• Nuclear Engineering and Technology
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• v.55 no.9
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• pp.3126-3132
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• 2023

To meet the increasing heating needs of the China Fusion Experimental Tokamak Reactor (CFETR), the output power in each Lower Hybrid Current Drive (LHCD) transmission line should be increased from 250 kW to 500 kW. Therefore, a new high-power water load must be developed for the 4.6 GHz 500 kW LHCD system. This paper aims to report the most recent research progress of the water load: aluminum nitride (AlN) ceramic is used as the media material to isolate the water and vacuum, and the radio frequency (RF) simulation results show that the return loss of the water load is less than -25dB at 4.6 GHz over a wide temperature range. Under 500 kW continuous wave (CW) operation, the maximum temperatures of the ceramic and water are separately 67 ℃ and 62 ℃, resulting in thermal deformation of the ceramic of approximately 0.003 mm. Moreover, the AlN water load was tested on the 4.6 GHz 250 kW high-power test bench and found to work well with low reflected power.

• Nuclear Engineering and Technology
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• v.55 no.3
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• pp.927-938
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• 2023

The blanket plays an important role in fusion reactor and stands extremely high thermal and electromagnetic loads during operation situation and plasma disruption event, brings the need for precise thermal and electromagnetic analysis. Since the thermal field and EM field interact with each other nonlinearly, we develop a method of electromagnetic-thermal two-way coupling by using finite element software COMSOL. The coupling analyses of blanket under steady state and MD event are implemented and the results are analyzed. For steady state, the influences of coupling effects are relatively small but still recommended to be considered for a high precision analysis. The influence of thermal field on EM field can't be ignored under MD events. The variation of force density could cause a significant change in stress in certain parts of blanket. The influence of Joule heat during MD event is negligible, yet the potential temperature rise caused by induced current after MD event still needs to be researched.

• Korean Journal of Metals and Materials
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• v.49 no.8
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• pp.628-634
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• 2011

SA508 Gr.4N Ni-Cr-Mo low alloy steel, in which Ni and Cr contents are higher than in commercial SA508 Gr.3 Mn-Mo-Ni low alloy steels, may be a candidate reactor pressure vessel (RPV) material with higher strength and toughness from its tempered martensitic microstructure. The inner surface of the RPV is weld-cladded with stainless steels to prevent corrosion. The goal of this study is to evaluate the microstructural properties of the clad interface between Ni-Cr-Mo low alloy steel and stainless weldment, and the effects of post weld heat treatment (PWHT) on the properties. The properties of the clad interface were compared with those of commercial Mn-Mo-Ni low alloy steel. Multi-layer welding of model alloys with ER308L and ER309L stainless steel by the SAW method was performed, and then PWHT was conducted at $610^{\circ}C$ for 30 h. The microstructural changes of the clad interface were analyzed using OM, SEM and TEM, and micro-Vickers hardness tests were performed. Before PWHT, the heat affected zone (HAZ) showed higher hardness than base and weld metals due to formation of martensite after welding in both steels. In addition, the hardness of the HAZ in Ni-Cr-Mo low alloy steel was higher than that in Mn-Mo-Ni low alloy steel due to a comparatively high martensite fraction. The hardness of the HAZ decreased after PWHT in both steels, but the dark region was formed near the fusion line in which the hardness was locally high. In the case of Mn-Mo-Ni low alloy steel, formation of fine Cr-carbides in the weld region near the fusion line by diffusion of C from the base metal resulted in locally high hardness in the dark region. However, the precipitates of the region in the Ni-Cr-Mo low alloy steel were similar to that in the base metal, and the hardness in the region was not greatly different from that in the base metal.

• Proceedings of the Korea Committee for Ocean Resources and Engineering Conference
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• 2004.05a
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• pp.337-341
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• 2004

Engineering ceramics have superior heat resistance, corrosion resistance, and wear resistance. Consequently, these art significant candidates for hot-section structural components of heat engine and the inner containment of nuclear fusion reactor. Besides, some of them have the ability to heal cracks and great benefit can be anticipated with great benefit the structural engineering field. Especially, law fracture toughness of ceramics supplement with self-healing ability. In the present study, we have been noticed some practically important points for the healing behavior of silicon nitride, alumina, mullite with SiC particle and whisker. The presence of silicon carbide (SiC) in ceramic compound is very important for crack-healing behavior. However, self-healing of SiC has not been investigated well in detail yet. In this study, commercial SiC was selected as sample, which can be anticipated in the excellent crack healing ability. The specimens were produced three-point bending specimen with a critical semi-circular crack of which size that is about $50-700{\mu}m$. Three-point bending test and static fatigue test were performed cracked and healed SiC specimens. A monotonic bending load was applied to cracked specimens by three-point loading at different temperature. The purpose of this paper is to report Strength Properties and Elastic Waves Characteristics of Silicon Carbide with Crack Healing Ability.

• 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.

• Transactions of the Korean Society of Mechanical Engineers
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• v.19 no.1
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• pp.79-86
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• 1995

The structural materials for cryogenic technology have been recently developed to support the many modern large-scale application from superconducting magnets for nuclear fusion reactor, magnetic levitation railway to LNG tankers. However it is pointed out that quenching phenomenon is one of the serious problems for the integrity of these applications, which is mainly attributed to the rapid temperature rising in the material due to some extrinsic factors of structures. From the viewpoint of fracture mechanics, it is therefore very important to clarify the mechanism of temperature rising of structural material due to cyclic loading at cryogenic temperature. From this purpose, fatigue test was carried out for high manganese steel at liquid helium temperature(4.2K) using triangular stress waveform to identify both the mechanism of temperature rising near crack tip and the effect of loading stress waveform on temperature rising near crack tip and the effect of loading stress waveforms on temperature rising. As the results, two types of temperature rising, that is, regular and burst types were observed. And a periodical temperature rising corresponding to the stress waveforms was also found. The peaks of the temperature rising were recorded near both the maximum and the minimum values of the applied stress. The sudden temperature rises, which indicated the higher values than those of periodical temperature rises under the repetition of stress, were observed at the final region of crack growth. It was shown that the peak values of the temperature rising increased with stress intensity factor range.

• Journal of the Korean Society of Radiology
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• v.9 no.1
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• pp.55-59
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• 2015

Research about the proton nuclear reaction is actively achieving on the proton therapy including material development of fusion reactor. The proton induced gamma ray energy(2754, 1386 keV) spectrum of 27Al(p,3p+n)24Na reaction was measured with 100 MeV high energy proton beam. The proton beam in the experiment was derived from 100 MeV proton linear accelerator in the KOMAC. We measured the gamma ray intensity ratio of the decay level from the energy spectrum. The previous results have been compared with the current result. Strength of measured gamma rays will provide very important information though decide high energy gamma radiation detection efficiency.

• Transactions of the Korean hydrogen and new energy society
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• v.26 no.2
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• pp.114-119
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• 2015

Nowdays, fossil fuels have been used as an important resource in development of industry. But it is limited and caused climate change such as pollution and global warming. So nuclear fusion research is being issued with tritium to develop eco-friendly and sustainable energy. Republic of Korea is in charge of Storage and Delivery System (SDS) in the International Thermonuclear Experimental Reactor (ITER), weld present in the SDS bottles are easily exposed to the hydrogen embrittlement of special characteristics of the hydrogen in hydrogen atmosphere, When the hydrogen embrittlement is rapidly progresses, the cracking is generated in the weld zone. Due to this cracking, the risk of leakage of tritium into the atmosphere occurs. In this study, hydrogen heat treatment was processed through the Pressure-Composition-Temperature (PCT) device according to the time variation. Also mechanical properties such as rupture strength test, three point bend test and hardness test in accordance with the respective time have been conducted and the fracture was observed by scanning electron microscopy(SEM) after the mechanical properties evaluation.

• Journal of the Earthquake Engineering Society of Korea
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• v.24 no.3
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• pp.123-128
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• 2020

Analysis of the 2016 Gyeongju earthquake and the 2017 Pohang earthquake showed the characteristics of a typical high-frequency earthquake with many high-frequency components, short time strong motion duration, and large peak ground acceleration relative to the magnitude of the earthquake. Domestic nuclear power plants were designed and evaluated based on NRC's Regulatory Guide 1.60 design response spectrum, which had a great deal of energy in the low-frequency range. Therefore, nuclear power plants should carry out seismic verification and seismic performance evaluation of systems, structures, and components by reflecting the domestic characteristics of earthquakes. In this study, high-frequency amplification factors that can be used for seismic verification and seismic performance evaluation of nuclear power plant systems, structures, and equipment were analyzed. In order to analyze the high-frequency amplification factor, five sets of seismic time history were generated, which were matched with the uniform hazard response spectrum to reflect the characteristics of domestic earthquake motion. The nuclear power plant was subjected to seismic analysis for the construction of the Korean standard nuclear power plant, OPR1000, which is a reactor building, an auxiliary building assembly, a component cooling water heat exchanger building, and an essential service water building. Based on the results of the seismic analysis, a high-frequency amplification factor was derived upon the calculation of the floor response spectrum of the important locations of nuclear power plants. The high-frequency amplification factor can be effectively used for the seismic verification and seismic performance evaluation of electric equipment which are sensitive to high-frequency earthquakes.

• Proceedings of the KSME Conference
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• 2001.06a
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• pp.181-186
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• 2001

JLF-1 steel (Fe-9Cr-2W-V-Ta), low activation ferritic steel, is one of the promising candidate materials fer fusion reactor applications. High temperature fatigue life and tensile strength of JLF-1 steel and its TIG welded joints were investigated at the room temperature and $400^{\circ}C$. The strength of base metal (JLF-1) is in between those of weld metal and the HAZ. When the test temperature was increased from room temperature to $400^{\circ}C$, both strength and ductility decreased for base metal, weld metal and the HAZ. The longitudinal specimens of base metal showed similar strength and ductility compared with those of the transverse specimens at room temperature and $400^{\circ}C$. Little anisotropy was observed in the JLF-1 steel base metal in terms of rolling direction. Fatigue limit of weld metal which was obtained from cross-weld specimen is 495MPa. Thus, the weld metal showed the higher fatigue limit than those of base metal at both room temperature and $400^{\circ}C$. Little anisotropy of fatigue properties was observed for JLF-1 base metal in terms of rolling direction. When the test temperature was increased from room temperature to $400^{\circ}C$, the fatigue limit of both base metal and weld metal decreased substantially.