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Journal Articles

Analysis on ex-vessel loss of coolant accident for a water-cooled fusion DEMO reactor

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.

Journal Articles

Study of safety features and accident scenarios in a fusion DEMO reactor

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:10 Percentile:68.92(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.

Journal Articles

Key aspects of the safety study of a water-cooled fusion DEMO reactor

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.

Journal Articles

Welding technology R&D on port joint of JT-60SA vacuum vessel

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:21.27(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.

Journal Articles

Safety design concepts for ITER-tritium facility; Toward construction in Japan

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

Journal Articles

Study on decay heat removal of compact ITER

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:3 Percentile:28.03(Nuclear Science & Technology)

no abstracts in English

Journal Articles

Role of the impurities in production rates of radiation-induced defects in silicon materials and solar cells

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:50 Percentile:85.51(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.

Journal Articles

Safety activities in JAERI related to ITER

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:28.03(Nuclear Science & Technology)

no abstracts in English

JAEA Reports

The Design research and development of JT-60 neutral beam injector

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

JAERI-M-87-169.pdf:5.53MB

no abstracts in English

Journal Articles

The JT-60 neutral beam injection system

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:22 Percentile:87.46(Nuclear Science & Technology)

no abstracts in English

Oral presentation

Buckling analysis of gravity support legs for JT-60SA vacuum vessel

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.

Oral presentation

Completion of vacuum vessel sector manufacturing and subsequent torus assembly for the JT-60SA

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

Oral presentation

Study of a loss of coolant accident in a tokamak DEMO

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.

Oral presentation

Gravity support design and manufacturing of the JT-60SA vacuum vessel

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.

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