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Watanabe, Kazuhito; Nakamura, Makoto; Tobita, Kenji; Someya, Yoji; Tanigawa, Hisashi; Uto, Hiroyasu; Sakamoto, Yoshiteru; Araki, Takao*; Asano, Shiro*; Asano, Kazuhito*
Proceedings of 26th IEEE Symposium on Fusion Engineering (SOFE 2015), 6 Pages, 2016/06
Safety studies of a water-cooled fusion DEMO reactor have been performed. In the event of the blanket cooling pipe break outside the vacuum vessel, i.e. ex-vacuum vessel loss of coolant accident (ex-VV LOCA), the pressurized steam and air may lead to damage reactor building walls which have confinement function, and to release the radioactive materials to the environment. In response to this accident, we proposed three cases of confinement strategies. In each case, the pressure and thermal loads to the confinement boundaries and total mass of tritium released to outside the boundaries were analyzed by accident analysis code MELCOR modified for fusion reactor. These analyses developed design parameters to maintain the integrity of the confinement boundaries.
Nakamura, Makoto; Tobita, Kenji; Gulden, W.*; Watanabe, Kazuhito*; Someya, Yoji; Tanigawa, Hisashi; Sakamoto, Yoshiteru; Araki, Takao*; Matsumiya, Hisato*; Ishii, Kyoko*; et al.
Fusion Engineering and Design, 89(9-10), p.2028 - 2032, 2014/10
Times Cited Count：13 Percentile：70.2(Nuclear Science & Technology)After the Fukushima Dai-ichi nuclear accident, a social need for assuring safety of fusion energy has grown gradually in the Japanese (JA) fusion research community. DEMO safety research has been launched as a part of BA DEMO Design Activities (BA-DDA). This paper reports progress in the fusion DEMO safety research conducted under BA-DDA. Safety requirements and evaluation guidelines have been, first of all, established based on those established in the Japanese ITER site invitation activities. The amounts of radioactive source terms and energies that can mobilize such source terms have been assessed for a reference DEMO, in which the blanket technology is based on the Japanese fusion technology R&D programme. Reference event sequences expected in DEMO have been analyzed based on the master logic diagram and functional FMEA techniques. Accident initiators of particular importance in DEMO have been selected based on the event sequence analysis.
Nakamura, Makoto; Tobita, Kenji; Someya, Yoji; Tanigawa, Hisashi; Gulden, W.*; Sakamoto, Yoshiteru; Araki, Takao*; Watanabe, Kazuhito*; Matsumiya, Hisato*; Ishii, Kyoko*; et al.
Plasma and Fusion Research (Internet), 9, p.1405139_1 - 1405139_11, 2014/10
Key aspects of the safety study of a water-cooled fusion DEMO reactor is reported. Safety requirements, dose target, DEMO plant model and confinement strategy of the safety study are briefly introduced. The internal hazard of a water-cooled DEMO, i.e. radioactive inventories, stored energies that can mobilize these inventories and accident initiators and scenarios, are evaluated. It is pointed out that the enthalpy in the first wall/blanket cooling loops, the decay heat and the energy potentially released by the Be-steam chemical reaction are of special concern for the water-cooled DEMO. An ex-vessel loss-of-coolant of the first wall/blanket cooling loop is also quantitatively analyzed. The integrity of the building against the ex-VV LOCA is discussed.
Shibama, Yusuke; Masaki, Kei; Sakurai, Shinji; Shibanuma, Kiyoshi; Sakasai, Akira; Onawa, Toshio*; Araki, Takao*; Asano, Shiro*
Fusion Engineering and Design, 88(9-10), p.1916 - 1919, 2013/10
Times Cited Count：2 Percentile：18.63(Nuclear Science & Technology)This presentation focuses on the welding technology R&D between the JT-60SA vacuum vessel and the ports. The vacuum vessel is designed to allow port bore penetration to access the vessel inside for plasma diagnostics, and so on. There are various types of 73 ports and these are categorized by their locations; the upper/lower vertical, the upper/lower oblique, and the horizontal. Ports are onsite-welded onto the VV port stub after the assembly of the VV. This assembly sequence involves the out-vessel components such as VV thermal shield and toroidal field magnets, so that these ports welding are accessed from the inside of the vessel and limited by the internal port wall. The one of the most difficult ports are the upper vertical port with corner radius of 50 mm under narrow space, and it is necessary to clarify mobility of the weld torch head. The port weldability is discussed with the mock-up trial, which consists of the partial test pieces of the product size. The TIG welding manipulator, optimized for this R&D, is prepared by its operational simulation and examined not to interfere with the internal port wall.
Ohira, Shigeru; Tada, Eisuke; Hada, Kazuhiko; Neyatani, Yuzuru; Maruo, Takeshi; Hashimoto, Masayoshi*; Araki, Takao*; Nomoto, Kazuhiro*; Tsuru, Daigo; Ishida, Toshikatsu*; et al.
Fusion Science and Technology, 41(3), p.642 - 646, 2002/05
no abstracts in English
Tsuru, Daigo; Neyatani, Yuzuru; Araki, Takao*; Nomoto, Kazuhiro*; Ohira, Shigeru; Maruo, Takeshi; Hashimoto, Masayoshi*; Hada, Kazuhiko; Tada, Eisuke
Fusion Engineering and Design, 58-59, p.985 - 989, 2001/11
Times Cited Count：4 Percentile：33.39(Nuclear Science & Technology)no abstracts in English
Khan, A.*; Yamaguchi, Masafumi*; Oshita, Yoshio*; Dharmarasu, N.*; Araki, Kenji*; Abe, Takao*; Ito, Hisayoshi; Oshima, Takeshi; Imaizumi, Mitsuru*; Matsuda, Sumio*
Journal of Applied Physics, 90(3), p.1170 - 1178, 2001/08
Times Cited Count：56 Percentile：86.43(Physics, Applied)1MeV-electron and 10MeV-proton irradiations into Si doped various impurities such as B, Ga, O and C were performed and residual defects in the Si were studied using DLTS and C-V measurements.It was revealed that Ci-Oi whose level is Ev-0.36 eV and Bi-Oi whose energy is Ec-0.18eV were generated. In Ga-doped Si, the generation of Ci-Oi was suppressed. Since Ci-Oi acts as scattering center, this result indicates that the radiation resistance of solar cells is improved by using Ga-doped Si substrates.Furthermore, a new defect level (Ev+18eV) was observed in Ga-dpoed Si by irradiation. This defect level was annealed out above 350 C.
Ohira, Shigeru; Tada, Eisuke; Hada, Kazuhiko; Neyatani, Yuzuru; Maruo, Takeshi; Hashimoto, Masayoshi*; Araki, Takao*; Nomoto, Kazuhiro*; Tsuru, Daigo; Ishida, Toshikatsu*; et al.
Fusion Engineering and Design, 54(3-4), p.515 - 522, 2001/04
Times Cited Count：3 Percentile：27.1(Nuclear Science & Technology)no abstracts in English
Kuriyama, Masaaki; Akiba, Masato; ; Araki, Masanori; Dairaku, Masayuki; ; Horiike, Hiroshi; Ito, Takao; Inoue, Takashi; Kawai, Mikito; et al.
JAERI-M 87-169, 182 Pages, 1987/10
no abstracts in English
Matsuda, Shinzaburo; Akiba, Masato; Araki, Masanori; Dairaku, Masayuki; ; Horiike, Hiroshi; Ito, Takao; *; Kawai, Mikito; Komata, Masao; et al.
Fusion Engineering and Design, 5, p.85 - 100, 1987/00
Times Cited Count：23 Percentile：87.82(Nuclear Science & Technology)no abstracts in English
Ejiri, Mitsuru*; Kitamura, Kazunori*; Araki, Takao*; Omori, Junji*; Asano, Shiro*; Hayakawa, Atsuro*; Shibama, Yusuke; Masaki, Kei; Sakasai, Akira
no journal, ,
In the operation of tokamak, such loads as electromagnetic and seismic are assumed to be imposed on the vacuum vessel (VV), and not a little thermal expansion takes place when VV is baked. The gravity support leg (GS) has to support the loads described above in addition to the dead weight of VV including in-vessel components and compensate deformation. The GS is equipped with plate spring (PS) to have both stiffness and flexibility. In this study, the buckling strength of the PSs was evaluated. The effect of the initial imperfection of the PSs which is assumed to result from machining or welding process on the buckling strength was also studied. It is concluded that GS has sufficient buckling strength against assumed initial imperfections.
Asano, Shiro*; Okuyama, Toshihisa*; Ejiri, Mitsuru*; Mizumaki, Shoichi*; Mochida, Tsutomu*; Hamada, Takashi*; Araki, Takao*; Hayakawa, Atsuro*; Sagawa, Keiich*; Kai, Toshiya*; et al.
no journal, ,
no abstracts in English
Nakamura, Makoto; Tobita, Kenji; Someya, Yoji; Tanigawa, Hisashi; Araki, Takao*; Watanabe, Kazuhito*; Kittaka, Daigo*; Ishii, Kyoko*; Matsumiya, Hisato*
no journal, ,
Recent findings on safety characteristics of a tokamak DEMO reactor are reported in the case where all the coolant water is lost completely and instantaneously. Assuming that there are neither off-site power nor active emergency cooling, we have analyzed temporal histories of the temperatures of the reactor components using the fusion reactor thermo-hydraulic analysis code MELCOR-fus. We have found that even in such an extremely severe case, the temperatures of the vacuum vessel and in-vessel components do not reach their melting points.
Ejiri, Mitsuru*; Asano, Shiro*; Omori, Junji*; Okuyama, Toshihisa*; Takahashi, Nobuji*; Yamada, Masahiro*; Araki, Takao*; Kai, Toshiya*; Shibama, Yusuke; Masaki, Kei; et al.
no journal, ,
In the operation of Tokamak device, such loads as electromagnetic and seismic are assumed to be imposed on the vacuum vessel (VV), and not a little thermal expansion takes place when VV is baked. The gravity support (GS) has to support the loads described above in addition to the dead weight of VV including in-vessel components and compensate deformation. The GS is equipped with leaf spring that has both stiffness and flexibility. In this study, the FEM analysis-based design and assembly procedure of the GS is reported. The manufacturing process of GS components is also reported with trial manufacturing results.
