• Nuclear Engineering and Technology
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• v.54 no.10
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• pp.3717-3723
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• 2022

For the first time, two-dimensional temporal behavior of the edge localized mode (ELM) filament is measured in the edge tokamak plasma with a multi-channel electrical probe array (MCEP). MCEP, which has 16 floating probes (4 × 4), is mounted at the far scrape-off layer (SOL) region in the KSTAR. An electron temperature and an ion flux are measured by sideband method (SBM), which can achieve two-dimensional measurements with high time resolution. Furthermore, temporal evolutions of the electron temperature and the ion flux are obtained during the ELM occurrence. In the H-mode period, short spikes from ELM bursts are observed in measured plasma parameters, and the trend is similar to that of typical Hα signal. Interestingly, when blob-like ELM filaments crash the probe, the heat flux is significantly higher in a local region of the probe array. The results show that our probe array using the SBM can measure the ELM behavior and the plasma parameters without the effect of the stray current caused by the huge device. This study can provide valuable data needed to understand the interaction between the SOL plasma and the plasma facing components (PFCs).

• Analytical Science and Technology
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• v.23 no.6
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• pp.511-517
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• 2010

Dissolution procedure of vanadium pentoxide, which is widely used as a catalyst for production of sulfuric acid or an oxide reaction of the numerous organic compounds, was investigated. Reagent of vanadium pentoxide was completely dissolved in aqua regia-$H_2O_2$-HF solution, but plate type of vanadium pentoxide sample was not clearly dissolved with mixed acids. Thus, in order to establish the dissolution procedure for plate type of vanadium pentoxide, the solubility of vanadium pentoxide was investigated through comparison of acid treatment-fusion and microwave digestion methods. The optimized acid for dissolution of vanadium compound was found to be mixing acids of aqua regia, $H_2O_2$ and HF. Acid-fusion and microwave digestion methods have a similar property in the solubility of vanadium compound, but the latter was more quick and convenient procedure. The content of vanadium pentoxide was found to be $97.9{\pm}0.9%$ using an inductively coupled plasma atomic emission spectrometer after dissolution of a sample with the microwave digestion system.

• Journal of Technology Innovation
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• v.22 no.4
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• pp.145-164
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• 2014

Nuclear fusion is one of the most promising options for generating large amounts of carbon-free energy in the future. Major countries such as China, EU, and Japan have established a national plan for DEMO construction and they are implementing it. Korea has started a nuclear fusion research and development by the KSTAR project started in 1995. There are matured needs for a full-scale research and development initiatives to ensure competition with the major countries for DEMO as well as achieve the final goal to commercialize fusion energy. In this paper, we apply the TRL and AHP methods in order to identify the key technologies to conduct DEMO R&D. We propose the priorities of future R&D on DEMO by deriving a core technology in the field. At first, we review the scientific theory of fusion and trend of progress of DEMO activities in major countries. For previous studies, we review TRL and AHP methods to examine the technology classification system of DEMO and identify key technologies. We apply TRL method to identify readiness level of DEMO technologies and AHP to compensate shortcoming of TRL. The key technologies of DEMO to be secured from a synthesis result of the TRL and AHP are burning plasma, plasma facing material, structural material, high frequency heating, neutral particle beam, safety, plasma diagnostic, and simulation technologies.

• Transactions of the Korean hydrogen and new energy society
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• v.22 no.3
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• pp.372-379
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• 2011

A nuclear fusion fuel cycle plant is composed of various subsystems such as a hydrogen isotope storage and delivery system, a tokamak exhaust processing system, and a hydrogen isotope separation system. Korea shares in the construction of its ITER fuel cycle plant with the EU, Japan, and the US, and is responsible for the development and supply of the storage and delivery system. The authors thus present details on the development status of hydrogen isotope storage technologies for nuclear fusion fuel cycle plants. We have developed various hydride beds of different size. We have realized a hydrogen delivery rate of 12.5 $Pam^3/s$ with a typical 1242g-ZrCo bed.

• Transactions of the Korean hydrogen and new energy society
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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.

• Journal of the Korean Ceramic Society
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• v.51 no.5
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• pp.435-441
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• 2014

Silicon carbide-based ceramics and their composites have been studied for application to fusion and advanced fission energy systems. For fission reactors, $SiC_f$/SiC composites can be applied to core structural materials. Multilayered SiC composite fuel cladding, owing to its superior high temperature strength and low hydrogen generation under severe accident conditions, is a candidate for the replacement of zirconium alloy cladding. The SiC composite cladding has to retain its mechanical properties and original structure under the inner pressure caused by fission products; as such it can be applied as a cladding in fission reactor. A hoop strength test using an expandable polyurethane plug was designed in order to evaluate the mechanical properties of the fuel cladding. In this paper, a hoop strength test of the multilayered SiC composite tube for nuclear fuel cladding was simulated using FEA. The stress caused by the plug was distributed nonuniformly because of the friction coefficient difference between the inner surface of the tube and the plug. Hoop stress and shear stress at the tube was evaluated and the relationship between the concentrated stress at the inner layer of the tube and the fracture behavior of the tube was investigated.

• Journal of Powder Materials
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• v.21 no.4
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• pp.271-276
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• 2014

15Cr-1Mo base oxide dispersion strengthened (ODS) steel which is considered to be as a promising candidate for high- temperature components in nuclear fusion and fission systems because of its excellent high temperature strength, corrosion and radiation resistance was fabricated by using mechanical alloying, hot isostatic pressing and hot rolling. Torsion tests were performed at room temperature, leading to two different shear strain routes in the forward and reverse directions. In this study, microstructure evolution of the ODS steel during simple shearing was investigated. Fine grained microstructure and a cell structure of dislocation with low angle boundaries were characterized with shear strain in the shear deformed region by electron backscattered diffraction (EBSD). Grain refinement with shear strain resulted in an increase in hardness. After the forward-reverse torsion, the hardness value was measured to be higher than that of the forward torsion only with an identical shear strain amount, suggesting that new dislocation cell structures inside the grain were generated, thus resulting in a larger strengthening of the steel.

• Nuclear Engineering and Technology
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• v.45 no.3
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• pp.311-322
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• 2013

High-Cr martensitic steel HT-9 is one of the candidate materials for advanced nuclear energy systems. Thanks to its excellent thermal conductivity and irradiation resistance, ferritic/martensitic steels such as HT-9 are considered for in-core applications of advanced nuclear reactors. The harsh neutron irradiation environments at the reactor core region pose a unique challenge for structural and cladding materials. Microstructural and microchemical changes resulting from displacement damage are anticipated for structural materials after prolonged neutron exposure. Consequently, various irradiation effects on the service performance of in-core materials need to be understood. In this work, the fundamentals of radiation damage and irradiation effects of the HT-9 martensitic steel are reviewed. The objective of this paper is to provide a background introduction of displacement damage, microstructural evolution, and subsequent effects on mechanical properties of the HT-9 martensitic steel under neutron irradiations. Mechanical test results of the irradiated HT-9 steel obtained from previous fast reactor and fusion programs are summarized along with the information of irradiated microstructure. This review can serve as a starting point for additional investigations on the in-core applications of ferritic/martensitic steels in advanced nuclear reactors.

• Transactions of the Korean hydrogen and new energy society
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• v.24 no.2
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• pp.128-135
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• 2013

The nuclear fuel cycle plant is composed of various subsystems such as a fuel storage and delivery system (SDS), a tokamak exhaust processing system, a hydrogen isotope separation system, and a tritium plant analytical system. Korea is sharing in the construction of the International Thermonuclear Experimental Reactor (ITER) fuel cycle plant with the EU, Japan, and the US, and is responsible for the development and supply of the SDS. Hydrogen isotopes are the main fuel for nuclear fusion reactors. Metal hydrides offer a safe and convenient method for hydrogen isotope storage. The storage of hydrogen isotopes is carried out by absorption and desorption in a metal hydride bed. These reactions require heat removal and supply respectively. Accordingly, the rapid storage and delivery of hydrogen isotopes are enabled by a rapid cooling and heating of the metal hydride bed. In this study, we designed and manufactured a vertical-type hydrogen isotope storage bed, which is used to enhance the cooling performance. We present the experimental details of the cooling performances of the bed using various cooling parameters. We also present the modeling results to estimate the heat transport phenomena. We compared the cooling performance of the bed by testing different cooling modes, such as an isolation mode, a natural convection mode, and an outer jacket helium circulation mode. We found that helium circulation mode is the most effective which was confirmed in our model calculations. Thus we can expect a more efficient bed design by employing a forced helium circulation method for new beds.

• Nuclear Engineering and Technology
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• v.55 no.3
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• pp.981-988
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• 2023

A seismic event caused an accident at the Fukushima Nuclear Power Plant, which further resulted in simultaneous accidents at several units. Consequently, this incident has aroused great interest in the safety of nuclear power plants worldwide. A reasonable safety evaluation of such an external event should appropriately consider the correlation between SSCs (structures, systems, and components) and the probability of failure. However, a probabilistic safety assessment in current nuclear industries is performed conservatively, assuming that the failure correlation between SSCs is independent or completely dependent. This is an extreme assumption; a reasonable risk can be calculated, or risk-based decision-making can be conducted only when the appropriate failure correlation between SSCs is considered. Thus, this study analyzed the effect of the failure correlation of SSCs on the safety of the system to realize rational safety assessment and decision-making. Consequently, the impact on the system differs according to the size of the failure probability of the SSCs and the AND and OR conditions.