• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.10 no.3
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• pp.199-208
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• 2012

To develop site characterization technologies for a radioactive waste disposal research in KAERI, the geological and hydrogeological investigations have been carried out since 1997. In 2006, the KURT (KAERI Underground Research Tunnel) was constructed to study a solute migration, a microbiology and an engineered barrier system as well as deeply to understand geological environments in in-situ condition. This study is performed as one of the site characterization works around KURT. Several investigations such as a lineament analysis, a borehole/tunnel survey, a geophyscial survey and logging in borehole, were used to construct the geological model. As a result, the geological model is constructed, which includes the lithological model and geo-structural model in this study. Moreover, from the results of the in-situ hydraulic tests, the hydrogeological properties of elements in geological model were evaluated.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.5 no.2
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• pp.145-154
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• 2007

In hydrothermal reaction tests, smectite-to-illite conversion was identified using a domestic bentonite which is favorably considered as a buffer material, and its dependency on various hydrothermal conditions was investigated. The analysis results of the XRD and Si concentration indicated that the smectite-to-illite conversion was a major process of bentonite alteration under the hydrothermal conditions. The temperature, potassium concentration in solution, and pH were observed to significantly affect the smectite-to illite conversion. A model of conversion reaction rate was suggested to evaluate the long-term stability of smectite composing a major constituent of bentonitic buffer. It was expected from the evaluation results that the smectite would keep its integrity for very long disposal time under a normal condition, whitens it might be converted to illite by 50 percent after over $5{\times}10^4$ year of disposal time under a conservative condition and consequently lose its swelling capacity as a buffer material of a repository.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.10 no.2
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• pp.125-132
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• 2012

The dehydration schemes of rare earth (La, Ce, Nd, Pr, Sm. Eu, Gd, Y) chloride hydrates was investigated by using a dehydration apparatus. To prevent the formation of the rare earth oxychlorides, the operation temperature was changed step by step ($80{\rightarrow}150{\rightarrow}230^{\circ}C$) based on the TGA (thermo-gravimetric analysis) results of the rare earth chloride hydrates. A vacuum pump and preheated Ar gas were used to effectively remove the evaporated moisture and maintain an inert condition in the dehydration apparatus. The dehydration temperature of the rare earth chloride hydrate was increased when the atomic number of the rare earth nuclide was increased. The content of the moisture in the rare earth chloride hydrate was decreased below 10% in the dehydration apparatus.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.10 no.2
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• pp.97-104
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• 2012

A porous $TiO_2$ pellet was electrochemically converted to the metallic titanium by using a $CaCl_2$ molten salt system at $850^{\circ}C$. Ni-$TiO_2$ and graphite electrodes were used as cathode and anode, respectively. The electrochemical behaviour of $TiO_2$ pellet was determined by a constant voltage control electrolysis. Various reaction intermediates such as $CaTiO_3$, $Ti_2O$ and $Ti_6O$ were observed by XRD analysis during electrolysis of the pellet. Once $TiO_2$ pellet was converted to a porous metallic structure, the porous structure disappeared by sintering and shrinking with increasing the reaction time at high temperature.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.14 no.2
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• pp.135-147
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• 2016

The buffer in geological disposal system is one of the major elements to restrain the release of radionuclide and to protect the container from the inflow of groundwater. The buffer material requires long-term stability, low hydraulic conductivity, low organic content, high retardation of radionuclide, high swelling pressure, and high thermal conductivity. These requirements could be determined by the quantitative analysis results. In case of South Korea, the bentonites produced in Gyeongju area have been regarded as candidate buffer/backfill materials at KAERI (Korea Atomic Energy Research Institute) since 1997. According to the study on several physical and chemical characteristics of domestic bentonite in the same district, this is the Ca-type bentonite with about 65% of montmorillonite content. Through this study, we present the criteria for the performance evaluation items and methods when collecting new buffer/backfill materials.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.6 no.4
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• pp.257-263
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• 2008

We analyzed the statistical properties for the conductive background fractures in the Low and Intermediate Level Waste(LILW) disposal site to conceptualize of its groundwater flow system. The background fractures were classified to fracture sets based on their trends and plunges that were obtained from the borehole logging data, and then the fracture transmissivity distribution was inferred from the fixed interval hydraulic test results. The fracture size distribution of each fracture set was estimated using the fracture density and fracture mapping data. To verify the analyzed results, we compared observed field data to simulated one from the DFN model that was constructed with the analyzed statistical properties of the background fractures, and they showed a good agreement.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.12 no.2
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• pp.171-178
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• 2014

After the expiration of operating license of a plant, all infrastructures within the plant must be safely dismantled to the point that it no longer requires measures for radiation protection. Despite the fact that Kori 1 and Wolsong 1 are close to the expiration of their operating license, sufficient technologies for radiological characterization, decontamination and dismantling is still under development. The purpose of this study is to develop one of methods for radiological inventory assessment on measuring object by using direct measure of large component with In-Situ measurement technique. Radiological inventory was assessed by analyzing nuclide using portable gamma spectroscopy without dismantling reactor head, and the result of direct measurement was supplemented by performing indirect measurement. Radiochemical analysis were performed on surface contamination samples as well. During the study, radiological inventory of reactor vessel calculated expanding the result. Based on the result and the radioactivity variation of each radionuclides time frame for decommissioning can be decided. Thus, it is expected that during the decommissioning of plants, the result of this study will contribute to the reduction of radiation exposure to workers.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.18 no.3
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• pp.383-393
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• 2020

The corrosion behavior of Hastelloy C-276 was investigated to identify its applicability for carbon-anode-based oxide reduction (OR), in which Cl₂ and O₂ are simultaneously evolved at the anode. Under a 30 mL·min^-1 Cl₂ + 170 mL·min^-1 Ar flow, the corrosion rate was less than 1 g·m^-2·h^-1 up to 500℃, whereas the rate increased exponentially from 500 to 700℃. The effects of the Cl₂-O₂ composition on the corrosion rate at flow rates of 30 mL·min^-1 Cl₂, 20 mL·min^-1 Cl₂ + 10 mL·min^-1 O₂, and 10 mL·min^-1 Cl₂ + 20 mL·min^-1 O₂ with a constant 170 mL·min^-1 Ar flow rate at 600℃ was analyzed. Based on the data from an 8 h reaction, the fastest corrosion rate was observed for the 20 mL·min^-1 Cl₂ + 10 mL·min^-1 O₂ case, followed by 30 mL·min^-1 Cl₂ and 10 mL·min^-1 Cl₂ + 20 mL·min^-1 O₂. The effects of the chlorine flow rate on the corrosion rate were negligible within the 5-30 mL·min^-1 range. A surface morphology analysis revealed the formation of vertical scratches in specimens that reacted under the Cl₂-O₂ mixed gas condition.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.18 no.3
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• pp.355-362
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• 2020

The corrosion behavior of the Inconel X-750 alloy was investigated for its potential application under a Cl₂-O₂ mixed gas flow in an Ar atmosphere. The corrosion rate was found to be negligible at temperatures up to 400℃ under a flow rate of 30 mL·min^-1 Cl₂ + 170 mL·min^-1 Ar, whereas an exponential increase was observed in the corrosion rate at temperatures greater than 500℃. The suppression of the corrosion reaction due to the presence of O₂ was verified experimentally at flow rates of 30 mL·min^-1 Cl₂ (4.96 g·m^-2·h^-1), 20 mL·min^-1 Cl₂ + 10 mL·min^-1 O₂ (2.02 g·m^-2 ·h^-1), and 10 mL·min^-1 Cl₂ + 20 mL·min^-1 O₂ (1.34 g·m^-2·h^-1) under a constant Ar flow rate of 170 mL·min^-1 at 600℃ for 8 h. The surface morphology analysis results revealed that porous surfaces with tunnel-type holes were produced under the Cl₂-O₂ mixed-gas condition. Furthermore, the effects of the Cl₂ flow rate on the corrosion rate were investigated, indicating that its impact was negligible within the range of 5-30 mL·min^-1 Cl₂ at 600℃.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.8 no.3
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• pp.201-205
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• 2010

Most of iodine was captured by the block when NaI solution flowed through a bentonite block sorbed silver to retard the migration of iodine released from high-level radioactive wastes. In order to understand in detail the mechanism for the retardation of the iodine by the silver ion, X-ray Absorption Near Edge Structure (XANES) and Extended X-ray Absorption Fine Structure (EXAFS) spectra of the silver sorbed bentonite before and after the contact with iodide were compared with those of AgO, $Ag_2O$ and AgI as references. This examination suggests that the silver ion sorbed on the bentonite is desorbed, and then it retards the migration of iodine by forming the cluster of AgI precipitate.