• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
• /
• v.18 no.1
• /
• pp.83-90
• /
• 2020

The Graphite Isotope Ratio Method (GIRM) can verify non-proliferation of nuclear weapon by estimating the total plutonium production in a graphite-moderated reactor. Using the reactor, plutonium is generated and accumulated through the ²³⁸U neutron capture reaction, and impurities in the graphite are converted to nuclides due to the nuclear reaction. Therefore, the amount of plutonium production and concentration of the impurities are correlated. However, the plutonium production cannot be predicted using only the absolute concentration of the impurities. It can only be predicted when the initial concentration of the impurities is obtained because the concentration, at a certain time, depends on it. Nevertheless, the ratios of the isotopes in an element are known regardless of the impurity of an element in the graphite moderator. Thus, the correlation between the isotope ratio and amount of plutonium produced helps predict plutonium production in a graphite-moderated reactor. Boron, Lithium, Chlorine, Titanium, and Uranium are known as indicator elements in the GIRM. To assess whether the correlation between the indicator isotope and amount of plutonium produced is independent of the initial concentration of the impurities, four different impurity compositions of graphite were used. ¹⁰B/¹¹B, ³⁶Cl/³⁵Cl, ⁴⁸Ti/⁴⁹Ti, and ²³⁵U/²³⁸U had a consistent correlation with the cumulative plutonium production, regardless of the initial impurity concentration of the graphite, because these isotopes were not generated through the nuclear reaction of other elements. On the other hand, the correlation between ⁶Li/⁷Li and plutonium production depended on the initial concentration of the impurities in graphite. Although ⁷Li can be produced through the neutron capture reaction of ⁶Li, the (n, α) reaction of ¹⁰B was the major source of ⁷Li. Therefore, the initial concentration of ¹⁰B affected the production of ⁷Li, making Li unsuitable as an indicator element for the GIRM.

• Journal of the Korean Chemical Society
• /
• v.26 no.3
• /
• pp.160-164
• /
• 1982

A guanidinium salt of $[U(PW_{11}O_{39})_2]^{10-}$, the solubility of which is adequate for crystal growing, has been synthesized. Using this salt or potassium salt, we have measured the stability of $[U(PW_{11}O_{39})_2]^{10-}$as a function of pH of the solution and found that the anion is stable for the pH range 3~7. We have developed a colorimetric method for determining the concentration of $U^{4+}$. In this method$PW_{11}O_{39}^{7-}$ is added to$U^{4+}$ in such a quantity that the mole ratio $PW_{11}O_{39}^{7-}/ U^{4+}$exceeds 2 and the intensity of the 22.7kK band (${\varepsilon}$1030 M-1cm-1) is measured. In order to develop a continuous method to recover uranium, we have determined the amount of recoverd$PW_{11}O_{39}^{7-}$ after decomposing $[U(PW_{11}O_{39})_2]^{10}$- by adding either a base or an oxidizing agent. The percentage of $PW_{11}O_{39}^{7-}$recovered was approximately 70% when a base was used and approximately 80% when$K_2S_2O_8$ was used. A colorimetric method for determining $PW_{11}O_{39}^{7-}$ has also been developed.

• Journal of Radiation Protection and Research
• /
• v.24 no.4
• /
• pp.187-193
• /
• 1999

The specific radioactivity concentrations in the coal fly ash obtained from heat producing stations in Korea were analyzed and its radiological hazard for reuse in construction purpose was evaluated. The concentrations of uranium isotopes in the real fly ash measured by TBP solvent extraction method and $\alpha$-spectrometer were found to be about 116.1 Bq $kg^{-1}$ for $^{238}U$, 5.01 Bq $kg^{-1}$ for $^{235}U$, and 121.2 Bq $kg^{-1}$ for $^{234}U$, respectively. The activity ratio of $^{234}U/^{238}U$, in the coal fly ash was in $1.04\;{\pm}\;0.03$, which is similar to that of uncontaminated Korean soil in natural conditions (1.14). The specific radioactivities of $^{226}Ra,\;^{232}Th,\;and\;^{40}K$ in the coal fly ash were also determined using $\gamma$-spectrometer with a HPGe detector The results showed that $^{226}Ra,\;^{232}Th,\;and\;^{40}K$ in the coal fly ash were in concentrations of $101.7{\sim}113.9$, $39.5{\sim}54.2\;and\;315.0{\sim}990.6$ Bq $kg^{-1}$, respectively. With the specific radioactivities obtained from $\gamma$-spectrometric measurements of the coal fly ash, its radiological hazard for reuse was evaluated. The result showed that the radioactivity of the coal fly ash was in permissible level.

• Nuclear Engineering and Technology
• /
• v.15 no.3
• /
• pp.197-206
• /
• 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
• /
• v.56 no.5
• /
• pp.603-618
• /
• 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)
• /
• v.27 no.3
• /
• pp.242-248
• /
• 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)
• /
• v.16 no.3
• /
• pp.309-313
• /
• 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
• /
• v.6 no.2
• /
• pp.163-177
• /
• 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
• /
• v.48 no.6
• /
• pp.1330-1337
• /
• 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
• /
• v.31 no.3
• /
• pp.335-343
• /
• 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.