• 한국원자력학회:학술대회논문집
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• 한국원자력학회 2005년도 춘계학술발표회
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• pp.43-44
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• 2005

• 한국세라믹학회지
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• 제43권7호
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• pp.439-444
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• 2006

Artificial graphite generally manufactured by carbonization sintering of shape-body of kneaded mixture using granular cokes as filler and pitch as binder, going through pitch impregnation process if necessary and finally applying graphitization heat treatment. Graphite materials are used for core internal structural components of the High-Temperature Gas-cooled Reactors (HTGR) because of their excellent heat resistibility and resistance of crack progress. The HTGR has a core consisting of an array of stacked graphite fuel blocks are machined from IG-110, a high-strength, fine-grained isotropic graphite. In this study, crack stabilization and micro-structures were measured by bend strength and fracture toughness of isotropic graphite grade IG-110. It is important to the reactor designer as they may govern the life of the graphite components and hence the life of the reactor. It was resulted crack propagation, bend strength, compressive strength and micro-structures of IG-110 graphite by scanning electron microscope and universal test machine.

• Carbon letters
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• 제4권1호
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• pp.36-39
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• 2003

Carbon and carbon-based materials are used in nuclear reactors and there has recently been growing interest to develop graphite and carbon based materials for high temperature nuclear and fusion reactors. Efforts are underway to develop high density carbon materials as well as amorphous isotropic carbon for the application in thermal reactors. There has been research on coated nuclear fuel for high temperature reactor and research and development on coated fuels are now focused on fuel particles with high endurance during normal lifetime of the reactor. Since graphite as a moderator as well as structural material in high temperature reactors is one of the most favored choices, it is now felt to develop high density isotropic graphite with suitable coating for safe application of carbon based materials even in oxidizing or water vapor environment. Carboncarbon composite materials compared to conventional graphite materials are now being looked into as the promising materials for the fusion reactor due their ability to have high thermal conductivity and high thermal shock resistance. This paper deals with the application of carbon materials on various nuclear reactors related issues and addresses the current need for focused research on novel carbon materials for future new generation nuclear reactors.

• Carbon letters
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• 제7권1호
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• pp.19-26
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• 2006

Three point bending tests of single edge notched beam (SENB) specimens were carried out to evaluate the fracture behavior of the fine-grain isotropic nuclear grade graphite, IG-11. To measure the crack initiation point and the subsequent crack growth, the direct current potential drop (DCPD) method and a traveling microscope were used. The effects of test variables like initial crack length, specimen thickness, notch type and loading rate on the measured fracture toughness, $K_Q$, were investigated. Based on the test results, the ranges of the test variables to measure the reliable fracture toughness value were proposed. During the crack growth, the rising R-curve behavior was observed in IG-11 graphite when the superficial crack length measured on the specimen surface was used. The increase of crack growth resistance was discussed in terms of crack bridging, crack meandering, crack branching, microcracking and crack deflection, which increase the surface energy and friction force.

• 한국세라믹학회지
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• 제43권10호
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• pp.613-618
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• 2006

Graphite is suitable materials as a moderator, reflector, and supporter of a nuclear reactor because of high tolerance to the high temperature and neutron irradiations. Because graphite is so weak to the oxidation, its oxidation study is essentially demanded for the operation and design of the nuclear reactor. This work focuses on the effect of the surface oxidation of graphite according to the surface treatment. With thermogravimeter (TG), oxidation characteristics of the isotropic graphite are measured at the three temperature areas, and oxidation ratio and amounts are estimated as changing the surface roughness. Furthermore, the polished graphite surface produced fom the surface treatment is investigated with the Raman spectroscopic study. Oxidation behaviors of the surface are also evaluated as elimination the polished layer by washing with strong sonication.

• 한국세라믹학회지
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• 제43권12호
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• pp.833-838
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• 2006

Graphite is suitable for high temperature structural materials because of chemical stability as well as unique crystal structure. Especially, graphite can be used as a part of a nuclear reactor due to high tolerance at the extreme conditions of high temperature and neutron irradiations. Although study of oxidation properties or behaviors of graphite are very important and essential for the life and stability of the nuclear reactor, most of studies treat this theme lightly. This work focuses on the oxidation characteristics of several grade isotropic graphite of the nuclear reactor.

• Carbon letters
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• 제7권3호
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• pp.196-200
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• 2006

Graphite has hexagonal closed packing structure with two bonding characteristics of van der Waals bonding between the carbon layers at c axis, and covalent bonding in the carbon layer at a and b axis. Graphite has high tolerant to the extreme conditions of high temperature and neutron irradiations rather than any other materials of metals and ceramics. However, carbon elements easily react with oxygen at as low as 400C. Considering the increasing production of today of hydrogen and electricity with a nuclear reactor, study of oxidation characteristics of graphite is very important, and essential for the life evaluation and design of the nuclear reactor. Since the oxidation behaviors of graphite are dependent on the shapes of testing specimen, critical care is required for evaluation of nuclear reactor graphite materials. In this work, oxidation rate and amounts of the isotropic graphite (IG-110, Toyo Carbon), currently being used for the Koran nuclear reactor, are investigated at various temperature. Oxidation process or principle of graphite was figured out by measuring the oxidation rate, and relation between oxidation rate and sample shape are understood. In the oxidation process, shape effect of volume, surface area, and surface to volume ratio are investigated at $600^{\circ}C$, based on the sample of ASTM C 1179-91.

• 대한금속재료학회지
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• 제49권7호
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• pp.557-564
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• 2011

This paper reports the relationship between the surface roughness and thermal emissivity of graphite (IG-11) in nuclear reactors. The roughness was controlled by changing the oxidization time, resulting in 0, 6, and 11% losses of mass. The levels of roughness were 0.40, 0.72 and 1.09${\mu}m$ for the weight loss of 0, 6 and 11%, respectively. The binders and graphite fillers were found to have sequentially oxidized with a higher thermal emission for the highly oxidized sample, but with a lower emission when measured at a higher temperature. Our study suggests a method for predicting the thermal emission rate of graphite in a nuclear reactor based on roughness measurement.

• Carbon letters
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• 제16권3호
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• pp.135-146
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• 2015

Graphite can be classified into natural graphite from mines and artificial graphite. Due to its outstanding properties such as light weight, thermal resistance, electrical conductivity, thermal conductivity, chemical stability, and high-temperature strength, artificial graphite is used across various industries in powder form and bulk form. Artificial graphite of powder form is usually used as anode materials for secondary cells, while artificial graphite of bulk form is used in steelmaking electrode bars, nuclear reactor moderators, silicon ingots for semiconductors, and manufacturing equipment. This study defines artificial graphite as bulk graphite, and provides an overview of bulk graphite manufacturing, including isotropic and anisotropic materials, molding methods, and heat treatment.

• Nuclear Engineering and Technology
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• 제52권6호
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• pp.1243-1251
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• 2020

Irradiation-induced damage of binderless nanoporous-isotropic graphite (NPIG) prepared by isostatic pressing of mesophase carbon microspheres for molten salt reactor was investigated by 3.0 MeV He⁺ irradiation at room temperature and high temperature of 600 ℃, and IG-110 was used as the comparation. SEM, TEM, X-ray diffraction and Raman spectrum are used to characterize the irradiation effect and the influence of temperature on graphite radiation damage. After irradiation at room temperature, the surface morphology is rougher, the increase of defect clusters makes atom flour bend, the layer spacing increases, and the catalytic graphitization phenomenon of NPIG is observed. However, the density of defects in high temperature environment decreases and other changes are not obvious. Mechanical properties also change due to changes in defects. In addition, SEM and Raman spectra of the cross section show that cracks appear in the depth range of the maximum irradiation dose, and the defect density increases with the increase of irradiation dose.