| U.S. Nuclear Regulatory Commission
Operations Center Event Reports For
03/29/2013 - 04/01/2013 ** EVENT NUMBERS ** |
 | Part 21 | Event Number: 47975 | Rep Org: ABB INC
Licensee: ABB INC
Region: 1
City: CORAL SPRINGS State: FL
County:
License #:
Agreement: Y
Docket:
NRC Notified By: DENNIS BATOVSKY
HQ OPS Officer: MARK ABRAMOVITZ | Notification Date: 05/29/2012
Notification Time: 17:07 [ET]
Event Date: 03/29/2012
Event Time: [EDT]
Last Update Date: 04/01/2013 | Emergency Class: NON EMERGENCY
10 CFR Section:
21.21(a)(2) - INTERIM EVAL OF DEVIATION
| Person (Organization):
HIRONORI PETERSON (R3DO)
PART 21 GROUP ()
| Event Text PART 21 NOTIFICATION - PROTECTIVE RELAYS MAY NOT BE QUALIFIED FOR HARSH ENVIRONMENTS
The following report was received via fax:
"During the commercial grade dedication process for a unit that was returned for repair, the unit was found to be in nonconformance with ABB specifications. The ABB specifications require that two (2) particular components, integrated circuits (ICs) of plastic construction, are replaced with 2 ICs of ceramic construction during the assembly process. The chips found on the harmonic filter circuit board (HF Board) of the relays were of plastic construction. While plastic ICs are approved for use in commercial relays, they have not been qualified for safety-related applications. Relays in this condition will function normally in mild environments, but have not been qualified for harsh environments, or for elevated radiation environments."
The affected solid state relays are 27N and 59G shipped between August 1, 2010 and October 15, 2010.
"Eighteen (18) Relays (Material number: 211T4175-HF-1E) were sold to ABB Florence who in turn shipped the same to Detroit Edison's Fermi 2 Power Plant."
* * * UPDATE FROM TAUZER TO SNYDER AT 1730 EDT ON 4/1/13 * * *
The supplier has concluded that the solid state relays with the plastic encapsulate are qualified for Class 1E applications.
"ABB Coral Springs is providing this letter to close the interim report and notice of deviation from specification requirements associated with Solid State Relays 27N and 59G dated May 29, 2012 (Event ML12153A030).
"As stated in the interim report, one of the actions to be completed was the qualification and testing of representative plastic ICs to confirm acceptability and use in safety-related applications.
"This letter is a follow-up notification that the plastic IC qualification and testing is complete, and based upon the results, we have concluded that the ICs with the plastic encapsulate are qualified for Class 1E applications specific to the subject relays.
"Affected customers have been notified."
Notified R3DO (Daley) and Part 21 Reactors (Email). |
| Part 21 | Event Number: 48660 | Rep Org: MPR ASSOCIATES, INC ENGINEERS
Licensee: BASLER ELECTRIC
Region: 1
City: ALEXANDRIA State: VA
County:
License #:
Agreement: Y
Docket:
NRC Notified By: PAUL DAMERELL
HQ OPS Officer: BILL HUFFMAN | Notification Date: 01/09/2013
Notification Time: 18:09 [ET]
Event Date: 10/27/2012
Event Time: [EST]
Last Update Date: 03/29/2013 | Emergency Class: NON EMERGENCY
10 CFR Section:
21.21(a)(2) - INTERIM EVAL OF DEVIATION
| Person (Organization):
CHRISTOPHER NEWPORT (R1DO)
MARK FRANKE (R2DO)
JOHN GIESSNER (R3DO)
VINCENT GADDY (R4DO)
PART 21 REACTORS GRP (E-MA)
| Event Text PART 21 INTERIM REPORT ON THE FAILURE OF AN EMERGENCY DIESEL GENERATOR EXCITATION SYSTEM
The following report was received from MPR Associates via facsimile:
"MPR Associates (MPR) is investigating the failure of a replacement emergency diesel generator excitation system that MPR supplied to Cooper Nuclear Station. The root cause investigation is still in-process and will not be completed within 60 days of discovery as defined by 10 CFR Part 21.
"The 10 CFR Part 21 Interim Report [below] provides the information known at this time. An updated report will be provided once the root cause investigation is completed.
IDENTIFICATION OF THE BASIC COMPONENT THAT FAILED
"The basic component is a Basler Electric SBSR emergency diesel generator (EDG) excitation system that was supplied as a replacement system to Cooper Nuclear Station (CNS). The replacement system included design changes relative to the original CNS excitation system, which is also a Basler Electric SBSR design. The design changes included larger magnetic components, which were intended to allow for continuous operation of the new exciter at the EDG overload rating.
IDENTIFICATION OF THE SUPPLIER
"The excitation system was supplied by MPR Associates (headquarters in Alexandria, VA).
"Basler Electric (headquarters in Highland, IL) designed and fabricated the system under a commercial grade program certified to ISO 9001:2008. MPR Associates dedicated the commercial grade item for nuclear use under the MPR Nuclear QA Program, which complies with 10 CFR 50 Appendix B and ASME NQA-1.
NATURE OF THE FAILURE
"CNS installed the replacement excitation system in the Division 1 EDG system during refueling outage RE27. Prior to declaring the EDG operable, CNS manually terminated an EDG maintenance run due to erratic EDG reactive power indication, which was followed by a sudden drop of EDG reactive load and an indication of negative reactive power.
"Several rounds of troubleshooting and surveillance testing were performed unsuccessfully. The surveillance testing resulted in faults to ground and overheating and failure of components in the excitation system. Some of the failed equipment included the automatic voltage regulator (AVR), manual voltage control autotransformer (T60), rectifier power diode failure indication resistors and light-emitting diodes (LEDs), insulation on the control windings of two saturable transformers, and elements of the data acquisition equipment used to record data during the testing.
"The root cause of the failure has not yet been determined. However, on-site troubleshooting efforts at CNS by MPR and Basler Electric identified unexpected high voltages across the direct current (DC) control winding of the saturable transformers. These voltages are likely the cause of the failures experienced in-situ at CNS.
"Follow-up tests at Basler Electric on a similar replacement system designed for the Hatch Plant (but not yet installed in the plant) also identified higher than expected voltages across the DC control windings of the saturable transformers. MPR and Basler Electric recommended postponement of the Hatch Plant installation until the impact of this condition (i.e., the higher than expected voltages) is evaluated. Note that the testing on the Hatch replacement system to date did not result in failure of the system or abnormal function of components external to the saturable transformers.
NATURE OF THE FAILURE
"Testing of the replacement excitation system began on October 27, 2012. The maintenance run resulting in erratic EDG reactive power indication was performed on October 30, 2012. CNS, MPR, and Basler Electric discontinued troubleshooting efforts for the replacement excitation system on November 9, 2012.
"The replacement system was removed, and the original Basler Electric SBSR excitation system was re-installed. CNS declared the re-installed system operable on November 14, 2012.
"MPR formally documented the issue in the MPR corrective action program on November 13, 2012.
NUMBER AND LOCATION OF THE AFFECTED BASIC COMPONENTS
"Based on the information known to date, this 10 CFR Part 21 Interim Report affects the following SBSR type excitation systems that were dedicated and supplied by MPR.
Nuclear Plant Date Equipment Provided Items Supplied
Cooper Nuclear 2012 1 Systems (failed during installation)
Hatch 2012 5 Systems (not yet installed)
CORRECTIVE ACTION PLAN
"MPR is performing a failure analysis and root cause investigation to determine the extent of the condition, corrective actions, and actions to prevent recurrence. The root cause investigation is scheduled for completion by March 29, 2013.
ADVICE GIVEN TO PURCHASERS OR LICENSEES
"There are numerous Basler Electric SBSR type excitation systems in service at multiple plants throughout the nuclear industry. In addition to CNS and Hatch, MPR has dedicated and supplied SBSR excitation systems to the Beaver Valley, Davis-Besse, and Robinson plants. Each system supplied is custom designed for the generator that it is slated to control. Basler Electric SBSR excitation systems have demonstrated reliable service for many years.
"The replacement SBSR excitation system supplied to CNS was not identical to the original system. Specifically, there were design differences in some of the components, including larger transformers, which were intended to allow for continuous operation of the new exciter at the EDG overload rating. Although it was not foreseen (and not revealed by factory acceptance testing), it seems that the design changes in the replacement system led to its maloperation and failure when it was initially installed at CNS. This faulty operation and failure were readily observed as part of normal EDG surveillance testing. Upon re-installation, the original SBSR excitation system functioned properly. Hence, it appears that:
" - Differences between the replacement system and original system lead to the problem, and
" - The problem is readily detectable in normal surveillance testing.
"For these reasons, SBSR excitation systems installed at plants that have shown reliable operation during surveillance testing are in a satisfactory state and condition. MPR has no evidence that the mechanism or conditions that led to the failure at CNS will lead to failures at other installations. Therefore, plants with SBSR excitation systems installed should continue to use them and conduct normal surveillance testing."
* * * UPDATE FROM PAUL DAMERELL (VIA FAX) TO HOWIE CROUCH AT 1830 EDT ON 3/29/13 * * *
Reporting Individual: Paul Damerell, Principal Officer, MPR Associates, Inc., 320 King Street Alexandria, VA 22314.
As indicated in the interim report, the EDG excitation system supplied to Cooper Nuclear Station failed. The root cause analysis identified a manufacturing defect induced design error in the saturable transformers supplied with the excitation system. The manufacturing error allowed the transformer control winding coils to shift which weakened the dielectric strength of the transformer control winding insulation system such that it could not withstand the voltage induced in the control windings. Testing and analysis estimated induced voltages exceeded 10 kV peak at motor start.
MPR Associates also determined that H.B. Robinson has 6 saturable transformers installed with 4 spares and Farley has 3 spares with potential manufacturing defects. They noted that all installed transformers that have passed installation and surveillance testing to date have no immediate concerns but could have long-term reliability issues, if the manufacturing deficiency is present.
MPR and Basler are coordinating with Cooper and Hatch to replace saturable transformers supplied by MPR. Cooper has 3 and Hatch has 15 of the devices. Additionally, they will coordinate with Robinson and Farley to test their transformers to determine if the manufacturing defects exist.
All affected licensees have been notified by MPR Associates.
Notified R1DO (Krohn), R2DO (Seymour), R3DO (Lara), R4DO (Pick) and Part 21 Group (via email). |
| Agreement State | Event Number: 48839 | Rep Org: OHIO BUREAU OF RADIATION PROTECTION
Licensee: AKRON GENERAL MEDICAL CENTER
Region: 3
City: AKRON State: OH
County:
License #: 02120-78-0000
Agreement: Y
Docket:
NRC Notified By: KARL VONAHN
HQ OPS Officer: CHARLES TEAL | Notification Date: 03/21/2013
Notification Time: 09:13 [ET]
Event Date: 03/19/2013
Event Time: [EDT]
Last Update Date: 03/21/2013 | Emergency Class: NON EMERGENCY
10 CFR Section:
AGREEMENT STATE
| Person (Organization):
DAVE PASSEHL (R3DO)
FSME EVENT RESOURCE (EMAI)
| Event Text AGREEMENT STATE REPORT - Y-90 SIR-SPHERE UNDER DOSE OF 87%
"The licensee notified the Department [Ohio Department of Health] on March 20, 2013 of a medical event that occurred on March 19, 2013 involving a Y-90 SIR-Sphere patient treatment. The delivery system became clogged and delivered only 3.2 mCi (13%) of the 24.3 mCi prescribed activity. The cause of the event is under investigation. The licensee plans on retreating the patient."
Ohio Item Number: OH130002
A Medical Event may indicate potential problems in a medical facility's use of radioactive materials. It does not necessarily result in harm to the patient. |
| Non-Agreement State | Event Number: 48841 | Rep Org: EI DUPONT DE NEMOURS, PENCADER SITE
Licensee:
Region: 1
City: NEWARK State: DE
County:
License #:
Agreement: N
Docket:
NRC Notified By: SHIRLEY STEWART
HQ OPS Officer: PETE SNYDER | Notification Date: 03/21/2013
Notification Time: 16:25 [ET]
Event Date: 03/21/2013
Event Time: [EDT]
Last Update Date: 03/21/2013 | Emergency Class: NON EMERGENCY
10 CFR Section:
20.2201(a)(1)(i) - LOST/STOLEN LNM>1000X
| Person (Organization):
JAMES TRAPP (R1DO)
FSME EVENT RESOURCE (EMAI)
|
| This material event contains a "Less than Cat 3" level of radioactive material. | Event Text MISSING TRITIUM EXIT SIGN
EI DuPont de Nemours and Company Inc.(DuPont) during a routine periodic audit identified that one tritium exit sign was missing from a doorway at their Pencader Site. It is believed that the sign was misplaced during recent remodeling/construction in the area.
A thorough search has been conducted by both the demolition vendor and DuPont and the sign is presumed lost. The missing sign was a Betalux E-Series, Model 171, Serial Number C090859 with a manufacture date of 7/2012. There was 21.6 curies of tritium gas in the sign at the date of manufacture.
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. For additional information go to http://www-pub.iaea.org/MTCD/publications/PDF/Pub1227_web.pdf
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: 48861 | Facility: OYSTER CREEK
Region: 1 State: NJ
Unit: [1] [ ] [ ]
RX Type: [1] GE-2
NRC Notified By: BRYAN EAGAN
HQ OPS Officer: MARK ABRAMOVITZ | Notification Date: 03/28/2013
Notification Time: 10:15 [ET]
Event Date: 03/28/2013
Event Time: 08:04 [EDT]
Last Update Date: 03/29/2013 | Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(3)(xiii) - LOSS COMM/ASMT/RESPONSE
| Person (Organization):
PAUL KROHN (R1DO)
|
| 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 TECHNICAL SUPPORT CENTER VENTILATION OUT OF SERVICE FOR PLANNED MAINTENANCE
"A planned maintenance evolution at the Oyster Creek Generating Station has removed the Technical Support Center (TSC) ventilation system from service. The TSC ventilation system will be rendered non-functional during the course of the work activities. The TSC ventilation is expected to be out of service for approximately fourteen hours and will return to service at approximately 2200 [EDT] March 28, 2013.
"If an emergency is declared requiring TSC activation during this period, the TSC will be staffed and activated using existing emergency planning procedures unless the TSC becomes uninhabitable due to ambient temperature, radiological, or other conditions. If relocation of the TSC becomes necessary, the Emergency Director will relocate the TSC staff to an alternate location in accordance with applicable site procedures.
"This notification is being made in accordance with 10CFR50.72(b)(3)(xiii) due to potential loss of the TSC. An update will be provided once the TSC ventilation has been restored to normal operation."
The licensee informed the NRC Resident Inspector.
* * * UPDATE FROM ERIC SWAIN TO HOWIE CROUCH AT 1245 EDT ON 3/29/13 * * *
The TSC ventilation system was returned to service at 1240 EDT. The licensee has notified the NRC Resident Inspector.
Notified R1DO (Krohn). |
| Part 21 | Event Number: 48863 | Rep Org: INTEGRATED RESOURCES, INC.
Licensee: INVENSYS (FOXBORO METER CO.)
Region: 4
City: NEBRASKA CITY State: NE
County:
License #:
Agreement: Y
Docket:
NRC Notified By: JOHN F. BROSEMER
HQ OPS Officer: HOWIE CROUCH | Notification Date: 03/28/2013
Notification Time: 15:53 [ET]
Event Date: 03/27/2013
Event Time: 15:30 [CDT]
Last Update Date: 04/01/2013 | Emergency Class: NON EMERGENCY
10 CFR Section:
21.21(d)(3)(i) - DEFECTS AND NONCOMPLIANCE
| Person (Organization):
GREG PICK (R4DO)
PAUL KROHN (R1DO)
PART 21 GROUP (EMAI)
DEBORAH SEYMOUR (R2DO)
JULIO LARA (R3DO)
| Event Text PART 21 REPORT - FOXBORO POWER SUPPLY POTENTIAL FAILURES DUE TO DEFECTIVE TIE WRAPS AND HOLDERS
Mr. John F. Brosemer, President of Integrated Resources, Inc., reported discovery of repeated defects in Foxboro Meter Company's N-2ARPS-A6, Style D power supplies. When manufactured, the power supplies utilized Thomas and Betts TC105A aluminum wire tie holders in random numbers and placements. As the power supplies age, the tie wrap holder adhesive degrades and the tie wraps embrittle resulting in the separation of the tie wraps and loss of holder adhesion to the power supply enclosure. This causes the wraps and holders to fall to the bottom of the enclosure which could result in shorts when the aluminum comes in contact with electronic components. In one particular power supply, all tie wrap holders in use failed and separated from the enclosure.
The power supplies are used in Foxboro SPEC-200 cabinetry that are used throughout the industry. At the time of this notification, Integrated Resources has one power supply from Three Mile Island and two power supplies from Ft. Calhoun undergoing refurbishment. Integrated Resources will be following up this telephonic notification with a written report once their internal investigation is done.
Recommended corrective actions are for affected facilities to open and inspect all power supplies and remove the aluminum tie wrap holders and replace the tie wraps and holders with Teflon types.
* * * UPDATE FROM BROSEMER TO SNYDER AT 1530 EDT ON 4/1/13 * * *
"Suspecting this to be a common mode failure IRI [Integrated Resources, Inc.] opened and inspected two Foxboro N-2ARPS-A6 power supplies which were sent to IRI for refurbishment by Fort Calhoun Nuclear Station. Examination revealed that both of the power supplies have the same failures of the tie wrap aluminum mounting plates adhesive with the majority of the plates being held on the wire bundles by age embrittled nylon wire ties.
"Confirmation of the common mode failure by inspection of the Fort Calhoun Nuclear Stations was on or about 1530 CDT on March 27, 2013.
"IRI is not the OEM or Original supplier for this power supply and cannot provide the number nor locations of these components. However, by searching the RAPID database IRI has found the power supplies at the following:
"Arizona Public Service - Palo Verde Nuclear Generating Station; Constellation Energy - Nine Mile Point Nuclear Power Plant; Detroit Edison - Fermi 2 Nuclear Power Plant; Dominion Nuclear - Millstone Nuclear Power Plant; Dominion Nuclear - Kewaunee Nuclear Power Plant; Eletronnuclear - Angra Nuclear Power Plant; Entergy Nuclear - Arkansas Nuclear One; Entergy Nuclear - Indian Point Energy Center; Entergy Nuclear - Pilgrim Nuclear Power Plant; Entergy Nuclear - J. A. Fitzpatrick Nuclear Power Plant; Exelon Corporation - Three Mile Island Nuclear Plant; Exelon Corporation - Peach Bottom Atomic Power Station; NextEra Energy - Point Beach Nuclear Power Plant; Progress Energy Florida - Crystal River Nuclear Power Plant; Southern California Edison - San Onofre Nuclear Generating Station.
"IRI suspects several other utilities and units are affected by this report.
Corrective action taken: "IRI's preliminary suggestion is inspection and removal of failed tie wrap mounting plates which are being held on to wire bundles by aging nylon tie wraps. IRI also suggests replacement of age embrittled nylon tie wraps with Tefzel tie wraps."
Contact Information:
John F. Brosemer; President
Integrated Resources, Inc.
113 South 9th Street
Nebraska City, NE 68410
Notified R1DO (Dwyer), R2DO (Seymour), R3DO (Daley), R4DO (Kellar) and Part 21 Reactors (Email). |
| Power Reactor | Event Number: 48868 | Facility: COOK
Region: 3 State: MI
Unit: [1] [ ] [ ]
RX Type: [1] W-4-LP,[2] W-4-LP
NRC Notified By: GREGORY KANDA
HQ OPS Officer: DONALD NORWOOD | Notification Date: 03/31/2013
Notification Time: 09:08 [ET]
Event Date: 03/31/2013
Event Time: 04:14 [EDT]
Last Update Date: 03/31/2013 | Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(3)(ii)(A) - DEGRADED CONDITION
| Person (Organization):
JULIO LARA (R3DO)
|
| Unit | SCRAM Code | RX CRIT | Initial PWR | Initial RX Mode | Current PWR | Current RX Mode | | 1 | N | N | 0 | Cold Shutdown | 0 | Cold Shutdown | Event Text RCS FLOW INSTRUMENT LINE THROUGH WALL WELD DEFECT
"On 3/31/13 at 0414 EDT, dye penetrant testing identified a 10mm long relevant linear indication on a 3/4 inch RCS flow instrument line weld, between the RCS loop piping and the instrument isolation valve. This constitutes a weld defect in the primary coolant system that cannot be found acceptable per ASME Section XI. This section of piping is unisolable from the Reactor Coolant System. There is no evidence of active leakage currently or during the previous operating cycle, however, based on the dye penetrant testing this is considered a through wall leak.
"This event notification is being made in accordance with 10CFR50.72(b)(3)(ii)(A), as a condition that results in the condition of the nuclear power plant, including its principal safety barriers, being seriously degraded.
"Unit 1 is currently in a scheduled refueling outage in Mode 5 making preparations to transition to Mode 6. The plant is evaluating the appropriate repair method for this condition. The condition will be corrected during the current refueling outage. Unit 2 is in Mode 1 at 100% power and is unaffected by this condition.
"The NRC Resident Inspector has been notified." |
| Power Reactor | Event Number: 48869 | Facility: ARKANSAS NUCLEAR
Region: 4 State: AR
Unit: [1] [2] [ ]
RX Type: [1] B&W-L-LP,[2] CE
NRC Notified By: PHILLIP FORE
HQ OPS Officer: HOWIE CROUCH | Notification Date: 03/31/2013
Notification Time: 11:57 [ET]
Event Date: 03/31/2013
Event Time: 10:33 [CDT]
Last Update Date: 04/02/2013 | Emergency Class: UNUSUAL EVENT
10 CFR Section:
50.72(a) (1) (i) - EMERGENCY DECLARED
50.72(b)(2)(iv)(B) - RPS ACTUATION - CRITICAL
50.72(b)(3)(iv)(A) - VALID SPECIF SYS ACTUATION
50.72(b)(2)(xi) - OFFSITE NOTIFICATION
| Person (Organization):
GREG PICK (R4DO)
ART HOWELL (R4RA)
JENNIFER UHLE (NRR)
ALLEN HOWE (NRR)
WILLIAM GOTT (IRD)
|
| Unit | SCRAM Code | RX CRIT | Initial PWR | Initial RX Mode | Current PWR | Current RX Mode | | 1 | N | N | 0 | Refueling | 0 | Refueling | | 2 | A/R | Y | 100 | Power Operation | 0 | Hot Standby | Event Text NOTIFICATION OF UNUSUAL EVENT DECLARED DUE TO A BREAKER EXPLOSION IN THE PROTECTED AREA
"At 0750 [CDT] on 3/31/2013, during movement of the Unit 1 Main Turbine Generator Stator (~500 tons), the Unit 1 turbine temporary lift device failed. This caused a loss of all off-site power on Unit 1. The ANO Unit 1 #1 and #2 EDG [Emergency Diesel Generator] have started and are supplying A-3 4160V switchgear and A-4 4160V switchgear. P-4A Service Water pump and P-4C Service Water pump has been verified running. Unit 1 has entered [procedures] 1202.007 - Degraded Power, 1203.028 - Loss of Decay Heat, and 1203.050 - Spent Fuel Emergencies. Unit 1 is in MODE 6.
"ANO-1 entered TS 3.8.2 A, 'One Required Offsite Circuit Inoperable'. All required actions are complete. The event caused a loss of decay heat removal on ANO Unit 1 which was restored in 3 minutes and 50 seconds.
"Unit 2 tripped and is in MODE 3. Emergency Feed Water was initiated on Unit 2 and Unit 2 was in [Technical Specification] 3.0.3 from 0817 [CDT] to 0848 [CDT] due to Emergency Feedwater. Unit 2 is being powered by off-site. Unit 2 Startup 3 [transformer] lock out at 0921 [CDT]. [Bus] 2A1 is on Start up 2 [transformer] and [bus] 2A3 is on #2 EDG.
"10CFR50.72 (b)(3)(iv)(A) - 4-hr. notification due to the ES [Engineered Safeguard Feature] actuation on both Unit 1 and Unit 2.
10CFR50 72 (b)(2)(iv)(B) - 4-hr. notification due to RPS [Reactor Protection System] actuation on Unit 2.
10CFR50.72 (b)(2)(xi) - 4-hr. notification due to Government Notification.
29CFR1904.39a - [OSHA] 8-hr. notification due to death on site.
"At 1033 [CDT] on 3/31/2013, Unit 2 entered a Notification of Unusual Event based on EAL HU4 due to damage in 2A1 switchgear. Notification of the NUE will be made lAW Emergency Plan requirements. Follow-up notifications will be made as appropriate."
At this time, the full extent of structural damage on Unit 1 is not known. There was one known fatality and 4 known serious injuries to workers. The local coroner is on site for the fatality and the injured personnel have been transported offsite to local hospitals. Investigation into the cause of the failure and extent of damage is ongoing.
On Unit 2, all rods inserted during the trip. The core is being cooled via natural circulation. Decay heat is being removed via steam dumps to atmosphere. There is no known primary to secondary leakage.
The licensee has notified the State of Arkansas, local authorities, OSHA and the NRC Resident Inspector.
Notified DHS SWO, DHS NICC, FEMA and Nuclear NSSA (via email).
* * * UPDATE FROM DAVID THOMPSON TO HOWIE CROUCH AT 1934 EDT ON 3/31/13 * * *
The licensee terminated the NOUE at 1821 CDT. The basis for termination was that the affected bus (2A2) is de-energized and no other equipment on Unit 2 was damaged.
The licensee has notified the state and local authorities and will be notifying the NRC Resident Inspector.
Notified R4DO (Pick), NRR EO (Howe), IRD (Gott), DHS SWO, DHS NICC, FEMA and Nuclear SSA (via email).
* * * UPDATE FROM STEVE COFFMAN TO HOWIE CROUCH AT 1054 EDT ON 4/2/13 * * *
The licensee made the following edits to the third paragraph of their original report (edits in quotes):
Unit 2 tripped and is in MODE 3. Emergency Feed Water initiated on Unit 2. Unit 2 was in [Technical Specification] 3.0.3 from 0817 [CDT] to 0848 [CDT] due to Emergency Feedwater "being procedurally overridden." Unit 2 "was initially" being powered by off-site. Unit 2 Startup 3 Lock out occurred at 0921. 2A1 is now on Startup 2, and "2A4" is on #2 EDG.
Notified R4DO (Kellar) via email. |
| Power Reactor | Event Number: 48870 | Facility: PRAIRIE ISLAND
Region: 3 State: MN
Unit: [1] [2] [ ]
RX Type: [1] W-2-LP,[2] W-2-LP
NRC Notified By: WAYNE EPPEN
HQ OPS Officer: HOWIE CROUCH | Notification Date: 03/31/2013
Notification Time: 15:07 [ET]
Event Date: 03/30/2013
Event Time: 23:44 [CDT]
Last Update Date: 03/31/2013 | Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(3)(v)(C) - POT UNCNTRL RAD REL
| Person (Organization):
JULIO LARA (R3DO)
|
| 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 BOTH TRAINS OF AUXILIARY BUILDING SPECIAL VENTILATION SYSTEM DECLARED INOPERABLE
"During testing of the Auxiliary Building Ventilation System it was discovered that the outlet damper of one of the normal make-up fans did not close. This created a non-closable opening in the Auxiliary Bldg Special Vent System (ABSVS) boundary of greater than 10 square feet and required entry into TS [Technical Specification] 3.7.12, Condition B, for two ABSVS trains inoperable due to inoperable ABSVS boundary. This is considered a potential loss of safety function to control the release of material."
The licensee has notified the NRC Resident Inspector. |
| Power Reactor | Event Number: 48871 | Facility: WATTS BAR
Region: 2 State: TN
Unit: [ ] [2] [ ]
RX Type: [1] W-4-LP,[2] W-4-LP
NRC Notified By: GORDON ARENT
HQ OPS Officer: JOHN SHOEMAKER | Notification Date: 04/01/2013
Notification Time: 09:52 [ET]
Event Date: 02/11/2013
Event Time: [EDT]
Last Update Date: 04/01/2013 | Emergency Class: NON EMERGENCY
10 CFR Section:
50.55(e) - CONSTRUCT DEFICIENCY
| Person (Organization):
DEBORAH SEYMOUR (R2DO)
NRR 50.55 COORD (EMAI)
|
| Unit | SCRAM Code | RX CRIT | Initial PWR | Initial RX Mode | Current PWR | Current RX Mode | | 2 | N | N | 0 | Under Construction | 0 | Under Construction | Event Text INSTRUMENTATION LINES NOT INSPECTED COMPLETELY FOR PROPER SLOPE
"WBN [Watts Bar Nuclear] Unit 2 (under construction) determined that a portion of a number of instrumentation lines within multiple systems may have not been inspected completely for proper slope. This condition may have resulted in sense lines being installed with less than the 1/4" per foot minimum slope. No confirmed examples have been identified that would have created a substantial safety hazard at this time. However, walkdowns and evaluations are still underway to confirm that no substantial safety hazards exist. If any examples are found, they will be corrected prior to system turnover to Plant Operations. Therefore, at this time, the safety significance remains indeterminate. This issue has been documented in TVA's corrective action program as Problem Evaluation Report 680826 and is being conservatively reported as a programmatic breakdown by WBN Unit 2 under 10 CFR 50.55(e)."
The licensee has notified the NRC Resident Inspector. |
| Part 21 | Event Number: 48872 | Rep Org: ABB INC.
Licensee: ABB INC.
Region: 1
City: CORAL SPRINGS State: FL
County:
License #:
Agreement: Y
Docket:
NRC Notified By: DENNIS BATOVSKY
HQ OPS Officer: PETE SNYDER | Notification Date: 04/01/2013
Notification Time: 17:29 [ET]
Event Date: 02/04/2013
Event Time: [EDT]
Last Update Date: 04/01/2013 | Emergency Class: NON EMERGENCY
10 CFR Section:
21.21(d)(3)(i) - DEFECTS AND NONCOMPLIANCE
| Person (Organization):
JAMES DWYER (R1DO)
DEBORAH SEYMOUR (R2DO)
ROBERT DALEY (R3DO)
RAY KELLAR (R4DO)
PART 21 REACTORS (EMAI)
| Event Text POTENTIAL ISSUE CONCERNING THE ZPA RATING FOR 3 PHASE RELAY TYPE SSC-T
"This letter is to notify you of a potential issue concerning the ZPA [Zero Point Acceleration] rating for our three phase relay type SSC-T.
"During a Customer Audit (week of February 4, 2013 by ABB Inc., Florence, S.C.), we discovered that our Class 1E Certification database showed the ZPA rating of our three phase SSV -T and SSC-T relays was 5.6g, while our most recent qualification (dated June 14, 2012) determined that the ZPA rating was 4.79g.
"A review of our Purchaser records for the last ten years indicates that one (1) affected relay was sold to WESCO:
"SSC-T relay completed 10/5/2007, order number 3365-960825 and serial number 11835.
"ABB does not have the capability to perform the evaluation to determine if a defect exists, so we are informing the purchaser of this determination so that they may evaluate the deviation or failure to comply, pursuant to 10CFR 21.21(a).
"ABB recommends that the affected licensee evaluate their specific application and determine whether the deviation described in this notice affects their design basis. If the licensee determines that it does, the licensee should contact ABB to determine appropriate corrective action.
"If you have any questions regarding this notice, please contact ABB Technical Support at 954-752-6700." | |