| U.S. Nuclear Regulatory Commission
Operations Center Event Reports For
06/30/2006 - 07/03/2006 ** EVENT NUMBERS ** |
 | General Information or Other | Event Number: 42670 | Rep Org: SC DIV OF HEALTH & ENV CONTROL
Licensee: SUMMERVILLE MEDICAL CENTER
Region: 1
City: SUMMERVILLE State: SC
County:
License #: 210
Agreement: Y
Docket:
NRC Notified By: MARK L. WINDHAM
HQ OPS Officer: ARLON COSTA | Notification Date: 06/28/2006
Notification Time: 10:40 [ET]
Event Date: 06/27/2006
Event Time: 11:30 [EDT]
Last Update Date: 06/28/2006 | Emergency Class: NON EMERGENCY
10 CFR Section:
AGREEMENT STATE
| Person (Organization):
NEIL PERRY (R1)
GREG MORELL (NMSS)
|
| This material event contains a "Less than Cat 3" level of radioactive material. | Event Text SC AGREEMENT STATE REPORT - SHEARED END OF Pd-103 SEED
The State provided the following information via facsimile:
"The SC Department of Health and Environmental Control was notified on Tuesday, June 27, 2006, at 5:18 p.m., that a Pd-103 [1.33 mCi] seed had jammed in a Mick applicator at 11:30 a.m., on June 27, 2006. Medical Physicist [Name Deleted] stated that the Mick cartridge was unscrewed to release spring pressure, which subsequently sheared the end of the seed off. The cartridge was removed and placed in a lead container, the applicator was flushed out and the seed was placed in a lead container. All remaining seeds were then placed in a shielded container and the implant was completed. [The Medical Physicist] stated that the area was surveyed with a Ludlum Model 3A utilizing a scintillation probe. One towel and water drained in a basin indicated contamination. These items were bagged and taken to the hot lab for decay in storage.
"[The Medical Physicist] was advised by Mark L. Windham to submit a written report detailing this event to the Department within 30 days. The event is considered closed and pending the licensee's investigation and report to the Department, updates will be made through the national NMED system."
SC Report ID No.SC060009
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. |
| General Information or Other | Event Number: 42671 | Rep Org: NV DIV OF RAD HEALTH
Licensee: ACCLAIM MATERIAL TESTING AND INSPECTION
Region: 4
City: HENDERSON State: NV
County:
License #: 00-11-0470-01
Agreement: Y
Docket:
NRC Notified By: KAREN BECKLEY
HQ OPS Officer: JEFF ROTTON | Notification Date: 06/28/2006
Notification Time: 17:54 [ET]
Event Date: 06/27/2006
Event Time: [PDT]
Last Update Date: 06/28/2006 | Emergency Class: NON EMERGENCY
10 CFR Section:
AGREEMENT STATE
| Person (Organization):
GREG PICK (R4)
SCOTT FLANDERS (NMSS)
ILTAB (via email) ()
|
| This material event contains a "Less than Cat 3" level of radioactive material. | Event Text NEVADA AGREEMENT STATE REPORT - STOLEN MOISTURE DENSITY GAUGE
The State provided the following information via email:
"Event Report ID No. NV-006-04
"License name, address, license No.: Acclaim Material Testing and Inspection, 720 Susanna Way, Henderson, Nevada; License No. 00-11-0470-01 [phone number deleted]
" Date and time of occurrence June 7, 2006, approx. 6:30 pm [Determined by review of secured storage room card reader access records]
"Date notified of event by licensee or non-licensee: June 28, 2006, 2:30 pm. The missing gauge was discovered during a routine radiation safety audit by the Radiological Health Section on June 27, 2006.
"Radionuclide, activity: Cs-137, approximately 10 mCi; Am-241:Be, 50 mCi
"Any exposures (indicate short and long-term effects): None known
"Sealed source, device, etc. (make, model #, serial #): Campbell Pacific Nuclear Model MCDRP s/n MD60108156
"Leak test information, If applicable: unknown
"Persons involved, consequences: [name deleted]
.
"Cause and contributing factors: Gauge was removed from its permanent storage site. It appears that the gauge was properly secured.
" Licensee corrective actions: Unknown at this point
"Notifications: Local Law enforcement.
" Enforcement Actions. None contemplated pending licensees final report
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: 42674 | Facility: FARLEY
Region: 2 State: AL
Unit: [1] [ ] [ ]
RX Type: [1] W-3-LP,[2] W-3-LP
NRC Notified By: CHUCK BAREFIELD
HQ OPS Officer: PETE SNYDER | Notification Date: 06/30/2006
Notification Time: 09:43 [ET]
Event Date: 06/30/2006
Event Time: 08:19 [CDT]
Last Update Date: 06/30/2006 | Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(2)(i) - PLANT S/D REQD BY TS
| Person (Organization):
THOMAS DECKER (R2)
|
| Unit | SCRAM Code | RX CRIT | Initial PWR | Initial RX Mode | Current PWR | Current RX Mode | | 1 | N | Y | 100 | Power Operation | 98 | Power Operation | Event Text MAIN STEAM ISOLATION VALVE STUCK DURING TESTING
"[On June 30, 2006 the licensee initiated] a plant shutdown required by Technical Specifications. [The licensee] discovered during surveillance testing, that the Unit 1 Main Steam Isolation Valve (MSIV-3370B) for the B-loop, would not return to the fully open condition after the partial stroke test. The valve appeared to stroke smoothly to the test position. Trouble shooting of the condition has not determined the exact cause however it appears that there is possible binding."
The licensee has tried reasonable methods to free the valve but was unable to do so. The licensee declared the MSIV inoperable. The plant's Technical Specifications require a shutdown within 72 hours with one MSIV inoperable.
The licensee has no other safety significant limiting conditions for operation in effect. The safety related electrical buses are powered from offsite power and the grid is stable in the area.
The licensee notified the NRC Resident Inspector. |
| Power Reactor | Event Number: 42675 | Facility: SAINT LUCIE
Region: 2 State: FL
Unit: [ ] [2] [ ]
RX Type: [1] CE,[2] CE
NRC Notified By: PAUL CHRISTIANSEN
HQ OPS Officer: JOHN KNOKE | Notification Date: 06/30/2006
Notification Time: 14:18 [ET]
Event Date: 06/29/2006
Event Time: 20:30 [EDT]
Last Update Date: 06/30/2006 | Emergency Class: NON EMERGENCY
10 CFR Section:
INFORMATION ONLY
| Person (Organization):
THOMAS DECKER (R2)
|
| Unit | SCRAM Code | RX CRIT | Initial PWR | Initial RX Mode | Current PWR | Current RX Mode | | 2 | N | Y | 100 | Power Operation | 100 | Power Operation | Event Text MAXIMUM POWER LEVEL EXCEEDED AS ALLOWED BY OPERATING LICENSE
"On Thursday, June 29, 2006, at approximately 20:30 the Engineering Department notified the Plant Management that St. Lucie Unit 2 had exceeded the maximum power level allowed by the facility's Operating License (License Condition 3.A). The overpower condition was determined to be approximately 0.5%. This equates to 13.5 MWth reactor core power above the Rated Thermal Power Limit of 2700 MWth. Preliminary data indicates that the overpower condition could have been present between June 20, 2006 and June 29, 2006. This telephone notification is made in accordance with Renewed Facility Operating License No. NPF-16, License Condition 3.H. A written follow-up report will be submitted within 14 days.
"A Distributed Control System (DCS) software error was responsible for the overpower condition. The DCS calculated a slightly lower feedwater flow value which was used as input into the calorimetric power calculation. The error resulted in a non-conservative indicated power in the control room.
"When the feedwater flow calculation error was identified, it was apparent that an overpower condition may have occurred. Therefore, a controlled power reduction to approximately 98% was performed in accordance with plant procedures. St. Lucie Unit 2 will remain at a reduced power level until the software error is corrected. The software error is understood and is in the process of being resolved within the corrective action process. An extent of condition review has determined that St. Lucie Unit 1 does not have the same problem."
The licensee notified the NRC Resident Inspector. |
| Power Reactor | Event Number: 42676 | Facility: CALLAWAY
Region: 4 State: MO
Unit: [1] [ ] [ ]
RX Type: [1] W-4-LP
NRC Notified By: GREG BRADLEY
HQ OPS Officer: JOHN KNOKE | Notification Date: 06/30/2006
Notification Time: 14:59 [ET]
Event Date: 06/30/2006
Event Time: 13:00 [CDT]
Last Update Date: 06/30/2006 | Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(2)(xi) - OFFSITE NOTIFICATION
| Person (Organization):
MICHAEL SHANNON (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 | Event Text LOW LEVELS OF TRITIUM FOUND ALONG PLANT DISCHARGE LINE
"This report is being made pursuant to 10CFR50.72(b)(2)(xi), 4-hour Non-Emergency Report due to a press release being issued by Callaway Plant reporting that its monitoring program recently detected small amounts of tritium and other radionuclides along the plant discharge pipeline. The concentrations of radioactive material in water sampled from manholes on the discharge line range from less than 1% to about 20% of the NRC effluent release concentration limits. This radioactive material is contained on plant owned property.
"All known wells in the vicinity of the plant, as well as drinking water at the plant and in the nearby town of Portland, have been tested and shown to have no detectable radionuclides above background. Further testing of drinking water supplies and in areas around the discharge line is taking place to assure that no migration of the contaminated water occurs. Sampling results of the monitoring program have been provided to the NRC and the Missouri Department of Natural Resources (DNR) and a plan to remediate the contaminated manholes will be coordinated with the DNR.
"No Offsite Dose Calculation Manual (ODCM) and no federal effluent limits have been exceeded. Callaway Plant has not identified any health or safety risk to the public or onsite personnel. The DNR, Missouri Department of Health and Social Services, and local County Officials have been notified of the press release."
The licensee notified the NRC Resident Inspector, and other State and local agencies are being notified via the press release. |
| Power Reactor | Event Number: 42677 | Facility: PERRY
Region: 3 State: OH
Unit: [1] [ ] [ ]
RX Type: [1] GE-6
NRC Notified By: MIKE BROGAN
HQ OPS Officer: JOHN KNOKE | Notification Date: 06/30/2006
Notification Time: 17:55 [ET]
Event Date: 06/30/2006
Event Time: 15:18 [EDT]
Last Update Date: 06/30/2006 | Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(2)(xi) - OFFSITE NOTIFICATION
| Person (Organization):
JOHN MADERA (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 | Event Text NON-CREDIBLE SECURITY THREAT MADE AGAINST PLANT
"Notified by Security Shift Supervisor of a report by the Ashtabula County Sheriff that a 92 year old individual has been checked into the Ashtabula Clinic at 14:57 EDT, claiming to have an explosive device at his Kingsville home and at the Perry Power Plant. This is being considered as a non-credible security threat, based on his never having access to the site, ONI-P56-2 is not applicable, based on this report and assessment. Law enforcement officers are obtaining a warrant to access his home to investigate. The FBI and Lake County Sheriff's offices have been notified by the Security Shift Supervisor. This is an off site report from the Perry Plant to a government agency about the status of the Perry Plant."
The Lake County Sheriff's office completed their search of individual's residence and found no bombs or related material.
The licensee notified the NRC Resident Inspector. |
| General Information or Other | Event Number: 42678 | Rep Org: THERMO ELECTRON CORPORATION
Licensee: THERMO ELECTRON CORPORATION
Region: 4
City: Sugar Land State: TX
County:
License #:
Agreement: Y
Docket:
NRC Notified By: BARRY NICHOLSON
HQ OPS Officer: JOHN KNOKE | Notification Date: 06/30/2006
Notification Time: 22:13 [ET]
Event Date: 06/30/2006
Event Time: [CDT]
Last Update Date: 06/30/2006 | Emergency Class: NON EMERGENCY
10 CFR Section:
21.21 - UNSPECIFIED PARAGRAPH
| Person (Organization):
MICHAEL SHANNON (R4)
THOMAS DECKER (R2)
JOHN MADERA (R3)
PART 21 EMAIL (NRR)
| Event Text PART 21 - DEFECTIVE PAPERLESS DATA ACQUISATION RECORDERS
"On May 3, 2006, Progress Energy (CP&T) Brunswick Nuclear Plant notified Thermo Electron that, during preventive maintenance on a Safety Related CSVC recorder, the rechargeable backup battery was found to have separated. Progress Energy also indicated that a similar failure had been found on a Non-Safety Related recorder prior to this incident. Neither recorder had failed while in service. Both are believed to have been in service approximately six (6) years with the original hack-up batteries installed.
"Progress Energy requested that Thermo Electron evaluate the failures to determine a recommended course of action and determine whether or not a Part 21 Notification was appropriate. Thermo Electron organized an engineering team to immediately begin an investigation of the failures. Thermo engineers solicited the services of the battery manufacturer in their investigation.
"The back-up battery installed on the CSVC is a NiCad sealed package. As the batteries age, the potential exists whereby if the batteries were fully discharged, which could occur during an outage or a shut down condition, a reverse charging situation may occur when the recorders are brought back on-line. If a reverse charging condition occur, the byproduct is hydrogen gas which could be sufficient to cause the battery to split open. This is not a normal condition, but could occur in batteries that have been in service for a number of years. There is no apparent hazardous condition within the recorder as a result of this failure and the recorder would continue to function properly while under power.
"There have been no failures reported in any nuclear safety-related application of CSVC recorders while in operating mode. Should the battery fail, it will not cause the recorder to present or record any erroneous data. However, should a recorder lose power after such a battery failure has occurred, the system configuration would be lost.
"Corrective Action : Thermo Electron has determined that a recommended replacement interval for the rechargeable battery should be documented to prevent the potential for this condition to occur in the future. The battery packs as originally installed on the CSVC are no longer available. A replacement has been identified which has been tested and verified to be suitable for the back-up battery application. This replacement does not compromise the original qualifications of the CSVC recorder. The battery pack (Thermo Part Number 1063-0200-001, Revision A) should be available for purchase within thirty (30) days. Thermo Electron recommends that utilities with CSVC recorders that have been in service greater than three (3) years with original batteries installed consider replacing the batteries at the next maintenance interval. The recommended replacement cycle for the replacement battery pack shall be three (3) years from the date the recorder is placed in service or the battery was replaced.
"Utilities have the option of purchasing batteries for replacement at their facility or returning the CSVC Recorders to Thermo Electron for battery replacement. Customers may contact Thermo Electron for spare parts or to request return for service." |
| Power Reactor | Event Number: 42680 | Facility: DUANE ARNOLD
Region: 3 State: IA
Unit: [1] [ ] [ ]
RX Type: [1] GE-4
NRC Notified By: TIM ERGER
HQ OPS Officer: BILL HUFFMAN | Notification Date: 07/01/2006
Notification Time: 01:39 [ET]
Event Date: 06/30/2006
Event Time: 23:37 [CDT]
Last Update Date: 07/01/2006 | Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(3)(v)(B) - POT RHR INOP
| Person (Organization):
JOHN MADERA (R3)
|
| Unit | SCRAM Code | RX CRIT | Initial PWR | Initial RX Mode | Current PWR | Current RX Mode | | 1 | N | Y | 96 | Power Operation | 96 | Power Operation | Event Text COMMON CAUSE DEGRADATION TO RHR SERVICE WATER PUMP MOTOR COOLING COILS
"At 2337 of 06/30/2006, it was determined that 3 out of 4 Residual Heat Removal Service Water (RHRSW) pumps were inoperable due to a common cause. Each of the 3 affected RHRSW pumps contained motor cooling coils that had experienced flow induced erosion which had significantly reduced the wall thickness of the cooling coil material. In some cases, the thinning was sufficient to cause through wall leakage of water into the upper motor bearing oil reservoir and subsequently displace oil out of the reservoir. Currently, 2 of the 3 affected RHRSW pumps have been removed from service for repairs. 1 of those pumps is expected to be returned to service shortly. The 3rd affected pump remains available for use, but has been conservatively declared inoperable based on Engineering evaluation that indicates a common cause failure mechanism likely exists, based on common fabrication methods used for all RHRSW pump motor cooling coils. The 4th (unaffected) RHRSW pump remains Operable because its cooling coil was replaced approximately 1 month ago. Under these conditions, sufficient confidence that the affected RHRSW pumps could perform their post LOCA mission does not exist. With only 1 of the 4 RHRSW pumps remaining Operable, the RHRSW System is not capable of performing its safety function. This event is reportable under 10CFR50.72(b)(3)(v)(B) and 50.73(a)(2)(v)(B)."
With 3 out of the 4 RHRSW pumps considered inoperable, the licensee is currently in an 8 hour Tech Spec LCO action statement. However, with expected the return of one of these pumps to operable status within the next hour, the LCO will convert to a 7 day action statement.
The licensee notified the NRC Resident Inspector. |
| Power Reactor | Event Number: 42681 | Facility: PALO VERDE
Region: 4 State: AZ
Unit: [ ] [ ] [3]
RX Type: [1] CE,[2] CE,[3] CE
NRC Notified By: JAMES BLAZEK
HQ OPS Officer: BILL HUFFMAN | Notification Date: 07/02/2006
Notification Time: 01:23 [ET]
Event Date: 07/01/2006
Event Time: 19:28 [MST]
Last Update Date: 07/02/2006 | Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(2)(iv)(B) - RPS ACTUATION - CRITICAL
| Person (Organization):
MICHAEL SHANNON (R4)
|
| Unit | SCRAM Code | RX CRIT | Initial PWR | Initial RX Mode | Current PWR | Current RX Mode | | 3 | M/R | Y | 100 | Power Operation | 0 | Hot Standby | Event Text MANUAL REACTOR TRIP FOLLOWING LARGE LEAK IN CONDENSATE SYSTEM
"On July 1, 2006, at approximately 19:28 MST Palo Verde Unit 3 experienced a manual reactor trip from approximately 100% rated thermal power due to a large leak in the condensate system. The sight glass for the 'A' condensate demin vessel blew out of the vessel creating an approximate 6 inch diameter hole. Several thousand gallons of condensate were discharged to the Turbine Building and the environment. Radiation Protection has sampled the condensate water and have verified no activity is present. Unit 3 was at normal operating temperature and pressure prior to the trip. The 'A' Main feedwater pump tripped on low suction pressure and a Reactor Power Cutback Signal was received. Condenser hotwell levels continued to decrease. When the main control board annunciator for the 'B' Main Feedwater Pump Trip Circuit Energized was received the reactor was manually tripped. All CEAs fully inserted into the reactor core. The event was diagnosed as an uncomplicated reactor trip. No ESF actuations occurred and none were required. Safety related buses remained energized during and following the reactor trip. The offsite power grid is stable. No significant LCOs have been entered as a result of this event. There was no loss of normal heat removal capabilities, or loss of any safety functions associated with this event. No major equipment was inoperable prior to the event that contributed to the event. No primary or secondary safety valves lifted during the event. The event did not result in any challenges to fission product barriers and there were no adverse safety consequences as a result of this event. The event did not adversely affect the safe operation of the plant or the health and safety of the public.
"Unit 3 is stable at normal operating temperature and pressure in Mode 3. There is no estimated time and date for the Unit 3 restart at the time this call is being made."
AFW is being used to feed the steam generators and is discharging steam to the main condenser. The licensee stated that the leak has been secured and that there is no significant standing water in the turbine building lower levels. All the water has drained into sumps or out doors. There does not appear to be any electrical or personnel safety hazards as a result of the water leak.
The licensee notified the NRC Resident Inspector. | |