• Nuclear Engineering and Technology
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• 제15권3호
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• pp.197-206
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• 1983

The slightly hyperstoichiometric uranium dioxide, i.e. U $O_{2.005}$ and U $O_{2.01}$ within a range of the requirement for the use of a nuclear fuel, were sintered directly in an atmosphere of $CO_2$/CO mixture without any succeeding reduction process. The kinetics of sintering in the late stage were investigated for various O/U ratios. A sintering diagram, which show the relation of Temperature-Time-Density-Grain size, was established for each O/U ratio. Only by controlling the oxygen partial pressure in the sintering atmosphere, U $O_2$ pellet could be sintered very easily at low temperature 1050$^{\circ}$～120$0^{\circ}C$ with a density above 95% T.D. and average grain size above 7${\mu}{\textrm}{m}$. It was found that the rate of grain growth follows D=(Kt)$^{1}$4/ in the late stage of sintering. And the activation energies for grain growth in the final sintering stage were found to be 75, 64 and 62kca1/mo1 for U $O_{2.005}$, U $O_{2.01}$ and U $O_{2.10}$, respectively. Although no significant differences are obtained between the activation energies for different O/U ratios, the sinterability is enhanced considerably with increasing the oxygen partial pressure in the sintering atmosphere.tmosphere.

• Economic and Environmental Geology
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• 제56권5호
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• pp.603-618
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• 2023

This study focused on evaluating the suitability of the WRK (waste repository Korea) bentonite as a buffer material in the SNF (spent nuclear fuel) repository. The U (uranium) adsorption/desorption characteristics and the adsorption mechanisms of the WRK bentonite were presented through various analyses, adsorption/desorption experiments, and kinetic adsorption modeling at various pH conditions. Mineralogical and structural analyses supported that the major mineral of the WRK bentonite is the Ca-montmorillonite having the great possibility for the U adsorption. From results of the U adsorption/desorption experiments (intial U concentration: 1 mg/L) for the WRK bentonite, despite the low ratio of the WRK bentonite/U (2 g/L), high U adsorption efficiency (>74%) and low U desorption rate (<14%) were acquired at pH 5, 6, 10, and 11 in solution, supporting that the WRK bentonite can be used as the buffer material preventing the U migration in the SNF repository. Relatively low U adsorption efficiency (<45%) for the WRK bentonite was acquired at pH 3 and 7 because the U exists as various species in solution depending on pH and thus its U adsorption mechanisms are different due to the U speciation. Based on experimental results and previous studies, the main U adsorption mechanisms of the WRK bentonite were understood in viewpoint of the chemical adsorption. At the acid conditions (＜pH 3), the U is apt to adsorb as forms of UO₂²⁺, mainly due to the ionic bond with Si-O or Al-O(OH) present on the WRK bentonite rather than the ion exchange with Ca²⁺ among layers of the WRK bentonite, showing the relatively low U adsorption efficiency. At the alkaline conditions (>pH 7), the U could be adsorbed in the form of anionic U-hydroxy complexes (UO₂(OH)₃^-, UO₂(OH)₄^2-, (UO₂)₃(OH)₇^-, etc.), mainly by bonding with oxygen (O^-) from Si-O or Al-O(OH) on the WRK bentonite or by co-precipitation in the form of hydroxide, showing the high U adsorption. At pH 7, the relatively low U adsorption efficiency (42%) was acquired in this study and it was due to the existence of the U-carbonates in solution, having relatively high solubility than other U species. The U adsorption efficiency of the WRK bentonite can be increased by maintaining a neutral or highly alkaline condition because of the formation of U-hydroxyl complexes rather than the uranyl ion (UO₂²⁺) in solution,and by restraining the formation of U-carbonate complexes in solution.

• Polymer(Korea)
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• 제27권3호
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• pp.242-248
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• 2003

PP-g-(AN/Sty) was synthesized by grafting with acrylonitrile (AN) and styrene (Sty) onto PP staple fiber using an electron beam accelerator and followed by amidoximination and phosphorylation. Mole fraction of AN in the graft chain increased with the increase of the AN content in the monomer mixture. The highest AN grafting yield of 45% was obtained at a monomer ratio of 40 vol% AN/60 vol% Sty. Mole fraction of AN in the graft chain decreased with the increase of methanol amount used its solvent. As reaction temperature increased, the grafting yield of copolymer increased and reached equilibrium at 50$^{\circ}C$. Amount of amidoxime group in fibrous ion exchanger was increased as increasing amount of hydroxylamine, and the maximum content of amidoxime group was observed at 5.8 mmol/g with the 9 wt% hydroxylamine concentration. Content of phosphorous group in fibrous ion exchanger increased up to 0.5 N phosphoric acid concentration, and then leveled off. The adsorption ability of the copolymer for uranyl ion by the chelating adsorbents was in the following order : bifunctional PP-g-(AN/sty) > amidoximated PP-g-(AN/Sty) > phosphorylated PP-g-(AN/Sty).

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• 제16권3호
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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 KSNVE
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• 제6권2호
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• pp.163-177
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• 1996

An investigation on thermohydraulic stability of flow oscillations in the CANada Deuterium Uranium-600(CANDU-6) heat transport system has been conducted. Flow oscillations in reactor coolant loops, comprising two heat sources and two heat sinks in series, are possibly caused by the response of the pressure to extraction of fluid in two-phase region. This response consists of two contributions, one arising from mass and another from enthalpy change in the two-phase region. The system computer code used in the investigation os SOPHT, which is capable of simulating steady states as well as transients with varying boundary conditions. The model was derived by linearizing and solving one-dimensional, homogeneous single- and two-phase flow conservation equations. The mass, energy and momentum equations with boundary conditions are set up throughout the system in matrix form based on a node-link structure. Loop stability was studied under full power conditions with interconnecting the two compressible two phase regions in the figure-of-eight circuit. The dominant function of the interconnecting pipe is the transfer of mass between the two-phase regions. Parametric survey of loop stability characteristics, i. e., damping ratio and period, has been made as a function of geometrical parameters of the interconnection line such as diameter, length, height and orifice flow coefficient. The stability characteristics with interconnection line has been clarified to provide a simple criterion to be used as a guide in scaling of the pipe.

• Nuclear Engineering and Technology
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• 제48권6호
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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.

• Nuclear Engineering and Technology
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• 제31권3호
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• pp.335-343
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• 1999

The effects of N $b_2$ $O_{5}$ and oxygen potential on the densification and grain growth of U $O_2$ fuel have been investigated.0.3 wt% N $b_2$ $O_{5}$ -doped U $O_2$fuel pellets were sintered at 1$700^{\circ}C$ for 4 hours in sintering atmospheres which have various ratios of $H_2O$ to $H_2$ gas. Compared with those of undoped U $O_2$ pellets, the sintered density and grain size of the 0.3 wt% N $b_2$ $O_{5}$ -doped U $O_2$ pellet increase under the $H_2O$/ $H_2$ gas ratio of 5.0$\times$10$^{-3}$ to 1.0$\times$10$^{-2}$ and under the $H_2O$/ $H_2$gas ratio of 5.0$\times$10$^{-3}$ to $1.5\times$10$^{-2}$ , respectively. The sintering of U $O_2$fuel pellets containing 0.1 wt% to 0.5 wt% N $b_2$ $O_{5}$ was carried out at 168$0^{\circ}C$ for 4 hours. The enhancing effect of N $b_2$ $O_{5}$ on the sintered density and grain size becomes larger as the N $b_2$ $O_{5}$ content increases. The solubility limit of N $b_2$ $O_{5}$ in U $O_{2}$ seems to be between 0.3 wt% and 0.5 wt%, and beyond the solubility limit the second phase whose composition corresponds near to N $b_2$U $O_{6}$ is precipitated on grain boundary. The enhancement of densification and grain growth in U $O_2$ is attributed to the increased concentration of a uranium vacancy which is formed by the interstitial N $b^{4+}$ ion in the U $O_2$ lattice.

• Economic and Environmental Geology
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• 제22권1호
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• pp.35-63
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• 1989

Geochemical characteristics of the Guryongsan (Ogcheon) uraniferous black slate show that this is an analogue to the conventional Chattanooga and Alum shales in occurrences. Whereas, its highest enrichment ratio in metals including uranium, among others, is explained by the cyclic sedimentation of the black muds and quartz-rich silts, and the uniform depositional condition with some what higher pH condition compared to the conditions of the known occurrences. The cyclic sedimentation, caused by the periodic open and close of the silled basin, has brought about the flush-out) of the uranium depleted water and the recharge with the new metal-rich sea water, which consequently contributed to the high concentration of metals in mud. The metal-rich marine black muds, which mostly occur in the early to middle Palaeozoic times, is attributed by the geologic conditions which related to the atmospheric oxygen contents, and these are scarcely met in the late Precambrian and/or with the onset of Palaeozoic era in the geologic evolution of the earth.

• The Journal of Engineering Geology
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• 제31권4호
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• pp.659-669
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• 2021

The purpose of this study is to elucidate the relationship between occurrence of natural radioactive materials such as ²³⁸U and ²²²Rn and original mixing ratio of helium isotope of groundwater from various geology, and to suggest the underground aquifer environment from helium original mixing data. 9 groundwater samples were collected from five study areas, and ²³⁸U, Rn-222 and helium isotope were analyzed. A high ²³⁸U content of the range of 218~477 ㎍ /L in the groundwater occurs in the twomica granite. ⁴He air-crust mixing ratio and the Rn-222 content show a rough relation, that is, Rn-222 content increases according to the increase of ⁴He crust mixing ratio. Because of helium and radon are an inert gas, their behavior in underground environment is assumed as an analogous. The ²³⁸U content and He isotope in groundwater does not show any distinct correlation. The groundwater can be classified as three groups (air, air-crust mixing, crust-mantle mixing origin) on the diagram of ³He/⁴He vs ⁴He/²⁰Ne, which is composed of original mixing line from air-crust-mantle end members. This original mixing of helium can provide the information of underground aquifer characteristic such as the connection with surface environment or isolation condition from air environment.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• 제7권4호
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• pp.219-227
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• 2009

The effect of organic acids on the adsorption of U(VI) onto oxide surfaces ($TiO_2)$(anatase), $SiO_2$(amorphous) and $Al_2O_3$(amorphous)) has been investigated. Two different organic acids, salicylic and picolinic acids, were used. Changes of adsorption ratio of U(VI), which depend on the existence of organic acids in a sample, were measured as a function of pH. Quantities of adsorbed organic acids, which depend on the existence of U(VI) in a sample, were also measured as a function of pH. It is confirmed that the soluble complex formation of U(VI) with organic acids can deteriorate the adsorption of U(VI) onto $TiO_2$ surface. It is noteworthy that salicylic acid does not affect the adsorption of U(VI) onto $SiO_2$ surface, however, picolinic acid enhances the adsorption of U(VI) onto $SiO_2$ surface. The latter effect can be understood by considering the formation of a ternary surface complex on $SiO_2$ surface, which was confirmed by the co-adsorption of picolinic acid with U(VI) and the change in a fluorescence spectra of U(VI) on surface, In the case of $Al_2O_3$, organic acids themselves were largely adsorbed onto a surface without deteriorating the adsorption of U(VI). This would support the possibility of a ternary surface complex formation on the $Al_2O_3$ surface, and an additional spectroscopic study is required.