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Ishitsuka, Etsuo; Ho, H. Q.; Kitagawa, Kanta*; Fukuda, Takahito*; Ito, Ryo*; Nemoto, Masaya*; Kusunoki, Hayato*; Nomura, Takuro*; Nagase, Sota*; Hashimoto, Haruki*; et al.
JAEA-Technology 2023-013, 19 Pages, 2023/06
JAEA-Technology-2023-013.pdf:1.75MB
Eight people from five universities participated in the 2022 summer holiday practical training with the theme of "Technical development on HTTR". The participants practiced the feasibility study for nuclear battery, the burn-up analysis of HTTR core, the feasibility study for 
Cf production, the analysis of behavior on loss of forced cooling test, and the thermal-hydraulic analysis near reactor pressure vessel. In the questionnaire after this training, there were impressions such as that it was useful as a work experience, that some students found it useful for their own research, and that discussion with other university students was a good experience. These impressions suggest that this training was generally evaluated as good.
Fujimoto, Nozomu*; Fukuda, Kodai*; Honda, Yuki*; Tochio, Daisuke; Ho, H. Q.; Nagasumi, Satoru; Ishii, Toshiaki; Hamamoto, Shimpei; Nakano, Yumi*; Ishitsuka, Etsuo
JAEA-Technology 2021-008, 23 Pages, 2021/06
JAEA-Technology-2021-008.pdf:2.62MB
The effect of mesh division around the burnable poison rod on the burnup calculation of the HTTR core was investigated using the SRAC code system. As a result, the mesh division inside the burnable poison rod does not have a large effect on the burnup calculation, and the effective multiplication factor is closer to the measured value than the conventional calculation by dividing the graphite region around the burnable poison rod into a mesh. It became clear that the mesh division of the graphite region around the burnable poison rod is important for more appropriately evaluating the burnup behavior of the HTTR core..
Takizawa, Kentaro*; Kakehashi, Kazuyuki*; Fukuda, Toshiaki*; Kida, Toru*; Sawa, Kazuhiro; Sumita, Junya; Kato, Yutai*; Snead, L. L.*
Ceramic Materials for Energy Applications; Ceramic Engineering and Science Proceedings, Vol.32, No.9, p.13 - 19, 2011/11
Fine-grained isotropic graphite shows high strength making it a promising material for the graphite component of High Temperature Gas-cooled Reactor (HTGR) and Very High Temperature Reactor (VHTR). Service life of the graphite component is determined primarily by the residual strength after neutron irradiation in the reactor core. It is expected that development of a new nuclear grade graphite possessing higher strength will contribute toward added design margins and an extension of the service life of components, which likely improve the reactor economy very significantly. Tokai Carbon Co. LTD. has started the development of nuclear grade graphite for the graphite component of VHTR. G347S and G458S grades are fine-grained isotropic graphites having high tensile strength greater than 30 MPa. It is planned to carry out the neutron irradiation tests using High Flux Isotope Reactor at Oak Ridge National Laboratory up to the neutron fluence of 30 dpa and the irradiation temperatures of 300-900
C. The dimensional changes, elastic modulus, coefficient of thermal expansion, etc., will be studied. It is also planned to evaluate the non-irradiated mechanical/thermal properties and the irradiation effects in collaboration with Japan Atomic Energy Agency. This paper introduces our technical R&D plan for G347S and G458S. The initial results of the properties and the irradiation test plan are also shown.
Takizuka, Tomonori; Fukuda, Takeshi; Kamada, Yutaka; Kikuchi, Mitsuru; Matsuda, Toshiaki; Miura, Yukitoshi; Naito, Osamu; Tamai, Hiroshi; D.Boucher*; G.Bracco*; et al.
Fusion Energy 1996, 2, p.795 - 806, 1997/00
no abstracts in English
Nagaishi, Yoshihide*; Fukuda, Toshiaki*; Kondo, Akira*; Sumita, Junya; Sakaba, Nariaki
no journal, ,
Fine-Grained isotropic graphite shows higher strength making it a promising material for the graphite component of High Temperature Gas-cooled Reactor (HTGR) and Very High Temperature Reactor (VHTR). There are two kinds of considerations about lifetime of graphite components under neutron irradiation condition. One consideration is determined based on dimensional changes due to neutron irradiation. In this consideration, small dimensional change with much amount of neutron irradiation is preferable. The other consideration is determined based on a margin between the specified minimum ultimate strengths and the residual stresses induced by neutron irradiation. In this consideration, strength of graphite components is important property to decide lifetime. It is expected that development of new fine-grained isotropic nuclear grade graphite possessing higher strength will contribute toward added design margins and an extension of the lifetime of graphite components, which likely improve the reactor economy very significantly. Furthermore, from the viewpoint of graphite waste, it is also expected to decrease the exchange frequency of core components. Tokai Carbon Co., Ltd. has developed new nuclear grade graphite with higher strength and characterization of them is being carried out. This presentation shows the R&D plan and the initial results of mechanical and thermal properties of un-irradiated nuclear grade graphite.
Fujitsuka, Kunihiro*; Kondo, Akira*; Sumita, Junya; Sakaba, Nariaki; Fukuda, Toshiaki*
no journal, ,
High Temperature Gas-cooled Reactor (HTGR) including High Temperature engineering Test Reactor (HTTR) constructed by Japan Atomic Energy Agency (JAEA) and Very High Temperature Reactor (VHTR) are able to supply high temperature helium gas at the reactor outlet, and possess inherent safety. Fine-Grained isotropic graphite with high strength is used for the in-components of HTGR. In the case of severe accident, for example air ingress accident, it is desirable that oxidation-resistant graphite is used for the in-component of HTGR. One of the effective methods for oxidation-resistant property on graphite is to coat the surface with ceramics, such as SiC, which forms SiO
layer when it oxidizes. This method is one of the general industrial methods called chemical vapor deposition (CVD). Tokai Carbon Co., Ltd. (Tokai carbon) and JAEA launched R&Ds to develop the oxidation-resistant graphite. Tokai Carbon manufactured the oxidation-resistant graphite coated with SiC by CVD etc., and oxidation test and observation of surface condition before and after oxidation were carried out. Moreover, irradiation test using WWR-K research reactor in Institute of Nuclear Physics of the Republic of Kazakhstan (INP) was carried out and post-irradiation examination will begin soon. This presentation shows the oxidation behaviour of the oxidation-resistant graphite according to weight change of specimens, and the surface condition of the specimen according to surface and cross section observation of specimens before and after oxidation.
