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U.S. Nuclear Regulatory Commission
Operations Center
Event Reports For
08/21/2006 - 08/22/2006
** EVENT NUMBERS **
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| Power Reactor |
Event Number: 42135 |
Facility: HATCH
Region: 2 State: GA
Unit: [1] [2] [ ]
RX Type: [1] GE-4,[2] GE-4
NRC Notified By: PAUL UNDERWOOD
HQ OPS Officer: ARLON COSTA |
Notification Date: 11/10/2005
Notification Time: 15:28 [ET]
Event Date: 10/27/2005
Event Time: 17:00 [EST]
Last Update Date: 08/21/2006 |
Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(2)(xi) - OFFSITE NOTIFICATION
20.2201(a)(1)(ii) - LOST/STOLEN LNM>10X
74.11(a) - LOST/STOLEN SNM |
Person (Organization):
MIKE ERNSTES (R2)
WILLIAM BECKNER (NRR) |
| 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 |
Event Text
MATERIAL ACCOUNTABILITY DISCREPANCY AT PLANT HATCH
"This is a non-emergency Event Notification made in accordance with 10 CFR 20.2201(a)(1)(ii) to inform the NRC of a nuclear material accountability discrepancy amounting to, in the aggregate, a portion of a spent fuel rod used at Plant Hatch (HNP). In the process of reviewing records and physically verifying the contents of the spent fuel pool (SFP) as a part of activities associated with SNC's [Southern Nuclear Company] response to Bulletin 2005-01, SNC has identified discrepancies between fuel segments located in the SFP and segments indicated in plant records.
"The segments originated in the early 1980s during fuel reconstitution and inspection activities. The discrepancies call into question the location of segments of single spent fuel rods in each of three bundles. Characterization of three segments located in the SFP provided rod serial numbers which, in turn, were used to determine the bundles from which these segments originated. These bundles were then inspected, and the length of fuel in the location corresponding to each rod segment's intended location was determined. The aggregate in-bundle length found in these rod locations was combined with the lengths of the segments located in the SFP and compared to the design active fuel length of the three rods. This comparison results in a material discrepancy of approximately 55 inches, based on the length measurements. In addition to this discrepancy, historical records indicate two segments (totaling 13 inches of fuel length), which may not be within the inventory of segments identified to date in the SFP. When this amount is added to the length associated with the three rod locations, a discrepancy of approximately 68 inches results. When the planned supplemental inspection of select bundles and SFP locations is completed and photographs are evaluated to aid in the determination of special nuclear material present, this 68 inch estimate may increase or decrease.
"On June 16, 2005 SNC formed a team to identify and characterize material in the SFPs at Hatch in order to account for special nuclear material (SNM) at the level of detail requested by Bulletin 2005-01. A work scope was established and specialized resources were contracted to support the work activities. During the performance of these work activities, a number of items of interest were characterized as being either SNM or non-SNM items. On October 28, 2005, SNC submitted the interim status report to NRC.
"On November 4, 2005 the Hatch Plant Review Board (PRB) reviewed the SNM Issue Resolution Team's assessment of data produced by the records searches and physical cataloging of SNM in the SFPs. Based on that review, the PRB concurred that a discrepancy exists in material accounting for a portion of a spent fuel rod in each of three bundles and records, as noted above, which in the aggregate approximates 68 inches of fuel rod length. This length is equivalent to about 45% of the length of one intact fuel rod. Further physical examination of the SFP will include additional examination of SFP floor areas that have not been examined to date and selected fuel bundle inspections. The SFP floor areas are limited to a small number of locations that are under equipment or objects stored on the SFP floor. This expanded work scope is expected to be completed by December 15, 2005.
"Based on the nature of the fuel rod segments and radiation monitoring, a high degree of confidence exists that the segments are in a restricted area of the plant or otherwise under the control of a licensed facility such that the public health and safety has not been adversely affected. In addition, there is no evidence of theft or diversion.
"This notification satisfies the 30-day notification requirement of 10 CFR 20.2201(a)(1)(ii). A subsequent written report will be made in accordance with 10 CFR 20.2201(b).
"The licensee has informed the NRC Resident Inspector regarding the discrepancies.
"SNC will be making a press release describing the current status of this issue. Accordingly, this notification also satisfies the 4-hour notification requirement of 10 CFR 50.72(b)(2)(xi) with respect to issuance of the press release associated with this issue."
* * * UPDATE AT 15:48 ON 8/21/2006 FROM FRANK GORLEY TO ABRAMOVITZ * * *
"This is an update of non-emergency Event Notification 42135 that was previously made on November 10, 2005, in accordance with 10CFR 20.2201. This non-emergency Event Notification is made in accordance with 10 CFR 74.11 and informs the NRC of the loss of special nuclear material (SNM) from the historic breakage of several fuel rods used at Plant Hatch (HNP) in the early 1980s amounting to, in the aggregate, approximately 18 inches of a spent fuel rod. This amount is based on the available information, potentially affected by incomplete historic documentation. While reviewing records and physically verifying the contents of the spent fuel pool (SFP) in 2005 and 2006 associated with its response to Bulletin 2005-01, SNC identified discrepancies between fuel rod segments located in the SFP and segments indicated in plant records. This discrepancy was the subject of Event Notification 42135 and LER 2005-003 transmitted by letter NL-05-2262 dated 12/09/2005, including Rev. 1 of the LER, dated 04/14/2006 (transmitted by letter NL-06-0689).
"Additional locations in and around the fuel racks and additional fuel bundles were inspected between November 11, 2005, and July 21, 2006. Fuel vendor disorientation and plant SNM offsite shipping records reviews and quantity reconciliations were also recently completed. Based an the results of this expanded work scope, SNC concluded that some SNM material has been lost. This material either resides in some unidentified location in the SFP, resides in the bottom of the SFP as particles or small pieces, or was inadvertently shipped to a licensed low level waste processing facility.
"One SNM fragment, referred to as Item 30, was dropped during the physical activities in the pools and has not yet been recovered. This 4-1/2-inch fuel segment had been characterized and quantified and was dropped in the SFP during handling. A search was performed to look for item 30, but the intact segment was not found. During this search, a cladding segment with no appreciable SNM inside was located and identified as Item 32. It may be a portion of Item 30. Item 30's fuel length of 4-1/2 inches is included in the total characterized as lost.
"Based on the nature of the fuel rod segments, fragments, pellets, pellet chips, and small particles, and the barrier provided by in-plant radiation monitoring Instrumentation, a high degree of confidence exists that the lost SNM is either still in the SFP or was inadvertently shipped offsite to a licensed low level waste processing facility. Throughout its investigation and review, SNC has identified no evidence to indicate the possibility of theft or diversion of the missing quantity of SNM material.
"This notification satisfies the one-hour notification requirement of 10 CFR 74.11 (b). A subsequent written report will be made in accordance with 10CFR74.11(c). The licensee has informed the NRC Resident Inspector regarding the discrepancies and the conclusion regarding the lost material. SNC will be making a press release describing the current status of this issue. Accordingly, this notification also satisfies the 4-hour notification requirement of 10 CFR 50.72(b)(2)(xi) with respect to issuance of the press release associated with this issue."
Notified the R2DO (Lesser), PAO (Brenner), and NRR EO (Jung). |
| General Information or Other |
Event Number: 42573 |
Rep Org: GENERAL ELECTRIC COMPANY
Licensee: GENERAL ELECTRIC COMPANY
Region: 1
City: WILMINGTON State: NC
County:
License #:
Agreement: Y
Docket:
NRC Notified By: JASON POST
HQ OPS Officer: MIKE RIPLEY |
Notification Date: 05/12/2006
Notification Time: 22:36 [ET]
Event Date: 04/24/2006
Event Time: [EDT]
Last Update Date: 08/21/2006 |
Emergency Class: NON EMERGENCY
10 CFR Section:
21.21 - UNSPECIFIED PARAGRAPH |
Person (Organization):
ANTHONY DIMITRIADIS (R1)
JAMES MOORMAN (R2)
RICHARD SKOKOWSKI (R3)
OMID TABATABAI-EMAIL (NRR)
JACK FOSTER (EMAIL) (NRR) |
Event Text
PART 21 NOTIFICATION - BWR CORE SHROUD TIE ROD UPPER SUPPORT CRACKING
"Summary:
GE Energy, Nuclear (GE) has provided core shroud repairs using tie rods to the US BWR plants identified in Attachment 1 [of the Part 21 notification]. Recently it was discovered during an in-vessel visual inspection (IVVI) that tie rod upper supports at Hatch Unit 1 experienced cracking. The apparent root cause is Intergranular Stress Corrosion Cracking (IGSCC) in the Alloy X-750 tie rod upper support material. Alloy X-750 material is susceptible to IGSCC if subjected to sustained, large peak stress conditions. GE opened an internal evaluation to determine if the potential IGSCC in the X-750 tie rod structural components of other BWR shroud repairs designed by GE could be a reportable condition under 10CFR21.
"GE used the criterion provided in the BWR Vessels & Internals Project (BWRVIP-84) for the IGSCC susceptibility assessment of the X-750 components in the tie rod vertical load path. GE has concluded that it is not a reportable condition for the plants that were found to be within or not significantly exceed the BWRVIP-84 criterion. These US plants are identified as 'NR' in Attachment 2 [of the Part 21 notification]. GE determined that two US plants exceed the BWRVIP-84 criterion for the upper supports (in addition to the Hatch Unit 1 as-found condition). GE has not completed the evaluation for these plants to assess if a substantial safety hazard (SSH) exists. These plants have been provided a 60-Day Interim Report Notification under §21.21(a)(2) and are identified as '60-Day' in Attachment 2 [of the Part 21 notification].
"Safety Basis:
Cracking in the tie rod components made of X-750 may render the tie rod ineffective in maintaining core shroud configuration integrity during postulated accident conditions. Loss of core shroud integrity could impact the ability to maintain adequate core cooling for postulated design basis accident conditions. This condition would be reportable under 10 CFR 21 as a substantial safety hazard.
"Corrective Action:
The corrective action which has been, is being, or will be taken; the name of the individual or organization responsible for the action; and the length of time that has been or will be taken to complete the action (note, these are actions specifically associated with the identified deviation or failure to comply):
1. A preliminary cause evaluation has been performed. The apparent cause of the cracking is Intergranular Stress Corrosion Cracking (IGSCC). A material sample is being shipped to the GE Vallecitos Nuclear Center for examination to confirm the apparent cause. GE will report the results of the examination by August 21, 2006.
2. The issue has been communicated to the industry through the BWR Owners' Group and the Electric Power Research Institute (EPRI)/BWR Vessel and Internals Project (BWRVIP). The NRC was informed in a NRC management meeting with EPRI and the BWRVIP Executive Oversight Committee at the NRC offices, Rockville, on March 15, 2006.
3. GE has completed an evaluation of the susceptibility to IGSCC using the BWRVIP-84 criterion. Determination of whether any possible cracking could lead to a substantial safety hazard (i.e., loss of core shroud configuration integrity during a design basis accident condition) depends upon many factors, including the actual extent of cracking in the repair components. Until inspections are completed, the actual extent of cracking is not known. GE is developing a model to predict the postulated extent of tie rod upper support cracking for tie rods with upper supports made of Alloy X-750. For upper supports that exceed the BWRVIP-84 criteria significantly, the model will be used to postulate the extent of cracking. This prediction will be used to determine if a substantial safety hazard could exist. GE will report the results of the evaluation by October 9, 2006.
4. The original design basis stress reports will be reviewed to assess the available margin in the primary membrane + bending stress intensities of the upper supports with respect to ASME code allowable values. Where reasonable margin exists in the original design basis code evaluation (an existing margin of approximately 25 % will be considered as reasonable margin), the existing margin is deemed adequate to offset any engineering assumptions or judgments used in the original analysis. Where the original margin is less than 25%, further review will be performed (including finite element analysis, if necessary) to confirm that the upper support remains qualified. This review will be completed by October 9, 2006."
Affected US Plants per Attachments 1 and 2 of the Part 21 notification: Clinton, Nine Mile Point 1, Pilgrim, Dresden 2 & 3, Quad Cities 1 & 2, Hatch 1 & 2.
***** UPDATE ON 8/21/06 AT 1614 ET VIA E-MAIL FROM JASON POST TO MACKINNON ****
"GE Energy, Nuclear (GE) has completed the failure evaluation of the cracking discovered in the Hatch Unit 1 core shroud repair tie rod upper supports as committed in Reference 2, (GE Part 21 60-Day Interim Report Notification: Core Shroud Repair Tie Rod Upper Support Cracking, MFN 06-133, May 12, 2006). A preliminary cause evaluation had concluded that the apparent cause of the cracking is Intergranular Stress Corrosion Cracking (IGSCC). A material sample was shipped to the GE Vallecitos Nuclear Center for examination to confirm the apparent cause. GE committed to report the results of the examination by August 21, 2006.
"The fracture was examined by metallographic and scanning electron microscope (SEM) techniques including an analysis of the fracture surface. The examinations revealed the cracking mechanism to be IGSCC. Scanning electron microscopy (SEM) showed the fracture surface had a "rock candy" like appearance, consistent with an IGSCC mechanism. Images of the cross-section of the fracture surface further verified the IGSCC mechanism by showing the path of the crack following the grain boundaries. No hardness or microstrutural anomalies were observed.
"GE continues to work on the other action items that were committed in Reference 2. If you have any questions, on this information, please call me . . . "
R2DO (Mark Lesser) notified. E-mailed to NRR Part 21 (Omid Tabastabai & Jack Foster). |
| General Information or Other |
Event Number: 42780 |
Rep Org: MA RADIATION CONTROL PROGRAM
Licensee: NITRON CORPORATION
Region: 1
City: BILLERICA State: MA
County:
License #: 55-0238
Agreement: Y
Docket:
NRC Notified By: JOHN SUMARES
HQ OPS Officer: BILL HUFFMAN |
Notification Date: 08/16/2006
Notification Time: 09:58 [ET]
Event Date: 10/13/2004
Event Time: [EDT]
Last Update Date: 08/16/2006 |
Emergency Class: NON EMERGENCY
10 CFR Section:
AGREEMENT STATE |
Person (Organization):
RAYMOND LORSON (R1)
CINDY FLANNERY (NMSS) |
Event Text
MASSACHUSETTES AGREEMENT STATE REPORT OF LEAKING FE-55
The State provided the following information regarding a previously unreported event via facsimile:
"On October 15, 2004, Niton Corporation provided a report of a leaking sealed source that was discovered by Niton on October 13, 2004, during routine maintenance of a customer's model Xli 868p XRF device, S/N 5043. The sealed source is AEA Technology model IEC.A1, S/N 8765LG containing 10.3 mCi of Fe-55. The estimated leakage was 700,000 dpm or 0.3 microcuries.
"The source and a few contaminated device components were removed from the device and placed into a waste storage container. The area where the source removal was performed was surveyed and no contamination was found. The source manufacturer, AEA Technology, was notified of the leaking source. The source manufacturer, AEA Technology, has re-designed the source."
The cause and corrective action were determined and this event is considered closed by the Agency. |
| 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 |
Notification Date: 08/21/2006
Notification Time: 09:47 [ET]
Event Date: 08/19/2006
Event Time: 23:00 [EDT]
Last Update Date: 08/21/2006 |
Emergency Class: NON EMERGENCY
10 CFR Section:
20.2201(a)(1)(i) - LOST/STOLEN LNM>1000X |
Person (Organization):
RICHARD CONTE (R1)
MICHELE BURGESS (NMSS)
AARON DANIS (TAS) |
| This material event contains a "Less than Cat 3" level of radioactive material. |
Event Text
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.
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. |
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