Gyromagnetic bifurcation in a levitated ferromagnetic particle

Sato, Tetsuya*; Kato, Takeo*; Oue, Daigo*; Matsuo, Mamoru

Physical Review B, 107(8), p.L180406_1 - L180406_6, 2023/05

https://doi.org/10.1103/PhysRevB.107.L180406

Multiparticle tunneling transport at strongly correlated interfaces

Tajima, Hiroyuki*; Oue, Daigo*; Matsuo, Mamoru

Physical Review A, 106(3), p.033310_1 - 033310_8, 2022/09

https://doi.org/10.1103/PhysRevA.106.033310

Motion-induced spin transfer

Oue, Daigo*; Matsuo, Mamoru

Physical Review B, 105(2), p.L020302_1 - L020302_5, 2022/01

https://doi.org/10.1103/PhysRevB.105.L020302

Conceptual design of the SlimCS fusion DEMO reactor

Tobita, Kenji; Nishio, Satoshi*; Enoeda, Mikio; Nakamura, Hirofumi; Hayashi, Takumi; Asakura, Nobuyuki; Uto, Hiroyasu; Tanigawa, Hiroyasu; Nishitani, Takeo; Isono, Takaaki; et al.

JAEA-Research 2010-019, 194 Pages, 2010/08

JAEA-Research-2010-019-01.pdf:48.47MB
JAEA-Research-2010-019-02.pdf:19.4MB

This report describes the results of the conceptual design study of the SlimCS fusion DEMO reactor aiming at demonstrating fusion power production in a plant scale and allowing to assess the economic prospects of a fusion power plant. The design study has focused on a compact and low aspect ratio tokamak reactor concept with a reduced-sized central solenoid, which is novel compared with previous tokamak reactor concept such as SSTR (Steady State Tokamak Reactor). The reactor has the main parameters of a major radius of 5.5 m, aspect ratio of 2.6, elongation of 2.0, normalized beta of 4.3, fusion out put of 2.95 GW and average neutron wall load of 3 MW/m. This report covers various aspects of design study including systemic design, physics design, torus configuration, blanket, superconducting magnet, maintenance and building, which were carried out increase the engineering feasibility of the concept.

An Experimental validation of the guideline for inelastic design analysis through structural model tests

Watanabe, Daigo*; Chuman, Yasuharu*; Otani, Tomomi*; Shibamoto, Hiroshi*; Inoue, Kazuhiko*; Kasahara, Naoto

Nuclear Engineering and Design, 238(2), p.389 - 398, 2008/02

https://doi.org/10.1016/j.nucengdes.2006.09.021

In this paper, the inelastic analysis procedures for the improved design of future fast breeder reactors were validated through the structural model tests and the evaluation of the experimental results by the inelastic analyses. First, a thermal fatigue test of a 316FR hollow cylinder with two longitudinal weldments was conducted under the condition of combined constant axial load and cyclic movement of axial temperature distribution, which simulated the loading condition near the free surface of coolant sodium in the main vessel of fast breeder reactors (FBRs). Second, the inelastic analyses were carried out in accordance with the recommended procedure by using the measured results of oscillating temperature distribution. Finally, the results of inelastic analyses were compared with the experimental results and it was validated that the recommended practice gave a conservative result for the deformation and a good estimation of strain range for the fatigue life evaluation.

Concept of compact low aspect ratio Demo reactor, SlimCS

Tobita, Kenji; Nishio, Satoshi; Sato, Masayasu; Sakurai, Shinji; Hayashi, Takao; Shibama, Yusuke; Isono, Takaaki; Enoeda, Mikio; Nakamura, Hirofumi; Sato, Satoshi; et al.

Proceedings of 21st IAEA Fusion Energy Conference (FEC 2006) (CD-ROM), 8 Pages, 2006/10

http://www-naweb.iaea.org/napc/physics/FEC/FEC2006/html/node305.htm#60147

no abstracts in English

Measurement of thermal ratcheting strain on the structures by the laser speckle method

Watanabe, Daigo*; Chuman, Yasuharu*; Otani, Tomomi*; Shibamoto, Hiroshi; Inoue, Kazuhiko*; Kasahara, Naoto

Proceedings of 2006 ASME Pressure Vessels and Piping Division Conference (PVP 2006)/International Council on Pressure Vessel Technology (ICPVT-11) (CD-ROM), 7 Pages, 2006/00

Prevention of thermal ratcheting is an important problem for high temperature components of fast breeder reactors that are subjected to cyclic thermal loads. To clarify ratcheting behaviors, structural model tests were planned. Strain measurement is important for understanding the thermal ratcheting phenomenon, however the conventional measurement by strain gauge is difficult at high temperature. Then, Laser speckle strain measurement system using the dual-beam set-up was developed to apply to high temperature structural model tests. This system was applied to the thermal ratcheting tests, which demonstrated the actual operative conditions of reactor vessels. Through comparison with uniaxial test results obtained by extensometers, the laser speckle method was verified. Measured data of structural model tests were utilized to certify the guidelines of inelastic analysis for design, which provide prediction method of strain in components of fast reactor.