|
U.S. Nuclear Regulatory Commission
Operations Center
Event Reports For
10/06/2006 - 10/10/2006
** EVENT NUMBERS **
|
| 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: 10/10/2006 |
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).
* * * UPDATE ON 10/09/06 AT 2333 EDT VIA E-MAIL FROM JASON POST TO HUFFMAN * * *
Summary of Part 21 Notification: Completion of GE Evaluation on Core Shroud Repair Tie Rod Upper Support Cracking
Previous correspondence on this subject identified that GE had concluded that this is not a reportable condition for Hatch Unit 1 and for several other US plants that have core shroud repairs designed by GE. However, GE had not completed the evaluation for two other US plants (Pilgrim and NMP-1). GE committed to complete the evaluation and inform the NRC of the results by October 9, 2006. The purpose of this letter is to inform the NRC of the results of the Pilgrim and NMP-1 evaluations.
"The GE evaluation concluded the following:
"a. All other major Alloy X-750 components in the tie rod assembly vertical load path besides the previously identified upper support brackets for the plants with GE designed tie rod repairs are within the BWRVIP-84 criterion for IGSCC susceptibility. Furthermore, the upper supports remain qualified with respect to ASME code allowable values per the original design basis stress analyses.
"b. It is not a reportable condition for the as found Hatch Unit 1 condition based on the known extent of tie rod upper support cracking.
"c. It is not a reportable condition for plants that do not use X-750 material for the tie rod upper support brackets. This is applicable to Clinton.
"d. It is not a reportable condition for plants that have margin to the BWRVIP-84 IGSCC criterion for the tie rod upper support brackets (or which exceed the criteria by a very small amount). These plants are Hatch 2, Quad Cities 1/2, and Dresden 2/3.
"e. It is not a reportable condition for the plants that exceed the BWRVIP-84 IGSCC criterion for the tie rod upper supports based on a three-pronged composite approach of (1) the shroud repair upper support integrity based on assumed cracking, (2) the shroud horizontal welds integrity, and (3) the previous shroud cracking safety evaluations. These plants are NMP-1 and Pilgrim.
"Corrective/Preventive Actions:
"GE recommends that NMP-1 and Pilgrim replace the tie rod upper supports with an improved IGSCC-resistant design at the next regular scheduled outage.
R1DO (Decker) notified. E-mailed to NRR Part 21 (Vern Hodge, Ian Jung, John Thorp). |
| General Information or Other |
Event Number: 42874 |
Rep Org: TEXAS DEPARTMENT OF HEALTH
Licensee: GOOLSBY TESTING
Region: 4
City: HUMBLE State: TX
County:
License #: 03115
Agreement: Y
Docket:
NRC Notified By: ART TUCKER
HQ OPS Officer: BILL GOTT |
Notification Date: 10/04/2006
Notification Time: 11:01 [ET]
Event Date: 09/27/2006
Event Time: [CDT]
Last Update Date: 10/04/2006 |
Emergency Class: NON EMERGENCY
10 CFR Section:
AGREEMENT STATE |
Person (Organization):
GREG PICK (R4)
MICHELE BURGESS (NMSS) |
Event Text
AGREEMENT STATE REPORT - PERSONNEL OVEREXPOSURE
"A technician was performing radiography operations at an off site facility. The technician had completed an exposure and had returned the source to the camera. The technician noted the survey meter setting about six [feet] in front of the camera was going towards zero. Was setting up for the next exposure. The technician disconnected the guide tube from the camera and relocated it to the new location. Upon returning to pick up the camera, the technician noticed the source hanging out of the camera. The technician cranked the source into the camera. The technician notified his supervisor of the event. The supervisor instructed the technician to cease all work and to bring his film badge, alarming dosimeter, and dosimeter to the office. The film badge was overnighted to the licensee's processor. On 10/3/06, the processor reported the exposure for the film badge to be 16543 mrem DDE. The workers total for the year is reported as 19538 mrem DDE. A safety meeting was held for all radiographers at the licensee facility on October 3, 2006 to discuss proper performance of surveys when conducting radiography operations. It appears at this time that no other persons were exposed to the high dose rates."
Texas Incident #: I-8365 |
| General Information or Other |
Event Number: 42876 |
Rep Org: WA DIVISION OF RADIATION PROTECTION
Licensee: ISORAY
Region: 4
City: RICHLAND State: WA
County:
License #: WN-L0213-1
Agreement: Y
Docket:
NRC Notified By: ARDEN C. SCROGGS
HQ OPS Officer: JOHN MacKINNON |
Notification Date: 10/04/2006
Notification Time: 21:02 [ET]
Event Date: 10/04/2006
Event Time: 06:30 [PDT]
Last Update Date: 10/04/2006 |
Emergency Class: NON EMERGENCY
10 CFR Section:
AGREEMENT STATE |
Person (Organization):
GREG PICK (R4)
LAWRENCE KOKAJKO (NMSS) |
| This material event contains a "Less than Cat 3" level of radioactive material. |
Event Text
WASHINGTON AGREEMENT STATE REPORT - TWO DAMAGED SHIPPING PACKAGES CONTAINING CESIUM-131 CANCER THERAPY SEEDS.
Event sent to NRC Headquarters Operation Center via e-mail.
"ABSTRACT: (where, when, how, why; cause, contributing factors, corrective actions, consequences, Dept. of Health (DOH) on-site investigation; media attention):
"At 6:30 am, 4 October, the Spokane FedEx Terminal Manager discovered a flattened lead cap. A partial label on the cap indicated it came from one of two packages containing IsoRay, Cesium 131, therapy seeds. A second package was found crushed but essentially intact; the seeds in this package were all present and apparently undamaged. Scraps from the first box were found on the runway and other pieces on the floor of a tug (airport vehicle used to move cargo). It is thought the damage was caused when the boxes were caught between moving pieces of the cargo loading equipment. The actual incident appears to have happened 12 hours earlier (about 6 pm, 3 October). Dayshift FedEx staff had apparently placed the damaged packages on the floor on the passenger side of the tug cab.
"Four WA Department of Health, Health Physicists responded to the scene at about 8:00 am, 4 October. IsoRay also dispatched a team of three HP technicians arriving at noon with equipment including a decontamination kit. WA Department of Health staff were able to retrieved three of the sixty-three seeds that were in the one shredded package. Several areas of contamination were also found.
"Measurement on the floor of the tug's passenger side was reading 150 mR/hour with an Eberline RO2 ion chamber. Radiation measurements on the crushed pig lid were about 25 mR/hour with an RO2 and a contamination measurement of about 400 cpm with a GM instrument. Contamination reading on the crushed box was about 300 cpm with the GM instrument. A spot on the tarmac was found reading about 12 mR/hour with an RO2.
"The undamaged stainless steel pig reads about 5 mR/hour with an RO2.
"Night shift personnel were asked to return to the facility. No personnel contamination had been found at the writing of this report.
"WA Dept of Health and IsoRay staff continue to look for the remainder of the packaging and seeds.
"No news media attention yet.
"Notification Reporting Criteria: 10 CFR 30.50(b)(1)
"Isotope and Activity involved: 330 mCi, 12.2 Gigabecqerals of Cesium 131.
"Overexposures? (number of workers/members of the public; dose estimate; body part receiving dose; consequence): Exposures to be determined.
"Lost, Stolen or Damaged? (mfg., model, serial number): 3 of 63 missing seeds recovered, contamination found in tug, packaging and on the runway. The remainder of the activity is still being sought.
"Sealed Source and Device Registry: WA-1220-S-101-S
"Disposition/recovery: WA Dept of Health and the manufacturer staff are still on the scene to assist in recovery of seeds.
"Leak test? The seeds are leak tested prior to packaging and were within limits.
"Vehicle: (description; placards; Shipper; package type; Pkg. ID number) Airport tug, N/N FedEx air bill # 730 235 322 954.
"Release of activity? Three seeds found in the tug. Contamination found in the tug, on packaging and on the runway.
"Activity and pharmaceutical compound intended: N/A
"Misadministered activity and/or compound received: N/A
"Device (HDR, etc.) Mfg., Model; computer program: N/A
"Exposure (intended/actual); consequences: N/A
"Was patient or responsible relative notified? N/A
"Was written report provided? Not yet
"Was referring physician notified? N/A
"Consultant used? No."
Washington Incident Report - WA-06-053
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.
This source is not amongst those sources or devices identified by the IAEA Code of Conduct for the Safety & Security of Radioactive Sources to be of concern from a radiological standpoint. Therefore is it being categorized as a less than Category 3 source |
| Power Reactor |
Event Number: 42882 |
Facility: MILLSTONE
Region: 1 State: CT
Unit: [ ] [2] [ ]
RX Type: [1] GE-3,[2] CE,[3] W-4-LP
NRC Notified By: JANET NOVAK
HQ OPS Officer: MARK ABRAMOVITZ |
Notification Date: 10/07/2006
Notification Time: 03:49 [ET]
Event Date: 10/07/2006
Event Time: 01:51 [EDT]
Last Update Date: 10/07/2006 |
Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(3)(v)(D) - ACCIDENT MITIGATION |
Person (Organization):
PAUL KROHN (R1) |
| Unit |
SCRAM Code |
RX CRIT |
Initial PWR |
Initial RX Mode |
Current PWR |
Current RX Mode |
| 2 |
N |
N |
0 |
Hot Standby |
0 |
Hot Standby |
Event Text
SCAFFOLD PREVENTS MAIN STEAM ISOLATION VALVE CLOSURE
Temporary scaffolding was constructed near the loop-1 Main Steam Isolation Valve (MSIV) on 8/30/2006 while the Unit was in Mode 1. On 10/7/2006 surveillance testing on the MSIVs was conducted in Mode 3 and the loop-1 MSIV failed to fully stroke closed. It was determined that temporary scaffolding was interfering with full valve travel. The scaffolding was removed and the surveillance test verified full valve motion.
The licensee notified the NRC Resident Inspector and will notify state and local officials. |
| Power Reactor |
Event Number: 42883 |
Facility: LIMERICK
Region: 1 State: PA
Unit: [ ] [2] [ ]
RX Type: [1] GE-4,[2] GE-4
NRC Notified By: J.E. THOMPSON
HQ OPS Officer: STEVE SANDIN |
Notification Date: 10/07/2006
Notification Time: 11:05 [ET]
Event Date: 10/07/2006
Event Time: 06:57 [EDT]
Last Update Date: 10/07/2006 |
Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(3)(v)(D) - ACCIDENT MITIGATION |
Person (Organization):
PAUL KROHN (R1) |
| 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
HPCI DECLARED INOPERABLE
"A division 2 Primary Containment Isolation Instrument System component inoperable required the manual isolation of High Pressure Coolant Injection (HPCI) outboard Primary Containment Isolation valves (PCIV). This isolation was required by T/S action 3.3.2.b.1.
"This report made pursuant to 10CFR 50,72 (b)(3)(v)(D) for the inability of single train system to mitigate the consequences of an accident."
At approximately midnight on 10/7/06, Control Room Operators received a Div 2 Steam Leak Detection Trouble Alarm. Troubleshooting identified the failure as associated with the GE NUMAC drawer where temperature inputs are processed. The applicable 14-day T/S LCO Action Statement requires completion of repairs within 6 hours or isolation of HPCI 1 hour later. HPCI was isolated at 0657 EDT. There was no I&C work in progress at the time of the failure. Repair parts are available and onsite.
The licensee informed the NRC Resident Inspector. |
| Power Reactor |
Event Number: 42884 |
Facility: KEWAUNEE
Region: 3 State: WI
Unit: [1] [ ] [ ]
RX Type: [1] W-2-LP
NRC Notified By: DAVID KARST
HQ OPS Officer: JOHN KNOKE |
Notification Date: 10/07/2006
Notification Time: 14:16 [ET]
Event Date: 10/07/2006
Event Time: 07:53 [CDT]
Last Update Date: 10/07/2006 |
Emergency Class: NON EMERGENCY
10 CFR Section:
26.73 - FITNESS FOR DUTY |
Person (Organization):
PATTY PELKE (R3) |
| Unit |
SCRAM Code |
RX CRIT |
Initial PWR |
Initial RX Mode |
Current PWR |
Current RX Mode |
| 1 |
N |
N |
0 |
Refueling |
0 |
Refueling |
Event Text
FITNESS FOR DUTY - CONFIRMED POSITIVE FOR NON-LICENSED SUPERVISOR
A non-licensed supervisor had a confirmed positive for alcohol. The supervisor's access to the plant has been denied while a review of this matter is performed. Contact the Headquarters Operations Officer for additional details.
The NRC Resident Inspector has been notified. |
| Power Reactor |
Event Number: 42885 |
Facility: SURRY
Region: 2 State: VA
Unit: [1] [2] [ ]
RX Type: [1] W-3-LP,[2] W-3-LP
NRC Notified By: THOMAS FREDETTE
HQ OPS Officer: PETE SNYDER |
Notification Date: 10/07/2006
Notification Time: 14:31 [ET]
Event Date: 10/07/2006
Event Time: 13:40 [EDT]
Last Update Date: 10/07/2006 |
Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(3)(xiii) - LOSS COMM/ASMT/RESPONSE |
Person (Organization):
THOMAS DECKER (R2)
ANTHONY MCMURTRAY (EP) |
| 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
SITE ACCESS ROAD FLOODED
"At 1340 on 10/7/06; normal road access to Surry Power Station became flooded and impassable due to sustained heavy rainstorms. This flooding compromised the Surry off-site response capability.
"This event is reportable in accordance with 10 CFR 50.72(b)(3)(xiii), any event that results in a major loss of emergency assessment capability (e.g., offsite response capability)."
The licensee notified the NRC Resident Inspector. |
| Power Reactor |
Event Number: 42886 |
Facility: PALO VERDE
Region: 4 State: AZ
Unit: [ ] [2] [ ]
RX Type: [1] CE,[2] CE,[3] CE
NRC Notified By: JOHN GUNN
HQ OPS Officer: PETE SNYDER |
Notification Date: 10/07/2006
Notification Time: 16:33 [ET]
Event Date: 10/07/2006
Event Time: 11:40 [MST]
Last Update Date: 10/07/2006 |
Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(3)(ii)(A) - DEGRADED CONDITION |
Person (Organization):
GREG PICK (R4) |
| Unit |
SCRAM Code |
RX CRIT |
Initial PWR |
Initial RX Mode |
Current PWR |
Current RX Mode |
| 2 |
N |
N |
0 |
Refueling |
0 |
Refueling |
Event Text
AXIAL INDICATIONS IDENTIFIED ON PRESSURE BOUNDRY
"The following event description is based on information currently available. If through subsequent reviews of this event, additional information is identified that is pertinent to this event or alters the information being provided at this time, a follow-up notification will be made via the ENS or under the reporting requirements of 10CFR50.73.
"On October 7, 2006, at approximately 11:40 Mountain Standard Time (MST), Engineering personnel performing preplanned in-service examination of the reactor vessel head vent piping notified the Unit 2 Control Room that two axial indications had been discovered. The indications are located on the inner diameter surface of the pipe adjacent to the J-weld to the reactor head and are part of the Reactor Coolant System (RCS) pressure boundary. The indications are characterized as axial, estimated 0.020 to 0.030 inches deep, and approximately 0.2 inches long. The indications do not appear to be through wall and there is no evidence of RCS pressure boundary leakage.
"Technical Specifications (TS) Limiting Conditions for Operation (LCO) 3.4.14 permits no RCS pressure boundary leakage and there is no evidence of pressure boundary leakage. Nevertheless, the indications are being conservatively identified as abnormal degradation of the RCS pressure boundary and will necessitate taking corrective action to restore the barrier's capability. Therefore, the ENS notification of this event is in accordance with 10CFR50.72(b)(3)(ii)(A). Unit 2 is currently shutdown for its 13th refueling outage and is in Mode 6.
"No ESF actuations occurred and none were required. There were no structures, systems, or components that were inoperable at the time of discovery that contributed to this condition. There were no failures that rendered a train of a safety system inoperable and no failures of components with multiple functions were involved. The event did not result In the release of radioactivity to the environment and did not adversely affect the safe operation of the plant or health and safety of the public.
"An investigation of this event will be conducted in accordance with the PVNGS corrective action program."
The licensee notified the NRC Resident Inspector. |
| Power Reactor |
Event Number: 42887 |
Facility: PEACH BOTTOM
Region: 1 State: PA
Unit: [2] [ ] [ ]
RX Type: [2] GE-4,[3] GE-4
NRC Notified By: MIKE BERG
HQ OPS Officer: JOHN KNOKE |
Notification Date: 10/07/2006
Notification Time: 18:14 [ET]
Event Date: 10/07/2006
Event Time: 18:02 [EDT]
Last Update Date: 10/08/2006 |
Emergency Class: UNUSUAL EVENT
10 CFR Section:
50.72(a) (1) (i) - EMERGENCY DECLARED |
Person (Organization):
PAUL KROHN (R1DO)
PETER WILSON (IRD)
THEODORE QUAY (NRR)
MIKE WEBER (NRR)
MARC DAPAS (R1)
R. BOZZO (DHS)
G. KANUPP (FEMA) |
| Unit |
SCRAM Code |
RX CRIT |
Initial PWR |
Initial RX Mode |
Current PWR |
Current RX Mode |
| 2 |
N |
Y |
22 |
Power Operation |
22 |
Power Operation |
Event Text
UNUSUAL EVENT - LOSS OF CONTAINMENT INTEGRITY
During testing, a small crack (3"- 4" axial crack) was identified on a return line to the torus for both HPCI and RCIC (steam driven ECCS systems). The valve upstream of the leak has been closed, however, the leak is still unisolable between that valve and the torus. This plant condition does not threaten public safety.
The licensee notified the NRC Resident Inspector, as well as State and local agencies.
* * * UPDATE PROVIDED BY ROBERT STEIGERWALD TO JEFF ROTTON AT 0555 EDT ON 10/08/06 * * *
"This is a follow up notification to EN# 42887. Peach Bottom Atomic Power Station Terminated the Unusual Event at 0513 [EDT] on 10/08/06. Unit 2 is currently shutdown in Mode 4."
In Mode 4, Primary Containment is no longer required. The licensee notified the NRC Resident Inspector, State and Local agencies.
Notified R1DO (Krohn), R1 (Dapas), NRREO (Quay), NRR (Dyer, Weber), IRD (Blount), DHS (Gray), FEMA (Dunker). |
| Power Reactor |
Event Number: 42888 |
Facility: SURRY
Region: 2 State: VA
Unit: [ ] [2] [ ]
RX Type: [1] W-3-LP,[2] W-3-LP
NRC Notified By: ROBERT PORTLOCK
HQ OPS Officer: JOHN KNOKE |
Notification Date: 10/07/2006
Notification Time: 18:27 [ET]
Event Date: 10/07/2006
Event Time: 18:01 [EDT]
Last Update Date: 10/08/2006 |
Emergency Class: ALERT
10 CFR Section:
50.72(a) (1) (i) - EMERGENCY DECLARED |
Person (Organization):
THOMAS DECKER (R2DO)
THEODORE QUAY (NRR)
PETER WILSON (IRD)
MIKE WEBER (NRR)
V. MCREE (R2)
T. BARNES (DHS)
G. KANUPP (FEMA)
R. TURNER (DOE)
N. PYLES (HHS)
CREWS (EPA) |
| Unit |
SCRAM Code |
RX CRIT |
Initial PWR |
Initial RX Mode |
Current PWR |
Current RX Mode |
| 2 |
M/R |
Y |
100 |
Power Operation |
0 |
Hot Standby |
Event Text
MANUAL SCRAM DUE TO UNUSUAL NOISE IN TURBINE BUILDING
The main turbine cross-under safety relief valves lifted for no known reason and blew siding off the side of the Unit 2 Turbine building. This siding hit the feeder lines to the A & C Reserve Station Service Transformers (RSSTs). The operator manually scrammed the plant due to swings in steam generator level and unusual noise coming from the turbine building . Unit 2 shutdown currently de-energized A & C Reserve Station Transformers, which effects D & E transfer buses. This also effects 1J bus, which is de-energized, and 1H & 2J buses which are energized with #1 diesel and #3 diesel.
Decay heat removal is being performed thru the SG PORV's and auxiliary feedwater system, with forced cooling from the "B" RCP. Safety related systems are available if required.
Notified USDA (A. Jimenez) in addition to the other agencies already identified.
The licensee notified the NRC Resident Inspector, as well as State and local agencies.
* * * UPDATE ON 10/8/2006 AT 05:45 FROM MIKE CHRIS TO ABRAMOVITZ * * *
The site terminated the Alert at 05:40 due to having the "A" RSST in service with bus 1J being powered from its normal power supply. No damage was found from the displaced siding with the exception of the "C" RSST (which should be repaired around noon). The "C" RSST is currently tagged out for maintenance.
The licensee notified the NRC Resident Inspector, state, and local governments.
Notified: R2DO (Decker), R4DO (Pick), NRR (Dyer, Weber, Quay), IRD (Blount, Wilson), R2 (McCree), DHS (Gray), FEMA (Dunker), DOE (Steve Bailey), EPA (Allison), USDA (Dean Giles), and HHS (Lt. Smith). |
| Power Reactor |
Event Number: 42889 |
Facility: PEACH BOTTOM
Region: 1 State: PA
Unit: [2] [ ] [ ]
RX Type: [2] GE-4,[3] GE-4
NRC Notified By: ROBERT STEIGERWALD
HQ OPS Officer: JEFF ROTTON |
Notification Date: 10/07/2006
Notification Time: 23:49 [ET]
Event Date: 10/07/2006
Event Time: 17:50 [EDT]
Last Update Date: 10/08/2006 |
Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(2)(i) - PLANT S/D REQD BY TS
50.72(b)(2)(iv)(B) - RPS ACTUATION - CRITICAL
50.72(b)(3)(ii)(A) - DEGRADED CONDITION
50.72(b)(3)(v)(C) - POT UNCNTRL RAD REL |
Person (Organization):
PAUL KROHN (R1) |
| Unit |
SCRAM Code |
RX CRIT |
Initial PWR |
Initial RX Mode |
Current PWR |
Current RX Mode |
| 2 |
M/R |
Y |
22 |
Power Operation |
0 |
Hot Shutdown |
Event Text
TECHNICAL SPECIFICATION REQUIRED SHUTDOWN DUE TO LOSS OF PRIMARY CONTAINMENT INTEGRITY
"At 17:50, on 10/07/2006, the Peach Bottom Atomic Power Station identified a crack approximately 4 inches long on a Unit 2, Reactor Core Isolation Cooling (RCIC) test line, as the line penetrates the Suppression Pool of Primary Containment. The degraded piping represents a loss of primary containment integrity, placing one of the principle safety barriers in a 'Seriously Degraded' condition as defined by 10CFR50.72 (b)(3)(ii)(A). This condition required a Reactor Shutdown per the plants Technical Specifications (TS) [TS 3.6.1.1]. Unit 2 was manually scrammed, at 20:16, in order to shutdown the reactor and place the unit in Mode 3 per the Technical Specifications. The TS required shutdown is reportable per 10CFR50.72(b)(2)(i). The unplanned reactor scram is reportable per 10CFR50.72(b)(2)(iv)(B). The reactor scram and resultant Emergency Safety Feature actuations were completed as required. In addition, the loss of primary containment integrity represents a condition that 'could have prevented the fulfillment of a safety function of a structure required to control the release of radioactive material' and/or 'mitigate the consequences of an accident.' This is reportable per 10CFR50.72(b)(3)(v)(C) and (D). Unit 2 is currently shutdown, Mode 3, with an RPV cooldown in progress, with plans to Enter Mode 4 by 02:00 on 10/08/06."
All control rods fully inserted on the Manual Reactor Scram. The reactor is currently being fed from the condensate system with decay heat being removed to the condenser via the MSL drains. The electric plant is in a normal shutdown lineup.
See EN # 42887 for related NOTICE OF UNUSUAL EVENT.
Additional 10 CFR Section not listed above:
50.72(b)(3)(v)(D) ACCIDENT MITIGATION
The licensee notified the NRC Resident Inspector. |
| Power Reactor |
Event Number: 42890 |
Facility: SURRY
Region: 2 State: VA
Unit: [1] [ ] [ ]
RX Type: [1] W-3-LP,[2] W-3-LP
NRC Notified By: DAVID HERRING
HQ OPS Officer: JEFF ROTTON |
Notification Date: 10/08/2006
Notification Time: 10:13 [ET]
Event Date: 10/07/2006
Event Time: 17:11 [EDT]
Last Update Date: 10/08/2006 |
Emergency Class: NON EMERGENCY
10 CFR Section:
OTHER UNSPEC REQMNT |
Person (Organization):
THOMAS DECKER (R2)
THEODORE QUAY (NRR)
PETER WILSON (IRD) |
| Unit |
SCRAM Code |
RX CRIT |
Initial PWR |
Initial RX Mode |
Current PWR |
Current RX Mode |
| 1 |
N |
Y |
100 |
Power Operation |
75 |
Power Operation |
Event Text
DISCOVERY OF AFTER-THE-FACT-EMERGENCY CONDITION - UNUSUAL EVENT
"At 09:20 [EDT] on 10/8/06, it was discovered that a NOUE had not been declared on 10/7/06 as required by Surry EPIP-1.01, EAL Tab H-3. At 17:11 [EDT] on 10/7/06 Surry Power Station experienced a loss of offsite power to unit 1 Transfer Buses (both D & F), therefore in accordance with tab H-3, a Notification of Unusual Event [NOUE] should have been declared. Subsequent to this event, an Alert was declared [on Unit 2] at 18:01 [EDT] due to tab K-11 and the Alert was exited at 05:40 [EDT] on 10/8/06. This notification is for the discovery of an undeclared event that exceeded an Emergency Action Level (EAL) as specified in EPIP-1.01."
The licensee notified the NRC Resident Inspector. See EN # 42888 for details regarding Alert declared on Unit 2. |
| Power Reactor |
Event Number: 42892 |
Facility: GINNA
Region: 1 State: NY
Unit: [1] [ ] [ ]
RX Type: [1] W-2-LP
NRC Notified By: KEVIN McLAUGHLIN
HQ OPS Officer: MARK ABRAMOVITZ |
Notification Date: 10/09/2006
Notification Time: 10:52 [ET]
Event Date: 10/09/2006
Event Time: 04:00 [EDT]
Last Update Date: 10/09/2006 |
Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(3)(v)(D) - ACCIDENT MITIGATION |
Person (Organization):
PAUL KROHN (R1) |
| Unit |
SCRAM Code |
RX CRIT |
Initial PWR |
Initial RX Mode |
Current PWR |
Current RX Mode |
| 1 |
N |
N |
0 |
Hot Shutdown |
0 |
Cold Shutdown |
Event Text
CONTAINMENT SUMP INOPERABLE WHEN REQUIRED BY TECHNICAL SPECIFICATIONS
"On October 9, 2006 at or about 0400 EDT it was discovered that the containment Sump B grating had been covered with temporary lead shielding. The shielding was placed as part of refueling outage activities in containment. Investigation has revealed that the lead shielding was put in place at about 0230 hours. All lead was removed from the grating. This was confirmed by the Outage Control Center at 0930. At the time of discovery the plant was in MODE 4 with RCS pressure of 315 psig, RCS Temperature of 337 °F.
"The plant entered MODE 5 at 08:15 hours on 10/9/06."
The B containment sump is the suction point for the A and B RHR pumps. When required, the RHR pumps also provide suction flow to the Safety Injection pumps. The containment sump is require to be operable in modes 1-4.
The licensee notified the NRC Resident Inspector. |
| Power Reactor |
Event Number: 42893 |
Facility: OCONEE
Region: 2 State: SC
Unit: [1] [2] [3]
RX Type: [1] B&W-L-LP,[2] B&W-L-LP,[3] B&W-L-LP
NRC Notified By: RANDALL TODD
HQ OPS Officer: MARK ABRAMOVITZ |
Notification Date: 10/09/2006
Notification Time: 11:35 [ET]
Event Date: 01/02/2003
Event Time: [EDT]
Last Update Date: 10/09/2006 |
Emergency Class: NON EMERGENCY
10 CFR Section:
21.21 - UNSPECIFIED PARAGRAPH |
Person (Organization):
THOMAS DECKER (R2)
VERN HODGE (email) (NRR) |
| Unit |
SCRAM Code |
RX CRIT |
Initial PWR |
Initial RX Mode |
Current PWR |
Current RX Mode |
| 1 |
N |
N |
0 |
Cold Shutdown |
0 |
Cold Shutdown |
| 2 |
N |
Y |
100 |
Power Operation |
100 |
Power Operation |
| 3 |
N |
Y |
100 |
Power Operation |
100 |
Power Operation |
Event Text
PART 21 NOTIFICATION - VOIDS IN PLATE STEEL
"Facility Reporting: Oconee Nuclear Station, Duke Power Company
"Basic component: 3/4 in thick, 32 sq ft (4 ft by 8 ft) A36 steel plate manufactured in accordance with ASTM/ASME Section II; manufactured by CORUS; supplied by CONSOLIDATED POWER SUPPLY Purchase Order ON 46703, Heat number A2WT; Duke QC receipt inspection on 9/25/2001.
"192 sq-ft of plate was ordered on this Purchase Order, of this, 128 square feet of plate had the applicable heat number
"Nature of the defect or failure to comply and the safety hazard:
"During fabrication of hangers for use during a modification, pieces were cut from a sheet of 3/4 inch steel plate for use as hanger support base plates. Personnel noted that a base plate contained a lamination, resulting in the hanger strut being welded to only a thin layer of material. The piece was examined by engineering and a Ultrasonic Test examination of the remainder of the original 4 foot by 8 foot sheet found that the lamination area was approximately 15 inches by 36 inches and approximately 1/32 to 3/32 inches deep. This area was cut out and discarded and the remainder of the sheet was considered acceptable for use.
"The issue was entered into the Duke corrective action program and an evaluation for Part 21 reportability was initiated.
"The Part 21 evaluation concluded that, if the deflective material had been used for hangers as initially intended, the hangers may have failed during a design basis transient or accident due to failure of the thin lamination layer.
"Duke hanger designs are typically conservative such that failure of a single hanger will not result in the failure of the supported system/component, but within the rules for evaluation of a Part 21 defect, it must be assumed that this defective material may have resulted in a significant safety hazard and therefore must be reported under Part 21. The material known to be defective was not actually installed; therefore, this event is not reportable under 10 CFR 50.72 or 50.73.
"At this time Duke has no information to determine if this defect could extend to other material of the same heat or same manufacturer. Duke has no information on other possible customers.
"Duke did use the remaining sheets from the same purchase order in various applications, and is creating corrective actions to inspect the material with the same heat number for additional lamination areas: However due to the fact that no similar deficiencies were observed during fabrication and installation of the components made from this material or in use of similar materials, Duke considers this an isolated instance and does not consider this to be a current operability issue.
"Additional details will be provided in the required 30 day report."
A sheet (4'x8') of this heat of plate steel was shipped to Catawba. Catawba has been notified and will investigate their use.
The licensee notified the NRC Resident Inspector. |
| Power Reactor |
Event Number: 42894 |
Facility: SURRY
Region: 2 State: VA
Unit: [1] [2] [ ]
RX Type: [1] W-3-LP,[2] W-3-LP
NRC Notified By: THOMAS FREDETTE
HQ OPS Officer: JASON KOZAL |
Notification Date: 10/09/2006
Notification Time: 14:17 [ET]
Event Date: 10/09/2006
Event Time: 11:24 [EDT]
Last Update Date: 10/09/2006 |
Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(2)(xi) - OFFSITE NOTIFICATION |
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 |
100 |
Power Operation |
| 2 |
N |
N |
0 |
Hot Shutdown |
0 |
Hot Shutdown |
Event Text
POTENTIAL VIOLATION OF VIRGINIA POLLUTANT DISCHARGE ELIMINATION SYSTEM PERMIT
"As a result of heavy rains, influent flows to the Surry Sewage Treatment Plant (STP) were in excess of 193,349 gallons on Saturday, October 7. Road conditions prevented normal pumping service from coming onsite to remove this excess water.
"Resultant analysis showed a loss of 54%, 39% and 10% solids in STP aeration tanks 1, 2 and 3, respectively. These 'solid' parameters are normally an indication that the STP is able to maintain effective treatment. At 1124 on 10/09/06, the Virginia Department of Environmental Quality (DEQ) was notified that the loss of solids may indicate that the treatment capability of the STP is impacted, which may cause a violation of one of the parameters of Surry's Virginia Pollutant Discharge Elimination System (VPDES) Permit.
"This event is being reported in accordance with 10CFR50.72(b)(2)(xi) due to the fact that the Virginia Department of Environmental Quality was notified at 1124 hours."
The licensee notified the NRC Resident Inspector. |
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