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Shirai, Hiroshi; Barabaschi, P.*; Kamada, Yutaka; JT-60SA Team
Fusion Engineering and Design, 109-111(Part B), p.1701 - 1708, 2016/11
Times Cited Count：21 Percentile：88.83(Nuclear Science & Technology)The JT-60SA Project has shown steady progress toward the first plasma in 2019. JT-60SA is a superconducting tokamak designed to operate in the break-even conditions for a long pulse duration with a maximum plasma current of 5.5 MA. Design and fabrication of JT-60SA components shared by EU and Japan started in 2007. Assembly in the torus hall started in January 2013, and welding work of the vacuum vessel sectors is currently on going on the cryostat base. Other components such as TF coils, PF coils, power supplies, cryogenic system, cryostat vessel, thermal shields and so forth were or are being delivered to Naka site for installation, assembly and commissioning. This paper gives technical progress on fabrication, installation and assembly of tokamak components and ancillary systems, as well as progress of JT-60SA Research Plan being developed jointly by EU and Japanese fusion communities.
Tobita, Kenji
no journal, ,
Status of DEMO design conducted under the Broader Approach Activity is reported highlighting on Japanese design activity. Remote maintenance is one of the most critical design issues in DEMO in that a reasonable plant availability needs to be attained in severe radiation environment. Recent design study revealed that the application of "sector maintenance" scheme to a medium size DEMO with a major radius of about 8 m would lead to increase in the size of toroidal field coils, the required current of poloidal field coils and the sector weight. For this reason, "banana-type segment" scheme is under study instead of the sector scheme. In order to resolve diverter heat removal problem, use of copper alloy pipes in high heat flux and low dpa (displacements per atom) zones is under study. Besides, a possible management scenario of radioactive waste produced in periodic maintenance and the resulting waste-related facilities for DEMO are presented.
Higashijima, Satoru; JT-60SA Team
no journal, ,
no abstracts in English
Takase, Haruhiko; Tobita, Kenji; Sakamoto, Yoshiteru; Uto, Hiroyasu; Mori, Kazuo; Kudo, Tatsuya
no journal, ,
Analysis of plasma position control is one of important issues for design of DEMO reactor on Broader Approach (BA). Especially, plasma performance, blanket design and maintenance scheme influence the plasma position control mutually. Therefore, we made a numerical simulation code that consists of plasma equilibrium analysis, eddy current analysis and plasma motion analysis. Since we analyzed several cases of design using this numerical simulation code, the results will be shown.
Rivas, J. C.*; Nakamura, Makoto; Someya, Yoji; Takase, Haruhiko; Tobita, Kenji; de Blas, A.*; Dies, J.*; Fabbri, M.*; Riego, A.*
no journal, ,
Safety studies of plasma-wall transients have been performed with AINA code for the Japanese DEMO design (water cooled pebble bed). The AINA code has been adapted from its original mission of performing safety studies for ITER to this new mission. A breeding blanket model has been implemented in code. The configuration has been changed to implement the design parameters of DEMO reactor. First analyses performed show the behavior of the reactor during ex-vessel LOCA transients and during overpower events.
Ohira, Shigeru
no journal, ,
Under the Broader Approach (BA) activities, which are research and development activities jointly implemented by Japan and EURATOM in support of the ITER Project and an early realization of fusion energy for peaceful purpose, projects of the Engineering Validation and Engineering Design Activities for International Fusion Materials Irradiation Facility (IFMIF/EVEDA) and the International Fusion Energy Research Centre (IFERC) are being carried out in Rokkasho, Aomori, Japan. In this presentation, Status of research & development for DEMO under the Broader Approach Activities are introduced.
Kasugai, Atsushi; ILIC Unit*
no journal, ,
The prototype accelerator is being carried out for an engineering validation of the International Fusion Materials Irradiation Facility (IFMIF) as accelerator-driven-type neutron source for developing fusion reactor materials. This is a deuteron linear accelerator consisting of an injector, an RFQ, a superconducting RF linac, RF power systems, a beam dump and beam transport lines. The specification of the target is to produce a CW deuteron beam with the beam energy and current of 9 MeV/ 125 mA. From the spring of 2014, full installation of the injector was started at Rokkasho site and the injector beam test has been just began from November 2014 in order to obtain better beam qualities for successful injection and acceleration of the following accelerators such as RFQ.
Sakurai, Shinji
no journal, ,
The Broader Approach (BA) activities have been developed between EU and Japan to support ITER. JT-60SA is a joint project between the BA satellite tokamak and Japanese national program. The project mission is the followings, (1) Support for ITER using break-even-class high performance plasmas with long pulse, (2) Supplement of ITER toward DEMO with long sustainment of high pressure plasmas for establishment of DEMO operating scenarios. The JT-60SA Research Plan has been discussed widely in EU and Japan fusion research communities and completed with more than 300 co-authors. Construction of JT-60SA is in progress toward its first plasma in 2019. Procurements of components by EU and Japan are on schedule and assembly in Naka site has been started. Outline of the project and progress of manufacturing and assembly will be presented.
Takenaga, Hidenobu
no journal, ,
The broader approach (BA) activities in the field of fusion energy research have been progressed in collaboration between Japan and EU. The JAEA plan after the present BA activities which will be completed on May 2017 was reported. In order to contribute to demonstration of feasibility of fusion energy from scientific and engineering viewpoints and establishment of technology base necessary for assessment of transition condition to DEMO reactor phase, JAEA promotes three main activities, which are ITER project, advanced plasma research utilizing JT-60SA, and science and technology research utilizing facilities constructed in BA activities, with strong corporation and personnel exchange.
Rivas, J. C.*; Nakamura, Makoto; Someya, Yoji; Takase, Haruhiko; Tobita, Kenji; Dies, J.*; Blas, A. de*; Fabbri, M.*; Riego, A.*
no journal, ,
In the frame of JAPAN-EU collaborative work for development of AINA code in 2014-2016, a version of AINA code has been developed for the Japanese DEMO WCPB design. During 2014, the AINA code was adapted from ITER to this new mission. A breeding blanket model was implemented in code. The configuration was changed to implement the design parameters of DEMO reactor. Finally, safety studies of plasma-wall transients affecting blanket region were performed. During 2015, plasma models were improved both for plasma core and for divertor (improved SOL model). Safety analyses affecting divertor were performed, considering thermohydraulic accidents and plasma transients where loss of control function was assumed. First analyses performed for the Japanese DEMO design show the behavior of the reactor during Ex-Vessel LOCA and during overpower events. The preliminary conclusions point to the possibility of considering the plasma control system as a safety important component.
Takase, Haruhiko; Uto, Hiroyasu; Sakamoto, Yoshiteru; Mori, Kazuo; Kudo, Tatsuya; Tobita, Kenji
no journal, ,
Pre-conceptual design of DEMO reactor has been preceded under collaboration Japan and Europe (Broader Approach activities (BA)). In the case of DEMO reactor, it is important for design of plasma position control to take into account the actual shape of vacuum vessel and in-vessel components precisely since the design condition of DEMO reactor is different from current tokamak devices and ITER (for example, installation of maintenance ports). To consider the DEMO design condition, the numerical simulation that consists of three modules has been developed. As results, (1) The time constants of eddy current in the breeding blankets are less than 10ms and there is no influence to passive stabilization effect. (2) The stabilization effect of conducting components decreases by considering installation of vertical maintenance port. The adoption of vertical port is related to the choice of the maintenance scenario.
