| U.S. Nuclear Regulatory Commission
Operations Center Event Reports For
07/27/2015 - 07/28/2015 ** EVENT NUMBERS ** |
 | Agreement State | Event Number: 51183 | Rep Org: NE DIV OF RADIOACTIVE MATERIALS
Licensee: GREGG YOUNG CHEVROLET
Region: 4
City: OMAHA State: NE
County:
License #: NE GL-0488
Agreement: Y
Docket:
NRC Notified By: JULIA SCHMITT
HQ OPS Officer: MARK ABRAMOVITZ | Notification Date: 06/26/2015
Notification Time: 10:28 [ET]
Event Date: 12/31/2008
Event Time: [CDT]
Last Update Date: 07/27/2015 | Emergency Class: NON EMERGENCY
10 CFR Section:
AGREEMENT STATE
| Person (Organization):
VIVIAN CAMPBELL (R4DO)
NMSS_EVENTS_NOTIFICA (EMAI)
BARRY WRAY (ILTA)
|
| This material event contains a "Less than Cat 3 " level of radioactive material. | Event Text AGREEMENT STATE REPORT - MISSING TRITIUM EXIT SIGNS
The following was received from the State of Nebraska via email:
"On June 26, 2015, Gregg Young Chevrolet (GL0488) reported a loss of 10 exit signs containing tritium (Evenlite 201) via letter. Each sign originally contained 10.5 curies and current approximate activity is 3.852 curies. Our records show that the registrant possessed these signs in 2013 (latest inventory) however, the registrant assumes these signs were removed from their facility by a tenant sometime in 2008."
* * * UPDATE FROM HOWARD SHUMAN TO VINCE KLCO ON 7/27/2015 AT 0916 * * *
The following information was received from the State of Nebraska via email:
"No additional information has been submitted. The NMED event is closed."
Notified the R4DO (Warnick), ILTAB (Johnson) and NMSS Events Notification via email.
NE Item Number: NE150002
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 |
| Power Reactor | Event Number: 51265 | Facility: SEQUOYAH
Region: 2 State: TN
Unit: [1] [ ] [ ]
RX Type: [1] W-4-LP,[2] W-4-LP
NRC Notified By: KEVIN MICHAEL
HQ OPS Officer: JOHN SHOEMAKER | Notification Date: 07/27/2015
Notification Time: 13:44 [ET]
Event Date: 07/27/2015
Event Time: 10:43 [EDT]
Last Update Date: 07/27/2015 | 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):
RANDY MUSSER (R2DO)
|
| Unit | SCRAM Code | RX CRIT | Initial PWR | Initial RX Mode | Current PWR | Current RX Mode | | 1 | A/R | Y | 82 | Power Operation | 0 | Hot Standby | Event Text AUTOMATIC REACTOR/TURBINE TRIP
"At 1043 EDT on 7/27/2015, Sequoyah Unit 1 was at 82% power and continuing to perform a startup when the reactor/turbine automatically tripped.
"Following the reactor trip, all safety-related equipment operated as designed. Auxiliary Feedwater automatically initiated as expected from the Feedwater Isolation Signal.
"Unit 1 is currently being maintained in Mode 3, at NOT/NOP [normal operating temperature and normal operating pressure], approximately 545 F and 2235 psig, with auxiliary feedwater supplying the steam generators and decay heat removal via the condenser steam dumps.
"The immediate cause of the trip was an electrically-induced turbine trip. Due to fluctuating voltage the main generator voltage regulator was taken to manual; immediately after this the unit tripped.
"Current Temperature and Pressure - Reactor Coolant System (RCS) temperature is 545 degrees F and stable and RCS pressure is 2235 psig and stable. There is no indication of any primary/secondary leakage. All rods fully inserted on the reactor trip and remain inserted. The electrical alignment is normal with shutdown power supplied from off-site power. There is no operational impact to Unit 2. Unit 2 continues to operate in Mode 1 at 100%.
"There was no impact on public health and safety. Post-trip investigation is in progress and planned restart timeline has not yet been determined.
"The licensee notified the NRC Resident Inspector." |
| Power Reactor | Event Number: 51266 | Facility: PRAIRIE ISLAND
Region: 3 State: MN
Unit: [1] [2] [ ]
RX Type: [1] W-2-LP,[2] W-2-LP
NRC Notified By: JEFFREY HUMAN
HQ OPS Officer: JEFF ROTTON | Notification Date: 07/27/2015
Notification Time: 14:28 [ET]
Event Date: 07/27/2015
Event Time: 09:02 [CDT]
Last Update Date: 07/27/2015 | Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(3)(xiii) - LOSS COMM/ASMT/RESPONSE
| Person (Organization):
CHRISTINE LIPA (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 PLANNED OUTAGE FOR PLANT COMPUTER SYSTEMS
"On July 27, 2015 at 0902 [CDT], [the site commenced] a planned outage of the Emergency Response Data System (ERDS) and Safety Parameter Display System (SPDS), referred to as 'plant computer'. The unavailability of ERDS and SPDS could significantly affect the site's ability to respond to an emergency if one were to occur. During this time, Operations will be utilizing the site's procedures 1C1.5 and 2C1.5, 'OPERATION WITHOUT COMPUTER', which requires additional operators for monitoring of equipment affected by the loss of the plant computer. Additionally, as this is a planned outage, the work week schedule has been modified to ensure limited interactions required by Operations during this time frame. The site expects ERDS and SPDS to be operational 1200 July 28, 2015.
"This event is reportable under 10 CFR 50.72(b)(3)(xiii), Any event that results in a major loss of emergency assessment capability, offsite response capability, or offsite communications capability (e.g., significant portion of Control Room indication, Emergency Notification System (ENS), or Offsite Notification System). The ENS and Offsite Notification System are not affected by this planned outage.
"The health and safety of the public are not impacted by this planned outage.
"The NRC Resident Inspector has been informed." |
| Power Reactor | Event Number: 51267 | Facility: PILGRIM
Region: 1 State: MA
Unit: [1] [ ] [ ]
RX Type: [1] GE-3
NRC Notified By: GRANT FLYNN
HQ OPS Officer: JOHN SHOEMAKER | Notification Date: 07/27/2015
Notification Time: 16:05 [ET]
Event Date: 07/27/2015
Event Time: 12:29 [EDT]
Last Update Date: 07/27/2015 | Emergency Class: NON EMERGENCY
10 CFR Section:
26.719 - FITNESS FOR DUTY
| Person (Organization):
RAY MCKINLEY (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 FITNESS FOR DUTY REPORT INVOLVING A LICENSED EMPLOYEE
"A licensed employee violated the site Fitness-for-Duty (FFD) policy. The employee's plant access has been denied.
"The NRC Resident Inspector has been informed." |
| Power Reactor | Event Number: 51268 | Facility: OCONEE
Region: 2 State: SC
Unit: [ ] [2] [ ]
RX Type: [1] B&W-L-LP,[2] B&W-L-LP,[3] B&W-L-LP
NRC Notified By: BYRON LECROY
HQ OPS Officer: JOHN SHOEMAKER | Notification Date: 07/27/2015
Notification Time: 16:31 [ET]
Event Date: 07/27/2015
Event Time: 09:56 [EDT]
Last Update Date: 07/27/2015 | Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(3)(iv)(A) - VALID SPECIF SYS ACTUATION
| Person (Organization):
RANDY MUSSER (R2DO)
|
| Unit | SCRAM Code | RX CRIT | Initial PWR | Initial RX Mode | Current PWR | Current RX Mode | | 2 | N | Y | 17 | Power Operation | 16 | Power Operation | Event Text VALID ACTUATION OF UNIT 2 EMERGENCY FEEDWATER SYSTEM DURING STARTUP
"At approximately 0956 EDT on July 27, 2015, Oconee Nuclear Station Unit 2 experienced a valid actuation of the Emergency Feedwater System [EFW]. At the time of the event, Unit 2 was in Mode 1 at approximately 17% power and increasing with preparations in progress for placing the main turbine on line during a unit startup. The [EFW] actuation was due to a low level on the 2B steam generator, which resulted from failure of 2B Main Feedwater Block Valve 2FDW-40 to automatically open upon demand. All systems operated as expected with no problems observed. Unit 2 is currently stable at approximately 16% power while troubleshooting valve 2FDW-40 [and the 2B Steam Generator level stable at the normal operating level]. Units 1 and 3 were unaffected and remain on line and stable at 100% power. Public health and safety were not impacted by this event.
"This event is being reported as an 8 hour non-emergency in accordance with 10 CPR 50.72(b)(3)(iv) for a valid actuation of the Emergency Feedwater System.
"The NRC Resident Inspector has been notified.
"Corrective Action: Troubleshooting of valve 2FDW-40 is on-going." |
| Power Reactor | Event Number: 51269 | Facility: SUSQUEHANNA
Region: 1 State: PA
Unit: [1] [2] [ ]
RX Type: [1] GE-4,[2] GE-4
NRC Notified By: JASON WILLIS
HQ OPS Officer: JEFF ROTTON | Notification Date: 07/27/2015
Notification Time: 17:50 [ET]
Event Date: 07/27/2015
Event Time: 11:18 [EDT]
Last Update Date: 07/27/2015 | Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(3)(v)(C) - POT UNCNTRL RAD REL
| Person (Organization):
RAY MCKINLEY (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 | | 2 | N | Y | 100 | Power Operation | 100 | Power Operation | Event Text LOSS OF SECONDARY CONTAINMENT DUE TO OPENING SINGLE DOOR WITHOUT PROPER AUTHORIZATION
"On July 27, 2015 at 1118 [EDT], Secondary Containment became inoperable requiring a Technical Specification 3.6.4.1 entry for failure to meet SR [Surveillance Requirement] 3.6.4.1.1 on Unit 1 and Unit 2.
"The inoperability was caused by Zone 2 differential pressure lowering to less than 0.25 inches WC when a secondary containment door was opened without appropriate authorization.
"The secondary containment door was closed at 1149 and secondary containment D/P verified greater than 0.25 inches WC at 1205.
"This event is being reported under 10 CFR 50.72(b)(3)(v)(c) and per the guidance of NUREG 1022 Rev. 3 section 3.2.7 as a loss of a Safety Function. There is no redundant Susquehanna Secondary Containment system."
The loss of secondary containment occurred due to multiple openings of Door 104R which provides access to area of the building that provides alternate access to the building roof, but this door is not the normal access to the building roof
The licensee notified the NRC Resident Inspector. |
| Power Reactor | Event Number: 51270 | Facility: FORT CALHOUN
Region: 4 State: NE
Unit: [1] [ ] [ ]
RX Type: (1) CE
NRC Notified By: JULIE BISSEN
HQ OPS Officer: JOHN SHOEMAKER | Notification Date: 07/27/2015
Notification Time: 18:01 [ET]
Event Date: 07/22/2015
Event Time: 13:30 [CDT]
Last Update Date: 07/27/2015 | Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(3)(ii)(A) - DEGRADED CONDITION
| Person (Organization):
GREG WARNICK (R4DO)
|
| Unit | SCRAM Code | RX CRIT | Initial PWR | Initial RX Mode | Current PWR | Current RX Mode | | 1 | N | N | 0 | Hot Shutdown | 0 | Hot Shutdown | Event Text DEGRADED CONDITION DUE TO REACTOR COOLANT SYSTEM LEAK
"On July 8, 2015, Fort Calhoun Station was in Mode 1, 100% when personnel identified an increase in the Reactor Coolant System unidentified leakage rate. As a result, personnel performed a containment entry and identified the source coming from the Reactor Coolant Pump 3A seal area. Based on this observation, a monitoring plan was established and on July 20, 2015 the leak rate exceeded the pre-establish leak limit and operators manually shutdown the reactor. On July 22, 2015, at approximately 1330 CDT, personnel identified the source of the seal leak as a crack on the middle seal inlet line, which is part of the reactor coolant system boundary. Maintenance personnel have since repaired the seal line and it has passed post-maintenance testing.
"The 8-hour verbal report is being made post event due to additional review of the leakage condition identifying the leakage constituted a degraded condition due to be material defects in the primary coolant system."
The leak rate was between 1 and 2 gpm.
The licensee notified the NRC Resident Inspector. |
| Power Reactor | Event Number: 51272 | Facility: QUAD CITIES
Region: 3 State: IL
Unit: [1] [2] [ ]
RX Type: [1] GE-3,[2] GE-3
NRC Notified By: NATE CLEVELAND
HQ OPS Officer: MARK ABRAMOVITZ | Notification Date: 07/28/2015
Notification Time: 01:08 [ET]
Event Date: 07/27/2015
Event Time: 17:30 [CDT]
Last Update Date: 07/28/2015 | Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(3)(v)(D) - ACCIDENT MITIGATION
50.72(b)(3)(xiii) - LOSS COMM/ASMT/RESPONSE
| Person (Organization):
CHRISTINE LIPA (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 CONTROL ROOM EMERGENCY VENTILATION SYSTEM INOPERABLE
"On July 27, 2015, at 1730 hours [CDT], the Control Room Emergency Ventilation (CREV) system was declared inoperable due to the 'B' Air Filtration Unit (AFU) Booster Fan discharge damper stuck open in mid-position. In this condition, the CREV system cannot be guaranteed to achieve required design flow rate. As a result, Technical Specification 3.7.4, Condition A, was entered.
"The CREV system maintains a habitable control room environment and ensures the operability of components in the control room emergency zone during accident conditions as well as protection of the operators from a high dose environment assumed during a design basis accident.
"This notification is being made in accordance with 10 CFR 50.72(b)(3)(v)(D) because the CREV system is a single train system, and loss of the CREV system could impact the plant's ability to mitigate the consequences of an accident as stated in Chapter 6 of the UFSAR [Updated Final Safety Analysis Report]. This event is also reportable under 10 CFR 50.72(b)(3)(xiii) since this condition also impacts the control room as an Emergency Response Facility.
"The NRC Resident Inspector has been notified."
Both units are in a seven day technical specification for troubleshooting and repairs. If the control room became uninhabitable, procedure "Complete Loss of Control Room HVAC" would be entered. | |