| U.S. Nuclear Regulatory Commission
Operations Center Event Reports For
09/25/2014 - 09/26/2014 ** EVENT NUMBERS ** |
 | Agreement State | Event Number: 50466 | Rep Org: ILLINOIS EMERGENCY MGMT. AGENCY
Licensee: PDV MIDWEST REFINING
Region: 3
City: LEMONT State: IL
County:
License #: IL-01603-01
Agreement: Y
Docket:
NRC Notified By: DAREN PERRERO
HQ OPS Officer: MARK ABRAMOVITZ | Notification Date: 09/18/2014
Notification Time: 10:35 [ET]
Event Date: 09/16/2014
Event Time: [CDT]
Last Update Date: 09/18/2014 | Emergency Class: NON EMERGENCY
10 CFR Section:
AGREEMENT STATE
| Person (Organization):
BILLY DICKSON (R3DO)
FSME EVENTS RESOURCE (E-MA)
| Event Text AGREEMENT STATE REPORT - FIXED GAUGE WITH A STUCK OPEN SHUTTER
The following was received via email:
"The licensee's radiation safety officer called the Agency [Illinois Emergency Management Agency] to advise that on 9/16/14, an Ohmart Vega, m/n SH-F1-45 fixed gauge containing 15 milliCi of Cs-137 (as of April 2002) had a shutter which could not be closed. They became aware of the situation during the performance of routine 6 month operability checks. The manufacturer was immediately contacted to schedule an evaluation in the coming weeks and for advice. The licensee was advised to lubricate the shutter pinion and allow it to sit overnight before attempting to close the shutter again the next day. The licensee intends to have the manufacturer's field engineer come on site to inspect the device.
"The gauge is located on a vessel which is still in active use for sulfur production and is exposed to ambient weather conditions. The gauge is located approximately 15 feet overhead on a platform which is not a routine work location for day to day operations. A lock out/tag out procedure is actively in use at the site should any work on/in the vessel be necessary before repairs can be affected. Notifications to the acting RSO are part of that procedure. The licensee is aware of the 30 [day] written reporting requirement."
This is the second time that the shutter has stuck open recently.
Illinois Event #: IL14019 |
| Agreement State | Event Number: 50469 | Rep Org: COLORADO DEPT OF HEALTH
Licensee: ANALYTICAL INSTRUMENT RECYCLE, INC.
Region: 4
City: GOLDEN State: CO
County:
License #: 974-01
Agreement: Y
Docket:
NRC Notified By: JAMES JARVIS
HQ OPS Officer: HOWIE CROUCH | Notification Date: 09/18/2014
Notification Time: 14:24 [ET]
Event Date: 02/01/2014
Event Time: [MDT]
Last Update Date: 09/18/2014 | Emergency Class: NON EMERGENCY
10 CFR Section:
AGREEMENT STATE
| Person (Organization):
MICHAEL VASQUEZ (R4DO)
FSME EVENTS RESOURCE (EMAI)
| Event Text AGREEMENT STATE REPORT - DISCOVERY OF THREE LEAKING SOURCES
The following information was received from the state of Colorado via email:
"The following is a notification not previously identified to NRC.
"On or about April 24, 2014, during licensing renewal activities, the [Colorado Radiation Control Program] Agency licensing staff received information that Colorado specific licensee Analytical Instrument Recycle, Inc. (AIR, 15860 West 6th Ave., Golden, CO 80401; CO License #974-01) had reportedly discovered that three generally licensed (GL) electron capture devices (ECDs) in their possession that had failed a leak test approximately two months prior (in ~February 2014). (The licensee refurbishes and resells analytical instruments containing ECDs, although the licensee performs leak tests, it is not authorized to work on or open ECDs or the sources contained within them.)
"The licensee's Radiation Safety Officer (RSO) indicated that he had previously sent notification regarding the leaking sources to the Agency (Colorado Department of Public Health and Environment) in early February, 2014 and provided a copy of an undated letter during the investigation.
"The licensee reported the leak test results/analysis data as follows:
1) Hewlett-Packard/Agilent model 19233 serial number L4716 (Contamination level of 0.064uCi)
2) Hewlett-Packard/Agilent model G1533A serial number K3545 (Contamination level of 0.013uCi)
3) Hewlett-Packard/Agilent model G1223A serial number F1219 (Contamination level of 0.008uCi)
"Each of the leaking devices contains a nominal activity of 15 mCi of Ni63, although the devices are authorized to contain up to 18 mCi of Ni63.
"The licensee reported that areas where the devices were handled had been surveyed and indicated that no contamination above background present in the work area or facility. The licensee reported to the Agency that the leaking ECDs were returned to the manufacturer on March 7, 2014.
"No further action is required and the Agency considers this matter closed." |
| Power Reactor | Event Number: 50492 | Facility: PRAIRIE ISLAND
Region: 3 State: MN
Unit: [1] [ ] [ ]
RX Type: [1] W-2-LP,[2] W-2-LP
NRC Notified By: STEVE INGALLS
HQ OPS Officer: DANIEL MILLS | Notification Date: 09/25/2014
Notification Time: 00:04 [ET]
Event Date: 09/24/2014
Event Time: 15:33 [CDT]
Last Update Date: 09/25/2014 | Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(3)(v)(D) - ACCIDENT MITIGATION
| Person (Organization):
NICK VALOS (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 UNIT 1 TRAIN B RVLIS AND ICCM INOPERABLE DUE TO DATA ERRORS
"At 1533 CDT on September 24, during a Past Operability review for the Unit 1 Train B Inadequate Core Cooling Monitor (ICCM), it was discovered that due to data errors, the as left VDC [voltage DC] value for the Remote Display Power Supply was outside of acceptance criteria during the last performance of SP 1213B 'Inadequate Core Cooling Monitor Calibration,' completed on 8/25/2014. The Reactor Vessel Level Instrumentation System (RVLIS) and the Core Exit Thermocouples (CETs) are subsystems of the ICCM System and were determined to be inoperable during this timeframe. Therefore Tech Spec 3.3.3 Condition A for Train B RVLIS and Tech Spec 3.3.3 Condition B for CETs were not met. Also during this time, Unit 1 Train A ICCM was out of service from 1110 [CDT] on 9/8/2014 to 1530 [CDT] on 9/12/2014 for calibration. Thus, both trains of Core Exit Thermocouples and RVLIS were inoperable during this 4 day, 4 hour and 20 minute time period.
"ICCM System indication is used for Safety Injection Termination and Reinitiation criteria in the Emergency Operating Procedures (EOPs). During the time when both trains of ICCM were inoperable, B Train remained in service and was thought to be operable. This provided the potential for operators to perform an untimely SI Termination or Reinitiation during an accident based on inaccurate information. The Emergency Response Computer System was available to provide accurate backup information to the operators and for timely Emergency Action Limit classification
"This is reportable under 10 CFR 50.72 (b)(3)(v)(D), Event or Condition that Could Have Prevented Fulfillment of a Safety Function to Mitigate the Consequences of an Accident.
"Based on new information currently being received from Westinghouse, the ICCM Remote Display Power Supply only functions to communicate data to the display and has no effect on the data values. Thus the ICCM System may have been functioning properly. An update to this information will be provided once it is available
"The protection of the health and safety of the public was not affected by this issue.
"The Unit 1 Train B ICCM System was returned to operable status at 2300 [CDT] on 9/24/2014.
"The license has notified the NRC Senior Resident Inspector." |
| Power Reactor | Event Number: 50495 | Facility: OYSTER CREEK
Region: 1 State: NJ
Unit: [1] [ ] [ ]
RX Type: [1] GE-2
NRC Notified By: JEREMY SHARKEY
HQ OPS Officer: MARK ABRAMOVITZ | Notification Date: 09/25/2014
Notification Time: 16:35 [ET]
Event Date: 09/23/2014
Event Time: 15:55 [EDT]
Last Update Date: 09/25/2014 | Emergency Class: NON EMERGENCY
10 CFR Section:
21.21(d)(3)(i) - DEFECTS AND NONCOMPLIANCE
| Person (Organization):
MEL GRAY (R1DO)
NICK VALOS (R3DO)
PART 21 GROUP (EMAI)
|
| Unit | SCRAM Code | RX CRIT | Initial PWR | Initial RX Mode | Current PWR | Current RX Mode | | 1 | N | N | 0 | Refueling | 0 | Refueling | Event Text PART 21 REPORT - ELECTROMATIC RELIEF VALVE EXCESSIVE WEAR
The following report was received via e-mail:
"This is a non-emergency notification from Oyster Creek Nuclear Generating Station (OCNGS) required under 10 CFR Part 21 concerning the design of Electromatic Relief Valve (EMRV) actuators.
"On June 20, 2014, during as-found bench (stroke) testing of the EMRV actuators removed from the plant during refueling outage 1R24 (October 2012), two of five EMRV actuators failed to operate. Subsequent inspection of these actuators found unexpected wear of the posts (grooves approximately 1/2 inch from the top), springs (thinned and broken at the top), and guides (grooves inside), with one spring having a piece axially wedged between the post and the guide.
"The root cause of this failure was determined to be the inadequate design of the EMRV actuators in that when placed in an environment where the actuator is subject to the vibration associated with plant operation, the allowed installation tolerances between posts and guides can create a condition where the springs can jam the actuator plunger assembly by wedging between the guides and the posts. If the EMRV actuators are set up in a condition where the posts are not optimally aligned, preferential wear of the post is observed due to interaction of the post, spring, and guide. Additionally, the vendor guidance for refurbishment of the EMRV actuator does not provide the necessary acceptance criteria for alignment of the posts to guides to ensure that the springs, posts, and guides do not interact in a way that causes preferential wear of the post allowing the jamming mechanism to exist.
"By OCNGS process, the EMRV actuators are refurbished with new springs, posts, guides, and microswitches every 24 months during refueling outages due to the known wear of these parts. The actuator inspection/refurbishment frequency of 24 months exceeds the manufacturer's (i.e., Dresser Industries) recommended frequency of 36 months (per Vendor Manual VM-OC-0030, Installation and Maintenance Manual for Electromatic Relief Valves, Revision 1, Section VII, Ref. 4.5). In addition, in 2008, the station implemented the manufacturer's recommended material changes intended to minimize part wear, and prevent potential actuator failures.
"Identification of Facility and Component: Oyster Creek Nuclear Generating Station / Electromatic Relief Valve Actuator, Dresser Valve Type 6 inches Model 1525-VX
"Identification of Component Manufacturer and/or Supplier: Dresser Industries.
"Nature of Defect: Inadequate design of the EMRV actuators in that when placed in an environment where the actuator is subject to the vibration associated with plant operation, the allowed installation tolerances between posts and guides can create a condition where the springs can jam the actuator plunger assembly by wedging between the guides and the posts.
"Safety Significance (e.g., substantial hazard that is or could be created): Identified condition is a Substantial Safety Hazard since it could cause EMRV to be inoperable, which could result in a loss of safety function.
"Date of Discovery of Initial Condition (taken from the IR): 06/20/14
"Date of Discovery of the Substantial Safety Hazard (date of approval of the technical evaluation): 09/22/14
"Recommended Actions: Change the design to mitigate the cause of the actuator parts (i.e., springs, guides and posts) interacting which results in unexpected wear due to vibration induced fretting.
"Number and Locations of All Defective Components: Two EMRVs located in the Drywell on the Main Steam piping.
"Any Advice Related to the Defect: Consider actuator design change if vibration conditions at the valve location results in unexpected wear of the EMRV actuators. Perform actuator as-found stroke testing before removal for refurbishment to determine in situ operability of the component. With current design, ensure posts are aligned such that preferential wear does not occur. Spring should be optimally centered on the guide and the post and guide should be equidistant around the full circumference of the bushing.
"Contacts (Name, Title, Location, Phone Number, etc.): Sylvain L. Schwartz, Senior Staff Engineer, Oyster Creek Nuclear Generating Station, Phone: (609) 971-4558, Email: sylvain.schwartz@exeloncorp.com"
The licensee notified the NRC Resident Inspector.
Plants with similar Dresser EMRVs: Nine Mile Point, Quad Cities, and Dresden. |
| Power Reactor | Event Number: 50496 | Facility: MONTICELLO
Region: 3 State: MN
Unit: [1] [ ] [ ]
RX Type: [1] GE-3
NRC Notified By: KIRK DUEA
HQ OPS Officer: DANIEL MILLS | Notification Date: 09/26/2014
Notification Time: 02:53 [ET]
Event Date: 09/25/2014
Event Time: 22:00 [CDT]
Last Update Date: 09/26/2014 | Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(3)(v)(C) - POT UNCNTRL RAD REL
50.72(b)(3)(v)(D) - ACCIDENT MITIGATION
50.72(b)(3)(xiii) - LOSS COMM/ASMT/RESPONSE
| Person (Organization):
NICK VALOS (R3DO)
|
| Unit | SCRAM Code | RX CRIT | Initial PWR | Initial RX Mode | Current PWR | Current RX Mode | | 1 | N | Y | 69 | Power Operation | 69 | Power Operation | Event Text CONTAINMENT ISOLATION DECLARED INOPERABLE DUE TO RELAY AGE
"At 2200 CDT on September 25, 2014, the Duty Shift Manager was notified that Agastat relays associated with Primary Containment Isolation valves on the Hydrogen-Oxygen Analyzing System are beyond the analyzed shelf life for relays that are in the normally energized state and are considered INOPERABLE. This affected both primary containment isolation valves for a containment penetration on multiple flow paths. This issue was determined to be reportable under [10 CFR] 50.72 (b)(3)(v)(C) & (D) for an event or condition that at the time of discovery could have prevented the fulfillment of the safety function of structures or systems that are needed to control the release of radioactive material and mitigate the consequences of an accident.
"Additionally, the required actions involved isolating six flow paths via manual isolation valves. This action rendered the Hydrogen-Oxygen Analyzers non-functional for both trains and constitutes a loss of Emergency Preparedness and Accident Assessment Capability. This is reportable under [10 CFR] 50. 72(b)(3)(xiii).
"The Primary Containment Isolation Valves have been, and remain, in their closed position to satisfy their Primary Containment Function and protect the health and safety of the public.
"The NRC Senior Resident Inspector has been notified."
The licensee will notify the state on Minnesota.
The relays of concern were manufactured 19 years ago and have been in operation for 11 years, versus a manufacturer assumption of a 10 year operational lifespan. | |