Design study of National Centralized Tokamak facility for the demonstration of steady state high-β plasma operation

Tamai, Hiroshi; Akiba, Masato; Azechi, H.; Fujita, T.; Hamamatsu, K.; Hashizume, Hidetoshi; Hayashi, Naoki; Horiike, Hiroshi; Hosogane, N.; Ichimura, M.; Ida, K.; Imai, T.; Ishida, S.; Itoh, Susumu; Kamada, Y.; Kawashima, H.; Kikuchi, M.; Kimura, Akihiko; Kizu, K.; Kubo, H.; Kudo, Yusuke; Kurihara, Kenichi; Kurita, G.; Kuriyama, M.; Masaki, K.; Matsukawa, M.; Matsuoka, M.; Miura, Y.; Miya, N.; Morioka, A.; Nakamura, K.; Ninomiya, H.; Nishimura, A.; Okano, Kunihiko; Okuno, K.; Sagara, A.; Sakamoto, M.; Sakurai, S.; Sato, Kohnosuke; Shimada, Ryuichi; Shimizu, Akihiko; Suzuki, Takahiro; Takahashi, H.; Takase, Y.; Takechi, M.; Tanaka, Satoru; Tsuchiya, Kazuyoshi; Tsutsui, Hiroaki; Uesugi, Yoshihiko; Yatsu, K.; Yoshida, N.

doi:10.1088/0029-5515/45/12/023

Cited by 16 publications

(8 citation statements)

References 23 publications

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“…The up-to-date TF coil design is globally derived from the previous version issued from the NCT project [5], modified according to optimization considerations on performances, manufacture feasibility and costs. It consists of 18 D-shaped magnets, constituting a self-supporting system.…”

Section: Tf Coil and Conductor Designmentioning

confidence: 99%

JT-60SA Toroidal Field Magnet System

Pizzuto

Semeraro

Zani

et al. 2008

IEEE Trans. Appl. Supercond.

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The Broader Approach agreement between Europe and Japan includes the construction of a fully superconducting tokamak, the JT-60 Super Advanced (JT-60SA), as a satellite experiment to ITER. In particular, the whole Toroidal Field magnet system, described in this paper, will be provided to Japan by the EU. All the TF coil main constituents, i.e. conductor, winding pack, joints, casing, current leads, are here presented and discussed as well as the design criteria adopted to fulfil the machine requirements. The results of the analyses performed by the EU and JA to define and assess the TF magnet system conceptual design are reported and commented. Future work plan is also discussed.

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Section: Tf Coil and Conductor Designmentioning

confidence: 99%

JT-60SA Toroidal Field Magnet System

Pizzuto

Semeraro

Zani

et al. 2008

IEEE Trans. Appl. Supercond.

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“…Experimental data in this region is scarce as well. National centralized tokamak (NCT) [12] planned in Japan is designed to aim at high β N (= 3.5-5.5) at low-A (2.6-3.1) and make the proper stepping for "slim CS" experimentally. Taking advantage of low-A on β N , the required β N for "slim CS" is rather conservative compared to the existing designs.…”

Section: Main Option "Slim Cs"mentioning

confidence: 99%

Design study of fusion DEMO plant at JAERI

Tobita¹,

Nishio²,

Enoeda³

et al. 2006

Fusion Engineering and Design

139

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“…addressing key physics issues for ITER and DEMO. JT-60SA is a tokamak with superconducting magnets that will be built in the present JT-60U torus hall [1,2]. Many components of the present JT-60U facility including heating and current drive (CD) systems and some power supplies will be utilized.…”

Section: Introductionmentioning

confidence: 99%

“…The plasma major radius will be about 3 m and the plasma current will be up to 5.5 MA, and hence a high-performance deuterium plasma is expected with confinement close to the equivalent break-even condition. The device is characterized by enhanced flexibility in the aspect ratio (A) and the plasma shape [1,2]. The heating and CD system has been upgraded to make a total power of 41 MW available for 100 s [3,4], compared with 15 MW for 100 s (25 MW for 50 s) in the previous design.…”

Section: Introductionmentioning

confidence: 99%

Design optimization for plasma performance and assessment of operation regimes in JT-60SA

et al. 2007

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The design of the modification of JT-60U, JT-60SA has been optimized from the viewpoint of plasma performance, and operation regimes have been evaluated with the latest design. Upper and lower divertors with different geometries will be prepared for flexibility of the plasma shape, which will enable both low aspect ratio (A ∼ 2.65) and ITER shape (A = 3.1) configurations. The beam lines of negative-ion neutral beam injection will be shifted downwards by ∼0.6 m for the off-axis current drive (CD), in order to obtain a weak/reversed shear plasma, as well as having the capability of heating the central region. The feedback control coils along the openings in the stabilizing plate are found effective in suppressing the resistive wall mode and sustaining high βN close to the ideal wall limit. Sustainment of plasma current of 3–3.5 MA for 100 s will be possible in ELMy H-mode plasmas with moderate heating power, βN, and density within an available flux swing. It is also expected that higher βN, high-density ELMy H-mode plasmas will be maintained for 100 s with higher heating power. The expected regime of full CD operation has been extended with upgraded heating and CD power. Full CD operation for 100 s with reactor-relevant high values of normalized beta and bootstrap current fraction (Ip = 2.4 MA, βN = 4.3, fBS = 0.69, , HH98y2 = 1.3) is expected in a highly-shaped low-aspect-ratio configuration (A = 2.65).

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Design study of National Centralized Tokamak facility for the demonstration of steady state high-β plasma operation

Cited by 16 publications

References 23 publications

JT-60SA Toroidal Field Magnet System

JT-60SA Toroidal Field Magnet System

Design study of fusion DEMO plant at JAERI

Design optimization for plasma performance and assessment of operation regimes in JT-60SA

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