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Event Notification Report for December 9, 2014

U.S. Nuclear Regulatory Commission
Operations Center

Event Reports For
12/08/2014 - 12/09/2014

** EVENT NUMBERS **


50532 50644 50646 50658

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!!!!! THIS EVENT HAS BEEN RETRACTED. THIS EVENT HAS BEEN RETRACTED !!!!!
Power Reactor Event Number: 50532
Facility: FITZPATRICK
Region: 1 State: NY
Unit: [1] [ ] [ ]
RX Type: [1] GE-4
NRC Notified By: THOMAS YURKON
HQ OPS Officer: VINCE KLCO
Notification Date: 10/14/2014
Notification Time: 01:30 [ET]
Event Date: 10/13/2014
Event Time: 19:35 [EDT]
Last Update Date: 12/08/2014
Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(3)(v)(D) - ACCIDENT MITIGATION
Person (Organization):
FRED BOWER (R1DO)

Unit SCRAM Code RX CRIT Initial PWR Initial RX Mode Current PWR Current RX Mode
1 N Y 100 Power Operation 60 Power Operation

Event Text

HIGH PRESSURE CORE INJECTION DEGRADED ACCIDENT MITIGATION CAPABILITY

"During the plant response to the trip of the B Recirculating water pump, reactor water level rose to the HPCI [High Pressure Core Injection] high water level trip setpoint as indicated on the associated instrumentation. With this high water level trip actuated, the HPCI high drywell pressure initiation signal would not have allowed the HPCI system to perform its intended safety function if required. If the HPCI system received the low water level initiation signal, the system would have been able to perform Its intended safety function. This high water level signal was actuated from 1935 [EDT] until reset at 1940 [EDT]. This is reportable under 50.72(b)(3)(v)."

The licensee notified NRC Resident Inspector.

* * * RETRACTION PROVIDED BY DAVID CALLAN TO JEFFREY HERRERA AT 1404 EDT ON 12/08/14 * * *

"Further review has determined that the condition was not a result of procedural errors/inadequacies, equipment failures, or design / analysis inadequacies. Plant systems responded as per design when the HPCI system high water level trip actuated when reactor vessel water level rose to the HPCI high water level trip setpoint. HPCI initiation has two logics: one for low-low vessel water level and the other for a high drywell pressure. A vessel low-low water level is an indication that reactor coolant is being lost with a need for HPCI injection for core cooling. High drywell pressure could indicate a line break in the Reactor Coolant Pressure Boundary inside the drywell. The HPCI level instrumentation is designed to shut down the HPCI system upon high water level to prevent HPCI turbine damage due to gross moisture carryover and will re-initiate HPCI if vessel water level drops to the initiation water level setpoint. A HPCI high drywell pressure initiation signal, above setpoint, would have made up the logic for HPCI initiation and as per design, HPCI would have injected at the vessel low low level setpoint without operator action to reset the trip. In this instance, the trip was reset as prescribed by station procedures. HPCI was capable of performing its safety function after the high water level trip reset either by operator action or instrumentation (low low level initiation)."

The licensee will be notifying the NRC Resident Inspector.

Notified R1DO (Rogge).

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Agreement State Event Number: 50644
Rep Org: OHIO BUREAU OF RADIATION PROTECTION
Licensee: UNIVERSITY OF CINCINNATI MEDICAL CENTER
Region: 3
City: CINCINNATI State: OH
County:
License #: OH 2110-31-00
Agreement: Y
Docket:
NRC Notified By: MARK LIGHT
HQ OPS Officer: DONG HWA PARK
Notification Date: 11/28/2014
Notification Time: 10:57 [ET]
Event Date: 11/26/2014
Event Time: 11:29 [EST]
Last Update Date: 11/28/2014
Emergency Class: NON EMERGENCY
10 CFR Section:
AGREEMENT STATE
Person (Organization):
ERIC DUNCAN (R3DO)
NMSS EVENTS NOTIFICA (EMAI)

Event Text

AGREEMENT STATE REPORT - Y-90 THERASPHERE UNDERDOSE TO LIVER

The following email was received from the State of Ohio:

"On Wednesday, November 26, 2014 at 1129 [EST], the Interventional Radiologist Authorized User [AU] physician began to administer the prescribed activity of 1.38 GBq Y-90 micro-spheres (Theraspherer) for an estimated dose of 125 Gy to segment 7 of the right liver lobe. After a couple of 20 cc syringes of saline had been pushed through the administration kit, the Nuclear Medicine technologist made an ion chamber measurement of the dose vial and tubing set and observed a very high reading of 2.5 Roentgens/hour using a Keithley 451B/RYR, serial number 2935, ion chamber with the beta window open. This was a very high measurement, so the AU physician continued to push an additional 6 syringes of saline for a total of 8 syringes through the administration set with very little reduction in the ion chamber reading. At that point the AU physician decided to call an end to the therapy.

"The administration set was moved out of the room and using the manufacturer's template, the Nalgene jar containing the dose vial and tubing set in the plastic beta shield was measured as 0.55 mR/hr versus a pre-therapy measurement of 2.3 mR/hr. This meant that approximately 24 percent of the dose remained in the dose vial and/or tubing set and micro-catheter.

"The RSO then asked the Nuclear Medicine technologist to take the Nalgene jar to the hot lab to measure the dose vial, tubing set and micro-catheter individually in the dose calibrator. Each of the items were carefully wrapped and measured in the dose calibrator using a calibration factor of 048 for the dose vial and 086 for the tubing sets. The Capintec CRC 15-R, serial number 156151, dose calibrator measurements were 0.009 GBq in the dose vial, 0.31 GBq in the 'D' line tubing set and 0.038 in the micro-catheter for a total of 0.357 GBq of the 1.44 GBq decayed to the time of administration which equals about 24.8 percent of the total dose.

"The prescribed dose was 1.38 GBq which would deliver about 125 Gy to the segment 7 of the right liver lobe; subtracting the 0.357 GBq the calculated dose to the liver segment was 98.15 Gy or about 78 percent of the prescribed dose.

"The Radiation Safety Officer contacted BTG International (Canada) to file an incident report and provided all of the specific details of the incident (Incident Report Number: OTT-PC-14-0073). The BTG representative asked to check with the AU physician to see if he administered the saline at approximately 20 cc/min, and not less than 10 cc/min so that the micro-spheres would not settle out of suspension during the administration. The tubing set and dose vial will be stored for decay-in-storage and returned to the manufacturer for further analysis.

"After interviewing the Authorized User physician following the case, he said that there was no medical reason related to the patient that lead to the under dose, and that he intended to administer the entire 125 Gy to the liver segment throughout the therapy. He also mentioned that the delivered 98.15 Gy to a small segment of the liver (segment 7) was therapeutic and provided optimal treatment dose to the tumor. He explained that the delivered dose that differed by more than 20 percent from the prescribed dose had no consequences to the patient and there is no requirement to repeat the treatment. He also mentioned that he pushed the saline with at least a rate of 10 cc/min or higher not exceeding 20 cc/min.

"According to the Ohio Administrative Code, 3701:1-58-101, Report and Notification of a Medical Event, the total dose exceeded 0.5 Sv (50 rem) to an organ and differed from the prescribed dose by more than 20 percent, and therefore, met the reporting criteria. A telephonic report was made by the Radiation Safety Officer to the Ohio Department of Health at 1615 [EST] on 11/26/2014 regarding this incident. The Authorized User physician that administered the therapy was the referring physician and notified the patient the day of the procedure. The Nuclear Medicine imaging following the therapy indicated that the microspheres were administered to the correct liver region."

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.

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Non-Agreement State Event Number: 50646
Rep Org: HYDROLAKE INC
Licensee: HYDROLAKE INC
Region: 3
City: MCBAIN State: MI
County:
License #: GL644751
Agreement: N
Docket:
NRC Notified By: JESS ROLSTON
HQ OPS Officer: DANIEL MILLS
Notification Date: 12/01/2014
Notification Time: 16:44 [ET]
Event Date: 11/28/2014
Event Time: 06:33 [EST]
Last Update Date: 12/01/2014
Emergency Class: NON EMERGENCY
10 CFR Section:
30.50(a) - PROTECTIVE ACTION PREVENTED
Person (Organization):
ERIC DUNCAN (R3DO)
NMSS EVENT NOTIFICAT (EMAI)

Event Text

FIRE DAMAGED LABORATORY CONTAINING NRC GENERAL LICENSED MATERIAL

An NRC general licensee reported that a fire damaged their facility. Inside this facility was an ASOMA Model 200 (Serial # 4649) device containing 13 mCi of Cm-244. The device is assumed to be damaged or destroyed. The licensee has limited access to the area and is contacting a qualified contractor to assist with disposal of the device.

The ASOMA Model 200 is an Energy Dispersive-X-Ray Fluorescent (ED-XRF) benchtop analyzer.

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Power Reactor Event Number: 50658
Facility: BROWNS FERRY
Region: 2 State: AL
Unit: [1] [ ] [ ]
RX Type: [1] GE-4,[2] GE-4,[3] GE-4
NRC Notified By: TODD CHRISTENSEN
HQ OPS Officer: JOHN SHOEMAKER
Notification Date: 12/08/2014
Notification Time: 12:23 [ET]
Event Date: 10/07/2014
Event Time: 09:35 [CST]
Last Update Date: 12/08/2014
Emergency Class: NON EMERGENCY
10 CFR Section:
50.73(a)(1) - INVALID SPECIF SYSTEM ACTUATION
Person (Organization):
SCOTT FREEMAN (R2DO)

Unit SCRAM Code RX CRIT Initial PWR Initial RX Mode Current PWR Current RX Mode
1 N N 0 Refueling 0 Refueling

Event Text

INVALID SPECIFIED SYSTEM ACTUATION

"This 60-day telephone notification is being made per the reporting requirements specified by 10 CFR 50.73(a)(2)(iv)(A) and 10 CFR 50.73(a)(1) to describe an invalid actuation of general containment isolation signals affecting containment isolation valves in more than one system.

"On October 7, 2014, at 2135 [CDT], while in a refueling outage with the reactor non-critical (Mode 5), work activities were in progress that included replacement of an excess flow check valve and execution of a Technical Specification Surveillance Procedure on the Automatic Depressurization System. Subsequent to valving in a level transmitter (LT), water levels in both the variable and reference legs of the LT were disturbed resulting in a Unit 1 full scram and Primary Containment Isolation System (PCIS) Groups 2, 3, 6, and 8 isolation signals due to receipt of an invalid low reactor water level signal. The PCIS Groups 2, 3, 6, and 8 isolations caused the initiation of Trains A, B, and C of the Standby Gas Treatment System and Control Room Emergency Ventilation Subsystem 'A'. The Reactor and Refuel Zone ventilation fans tripped and the secondary containment dampers isolated.

"Operations personnel responded to the PCIS initiation, ensured all equipment operated as designed, and placed affected systems back in service.

"Plant conditions which initiate PCIS Group 2 actuations are Reactor Vessel Low Water Level (Level 3) or High Drywell Pressure. The PCIS Group 3 actuations are initiated by Reactor Vessel Low Water Level (Level 3) or Reactor Water Cleanup Area High Temperature. The PCIS Group 6 actuations are initiated by Reactor Vessel Low Water Level (Level 3), High Drywell Pressure, or Reactor Building Ventilation Exhaust High Radiation (Reactor Zone or Refuel Zone). The PCIS Group 8 actuations are initiated by Low Reactor Vessel Water Level (Level 3) or High Drywell Pressure. At the time of the event, these conditions did not exist; therefore, the actuation of the PCIS was invalid.

"There were no safety consequences or impact to the health and safety of the public as a result of this event.

"This event was entered into the Corrective Action Program as Problem Evaluation Report 943038.

"The NRC Resident Inspector has been notified of this event."

Page Last Reviewed/Updated Tuesday, December 09, 2014
Tuesday, December 09, 2014