 09:47 [ET]
Event Date: 08/19/2006
Event Time: 23:00 [EDT]
Last Update Date: 08/23/2006
Emergency Class: NON EMERGENCY
10 CFR Section:
20.2201(a)(1)(i) - LOST/STOLEN LNM>1000X
10 CFR Section:
20.2201(a)(1)(i) - LOST/STOLEN LNM>1000X
Person (Organization):
RICHARD CONTE (R1)
MICHELE BURGESS (NMSS)
AARON DANIS (TAS)
RICHARD CONTE (R1)
MICHELE BURGESS (NMSS)
AARON DANIS (TAS)
LOST MOISTURE DENSITY GAUGE
On Saturday August 19, 2006 at approximately 11:00 p.m., an employee of Pennoni Associates loaned his private, passenger vehicle to his sister to perform a short errand. Inside the trunk, the vehicle contained a Humboldt Model 5001 moisture density gauge, serial number 4746, with 10 milli Curies of Cs-137 and 40 milli Curies of Am-241/Be. The employee has not seen his vehicle since and contacted the Bethlehem City Police Department. The licensee is continuing the investigation as to the whereabouts of the gauge and the vehicle.
The licensee contacted the NRC Region I office.
*** UPDATE FROM CHARLES SNYDER TO KNOKE AT 13:07 EDT ON 8/23/06 ***
The RSO (Snyder) indicated that the lost density gauge (sn 4746) was found. The employee, who lent his car to his sister-in-law, retrieved the gauge and returned it to Pennoni Associates. There was no damage to the gauge and the sources were not tampered with. The RSO stated that the sister-in-law emptied the contents out of the car, including the gauge, and then took off with the car without permission. She notified the employee where she dumped the contents, and he came and retrieved it. LLEA is following up on the stolen vehicle.
Contacted the R1DO (Conte) and NMSS ( Burgess) and emailed ILTAB (Danis)
THIS MATERIAL EVENT CONTAINS A "LESS THAN CAT 3" LEVEL OF RADIOACTIVE MATERIAL
Sources that are "Less than IAEA Category 3 sources," are either sources that are very unlikely to cause permanent injury to individuals or contain a very small amount of radioactive material that would not cause any permanent injury. Some of these sources, such as moisture density gauges or thickness gauges that are Category 4, the amount of unshielded radioactive material, if not safely managed or securely protected, could possibly - although it is unlikely - temporarily injure someone who handled it or were otherwise in contact with it, or who were close to it for a period of many weeks.
On Saturday August 19, 2006 at approximately 11:00 p.m., an employee of Pennoni Associates loaned his private, passenger vehicle to his sister to perform a short errand. Inside the trunk, the vehicle contained a Humboldt Model 5001 moisture density gauge, serial number 4746, with 10 milli Curies of Cs-137 and 40 milli Curies of Am-241/Be. The employee has not seen his vehicle since and contacted the Bethlehem City Police Department. The licensee is continuing the investigation as to the whereabouts of the gauge and the vehicle.
The licensee contacted the NRC Region I office.
*** UPDATE FROM CHARLES SNYDER TO KNOKE AT 13:07 EDT ON 8/23/06 ***
The RSO (Snyder) indicated that the lost density gauge (sn 4746) was found. The employee, who lent his car to his sister-in-law, retrieved the gauge and returned it to Pennoni Associates. There was no damage to the gauge and the sources were not tampered with. The RSO stated that the sister-in-law emptied the contents out of the car, including the gauge, and then took off with the car without permission. She notified the employee where she dumped the contents, and he came and retrieved it. LLEA is following up on the stolen vehicle.
Contacted the R1DO (Conte) and NMSS ( Burgess) and emailed ILTAB (Danis)
THIS MATERIAL EVENT CONTAINS A "LESS THAN CAT 3" LEVEL OF RADIOACTIVE MATERIAL
Sources that are "Less than IAEA Category 3 sources," are either sources that are very unlikely to cause permanent injury to individuals or contain a very small amount of radioactive material that would not cause any permanent injury. Some of these sources, such as moisture density gauges or thickness gauges that are Category 4, the amount of unshielded radioactive material, if not safely managed or securely protected, could possibly - although it is unlikely - temporarily injure someone who handled it or were otherwise in contact with it, or who were close to it for a period of many weeks.
Power Reactor
Event Number: 42878
Facility: BROWNS FERRY
Region: 2 State: AL
Unit: [] [] [3]
RX Type: [1] GE-4,[2] GE-4,[3] GE-4
NRC Notified By: DONALD C. SMITH
HQ OPS Officer: PETE SNYDER
Region: 2 State: AL
Unit: [] [] [3]
RX Type: [1] GE-4,[2] GE-4,[3] GE-4
NRC Notified By: DONALD C. SMITH
HQ OPS Officer: PETE SNYDER
Notification Date: 10/05/2006
Notification Time: 14:18 [ET]
Event Date: 08/19/2006
Event Time: 20:02 [CDT]
Last Update Date: 10/05/2006
Notification Time: 14:18 [ET]
Event Date: 08/19/2006
Event Time: 20:02 [CDT]
Last Update Date: 10/05/2006
Emergency Class: NON EMERGENCY
10 CFR Section:
50.73(a)(1) - INVALID SPECIF SYSTEM ACTUATION
10 CFR Section:
50.73(a)(1) - INVALID SPECIF SYSTEM ACTUATION
Person (Organization):
THOMAS DECKER (R2)
THOMAS DECKER (R2)
| Unit | SCRAM Code | RX Crit | Initial PWR | Initial RX Mode | Current PWR | Current RX Mode |
|---|---|---|---|---|---|---|
| 3 | N | N | 0 | Hot Shutdown | 0 | Hot Shutdown |
INVALID PRIMARY CONTAINMENT ISOLATION SYSTEM ACTUATION
"This 60-day telephone notification is being made under reporting requirements specified by 10 CFR 50.73(a)(2)(iv)(A) and 10 CFR 50.73(a)(1) to describe an invalid actuation of multiple main steam isolation valves.
"On August 19, 2006, at 2002 hours CDT, with Unit 3 shutdown in Mode 3 following an earlier manual scram of Unit 3 (reference Event Number 42787), the EHC System along with the turbine bypass valves was being used to cool down/depressurize the reactor. The EHC system is used to supervise the turbine control valves and turbine bypass valves to control reactor pressure. The EHC system was subsequently removed from service for the repair of a previously identified system fluid leak.
"Following the shutdown of EHC system, reactor pressure and temperature were allowed to slowly increase. At 2240 hours CDT upon completion of the leak repair, operations personnel placed an EHC pump back in service in accordance with the system operating instruction to support post-maintenance testing activities. However, when the EHC system was returned to service, the pressure control set point was lower than the reactor pressure. The significance of the delta between the EHC setpoint and the actual reactor pressure was not recognized by operations. Because the actual reactor pressure was higher than the existing control set point, the EHC system responded by opening turbine bypass valves to lower the reactor pressure. Operations personnel observed the bypass valve response, identified the cause, and raised the pressure control set point. This action caused the bypass valves to rapidly close. The abrupt cessation of steam flow caused by the rapid closure of the bypass valves initiated a reactor pressure transient that affected the reactor water level instrumentation. The affected level instruments' output signals exhibited a ringing effect of a magnitude sufficient to reach the low level set point for primary containment isolation system (PCIS) Group 1 actuation. The Group 1 isolation logic actuated in accordance with its design, and the main steam isolation valves and the main steam line drain valves automatically closed.
"Designed time delays in other logic circuits affected by these water level signals prevented additional equipment actuation during this event. Actual reactor water level did not change, remaining within the normal level band; therefore, the isolation signal is considered invalid.
"All equipment responded in accordance with the plant design. Upon verification that no actual water level anomaly existed and that the transient instrumentation response had stabilized, the affected PCIS logic was reset, and equipment was realigned as appropriate.
"There were no safety consequences or impacts on the health and safety of the public. The event was entered into TVA's corrective action program for evaluation and resolution.
"The NRC senior resident inspector has been notified of this report.
"Reference corrective action document PER 109118."
"This 60-day telephone notification is being made under reporting requirements specified by 10 CFR 50.73(a)(2)(iv)(A) and 10 CFR 50.73(a)(1) to describe an invalid actuation of multiple main steam isolation valves.
"On August 19, 2006, at 2002 hours CDT, with Unit 3 shutdown in Mode 3 following an earlier manual scram of Unit 3 (reference Event Number 42787), the EHC System along with the turbine bypass valves was being used to cool down/depressurize the reactor. The EHC system is used to supervise the turbine control valves and turbine bypass valves to control reactor pressure. The EHC system was subsequently removed from service for the repair of a previously identified system fluid leak.
"Following the shutdown of EHC system, reactor pressure and temperature were allowed to slowly increase. At 2240 hours CDT upon completion of the leak repair, operations personnel placed an EHC pump back in service in accordance with the system operating instruction to support post-maintenance testing activities. However, when the EHC system was returned to service, the pressure control set point was lower than the reactor pressure. The significance of the delta between the EHC setpoint and the actual reactor pressure was not recognized by operations. Because the actual reactor pressure was higher than the existing control set point, the EHC system responded by opening turbine bypass valves to lower the reactor pressure. Operations personnel observed the bypass valve response, identified the cause, and raised the pressure control set point. This action caused the bypass valves to rapidly close. The abrupt cessation of steam flow caused by the rapid closure of the bypass valves initiated a reactor pressure transient that affected the reactor water level instrumentation. The affected level instruments' output signals exhibited a ringing effect of a magnitude sufficient to reach the low level set point for primary containment isolation system (PCIS) Group 1 actuation. The Group 1 isolation logic actuated in accordance with its design, and the main steam isolation valves and the main steam line drain valves automatically closed.
"Designed time delays in other logic circuits affected by these water level signals prevented additional equipment actuation during this event. Actual reactor water level did not change, remaining within the normal level band; therefore, the isolation signal is considered invalid.
"All equipment responded in accordance with the plant design. Upon verification that no actual water level anomaly existed and that the transient instrumentation response had stabilized, the affected PCIS logic was reset, and equipment was realigned as appropriate.
"There were no safety consequences or impacts on the health and safety of the public. The event was entered into TVA's corrective action program for evaluation and resolution.
"The NRC senior resident inspector has been notified of this report.
"Reference corrective action document PER 109118."