Design Optimization Study of the JT-60 Superconducting Coil Modification

Matsukawa, M.

doi:10.1109/tasc.2004.830621

Cited by 3 publications

(3 citation statements)

References 8 publications

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“…The radial cross section of TFC in the midplane is illustrated in figure 1. In the first design, NCT-1, the inboard and the outboard side of the TFC are moved outwards by 5 cm, and by 50 cm, respectively [11]. By this extension the minor radius increases up to a p ∼ 1.1 m, and the aspect ratio of A ∼ 2.8 is enabled.…”

Section: Modification Of Tfc Designmentioning

confidence: 99%

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

Tamai¹,

Akiba²,

Azechi³

et al. 2005

Nucl. Fusion

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This is a preprint of a paper intended for presentation at a scientific meeting. Because of the provisional nature of its content and since changes of substance or detail may have to be made before publication, the preprint is made available on the understanding that it will not be cited in the literature or in any way be reproduced in its present form. The views expressed and the statements made remain the responsibility of the named author(s); the views do not necessarily reflect those of the government of the designating Member State(s) or of the designating organization(s). In particular, neither the IAEA nor any other organization or body sponsoring this meeting can be held responsible for any material reproduced in this preprint.

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Section: Modification Of Tfc Designmentioning

confidence: 99%

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

Tamai¹,

Akiba²,

Azechi³

et al. 2005

Nucl. Fusion

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“…T HE NATIONAL centralized tokamak (NCT), a new name of the former JT-60SC [1], and is aiming at contributing to DEMO as well as the ITER. In the DEMO reactor design, high-beta steady-state operation is a key to realize the nuclear fusion power plant economically competitive to other energy sources, because the fusion power is approximately proportional to the square of normalized plasma beta value and the cost of electricity is determined dominantly by the capital investment of the machine construction.…”

Section: Introductionmentioning

confidence: 99%

Poloidal Field Coil Configuration and Plasma Shaping Capability in NCT

Matsukawa

Tamai

Fujita

et al. 2006

IEEE Trans. Appl. Supercond.

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This paper describes the latest design of poloidal field coil configuration in the national centralized tokamak (NCT), that is a new design based on the former superconducting coil tokamak JT-60SC. The most notable design change was made for the outer equilibrium field (EF) coils to increase a plasma shaping parameter S(= q 95 I p a B t ) and to make off-axis heating using N-NBI possible. As a result, the maximum plasma shaping parameter of 7 was obtained by double null divertor configuration with a lower aspect ratio of A 2.6. It was confirmed that two additional EF coils are useful for the plasma squareness control, but only one is finally adopted for ITER plasma simulation and to keep compatibility of off-axis N-NBI heating.

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“…From 2004, the design of is changed in order to keep the broad operational space for shape parameter and aspect ratio of plasmas through the nation-wide collaboration in Japan, so that this devise is named National Centralized Tokamak (NCT) [3]- [5]. NCT features the superconductor to the coil systems that produce the field for the confinement and the configuration control of plasmas.…”

Section: Introductionmentioning

confidence: 99%

Stress Analyses of the Support Structure and Winding Pack of the Superconducting TF Coil in National Centralized Tokamak

Tsuchiya¹,

Kizu²,

Takahashi

et al. 2006

IEEE Trans. Appl. Supercond.

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The design activity of the support structure of Toroidal Field (TF) coil is progressed for the National Centralized Tokamak (NCT). For this device, the new support structures of the TF coil is proposed. In the new design, connection keys between coil case and shear panel are adopted in order to decrease the movement of shear panel from coil case which induces the displacement of TF coil during coil operation. For the stress analysis of new structure, FEM model with full-solid elements are developed to evaluate the behavior around the connection keys. The electromagnetic force induces strain and stress on the winding pack which affect the performance of superconductor such as Nb 3 Al that is adopted for TF coil in NCT. Therefore, longitudinal strain and transverse stress on the winding pack during coil operation are inspected. For the new design of TF coil, the maximum displacement is 6.4 mm around the upper shear panel and maximum Tresca stress with 415 MPa appears on the shear panel at the midplane. This value is within the allowable stress of SS316LN at 4 K, 547 MPa. For the winding pack, longitudinal strain is evaluated with 0.03 to +0.07%, and transverse stress is evaluated with 40 MPa at the place with the maximum field.These values are too small to affect the critical current of Nb 3 Al conductor. Therefore, it is confirmed that the stable operation of superconductor is insured on the new structure of TF coil.

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Design Optimization Study of the JT-60 Superconducting Coil Modification

Cited by 3 publications

References 8 publications

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

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

Poloidal Field Coil Configuration and Plasma Shaping Capability in NCT

Stress Analyses of the Support Structure and Winding Pack of the Superconducting TF Coil in National Centralized Tokamak

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