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| Discovered date | Reporting criterion | Title | Description | LER | |
|---|---|---|---|---|---|
| ENS 55855 | 21 July 2021 05:00:00 | Agreement State | Damaged Portable Gauge | The following was received from the state of Kansas via email: On 4/21/22 during the course of an inspection of the facility, the State of Kansas discovered that an incident involving one of their portable gauges occurred on 7/21/2021. This incident was never reported to Kansas and was discovered through the inspection process. Licensee Kansas State University (# 38-C011-01) had a Campbell Pacific Nuclear model 503 portable gauge (serial number 50505) damaged while being used in a field at the Hutchinson, Kansas field research station. The gauge contained 50 mCi of AmBe. The gauge was run over when the student who was using the gauge under the oversight of the local RSO (Radiation Safety Officer) (unknown at this time if the local RSO was present at the site) backed a vehicle over it. At this time Kansas has not been able to determine if the student left the gauge unattended for a brief time or if the student did not properly secure the gauge into the vehicle and it fell out. The gauge was inspected immediately after the incident, and it was found that, though the gauge shielding appeared to be intact, the shipping case was damaged. Immediately following the incident, the student contacted their Primary Investigator (PI), who is a university instructor overseeing the student's project, to inform him of the incident, but it was reported that the PI asked if it was urgent and the student said no. The gauge was discovered damaged by the PI a week later on 7/28/2021. Upon discovery, the PI reported that he ordered a new shipping case and ordered leak tests. The leak tests were performed on 7/29/2021 and did not show damage to the source. The damage to the gauge housing was on the opposite side of the machine from the source and did not interfere with the source's insertion or retraction. Because of this, the licensee stated that they decided it was not reportable to Kansas. An investigation is underway to determine what steps were taken by the licensee, including possible repairs to the unit. Follow-up information will be provided as it is obtained. | |
| ENS 54421 | 5 December 2019 00:00:00 | Non | Annual Surveillance Not Completed | The following was received via email from Kansas State University Nuclear Reactor Facility: Per Technical Specification 4.7.3, fuel elements comprising approximately 1/3 of the core shall be visually inspected annually for corrosion and mechanical damage such that the entire core shall be inspected at a 3-year intervals. Due to an inspection tracking sheet sorting error, four fuel elements were not marked to be inspected and are currently outside of the required surveillance frequency. Failure to perform a surveillance within the required time interval shall result in the component being inoperable. The reactor is and will remain shutdown until review by the Reactor Safeguards Committee of corrective actions in accordance with the Technical Specifications. A written report will be submitted within ten days summarizing the reportable occurrence. | |
| ENS 53805 | 22 December 2018 06:00:00 | Non | Surveillance Not Done Due to Inadequate Procedure | The following was received via email: (The licensee notified) the NRC of a reportable occurrence identified on December 22, 2018 due to violation of an LCO (limiting condition for operation). While performing testing and surveillance following control rod maintenance, the Reactor Supervisor identified an interlock check was not being completed due to an inadequate procedure. The inadequacy was identified as part of additional documentation checks that are being implemented at the facility. Per Technical Specification 3.4.3 - Table 2, the CONTROL ROD (STANDARD) position interlock must be operable in PULSE MODE. The interlock prevents withdrawal of the standard control rods while in PULSE MODE. The surveillance to ensure the interlock is operable is required on a SEMIANNUAL frequency as specified in Technical Specifications 4.4.2. As currently written, the procedure used to satisfy this surveillance does not direct the operator to enter PULSE MODE to check the interlock. Instead, the procedure activated a similarly named interlock (Pulse Power) that indirectly prevents withdrawal of the standard control rods by engaging the source interlock; however it is not specified to enter PULSE MODE in the procedure steps. The CONTROL ROD (STANDARD) position interlock was subsequently tested to be operable in that it functioned properly during the test and no operational history would suggest it was non-functioning otherwise. Based on the surveillance not being performed in PULSE MODE, the interlock was technically not operable during pulsing operations. The reactor is and will remain shutdown until a review by the Reactor Safeguards Committee of corrective actions in accordance with Technical Specifications. A written report will be submitted within ten days summarizing the reportable occurrence. | |
| ENS 53520 | 21 July 2018 05:00:00 | 10 CFR 50.72(a)(1)(i), Emergency Class Declaration | Unusual Event Declared at a Non-Power Reactor Site Due to a Campus Gas Leak | The licensee declared an Unusual Event on 7/21/2018 at 2000 CDT due to a detected gas leak near a campus dormitory. The reactor was not impacted. Local authorities and the gas company are on location in order to identify and repair the source of the gas leak. The Kansas State University non-power reactor is in a safe shutdown condition.
The Unusual Event was terminated on July 21, 2018 at 2348 CDT. The source of the gas leak was identified and isolated. Notified NRR (Evans), NRR PM (Traiforos), NRR (Reed), IRD (Grant), DHS SWO, DHS NICC, FEMA OPS AND FEMA NATIONAL WATCH CENTER, FEMA NRCC SASC AND NUCLEAR SSA via email. | |
| ENS 52980 | 19 September 2017 19:49:00 | Non | Reactor Bay Ventilation Inoperable | The following was received via email: The exhaust plenum monitor (EPM) was sent out for calibration a couple of weeks ago. It was received back on September 17, 2017, and returned to service. (On 9/20/17,) one of the senior reactor operators noticed that the noble gas detector was not responding properly. It was turned off. A relay on the EPM controls reactor bay ventilation, which is required to be operable per Technical Specifications. If the EPM alarms, ventilation is turned off. This relay was rewired to use only the two operating EPM detectors. Despite this, the reactor bay ventilation was observed to not operate as expected. The problem was traced back to a separate faulty relay, which was subsequently bypassed. The senior reactor operator had access to a control room breaker, and planned to manually cut power to the ventilation system if the EPM or continuous air monitor alarmed. An experiment was installed that has the potential to release radioactive gases; the reactor was started up at 1449 CDT and shut down at 1743 CDT. No increases in radiation levels were observed during or after operations. The ventilation system was operating during reactor operations. As previously mentioned, reactor bay ventilation is required to be operable per Technical Specifications. Operable is defined as being capable of performing it's intended function in a normal manner. Manually cutting power to the ventilation system via control room breaker is not considered normal operation for the ventilation system. According to Technical Specifications, the reactor may be operated with the ventilation system inoperable, but reactor experiment operations with the potential to release radioactive gases or aerosols must be secured. While the apparatus was secured according to the definition in Section 1 of the Technical Specifications, the experiment was still performed, and so experiment operations were not secured in accordance with the Limiting Condition of Operation. | |
| ENS 49117 | 13 June 2013 20:39:00 | Non | Failure of In-Core Thermocouple | 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 46982 | 24 June 2011 18:53:00 | 10 CFR 50.72(a)(1)(i), Emergency Class Declaration | Unusual Event Declared Due to Indications of Fuel Damage | The reactor operated at 10 W to 25 kW steady-state power for a total of 25 minutes. At 1257 CDT the Reactor Manager noticed that the radiation monitor (AMS-4) was reading approximately 80 times the Derived Air Concentration (DAC) for Iodine. A second continuous air monitor and a portable thin-window ionization chamber both read background levels of radiation. A high-volume air sampler was used to sample the air near the sampling location of the AMS, which draws air from the reactor deck. The filter from the air sampler read 3700 DPM after drawing 50 cubic feet of air, indicating elevated levels of activity. An HPGe (High Purity Germanium Detector) spectrograph indicated some Cs-137 in the air filter sample, but no I-131. Primary water activity and conductivity were both at normal levels. The indicated DAC of Iodine increased to a value of 149 times DAC before decreasing to 145 times DAC. The Reactor Manager and Radiation Safety Officer agreed that fuel damage was unlikely, but the results of testing were inconclusive, and that it would be prudent to declare an Unusual Event.
HPGe spectroscopy indicated only background levels of Cs-137 for a air sample, a primary water sample, and a swipe taken from the reactor deck. None of these samples showed I-131. A second portable air sample taken at the reactor deck, but on the opposite side of the deck from the radioactive sampling handling table, did not have elevated counts. The Reactor Manager and Radiation Safety Officer agreed that the AMS-4 was most likely mis-calibrated, and the elevated levels in the first high-volume air sample were probably due to the proximity of the radioactive sample handling table to the sampling location. Notified NRR (Grobe), NRR EO (Evans), R4DO (Deese), IRD (Grant), DHS (Hill), FEMA (Blankenship) | |
| ENS 46283 | 22 September 2010 12:40:00 | Non | Unexpected Changes in Restricted Area Dose Rates | 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 44289 | 12 June 2008 05:21:00 | 10 CFR 50.72(a)(1)(i), Emergency Class Declaration | Alert Declared Due to Tornado Impacting Reactor Building Onsite | The research reactor building at the Kansas State University was damaged due to a tornado onsite. There is extensive damage to the building. The reactor was shutdown properly earlier in the day. There is no damage to the reactor. There is a loss of power onsite however, this type of reactor requires no active cooling. The reactor is shutdown and stable. * * * UPDATE AT 0555 ON 6/12/08 FROM PAUL WHALEY TO PETE SNYDER * * * The Alert condition at the site has been terminated. Conditions requiring entry into the Alert have been rectified. Notified IRD (Gott), R4DO(Miller), EDO (Mallett), DHS (Jason Craig), FEMA (Mike LaForty), DOE (Thomas Yates), USDA (Rob Leadbetter), and HHS (Harry Peagler). |