• Nuclear Engineering and Technology
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• v.48 no.1
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• pp.274-284
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• 2016

Purpose: The purpose of this work is to describe the spectrometric analysis of gaseous cloud formation over reactor mixed uranium-and-plutonium (UP) fuel $(U_{0.8}Pu_{0.2})O_2$ samples heated to a temperature $>2,000^{\circ}C$, and thus forecast and evaluate radiation hazards threatening humans who cope with the consequences of any accident at a fission reactor loaded by UP mixed oxide $(U_{0.8}Pu_{0.2})O_2$, such as a mixture of 80% U and 20% Pu in weight. Materials and methods: The UP nuclear fuel samples were heated up to a temperature of over $2,000^{\circ}C$ in a suitable assembly (apparatus) at out-of-pile experiments' implementation, the experimental in-depth study of metabolism of active materials in living organisms by means of artificial irradiation of pigs by plutonium. Spectrometric measurements were carried out on the different exposed organs and tissues of pigs for the further estimation of human internal exposure by nuclear materials released from the core of a fission reactor fueled with UP mixed oxide. Results: The main results of the research described are the following: (1) following the research on the influence of mixed fuel fission products (radioactive isotopes being formed during reactor operation as a result of nuclear decay of elements included into the fuel composition) on living organisms, the authors determined the quantities of plutonium dioxide ($PuO_2$) that penetrated into blood and lay in the pulmonary region, liver, skeleton and other tissues; and (2) experiments confirmed that the output speed of plutonium out of the basic precipitation locations is very small. On the strength of the experimental evidence, the authors suggest that the biological output of plutonium can be disregarded in the process of evaluation of the internal irradiation doses.

• Economic and Environmental Geology
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• v.12 no.1
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• pp.41-49
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• 1979

A literature review is made on the physical and chemical characteristics of clay minerals in acidic solutions from the mineralogical and hydrometallurgical viewpoints. Some of the important characteristics of clays are their ability to cation exchange, swelling, and incongruent dissolution in acidic solutions. Various clay minerals can take up metallic ions from solution via cation exchange mechanism. Generally, cation exchange capacity increases in the following order : kaolinite, halloysite, illite, vermiculite, and montmorillonite. In acidic solutions, the cation uptake such as copper by clay minerals is strongly inhibited by hydrogen and aluminum ions and thus is not economically significant factor for recovery of metals such as uranium and copper. In acidic solutions, the cation uptake is substial. Swelling is minimal at lower pH, possibly due to lattice collapse. Swelling may be controllable with montmorillonite type clays by exchanging interlayer sodium with lithium and/or hydroxylated aluminum species. The effect of add on clay minerals are : 1. Division of aggregates into smaller plates with increase in surface area and porosity. 2. Clay-acid reactions occur in the following order: (i) $H^+$ replacement of interlayer cations, (ii) removal of octahedral cations, such as Al, Fe, and Mg, and (iii) removal of tetrahedral Al ions. Acid attack initiates, around the edges of the clay particles and continued inward, leaving hydrated silica gel residue around the edges. 3. Reaction rates of (ii) and (iii) are pseudo-1st order and proportional to acid concentration. Rate doubles for every temperature increment of $10^{\circ}C$. Implications in in-situ leaching of copper or uranium with acid are : 1. Over the life span of the operation for a year or more, clays attacked by acid will leave silica gel. If such gel covers the surface of valuable mineral surfaces being leached, recovery could be substantially delayed. 2. For a copper deposit containing 0.5% each of clay minerals and recoverable copper, the added cost due to clay-acid reaction is about 1.5c/lb of copper (or 0.93 lbs of $H_2SO_4/1b$ of copper). This acid consumption by clay may be a factor for economic evaluation of in-situ leaching of an oxide copper deposit.

• Economic and Environmental Geology
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• v.41 no.6
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• pp.773-795
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• 2008

The major import minerals of South Korea are copper ore, lead-zinc ore, iron ore, manganese ore and molybdenum ore. Oversea resources development of South Korea have 92 projects in 14 nations of Asia, 29 projects in 10 nations of America and Europe, and 14 projects in 9 nations of Middle Asia and Africa. But, most projects of them are found in Australia, China, Mongolia and Indonesia. The most projects of the Australia, China and Indonesia are interested in coal and a little projects of them have manganese, iron, lead-zinc, nickel, copper, gold, molybdenum, rare earth elements and uranium. The most projects of the Mongolia are interested in gold and rare earth elements. Representative ore deposits models of metal resources are Orogenic lode deposits, Volcanogenic massive sulphide deposits, Porphyry deposits, Sedimentary exhalative deposits, Mississippi valley type deposits, Iron oxide copper-gold deposits and Magmatic nickel-copper-platinum group element deposits based on global distribution, reverses and grades of their deposits models. If oversea mineral resources will be examined the mineral reserves, mineral mine production and ore deposits models of nations and then survey and investigate of mineral resources, we may be maintained ore body of high grade at survey area and decrease the investment risk.

• Analytical Science and Technology
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• v.11 no.3
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• pp.189-193
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• 1998

Binary alloys, $MoRu_3$ and $MoRh_3$, have been prepared using arc melting furnace. Mo and the noble metals Ru and Rh are the constituents of metallic insoluble residues, which were found in the early days of the post-irradiation studies on uranium oxide fuels. Detailed structural informations about these alloys have not been reported on JCPDS files of ICDD (International Centre for Diffraction Data). The results of X-ray diffraction study showed that the alloy was crystallized in hexagonal close-packing, well known as ${\varepsilon}$-phase. The X-ray diffraction patterns of these alloys matched well to that of $WRh_3$ with $P6_3/mmc$ of space group. The lattice parameters, a and c, were calculated using the least squares extrapolation. It was found from X-ray photoelectron spectroscopic measurements that Mo on the surface of the alloy was oxidized to Mo(6+), which could be removed by sputtering with Ar ions for approximately 15 minutes. The changes in binding energy of Mo, Ru, and Rh on the surface of the alloy were not observed. Magnetic susceptibility measurements resulted in the typical Pauli-paramagnetic behavior in the temperature range of 2 to 300 K.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.16 no.1
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• pp.59-64
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• 2018

In case of high contents of rare earths in the LiCl-KCl salt, it is not easy to recover U and TRU metals as a usable resource form from LiCl-KCl eutectic salts generated from the pyroprocessing of spent nuclear fuel. In this study, a conversion of $UCl_3$ into an oxide form using $K_2CO_3$ and an electrodeposition of $NdCl_3$ into a metal form in $LiCl-KCl-UCl_3-NdCl_3$ system were conducted to resolve the problem. Before conducting the conversion, experimental conditions for the conversion were determined by performing a thermodynamic equilibrium calculation. In this study, almost all of $UCl_3$ disappeared in the LiCl-KCl salt when the injection of $K_2CO_3$ reached theoretical equivalent for the conversion, and then $NdCl_3$ was effectively electrodeposited as a metal form using liquid zinc cathode. After that, the LiCl-KCl salt became transparent, and uranium oxides were precipitated to the bottom of the LiCl-KCl salt. These results will be utilized in designing a process to separate U and rare earths in LiCl-KCl salt.

• Journal of the Korean Ceramic Society
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• v.40 no.9
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• pp.916-920
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• 2003

The change of structure and morphology in ( $U_{0.913}$G $d_{0.087}$) $O_2$ during oxidation at 475$^{\circ}C$ and heat treatment at 130$0^{\circ}C$ in air were investigated using XRD, SEM, and EPMA. The ( $U_{0.913}$G $d_{0.087}$) $O_2$ cubic phase converted to ( $U_{0.913}$G $d_{0.087}$)$_3$ $O_{8}$ orthorhombic phase by oxidation at 475$^{\circ}C$ in air. The XRD and EPMA result of the 130$0^{\circ}C$ heat treated powder revealed that ( $U_{0.913}$G $d_{0.087}$)$_3$ $O_{8}$ orthorhombic phase was separated into $U_3$ $O_{8}$ and ( $U_{0.67}$G $d_{0.33}$) $O_{2+}$x/ cubic phase. The weight variations of (U,Gd) $O_2$ with various Gd contents were measured using TGA at the same heat treated condition. The weight variation during the heat treatment of Gd dissolve (U,Gd) $O_2$ in air can be expressed in terms of phase reaction equations related with oxidation and phase separation. Based on these phase reaction, a initial content of Gd dissolved in (U,Gd) $O_2$ can be exactly calculated by measuring the weight change during the heat treatment.

• Nuclear Engineering and Technology
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• v.49 no.5
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• pp.1044-1062
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• 2017

The construction project of the Kijang research reactor (KJRR), which is the second research reactor in Korea, has been launched. The KJRR was designed to use, for the first time, U-Mo fuel. Plate-type U-7 wt.% Mo/Al-5 wt.% Si, referred to as U-7Mo/Ale5Si, dispersion fuel with a uranium loading of $8.0gU/cm^3$, was selected to achieve higher fuel efficiency and performance than are possible when using $U_3Si_2/Al$ dispersion fuel. To qualify the U-Mo fuel in terms of plate geometry, the first miniplates [HANARO Miniplate (HAMP-1)], containing U-7Mo/Al-5Si dispersion fuel ($8gU/cm^3$), were fabricated at the Korea Atomic Energy Research Institute and recently irradiated at HANARO. The PIE (Post-irradiation Examination) results of the HAMP-1 irradiation test were analyzed in depth in order to verify the safe in-pile performance of the U-7Mo/Al-5Si dispersion fuel under the KJRR irradiation conditions. Nondestructive analyses included visual inspection, gamma spectrometric mapping, and two-dimensional measurements of the plate thickness and oxide thickness. Destructive PIE work was also carried out, focusing on characterization of the microstructural behavior using optical microscopy and scanning electron microscopy. Electron probe microanalysis was also used to measure the elemental concentrations in the interaction layer formed between the U-Mo kernels and the matrix. A blistering threshold test and a bending test were performed on the irradiated HAMP-1 miniplates that were saved from the destructive tests. Swelling evaluation of the U-Mo fuel was also conducted using two methods: plate thickness measurement and meat thickness measurement.

• Nuclear Engineering and Technology
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• v.53 no.10
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• pp.3367-3378
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• 2021

The design phase and safety assessment of Generation IV liquid metal-cooled fast reactors calls for the improvement of fuel pin performance codes, in particular the enhancement of their predictive capabilities towards uranium-plutonium mixed oxide fuels and stainless-steel cladding under irradiation in fast reactor environments. To this end, the current capabilities of fuel performance codes must be critically assessed against experimental data from available irradiation experiments. This work is devoted to the assessment of three European fuel performance codes, namely GERMINAL, MACROS and TRANSURANUS, against the irradiation of two fuel pins selected from the SUPERFACT-1 experimental campaign. The pins are characterized by a low enrichment (~ 2 wt.%) of minor actinides (neptunium and americium) in the fuel, and by plutonium content and cladding material in line with design choices envisaged for liquid metal-cooled Generation IV reactor fuels. The predictions of the codes are compared to several experimental measurements, allowing the identification of the current code capabilities in predicting fuel restructuring, cladding deformation, redistribution of actinides and volatile fission products. The integral assessment against experimental data is complemented by a code-to-code benchmark focused on the evolution of quantities of engineering interest over time. The benchmark analysis points out the differences in the code predictions of fuel central temperature, fuel-cladding gap width, cladding outer radius, pin internal pressure and fission gas release and suggests potential modelling development paths towards an improved description of the fuel pin behaviour in fast reactor irradiation conditions.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.7 no.2
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• pp.93-100
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• 2009

This study has been carried out to look into the characteristics of an oxidative-dissolution of fission products (FP) co-dissolved with uranium (U) in a $(NH_4)_2CO_3$ carbonate solution. Simulated FP-oxides which contained 12 components have been added to the solution to examine their dissolution characteristics. It is found that $H_2O_2$ is an effective oxidant to minimize the oxidative-dissolution of FP. In the 0.5 M $(NH_4)_2CO_3$-0.5 M $H_2O_2$ solution, some elements such as Re, Te, Cs and Mo seem to be dissolved together with U, while 98${\pm}$2% for Re and Te, 94${\pm}$2% for Cs, and 29${\pm}$2 % for Mo are dissolved for 2 hours. It is revealed that dissolution rates of Re, Te and Cs are high (completely dissolved within 10${\sim}$20 minutes) due to their high solubility in the $(NH_4)_2CO_3$ solution regardless of the addition of $H_2O_2$, and independent of the concentrations of $Na_2CO_3$ and $H_2O_2$. However, the dissolution ratio of Mo seems to be slightly increased with time and about 33 % for 4 hours, indicating a very slow dissolution rate and also independent of the $(NH_4)_2CO_3$ concentration. It is found that the most important factor for the oxidative-dissolution of FP is the pH of the solution and an effective dissolution is achieved at a pH between 9${\sim}$10 in order to minimize the dissolution of FP.