• Proceedings of the Korean Nuclear Society Conference
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• 2004.10a
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• pp.377-378
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• 2004

• Journal of Hydrogen and New Energy
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• v.27 no.2
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• pp.163-168
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• 2016

Korea has been developing nuclear fusion fuel storage and delivery system (SDS) technologies including a basic scientific study on hydrogen storage. To develop nuclear fusion technology, it is necessary to store and supply hydrogen isotopes needed for Tokamak operation. SDS is used for storing hydrogen isotopes as a metal hydride form. The rapid hydriding of tritium is very important not only for safety reasons but also for the economic design and operation of the SDS. In this study, we designed and fabricated a medium-scale getter bed of depleted uranium (DU). The hydriding of DU has been measured by varying the initial temperature ($100-300^{\circ}C$) of the DU getter bed to investigate the influence of the cooling temperature. Furthermore, we analyzed the effect of a helium blanket on the hydriding performance with 0 - 12% helium content in hydrogen.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.28 no.2
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• pp.189-195
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• 2004

The pressure tubes, which contain high temperature heavy water and fuel, are within the core of a CANDU nuclear reactor, and are thus subjected to high stresses, temperature gradient, and neutron flux. Further, it is well known that pressure tubes of cold-worked Zr-2.5Nb materials result in hydrogen diffusion, which create fully-hydrided regions (frequently called Blister). Thus a proper investigation of hydrogen diffusion within zirconium-alloy nuclear components, such as CANDU pressure tube and fuel channels is essential to predict the structural integrity of these components. In this respect, this paper presents numerical investigation of hydrogen diffusion to quantify the hydrogen concentration fur blister growth of CANDU pressure tube. For this purpose, coupled temperature-hydrogen diffusion analyses are performed by means of two-dimensional finite element analysis. Comparison of predicted temperature field and blister with published test data shows good agreement.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.19 no.2
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• pp.205-214
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• 2021

Currently, the interim storage pools of spent fuels in South Korea are expected to become saturated from 2024. It is required to prepare an operation plan of a domestic dry storage facility during a long-term period, with the researches on safety evaluation methods. This study modified the FRAPCON code to predict the spent fuel integrity evaluation such as the axial cladding temperature, the hoop stress and hydrogen distribution in dry storage. The cladding temperature in dry storage was calculated using the COBRA-SFS code with the burnup information which was calculated using the FRAPCON code. The hoop stress was calculated using the ideal gas equation with spent fuel information such as rod internal pressure. Numerical analysis method was used to calculate the degree of hydrogen diffusion according to the hydrogen concentration and temperature distribution during a dry storage period. Before 50 years of dry storage, the cladding temperature and hoop stress decreased rapidly. However, after 50 years, they decreased gradually and the cladding temperature was below 400 K. The initial temperature distribution and hydrogen concentration showed a parabolic line, but hydrogen was transferred by the hydrogen concentration and temperature gradient over time.

• Nuclear Engineering and Technology
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• v.51 no.3
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• pp.792-799
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• 2019

The present study aims to investigate the fluorination of thorium oxide ($ThO_2$) by ammonium hydrogen difluoride ($NH_4HF_2$). Fluorination was performed at room temperature by mixing $ThO_2$ and $NH_4HF_2$ at different molar ratios, which was then left to react for 20 days. Next, the mixtures were analyzed using X-ray diffraction (XRD) at the intervals of 5, 10, 15, and 20 days, followed by the heating of the mixtures at $450-750^{\circ}C$ with argon gas flow. The characterization of $ThF_4$ was established using X-ray diffraction (XRD) and scanning electron microscopy-dispersion X-ray spectroscopy (SEM-EDX). In this study, ammonium thorium fluoride was synthesized through the fluorination of $ThO_2$ at room temperature. The optimum molar ratio in synthesizing ammonium thorium fluoride was 1.0:5.5 ($ThO_2:NH_4HF_2$) with 5 days reaction time. In addition, the heating of ammonium thorium fluoride at $450^{\circ}C$ was sufficient to produce $ThF_4$. Overall, this study proved that $NH_4HF_2$ is one of the fluorination agents that is capable of synthesizing $ThF_4$.

• Nuclear Engineering and Technology
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• v.52 no.9
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• pp.2047-2053
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• 2020

Single-tube burst tests on hydrogenated Zircaloy-4 nuclear fuel cladding under simulated loss-of-coolant accident are conducted to evaluate the impact of hydrogen on burst parameters. The heating rate and initial pressure are varied from 5 K/s to 150 K/s and 5 bar-80 bar, respectively. The hydrogen concentration in the cladding is in the range of 0-2000 wppm. Burst stress is lower for hydrogenated cladding in α-phase. A significant loss of ductility is observed in α-phase and lower α + β-phase for hydrogenated cladding. However, the burst strain is higher for hydrogenated cladding in β-phase. There is a sigmoidal dependency of rupture area with initial stress and rupture area is larger for hydrogenated cladding. A novel burst stress correlation for hydrogenated Zircaloy-4 cladding has been proposed.

• Nuclear Engineering and Technology
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• v.52 no.8
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• pp.1742-1748
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• 2020

TSSD, TSSP, and TSSP2 of hydrogen for optimized-Zirlo (Zirlo™) alloy were measured by DSC in the range of 53-457 wppm. Solvus curves of the TSSs are derived and proposed in this study. The results show that the temperature gap between TSSD and TSSP solvus lines of Zirlo™ are similar to those of other zirconium alloys, but another gap between the TSSD and TSSP2 line differs significantly. In particular, the TSSP2 solvus line becomes closer to the TSSD solvus line than to TSSP unlike Zircaloy-4, so ΔT_TSSD-TSSP2 of Zirlo™ decreases with decreasing temperature. This implies that hydride reorientation can take place more significantly in Zirlo™ than in Zircaloy-4, and the limited temperature variation of 65 ℃ during the vacuum drying and the cooling-down process may not be sufficient to prevent the triggering of hydride reorientation in Zirlo™ cladding under long-term dry storage.

• Journal of Hydrogen and New Energy
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• v.26 no.5
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• pp.402-408
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• 2015

In D-T fusion reaction, $D_2$ (duterium) and $T_2$(tritium) are used as fuel gas. The exhaust gas of nuclear fusion includes hydrogen isotopes $Q_2$ (Q means H, D or T), tritiated components ($CQ_4$ and $Q_2O$), CO, $CO_2$, etc. All of hydrogen isotopes should be recovered before released to the atmosphere. This study focused on the recovery of hydrogen isotopes from $CQ_4$ and $Q_2O$. Two kinds of experiments were conducted to investigate the catalytic reaction characteristics of SMR (Steam Methane Reforming) and WGS (Water Gas Shift) reactions using Pt catalyst. First test was performed to convert $CH_4$ into $H_2$ using 6% $CH_4$, 6% CO/Ar feed gas. In the other test, 100% CO gas was used to convert $H_2O$ into $H_2$ at various reaction conditions (reaction temperature, S/C ratio, GHSV). As a result of the first test, $CH_4$ and CO conversion were 41.6%, 57.8% respectively at $600^{\circ}C$, S/C ratio 3, GHSV $2000hr^{-1}$. And CO conversion was 72% at $400^{\circ}C$, S/C ratio 0.95, GHSV $333hr^{-1}$ in the second test.

• Proceedings of the Korean Nuclear Society Conference
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• 1998.10a
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• pp.187-187
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• 1998

• Proceedings of the Korean Nuclear Society Conference
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• 2002.10a
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• pp.436.2-436
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• 2002