• Korean Chemical Engineering Research
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• v.52 no.3
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• pp.279-288
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• 2014

Nuclear energy is expected to meet the growing energy demand while avoiding CO2 emission. However, the problem of accumulating spent fuel from current nuclear power plants which is mainly composed of uranium oxides should be addressed. One of the most practical solutions is to reduce the spent oxide fuel and recycle it. Next-generation fuel cycles demand innovative features such as a reduction of the environmental load, improved safety, efficient recycling of resources, and feasible economics. Pyroprocessing based on molten salt electrolysis is one of the key technologies for reducing the amount of spent nuclear fuel and destroying toxic waste products, such as the long-life fission products. The oxide reduction process based on the electrochemical reduction in a LiCl-$Li_2O$ electrolyte has been developed for the volume reduction of PWR (Pressurized Water Reactor) spent fuels and for providing metal feeds for the electrorefining process. To speed up the electrochemical reduction process, the influences of the feed form for the cathode and the type of anode shroud on the reduction rate were investigated.

• Journal of the Korean Ceramic Society
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• v.32 no.4
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• pp.471-481
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• 1995

Unirradiated UO2 pellets were pulverized by oxidation in air at 40$0^{\circ}C$, and the oxidized powders were reduced in H2 and CO atmospheres at $600^{\circ}C$. During the oxidation of UO2 at 40$0^{\circ}C$, intergranular cracks which caused the spallation were mainly developed by the volume contraction due to the formation of intermediate phase (U4O9 or U3O7). As oxidation proceeded, U3O8 finally formed. As the oxidation/reduction cycles were repeated, the powder surface became coarser, specific surface area was increased and average particle size was decreased. The sintered densities of the powder were increased by the oxidation/reduction cycle due to the characteristic changes of the powder.

• Nuclear Engineering and Technology
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• v.52 no.4
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• pp.674-680
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• 2020

We used MCNP6 computer code to model HTR-10 core reactor. We used two types of fuel; UO₂ and (Th+Pu)O₂ mixture. We determined the critical height at which the reactor approached criticality in both two cases. The neutronic and burnup parameters were investigated. The results indicated that the core fueled with mixed (Th+Pu)O₂, achieved about 24% higher fuel cycle length than the UO₂ case. It also enhanced safeguard security by burning Pu isotopes. The results were compared with previously published papers and good agreements were found.

• Proceedings of the Korean Institute of Building Construction Conference
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• 2018.05a
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• pp.224-225
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• 2018

Modern people are more interested in the indoor environment as they spend more time indoors than in the past. Among the air pollutants in the indoor air, ladon gas is a colorless, tasteless, odorless, inert gas produced by nuclear decomposition of naturally occurring uranium in rocks and soils. It has been proven that ladon gas is introduced into the room through cracks on the floor of the building or basement wall, and it causes various diseases such as lung cancer when exposed to radon during human breathing. The US Environmental Protection Agency (EPA) specifies 4pCi / L as a necessary measure for radon, and the Korea Environmental Protection Agency has implemented comprehensive indoor radon management measures since 2007. Therefore, in this study, we intend to adsorb and reduce radon in indoor air pollutants.

• Nuclear Engineering and Technology
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• v.46 no.6
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• pp.863-868
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• 2014

The evaluation of thermal properties of actinide oxide fuels is a problem of high importance for the development of new generation reactors. In the present study, an expression obtained for n-dimensional Debye functions is used to derive a simple analytical expression for the specific heat capacity of nuclear fuels. To test the validity and reliability of this expression, the analytical expression is applied to $UO_2$, $NpO_2$, $ThO_2$, and $PuO_2$. It is seen that the formula was in agreement with the experimental and theoretical results reported in the literature.

• Bulletin of the Korean Chemical Society
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• v.21 no.12
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• pp.1187-1192
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• 2000

Ternary (Mo-Ru-Pd) and binary (Mo-Ru, Mo-Pd) alloys have been prepared using an Ar arc melting furnace. Mo and the noble metals, Ru and Pd, are the constituents of metallic insoluble residues, which were found in the early days of post-irradiation studies on uranium oxide fuels. In the present study, the structure of the alloys was evaluated using a powder X-ray diffractometer. Unit cell parameters were determined by least squares refinements of powder X-ray diffraction data. Scanning electron microscopic analyses of the surface of the alloys indicated that surface morphology was dependent on the crystallographic structure as well as its composition. Measurements of the magnetic susceptibility of the alloys showed evidence of superconducting transition from 3 to 9.2 K. Among the ternary and binary alloys, the ${\sigma}-phase$ showed the highest superconducting transition temperature,~9.2 K.

• Nuclear Engineering and Technology
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• v.54 no.5
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• pp.1541-1548
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• 2022

The thermal conductivity of uranium oxide (UO₂) containing pores and grain boundaries is investigated using continuum-level simulations based on the finite-difference method in two and three dimensions. Steady-state heat conduction is solved on microstructures generated from the phase-field model of the porous polycrystal to calculate the effective thermal conductivity of the domain. The effects of porosity, pore size, and grain size on the effective thermal conductivity of UO₂ are quantified. Using simulation results, a new empirical model is developed to predict the effective thermal conductivity of porous polycrystalline UO₂ fuel as a function of porosity and grain size.

• Nuclear Engineering and Technology
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• v.51 no.2
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• pp.631-640
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• 2019

Thorium ($^{232}Th$) is four times more abundant than uranium in nature and has become a new important source of energy in the future. This is due to the ability of thorium to undergo the bombardment of neutron to produce uranium-233 ($^{233}U$). The aim of this study is to investigate the production cost of thorium oxide ($ThO_2$) resulted from the thorium extraction process. Four main parameters were studied which include raw material and chemical cost, total capital investment, direct cost and indirect cost. These parameters were justified to obtain the final production cost for the thorium extraction process. The result showed that the raw material costs were $63,126.00 - $104,120.77 (0.5 ton), $126,252.00 - $178,241.53 (1.0 ton), and $1,262,520.00 - $1,782,415.33 (10.0 tons). The total installed equipment and total cost investment were estimated to be approximately $11,542,984.10 and $13,274,431.715 respectively. Hence, the total costs for producing 1 kg $ThO_2$ were $6829.79 - $6911.78, $3540.95 - $3592.94, and $501.18 - $553.17 for 0.5, 1.0, and 10.0 tons respectively. The result concluded that with higher mass production, the cost of 1 kg $ThO_2$ would be reduced which in this scenario, the lowest production cost was $$501.18kg^{-1}$-$$553.17kg^{-1}$ for 10.0 tons of $ThO_2$ production.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.8 no.1
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• pp.19-32
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• 2010

Electrolytic reduction technology is essential for the purpose of adopting pyroprocessing into spent oxide fuel as an alternative option in a back-end fuel cycle. Spent fuel consists of various metal oxides, and each metal oxide releases an oxygen element depending on its chemical characteristic during the electrolytic reduction process. In the present work, an electrolytic reduction behavior was estimated for voloxidized spent fuel based on the assumption that each metal-oxygen system is independent and behaves as an ideal solid solution. The electrolytic reduction was considered as a combination of a Li recovery and chemical reactions between the metal oxides such as uranium oxide and the produced Li metal. The calculated result revealed that most of the metal oxides were reduced by the process. It was evaluated that a reduced fraction of lanthanide oxides increased with a decreasing $Li_2O$ concentration. However, most of the lanthanides were expected to be stable in their oxide forms. In addition, a semi-empirical model for describing $U_3O_8$ electrolytic reduction behavior was proposed by considering Li diffusion and a chemical reaction between $U_3O_8$ and Li. Experimental data was used to determine model parameters and, then, the model was applied to calculate the reduction yield with time and to estimate the required time for a 99.9% reduction.

• Nuclear Engineering and Technology
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• v.48 no.6
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• pp.1330-1337
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• 2016

A visualization test of the fuel-coolant interaction in the Test for Real cOrium Interaction with water (TROI) test facility was carried out. To experimentally simulate the In-Vessel corium Retention (IVR)- External Reactor Vessel Cooling (ERVC) conditions, prototypic corium was released directly into the coolant water without a free fall in a gas phase before making contact with the coolant. Corium (34.39 kg) consisting of uranium oxide and zirconium oxide with a weight ratio of 8:2 was superheated, and 22.54 kg of the 34.39 kg corium was passed through water contained in a transparent interaction vessel. An image of the corium jet behavior in the coolant was taken by a high-speed camera every millisecond. Thermocouple junctions installed in the vertical direction of the coolant were cut sequentially by the falling corium jet. It was clearly observed that the visualization image of the corium jet taken during the fuel-coolant interaction corresponded with the temperature variations in the direction of the falling melt. The corium penetrated through the coolant, and the jet leading edge velocity was 2.0 m/s. Debris smaller than 1 mm was 15% of the total weight of the debris collected after a fuel-coolant interaction test, and the mass median diameter was 2.9 mm.