• Applied Chemistry for Engineering
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• v.24 no.2
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• pp.165-170
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• 2013

An external gelation method in place of an internal gelation method applied to the fabrication process of an intermediated compound of Uranium Oxy-Carbide (UCO) kernel spheres for Very High Temperature Reactor (VHTR) fuel preparation is under development in Korea. For the preliminary experiments of the UCO kernel sphere preparation using an external gelation method, the carbon black dispersion experiments were carried out using a simulated broth solution. From the selection experiments of various kinds of carbon black through dispersion experiments in a viscous metal salt solution, Cabot G carbon black was selected owing to its dispersion stability, and the homogeneous dispersing state of carbon black particles in our system. For the effective dispersion of nano-size aggregated carbon black particles in a high viscous liquid, the carbon black particles in a metal salt solution were first de-aggregated with ultrasonic force. The mixed solution was then dispersed secondly by the use of the extremely high-speed agitation with a mechanical mixer of 6000 rpm after feeding the Poly Vinyl Alcohol (PVA) in the solution. This results in the broth solution with good stability and homogeneity alongside no further changes in physical properties.

• Nuclear Engineering and Technology
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• v.34 no.1
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• pp.80-89
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• 2002

A laboratory investigation of the behavior of radioactive metals under the various waste combustion atmospheres was conducted to predict the parameters that influence their partitioning behavior during waste incineration. Neodymium, samarium, cerium, gadolinium, cesium and cobalt were used as non-radioactive surrogate metals that are representative of uranium, plutonium, americium, curium, radioactive cesium, and radioactive cobalt, respectively. Except for cesium, all of the investigated surrogate metal compounds converted into each of their stable oxides at medium temperatures from 400 to 90$0^{\circ}C$, under oxygen- deficient and oxygen-sufficient atmospheres (0.001-atm and 0.21-atm $O_2$). At high temperatures above 1,40$0^{\circ}C$, cerium, neodymium and samarium in the form of their oxides started to vaporize but the vaporization rates were very slow up to 150$0^{\circ}C$ . Inorganic chlorine (NaCl) as well as organic chlorine (PVC) did not impact the volatility of investigated Nd$_2$O$_3$, CoO and Cs$_2$O. The results of laboratory investigations suggested that the combustion chamber operating parameters affecting the entrainment of particulate and filtration equipment operating parameters affecting particle collection efficiency be the governing parameters of alpha radionuclides partitioning during waste incineration.

• Transactions of the Korean hydrogen and new energy society
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• v.24 no.4
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• pp.295-301
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• 2013

Long-term global energy-demand growth is expected to increase driven by strong energy-demand growth from developing countries. Fusion power offers the prospect of an almost inexhaustible source of energy for future generations, even though it also presents so far insurmountable scientific and engineering challenges. One of the challenges is safe handling of hydrogen isotopes. Metal hydrides such as depleted uranium hydride or ZrCo hydride are used as a storage medium for hydrogen isotopes reversibly. The metal hydrides bind with hydrogen very strongly. In this paper, we carried out a modeling and simulation work for absorption/desorption of hydrogen by ZrCo in a horizontal annulus cylinder bed. A comprehensive mathematical description of a metal hydride hydrogen storage vessel was developed. This model was calibrated against experimental data obtained from our experimental system containing ZrCo metal hydride. The model was capable of predicting the performance of the bed for not only both the storage and delivery processes but also heat transfer operations. This model should thus be very useful for the design and development of the next generation of metal hydride hydrogen isotope storage systems.

• Analytical Science and Technology
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• v.16 no.2
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• pp.117-124
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• 2003

Oxygen to metal ratio has been measured by wet and dry chemical analysis to study the properties of sintered $UO_2$ pellets and $U_3O_8$ in the lithium reduction process of spent pressurized water reactor fuels. Uranium dioxide pellets simulated for the spent PWR fuels with burnup values of 20,000~60,000 MWd/MtU were prepared by mixing $UO_2$ powder and oxides of fission product elements, pelleting the powder mixture and sintering it at $1,700^{\circ}C$ under a hydrogen atmosphere. For wet chemical analysis, the simulated spent fuels were dissolved with mixed acid (10 M HCl : 8 M $HNO_3$, 2.5 : 1, v/v) using acid digestion bomb technique. The total amount of uranium and fission products added in the simulated spent fuels were measured using inductively coupled plasma atomic emission spectrometry. Weight change of the simulated fuel during its oxydation was measured by thermogravimetry and then the O/M ratio result was compared to that obtained by wet chemical analysis. Influence of $Mo_{0.4}-Ru_{0.4}-Rh_{0.1}-Pd_{0.1}$, quaternary alloy, on the determination of O/M ratio was investigated.

• Proceedings of the Korean Society for Technology of Plasticity Conference
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• 2003.05a
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• pp.303-306
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• 2003

To reduce the storage space of spent fuels used at the atomic power plants all over the world, the uranium elements contained in the spent fuels is being extracted and effectively stored. For this, the spent fuel are oxidized and deoxidized. In this study, it is produced conceptual design specification about the spent fuel management technology research and test facilities have been produced. The first considered processes in the facilities is the metal conversion furnace in the dry environment. Since this process is operates at the high temperature range, we have to consider heat-resisting designs for the device. For the heat-resisting designs, we have surveyed and analyzed technical references for material properties. Also, we have determined the temperature distribution condition of the device based on experimental results. We have calculated thermal stress and strain of each devices by the commercial analysis software, I-DEAS. By using the results, we have analyzed design configurations of the point at issue by thermal effects, and suggested alternative design configurations. It is experimented for inspecting confidence rate of heat strain. Based on these results, necessary design specifications for heat-resisting design have been produced.

• Nuclear Engineering and Technology
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• v.42 no.2
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• pp.183-192
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• 2010

An electrochemical reduction of a mixture of metal oxides was conducted in a LiCl molten salt containing 3 wt% $Li_2O$ at $650^{\circ}C$. The oxide reduction was carried out by applying a current to an electrolysis cell, and the $Li_2O$ concentration was analyzed during each run. The concentration of $Li_2O$ in the electrolyte bulk phase gradually decreases according to Faraday's law due to a slow diffusion of the $O^{2-}$ ions. A hindrance effect of the unreduced metal oxides was observed for the reduction of the uranium oxide. Cs, Sr, and Ba of high heat-load fission products were diffused into and accumulated in the salt phase as predicted with thermodynamic consideration.

• Journal of environmental and Sanitary engineering
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• v.15 no.1
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• pp.77-87
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• 2000

Several new ion exchange resins have been synthesized from chloromethyl styrene-l,4-divinylbenzine with 1%, 2%, 10% and 20%-crosslink and macrocyclic ligands of cryptand type by interpolymerization method. The adsorption characteristics and the pH, time, solvents and concentration dependence of the adsorption of metal ions by this resin were studied. The correlation between the separation characteristics of uranium and transition metal on the resins and the stability constants of complexes with macrocyclic ligands have been examined. The resins were very stable in both acidic and basic media and have good resistance to heat. The $UO_2^{2+}$ was not adsorbed on the resins below pH 3.0, but the power of adsorption of $UO_2^{2+}$ increased rapidly above pH 4.0. The optimum equilibrium time for adsorption of metallic ions was two hours and adsorptive power decreased in proportion to crosslink size of the resins and order of dielectric constants of solvents used and the selective sequence for metal cations was in the order of $UO_2^{2+}$, $Ni{2+}$ and $Nd{3+}$.

• Nuclear Engineering and Technology
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• v.54 no.8
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• pp.2792-2800
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• 2022

In the core of the WWR-K reactor, a long-term irradiation of tristructural isotopic (TRISO)-coated fuel particles (CFPs) with a UO₂ kernel was carried out under high-temperature gas-cooled reactor (HTGR)-like operating conditions. The temperature of this TRISO fuel during irradiation varied in the range of 950-1100 ℃. A fission per initial metal atom (FIMA) of uranium burnup of 9.9% was reached. The release of gaseous fission products was measured in-pile. The release-to-birth ratio (R/B) for the fission product isotopes was calculated. Aspects of fuel safety while achieving deep fuel burnup are important and relevant, including maintaining the integrity of the fuel coatings. The main mechanisms of fuel failure are kernel migration, silicon carbide corrosion by palladium, and gas pressure increase inside the CFP. The formation of gaseous fission products and carbon monoxide leads to an increase in the internal pressure in the CFP, which is a dominant failure mechanism of the coatings under this level of burnup. Irradiated fuel compacts were subjected to electric dissociation to isolate the CFPs from the fuel compacts. In addition, nondestructive methods, such as X-ray radiography and gamma spectrometry, were used. The predicted R/B ratio was evaluated using the fission gas release model developed in the high-temperature test reactor (HTTR) project. In the model, both the through-coatings of failed CFPs and as-fabricated uranium contamination were assumed to be sources of the fission gas. The obtained R/B ratio for gaseous fission products allows the finalization and validation of the model for the release of fission products from the CFPs and fuel compacts. The success of the integrity of TRISO fuel irradiated at approximately 9.9% FIMA was demonstrated. A low fuel failure fraction and R/B ratios indicated good performance and reliability of the studied TRISO fuel.

• Explosives and Blasting
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• v.19 no.1
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• pp.101-110
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• 2001

Depleted uranium (DU) outperforms tungsten heavy alloys (WHA) by about 10%. Because of environmental and hence, political concerns, there is a need to improve WHA performance, in order to replace the DU penetrators. A technique of metal powder compaction by the detonation of an explosive has been applied to tungsten-titanium(W-Ti) powder materials that otherwise may be difficult to fabricate conventionally or have dissimilar, nonequilibrium, or unique me1astab1e substructures. However, the engineering properties of compacted materials are not widely reported and are little known especially for the "unique" composition of W-Ti alloy. To develop high-performance tungsten composites with superior ballistic attributes, it is necessary to understand, carefully document controlled experimental results, and develop basic computational models for potential composites with controlled microstructures. A detailed understanding and engineering application of W-Ti alloy can lead to the development of new structural design for engineering components and materials.

• Economic and Environmental Geology
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• v.31 no.5
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• pp.365-374
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• 1998

Economic U- and Th-bearing pegmatite deposits occur in the Kyemyeongsan Formation, and are spatially closely associated with the Carboniferous alkali granite. The pegmatite is lithochemically alkaline and peralumious, and consists mainly of potassic feldspar and quartz with allanite and U- and Th-bearing minerals. Paragenetic stages of mineralization in the pegmatite are divided as follows: early silicate mineralization, main rare metal mineralization, and late silicate mineralization. Thorite, euxenite, fergusonite and uranpyrochlore are the predominant U- and Th-bearing minerals. Both the enrichments of Nb, Y, Th, U, and Ta and the depletions of Hf, Ba, and Rb in the pegmatite were resulted from magmatic differentiation. The increases of Na and Ca in uranpyrochlore, of Th and U in fergusonite, of Si, Th, U and Pb in thorite, and of Nb and Y in euxenite were possibly resulted from both later internal fractionation and hydrothermal alteration. The variation of chemical composition in a mineral species reflects the different pysico-chemical conditions during the crystallization.