Event Notification Report for November 14, 2013

FIRE DAMAGE OF A GENERAL LICENSED DEVICE

"On the afternoon of November 15, 2013 at 1530 CST, the Alabama Office of Radiation Control was notified that on November 14, 2013 a coal bed fire occurred at Nichols Aluminum Alabama in Decatur, AL. This fire occurred in the vicinity of a general licensed radioactive material device causing damage to the device. The device was identified as an Accuracy model 7000M, serial number 4084600. The device contained a Sr-90 sealed source model USS-18, source serial number LB-650 with an original activity of 11.1 milliCurie. Initial action taken by the company was to restrict access around the device at a 100 feet radius. A consultant from Phillips Group evaluated the radiation hazard on November 16, 2013 taking radiation measurements and a leak test of the device. Preliminary test indicate that the sealed source capsule remained intact with no leakage. The device shutter could not be completely closed but the device has been secured and presents no hazard to workers. The general licensee plans to contract with the manufacturer to package and dispose of the device. The written report from the general licensee is pending."

Alabama Incident #13-52

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]."

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.

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."


* * * UPDATE FROM WARREN MOORE TO DANIEL MILLS ON 11/18/13 AT 0950 EST * * *

A subsequent licensee evaluation determined that there were ten dropped control rods.

Notified the R1DO (Desai).

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."

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.

Y-90 MICROSHPERE DOSE LESS THAN PRESCRIBED

"On the afternoon of 11/14/2013, a patient was scheduled for a Y-90 microsphere radioembolization treatment at Indiana University Medical Center under NRC License 13-02752-03. The treatment consisted of two separate doses of Y-90 for which two separate written directives were prepared. Segment 4 of the patient's liver was prescribed a dosage of 27.0 mCi, and the right lobe of the liver was prescribed a dosage of 88.0 mCi. At 13:50 on 11/14/2013, following measurement of the remaining activity after injection of the Y-90 microspheres, it was determined that a dose of 19.5 mCi was delivered to segment 4 (72.2% of the intended dose). Shortly thereafter, at 14:06, a dose of 87.5 mCi of Y-90 was delivered to the right lobe (99.4% of the intended dose).

"Both procedures appeared to proceed in accordance with standard operating procedures, and no abnormalities were identified during the procedure by the Interventional Radiology Physician or the Health Physicist supporting the procedure. Following the procedure, personnel and area surveys were performed using an SE International GM meter and no contamination of personnel, the room or equipment was identified. The container holding residual activity from the segment 4 treatment has been set aside for decay and further analysis.

"As the activity delivered to segment 4 of the liver meets the criteria in 10 CFR 35.3045(a)(1) and 10 CFR 35.3045(a)(1)(i), a report to the NRC Operations Center shall be made in accordance with 10 CFR 35.3045(c). The attending physician and patient will be contacted in accordance with 10 CFR 35.3045(e)."

A Medical Event may indicate potential problems in a medical facility's use of radioactive materials. It does not necessarily result in harm to the patient.

AGREEMENT STATE REPORT - DAMAGED PORTABLE GAUGE

The following was received via facsimile from the Utah Division of Radiation Control:

"The Operations Manager of Hales Sand and Gravel reported to the Division of Radiation Control that one of the licensee's gauge operators was at a temporary job site to perform soil moisture density measurements on November 14, 2013. After taking a moisture density measurement, the gauge operator left the gauge sitting on the ground while he walked away to talk to the roller operator. The gauge was left in the backing up path of Hales Sand Gravel's grader (heavy equipment) and was run over.

"After everyone was cleared out of the area, the Radiation Safety Officer of Jones & Demille Engineering was contacted to provide a survey instrument. At the time of the incident the radioactive sources were in the safe shielded position. After the gauge was run over, the gauge was broken into two pieces; the source rod was separated from the shielding block but was still in one piece. Upon arrival, the Radiation Safety Officer verified that the radioactive sources were still intact and attached to the source rod.

"The Radiation Safety Officer added additional shielding to the source rod to prevent the sources from becoming detached. The Radiation Safety Officer then put the damaged gauge and its pieces into the transportation container. After the gauge was loaded in the truck and removed from the job site, the Radiation Safety Officer performed a survey of the area the accident occurred at to verify that no contamination was present. No contamination was found, and the gauge was returned to its storage area in Elsinore, Utah.

"On November 15, 2013, the Division of Radiation Control's inspector arrived at Hales Sand and Gravel's facility at approximately 1230 [MST]. The inspector interviewed personnel involved in the accident and collected statements. The inspector took photographs of the damaged gauge, collected wipe tests, and took surveys of the damaged gauge. The inspector also visited the site of the accident to perform a contamination survey. No readings were distinguishable from background.

"Gauge information: Troxler 3430, s/n 31986"

Event Report ID Number: UT 130004

PART 21 REPORT REGARDING IDENTIFIED DISCREPANCY WITH INDIVIDUAL NDT LEVEL II CERTIFICATION

Received the following information was obtained via fax:

"Columbiana HI Tech, LLC is performing an evaluation to determine a potentially reportable condition per 10 CFR Part 21 requirements. Description of Noncompliance: A review of NDE certifications revealed discrepant training and work experience records from previous employment provided by a Columbiana HI Tech, LLC employee at the time of certification. This condition was first discovered on 11/14/13. The employee was certified under the Columbiana HI Tech, LLC Qualification & Certification of Nondestructive Examination Personnel Procedure passing all written and practical examinations required for Visual (VT) and Liquid Penetrant (PT) Inspection methods. The discrepant experience and training records from a previous employer were used to help fulfill the training and experience requirements for certification.

"Items affected: Columbiana Hi Tech, LLC Licensed components: Package USA/9196/B(U)F-96; certification number 9196 Rev. 26 UX-30 Overpack. A list of units affected will be compiled.

"Columbiana HI Tech, LLC CUSTOMER components: A list of components manufactured and inspected during the effected time is being reviewed to allow CUSTOMER Notification. Full disclosure will be made to all effected customers as soon as identified and within the reporting requirements of 10 CFR Part 21."


* * * RETRACTION - FROM LARRY WALKER TO GEROND GEORGE AT 1812 EST ON 01/09/2014 * * *

"Columbiana Hi Tech, LLC has concluded that the extent of the issue associated with this inspector is limited to his certification and did not negatively impact his competency for performance of Nondestructive Examination.

"Columbiana Hi Tech has determined this is not a reportable condition per 10 [CFR] Part 21."

Notified R2DO(Hopper), Part 21 Group, and NMSS EO(Silva) via email.