• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.12 no.1
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• pp.19-35
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• 2014

A parametric sensitivity to the annual exposure dose rate to the farming exposure group has been probabilistically carried out for three principal elements associated with the nuclide transport behavior in the near-field of the pyroprocessed waste repository system. Credit time for both metal and ceramic containers, annual nuclide release rete, and the degree of loss of bentonite buffer around the container are selected as the elements and investigated for important nuclides. All the elements are shown to be sensitive to the results. Methodology studied through this study and the results are expected to make a good feedback to the repository design. As a follow-up study, separated in Part 2, the A-KRS will be deterministically assessed and then compared among each other with the normal, the worst, and the best case scenarios associated with their extreme values these elements could have.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.16 no.3
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• pp.309-313
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• 2018

In this study, we prepared Li-K-Cd alloy, which meets the requirement of eutectic ratio of Li:K, to maintain the operating temperature of the drawdown process at $500^{\circ}C$ and to achieve the reuse of LiCl-KCl molten salt. The prepared Li-K-Cd alloys were added to LiCl-KCl salt bearing U and Nd at $500^{\circ}C$ to investigate the removal of $UCl_3$ in the salt. The reduction of $UCl_3$ in the salt was examined by measuring the OCP value of salt and analyzing the salt composition by ICP-OES. Reduction was also visually confirmed by change of salt color from dark purple to white. The experimental results reveal that the prepared Li-K-Cd alloy has reductive extractability for $UCl_3$ in salt. By improving the preparation method, the Li-K-Cd alloy can be applied to the drawdown process.

• Journal of the Korean Ceramic Society
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• v.39 no.6
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• pp.576-581
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• 2002

In order to simultaneously vitrify the ton Exchange Resin(IER) and Dry Active Waste(DAW) generated from the Nuclear Power Plants, a vitrification pilot test was conducted using an induction cold crucible melter. The PCT result evaluating the chemical durability of the vitrified from showed that the final glass was more durable than the benchmark glass. Liquidus temperature for the final vitrified form was 1048 K(775$\^{C}$) fur heat treatment experiments. The value of the compressive strength for the vitrified form was ninety times higher than the regulation limit, 34 kg/㎠. The glasses on bottom, middle and top of the CCM were homogeneous with no secondary phase. The precipitation of the magnetic metal phase was able to be avoided by simultaneously fEeding of DAW with IER containing strongly reducing organics. Volume reduction factor of 74 was achieved through the vitrification Pilot test for mixed waste.

• Nuclear Engineering and Technology
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• v.43 no.6
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• pp.557-566
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• 2011

A cold spray coating (CSC) of copper was studied for its application to a high-level radioactive waste (HLW) disposal canister. Several copper coatings of 10 mm thick were fabricated using two kinds of copper powders with different oxygen contents, and SS 304 and nodular cast iron were used as their base metal substrates. The fabricated CSC coppers showed a high tensile strength but were brittle in comparison with conventional non-coating copper, hereinafter defined to as "commercial copper". The corrosion behavior of CSC coppers was evaluated by comparison with commercial coppers, such as extruded and forged coppers. The polarization test results showed that the corrosion potential of the CSC coppers was closely related to its purity; low-purity (i.e., high oxygen content) copper exhibited a lower corrosion potential, and high-purity copper exhibited a relatively high corrosion potential. The corrosion rate converted from the measured corrosion current was not, however, dependent on its purity: CSC copper showed a little higher rate than that of commercial copper. Immersion tests in aqueous HCl solution showed that CSC coppers were more susceptible to corrosion, i.e., they had a higher corrosion rate. However, the difference was not significant between commercial copper and high-purity CSC copper. The decrease of corrosion was observed in a humid air test presumably due to the formation of a protective passive film. In conclusion, the results of this study indicate that CSC application of copper could be a useful option for fabricating a copper HLW disposal canister.

• Journal of the Korean Society of Systems Engineering
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• v.17 no.1
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• pp.11-20
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• 2021

Decontamination of systems, structures and components (SSC) during the decommissioning of a Nuclear Power Plant (NPP) can be for a variety of reasons. The main reasons for decontamination are: to reduce the contamination of SSC to a reasonably low level, to reduce the potential for the spread of contaminants into the environment and to reduce the cost of disposal due to the reduced level of contamination in a particular SSC. The decontamination technique can be aggressive or non-aggressive depending on the intent after the decontamination process. Aggressive decontamination technique is used when the intent is not to reuse the SSC while a non-aggressive decontamination technique is used with the intent of SSC reuse. For different SSCs there are different decontamination techniques that can be used, each having its own advantages and drawbacks. Metal components such as pipes in the nuclear power plant account for a large amount of nuclear wastes generated. Some of these wastes can be reused if the contaminant level is reduced to an acceptable level. Laser ablation is a non-aggressive decontamination technique that can be used to reduce the contamination in pipes to an acceptable level with no secondary waste generated during the process. The operation and control of a laser ablation device must be precise to achieve a high decontamination factor. This precision can be achieved by a well-designed motion control system. For this purpose, a motion control system was developed consisting of two parts: the first part being the precise control of the laser ablation device inside the pipe and the second part is the control of the laser ablation device outside the pipe. This paper describes the Systems Engineering approach for the development process of a motion control system for the Laser decontamination system.

• Korean Journal of Environmental Biology
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• v.34 no.2
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• pp.107-115
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• 2016

Currently, there are 442 operating nuclear power plants in the world, and 62 more are under construction. According to this reasoning, the treatment of radioactive waste is important to prevent the environmental ecosystem including humans, animals, and plants. Especially, a leakage of radioactive waste causes not only regional problem but also serious global one. In this study, we demonstrate the effect of radioisotopes (e.g., cesium, strontium, and cobalt) on a 3D culture cell. To develop the 3D cell culture system, we used a 96-well-culture plate with biocompatible agarose hydrogel. Using this method, we can perform the 3D cell culture system with three different cell lines such as HeLa, HepG2, and COS-7. In addition, we conducted a cell viability test in the presence of radioisotopes. Interestingly, the 3D morphological cells showed 42% higher cell viability than those on the 2D against cesium. This result indicates that the 3D platform provides cells morphological and physiological characteristic similar to in vivo grown tissues. Moreover, it overcomes the limitation of conventional cell culture system that can't reflect in vivo systems. Finally, we believe that the proposed approach can be applied a new strategy for simple high-throughput screening and accurate evaluation of metal toxicity assay.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2005.06a
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• pp.137-144
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• 2005

Silica in primary system combines with an alkali grammatical particle metal and forms the zeolite layer which is hindering the heat transfer on the surface of the cladding. Zeolite layer becomes the cause of the damage in this way. The problems of the NPP's primary system have been issued steadily by EPRI. Through a series of experiments of the laboratory scale, we confirmed the applicability of NF membrane for silica removal, as silica rejection rate of NF membrane is about $60\;{\sim}\;70\%$ and boron rejection rate is about $10\;{\sim}\;20\%$. We accomplished a site experiment about four NF membranes manufactured by FilmTec and Osmonics Inc. In experiment using 400L of SFP water, when operation pressure is $10kg_{f}/cm^2$, we confirmed that the silica rejection rate of NF90-2540 manufactured by FilmTec Inc. is about $98\%$, boron rejection rate is about $43\%$. The silica rejection rate of NF270-2540 is about $38\%$, boron rejection rate is about $3.5\%$. Afterward, through additional experiments, such as long term characteristic experiments, we are going to design a optimum NF membrane system for silica removal.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.12 no.3
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• pp.179-189
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• 2014

Pyroprocessing has been developed for the purpose of resolving the current spent nuclear fuel management issue and enhancing the recycle of valuable resources. An electrolytic reduction of the pyroprocessing is a process to reduce oxides into metals using LiCl as an electrolyte and requires a post-treatment process due to the inclusion of residual salt in porous metal products. A vacuum distillation has been adopted for various molten salt systems and could be applied to the post-treatment process of the electrolytic reduction. The residual salt in the metal products includes LiCl, alkali chlorides, and alkaline earth chlorides. In this paper, vapor pressures of chlorides have been estimated and the composition changes on the residual liquid during the vacuum distillation process have been calculated. A model combining a material balance and vapor-liquid equilibrium relations has been proposed under a constant vapor discharging flow rate and liquid composition changes have been calculated using the vapor pressures with respect to a dimensionless time. The behaviors have been compared with temperature and molten salt composition changes to simulate the process condition variation. The distillation of the residual salt has been dominated by LiCl which is the main component of the salt and CsCl of which vapor pressure is higher than that of LiCl would be readily removed. RbCl exhibits similar vapor pressure with LiCl and maintains its composition. However, $SrCl_2$ and $BaCl_2$ of which vapor pressures are much lower than that of LiCl are concentrated with time and expected to be possibly precipitated during the distillation when the initial compositions are increased.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.3 no.2
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• pp.105-112
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• 2005

The electrolytic reduction of uranium oxide in a LiCl-Li$_{2}$O molten salt system has been studied in a 10 g U$_{3}$O$_{8}$ /batch-scale experimental apparatus with an integrated cathode assembly at 650$^{\circ}C$. The integrated cathode assembly consists of an electric conductor, the uranium oxide to be reduced and the membrane for loading the uranium oxide. From the cyclic voltammograms for the LiCl-3 wt$\%$ Li$_{2}$O system and the U$_{3}$O$_{8}$-LiCl-3 wt$\%$ Li$_{2}$O system according to the materials of the membrane in the cathode assembly, the mechanisms of the predominant reduction reactions in the electrolytic reactor cell were to be understood; direct and indirect electrolytic reduction of uranium oxide. Direct and indirect electrolytic reductions have been performed with the integrated cathode assembly. Using the 325-mesh stainless steel screen the uranium oxide failed to be reduced to uranium metal by a direct and indirect electrolytic reduction because of a low current efficiency and with the porous magnesia membrane the uranium oxide was reduced successfully to uranium metal by an indirect electrolytic reduction because of a high current efficiency.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.6 no.1
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• pp.17-23
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• 2008

In this study, plasma processing of metal surface is experimentally investigated to enhance the surface decontamination efficiency and to find out the reaction mechanism. Cobalt, the major contaminant in the nuclear facilities, and three fluorine-containing gases, $CF_4/O_2$, $SF_6/O_2$, and $NF_3$ are chosen for the investigation. Thin metallic disk specimens are prepared and their surface etching reactions with the three plasma gases are examined. Results show that the maximum etching rate of $17.2\;{\mu}m/min.$ is obtained with NF3 gas at $420^{\circ}C$, while with $CF_4/O_2$, $SF_6/O_2$ gas plasmas those of $2.56\;{\mu}m/min.$ and $1.14\;{\mu}m/min.$ are obtained, respectively. Along with etching experiments, constituent elements of the reaction products are identified to be cobalt, oxygen, and fluorine by AES (Auger Electron Spectroscopy) analysis. It turns out that the oxygen atoms are physically adsorbed ones to the surface from the ambient not participation ones during the analysis after reaction, which supports that the surface reaction of cobalt is mainly to be a fluorination reaction.