• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.7 no.3
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• pp.183-190
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• 2009

The basic function of HLW disposal system is to prevent excessive radio-nuclides being leaked from the repository in a short time. To do this, many technical standards should be developed and established on the components of disposal system. Safety assessment of a repository is considered as one of technical standards, because it produces quantitative results of the future evolution of a repository based on a reasonably simplified model. In this paper, we investigated other countries' regulations related to safely assessment focused on the assessment period, radiation dose limits and uncertainties of the assessment. Especially, in the investigation process of the USA regulations, the USA regulatory bodies' approach to assessment period and peak dose is worth taking into account in case of a conflict between peak dose from safety assessment and limited value in regulation.

• 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.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.1
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• pp.85-89
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• 2010

The uranium ingot casting process is one of the steps which consolidate uranium deposits produced by electrorefiner as an ingot form in a pryprocessing technique. This paper introduces new design concept of the ingot casting equipment and the performance test results of the lab-scale ingot casting equipment fabricated based on the design concept. Casting equipment produces the uranium ingot by pouring an uranium melt into a mold by tilting a melting crucible. Also it is equipped with a cup which is able to continuously feed uranium deposits into a melting crucible. The productivity could be significantly enhanced by introducing the continuous operation concept.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.8 no.1
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• pp.71-76
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• 2010

The electrorefining process, one of main processes which are composed of pyroprocess to recover the useful elements from spent fuel, and the domestic development of electrorefiner have been reviewed. The electrorefiner is composed of an anode basket containing reduced spent fuel such as uranium, transuranic and rare earth elements, and a solid cathode, which are in LiCl-KCl eutectic electrolyte. Oxidation (dissolution) reaction occurs on the anode and a pure uranium is electrochemically reduced (deposited) on the solid cathode. By application of graphite cathode, which has a self-scrapping characteristics for the uranium deposits, and a recovery of the fallen deposits by a screw conveyer, a high-throughput continuous electrorefiner with a capacity of 20 kgU/day has been developed.

• Journal of Nuclear Fuel Cycle and Waste Technology
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• v.1 no.1
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• pp.65-73
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• 2013

Temperature-dependent hydrolysis behaviors of aqueous U(VI) species were investigated with time-resolved laser fluorescence spectroscopy (TRLFS) in the temperature range from 15 to $75^{\circ}C$. The formation of four different U(VI) hydrolysis species was measured at pHs from 1 to 7. The predominant presence of $UO{_2}^{2+}$, $(UO_2)_2(OH){_2}^{2+}$, $(UO_2)_3(OH){_5}^+$, and $(UO_2)_3(OH){_7}^-$ species were identified based on the spectroscopic properties such as fluorescence wavelengths and fluorescence lifetimes. With an increasing temperature, a remarkable decrement in the fluorescence lifetime for all U(VI) hydrolysis species was observed, representing the dynamic quenching behavior. Furthermore, the increase in the fluorescence intensity of the further hydrolyzed U(VI) species was clearly observed at an elevated temperature, showing stronger hydrolysis reactions with increasing temperatures. The formation constants of the U(VI) hydrolysis species were calculated to be $log\;K{^0}_{2,2}=-4.0{\pm}0.6$ for $(UO_2)_2(OH){_2}^{2+}$, $log\;K{^0}_{3,5}=-15.0{\pm}0.3$ for $(UO_2)_3(OH){_5}^+$, and $log\;K{^0}_{3,7}=-27.7{\pm}0.7$ for $(UO_2)_3(OH){_7}^-$ at $25^{\circ}C$ and I = 0 M. The specific ion interaction theory (SIT) was applied for the extrapolation of the formation constants to infinitely diluted solution. The results of temperature-dependent hydrolysis behavior in terms of the U(VI) fluorescence were compared and validated with those obtained using computational methods (DQUANT and constant enthalpy equation). Both results matched well with each other. The reaction enthalpies and entropies that are vital for the computational methods were determined by a combination of the van't Hoff equation and the Gibbs free energy equation. The temperature-dependent hydrolysis reaction of the U(VI) species indicates the transition of a major U(VI) species by means of geothermal gradient and decay heat from the radioactive isotopes, representing the necessity of deeper consideration in the safety assessment of geologic repository.

• 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.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.8 no.2
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• pp.107-113
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• 2010

In this study, the adsorption reactions in the binary component system such as kaolinite-humic acid, kaolinite-americium and humic acid-americium were investigated. After performing the basic physico-chemical properties of the kaolinite, the adsorption reactions of the humic acid on the kaolinite were carried out with varying concentration of humic acid and ion strength, and pH. With increasing HA concentration and pH, the sorption of HA onto KA decreased, while the sorption of HA onto KA increased with increasing ionic stre ngth. Also, with varying pH, the adsorption reactions of the americium-kaolinite and americium-humic acid were studied. In the acid and neutral region, Am easily adsorbed on the HA, while the sorption of Am on the HA in the alkali region decreased because of electrostatic repulsion. The results from these studies make it possible to understand the characteristics of adsorption behaviour of the americium by the humic acid in the water environment.