| U.S. Nuclear Regulatory Commission
Operations Center Event Reports For
11/14/2013 - 11/15/2013 ** EVENT NUMBERS ** |
 | !!!!! THIS EVENT HAS BEEN RETRACTED. THIS EVENT HAS BEEN RETRACTED !!!!! | | Power Reactor | Event Number: 49495 | Facility: PRAIRIE ISLAND
Region: 3 State: MN
Unit: [1] [ ] [ ]
RX Type: [1] W-2-LP,[2] W-2-LP
NRC Notified By: STEPHEN SEILHYMER
HQ OPS Officer: HOWIE CROUCH | Notification Date: 11/01/2013
Notification Time: 17:31 [ET]
Event Date: 11/01/2013
Event Time: 12:06 [CDT]
Last Update Date: 11/14/2013 | Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(3)(v)(C) - POT UNCNTRL RAD REL
| Person (Organization):
ROBERT DALEY (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 | Event Text BOTH TRAINS OF AUXILIARY BUILDING SPECIAL VENTILATION SYSTEM DECLARED INOPERABLE
"During Unit 2 refueling outage [currently defueled] preventative maintenance on CV-31117, Loop B Main Steam Isolation Valve (MSIV), an opening in the valve was discovered without direct administrative controls to ensure the opening could be closed within 6 minutes following a Loss of Coolant Accident on Unit 1. Addition of this opening to other openings created a total of greater than 10 square feet of non-closable openings and required declaring the Auxiliary Building Special Ventilation System (ABSVS) boundary inoperable. The inoperable ABSVS boundary caused both trains of ABSVS to be declared inoperable and required entry into Technical Specification (TS) 3.7.12, Condition B. This could have prevented the ability to control the release of radioactive material and is considered a potential loss of safety function per 10CFR50.72(b)(3)(v)(C).
"Administrative controls in the ABSVS Boundary were re-established by installing a closure device in the opening on CV-31117 at 1246 CDT. The ABSVS boundary and both trains of ABSVS were declared operable at that time."
During the time that the ventilation system was out of service, no evolutions were in progress that could have resulted in an unmonitored release.
* * * RETRACTION ON 11/14/13 AT 1648 EST FROM STEPHEN SEILHYMER TO NESTOR MAKRIS * * *
"After further evaluation, the non-closable boundary openings for the Auxiliary Building Special Ventilation Zone (ABSVZ) including CV-31117, Loop B Main Steam Isolation Valve (MSIV), at the time of discovery is calculated to be 9.48 square feet. This is below the 10 square feet required in TS 3.7.12, thus both trains of ABSVS were operable and no loss of safety function existed.
"The licensee has notified the NRC Resident Inspector."
Notified the R3DO (Cameron). |
| Agreement State | Event Number: 49508 | Rep Org: NC DIV OF RADIATION PROTECTION
Licensee: S&ME - GREENSBORO OFFICE
Region: 1
City: GREENSBORO State: NC
County:
License #: 0922-3
Agreement: Y
Docket:
NRC Notified By: CHRIS FIDALGO
HQ OPS Officer: PETE SNYDER | Notification Date: 11/06/2013
Notification Time: 11:08 [ET]
Event Date: 11/06/2013
Event Time: 08:30 [EST]
Last Update Date: 11/06/2013 | Emergency Class: NON EMERGENCY
10 CFR Section:
AGREEMENT STATE
| Person (Organization):
JAMES NOGGLE (R1DO)
FSME EVENTS RESOURCE (EMAI)
| Event Text AGREEMENT STATE REPORT - DAMAGED MOISTURE DENSITY GAUGE
The following Agreement State Report was received via facsimile:
"The licensee contacted [the NC Radiation Protection] agency at approximately 0930 EST with the following info: A dump truck at a construction site ran over an lnstroTek Model 3500 Xplorer moisture density gauge. The area of the accident was cordoned off and the licensee RSO [was sent to] report back to agency with survey readings and leak test results. The gauge sources contain 11 mCi of Cs-137 and 44 mCi of Am-241/Be.
"A survey by the licensee [RSO] showed a maximum reading of 0.5 mrem/hr. The instrument make/model was of the survey meter was not specified. Sources appear to be intact. The gauge and debris were returned to the gauge case (doubles as shipping container- DOT 7A Yellow II). The gauge will be shipped to lnstroTek corporation for repair or disposition."
NC State Report Number: ICD 13-20 |
| Agreement State | Event Number: 49513 | Rep Org: PA BUREAU OF RADIATION PROTECTION
Licensee: UNKNOWN
Region: 1
City: COATESVILLE State: PA
County:
License #:
Agreement: Y
Docket:
NRC Notified By: JOSEPH M. MELNIC
HQ OPS Officer: DONG HWA PARK | Notification Date: 11/07/2013
Notification Time: 14:40 [ET]
Event Date: 10/31/2013
Event Time: [EST]
Last Update Date: 11/07/2013 | Emergency Class: NON EMERGENCY
10 CFR Section:
AGREEMENT STATE
| Person (Organization):
JAMES NOGGLE (R1DO)
FSME EVENTS RESOURCE (EMAI)
| Event Text ORPHAN SEALED SOURCE RECOVERED AT A SCRAP YARD
The following Agreement State Report was received via facsimile:
"Notifications: The Southeast Regional (SER) Office was contacted by Coatesville Scrap on October 31, 2013 regarding a radiation alarm on an outbound trailer load. The event is reportable within 24 hours per 10 CFR 20.2201(a)(1)(i).
"Event Description: On Thursday, October 31, 2013, a radioactive source was detected when a radiation alarm sounded on an outbound trailer load. The alarm was caused by a small metal object. It was located, placed in a lead pipe, crimped and moved to a remote area of the scrapyard behind a large steel piece of equipment. A SER inspector was sent to the site to determine the isotope and activity. It was identified as cesium-137 and a measurement on the surface of the lead pipe was 320 mR/hr. On Wednesday, November 6th the activity was estimated at approximate 10 mCi. It was determined that three members of the public were involved in locating the source, however one had the longest contact with the source; approximately 2.5 hours on and off. It took site staff some time to find the material via shoveling through the solid material and metal of the load. When found, the radioactive source was carried by a shovel approximately 150 to 200 feet away. An individual then carried the source by hand to an adjoining property and placed it in a lead pipe, which was crimped. This took approximately 10 to 15 minutes. A whole body dose documented from the uncalibrated dosimetry that was being worn by this individual was 364 micro-roentgens (microR). SER staff estimate a possible 10 rad dose to this individual's hand. Hands and feet of all parties were surveyed, with no contamination found. Caution tape was used to create a boundary to help notify other employees to stay away from the area. Meter readings at the caution tape boundary were approximately 60 microR/hr.
"CAUSE OF THE EVENT: Loss of control of a Cs-137 sealed source.
"ACTIONS: The DEP [Department of Environmental Protection] plans a full investigation of this event. The scrap yard has hired a consultant health physicist to assist with this event and Cs-137 source. The DEP will recommend that the individual who handled the source have their hands photographed and be examined by a physician."
Event Report ID No: PA130026 |
| Power Reactor | Event Number: 49535 | Facility: DRESDEN
Region: 3 State: IL
Unit: [ ] [2] [ ]
RX Type: [1] GE-1,[2] GE-3,[3] GE-3
NRC Notified By: KYLE KOLLER
HQ OPS Officer: JOHN SHOEMAKER | Notification Date: 11/14/2013
Notification Time: 11:01 [ET]
Event Date: 11/14/2013
Event Time: 03:00 [CST]
Last Update Date: 11/14/2013 | Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(3)(ii)(A) - DEGRADED CONDITION
| Person (Organization):
JAMNES CAMERON (R3DO)
|
| Unit | SCRAM Code | RX CRIT | Initial PWR | Initial RX Mode | Current PWR | Current RX Mode | | 2 | N | N | 0 | Refueling | 0 | Refueling | Event Text PRIMARY CONTAINMENT ISOLATION VALVE LEAK TEST FAILURE
"On November 14, 2013, both the 2-220-58B Feed Water Inboard Check Valve and the 2-220-62B Feed Water Outboard Check Valve failed Local Leak Rate Testing (LLRT) acceptance criteria. These valves are considered primary containment isolation valves and, as such, are required to ensure that an adequate primary containment boundary is maintained.
"Technical Specification (TS) 5.5.12, 'Primary Containment Leakage Rate Testing Program,' establishes limits for Primary Containment leakage. Based upon the results of the LLRT, Dresden, Unit 2, may not have met the limits for primary containment leakage during the last operating cycle as specified in TS 5.5.12.c.
"Dresden Unit 2 is currently in Mode 5 for a refueling outage and per Dresden TS 3.6.1.1, 'Primary Containment,' Primary Containment is not required in the current mode of operation (i.e., Mode 5). However, in accordance with 10 CFR 50.72(b)(3)(ii)(A), this event is reportable as a condition that resulted in a principal safety barrier being seriously degraded.
"The NRC Resident Inspector has been notified [by the licensee]." |
| Power Reactor | Event Number: 49536 | Facility: SAINT LUCIE
Region: 2 State: FL
Unit: [ ] [2] [ ]
RX Type: [1] CE,[2] CE
NRC Notified By: ANDREW TEREZAKIS
HQ OPS Officer: NESTOR MAKRIS | Notification Date: 11/14/2013
Notification Time: 14:57 [ET]
Event Date: 11/14/2013
Event Time: 12:18 [EST]
Last Update Date: 11/14/2013 | 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):
BINOY DESAI (R2DO)
|
| 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 REACTOR TRIP DUE TO LOW LEVEL IN THE 2B STEAM GENERATOR
"On November 14, 2013 at 1218 EST, Unit 2 was manually tripped due to a lowering 2B Steam Generator level caused by the spurious closure of 2B Main Feedwater Isolation Valve HCV-09-2A.
"All CEAs [Control Element Assemblies] fully inserted into the core. All safety systems responded as expected with the 2B Train Auxiliary Feedwater Actuation System Channel 2 (AFAS 2) actuating on low 2B Steam Generator level. Decay Heat Removal is from Main Feedwater to the 2A Steam Generator and Auxiliary Feedwater to the 2B Steam Generator with Steam Bypass to the Main Condenser.
"This event is reportable pursuant to 10CFR 50.72(b)(2)(iv)(B) for the Reactor Trip and 10CFR 50.72(b)(3)(iv)(A) for the AFAS 2 actuation."
The plant is in its normal shutdown electrical lineup. No safeties or relief valves lifted during this event.
The NRC Resident Inspector has been notified by the licensee. |
| Power Reactor | Event Number: 49538 | Facility: MCGUIRE
Region: 2 State: NC
Unit: [1] [ ] [ ]
RX Type: [1] W-4-LP,[2] W-4-LP
NRC Notified By: SCOTT FORTIN
HQ OPS Officer: NESTOR MAKRIS | Notification Date: 11/14/2013
Notification Time: 16:03 [ET]
Event Date: 11/14/2013
Event Time: 13:13 [EST]
Last Update Date: 11/14/2013 | 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):
BINOY DESAI (R2DO)
|
| Unit | SCRAM Code | RX CRIT | Initial PWR | Initial RX Mode | Current PWR | Current RX Mode | | 1 | M/R | Y | 100 | Power Operation | 0 | Hot Standby | Event Text MANUAL REACTOR TRIP DUE TO INDICATION OF DROPPED CONTROL RODS
"On November 14, 2013, at 1313 Eastern Standard Time, Unit 1 was manually tripped from 100% power due to indications of [four] dropped control rods. This manual reactor protection system actuation is reportable per 10 CFR 50.72(b)(2)(iv)(B). The cause of the dropped rods is not confirmed at this time, but may be related to maintenance in a Rod Control Power Cabinet ongoing at the time of the event.
"The Operations crew entered the reactor trip procedure and stabilized Unit 1 in Mode 3 at normal operating temperature and pressure. All control rods fully inserted into the core following the reactor trip and all plant systems operated as designed.
"The Auxiliary Feedwater (AFW) system [1A and 1B motor-driven pumps] was manually started for steam generator level control following reactor trip. The start of the AFW system is reportable per 10 CFR 50.72 (b)(3)(iv)(A) for a valid system actuation.
"Decay heat is being removed via the steam generators [via steam dumps to the main condenser]. This event does not impact public health and safety.
"Unit 2 was not affected by this event.
"The licensee notified the NRC Resident Inspector." |
| Power Reactor | Event Number: 49539 | Facility: SALEM
Region: 1 State: NJ
Unit: [1] [2] [ ]
RX Type: [1] W-4-LP,[2] W-4-LP
NRC Notified By: KELLY JOHNSON
HQ OPS Officer: BILL HUFFMAN | Notification Date: 11/14/2013
Notification Time: 16:08 [ET]
Event Date: 11/14/2013
Event Time: 13:22 [EST]
Last Update Date: 11/14/2013 | Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(3)(ii)(B) - UNANALYZED CONDITION
| Person (Organization):
JON LILLIENDAHL (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 POSTULATED HOT SHORT FIRE EVENT THAT COULD ADVERSELY IMPACT SAFE SHUTDOWN EQUIPMENT
"A review of industry operating experience regarding the impact of unfused Direct Current (DC) ammeter circuits has determined the described condition is applicable to the Salem Nuclear Power Plant resulting in an unanalyzed condition with respect to 10 CFR 50 Appendix R. The original plant wiring design and associated analysis for ammeters associated with the station batteries are not provided with overcurrent protection features to limit the fault current.
"A postulated fire that results in a short to ground concurrent with an opposite polarity short from the same battery could result in excessive current flow (i.e., heating) in the ammeter wiring. This excessive current could result in a secondary fire in another fire area. The secondary fire could adversely affect safe shutdown equipment and cause loss of the ability to conduct a safe shutdown as required by 10 CFR 50 Appendix R. The areas affected are the Control Room, Relay Rooms and 460 Volt Switchgear Rooms.
"This condition is being reported in accordance with 10 CFR 50.72(b)(3)(ii)(B). Interim compensatory measures (i.e., fire watches) have been implemented for the affected areas of the plant.
"The licensee has notified the NRC Resident Inspector." |
| Power Reactor | Event Number: 49541 | Facility: WATTS BAR
Region: 2 State: TN
Unit: [1] [ ] [ ]
RX Type: [1] W-4-LP,[2] W-4-LP
NRC Notified By: CHARLES BROESCHE
HQ OPS Officer: BILL HUFFMAN | Notification Date: 11/14/2013
Notification Time: 20:35 [ET]
Event Date: 11/14/2013
Event Time: 16:00 [EST]
Last Update Date: 11/14/2013 | Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(3)(ii)(B) - UNANALYZED CONDITION
50.72(b)(3)(v)(A) - POT UNABLE TO SAFE SD
50.72(b)(3)(v)(C) - POT UNCNTRL RAD REL
50.72(b)(3)(v)(D) - ACCIDENT MITIGATION
| Person (Organization):
BINOY DESAI (R2DO)
|
| 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 POSTULATED FIRE INDUCED FAILURE OF CENTRIFUGAL CHARGING PUMPS
"During analysis of Watts Bar Nuclear (WBN) Unit 2 fire protection features, it was revealed that a potential fire induced failure of centrifugal charging pumps could occur in Unit 1.
"Specifically, a potential fire induced failure of both Unit 1 Chemical and Volume Control System centrifugal charging pumps (CCPs) (1-PMP-62-108-A and 1-PMP-62-104-B) could occur due a fire in either auxiliary building room 737.0-A1 (general area for elevation 737.0) or 757.0-A2 (6.9 kV and Shutdown Board Room A). It is postulated that a fire in these rooms could cause a spurious closure of the CCP suction valve (1-LCV-62-133-B) from the volume control tank (VCT) (1-TANK-62-129) and could disable the control circuit which opens the flow from the refueling water storage tank (RWST) suction valve (1-LCV-62-135-A).
"The fire safe shutdown analysis (Fire Protection Report, Part VI) currently addresses this occurrence via the performance of a prompt main control room operator action to open the RWST suction path. However, this procedurally directed action may require several minutes to complete and due to the potentially short duration (possibly as short as a few seconds) for CCP survivability without suction flow, the action has now been determined to be unacceptable. As a result, the loss of charging flow could result in a loss of injection to the reactor coolant pump (RCP) seals which could subsequently lead to a RCP seal failure and a small break loss of coolant event. WBN engineering is continuing to validate whether the CCP minimum flow recirculation would protect the pumps with both suction paths (VCT and RWST) isolated and with the reactor at normal operating pressure.
"WBN has established compensatory measures to ensure that a fire in affected rooms will not cause a spurious closure of the CCP suctions valves."
The licensee has notified the NRC Resident Inspector. | |