Event Notification Report for February 06, 2002
U.S. Nuclear Regulatory Commission
Operations Center
EVENT REPORTS FOR
02/05/2002 - 02/06/2002
EVENT NUMBERS
38682386753867638829
General Information or Other
Event Number: 38682
Rep Org: WA DIVISION OF RADIATION PROTECTION
Licensee: TRANSALTA MINING CENTRALIA
Region: 4
City: CENTRALIA State: WA
County:
License #: WN-I0241-1
Agreement: Y
Docket:
NRC Notified By: TERRY FRAZEE
HQ OPS Officer: STEVE SANDIN
Licensee: TRANSALTA MINING CENTRALIA
Region: 4
City: CENTRALIA State: WA
County:
License #: WN-I0241-1
Agreement: Y
Docket:
NRC Notified By: TERRY FRAZEE
HQ OPS Officer: STEVE SANDIN
Notification Date: 02/07/2002
Notification Time: 17:29 [ET]
Event Date: 02/06/2002
Event Time: 00:00 [PST]
Last Update Date: 02/07/2002
Notification Time: 17:29 [ET]
Event Date: 02/06/2002
Event Time: 00:00 [PST]
Last Update Date: 02/07/2002
Emergency Class: NON EMERGENCY
10 CFR Section:
10 CFR Section:
Person (Organization):
JOHN PELLET (R4)
M. WAYNE HODGES (NMSS)
JOHN PELLET (R4)
M. WAYNE HODGES (NMSS)
AGREEMENT STATE REPORT INVOLVING POTENTIAL PERSONNEL OVEREXPOSURE
This is notification of an event in Washington state as investigated by the WA Department of Health, Division of Radiation Protection.
STATUS: new
Licensee: TRANSALTA MINING CENTRALIA
City and state: CENTRALIA, WA
License number: WN-10241-1
Type of license: FIXED GAUGE
Date of event: February 6, 2002
Location of Event: Centralia, WA
ABSTRACT: (where, when, how, why; cause, contributing factors, corrective actions, consequences, DOH on-site investigation; media attention) The licensee reported a serious mechanical failure of a Ronan Engineering
Company model SA-4 gauging device. This is the same device which previously experienced several failures of the source rod becoming disconnected from the source. This failure is the most serious to date. The operator pulled the source rod, complete with source, entirely out of the shield while conducting a routine end-of-shift securing of the source. The RSO was informed and, while moving what he thought was just the source rod to a separate storage area, decided to check whether the source was still connected. An old CDV instrument had been provided to the licensee following the previous failures. This instrument confirmed that significant radiation exposure was occurring. The RSO acted quickly to isolate the source and notify the state. The state Division of Radiation Protection arrived within several hours and assisted the licensee in assessing the need for additional shielding. Initial readings with minimal shielding in place indicated a surface dose rate that was "off scale" (greater than 5000 mrem/hour) and about 3000 mrem/hour at 6 inches. Bags of lead shot were obtained and the maximum dose rate on the exterior of the storage building went from 50 mrem/hour to 0.25 mrem/hour. State staff are on-site February 7 to conduct interviews, time-motion studies, and additional exposure measurements to assess personnel exposure.
What is the notification or reporting criteria involved? WAC 246-221-250
Notification of incidents (equipment malfunction; personnel exposure)
Activity and Isotope(s) involved: 1.7 Ci Cesium-137
Overexposures? (number of workers/members of the public; dose estimate; body part receiving dose; consequence) Two workers exposed. Doses unknown at this time.
Lost, Stolen or Damaged? (mfg., model, serial number) Ronan Engineering Company Model SA-4 gauging device, serial number 0973GK, with Amersham Model CDC.711M source.
This is notification of an event in Washington state as investigated by the WA Department of Health, Division of Radiation Protection.
STATUS: new
Licensee: TRANSALTA MINING CENTRALIA
City and state: CENTRALIA, WA
License number: WN-10241-1
Type of license: FIXED GAUGE
Date of event: February 6, 2002
Location of Event: Centralia, WA
ABSTRACT: (where, when, how, why; cause, contributing factors, corrective actions, consequences, DOH on-site investigation; media attention) The licensee reported a serious mechanical failure of a Ronan Engineering
Company model SA-4 gauging device. This is the same device which previously experienced several failures of the source rod becoming disconnected from the source. This failure is the most serious to date. The operator pulled the source rod, complete with source, entirely out of the shield while conducting a routine end-of-shift securing of the source. The RSO was informed and, while moving what he thought was just the source rod to a separate storage area, decided to check whether the source was still connected. An old CDV instrument had been provided to the licensee following the previous failures. This instrument confirmed that significant radiation exposure was occurring. The RSO acted quickly to isolate the source and notify the state. The state Division of Radiation Protection arrived within several hours and assisted the licensee in assessing the need for additional shielding. Initial readings with minimal shielding in place indicated a surface dose rate that was "off scale" (greater than 5000 mrem/hour) and about 3000 mrem/hour at 6 inches. Bags of lead shot were obtained and the maximum dose rate on the exterior of the storage building went from 50 mrem/hour to 0.25 mrem/hour. State staff are on-site February 7 to conduct interviews, time-motion studies, and additional exposure measurements to assess personnel exposure.
What is the notification or reporting criteria involved? WAC 246-221-250
Notification of incidents (equipment malfunction; personnel exposure)
Activity and Isotope(s) involved: 1.7 Ci Cesium-137
Overexposures? (number of workers/members of the public; dose estimate; body part receiving dose; consequence) Two workers exposed. Doses unknown at this time.
Lost, Stolen or Damaged? (mfg., model, serial number) Ronan Engineering Company Model SA-4 gauging device, serial number 0973GK, with Amersham Model CDC.711M source.
Fuel Cycle Facility
Event Number: 38675
Facility: PORTSMOUTH GASEOUS DIFFUSION PLANT
Region: 3 State: OH
Unit: [] [] []
RX Type: URANIUM ENRICHMENT FACILITY
Comments: 2 DEMOCRACY CENTER
6903 ROCKLEDGE DRIVE
BETHESDA, MD 20817 (301)564-3200
NRC Notified By: KURT SISLER
HQ OPS Officer: BOB STRANSKY
Region: 3 State: OH
Unit: [] [] []
RX Type: URANIUM ENRICHMENT FACILITY
Comments: 2 DEMOCRACY CENTER
6903 ROCKLEDGE DRIVE
BETHESDA, MD 20817 (301)564-3200
NRC Notified By: KURT SISLER
HQ OPS Officer: BOB STRANSKY
Notification Date: 02/06/2002
Notification Time: 20:51 [ET]
Event Date: 02/06/2002
Event Time: 19:19 [EST]
Last Update Date: 02/06/2002
Notification Time: 20:51 [ET]
Event Date: 02/06/2002
Event Time: 19:19 [EST]
Last Update Date: 02/06/2002
Emergency Class: NON EMERGENCY
10 CFR Section:
10 CFR Section:
Person (Organization):
THOMAS KOZAK (R3)
M. WAYNE HODGES (NMSS)
THOMAS KOZAK (R3)
M. WAYNE HODGES (NMSS)
OFFSITE NOTIFICATION OF OPACITY EXCEEDANCE
"On 2/6/02, while placing X-600 Steam Plant #2 boiler in service, two opacity readings (six minute average) were above 20% within one hour. This requires a verbal notification to the Ohio Environmental Protection Agency (OEPA).
"At 1919 hours, the Plant Shift Superintendent notified OEPA of the X-600 Steam Plant opacity exceedance. UE2-RA-RE1030, App. D, section P requires a four hour NRC event when other government agencies are notified."
The NRC resident inspector has been informed of this notification.
"On 2/6/02, while placing X-600 Steam Plant #2 boiler in service, two opacity readings (six minute average) were above 20% within one hour. This requires a verbal notification to the Ohio Environmental Protection Agency (OEPA).
"At 1919 hours, the Plant Shift Superintendent notified OEPA of the X-600 Steam Plant opacity exceedance. UE2-RA-RE1030, App. D, section P requires a four hour NRC event when other government agencies are notified."
The NRC resident inspector has been informed of this notification.
!!!!! THIS EVENT HAS BEEN RETRACTED !!!!!
!!!!! THIS EVENT HAS BEEN RETRACTED !!!!!
Power Reactor
Event Number: 38676
Facility: LIMERICK
Region: 1 State: PA
Unit: [1] [] []
RX Type: [1] GE-4,[2] GE-4
NRC Notified By: PETE ORPHANOS
HQ OPS Officer: BOB STRANSKY
Region: 1 State: PA
Unit: [1] [] []
RX Type: [1] GE-4,[2] GE-4
NRC Notified By: PETE ORPHANOS
HQ OPS Officer: BOB STRANSKY
Notification Date: 02/06/2002
Notification Time: 22:37 [ET]
Event Date: 02/06/2002
Event Time: 17:22 [EST]
Last Update Date: 02/08/2002
Notification Time: 22:37 [ET]
Event Date: 02/06/2002
Event Time: 17:22 [EST]
Last Update Date: 02/08/2002
Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(3)(v)(D) - ACCIDENT MITIGATION
10 CFR Section:
50.72(b)(3)(v)(D) - ACCIDENT MITIGATION
Person (Organization):
LAWRENCE DOERFLEIN (R1)
LAWRENCE DOERFLEIN (R1)
| Unit | SCRAM Code | RX Crit | Initial PWR | Initial RX Mode | Current PWR | Current RX Mode |
|---|---|---|---|---|---|---|
| 1 | N | Y | 94 | Power Operation | 94 | Power Operation |
HPCI DECLARED INOPERABLE
"During the completion of scheduled testing on the Unit 1 High Pressure Coolant injection (HPCI) system, a valve designed to provide minimum flow protection for the associated pump did not automatically open in the time frame expected during pump shutdown. The valve did respond correctly to Operator action.
"HPCI was declared inoperable as of 1722 hours on February 6th, 2002. Troubleshooting on the minimum flow protection circuitry is currently in progress.
"This report is being made in pursuant to 10CFR50.72(b)(3)(V)(D) for failure of a single train accident mitigation system."
The NRC resident inspector has been informed of this event by the licensee.
* * * UPDATE 1508EST ON 2/8/02 FROM STAN GAMBLE TO S. SANDIN * * *
The licensee is retracting this report based on the following:
"This is a retraction of the event notification made on 2/6/02 at 22:37 hours. This event (#38676) was initially reported as a safety system functional failure under the requirement of 10CFR50.72(b)(3)(v)(D), after the Unit 1 High Pressure Coolant Injection (HPCI) minimum flow valve did not automatically open during system testing.
"Subsequent testing verified that the valve would have opened automatically without operator action following the low flow alarm actuation. A delay in response is expected when system flow is reduced to a value within the range of the flow transmitter. This delay in response, and the short duration of pump operation at a no flow condition is not detrimental to the HPCI pump.
"The system response was verified to be as designed considering the test conditions. A condition did not exist at the time of discovery that could have prevented the fulfillment of the safety function."
The licensee informed the NRC Resident Inspector. Notified R1DO(Doerflein).
"During the completion of scheduled testing on the Unit 1 High Pressure Coolant injection (HPCI) system, a valve designed to provide minimum flow protection for the associated pump did not automatically open in the time frame expected during pump shutdown. The valve did respond correctly to Operator action.
"HPCI was declared inoperable as of 1722 hours on February 6th, 2002. Troubleshooting on the minimum flow protection circuitry is currently in progress.
"This report is being made in pursuant to 10CFR50.72(b)(3)(V)(D) for failure of a single train accident mitigation system."
The NRC resident inspector has been informed of this event by the licensee.
* * * UPDATE 1508EST ON 2/8/02 FROM STAN GAMBLE TO S. SANDIN * * *
The licensee is retracting this report based on the following:
"This is a retraction of the event notification made on 2/6/02 at 22:37 hours. This event (#38676) was initially reported as a safety system functional failure under the requirement of 10CFR50.72(b)(3)(v)(D), after the Unit 1 High Pressure Coolant Injection (HPCI) minimum flow valve did not automatically open during system testing.
"Subsequent testing verified that the valve would have opened automatically without operator action following the low flow alarm actuation. A delay in response is expected when system flow is reduced to a value within the range of the flow transmitter. This delay in response, and the short duration of pump operation at a no flow condition is not detrimental to the HPCI pump.
"The system response was verified to be as designed considering the test conditions. A condition did not exist at the time of discovery that could have prevented the fulfillment of the safety function."
The licensee informed the NRC Resident Inspector. Notified R1DO(Doerflein).
Power Reactor
Event Number: 38829
Facility: CATAWBA
Region: 2 State: SC
Unit: [1] [2] []
RX Type: [1] W-4-LP,[2] W-4-LP
NRC Notified By: JEFF BRADLEY
HQ OPS Officer: MIKE NORRIS
Region: 2 State: SC
Unit: [1] [2] []
RX Type: [1] W-4-LP,[2] W-4-LP
NRC Notified By: JEFF BRADLEY
HQ OPS Officer: MIKE NORRIS
Notification Date: 04/04/2002
Notification Time: 09:25 [ET]
Event Date: 02/06/2002
Event Time: 09:30 [EST]
Last Update Date: 04/04/2002
Notification Time: 09:25 [ET]
Event Date: 02/06/2002
Event Time: 09:30 [EST]
Last Update Date: 04/04/2002
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):
RANDY MUSSER (R2)
RANDY MUSSER (R2)
| Unit | SCRAM Code | RX Crit | Initial PWR | Initial RX Mode | Current PWR | Current RX Mode |
|---|---|---|---|---|---|---|
| 1 | N | Y | 100 | Power Operation | 100 | Power Operation |
| 2 | N | Y | 100 | Power Operation | 100 | Power Operation |
60 DAY REPORT-INVALID SPECIFIED SYSTEM ACTUATION
"This 60 -day optional report as allowed by 10 CFR 50.73(a)(1), is being made under the reporting requirement in 10 CFR 50.73(a)(2)(iv)(A) to describe an invalid actuation of a specified system, specifically the Nuclear Service Water System (NSWS). On February 6, 2002, at 0930, with Unit 1 in MODE 1 at 100% power and Unit 2 in MODE 1 at 100% power, an invalid actuation of the NSWS suction transfer on Low Pit Level was generated from NSWS train B on Unit 2. Maintenance technicians were in the process of implementing a plant modification to replace a relay in the Unit 2, Train B, NSWS suction transfer circuit. As a part of the plant modification, an isolation plan had been developed by engineering to prevent any actuation of this circuit. However, upon removal of the relay, the NSWS suction transfer on low pit level occurred. The control room operators entered appropriate procedures and verified proper equipment operation for the exiting plant conditions."
"This was a complete actuation for Train B and the system performed as designed for the existing plant conditions. At the time of the invalid signal, the NSWS system had been transferred from the normal supply (Lake Wylie) to the backup supply [Standby Nuclear Service Water Pond (SNSWP)] and appropriate valves had been already repositioned. This event has been entered into the site specific corrective action program for resolution."
The Licensee has notified the NRC Resident Inspector. The Licensee will be notifying the states of NC and SC, and will be notifying local (York, Gaston and Meekenburg) counties.
"This 60 -day optional report as allowed by 10 CFR 50.73(a)(1), is being made under the reporting requirement in 10 CFR 50.73(a)(2)(iv)(A) to describe an invalid actuation of a specified system, specifically the Nuclear Service Water System (NSWS). On February 6, 2002, at 0930, with Unit 1 in MODE 1 at 100% power and Unit 2 in MODE 1 at 100% power, an invalid actuation of the NSWS suction transfer on Low Pit Level was generated from NSWS train B on Unit 2. Maintenance technicians were in the process of implementing a plant modification to replace a relay in the Unit 2, Train B, NSWS suction transfer circuit. As a part of the plant modification, an isolation plan had been developed by engineering to prevent any actuation of this circuit. However, upon removal of the relay, the NSWS suction transfer on low pit level occurred. The control room operators entered appropriate procedures and verified proper equipment operation for the exiting plant conditions."
"This was a complete actuation for Train B and the system performed as designed for the existing plant conditions. At the time of the invalid signal, the NSWS system had been transferred from the normal supply (Lake Wylie) to the backup supply [Standby Nuclear Service Water Pond (SNSWP)] and appropriate valves had been already repositioned. This event has been entered into the site specific corrective action program for resolution."
The Licensee has notified the NRC Resident Inspector. The Licensee will be notifying the states of NC and SC, and will be notifying local (York, Gaston and Meekenburg) counties.