• Nuclear Engineering and Technology
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• v.51 no.5
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• pp.1390-1397
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• 2019

SiC coating and SiC/glassy carbon composite coating were prepared on IG-110 nuclear graphite (Toyo Tanso Co., Ltd., Japan) to strengthen its inertness to molten fluoride salt used in molten salt reactor (MSR). Two kinds of modified graphite were obtained and correspondingly named as IG-110-1 and IG-110-2, which referred to modified IG-110 with a single SiC coating and a SiC/glassy carbon composite coating, respectively. Both structure and property of modified graphite were carefully researched and contrasted with virgin IG-110. Results indicated that modified graphite presented better comprehensive properties such as more compact structure and higher resistance to molten salt infiltration. With the protection of coatings, the infiltration amounts of fluoride salt into modified graphite were much less than that into virgin IG-110 at the same circumstance. Especially, the infiltration amount of fluoride salt into IG-110-2 under 5 atm was merely 0.26 wt%, which was much less than that into virgin IG-110 under 1.5 atm (13.5 wt%) and the critical index proposed for nuclear graphite used in MSR (0.5 wt%). The SiC/glassy carbon composite coating gave rise to highest resistance to molten salt infiltration into IG-110-2, and thus demonstrated it could be a promising protective coating for nuclear graphite used in MSR.

• Journal of the Korean Ceramic Society
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• v.43 no.7 s.290
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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.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.38 no.5
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• pp.469-474
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• 2014

The high temperature gas-cooled reactor (HTR-10) is designed to produce electricity and hydrogen. Graphite is used as reflector, support structures, and a moderator in reactor core; it has good resistance to neutron and is a suitable material at high temperatures. Friction is generated in the graphite structures for the core reflector, support structures, and moderator because of vibration from the HTR-10 fuel cycle flow. In this study, the wear characteristics of the isotropic graphite IG-110 used in HTR-10 were evaluated. The reciprocating wear test was carried out for graphite against graphite. The effects of changes in the contact load and sliding speeds at room temperature and $400^{\circ}C$ on the coefficient of friction and specific wear rate were evaluated. The wear behavior of graphite IG-110 was evaluated based on the wear surfaces.

• Nuclear Engineering and Technology
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• v.54 no.7
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• pp.2359-2366
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• 2022

Selecting graphite grades with superior irradiation characteristics is important task for designers of graphite moderation reactors. To provide reference information and data for graphite selection, the effects of irradiation on three fine-grained, iso-molded nuclear grade graphites, ETU-10, IG-110, and NBG-25, were compared based on irradiation-induced changes in volume, thermal conductivity, dynamic Young's modulus, and coefficient of thermal expansion. Data employed in this study were obtained from reported irradiation test results in the high flux isotope reactor (HFIR)(ORNL) (ETU-10, IG-110) and high flux reactor (HFR)(NRL) (IG-110, NBG-25). Comparisons were made based on the irradiation dose and irradiation temperature. Overall, the three grades showed similar irradiation-induced property change behaviors, which followed the historic data. More or less grade-sensitive behaviors were observed for the changes in volume and thermal conductivity, and, in contrast, grade-insensitive behaviors were observed for dynamic Young's modulus and coefficient of thermal expansion changes. The ETU-10 of the smallest grain size appeared to show a relatively smaller VC to IG-110 and NBG-25. Drastic decrease in the difference in thermal conductivity was observed for ETU-10 and IG-110 after irradiation. The similar irradiation-induced properties changing behaviors observed in this study especially in the DYM and CTE may be attributed to the assumed similar microstructures that evolved from the similar size coke particles and the same forming method.

• 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.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.

• Proceedings of the Korean Nuclear Society Conference
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• 2005.05a
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• pp.41-42
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• 2005

• Proceedings of the Korean Nuclear Society Conference
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• 2004.10a
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• pp.1005-1006
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• 2004

• Proceedings of the Korean Nuclear Society Conference
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• 2004.10a
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• pp.922-923
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• 2004

• Proceedings of the Korean Nuclear Society Conference
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• 2004.05a
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• pp.329-329
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• 2004