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[en] The widely used medical isotope technetium-99 m (99mTc) is a daughter of Molybdenum-99 (99Mo), which is mainly produced using dedicated research reactors from the nuclear fission of uranium-235 (235U). 99mTc has been used for several decades, which covers about 80% of the all the nuclear diagnostics procedures. Recently, the instability of the supply has become an important topic throughout the international radioisotope communities. The aging of major 99Mo production reactors has also caused frequent shutdowns. It has triggered movements to establish new research reactors for 99Mo production, as well as the development of various 99Mo production technologies. In this context, a new research reactor project was launched in 2012 in Korea. At the same time, the development of fission-based 99Mo production process was initiated by Korea Atomic Energy Research Institute (KAERI) in 2012 in order to be implemented by the new research reactor. The KAERI process is based on the caustic dissolution of plate-type LEU (low enriched uranium) dispersion targets, followed by the separation and purification using a series of columns. The development of proper waste treatment technologies for the gaseous, liquid, and solid radioactive wastes also took place. The first stage of this process development was completed in 2018. In this paper, the results of the hot test production of fission 99Mo using HANARO, KAERI's 30 MW research reactor, was described
[en] The irradiation capsules have been developed at HANARO for new alloy and fuel developments and the lifetime estimation of Nuclear Power Plants (NPPs). Among the irradiation facilities, the capsule is the most useful device for coping with the various test requirements at HANARO. There are several vertical test holes such as CT, IR and OR in the core of HANARO. An irradiation capsule is installed in these holes to evaluate the irradiation performance of nuclear fuels and materials at HANARO. The fuel capsule is applicable to research into the irradiation characteristics of fuel pellets and to obtain the in-core performance and the design data of nuclear fuel at HANARO. The thermal margin of plate type U-Mo fuel capsule was calculated for safety assessment using thermal-hydraulic system code RELAP5/MOD3.3. The analysis results lead to the following conclusions: 1) Fuel centerline temperature is a key concern during the forced convection. 2) In case of CT and OR hole, heat flux is limited at 43.96 W/cm2 and 42.58 W/cm2, respectively. 3) The Ob margin is dominant at low flow rate. 4) Heat flux is limited at 13.74 W/cm2 in case of IP hole.
[en] The interest in the safety of nuclear fuel has increased due to the Fukushima accident so that the accident tolerant fuel (ATF) has been actively developed. The candidate research method to evaluate the in-core performance of fuel is the utilization of research reactor, which has the advantage of simulating normal operating or accident condition of nuclear power plant. In Korea, although HANARO has been used to conduct the in-core performance test of fuel, the fission power of the test fuel during the irradiation could not be intentionally controlled, and it was only dependent on the change of the core variation such as the height of control absorber rod (CAR) and the depletion of HANARO fuel. In this study, the evaluation results for the test method that can control the fission power of test fuel during the irradiation were described to apply in HANARO. There are several methods in the research reactor that can change the fuel power during the irradiation test, such as reactor power control, moving the position of test fuel and the utilization of neutron absorber like He-3. Halden reactor can conduct the test with the control of fuel power by moving fuel position and the utilization of He-3. Jules Horowitz Reactor (JHR), which is currently under construction, will prepare a facility called ADELINE and plan to use the method of inserting and withdrawing the test fuel into the core. Therefore, the control method of fuel power is different for each reactor so that it is necessary to evaluate which method should be applied considering the HANARO characteristics. If the control of fuel power is possible in HANARO, it will be possible to carry out the dynamic safety research, the study of load following behavior and various basic studies.
[en] Since 1990, improving fuel cycle economy has required an extension of the fuel cycle and higher burnup operation. The modern PWR tend to be operated in the severe operating conditions. Moreover, hydrogen from the dissociation of water molecules during oxidation process can be absorbed in Zirconium alloys resulting in hydrogen embrittlement. To date, some systematical research has been conducted regarding hydrogen behavior in Zirconium alloy, including hydrogen absorption and embrittlement. This research indicated that hydrogen absorption has no great influence on the oxidation of cladding at LOCA (Loss of Coolant Accident). However, most of these studies considered only oxidation temperature and time, regardless of the effect of hydride precipitation under high steam pressure. Therefore, this study investigated the oxidation behavior of hydrogen charged Zirconium alloy cladding under high temperature and high steam pressure. This data will give an exact evaluation of the corrosion of fuel cladding. The results can also be used to prevent the problems arising from hydride precipitation in the nuclear fuel cladding materials
[en] Zirconium-alloy cladding is the most important barrier against releasing radioactive fission products in the fuel. At the accidental situation, the inner pressure of the fuel might be higher than the reactor pressure. Once the cladding is under the tension state at high temperatures (i.e., accident cases), the cladding starts to be oxidized with undergoing creep. Hence, creep enhanced oxidation is expected during accidents of nuclear reactor. Nuclear Regulatory Commission (NRC) regulation set LOCA accident criteria. The oxidation amount of cladding during an accident is generally calculated by the Baker-Just (B-J) correlation. But, B-J correlation only considers oxidation temperatures, but, does not considered the effect of creep. Generally it is known that the oxidation rate is accelerated under loaded condition, however; no detail measurements on high temperature oxidation with load have been done so far. Therefore, the goal of this study is to investigate the oxidation behaviors of claddings at high temperature under load. In this study, specifically load is uniaxial
[en] ○ Low temperature (of less than 100℃) irradiation capsule was developed for an irradiation of research reactor core materials (previous lower limit was 250℃). ○ High temperature (up to 900℃) irradiation capsule was developed for an irradiation of future reactor fuels and materials (previous upper limit was 700℃). ○ Long-term irradiation technology using capsule was developed and successfully irradiated for 8 cycles (about 200days, 3 dpa) (previous limit was 4 cycles). ○ The developed long-term and low temperature capsule was utilized for an evaluation of the neutron irradiation properties of the core materials of a research reactor for the National Project of 'Research Reactor Development'. ○ Irradiation analysis technology was improved by evaluation of test variables (Test hole characteristics, position of control rod) and by improvement of analysis code (HANAFMS, MCNP). ○ High performance instrument(LVDT) and related technologies were developed and patented to contribute localization of domestic irradiation instrumentation technology. ○ Neutron irradiation effect on the electro-magnetic properties of the advanced materials (semiconductors(Si,Ge, SiC), superconductor(MgB2), multiferroics(BiFeO)) was investigated and the research possibility of neutron irradiation was discussed.
[en] Nuclear energy currently becomes revitalized as a main energy source due to the rapid rise of oil prices and the growing demand for clean energy sources that do not produce carbon dioxide. However, nuclear energy indispensably produces radioactive nuclear wastes which should be very carefully kept in separation from the boundary of human environments for a long time. Radioactive nuclear wastes should be maintained in a good stability against leaking out for a long period. For this purpose, nuclear power plants generally use a method of the solidification of radioactive wastes for disposal of intermediate and low level radioactive waste. Two forms of solidification are used- cement and paraffin. These waste forms must have good structural durability to allow for permanent disposal under even unfavorable conditions. In contrast to the cement waste form, the paraffin waste form has a high leaching rate. A relatively low stability of the paraffin waste form makes it difficult to use for long-term storage. Therefore in order to solve this leaching problem, it may be recommended to separate the paraffin from the radioactive sludge in paraffin waste form above all. Then, the radioactive sludge can be processed into a more stable form of disposable wastes such as cement waste forms
[en] It is important to maintain the integrity of zirconium alloy fuel cladding because that is a barrier against radioactive fission products. Many researchers investigated the behavior of cladding under loss-of coolant accident (LOCA). When LOCA occurs, cladding temperature rise rapidly and primary system pressure drop to atmospheric range. However, the internal pressure of cladding is preserved ∼MPa because of the generation of fission products and the swelling of fuel pellets. Therefore, the difference of pressure can induce the local ballooning and burst. Some investigators conducted ballooning-burst tests. F. J. Erbacher and S. Leistikow researched the oxidation and embrittlement behavior of zircaloy cladding by simulating a LOCA. F. Erbacher et al. studied single rod and bundle tests in the fuel rod simulator. P. Hofmann investigated the influence of iodine on the burst test, and Jun Hwan Kim et al. studied the deformation of Zircaloy-4(Zry-4) under isothermal and transient-heating tests. Only Zry-4 was used in their study, except for an Nb-added alloy. Also most of the investigations were tested under transient temperatures to simulate a LOCA. Therefore, fundamental data of the ballooning-burst on the uniform temperature is insufficient. In this study, the ballooning-burst behavior of Zry-4 and an Nb-added alloy was studied and fundamental data was produced
[en] The integrity of reactor core materials of commercial nuclear power plant and research reactor is also interested to enhance safety and to produce licensing database. The higher neutron fluence than before are needed to demonstrate the in-core performance of above materials. Since the neutron flux of HANARO, which has been used for the irradiation testing, is respectively low, it is expected to take a long time to meet the required neutron fluence. o increase the neutron flux for the reduction of irradiation period, the application of booster fuel was proposed. The candidate fuels were selected by the consideration of irradiation experience and in-core performance. We conceptually designed booster fuel and irradiation device. It was used as the basic information for the performance evaluation during irradiation. From this evaluation, the performance of rod-type fuel is more effective than the plate-type as the booster fuel. To determine the limitation of uranium loading of rod-type booster fuel, more detailed performance evaluations are needed.
[en] SPND (Self-Powered Neutron Detector) has been developed to extend its lifespan. ENFMS (Ex-Core Flux Monitoring System) of pressurized water reactor has been also improved. After the development and improvement, their performance must be verified under the neutron irradiation environment. We used a research reactor for the performance verification of neutron detector and associated system because the research reactor can meet the neutron flux level of commercial nuclear reactor. In this paper, we report the performance verification method and result for the SPND and ENFMS using the research reactor. The performance tests for the SPND and ENFMS were conducted using UCI TRIGA reactor. The test environment of commercial reactor’s neutron flux level must be required. However, it is difficult to perform the test in the commercial rector due to the constraint of time and space. The research reactor can be good alternative neutron source for the test of neutron detectors and associated system