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| Entered date | Event description | |
|---|---|---|
| ENS 46283 | 27 September 2010 13:47:00 | On Wednesday morning, September 22, 2010, the Senior Reactor Operator (SRO) pulled some oil samples, housed in aluminum racks, out of the well groove around the reflector near the top of the reactor where the samples had been exposed to radiation for 8 hours and then allowed to decay for 12 hours. When the samples and sample holder were removed from the pool, three detectors alarmed: two were expected to alarm, and a third, the Control Room Evacuation Alarm, was not expected to alarm. The Control Room Evacuation Alarm has a conservative, audible, alarm set point of 2.5 mR/h which is well below the actual evacuation threshold of 100 mR/h. The Control Room Evacuation Alarm alerted the SRO that conditions were not normal. Within 30 seconds, the SRO placed the aluminum sample holder behind a lead shield and the oil samples behind a beta shield. According to facility procedures, the facility did not need to be evacuated because the easily visible Control Room Evacuation Alarm meter had not exceeded the evacuation threshold of 100 mR/h. The SRO estimated his dose as being approximately 1 Rem Total Effective Dose Equivalent (TEDE) which is less than the 5 Rem/yr, Part 20 limit; and approximately 15 Rem extremity dose which is less than 50 Rem/yr Part 20 limit. Follow-up dosimetry reports indicated that the SRO only received 147 mRem deep dose and 148 mRem shallow dose. No members of the public were in the vicinity at the time of the event. The facility believes that the cause of the high radiation field was possibly due to a high level of impurities in the aluminum sample holder and not from the oil samples. However, the oil samples received greater than normal neutron exposure and shorter than normal decay before removal from the reactor pool. This particular experiment required that these samples decay for only 12 hours, whereas normally, a decay time of 2 to 5 days is required before the samples are removed from the pool. The Reactor Safeguards Committee has recommended that the facility perform irradiation tests on each of its sample holders to prevent unexpectedly high radioactivity from unknown impurities. Kansas State University is making this report per their license requirements and Technical Specification 6.9.a.6, an observed inadequacy in the implementation of either administrative or procedural controls, such that the inadequacy has caused the existence or development of an unsafe condition in connection with the operations of the reactor. The State of Kansas has been notified of this event. |
| ENS 49117 | 14 June 2013 11:20:00 | Description of Event: During operations at full power (500 kWth), the Senior Reactor Operator (SRO) on duty noticed that the fuel temperature thermocouple reader indicated a temperature of 202 (degrees) C, approximately 60 - 70 (degrees) C below the expected value. The SRO recognized that the problem was likely caused by a fuel thermocouple wire grounding to its conduit. A trainee was instructed to move the wires to avoid grounding. Following this action, the thermocouple reader indicated the proper value. Upon review of the log book, the SRO noticed that the faulty fuel temperature reading had been logged for several days without corrective action. The facility Technical Specifications (TS) require at least one fuel temperature indication to be operable during operation, and define a system as 'operable' when it is capable of performing its intended function in a normal manner. Therefore the fuel temperature indication was not operable as defined in the TS. Since the reactor was not immediately secured nor was the indication immediately fixed, the event constitutes a Reportable Occurrence per facility TS 6.9.2. Background: For the four days of reactor operations during which the problem with the thermocouple existed, the reactor was operated for short amounts of time (approximately 10 -15 minutes) at various power levels in order to characterize a new beam port configuration and test an experimental apparatus. The total time above 100 kWth was 2 hours and 51 minutes. The reactor was typically staffed by a trainee at the panel, supervised by a licensed reactor operator (RO). The indicated temperature is the average reading of three thermocouples. One thermocouple is at the fuel midplane, one is 2 (inches) above the midplane, and one is 2 (inches) below the midplane. The fuel temperature indication, when partially grounded, is approximately correct below the point of adding heat (approximately 10 kWth). It differs from the expected value by approximately 15 (degrees) C at 100 kWth, and by approximately 65 (degrees) C at 500 kWth. The fuel temperature readout at the control panel is used to provide an automatic scram at 400 (degrees) C. This setpoint is set well below the Safety Limit of 750 (degrees) C fuel temperature during steady state reactor operations. The fuel temperature scram is NOT required by TS. Normal operations at the reactor do not approach this scram setpoint or the Safety Limit. The logbooks are reviewed daily as part of the pre-operation reactor checkout procedure. The staff is trained to review logs back to the most recent time they were on duty, to check for changes to the reactor, problems with instruments, etc. The staff is not trained to audit the previous days' logs for anomalous readings. Timeline: The following timeline of operations is taken from the reactor logbook. Only operations at or above 100 kWth are listed, because the difference between measured and expected temperature is small at lower power levels. All times are local (Central Daylight Time). Date Time Power T (Measured, (degrees) C) T (Expected, (degrees) C) 6/7/2013 0946-1001 500kW 200 265 6/7/2013 1037 - 1050 500kW 202 265 6/7/2013 1441 - 1446 500kW 202 265 6/10/2013 0937 - 0950 100kW 83 100 6/10/2013 1005 - 1015 100kW 87 100 6/10/2013 1029-1041 100kW 84 100 6/12/2013 0906 - 0919 100kW 84 100 6/12/2013 0934 - 0959 530kW 217 270 6/12/2013 1611 - 1633 100kW 74* 100 6/13/2013 1103 - 1143 500kW 201 265 6/13/2013 1352 - 1406 530kW 209 270
6/13/2013 - 1559 - Problem observed and corrected by repositioning thermocouple wires. 6/14/2013 - 1020 - Reportable occurrence reported to NRC Headquarters Operations Center. Causes: The facility has identified the following as contributing causes to the event. 1. Licensed operators were not sufficiently attentive when supervising trainees at the control panel. The licensed operators were focused on reactor power indications and did not pay sufficient attention to other TS related indications. 2. The log book review required prior to daily operations was not conducted with sufficient rigor to detect the improper thermocouple readings logged the previous day. The review was instead focused on noting changes to the reactor facility and problems with instrumentation since the operators' previous duty at the panel. 3. Only one functioning instrumented fuel element was used to provide the required fuel temperature indication channel. Therefore no redundant readout was available to check against the indicated fuel temperature. 4. The sharp edge on the instrumented fuel element conduit can cut through the insulation on the thermocouple wires, causing grounding. Corrective Actions: The facility will perform the following corrective actions. All changes to the reactor systems, such as thermocouple wire insulation, are subject to review per the requirements of 10CFR50.59. Time: Prior to operation; Action: Attempt to improve insulation on thermocouple wires, using electrical tape, shrink tubing, or spray-on insulation. Time: Prior to operation; Action: Attempt to repair thermocouple wires for a currently installed but non-functional instrumented fuel element to provide an independent fuel temperature indication channel. Time: Prior to operation; Action: Install a new instrumented fuel element. This will bring the total number of independent channels of fuel temperature indication to 2 - 3. Time: Prior to operation and as part of requalification training program; Action: Train reactor staff on the importance of vigilance when supervising trainees and the importance of attentiveness to all channels of information at the control console, as opposed to focusing on a few specific indicators, such as reactor power channels. Time: Prior to operation and as part of requalification training program; Action: Train reactor staff to check for anomalous values in the prior days' log entries during the daily reactor checkout. Time: Upon approval by Reactor Safeguards Committee, but not necessarily prior to operation; Action: Append Procedure 15 - Reactor Startup with a list of observed instrument values for different reactor power levels to be used as a reference by trainees and licensed staff. A copy of this report will be provided to the Kansas State University Reactor Safeguards Committee for review. |
| ENS 55632 | 8 December 2021 14:22:00 | On 12/7/2021 at 1248 EST, the Breazeale reactor (50-005) (a TRIGA reactor) was operating for a NUCE 451 lab. The reactor operator was trying to perform a $0.75 (measure of reactivity) square wave with a setpoint of 500 kW. When the operator entered the setpoint, he did not hit ENTER so the 500 kW was not stored in the console, which defaulted to the current power level (100 W) as the setpoint. Once the square wave was executed, the control rods immediately began to move in to counter the $0.75 of reactivity from the pulse rod and maintain the power level at 100 W. The operator noticed that the setpoint was incorrect, and after 9 seconds, changed the power setpoint to 500 kW. At this time, the rod bank began to move out, adding $1.20 over the course of 4 seconds. (The maximum total reactivity beyond critical is estimated to be ~$1.10). The reactor scrammed based on high log range (fission chamber) power and high wide range (GIC) power. The last two points of data from the data historian indicate that the period was +0.25 seconds. Based on the rod insertion speed and differential rod worth at the position from which the scram was initiated, it was estimated that the maximum power following the scram (setpoint = 1.08 MW) was approximately 1.29 MW. The highest data point recorded by the historian was 1.38 MW (log fission chamber data), which is corroborated by the estimate calculated based on rod speed, and 1.38 MW represents the best estimate of the maximum reactor power. The reactor technical specifications (TS) dictate that: "The maximum power level SHALL be no greater than 1.1 MW (thermal)." (TS 3.1.1.b). This condition applies to non-pulse operation. According to the TS definitions, the reactor is neither "secured" nor "shut down" during a scram, and therefore must be considered to be operating while the rods are in motion after the scram is initiated. Therefore, this event resulted in the violation of TS 3.1.1.b by allowing power to reach 1.38 MW, higher than the 1.1 MW scram setpoint. It is worth noting that TRIGA reactors like the Breazeale reactor are designed to be pulsed to several gigawatts of power, and the 1.1 MW limit is based on steady state power analysis, not power transient analysis. The fuel reached a maximum temperature of 42 C, far below the safety limit of 1150 C (TS 2.1). This event is reportable due to: exceeding an LCO in the technical specifications (1.1 MW power limit) and an unanticipated change in reactivity greater than $1 when the rod bank drove out following the change in power setpoint. The root cause of this event was operator error. The operator failed to follow best practices by checking that the setpoint was entered correctly, and then acted outside of procedure to attempt to correct the setpoint. The reactor was immediately secured and tagged out pending corrective actions identified in the event evaluation document, AP4 2021-03. The immediate corrective actions, completed on 12/8/21, were to: 1 - add a pen-and-ink revision to SOP-1 instructing the operator to verify the power setpoint; 2 - hold a reactor staff training on the event, its causes, and the importance of following procedure and checking values entered into the console; 3 - implement an administrative prohibition on square waves until the console software can be changed to add a feature to prevent recurrence of this event. Following these corrective actions reactor operation was approved by the ADO (Level 2). A detailed written report will be sent to the Reactor Safeguards Committee and NRC by December 21st. The licensee will notify the Non-Power Production or Utilization Facility (NPUF) Licensing Branch Project Manager. |