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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.
Shibata, Taiju; Sumita, Junya; Makita, Taiyo*; Takagi, Takashi*; Kunimoto, Eiji; Sawa, Kazuhiro
Ceramics in Nuclear Applications, Vol.30, Issue 10, p.19 - 32, 2009/11
Japan Atomic Energy Agency (JAEA) carries out R&Ds for Very High Temperature Reactor (VHTR). Since the core components in VHTR will be used at severer condition than in the High Temperature Engineering Test Reactor (HTTR), it is important to develop heat-resistant ceramic composite. C/C and SiC/SiC composites are major candidates substitute for metallic control rod. JAEA studies on the application of the composite materials for the VHTR control rod following the scheme; (1) Database establishment, (2) Design and (3) Demonstration test by HTTR. In this paper, the development plan was explained. It was expressed that the irradiation effects on graphitized C/C composite showed similar trend with that of graphite. It is possible to evaluate the irradiation effects based on the existing graphite irradiation database. The HTTR can irradiate large component. It is possible to demonstrate the structural integrity of the mock-up of the composite control rod being necessary for the final phase.
Nozawa, Takashi; Tanigawa, Hiroyasu; Park, J.-S.*; Koyama, Akira*
Proceedings of 33rd International Conference on Advanced Ceramics and Composites (CD-ROM), p.65 - 76, 2009/09
SiC/SiC composites are promising candidate materials for various energy industries. One of key issues for their practical application is to develop design basis of quasi-ductile composites. Understanding the crack propagation behavior is therefore essential. For this purpose, determining a fracture resistances undoubtedly important and this paper aims to evaluate this using various types of notched specimens. With the preliminary analysis by the authors, a new methodology to determine the fracture resistance has been developed based on the non-linear fracture mechanics, separately discussing the effect of irreversible energies. By adopting this method for nano-infiltration transient-eutectic-phase sintered (NITE) SiC/SiC composites, several key findings were obtained: (1) notch insensitivity, (2) no significant size effect, and (3) superior crack tolerance.
Nozawa, Takashi; Jitsukawa, Shiro; Tanigawa, Hiroyasu
no journal, ,
SiC/SiC composites are candidate materials of the advanced fusion DEMO reactor. For the practical application, it is essential to have a reliable and practical database and relevant test technologies. Understanding irradiation effects on mechanical and physical properties is also required. Accordingly, these issues have recently been adopted as one of the key R&Ds for the DEMO development under the IFERC project in BA activities between EU and Japan. The R&D has two tasks: (1) R&D on mechanical properties and (2) R&D on physical/chemical properties. Specifically, in Japan, (1) generation of mechanical property database such as lifetime and strength anisotropy with development of small specimen test techniques, (2) generation of physical property database with development of in-situ measurement technique under irradiation, and (3) development of a basic model considering irradiation effects on mechanical properties are emphasized.
