J-PARC E19 experiment; Pentaquark search in hadronic reaction at J-PARC

Takahashi, Tomonori*; Ekawa, Hiroyuki; Hayakawa, Shuhei; Hosomi, Kenji; Ichikawa, Yudai; Imai, Kenichi; Sako, Hiroyuki; Sato, Susumu; Sugimura, Hitoshi; Tanida, Kiyoshi; et al.

JPS Conference Proceedings (Internet), 8, p.022011_1 - 022011_6, 2015/09

https://doi.org/10.7566/JPSCP.8.022011

Overview of national centralized tokamak program; Mission, design and strategy to contribute ITER and DEMO

Ninomiya, Hiromasa; Akiba, Masato; Fujii, Tsuneyuki; Fujita, Takaaki; Fujiwara, Masami*; Hamamatsu, Kiyotaka; Hayashi, Nobuhiko; Hosogane, Nobuyuki; Ikeda, Yoshitaka; Inoue, Nobuyuki; et al.

Journal of the Korean Physical Society, 49, p.S428 - S432, 2006/12

To contribute DEMO and ITER, the design to modify the present JT-60U into superconducting coil machine, named National Centralized Tokamak (NCT), is being progressed under nationwide collaborations in Japan. Mission, design and strategy of this NCT program is summarized.

Overview of the national centralized tokamak programme

Kikuchi, Mitsuru; Tamai, Hiroshi; Matsukawa, Makoto; Fujita, Takaaki; Takase, Yuichi*; Sakurai, Shinji; Kizu, Kaname; Tsuchiya, Katsuhiko; Kurita, Genichi; Morioka, Atsuhiko; et al.

Nuclear Fusion, 46(3), p.S29 - S38, 2006/03

https://doi.org/10.1088/0029-5515/46/3/S05

The National Centralized Tokamak (NCT) facility program is a domestic research program for advanced tokamak research to succeed JT-60U incorporating Japanese university accomplishments. The mission of NCT is to establish high beta steady-state operation for DEMO and to contribute to ITER. The machine flexibility and mobility is pursued in aspect ratio and shape controllability, feedback control of resistive wall modes, wide current and pressure profile control capability for the demonstration of the high-b steady state.

Fracture mechanics analysis including the butt joint geometry for the superconducting conductor conduit of the national centralized tokamak

Takahashi, Hiroyuki*; Kudo, Yusuke; Tsuchiya, Katsuhiko; Kizu, Kaname; Ando, Toshinari*; Matsukawa, Makoto; Tamai, Hiroshi; Miura, Yukitoshi

Fusion Engineering and Design, 81(8-14), p.1005 - 1011, 2006/02

https://doi.org/10.1016/j.fusengdes.2005.07.028

This paper presents dependence of the stress intensity factor, around the defect in the butt joint welding of a superconducting conductor conduit, on a geometrical factor estimated by fracture mechanics analysis. The stress intensity factor can be estimated by the Newman-Raju equation about CICC section, but the effect of the difference between the geometry assumed in the equation and CICC has not been clarified yet. Therefore, the three-dimensional finite element method (3D-FEM) is performed to estimate the geometrical factor. As a result, the Newman-Raju equation is considered to be available for the assessment of the fracture toughness of the conduit of rectangular shape because the maximum stress intensity factor by 3-D FEM is only 3% larger than that by the Newman-Raju equation in the maximum postulated defect.

Aging deterioration test of seismic isolation applied to fusion experimental reactor

Takeda, Nobukazu; Nakahira, Masataka; Kakudate, Satoshi; Takahashi, Hiroyuki*; Shibanuma, Kiyoshi; Yabana, Shuichi*; Matsuda, Akihiro*

Proceedings of 9th World Seminar on Seismic Isolation, Energy Dissipation and Active Vibration Control of Structures (CD-ROM), p.299 - 306, 2005/00

For the ITER, a fusion experimental reactor, it is planned to use rubber bearings in order to enhance the reliability of integrity with a sufficient margin even for the earthquakes beyond the design basis earthquake. In application for nuclear plants, the vertical compression of the isolator is 25 MPa and there is no experience for such a high compression as 10 MPa to be used for the ITER. Therefore, there is not enough design data of the rubber bearings with high compression, and thus a detailed estimation of performance is necessary. As a result of the endurance test after aging, it was validated that the bearing can be applied safely until 400th cycle even after 40 years of aging. On the other hand, the residual deformation was found at the 246th cycle. This means that the residual deformation can be observed enough earlier than the change of the macroscopic mechanical parameter such as stiffness. Therefore, it is possible to prevent break of the bearing during operation by sensing a sign of break with a periodical visual inspection.

Tensile and fatigue strength of a through-wall-electron-beam-welded joint for the vacuum vessel of a fusion reactor

Suzuki, Takayuki*; Usami, Saburo*; Kimura, Takae*; Koizumi, Koichi; Nakahira, Masataka; Takahashi, Hiroyuki*

Proceedings of 55th Annual Assembly of International Institute of Welding (IIW2002), 16 Pages, 2002/06

A new type of welded joint for the outer wall and rib of a double-walled vacuum vessel of a fusion reactor has been developed. The joint is manufactured by through-wall electron-beam welding (TW-EBW), in which the beam is injected from the outside of the outer wall. Static and fatigue tests are carried out on one-bead-specimens under an axial load and two-bead-specimens under a bending load. The experimental results are analytically investigated by FEM. Although this joint is partially penetrated, the net yield strength of the bead is increased by the plastic constraint due to triaxial tensile stress in the weldment. This phenomenon reduces the mean equivalent stress on the bead cross section, and the gross strength of the joint is close to that of a full thickness welded joint. The fatigue-strength reduction factor for low-cycle fatigue life is a little larger than four. The calculated fatigue-crack growth rate in the joint is conservatively calculated by using the maximum stress intensity factor of the crack and the fatigue-crack growth rate given in ASME Code Section XI.

Characteristic evaluation of high compression seismic isolator for International Thermonuclear Experimental Reactor (ITER); Verification test of sub-scaled rubber bearings (Contract research)

Takahashi, Hiroyuki*; Nakahira, Masataka; Yabana, Shuichi*; Matsuda, Akihiro*; Otori, Yasuki*

JAERI-Tech 2001-064, 111 Pages, 2001/11

JAERI-Tech-2001-064.pdf:8.96MB

no abstracts in English

Mechanical properties of type 316L stainless steel welded joint for vacuum vessel of ITER, 2; Neutron irradiation tests and post-irradiation experiments

Saito, Shigeru; Fukaya, Kiyoshi; Ishiyama, Shintaro; Amezawa, Hiroo; Yonekawa, Minoru; Takada, Fumiki; Kato, Yoshiaki; Takeda, Takashi; Takahashi, Hiroyuki*; Koizumi, Koichi

JAERI-Tech 2001-035, 81 Pages, 2001/06

JAERI-Tech-2001-035.pdf:18.91MB

no abstracts in English

Mechanical properties of type 316L stainless steel welded joint for ITER vacuum vessel, 1; Experiment of unirradiated welded joint

Saito, Shigeru; Fukaya, Kiyoshi; Ishiyama, Shintaro; Takahashi, Hiroyuki*; Koizumi, Koichi

JAERI-Tech 2000-075, 98 Pages, 2001/01

JAERI-Tech-2000-075.pdf:21.85MB

no abstracts in English

ITER cryostat thermal shield detailed design

*; Nakahira, Masataka; Hamada, Kazuya; Takahashi, Hiroyuki*; Tada, Eisuke; Kato, Takashi; *

JAERI-Tech 99-027, 113 Pages, 1999/03

JAERI-Tech-99-027.pdf:7.4MB

no abstracts in English