| U.S. Nuclear Regulatory Commission
Operations Center Event Reports For
02/07/2014 - 02/10/2014 ** EVENT NUMBERS ** |
 | !!!!! THIS EVENT HAS BEEN RETRACTED. THIS EVENT HAS BEEN RETRACTED !!!!! | | Power Reactor | Event Number: 49727 | Facility: QUAD CITIES
Region: 3 State: IL
Unit: [ ] [2] [ ]
RX Type: [1] GE-3,[2] GE-3
NRC Notified By: JAMES BURKE
HQ OPS Officer: STEVE SANDIN | Notification Date: 01/15/2014
Notification Time: 15:47 [ET]
Event Date: 01/15/2014
Event Time: 07:23 [CST]
Last Update Date: 02/07/2014 | Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(3)(v)(C) - POT UNCNTRL RAD REL
| Person (Organization):
RICHARD SKOKOWSKI (R3DO)
|
| 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 DRYWELL RADIATION MONITOR INOPERABLE
"At 0723 [CST] hours on January 15, 2014, the 2A Drywell Radiation Monitor was declared inoperable after exhibiting erratic performance. This monitor provides the input into one division of the primary containment isolation logic for a Group II isolation. As a result, the channel was placed in a tripped condition in accordance with Technical Specification 3.3.6.1, Condition B.
"Given both divisions are required to complete the Group II isolation logic, this condition is reportable in accordance with 10 CFR 50.72(b)(3)(v)(C) as an event or condition that could have prevented the fulfillment of a safety function.
"The 2A Drywell Radiation Monitor has been repaired and restored to an operable condition.
"The NRC Resident Inspector has been notified."
* * * RETRACTION FROM JAMES BURKE TO JOHN SHOEMAKER AT 1447 EST ON 2/7/14 * * *
"The purpose of this notification is to retract the ENS notification made on January 15, 2014 (ENS 49727). Upon further investigation it was verified that the Drywell Radiation Monitors do not mitigate the consequences of an accident as discussed in Chapters 6 and 15 of the Updated Final Safety Analysis Report and the Technical Specifications Bases for 3.3.6.1. Therefore, the threshold for reporting the issue as an event or condition that could have prevented the fulfillment of a safety function was not met (NUREG 1022 Revision 3 - Event Report Guidelines Section 3.2.7).
"The NRC Resident Inspector has been notified."
Notified the R3DO (Orlikowski). |
| Agreement State | Event Number: 49782 | Rep Org: TEXAS DEPARTMENT OF HEALTH
Licensee: TEXAS GAMMA RAY LLC
Region: 4
City: PASADENA State: TX
County:
License #: 05561
Agreement: Y
Docket:
NRC Notified By: ART TUCKER
HQ OPS Officer: DANIEL MILLS | Notification Date: 01/31/2014
Notification Time: 15:02 [ET]
Event Date: 01/31/2014
Event Time: [CST]
Last Update Date: 01/31/2014 | Emergency Class: NON EMERGENCY
10 CFR Section:
AGREEMENT STATE
| Person (Organization):
NEIL OKEEFE (R4DO)
FSME EVENT RESOURCE (EMAI)
| Event Text AGREEMENT STATE REPORT - RADIOGRAPHY SOURCE COULD NOT BE RETRACTED
The following was received from the state of Texas via email:
"On January 31, 2014, the Agency [Texas Department of Health] was notified by the licensee that while performing radiography at a field location a radiography crew was unable to retract a 19 curie iridium - 192 source into a SPEC 150 exposure device. The radiographers were testing a pipe when a second pipe fell on the guide tube and crimped it enough that the source could not pass by it. The radiographers contacted their manager and informed them of the problem. Two individuals qualified to retrieve sources went to the location. They were able to reshape the guide tube enough to retract the source. The work was being conducted in a shielded shooting bay, therefore there was no exposure to any member of the general public. The highest exposure to anyone involved with the event was 60 millirem. No overexposure occurred. The guide tube was disposed of. The exposure device was tested with a new guide tube and functioned normally. The exposure device was returned to service. Additional information will be provided as it is received in accordance with SA-300."
Texas incident # I-9152 |
| Power Reactor | Event Number: 49803 | Facility: COOK
Region: 3 State: MI
Unit: [1] [2] [ ]
RX Type: [1] W-4-LP,[2] W-4-LP
NRC Notified By: KELLY BAKER
HQ OPS Officer: PETE SNYDER | Notification Date: 02/07/2014
Notification Time: 03:04 [ET]
Event Date: 02/07/2014
Event Time: 03:00 [EST]
Last Update Date: 02/07/2014 | Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(3)(xiii) - LOSS COMM/ASMT/RESPONSE
| Person (Organization):
ROBERT ORLIKOWSKI (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 UNAVAILABILIITY OF TSC VENTILATION SYSTEM DUE TO SCHEDULED MAINTENANCE
"At 03:00 EST on Friday, February 7, the Cook Nuclear Plant (CNP) Technical Support Center (TSC) air conditioning and charcoal filtration systems will be removed from service for scheduled maintenance.
"Under certain accident conditions the TSC may become unavailable due to the inability of the air conditioning and charcoal filtration systems to maintain a habitable atmosphere. Compensatory measures exist to relocate TSC personnel to the unaffected unit's control room if necessary.
"TSC ventilation system maintenance and post maintenance testing is scheduled to be completed by 16:00 EST on
Friday, February 7.
"The licensee has notified the NRC Resident Inspector.
"This notification is being made in accordance with 10 CFR 50.72 (b)(3)(xiii) due to the loss of an emergency response facility."
* * * UPDATE PROVIDED BY BRAD LEWIS TO JEFF ROTTON AT 1605 ON 02/07/2014 * * *
"TSC ventilation system was returned to service following successful maintenance and post maintenance testing at 14:30 EST on Friday, February 7, 2014. The licensee has notified the NRC Resident Inspector."
Notified R3DO (Orlikowski) |
| Power Reactor | Event Number: 49804 | Facility: PERRY
Region: 3 State: OH
Unit: [1] [ ] [ ]
RX Type: [1] GE-6
NRC Notified By: ROBERT KIDDER
HQ OPS Officer: PETE SNYDER | Notification Date: 02/07/2014
Notification Time: 08:19 [ET]
Event Date: 02/07/2014
Event Time: 06:00 [EST]
Last Update Date: 02/07/2014 | Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(2)(i) - PLANT S/D REQD BY TS
| Person (Organization):
ROBERT ORLIKOWSKI (R3DO)
|
| Unit | SCRAM Code | RX CRIT | Initial PWR | Initial RX Mode | Current PWR | Current RX Mode | | 1 | N | Y | 100 | Power Operation | 45 | Power Operation | Event Text PLANT SHUTDOWN REQUIRED BY TECHNICAL SPECIFICATIONS DUE TO CONTAINMENT ISOLATION VALVE OVER LEAK RATE LIMIT
"This event is being reported in accordance with 10CFR 50.72(b)(2)(i), 'Initiation of a Shutdown Required by Technical Specifications.'
"At 2043 hours [EST] on February 06, 2014, the Perry Nuclear Power Plant entered Technical Specification 3.6.1.3 Primary Containment Isolation Valves (PCIVs), action C.1, due to leakage identified during local leak rate testing of the containment penetration for the Containment and Drywell Purge system. Leakage was identified on the outboard containment isolation valve resulting in the plant exceeding the limit for secondary containment bypass leakage. The Containment and Drywell Purge system penetration is normally isolated and remains isolated in accordance with Technical Specifications. Action C.1 requires restoration of the leakage rate within four hours. At 0043 hours on February 7, 2014, the plant entered Technical Specification 3.6.1.3, 'Primary Containment Isolation Valves (PCIVs)', action E as the leakage rate was not restored. Action E requires the plant be in Mode 3 in 12 hours and Mode 4 in 36 hours.
"At 0600 hours on February 07, 2014, the Perry Nuclear Power Plant initiated a shutdown in accordance with Technical Specification 3.6.1.3, action E. Repairs to restore the penetration leakage to within allowable limits are in progress."
The NRC Resident Inspector has been notified.
* * * UPDATE PROVIDED BY DAVE ODONNELL TO JEFF ROTTON AT 1220 EST ON 02/07/2014 * * *
"At 0943 hours [EST] the reactor shutdown to comply with Technical Specification 3.6.1.3 action E was terminated [with the reactor at 42% power]. A blind flange was installed downstream of the outboard containment isolation valve. Local leak rate testing of the containment penetration for the Containment and Drywell Purge system verified that leakage was within the limits for secondary containment bypass leakage.
"The NRC Resident Inspector has been notified."
The licensee has commenced increasing reactor power.
Notified R3DO (Orlikowski) |
| Power Reactor | Event Number: 49806 | Facility: PRAIRIE ISLAND
Region: 3 State: MN
Unit: [1] [2] [ ]
RX Type: [1] W-2-LP,[2] W-2-LP
NRC Notified By: STEVE INGALLS
HQ OPS Officer: JOHN SHOEMAKER | Notification Date: 02/07/2014
Notification Time: 12:16 [ET]
Event Date: 02/07/2014
Event Time: 06:20 [CST]
Last Update Date: 02/07/2014 | Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(3)(xiii) - LOSS COMM/ASMT/RESPONSE
| Person (Organization):
ROBERT ORLIKOWSKI (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 SHIELD BUILDING VENT GAS RADIATION MONITOR FAILURE
"At 0620 CST on February 7, 2014, 1R-22 Shield Building Vent Gas Radiation Monitor failed and was declared nonfunctional. This monitor has no compensatory measure that will allow timely classification of two Emergency Action Levels (EALs); NUE (Notification of Unusual Event) and Alert classifications when out of service. It is also used for offsite dose projection calculations. This results in a Loss of Emergency Assessment Capability while 1R-22 is out of service. This is a reportable condition in accordance with 10 CFR 50.72(b)(3)(xiii).
"Unit 1 Shield Building Ventilation Stack is also monitored by high range monitor, 1R-50, which is used for the same purpose in Site Area or General Emergency classifications. 1R-50 is being monitored and is indicating normal values. There are no radioactive leaks that will impact the Shield Building as evidenced by normal readings on 1R-22 prior to its failure. Corrective maintenance is in progress and will continue until the monitor is
returned to service. Maintenance will not result in the unplanned release of radioactivity to the environment and will not adversely affect the safe operation of the plant or health and safety of the public.
"The licensee has notified the NRC Resident Inspector." |
| Power Reactor | Event Number: 49807 | Facility: PALO VERDE
Region: 4 State: AZ
Unit: [ ] [2] [ ]
RX Type: [1] CE,[2] CE,[3] CE
NRC Notified By: SEAN DORNSEIF
HQ OPS Officer: JOHN SHOEMAKER | Notification Date: 02/07/2014
Notification Time: 16:31 [ET]
Event Date: 12/11/2013
Event Time: 20:11 [MST]
Last Update Date: 02/07/2014 | Emergency Class: NON EMERGENCY
10 CFR Section:
50.73(a)(1) - INVALID SPECIF SYSTEM ACTUATION
| Person (Organization):
GREG WERNER (R4DO)
|
| Unit | SCRAM Code | RX CRIT | Initial PWR | Initial RX Mode | Current PWR | Current RX Mode | | 2 | N | N | 0 | Hot Shutdown | 0 | Hot Shutdown | Event Text 60 DAY REPORT FOR AN INVALID MAIN STEAM LINE ACTUATION
"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 10 CFR 50.73.
"This telephone notification is being made pursuant to the reporting requirements of 10 CFR 50.73(a)(2)(iv)(A) to describe an invalid actuation of the Palo Verde Nuclear Generating Station Unit 2 'A' train Main Steam and Feedwater Isolation System (MSFIS) logic.
"On December 11, 2013, at approximately 2011, Mountain Standard Time, Unit 2 was shut down in Mode 4 with Reactor Coolant System (RCS) temperature at approximately 425 degrees Fahrenheit and 1900 pounds per square inch absolute. While starting up the plant following completion of repairs to the 1A reactor coolant pump, an invalid actuation of the 'A' train MSFIS fast close logic occurred due to a random electrical failure in the 'A' train MSFIS logic cabinet. This resulted in the closure of all four Main Steam Isolation Valves and the two 'A' train Economizer Feed Water Isolation Valves. No other engineered safety feature actuations occurred and none were required. The invalid actuation was initiated by a ground fault in a 30-volt direct current power supply in the 'A' train MSFIS Logic cabinet. Repairs were completed on December 12, 2013.
"This was a complete actuation of 'A' train MSFIS equipment and all the affected equipment responded as designed. RCS temperature and pressure were maintained stable using atmospheric dump valves. Feed water to the steam generators was being provided by Auxiliary Feed Water and was not affected by the event.
"The NRC Resident Inspector has been notified." |
| Power Reactor | Event Number: 49808 | Facility: MONTICELLO
Region: 3 State: MN
Unit: [1] [ ] [ ]
RX Type: [1] GE-3
NRC Notified By: JEREMY TANNER
HQ OPS Officer: JEFF ROTTON | Notification Date: 02/07/2014
Notification Time: 17:52 [ET]
Event Date: 02/07/2014
Event Time: 10:05 [CST]
Last Update Date: 02/07/2014 | Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(3)(ii)(B) - UNANALYZED CONDITION
50.72(b)(3)(v)(D) - ACCIDENT MITIGATION
| Person (Organization):
ROBERT ORLIKOWSKI (R3DO)
|
| Unit | SCRAM Code | RX CRIT | Initial PWR | Initial RX Mode | Current PWR | Current RX Mode | | 1 | N | Y | 70 | Power Operation | 70 | Power Operation | Event Text SINGLE DRYWELL TO TORUS VACUUM BREAKER NOT GOING FULLY CLOSED DURING SURVEILLANCE TEST
"After cycling AO-2382A (Drywell to Torus Vacuum Breaker) for surveillance testing, it did not indicate fully closed. The procedure for this condition was entered and after cycling the valve several times, the vacuum breaker indicated full closed. During the approximately eight minutes that the indication showed that it was not closed, the Technical Specification Limiting Condition for Operation (LCO) requirement was not met. After validation that the vacuum breaker had opened as required, and was closed successfully, the safety function was restored. The health and safety of the public was not jeopardized as the plant was in a normal condition and an initiating event was not in progress.
"The USAR [Updated Safety Analysis Report] assumes all eight vacuum breakers to be closed. This condition therefore put the nuclear power plant in an unanalyzed condition and is reportable per 10CFR50.72(b)(3)(ii)(B).
"This condition, at time of discovery, could have prevented the fulfillment of the safety function of structures or systems that are needed to mitigate the consequences of an accident and is reportable per 10CFR50.72(b)(3)(v)(D).
The licensee notified the NRC Resident Inspector, the State of Minnesota Duty Officer, and the local counties. |
| Part 21 | Event Number: 49809 | Rep Org: WEIR VALVES AND CONTROLS, USA INC
Licensee: WEIR VALVES AND CONTROLS, USA INC
Region: 1
City: IPSWICH State: MA
County:
License #:
Agreement: Y
Docket:
NRC Notified By: ARTHUR BUTTERS
HQ OPS Officer: JEFF ROTTON | Notification Date: 02/07/2014
Notification Time: 19:13 [ET]
Event Date: 02/07/2014
Event Time: [EST]
Last Update Date: 02/07/2014 | Emergency Class: NON EMERGENCY
10 CFR Section:
21.21(a)(2) - INTERIM EVAL OF DEVIATION
| Person (Organization):
GLENN DENTEL (R1DO)
PART 21 REACTORS GRP (EMAI)
| Event Text PART 21 INTERIM REPORTING REGARDING TRICENTRIC TRIPLE OFFSET BUTTERFLY VALVES
The following is a summary of information provided via facsimile:
"Based on recent testing conducted on a 24 inch Class 150 TRICENTRIC [triple offset butterfly valve] to evaluate bearing coefficients of friction (COF), it has been determined that there exists an unseating load that has not been accounted for on our TRICENTRIC Triple Offset Product Line.
"Weir has not experienced a field failure of the product, but recent testing by a customer indicated anomalies that required review by Weir. The review of the unseating anomaly relates to two inputs; the inputted seating torque and increases in differential pressure across the disc after initial closure. Weir has developed a conservative approach for calculating this additional loading.
"Weir has reviewed data from customers for several recent orders, and has determined that there is a minimal risk of failure of an item to perform its safety function. However, based on the safety related functions of these items, Weir cannot provide absolute assurance of operability on items in service without confirming with the end user. Weir will be releasing an industry notice that will address our initial findings, continue with the detailed investigation, and advise customers concerning the recommended corrective actions."
Known customer - PSEG |
| Power Reactor | Event Number: 49810 | Facility: PEACH BOTTOM
Region: 1 State: PA
Unit: [2] [3] [ ]
RX Type: [2] GE-4,[3] GE-4
NRC Notified By: DAN MERCER
HQ OPS Officer: VINCE KLCO | Notification Date: 02/09/2014
Notification Time: 08:10 [ET]
Event Date: 02/09/2014
Event Time: 01:00 [EST]
Last Update Date: 02/09/2014 | Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(3)(xiii) - LOSS COMM/ASMT/RESPONSE
| Person (Organization):
GLENN DENTEL (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 | | 3 | N | Y | 100 | Power Operation | 100 | Power Operation | Event Text LOSS OF OFFSITE COMMUNICATIONS CAPABILITY
"On 02/09/2014, at 0100 [EST], a major portion of Peach Bottom Atomic Power Station's [PBAPS] offsite communication capability was discovered to be non-functional due to an offsite communications equipment failure outside the control of PBAPS. Peach Bottom continues to investigate and resolve the cause of this loss of offsite communications. Peach Bottom Main Control Room ENS and offsite communication lines have been restored and are currently available. The Technical Support Center is limited to satellite phones for offsite communications. On-site communications were not affected.
"This report is being submitted pursuant to 10CFR 50.72(b)(3)(xiii) as a result of a major loss of offsite communications capability.
"The NRC [Resident Inspector] has been informed of this notification." | |