| U.S. Nuclear Regulatory Commission
Operations Center Event Reports For
11/18/2016 - 11/22/2016 ** EVENT NUMBERS ** |
 | !!!!! THIS EVENT HAS BEEN RETRACTED. THIS EVENT HAS BEEN RETRACTED !!!!! | | Power Reactor | Event Number: 52254 | Facility: INDIAN POINT
Region: 1 State: NY
Unit: [2] [ ] [ ]
RX Type: [2] W-4-LP,[3] W-4-LP
NRC Notified By: CHRIS HASSENBEIN
HQ OPS Officer: JEFF HERRERA | Notification Date: 09/21/2016
Notification Time: 09:20 [ET]
Event Date: 09/21/2016
Event Time: 02:21 [EDT]
Last Update Date: 11/18/2016 | Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(3)(v)(D) - ACCIDENT MITIGATION
| Person (Organization):
PAUL KROHN (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 DISCHARGE CHECK VALVE FAILURE TO SEAT CAUSES TRIP OF COMPONENT COOLING WATER PUMP
"At 0221 [EDT] on 9/21/16, Operators at Unit 2 Secured the 21 Component Cooling Water (CCW) Pump for planned maintenance while 22 and 23 CCW pumps were in operation. When the 21 pump was secured, the discharge check valve failed to seat. This resulted in a low system pressure and reverse rotation of the 21 CCW Pump due to the discharge of the 22 and 23 CCW pumps to a common header. When system pressure dropped below 107 psig the 21 CCW pump received an auto start signal. Due to the reverse rotation, the 21 CCW pump tripped on overcurrent. Reactor Operators directed Field Operators to manually shut the 21 CCW Pump discharge valve. The 21 CCW pump Discharge Valve was closed at 0223 [EDT]. This action was successful in stopping the reverse flow and restoring system parameters. During this two minute period the CCW system was declared inoperable and LCO 3.0.3 was entered. Unit 2 exited LCO 3.0.3 at 0223 [EDT] after observing system pressure and flow return to normal. The declaration of inoperability on the CCW system is considered a Loss of Safety Function for purposes of reporting under 50.72(b)(3)(v)(D). There was no reduction in power while in LCO 3.0.3 and no other issues arose."
The Licensee notified the NRC Resident Inspector.
The Licensee notified the Public Service Commission.
* * * RETRACTION FROM CHARLES ROKES TO HOWIE CROUCH AT 1108 EST ON 11/18/16 * * *
"Indian Point Unit 2 is retracting the 8-hour non-emergency notification made on September 21, 2016, at 0920 EDT (EN#52254). The notification on September 21, 2016, reported a safety system functional failure (SSFF) as a result of declaring the Component Cooling Water System (CCW) inoperable due to failure of the 21 CCW pump discharge check valve (761C) to close. This condition was discovered during planned maintenance after securing the 21 CCW pump while the 22 and 23 CCW pumps were in operation. When the 21 CCW pump was secured, the discharge check valve failed to seat. This resulted in a low system pressure and reverse rotation of the 21 CCW pump due to the discharge of the 22 and 23 CCW pumps to a common header. Condition was reported as a safety system functional failure (SSFF) under 10 CFR 50.72(b)(3)(v)(D).
"After further investigation of the condition, a revised calculation was prepared for the CCW hydraulic model which is used to analyze CCW system performance for normal and DBA [design basis accident] modes of operation and documented in a calculation. The new calculation included the as-found condition of the 21 CCW pump discharge check valve failure to seat. Based on the results of the new calculation, the CCW system is capable of performing its design basis heat removal function during a design basis accident. Calculated flow rates with CCW aligned for Post-LOCA recirculation demonstrates that with failed open check valve 761C, the 22 CCW pump and 23 CCW pump have adequate NPSH margin, are operating below analyzed pump run out and deliver flow to the CCW system that is significantly greater than the flow required for post-LOCA recirculation. Therefore the CCW system was operable and a safety system functional failure (SSFF) did not occur as a result of failed open 21 CCW pump discharge check valve 761C."
The licensee has notified the NRC Resident Inspector and will be notifying the New York Public Service Commission.
Notified R1DO (Bickett). |
| !!!!! THIS EVENT HAS BEEN RETRACTED. THIS EVENT HAS BEEN RETRACTED !!!!! | | Power Reactor | Event Number: 52260 | Facility: DIABLO CANYON
Region: 4 State: CA
Unit: [ ] [2] [ ]
RX Type: [1] W-4-LP,[2] W-4-LP
NRC Notified By: BRYAN GALVAN
HQ OPS Officer: DONG HWA PARK | Notification Date: 09/23/2016
Notification Time: 20:10 [ET]
Event Date: 09/20/2016
Event Time: 20:47 [PDT]
Last Update Date: 11/18/2016 | Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(3)(ii)(B) - UNANALYZED CONDITION
| Person (Organization):
RICK DEESE (R4DO)
|
| 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 STEAM LINE PIPE WHIP RESTRAINT COUPLING SLEEVE FOUND NOT ENGAGED
"On 9/20/2016, a coupling sleeve on a pipe whip restraint located on the 119 foot elevation of the turbine building associated with Unit 2 main steam line 4 was found to be not engaged. As a result of the detached coupling, the restraint was not capable of performing its restraint function for a postulated pipe whip event on the main steam line. The coupling was reconnected on 9/20/2016, restoring its functionality. An extent of condition walkdown was subsequently performed for the other Unit 1 and Unit 2 steam line restraints and no similar issues were identified.
"This concern did not result in any adverse effect on the radiological health and safety of the public.
"The purpose of this whip restraint is to restrain the steam line for a postulated loss at the G-line anchor (east side of Turbine Building above the 104 foot elevation). The restraint protects the floor slab at the 104 foot elevation, which extends over the Unit 2 component cooling water heat exchangers.
"With the detached coupling, equipment in the area may have been vulnerable to damage if a pipe whip event occurred. Further analysis is needed to conclude whether the heat exchangers and other equipment would have remained protected in such an event and whether this would have significantly affected the designed plant response to a pipe event.
"Based on the need for further analysis, this event is being reported as an unanalyzed condition that may have significantly degraded plant safety in accordance with 10 CFR 50.72(b)(3)(ii)(B).
"The NRC Senior Resident Inspector was notified."
* * * RETRACTION FROM FRANK LEE TO DONALD NORWOOD AT 1953 EST ON 11/18/2016 * * *
"The purpose of this notification is to retract a previous report made on EN #52260, reported 9/23/2016. NRC notification was initially made as a result of a condition that required further analysis to determine whether the condition would have significantly degraded plant safety in accordance with 10 CFR 50.72(b)(3)(ii)(B).
"Further analysis of the condition concluded that PG&E Design Class I equipment located inside the Turbine Building would have remained undamaged and capable of performing their safety functions. The Turbine Building would not have experienced failure of major structural elements or adverse impact to the overall building stability.
"Therefore, the coupling sleeve on a pipe whip restraint located on the 119 foot elevation of the Turbine Building associated with Unit 2 main steam line 4 that was found to be not engaged did not constitute an unanalyzed condition that may have significantly degraded plant safety in accordance with 10 CFR 50.72(b)(3)(ii)(B).
"The NRC Resident Inspector has been notified."
Notified R4DO (Azua) |
| Agreement State | Event Number: 52360 | Rep Org: GEORGIA RADIOACTIVE MATERIAL PGM
Licensee: THERAGENICS CORPORATION
Region: 1
City: BUFORD State: GA
County:
License #: UNKNOWN
Agreement: Y
Docket:
NRC Notified By: MONICA JOHNSON
HQ OPS Officer: JOHN SHOEMAKER | Notification Date: 11/10/2016
Notification Time: 10:33 [ET]
Event Date: 10/12/2016
Event Time: [EST]
Last Update Date: 11/10/2016 | Emergency Class: NON EMERGENCY
10 CFR Section:
AGREEMENT STATE
| Person (Organization):
RAY MCKINLEY (R1DO)
NMSS_EVENTS_NOTIFIC (EMAI)
ANGELA MCINTOSH (INES)
| Event Text AGREEMENT STATE REPORT - POTENTIAL OVEREXPOSURE
The following information was excerpted from a report received from the State of Georgia via email:
"On Wednesday, November 9th, it was reported to the Department [Georgia Radioactive Materials Program] by Theragenics Corporation in Buford, GA that an employee's dosimetry report indicated that she had exceeded the annual whole body dose limit of 5000 mRem. An investigation has been conducted by the employer and no possible explanation has been found as of yet. Our office has an ongoing investigation and further details will be provided as we receive them.
"[The licensee reports receiving] notification from their dosimetry processor, on October 12, an individual received a whole body dose of 5,215 mRem. [The licensee] spent the last month trying to recreate a situation that could have exposed this worker to a dose of this magnitude and can't make it happen. According to discussions, with the worker, she never left her dosimetry in a lab where it could have been exposed to an unshielded source of radiation. According to her supervisor, team leads in the lab in which she worked, and HP [Health Physics], she is a very good radiation worker and, there is never anything out of the normal at her work station.
"[The worker] did take her whole body dosimeter and both finger rings home with her one evening. Even more baffling is her ring dose is less than 200 mRem. [The licensee has] tried to recreate that scenario of taking the dosimeters home as well and have not had a dosimeter come back from processing that is above its minimal level of detection.
"It is [licensee's] interpretation of [the CFR,] Reportable Events, that this report of an overexposure is due to [the State of Georgia] 30 days after notification, or Friday, 11/11/16. [The licensee is] waiting on additional information [from the dosimetry processor] prior to report submittal." |
| Agreement State | Event Number: 52361 | Rep Org: OK DEQ RAD MANAGEMENT
Licensee: BUILDING AND EARTH SCIENCES
Region: 4
City: TULSA State: OK
County:
License #: OK-31032-01
Agreement: Y
Docket:
NRC Notified By: KEVIN SAMPSON
HQ OPS Officer: MARK ABRAMOVITZ | Notification Date: 11/10/2016
Notification Time: 15:30 [ET]
Event Date: 11/10/2016
Event Time: 14:00 [CST]
Last Update Date: 11/10/2016 | Emergency Class: NON EMERGENCY
10 CFR Section:
AGREEMENT STATE
| Person (Organization):
VIVIAN CAMPBELL (R4DO)
NMSS_EVENTS_NOTIFICA (EMAI)
| Event Text AGREEMENT STATE REPORT - DAMAGED MOISTURE DENSITY GAUGE
The following report was received via e-mail:
"Building and Earth Sciences had a Troxler Model 3430 portable gauge run over by construction equipment at a temporary job site in Tulsa, OK. The source rod has been retracted and the sources are shielded. A leak test has been collected and will be counted as soon as possible. The gauge will be returned to Troxler for repair or disposal." |
| Non-Agreement State | Event Number: 52362 | Rep Org: NORTHERN INDIANA PUBLIC SVC COMPANY
Licensee: NORTHERN INDIANA PUBLIC SVC COMPANY
Region: 3
City: VALPARAISO State: IN
County:
License #: 13-14984-01
Agreement: N
Docket:
NRC Notified By: ANDREW G. FISHMAN
HQ OPS Officer: STEVE SANDIN | Notification Date: 11/11/2016
Notification Time: 12:57 [ET]
Event Date: 11/10/2016
Event Time: 14:00 [CST]
Last Update Date: 11/11/2016 | Emergency Class: NON EMERGENCY
10 CFR Section:
30.50(b)(2) - SAFETY EQUIPMENT FAILURE
| Person (Organization):
HIRONORI PETERSON (R3DO)
NMSS_EVENTS_NOTIFICA (EMAI)
| Event Text SHUTTER FAILURE ON A FIXED GAUGE
At 1400 CST on 11/10/16, technicians performed a routine shutter check on a fixed gauge located at their Wheatfield, IN facility. The gauge, a Kay-Ray Model 7062BP, S/N 30340 containing 50 mCi Cs-137 as of 8/81, failed to properly operate with the shutter returning to the mid-position. The shutter is currently in the locked closed position with no potential for personnel exposure. The gauge is 30 ft. above floor level with the beam directed upward. The licensee will contact the manufacturer for repairs. |
| Agreement State | Event Number: 52365 | Rep Org: ILLINOIS EMERGENCY MGMT. AGENCY
Licensee: UNIVERSITY OF ILLINOIS AT URBANA-CHAMPAIGN
Region: 3
City: URBANA State: IL
County:
License #: IL-01271-01
Agreement: Y
Docket:
NRC Notified By: C. GIBB VINSON
HQ OPS Officer: JEFF HERRERA | Notification Date: 11/14/2016
Notification Time: 16:26 [ET]
Event Date: 11/11/2016
Event Time: [CST]
Last Update Date: 11/14/2016 | Emergency Class: NON EMERGENCY
10 CFR Section:
AGREEMENT STATE
| Person (Organization):
HIRONORI PETERSON (R3DO)
NMSS_EVENTS_NOTIFICA (EMAI)
|
| This material event contains a "Less than Cat 3 " level of radioactive material. | Event Text AGREEMENT STATE REPORT - LOST AM-241 SOURCE
The following report was received from the Illinois Emergency Management Agency via email:
"On October 14, 2016, the licensee's RSO [Radiation Safety Officer] called to report a lost Am-241, 45 mCi source that was part of a wave generator manufactured on 12/2/1982. The manufacturer and model number of the wave generator and radioactive component are not known. It was noted as lost during a recent inventory. An exhaustive search of nearby laboratories and waste areas did not reveal any clues of its whereabouts. During a radioactive material inventory in March 2016, records show this source was identified and accounted for. Access to the storage room is strictly controlled. It is only speculation to how the source became missing. First, it is possible that the source was mistakenly shipped as part of the University's participation in the Source Collection and Threat Reduction (SCATR) program. A vendor for the SCATR program removed nearly 200 sealed sources in December 2015. It is possible that the Am-241 source was mistakenly included in this shipment and the March 2016 inventory was in error. Another possibility is that the source was misplaced by staff during the summer clean-up of 2016. Personnel were questioned about this likelihood but there was no recollection of moving or discarding of such a unit. Inquiries with staff and SCATR are ongoing."
IL Item Number: IL 16012
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: 52377 | Facility: CLINTON
Region: 3 State: IL
Unit: [1] [ ] [ ]
RX Type: [1] GE-6
NRC Notified By: JACOB SERENO
HQ OPS Officer: BETHANY CECERE | Notification Date: 11/18/2016
Notification Time: 13:40 [ET]
Event Date: 11/17/2016
Event Time: 17:00 [CST]
Last Update Date: 11/18/2016 | Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(3)(ii)(B) - UNANALYZED CONDITION
| Person (Organization):
HIRONORI PETERSON (R3DO)
|
| Unit | SCRAM Code | RX CRIT | Initial PWR | Initial RX Mode | Current PWR | Current RX Mode | | 1 | N | Y | 99 | Power Operation | 99 | Power Operation | Event Text INCORRECT METHOD USED FOR CALCULATION OF CONTROL ROOM HABITABILITY
"During the NRC CDBI [Component Design Basis Inspection], it was identified that the calculation used to demonstrate Control Room Habitability following a Design Basis Accident (DBA) utilized an inappropriate methodology. Specifically, the calculation used dual air inlets for the emergency zones as the type of system used for Main Control [Room] Ventilation (VC) system. In order to use the dual inlet type system in the analysis, each of the VC subsystems is required to be single failure proof. The VC system is single failure proof, but the individual subsystems at the inlet, as designed, are not.
"The dual inlet type system allows for certain calculated dose concentrations to be reduced by a factor of 4. Elimination of this reduction factor results in higher calculated control room dose following a DBA which exceeds the 5 Rem limit.
"This event is reportable under 10 CFR 50.72(b)(3)(ii)(B), 'Any event or condition that results in the nuclear power plant being in an unanalyzed condition that significantly degrades plant safety.'"
A standing order has been issued for compensatory actions in the event of an emergency.
The licensee notified the NRC Resident Inspector. |
| Part 21 | Event Number: 52378 | Rep Org: WESTINGHOUSE
Licensee: WESTINGHOUSE
Region: 1
City: CRANBERRY TOWNSHIP State: PA
County:
License #:
Agreement: Y
Docket:
NRC Notified By: JAMES GRESHAM
HQ OPS Officer: MARK ABRAMOVITZ | Notification Date: 11/18/2016
Notification Time: 14:42 [ET]
Event Date: 11/18/2016
Event Time: [EST]
Last Update Date: 11/18/2016 | Emergency Class: NON EMERGENCY
10 CFR Section:
21.21(d)(3)(i) - DEFECTS AND NONCOMPLIANCE
| Person (Organization):
BRICE BICKETT (R1DO)
OMAR LOPEZ (R2DO)
PART 21/50.55 REACT (EMAI)
| Event Text PART 21 NOTIFICATION - CONTROL ROD GUIDE CARD EXCESSIVE WEAR
The following report is excerpted from Westinghouse fax LTR-NRC-16-74:
"Westinghouse nuclear steam supply system (NSSS) plants that are using iron nitride rod cluster control assemblies (RCCAs) in conjunction with Westinghouse 17x17 A or 17x17 AS style guide tubes (GTs), the GT guide card wear inspection guidance provided in Westinghouse report WCAP-17451-P, Revision 1, is not conservative.
"Based on the data, a determination was made as to whether an affected plant could exceed its permissible number of worn guide cards before completing its first guide card wear measurement (GCWM) inspection, the timing of which is based on recommendations documented in the WCAP. The concern associated with exceeding the permissible number of worn guide cards is that a rodlet might slip out of alignment with the GT guide card fingers and become stuck. For example, if an unsupported rodlet were to deflect as a result of applied lateral fluid forces during normal operation, it could get caught on the corner of a worn guide card finger and jam or buckle during rod motion (i.e., scram). Additionally,. if a GT with an excessive number of worn guide cards experiences a seismic event or loss-of-coolant accident (LOCA), an unsupported rodlet might become plastically deformed, which could prevent a RCCA from scramming properly. Should this affect multiple GT locations, it could prevent a plant from achieving reactor shutdown using only the RCCAs.
"Left uncorrected, the wear inspection guidelines in WCAP-17451-P, Revision 1, underestimate the 17x17 A and 17x17 AS style GT guide card wear attributed to the use of iron nitride RCCAs to the extent that, for the four plants listed, the WCAP recommended number of effective full-power years (EFPYs) until the first GCWM is performed may no longer have margin to safety. The guidelines presented in this WCAP do not result in a defect for the other plants to which the WCAP applies."
Affected reactors are: Catawba Units 1 and 2, McGuire Unit 2 , and Millstone Unit 3. |
| Power Reactor | Event Number: 52380 | Facility: FERMI
Region: 3 State: MI
Unit: [2] [ ] [ ]
RX Type: [2] GE-4
NRC Notified By: GREG MILLER
HQ OPS Officer: DONG HWA PARK | Notification Date: 11/20/2016
Notification Time: 04:56 [ET]
Event Date: 11/19/2016
Event Time: 21:50 [EST]
Last Update Date: 11/20/2016 | Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(3)(v)(C) - POT UNCNTRL RAD REL
| Person (Organization):
MARK JEFFERS (R3DO)
|
| Unit | SCRAM Code | RX CRIT | Initial PWR | Initial RX Mode | Current PWR | Current RX Mode | | 2 | N | Y | 96 | Power Operation | 100 | Power Operation | Event Text SECONDARY CONTAINMENT TECHNICAL SPECIFICATION NOT MET
"On November 19, 2016, starting at 2150 EST, high wind conditions encountered on site resulted in the Technical Specification (TS) for secondary containment pressure boundary not being met numerous times. The duration of time that the secondary containment Technical Specification was not met was approximately 1 second for each instance.
"All plant equipment responded as required to the changing environmental conditions and Reactor Building HVAC returned secondary containment pressure within TS limits. At 0430 EST, high wind conditions have subsided and secondary containment vacuum was greater than the TS operability limit of 0.125 inches of vacuum water gauge (TS SR 3.6.4.1.1) and steady, and the LCO was exited. There were no radiological releases associated with this event.
"Declaring secondary containment inoperable is reportable under 10 CFR 50.72(b)(3)(v)(C) as an event or condition that could have prevented the fulfillment of a safety function needed to control the release of radioactive material.
"The licensee has notified the NRC Resident Inspector."
* * * UPDATE ON 11/20/16 AT 1416 EST FROM BRETT JEBBIA TO BETHANY CECERE * * *
"On November 20, 2016, starting at 0654 EST, high wind conditions encountered on site resulted in the Technical Specification (TS) for secondary containment pressure boundary not being met on multiple different occasions as of event notification update time. The duration of time that the secondary containment Technical Specification was not met was approximately 1 second for each instance.
"Fermi 2 continues to remain in a gale force wind advisory for the local area of Lake Erie.
"All plant equipment responded as required to the changing environmental conditions and Reactor Building HVAC returned secondary containment pressure within TS limits. There were no radiological releases associated with this event.
"Declaring secondary containment inoperable is reportable under 10 CFR 50.72(b)(3)(v)(C) as an event or condition that could have prevented the fulfillment of a safety function needed to control the release of radioactive material.
"The licensee has notified the NRC Resident Inspector."
The R3DO (Jeffers) has been notified.
* * * UPDATE ON 11/20/16 AT 2104 FROM GREG MILLER TO BETHANY CECERE * * *
"On November 20, 2016, at 1426 EST, high wind conditions encountered on site resulted in the Technical Specification (TS) for secondary containment pressure boundary not being met. The duration of time that the secondary containment Technical Specification was not met was approximately 1 second.
"The Fermi 2 local area of Lake Erie is no longer in a gale force wind advisory and the high wind conditions have subsided.
"All plant equipment responded as required to the changing environmental conditions and Reactor Building HVAC returned secondary containment pressure within TS limits. There were no radiological releases associated with this event.
"Declaring secondary containment inoperable is reportable under 10 CFR 50.72(b)(3)(v)(C) as an event or condition that could have prevented the fulfillment of a safety function needed to control the release of radioactive material.
"The licensee has notified the NRC Resident Inspector."
The R3DO (Jeffers) has been notified. |
| Power Reactor | Event Number: 52381 | Facility: OYSTER CREEK
Region: 1 State: NJ
Unit: [1] [ ] [ ]
RX Type: [1] GE-2
NRC Notified By: JOSH MCGUIRE
HQ OPS Officer: JEFF ROTTON | Notification Date: 11/20/2016
Notification Time: 06:01 [ET]
Event Date: 11/20/2016
Event Time: 03:42 [EST]
Last Update Date: 11/20/2016 | Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(2)(iv)(B) - RPS ACTUATION - CRITICAL
| Person (Organization):
BRICE BICKETT (R1DO)
|
| Unit | SCRAM Code | RX CRIT | Initial PWR | Initial RX Mode | Current PWR | Current RX Mode | | 1 | A/R | Y | 90 | Power Operation | 0 | Hot Shutdown | Event Text AUTOMATIC REACTOR SCRAM DURING MAIN TURBINE TESTING
"At 0342 EST, an automatic reactor scram was processed during turbine valve testing. All rods inserted into the core as expected and all systems functioned as expected during the scram.
"The event is reportable within 4 hours per 10 CFR 50.72(b)(2)(iv)(B) - any event or condition that results in actuation of the Reactor Protection System (RPS) when the reactor is critical except when the actuation results from and is part of a preplanned sequence during testing or reactor operation."
The plant response to the reactor scram was uncomplicated. The main feedwater system is maintaining reactor water level and decay heat is being removed by the main turbine bypass valves to the main condenser. The unit is in a normal shutdown electrical lineup. No SRVs lifted during the scram. The licensee was testing the main turbine trip function just prior to the scram. The cause is under investigation.
The licensee notified the NRC Resident Inspector. |
| Power Reactor | Event Number: 52382 | Facility: COLUMBIA GENERATING STATION
Region: 4 State: WA
Unit: [2] [ ] [ ]
RX Type: [2] GE-5
NRC Notified By: ZACHARY DUNHAM
HQ OPS Officer: DONALD NORWOOD | Notification Date: 11/20/2016
Notification Time: 18:53 [ET]
Event Date: 11/20/2016
Event Time: 14:02 [PST]
Last Update Date: 11/20/2016 | 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
| Person (Organization):
RAY AZUA (R4DO)
|
| Unit | SCRAM Code | RX CRIT | Initial PWR | Initial RX Mode | Current PWR | Current RX Mode | | 2 | N | Y | 97 | Power Operation | 97 | Power Operation | Event Text SECONDARY CONTAINMENT DIFFERENTIAL PRESSURE LESS THAN TECHNICAL SPECIFICATION REQUIREMENT
"On November 20, 2016 at 1402 PST, Reactor Building Exhaust Air Fan 1B, REA-FN-1B, failed to start in manual which caused the Technical Specification (TS) for secondary containment pressure boundary to not be met. The duration of the time that the secondary containment TS was not met was approximately less than one minute. REA-FN-1B was being started in manual during a shift of Reactor Building Ventilation to support a post-maintenance support task on REA-FN-1B.
"Secondary containment differential pressure was restored within the TS requirement of greater than or equal to 0.25 inch of vacuum water gauge by restarting Reactor Building HVAC Train A.
"The cause of REA-FN-1B failing to start is currently under investigation.
"This condition is being reported under 10 CFR 50.72(b)(3)(v)(C) and 10 CFR 50.72(b)(3)(v)(D) for an event or condition that could have prevented fulfillment of a safety function needed to control the release of radioactive material and accident mitigation.
"The licensee has notified the NRC Resident Inspector." |
| Part 21 | Event Number: 52383 | Rep Org: SYSTEM ONE SOLUTIONS LLC
Licensee: SYSTEM ONE SOLUTIONS LLC
Region: 1
City: CHESWICK State: PA
County:
License #:
Agreement: Y
Docket:
NRC Notified By: WILLIAM STUCKEY
HQ OPS Officer: MARK ABRAMOVITZ | Notification Date: 11/21/2016
Notification Time: 11:43 [ET]
Event Date: 10/28/2016
Event Time: [EST]
Last Update Date: 11/21/2016 | Emergency Class: NON EMERGENCY
10 CFR Section:
21.21(d)(3)(i) - DEFECTS AND NONCOMPLIANCE
| Person (Organization):
AARON McCRAW (R3DO)
NICK TAYLOR (R4DO)
PART 21/50.55 REACTO (EMAI)
| Event Text PART-21 NOTIFICATION - WILLFUL MISCONDUCT AND FALSIFICATION OF RECORDS
The following report was excerpted from an e-mail:
"On September 28th, System One received an email from Curtis Wright that a 'forged' VT-1, 2, 3 Level II training record was being circulated by [an individual], who was seeking employment with Applied Technical Services.
"On September 28th 2016, System One voided the certification of [the individual] and proceeded to address the condition under the nonconformance reporting process.
"On October 11th, 2016, System One determined that [the individual], who was certified in VT-1, 2, 3 by the System One Quality Assurance Manager, did not fully meet the certification requirements of the System One Written Practice under CP-199 (in effect at the time) due to lack of acceptable VT [visual testing] Training (40 hours of required class room training) to support the VT-1, VT-2 and VT-3 certification issued by System One. A review of the original certification record found/confirmed there was indication of alteration on a training record used for the original certification.
"On October 28th, 2016, System One notified stakeholders (AREVA NP). [The individual was deployed to Areva on two occasions] based upon holding VT-1, 2, 3 certification.
AREVA NP - deployed in 10/7/2012 to Cooper with VT-1, 2, 3 Certification
AREVA NP - deployed in 3/31/2013 to DC Cook with VT-1, 2, 3 Certification
Carolina Energy Services - deployed in 8/23/2013 - certification conducted by CES not System One LLC.
AREVA NP - deployed in 10/27/2013 to Palo Verde with PT Level II and QC Mechanical (not deployed for VT-1, 2, 3)
"AREVA NP has informed System One that they have conducted/completed their extent of condition review regarding work performed by [the individual], and confirmed to System One, that [the individual] did not perform any safety related VT-1, 2, 3 work at either the D. C. Cook or Cooper facilities.
"Our internal investigation has determined this matter does not represent a breakdown in the mechanics of the System One Quality Assurance Program or procedures in effect at the time. We have established the event represents willful misconduct and the falsification of a record by an individual, and use of such forged record, under 10 CFR references contained in US NRC Information Notice 2013-15." |
| Part 21 | Event Number: 52384 | Rep Org: AMETEK SOLID STATE CONTROLS
Licensee: AMETEK SOLID STATE CONTROLS
Region: 3
City: COLUMBUS State: OH
County:
License #:
Agreement: Y
Docket:
NRC Notified By: ETHAN SALSBURY
HQ OPS Officer: MARK ABRAMOVITZ | Notification Date: 11/21/2016
Notification Time: 12:13 [ET]
Event Date: 10/26/2016
Event Time: [EST]
Last Update Date: 11/21/2016 | Emergency Class: NON EMERGENCY
10 CFR Section:
21.21(d)(3)(i) - DEFECTS AND NONCOMPLIANCE
| Person (Organization):
AARON McCRAW (R3DO)
PART 21/50.55 REACT (EMAI)
| Event Text PART-21 - INCORRECT CAPACITOR USED ON CIRCUIT BOARDS
The following report was excerpted from an e-mail:
"Product: Ametek Solidstate Controls Analog Oscillator, printed circuit board part number 80-9230404-90
"A single unit, Ametek part number 80-9230404-90, was shipped to Exelon Dresden Station on the reference purchase order has an incorrect capacitor installed. This error was identified during testing of a subsequent identical part. Ametek reviewed all printed circuit boards built under that run and verified all affected printed circuits boards less one, were still in house. This shipment was made on October 26, 2016. Ametek contacted Exelon Dresden immediately upon identification of the incorrect part installation and Exelon was able to have the board retrieved from their inventory. Ametek Solidstate Controls is submitting this notification as a precaution under the requirements of 10 CFR Part 21, but considers it a limited incident as the single defective part was isolated and the board is in the process of being returned. No further action is required and this notification does not affect any other installation, client, inventory, or equipment provided by Ametek." |
| Power Reactor | Event Number: 52388 | Facility: INDIAN POINT
Region: 1 State: NY
Unit: [2] [ ] [ ]
RX Type: [2] W-4-LP,[3] W-4-LP
NRC Notified By: STEVEN RADOMSKI
HQ OPS Officer: MARK ABRAMOVITZ | Notification Date: 11/21/2016
Notification Time: 21:22 [ET]
Event Date: 11/21/2016
Event Time: 17:38 [EST]
Last Update Date: 11/21/2016 | Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(3)(v)(C) - POT UNCNTRL RAD REL
| Person (Organization):
JAMES DWYER (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 SERVICE WATER LEAK INSIDE CONTAINMENT
"At 1738 [EST] on November 21, 2016, a leak was identified inside the Vapor Containment building on a Service Water line associated with 24 Fan Cooler Unit. The leak was isolated at 1743 by shutting the Service Water Isolation valves to 24 Fan Cooler Unit. This isolation meets the Technical Specifications of 3.6.1 Condition A Required Action. The leaking defect could have resulted in post-LOCA air leakage out of containment in excess of that allowed by Technical Specification 3.6.1 (Containment) which requires leakage rates to comply with 10 CFR 50, Appendix J. This event had no effect on the health and safety of the public.
"This event is being reported under 10 CFR 50.72(b)(3)(v) and the guidance of NUREG 1022, section 3.2.7 as a loss of safety function."
The service water was quantified at approximately 15 gpm.
The licensee notified the NRC Resident Inspector. | |