| U.S. Nuclear Regulatory Commission
Operations Center Event Reports For
06/06/2012 - 06/07/2012 ** EVENT NUMBERS ** |
 | Part 21 | Event Number: 47630 | Rep Org: GE HITACHI NUCLEAR ENERGY
Licensee: GE HITACHI NUCLEAR ENERGY
Region: 1
City: WILMINGTON State: NC
County:
License #:
Agreement: Y
Docket:
NRC Notified By: DALE PORTER
HQ OPS Officer: JOHN KNOKE | Notification Date: 02/01/2012
Notification Time: 15:33 [ET]
Event Date: 02/01/2012
Event Time: [EST]
Last Update Date: 06/06/2012 | Emergency Class: NON EMERGENCY
10 CFR Section:
21.21(a)(2) - INTERIM EVAL OF DEVIATION
| Person (Organization):
WILLIAM COOK (R1DO)
JONATHAN BARTLEY (R2DO)
JAMNES CAMERON (R3DO)
JEFF CLARK (R4DO)
PART21 GROUP ()
| Event Text PART 21 REPORT - FAILURE OF CRD COLLET RETAINER TUBE/OUTER TUBE WELD
The following information was received via facsimile:
"During a recent refurbishment of a Control Rod Drive (CRD) performed by GE Hitachi Nuclear Energy (GEH) for a domestic customer a 360 degree failure of the collet retainer tube fillet weld was identified. This weld is part of the CRD 919D258G003 Cylinder, Tube and Flange (CTF) assembly. The collet retainer tube fillet weld was performed in 1983 and subsequently assembled into a Group 003 part number 919D258G003 CTF. This G003 CTF assembly was assembled into a CRD in 1995 and placed into service in 1996. GEH continues to investigate the cause(s) of the failed fillet weld. Once the cause of the fillet weld failure is determined, GEH will review the extent of condition of this failure as well as the consequences to determine if a reportable condition exists.
"There were no adverse effects on the CRD's operation observed due to this failure.
"This 60-day interim notification, in accordance with 10CFR Part 21.21(a)(2), will be sent to all BWR/2-6 plants that utilize CRDs equipped with either 919D258G002 or 919D258G003 CTF assemblies."
The affected plants are: Nine Mile Point 1-2, Fermi 2, Columbia, Grand Gulf, River Bend, Fitzpatrick, Pilgrim, Vermont Yankee, Clinton, Dresden 2-3, LaSalle 1-2, Limerick 1-2, Oyster Creek, Peach Bottom 2-3, Quad Cities 1-2, Perry 1, Duane Arnold, Cooper, Susquehanna 1-2, Brunswick 1-2, Hope Creek, Hatch 1 - 2, Browns Ferry 1-3, Monticello, and Millstone.
* * * UPDATE FROM GE HITACHI VIA FAX AT 1259 EDT ON 6/6/12 * * *
"GEH has completed the evaluation of this condition and has determined that the failure of Control Rod Drive collet retainer tube fillet weld is not a Reportable Condition as defined by 10CFR Part 21."
Notified R1DO (Cahill), R2DO (Widmann), R3DO (Passehl), R4DO (Gepford) and Part 21 Group (via email). |
| !!!!! THIS EVENT HAS BEEN RETRACTED. THIS EVENT HAS BEEN RETRACTED !!!!! | | Power Reactor | Event Number: 47816 | Facility: SUSQUEHANNA
Region: 1 State: PA
Unit: [1] [2] [ ]
RX Type: [1] GE-4,[2] GE-4
NRC Notified By: ALEX McLELLAN
HQ OPS Officer: MARK ABRAMOVITZ | Notification Date: 04/09/2012
Notification Time: 02:07 [ET]
Event Date: 04/09/2012
Event Time: 01:02 [EDT]
Last Update Date: 06/07/2012 | Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(3)(v)(D) - ACCIDENT MITIGATION
| Person (Organization):
JOHN CARUSO (R1DO)
|
| Unit | SCRAM Code | RX CRIT | Initial PWR | Initial RX Mode | Current PWR | Current RX Mode | | 1 | N | N | 0 | Refueling | 0 | Refueling | | 2 | N | Y | 100 | Power Operation | 100 | Power Operation | Event Text BOTH CONTROL STRUCTURE CHILLERS DECLARED INOPERABLE WHILE SWITCHING POWER SUPPLIES
"On 4/9/2012, starting at 0102 EDT, the 'A' and 'C' Emergency Diesel Generators (EDG) were sequentially and briefly declared inoperable to switch their DC control power back to their normal supplies. Switching power to the normal supply is required by Unit 2 technical specification 3.8.4 following maintenance work on the U1 power supplies. Previously, at 18:35 EDT on 4/4/2012, the 'B' Control Structure Chiller was declared inoperable due to an unrelated problem. With the 'B' Control Structure Chiller inoperable coincident with the 'A' EDG or 'C' EDG inoperable, neither Control Structure Chiller would be available to perform its design function on a loss of offsite power. This is a condition that, at the time of discovery, could have prevented fulfillment of a Safety Function and is reportable under 50.72(b)(3)(v)(D) as an 8 hour notification.
"Switching the power supplies was a planned evolution. The duration of the loss of safety function was a total of eight minutes. As a mitigating action, operators were continuously available with communication to the control room. The associated diesel generator could have been returned to an operable condition promptly if required.
"Note that Technical Specifications allows four hours to correct the condition before further actions are required, i.e. declare the features ('A' Control Structure Chiller) supported by the inoperable diesel inoperable."
The licensee notified the NRC Resident Inspector.
***RETRACTION FROM RON FRY TO S. SANDIN ON 6/7/12 AT 0205 EDT***
The licensee is retracting this report based on the following:
"On April 9, 2012, Susquehanna reported that the 'A' and 'C' Emergency Diesel Generators (EDGs) were sequentially and briefly declared inoperable to switch their DC control power back to their normal supplies while the 'B' Control Structure Chiller was inoperable. The basis for the 8 hour notification, which was reported under Reporting Requirement 50.72(b)(3)(v)(D), was the conclusion that neither Control Structure Chiller would be available to perform its design function on a loss of offsite power.
"The reporting guidance in NUREG-1022, Revision 2 identifies events or conditions that are generally not reportable in accordance with 50.72(b)(3)(v). One of the identified conditions is 'removal of a system or part of a system from service as part of a planned evolution for maintenance or surveillance testing when done in accordance with an approved procedure and the plant's TS (unless a condition is discovered that could have prevented the system from performing its function).' After further review, Susquehanna has determined that an ENS report was not required for this event since the EDGs and the associated 'A' Chiller were removed from service as part of a planned evolution in accordance with approved procedures and the plant Technical Specifications and no condition was discovered that could have prevented the EDGs and associated 'A' chiller from performing their function.
"Based on the above information, this ENS report is retracted."
The licensee informed the NRC Resident Inspector. Notified R1DO(Cahill). |
| !!!!! THIS EVENT HAS BEEN RETRACTED. THIS EVENT HAS BEEN RETRACTED !!!!! | | Power Reactor | Event Number: 47844 | Facility: SUSQUEHANNA
Region: 1 State: PA
Unit: [ ] [2] [ ]
RX Type: [1] GE-4,[2] GE-4
NRC Notified By: ALEX MCLELLAN
HQ OPS Officer: JOHN SHOEMAKER | Notification Date: 04/17/2012
Notification Time: 21:02 [ET]
Event Date: 04/17/2012
Event Time: 15:40 [EDT]
Last Update Date: 06/07/2012 | Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(3)(v)(C) - POT UNCNTRL RAD REL
| Person (Organization):
JUDY JOUSTRA (R1DO)
|
| 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 UNIT 2 SECONDARY CONTAINMENT AFFECTED BY VIOLATION OF UNIT 1 SECONDARY CONTAINMENT INTEGRITY DURING OUTAGE
"At 1540 (EDT) on 4/17/12, with Unit 1 in mode 5 and Unit 2 in mode 1, the Work Control Center was notified that the U1 #2 Main Stop Valve (MSV) was disassembled. The U1 #2 MSV was required to be intact to maintain Unit 1 Secondary Containment. Ongoing work on the D Main Steam Line Outboard Valve created a pathway that violated Unit 1 secondary containment integrity. Unit 1 Secondary Containment is required to be operable for Unit 2 while Unit 1 Zone 1 is aligned to the Recirculation Plenum. Unit 1 Zone 1 was isolated from the recirculation plenum and Unit 2 Secondary Containment was restored at 1643 (EDT) on 4/17/12. Unit 2 Secondary Containment differential pressures were maintained throughout the event.
"This is considered a loss of an entire safety function and requires an 8 hour report per 10CFR50.72(b)(3)(v)(C)."
The licensee is still investigating the cause but it appears to be associated with recent administrative changes to the Reactor Vessel draining definition and work process procedures.
The licensee has notified the NRC Resident Inspector.
*** RETRACTION FROM RON FRY TO S. SANDIN ON 6/7/12 AT 0205 EDT ***
The licensee is retracting this report based on the following:
"On April 17, 2012, work on the Unit 1 'D' Main Steam Line Outboard Valve with the Unit 1 #2 Main Stop Valve disassembled created a pathway that violated Unit 1 secondary containment integrity. Since Unit 1 Secondary Containment is required to be operable for Unit 2 while Unit 1 Zone 1 would be aligned to the Recirculation Plenum in the event of a secondary containment isolation signal, the condition impacted Unit 2 Secondary Containment. Susquehanna considered the impact a loss of safety function and reported the impact in accordance with 10CFR50.72(b)(3)(v)(C).
"Following the ENS report, Susquehanna analyzed the impact of the opening. Calculations were performed that show secondary containment would have maintained the dose consequences to the public and control room operators within regulatory limits (10 CFR 50.67) assuming a Unit 2 design basis accident (Unit 1 was in a refueling outage at the time of the condition).
"Based on the above information, Susquehanna has determined that there was no loss of safety function and this ENS report is retracted."
The licensee informed the NRC Resident Inspector. Notified R1DO(Cahill). |
| Non-Agreement State | Event Number: 47974 | Rep Org: 3M CORPORATION
Licensee: 3M CORPORATION
Region: 3
City: SPRINGFIELD State: MO
County:
License #: 22-00057-03
Agreement: N
Docket:
NRC Notified By: MIKE LEWANDOWSKI
HQ OPS Officer: HOWIE CROUCH | Notification Date: 05/29/2012
Notification Time: 16:40 [ET]
Event Date: 05/21/2012
Event Time: 13:30 [CDT]
Last Update Date: 05/29/2012 | Emergency Class: NON EMERGENCY
10 CFR Section:
30.50(b)(2) - SAFETY EQUIPMENT FAILURE
| Person (Organization):
HIRONORI PETERSON (R3DO)
FSME VIA EMAIL ()
|
| This material event contains a "Less than Cat 3" level of radioactive material. | Event Text INDUSTRIAL GAUGE INOPERABLE DUE TO BENT SHUTTER
"On May 21, 2012, the shutter on an industrial thickness gauging device (NDC Infrared Systems model 102) was discovered to be bent so that when closed, it did not adequately attenuate the radiation beam from the device. The facility Radiation Safety Officer removed the gauging device from the production line and secured plate steel over the radiation source to provide supplemental shielding to the same degree as the undamaged shutter. The device was placed in secure storage pending shipment for repair.
"The device was shipped to 3M Corporate Health Physics in St. Paul, MN, on May 25, 2012. On May 29, 2012, the device was received and the shutter assembly replaced with an undamaged, spare shutter assembly. The repair was performed by [corporate radiation safety staff] under US NRC Radioactive Materials license no. 22-00057-03. The device was then returned to 3M Springfield.
"Conversations between 3M Corporate Health Physics and US NRC Region III (Geoffrey Warren) on May 29, 2012, indicated that this event is reportable under 10 CFR 30.50(b)(2).
"Location of event: 3M Springfield, 3211 East Chestnut Expressway, Springfield, MO 65802
"Isotope, quantity, chemical and physical form: Am-241, 150 milliCuries on February, 7, 1997, double encapsulated solid source (special form).
"The event resulted in no additional exposure to any individual."
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: 47997 | Facility: PILGRIM
Region: 1 State: MA
Unit: [1] [ ] [ ]
RX Type: [1] GE-3
NRC Notified By: DAVID NOYES
HQ OPS Officer: JOE O'HARA | Notification Date: 06/06/2012
Notification Time: 00:47 [ET]
Event Date: 06/06/2012
Event Time: 00:45 [EDT]
Last Update Date: 06/06/2012 | Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(2)(xi) - OFFSITE NOTIFICATION
| Person (Organization):
CHRISTOPHER CAHILL (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 OFFSITE NOTIFICATION REGARDING LABOR NEGOTIATIONS
"On June 6, 2012, at 0045 EDT hours and the reactor at 100% core thermal power the following press release being issued by Entergy Nuclear regarding the status of the ongoing labor negotiations.
"Plymouth, Mass. - Entergy Nuclear, the company that owns and operates the Pilgrim Nuclear Power Station is
implementing a contingency staffing plan after weeks of negotiations between the Company and the Utility Workers Union of America Local 369 leadership were unsuccessful in reaching an agreement for a new union contract.
"Local 369 has stated flatly that while its members are not on strike, they reserve the right to walk off the job at any time, without any notice, and leave the nuclear power plant critically understaffed and in violation of the plant's operating license. This disregard for public safety is unacceptable. Accordingly, the company's contingency plan is being implemented.
"The contingency plan for temporary alternate staffing is consistent with Nuclear Regulatory Commission regulations and is in the interest of safety and maintaining formal organizational controls on the plant site. Highly qualified individuals from within Pilgrim's management team as well as the Entergy nuclear fleet will be filling the necessary positions.
"The NRC Resident Inspector staff has been informed of this press release and notification.
"This notification is being made in accordance with 10 CFR 50.72(b)(2)(xi)."
The licensee will be notifying the Massachusetts Emergency Management Agency. |
| Power Reactor | Event Number: 47998 | Facility: WOLF CREEK
Region: 4 State: KS
Unit: [1] [ ] [ ]
RX Type: [1] W-4-LP
NRC Notified By: WARREN BRANDT
HQ OPS Officer: DONG HWA PARK | Notification Date: 06/06/2012
Notification Time: 05:19 [ET]
Event Date: 06/06/2012
Event Time: 03:27 [CDT]
Last Update Date: 06/06/2012 | Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(2)(i) - PLANT S/D REQD BY TS
| Person (Organization):
HEATHER GEPFORD (R4DO)
|
| 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 INADEQUATE COOLING AVAILABLE TO SUPPORT ELECTRICAL POWER SYSTEMS
"Wolf Creek commenced a nuclear plant shutdown in accordance with Technical Specifications. Engineering analysis has determined that with one non-functional Class IE A/C unit, there is inadequate cooling available to support electrical power systems required by Technical Specifications (TS) 3.8.4, 3.8.7, 3.8.9, during all postulated conditions. Actions are currently underway to restore the non-functional Class IE A/C unit to a functional status."
The non-functional Class IE A/C unit had a clogged oil pump strainer. Per TS 3.0.3, Wolf Creek began reducing power and was at 89% power at the time of the report.
The licensee has notified the NRC Resident Inspector.
* * * UPDATE FROM WARREN BRANDT TO DONG PARK AT 0700 EDT ON 06/06/12 * * *
"The A Class IE A/C unit was restored to functional at 6/6/2012 0505 CDT. The plant shutdown was terminated at 87 percent power. Actions are underway to restore the unit to 100 percent power."
Notified R4DO (Gepford). |
| Power Reactor | Event Number: 47999 | Facility: INDIAN POINT
Region: 1 State: NY
Unit: [2] [ ] [ ]
RX Type: [2] W-4-LP,[3] W-4-LP
NRC Notified By: VINCENT DeCLEMENTE
HQ OPS Officer: STEVE SANDIN | Notification Date: 06/06/2012
Notification Time: 09:38 [ET]
Event Date: 06/06/2012
Event Time: 06:12 [EDT]
Last Update Date: 06/06/2012 | Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(2)(iv)(B) - RPS ACTUATION - CRITICAL
50.72(b)(3)(iv)(A) - VALID SPECIF SYS ACTUATION
| Person (Organization):
CHRISTOPHER CAHILL (R1DO)
|
| Unit | SCRAM Code | RX CRIT | Initial PWR | Initial RX Mode | Current PWR | Current RX Mode | | 2 | A/R | Y | 100 | Power Operation | 0 | Hot Standby | Event Text UNIT 2 EXPERIENCED AN AUTOMATIC REACTOR TRIP ON TURBINE MAIN GENERATOR TRIP
"On June 6, 2012, at approximately 0612 hours, Indian Point Unit 2 reactor automatically tripped from 100% steady state power due to a turbine-main generator electrical trip. All rods fully inserted. Investigation into the generator trip is in progress. All systems responded as expected. The Auxiliary Feedwater system actuated and responded as expected and is maintaining steam generator levels. Decay heat removal is via the steam generators to the main condenser. Offsite power and plant electrical lineups are normal. No primary or secondary code safety relief valves lifted. The reactor is in Mode 3 and stable. Unit 3 was unaffected and remains at 100% power."
The NRC Resident Inspector has been notified, and the licensee issued a press release. |
| Power Reactor | Event Number: 48000 | Facility: DAVIS BESSE
Region: 3 State: OH
Unit: [1] [ ] [ ]
RX Type: [1] B&W-R-LP
NRC Notified By: ERIC HORVATH
HQ OPS Officer: DONG HWA PARK | Notification Date: 06/07/2012
Notification Time: 02:39 [ET]
Event Date: 06/06/2012
Event Time: 19:56 [EDT]
Last Update Date: 06/07/2012 | Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(3)(ii)(A) - DEGRADED CONDITION
| Person (Organization):
DAVE PASSEHL (R3DO)
|
| Unit | SCRAM Code | RX CRIT | Initial PWR | Initial RX Mode | Current PWR | Current RX Mode | | 1 | N | N | 0 | Hot Standby | 0 | Hot Standby | Event Text DEGRADED CONDITION DUE TO DISCOVERY OF PRESSURE BOUNDARY LEAKAGE
"On June 6, 2012, at 1956 EDT, with the Unit shutdown for refueling, leakage was identified from a 3/4-inch weld during Reactor Coolant System (RCS) walkdown inspections. The leakage amount was approximately 0.1 gpm pinhole spray.
"During the performance of MODE 3 engineering walkdown inspections in accordance with procedure DB-PF-03010 (ASME Section III, Class 1 and 2), with the RCS at Normal Operating Temperature and Pressure, a pressure boundary leak was identified on the Reactor Coolant Pump (RCP) 1-2 1st seal cavity vent line upstream weld of 3/4 inch small bore pipe socketweld at a 90 degree elbow between the RCP pump and valve RC-407 (1st Seal Cavity Vent Isolation). The plant was in MODE 3 at Normal Operating Pressure and Normal Operating Temperature (NOP/NOT) for the inspections.
"The plant entered Technical Specification (TS) Limiting Condition for Operation (LCO) 3.4.13, 'RCS Operational Leakage,' Condition B and procedure DB-OP-02522. 'Small RCS Leaks,' abnormal operating procedure. Plant cooldown to comply with LCO 3.4.13, Condition B, Required Action B.2 is in progress. The cause and resolution are under evaluation.
"This event is reportable within 8 hours under 10CFR50.72(b)(3)(ii)(A).
"The NRC Resident Inspector has been notified. This condition has been documented in the Davis-Besse Corrective Action program as Condition Report 2012-09381."
The plant is required to be in MODE 5 within 36 hours. | |