• Journal of the Korean Ceramic Society
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• v.22 no.6
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• pp.5-8
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• 1985

Graphite has been oxidized to graphite hydrogen sulfate in concentrated $H_2SO_4$. Anodic oxidation and chemical oxidation of graphite in $H_2SO_4$ generally leads to the formation of intercalation compounds of the ionic salt type through incorporation of $H_2SO_4^-$ions and $H_2SO_4$ molecules into the graphite. Several other reactions also accur at various points of the charging cycle. But there is no satisfactory kinetics and mechanism of intercalationin graphite. We have studied them with anodic oxidation and chemical oxidation. We found six distinct phenomena between 2nd stage and 1st stage in chemical oxidation. We examined them in detail by the following in the measurements electrical oxidation. X-ray diffractions UV-Vis spectroscopy density measurements. We could obtained a equation for kinetic according to the reaction rate from this results and mechanism of intercalation between 2nd stage and 1st stage with hydrogen sulfate in graphite. Three thesis were written for the mechanism of intercalation compounds in graphite with hydrogen sulfate ; first thesis is anodic oxidation second thesis is chemical oxidation and definition of transit phase between 2nd etc the third thesis is the kinetic mechanism of intercalation compounds in graphite with Hydrogen sulfate. This thesis is the first paper among three thesis as anodic oxidation.

• Journal of the Korean Ceramic Society
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• v.43 no.10 s.293
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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.

• Carbon letters
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• v.7 no.3
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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.

• Nuclear Engineering and Technology
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• v.48 no.1
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• pp.182-188
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• 2016

In this paper, a feasible strategy for the recycling of nuclear graphite is reported, based on the formation mechanism and the removal of carbon-14 by micro-oxidation. We investigated whether ground micro-oxidation graphite could be used as a filler to make new recycled graphite and which graphite/pitch coke ratio will give the recycled graphite outstanding properties (e.g., apparent density, flexural strength, compressive strength, and tensile strength). According to the existing properties of nuclear graphite, the ratio of graphite to pitch coke should not exceed 3. The recycled reactor graphite has been proven superior in density, strength, and thermal conductivity. The micro-oxidation process enhances the strength of the recycled graphite because there are more pores and unsmooth surfaces on the oxidized graphite particles, which is beneficial for the access of the pitch binder and leads to efficient joint adhesion among the graphite particles.

• Journal of the Korean Ceramic Society
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• v.36 no.5
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• pp.555-563
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• 1999

Impregnation of phosphorous additiers into graphite bulk was studied with the goal of enhancing the effectiveness of oxidationprotection. In addition graphite acid washing was carried out prior to the impregnation further to improve oxidation resistance. Observation of the oxidation rate for raw graphite(Raw) impregnated graphite with tri-butyl phsophate on raw block(RP) and impregnated graphite on acid-treated graphite(AP) in air are reported. The phsophorus residue adsorbed on the graphite surface at active sites was determined by FTIR, XRS, TGA techniques. AP with tri-butyl phosphate was found to result in both 30% reduction in oxidation rate at 1000$^{\circ}C$ compared to Raw and increase of 120$^{\circ}C$ in oxidation temperature From the samples of oxidation rate of each specimen in Arrhenius plot it can be said that the present oxidation resistance origninates from the change of chemical reaction modesw neigther by acid-washing treatment nor phsophate impregnation

• Journal of the Korean Ceramic Society
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• v.43 no.12 s.295
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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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• v.10 no.4
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• pp.300-304
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• 2009

Thermal emissivity of nuclear graphite was measured with its oxidation degree. Commercial nuclear graphites (IG-110, PECA, IG-430, and NBG-18) have been used as samples. Concave on graphites surface increased as its oxidation degree increased, and R value (Id/Ig) of the graphites decreased as the oxidation degree increased. The thermal emissivity increased depending on the decrease of the R (Id/Ig) value through Raman spectroscopy analysis. It was determined that the thermal emissivity was influenced by the crystallinity of the nuclear graphite.

• Journal of the Korean Ceramic Society
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• v.34 no.12
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• pp.1199-1204
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• 1997

SiC conversion layer was fabricated by the chemical vapor reaction between graphite substrate and silica powder. The CVR process was carried out in nitrogen atmosphere at 175$0^{\circ}C$ and 185$0^{\circ}C$. From the reduction of silica powder with graphite substrate, the SiO vapor was created, infiltrated into the graphite substrate, then, the SiC conversion layer was formed from the vapor-solid reaction of SiO and graphite. In the XRD pattern of conversion layer, it was confirmed that 3C $\beta$-SiC phase was created at 175$0^{\circ}C$ and 185$0^{\circ}C$. Also, in the back scattered image of cross-sectional conversion layer, it was found that the conversion layer was easily formed at 185$0^{\circ}C$, the interface of graphite substrate and SiC layer was observed. It was though that the coke particle size and density of graphite substrate mainly affect the XRD pattern and microstructure of SiC conversion layer. In the oxidation test of 100$0^{\circ}C$, the SiC converted graphites exhibited good oxidation resistance compared with the unconverted graphites.

• Nuclear Engineering and Technology
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• v.45 no.2
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• pp.211-218
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• 2013

Large quantities of irradiated graphite waste from graphite-moderated nuclear reactors exist and are expected to increase in the case of High Temperature Reactor (HTR) deployment [1,2]. This situation indicates the need for a graphite waste management strategy. Of greatest concern for long-term disposal of irradiated graphite is carbon-14 ($^{14}C$), with a half-life of 5730 years. Fachinger et al. [2] have demonstrated that thermal treatment of irradiated graphite removes a significant fraction of the $^{14}C$, which tends to be concentrated on the graphite surface. During thermal treatment, graphite surface carbon atoms interact with naturally adsorbed oxygen complexes to create $CO_x$ gases, i.e. "gasify" graphite. The effectiveness of this process is highly dependent on the availability of adsorbed oxygen compounds. The quantity and form of adsorbed oxygen complexes in pre- and post-irradiated graphite were studied using Time of Flight Secondary Ion Mass Spectrometry (ToF-SIMS) and Xray Photoelectron Spectroscopy (XPS) in an effort to better understand the gasification process and to apply that understanding to process optimization. Adsorbed oxygen fragments were detected on both irradiated and unirradiated graphite; however, carbon-oxygen bonds were identified only on the irradiated material. This difference is likely due to a large number of carbon active sites associated with the higher lattice disorder resulting from irradiation. Results of XPS analysis also indicated the potential bonding structures of the oxygen fragments removed during surface impingement. Ester- and carboxyl-like structures were predominant among the identified oxygen-containing fragments. The indicated structures are consistent with those characterized by Fanning and Vannice [3] and later incorporated into an oxidation kinetics model by El-Genk and Tournier [4]. Based on the predicted desorption mechanisms of carbon oxides from the identified compounds, it is expected that a majority of the graphite should gasify as carbon monoxide (CO) rather than carbon dioxide ($CO_2$). Therefore, to optimize the efficiency of thermal treatment the graphite should be heated to temperatures above the surface decomposition temperature increasing the evolution of CO [4].

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2004.06a
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• pp.105-114
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• 2004

This study investigated the thermal behavior of the nuclear graphite waste generated from the decommissioning of the Korean nuclear research reactor, The first part study investigated the decomposition rate of the nuclear graphite waste up to $1000^{\circ}C$ under various oxygen partial pressures using a thermo-gravimetric analyzer (TGA). Tested graphite waste sample not easily destroyed in the oxygen-deficient condition. However, the gas-solid oxidation reaction was found to be very effective in the presence of oxygen. No significant amount of the product of incomplete combustion was formed even in the limited oxygen concentration of 4% $O_2$. The influence of temperature and oxygen partial pressure was evaluated by the theoretical model analysis of the thermo-gravimetric data. The activation energy and the reaction order of graphite oxidation were evaluated as 128 kJ/mole and 1.1, respectively. The second part of this study investigated the behavior of radioactive elements under graphite oxidation atmosphere using thermodynamic equilibrium model. $^{22}Na$, $^{134}Cs$ and $^{137}Cs$ were found be the semi-volatile elements. Since volatile uranium species can be formulated at high temperatures above $1050^{\circ}C$, the temperature of incinerator furnace should be minimized. Other corrosion/activation products, fission products and uranium were found to be the non-volatile species.