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Event Notification Report for June 26, 2015

U.S. Nuclear Regulatory Commission
Operations Center

Event Reports For
06/25/2015 - 06/26/2015

** EVENT NUMBERS **


51162 51179 51181 51182

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Non-Agreement State Event Number: 51162
Rep Org: ECS MID-ATLANTIC, LLC
Licensee: ECS MID-ATLANTIC, LLC
Region: 1
City: ABERDEEN State: MD
County:
License #: 19-31269-01
Agreement: Y
Docket:
NRC Notified By: IRVIN FISCHER
HQ OPS Officer: DONALD NORWOOD
Notification Date: 06/17/2015
Notification Time: 14:23 [ET]
Event Date: 06/16/2015
Event Time: 15:00 [EDT]
Last Update Date: 06/17/2015
Emergency Class: NON EMERGENCY
10 CFR Section:
30.50(b)(4) - FIRE/EXPLOSION
Person (Organization):
BRICE BICKETT (R1DO)
NMSS_EVENTS_NOTIFICA (EMAI)

Event Text

MOISTURE/DENSITY GAUGE INVOLVED IN AUTOMOBILE FIRE

A licensee employee working at a job site in Claymont, Delaware went to leave the job site at 1500 EDT on June 16, 2015. The employee got in his vehicle and turned on the ignition. A fire started under and around the dashboard. The employee exited his vehicle and called 911 and the licensee Radiation Safety Officer. The local fire department responded. The front portion of the vehicle had been totally engulfed. The vehicle trunk, where the gauge was stored, received heat and smoke damage. The carrying case of the gauge was partially melted. No visible damage occurred to the gauge itself. The gauge was transported to a service vendor for inspection and repair if needed.

The gauge was a Troxler 3440 Moisture/Density gauge, serial number 20118, containing an 8 mCi Cesium-137 source and a 40 mCi Am-241source.

The licensee notified NRC R1(Ragland), Delaware Department of Natural Resources and Environmental Control, Division of Waste and Hazardous Substances, Emergency Prevention and Response Section, and Delaware Health and Social Services, Division of Public Health.

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Power Reactor Event Number: 51179
Facility: CLINTON
Region: 3 State: IL
Unit: [1] [ ] [ ]
RX Type: [1] GE-6
NRC Notified By: PAUL SANDERS
HQ OPS Officer: JEFF HERRERA
Notification Date: 06/25/2015
Notification Time: 09:27 [ET]
Event Date: 06/25/2015
Event Time: 03:01 [CDT]
Last Update Date: 06/25/2015
Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(3)(v)(C) - POT UNCNTRL RAD REL
Person (Organization):
DAVID HILLS (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

SECONDARY CONTAINMENT PRESSURE INCREASE DUE TO VOLTAGE TRANSIENT

"At approximately 0301 [CDT] on 6/25/15, the Main Control Room received numerous annunciators that indicated a trip of the Emergency Reserve Auxiliary Transformer (ERAT) Static VAR Compensator (SVC) caused by a voltage transient on the 138 kV feed due to thunderstorms in the area. The Division 1 Safety Bus was manually aligned from the reserve source to its normal source. As a result of the voltage transient, the Division 1 Fuel Building Ventilation (VF) system isolation dampers closed causing a trip of VF supply and exhaust fans. With no running VF fans, secondary containment differential pressure rose to slightly greater than 0 inches water gauge and which exceeded the Technical Specification requirement of greater than 0.25 inches vacuum water gauge. The Control Room entered EOP-8, Secondary Containment Control. This event is being reported as a condition that could have prevented fulfillment of a safety function under 10 CFR 50.72(b)(3)(v)(C). Secondary Containment differential pressure was restored within Technical Specification requirements at 0319 [CDT] by reopening the VF isolation dampers and restarting the VF supply and exhaust fans. The ERAT SVC was returned to service at 0457 [CDT].

"The NRC Resident Inspector has been notified."

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Non-Agreement State Event Number: 51181
Rep Org: ELECTRIC POWER RESEARCH INSTITUTE
Licensee: ELECTRIC POWER RESEARCH INSTITUTE
Region: 1
City: CHARLOTTE State: NC
County:
License #:
Agreement: Y
Docket:
NRC Notified By: TRACY WILSON
HQ OPS Officer: DONG HWA PARK
Notification Date: 06/25/2015
Notification Time: 15:40 [ET]
Event Date: 05/01/2015
Event Time: [EDT]
Last Update Date: 06/25/2015
Emergency Class: NON EMERGENCY
10 CFR Section:
21.21(a)(2) - INTERIM EVAL OF DEVIATION
Person (Organization):
JOHN ROGGE (R1DO)
KATHLEEN O'DONOHUE (R2DO)
VIVIAN CAMPBELL (R4DO)

Event Text

PART 21 REPORT - DEVIATION IN NOZZLE MODELING INTERNAL REPORTS

The following was received via facsimile:

"[This report pertains] to a deviation in a basic product (EPRI nozzle modeling internal reports) supplied by EPRI (Electric Power Research Institute) regarding Westinghouse Pressurizer Head Nozzle Inner Corner Region Ultrasonic Inspections. EPRI will complete all evaluation efforts and provide a determination of reportability in accordance with 10 CFR Part 21 no later than July 24, 2015.

"EPRI has conducted an evaluation to the basic product's actual use and determined that the ASME examination volume coverage for at least one of the pressurizer nozzles has changed and is now 90 percent or less. A 90 percent threshold is required by ASME Boiler & Pressure Vessel Code, Section XI.

"Design inputs used in EPRI modeling for ultrasonic scanning coverage for nuclear safety related component nozzles may have been inaccurate. In some cases, the upper and lower heads of Westinghouse pressurizers can be offset from the center of each nozzle (spray, safety, relief, surge). This offset results in a change in the thickness of the pressurizer head as compared to an on-axis pressurizer head with the same radial dimensions. Some of the computer models EPRI used to describe these pressurizer heads did not account for an increase in the thickness due to these offsets. As a result, in some cases the ultrasonic inspection parameters produced by these computer models may have produced inaccuracies in the examination volume coverage calculations.

"In the case of a basic component which contains a defect or falls to comply, the number and location of these components in use at, supplied for, being supplied for, or may be supplied for, manufactured, or being manufactured for one or more facilities or activities subject to the regulations in this part.

Utility Name/Plant Name
Exelon Corporation / Ginna
First Energy Nuclear Operating / Beaver Valley 1
Entergy / Indian Point 2
Entergy / Indian Point 3
Pacific Gas & Electric Co. / Diablo Canyon Unit 2
Dominion Generation / North Anna

"EPRI has reviewed the pressurizer upper and lower head drawings for the nozzles that it has modeled and determined if these offsets are present. For those cases that are potentially affected EPRI has recalculate the new examination volume coverage for the nozzle inspection detection techniques and provided this information to the corresponding licensees.

"EPRI staff shall develop a matrix or table to better define the necessary design inputs for computer modeling of nozzles. This should also include a question to the utility regarding any obstructions or thickness changes which would impact the ultrasonic inspection parameters. EPRI staff shall improve its documentation for review and approval of design inputs for computer modeling. Consideration shall also be given to including a review of design inputs by the member along with an acknowledgement from the member that the design inputs are appropriate for use. EPRI staff shall consider methods of including additional conservatism to the modeling results to better accommodate changes which may be observed in the field. The project quality plan and quality project instruction shall be updated as necessary to accommodate or clarify these improvements. Completion commitment date - 10/27/2015.

"The coverage calculations indicated in the notification letters would likely increase if the EPRI modeled scan plans are exceeded and or if additional inspection angles were implemented. Conversely, these coverage calculations would likely decrease if physical field limitations prevented the ultrasonic probe from executing the EPRI modeled scan pattern. It is on this basis that recipients of this letter must evaluate the condition pursuant to 10 CFR Part 21.21 to determine if it could represent a substantial safety hazard reportable under 10 CFR Part 21."

Potentially affected US plants include Ginna, Beaver Valley Unit 1, Indian Point Units 2 and 3, Diablo Canyon Unit 2, and North Anna.

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Power Reactor Event Number: 51182
Facility: COLUMBIA GENERATING STATION
Region: 4 State: WA
Unit: [2] [ ] [ ]
RX Type: [2] GE-5
NRC Notified By: QUOC VO
HQ OPS Officer: MARK ABRAMOVITZ
Notification Date: 06/26/2015
Notification Time: 04:38 [ET]
Event Date: 06/25/2015
Event Time: 22:00 [PDT]
Last Update Date: 06/26/2015
Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(3)(v)(A) - POT UNABLE TO SAFE SD
Person (Organization):
VIVIAN CAMPBELL (R4DO)

Unit SCRAM Code RX CRIT Initial PWR Initial RX Mode Current PWR Current RX Mode
2 N Y 2 Startup 2 Startup

Event Text

TWO REACTOR VESSEL LEVEL CHANNELS FAILED HIGH

"At 2200 PDT during startup from refueling outage 22, it was discovered that both level instruments used in reactor protection system (RPS) trip system 'A' for initiation of a reactor scram on low reactor pressure vessel (RPV) level were observed to have failed high. This resulted in the inability to generate a full reactor scram on low level (+13 inches). All remaining RPV level indications demonstrated that level was being maintained within normal operating bands. This constitutes a condition that at the time of discovery could have prevented the fulfillment of the safety function of structures or systems that are needed to shut down the reactor.

"The RPS trip logic at Columbia consists of two trip systems, RPS trip system 'A' and RPS trip system 'B'. There are two level instrument channels in each trip system. Columbia utilizes a 'one-out-of-two taken-twice' trip logic to generate a full scram signal. At least one channel in both trip systems must actuate to generate a full scram signal. With both level instruments in RPS system 'A' failed high, the RPS trip logic was unable to generate a full scram.

"At 2246 [PDT] and in accordance with TS LCO 3.3.1.1 Condition C, a half scram was generated on RPS trip system 'A' to restore full scram capability. The cause of the failure of the two level instruments associated with RPS Trip system 'A' is under investigation."

The level channels are being calibrated prior to changing to mode 1 (power operations). The licensee will notify the NRC Resident Inspector.

Page Last Reviewed/Updated Friday, June 26, 2015
Friday, June 26, 2015