Irradiation test plan of oxidation-resistant graphite in WWR-K research reactor

Sumita, Junya; Shibata, Taiju; Sakaba, Nariaki; Osaki, Hiroki*; Kato, Hideki*; Fujitsuka, Kunihiro*; Muto, Takenori*; Gizatulin, S.*; Shaimerdenov, A.*; Dyussambayev, D.*; et al.

Proceedings of 7th International Topical Meeting on High Temperature Reactor Technology (HTR 2014) (USB Flash Drive), 7 Pages, 2014/10

Graphite materials are used for the in-core components of High Temperature Gas-cooled Reactor(HTGR)which is a graphite-moderated and helium gas-cooled reactor. In the case of air ingress accident in HTGR, SiO protective layer is formed on the surface of SiC layer in TRISO CFP and oxidation of SiC does not proceed and fission products are retained inside the fuel particle. A new safety concept for the HTGR, called Naturally Safe HTGR, has been recently proposed. To enhance the safety of Naturally Safe HTGR ultimately, it is expected that oxidation-resistant graphite is used for graphite components to prevent the TRISO CFPs and fuel compacts from failure. SiC coating is one of candidate methods for oxidation-resistant graphite. JAEA and four graphite companies launched R&Ds to develop the oxidation-resistant graphite and the International Science and Technology Center(ISTC) partner project with JAEA and INP was launched to investigate the irradiation effects on the oxidation-resistant graphite. To determine grades of the oxidation-resistant graphite which will be adopted as irradiation test, a preliminary oxidation test was carried out. This paper described the results of the preliminary oxidation test, the plan of out-of-pile test, irradiation test and post-irradiation test(PIE)of the oxidation-resistant graphite.

Post irradiation examination results of hydride neutron absorber for fast reactor, 2; Weight measurement, X-ray diffraction

Harada, Akio; Hatakeyama, Yuichi; Honda, Junichi; Matsui, Hiroki; Kurosaki, Ken*; Konashi, Kenji*

no journal, , 

no abstracts in English

Evaluation of material properties of IG-110 and IG-430 from their microstructure

Sumita, Junya; Shibata, Taiju; Osaki, Hiroki*; Eto, Motokuni*; Konishi, Takashi*

no journal, , 

Graphite materials are used for the in-core components of High Temperature Gas-cooled Reactor (HTGR). The HTGR is particularly attractive due to its passive and inherent safety features. The Very High Temperature Reactor (VHTR) is one of the most promising candidates as the Generation-IV nuclear reactor systems. IG-110 graphite having high strength and resistance to oxidation is used in the HTTR of JAEA and HTR-PM in China. IG-110 graphite is also a major candidate for the in-core graphite components of VHTR, too. IG-430 graphite having the higher strength and resistance to oxidation than IG-110 is an advanced candidate for the VHTR. In this study, the elastic modulus and coefficient of thermal expansion of these graphites were measured and correlation of compressive strength and microstructure was evaluated. Moreover, the densification effects on the material properties were discussed from the standpoint of microstructure using X-ray tomography method.

Characterization of oxidation behavior of boron-doped graphite

Osaki, Hiroki*; Sumita, Junya; Shibata, Taiju; Konishi, Takashi*

no journal, , 

Boron-doped graphite is one of candidate materials for oxidation-resistant graphite. It was reported that BO, which was generated by oxidation of BC, prevented boron-doped graphite from oxidation. However, to apply boron-doped graphite to core support graphite structure of HTGR, it is necessary to measure the mechanical and thermal property of oxidized boron-doped graphite. This study reports results of oxidation test, bending strength test and surface observation using GB-210 fabricated by Toyo Tanso Co., Ltd. in order to measure the bending strength of oxidized boron-doped graphite.

Evaluation of material properties of IG-430 graphite for next generation high temperature gas-cooled reactor

Kunimoto, Eiji*; Sumita, Junya; Osaki, Takashi*; Osaki, Hiroki*; Yamaji, Masatoshi*; Konishi, Takashi*

no journal, , 

Graphite materials are used for the in-core components of High Temperature Gas-cooled Reactor (HTGR) which is a graphite-moderated and helium gas-cooled reactor. The major features of the HTGR are that the HTGR can take out the high-temperature helium gas at the reactor outlet and has inherent safety characteristics. The Very High Temperature Reactor (VHTR) is one of the most promising candidates as the Generation-IV nuclear reactor systems. IG-110 graphite having high strength and resistance to oxidation is used in the HTTR of JAEA and HTR-10 in China. Moreover, IG-110 graphite provides highly consistent quality and long-term stable supply. IG-110 graphite is a major candidate for the in-core graphite components of VHTR. IG-430 graphite having the higher strength and resistance to oxidation than IG-110 is an advanced candidate for the VHTR. However, a new material of IG-430 does not have enough databases for the design of HTGR. Therefore, preparation of the necessary database for the design, mechanical and thermal properties, irradiation effect on them, is underway. In this study, the tensile strength, compressive strength and fatigue strength of IG-430 were statistically evaluated and the applicability of IG-430 as HTGR graphite materials was discussed. Moreover, the Su value of tensile and compressive strength of IG-430 was evaluated and compared to that of IG-110. It was found that IG-430 has excellent properties.

Selective uranium adsorption from liquid waste using zeolites

Matsuura, Haruaki*; Aso, Hiroki*; Toyosaki, Ayaka*; Asanuma, Noriko*; Takahatake, Yoko; Hoshino, Takanori; Watanabe, So; Watanabe, Masayuki

no journal, ,