• Korean Journal of Mineralogy and Petrology
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• v.36 no.4
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• pp.289-302
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• 2023

The study area was Gangnim-myeon, Hoengseong-gun, Gangwon-do, composed of the Chiaksan gneiss complex, and it was revealed that the concentrations of uranium (U) and thorium (Th) within the groundwater of the study area exceeded their water quality standards. Hence, artificial weathering experiments were conducted to elucidate mineralogically the mechanisms of their leaching using drilling cores obtained from the corresponding groundwater aquifers. First of all, the mineralogical compositions of core samples were observed, and the results indicated that the content of clinochlore, a member of the chlorite group of minerals that can form through low- and intermediate-temperature metamorphisms, was relatively higher. In addition, the Th concentration was measured ten times higher than that of U. The results of artificial weathering experiments suggested that the Th concentrations gradually increased through the dissolution of radioactive-element-bearing minerals up to the first day, and then they tended to decrease. It could be attributed to the fact that Th was leached with the dissolution of thorite, which might be a secondary mineral, and then dissolved Th was re-precipitated as the various forms of salt, such as sulfate. Even though the U content was lower than that of Th in the core samples, the U concentration was one hundred times higher than that of Th after the weathering experiments. It is likely caused by the gradual dissolution and desorption of U included in intensively weathered thorite or adsorbed as a form of UO₂²⁺ on the mineral surface. In addition, the leaching tendency of U and Th was positively correlated with the bicarbonate concentration. However, the concentrations between U and Th in groundwater exhibited a relatively lower correlation, which might result from the fact that they occurred from different sources, as aforementioned. Among various kinetic models, the parabolic diffusion and pseudo-second-order kinetic models were confirmed to best fit the dissolution kinetics of both elements. The period that would be taken for the U concentration to exceed its drinking-water standard was inferred using the regressed parameters of the best-fitted models, and the duration of 29.4 years was predicted in the neutral-pH aquifers with relatively higher concentrations of HCO₃, indicating that U could be relatively quickly leached out into groundwater.

• Nuclear Engineering and Technology
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• v.53 no.11
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• pp.3653-3664
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• 2021

In PWR three-dimensional transient coupling calculation software CORCA-K, the nodal Green's function method and diagonal implicit Runge Kutta method are used to solve the spatiotemporal neutron dynamic diffusion equation, and the single-phase closed channel model and one-dimensional cylindrical heat conduction transient model are used to calculate the coolant temperature and fuel temperature. The LMW, NEACRP and PWR MOX/UO₂ benchmarks and FangJiaShan (FJS) nuclear power plant (NPP) transient control rod move cases are used to verify the CORCA-K. The effects of burnup, fuel effective temperature and ejection rate on the control rod ejection process of PWR are analyzed. The conclusions are as follows: (1) core relative power and fuel Doppler temperature are in good agreement with the results of benchmark and ADPRES, and the deviation between with the reference results is within 3.0% in LMW and NEACRP benchmarks; 2) the variation trend of FJS NPP core transient parameters is consistent with the results of SMART and ADPRES. And the core relative power is in better agreement with the SMART when weighting coefficient is 0.7. Compared with SMART, the maximum deviation is -5.08% in the rod ejection condition and while -5.09% in the control rod complex movement condition.

• Nuclear Engineering and Technology
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• v.53 no.7
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• pp.2095-2103
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• 2021

A small power ADS design using thorium oxide and diluent matrix reprocessed fuel is proposed for a high transmutation rate, small reactivity swing, and strong safety features. Two fuel matrices (CERCER and CERMET) and different recycled fuel compositions recovered from UO₂ spent fuels with 45 GWd/tU and 60 GWd/tU burnup were investigated to determine the suitable fuel for the ADS. It was found that the transmutation of each isotope depends on TRU initial loading amount. After examining the cores, the results show that CERCER fueled ADS has a negative coolant void reactivity (CVR) and a smaller radiotoxicity at discharge compared to that of CERMET core. It implies that CERCER fuel has enhanced safety features and more flavor in terms of radiotoxicity management. To increase fuel utilization and core operation efficiency, a simple assembly shuffling pattern for the CERCER fueled ADS is also proposed. Eigenvalue and burnup calculations were conducted using Serpent 2 with ENDF/B-VII.0 library in both kcode and external source modes, and it indicates that the results of transmutation analyses obtained by kcode only is reliable to discuss the transmutation potential of ADS. Burnup calculation with the fixed-source mode is essential to be used for more practical results of the transmutation by ADS.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.21 no.2
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• pp.193-204
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• 2023

To ensure radiological safety margin in the transport and storage of spent nuclear fuel, it is crucial to perform source term and shielding analyses in advance from the perspective of conservation. When performing source term analysis on UO₂ fuel, which is mostly used in commercial nuclear power plants, uranium and oxygen are basically considered to be the initial materials of the new fuel. However, the presence of impurities in the fuel and structural materials of the fuel assembly may influence the source term and shielding analyses. The impurities could be radioactive materials or the stable materials that are activated by irradiation during reactor power operation. As measuring the impurity concentration levels in the fuel and structural materials can be challenging, publicly available information on impurity concentration levels is used as a reference in this evaluation. To assess the effect of impurities, the results of the source term and shielding analyses were compared depending on whether the assumed impurity concentration is considered. For the shielding analysis, generic cask design data developed by KEPCO-E&C was utilized.

• Proceedings of the Mineralogical Society of Korea Conference
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• 2003.05a
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• pp.67-67
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• 2003

The chemical compositions and textural relationships of the Nb-Zr oxide minerals including pyrochlore [ideally (Ca,Na)$_2$Nb$_2$O$\sub$6/(OH,F), with up to 24% UO$_2$ and 16% Ta$_2$O$\sub$5/] and baddeleyite [ideally ZrO$_2$, with up to 6% Nb$_2$O$\sub$5/] in the Sokli carbonatite complex, Kola Peninsula, Arctic are described. These two minerals in carbonatites are the major hosts for the HFSEs such as U, Th, Ta, Nb, Zr and Hf and thus are interest both economically and petrologically. The Sokli carbonatite complex (360-370 Ma) in Northern Finland, which forms a part of the Paleozoic Kola Alkaline Province (KAP), is mainly composed of multi-stages of carbonatite and phoscorite associations (P1-C1 P2-C2, P3-C3, D4 and D5) surrounded by altered ultramafic rocks (olivinite and pyroxenite) and cut by numerous small dikes of ultramafic lamprophyre. The Sokli complex contains the highest concentration in niobium and probably in tantalum, which are economically very important to modern steel technology, among the ultramafic-alkaline complexes of the KAP. Pyrochlore and baddeleyite mostly concentrate in the phoscorites. Pyrochlores in the Sokli complex are generally rounded octahedra and cubes in shape, red brown to grey yellow in color, and 0.2 to 5 mm in size. They are found in all calcite carbonatites, phoscorites and dolomite carbonatites, except P1-C1 rocks. These pyrochlores display remarkable zonations which depend on host rock compositions, and have significant compositional variations with evolution of the Sokli complex. The common variation scheme is that (1) early pyrochlore is highly enriched in U and Ta; (2) these elements decrease abruptly in the intermediate stage, while Th and Ce increase, and (3) late stage pyrochlore is low in U, Ta, Th, and Ce, and correspondingly high in Nb. Baddeleyites in the Sokli complex occur in the early P1-C1 and P2-C2 rocks and rarely in P3. They crystallized earlier than pyrochlores, and occasionally show post-magmatic corrosion and replacement. The FeO and TiO$_2$ contents of baddeleyites are much lower than those of the other terrestrial and lunar baddeleyites, whereas Nb$_2$O$\sub$5/ and Ta$_2$O$\sub$5/ contents are the highest among the reported compositions. Ta/Nb and Zr/Nb ratios of pyrochlores and baddeleyites decrease towards later stage facies, which is in accordance with the whole rock compositions. The variation of Ta/Nb and Zr/Nb ratios of pyrochlores and baddeleyites is considered to be a good indicator to trace an evolution of the carbonatite complexes.

• Nuclear Engineering and Technology
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• v.52 no.8
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• pp.1714-1723
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• 2020

This study aims at characterizing the rheological properties of uranium oxide powders for nuclear fuel pellets manufacturing. The flowability of these powders must be compatible with a reproducible filling of press molds. The particle size distribution is known to have an impact on the rheological properties and fine particles (<100 ㎛) are suspected to have a detrimental effect. In this study, the impact of the particle size distribution on the rheological properties of UO₂ powders was quantified, focusing on the influence of fine particles. Two complementary approaches were used. The first approach involved characterizing the powder in a static state: density, compressibility and shear test measurements were used to understand the behavior of the powder when it is transitioned from a static to a dynamic state (i.e., incipient flow conditions). The second approach involved characterizing the behavior of the powder in a dynamic state. Two zones, corresponding to two characteristic behaviors, were demonstrated for both types of measurements. The obtained results showed the amount of fines should be kept below 10 % wt to ensure a robust mold filling operation (i.e., constant mass and production rate).

• Nuclear Engineering and Technology
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• v.53 no.9
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• pp.2830-2838
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• 2021

This work deals with the assessment of the burnup capabilities of the Serpent Monte Carlo code to predict spent nuclear fuel (SNF) isotopic concentrations for low-enriched uranium (LEU) fuel at different burnup levels up to 47 MWd/kgU. The irradiation of six UO₂ experimental samples in three different VVER-1000 reactor units has been simulated and the predicted concentrations of actinides up to ²⁴⁴Cm have been compared with the corresponding measured values. The results show a global good agreement between calculated and experimental concentrations, in several cases within the margins of the nuclear data uncertainties and in a few cases even within the reported experimental uncertainties. The differences in the performances of the JEFF3.1.1, ENDF/B-VII.1 and ENDF/B-VIII.0 nuclear data libraries (NDLs) have also been assessed and the use of the newly released ENDF/B-VIII.0 library has shown an increased accuracy in the prediction of the C/E's for some of the actinides considered, particularly for the plutonium isotopes. This work represents a step forward towards the validation of advanced simulation tools against post irradiation experimental data and the obtained results provide an evidence of the capabilities of the Serpent Monte-Carlo code with the associated modern NDLs to accurately compute SNF nuclide inventory concentrations for VVER-1000 type reactors.

• Nuclear Engineering and Technology
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• v.54 no.7
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• pp.2475-2490
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• 2022

The Accident Tolerant Fuel (ATF) is a new concept of fuel, which can not only withstand the consequences of the accident for a longer time, but also maintain or improve the performance under operating conditions. ISAA is a self-developed severe accident analysis code, which uses modular structures to simulate the development processes of severe accidents in nuclear plants. The basic version of ISAA is developed based on UO₂-Zr fuel. To study the potential safety gain of ATF cladding, an improved version of ISAA, referred to as ISAA-ATF, is introduced to analyze the station blackout accident of PWR using ATF cladding. The results show that ATF cladding enable the core to maintain a longer time compared to zirconium alloy cladding, thereby enhancing the accident mitigation capability. Meanwhile, the generation of hydrogen is significantly reduced and delayed, which proves that ATF can improve the safety characteristics of the nuclear reactor.

• Nuclear Engineering and Technology
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• v.56 no.7
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• pp.2625-2632
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• 2024

In this study, we investigated the possibility of employing accident tolerant fuel (ATF) in VVER-1200/V491 assembly without gadolinium-containing fuel rods using the Monte Carlo code Serpent 1.1.7 with ENDF/B-VII cross-section library. The analysis involves assembly design with reflective boundary conditions. To compare the neutronic performances, U-5Mo, U-7.5Mo, U-10Mo, and U-15Mo fuels were chosen in addition to ordinary UO₂ fuel. The concentration of ¹³⁵Xe, ¹⁴⁹Sm, fissile and fertile isotopes with burnup, reactivity feedback with fuel temperature variation, and β eff values were calculated. The results indicate that the fuel cycle length increases by 54.27% for U-5Mo, 32.6% for U-7.5Mo, and 13.8% for U-10Mo, while it decreases by 16.4% for U-15Mo fuel. Additionally, the effect of ⁹⁵Mo content in natural Mo was investigated by reducing the ⁹⁵Mo concentration. According to the results, each proposed fuel's fuel cycle length extended when the depletion ratio of ⁹⁵Mo increased. Additionally, the calculations for reactivity feedback guarantee safe operating conditions for all U-xMo fuels.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.3 no.1
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• pp.67-75
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• 2005

The size and main ingredient of hot particulate generated during the nuclide experiment in hot cells of nuclear facilities were 0.5300 $\mu$m and UO$\_2$. A cyclone filter equipment which consists of a cyclone and Bag/HEPA filter was devised to remove hot particulate generated during the nuclide experiment in hot cells of nuclear facilities. The experimental conditions to maximize the collection efficiency of hot particulate were suggested through experiments done with the cyclone filter equipment. With the large size of simulated particulate, the collection efficiency of the particulate was high. When the size of simulated particulate was more than 5 $\mu$m, the collection efficiency of the particulate was more than $80\%$ and when the size of simulated particulate was less than 1.0 urn, the collection efficiency decreased by less than $70\%$. If the inflow velocity of simulated particulate was increased, the collection efficiency of the particulate was also increased. When the inflow velocity of simulated particulate was more than 12m/sec, the collection efficiency was higher than $70\%$, but after 17 m/sec inflow velocity, no change observed. The collection efficiency of the simulated particulate can be enhanced with the length of vortex finder inside the chamber. With the length of vortex finder, 7.2cm, the observed collection efficiency of the particulate was the maximum. Moreover, when the sub-cone was attached under the cyclone, the collection efficiency of cyclone increased $2\%$. It was found that effect by attachment of sub-cone was not serious.