Plasma design considerations of near term tokamak fusion experimental reactor.

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概要

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• Wide range of parameter surveys are made on the DT fusion tokamak experimental reactor next to JT-60. Various physics and engineering requirements are taken into account, <I>e</I>.<I>g</I>. self-ignition, available maximum toroidal β value, α-particle confinement, total fusion power, neutron wall loading, heat flux to divertor plate, structural restriction on major radius, device size, maximum toroidal magnetic field, poloidal field power supply and so on. Theoretical scaling law for the available maximum toroidal β value deter-mined by ballooning mode instability is used. The toroidal magnetic field on plasma axis can be expressed by the aspect ratio A for a given maximum field at the toroidal field coil conductor. Empirical scaling law for the electron energy confinement and neoclassical heat conductivity for the ion are employed. These confinement times can be expressed by the plasma minor radius a and A through the maximum available Β value and the toroidal field on axis. In the similar way, most of the physics and engineering requirements can be mapped on the α-A diagram. This diagram enables us to make systematic and wide range of parameter surveys of the device. In particular, this offers a clear perspective on the device parameters, which can mitigate the engineering difficulties and can also realize the required plasma performances.

• 一般社団法人 日本原子力学会の論文

著者

• Sugihara Masayoshi

  Division Of Large Tokamak Development Japan Atomic Energy Research Institute

• SHIMADA Ryuichi

  Division of Large Tokamak Development, Japan Atomic Energy Research Institute

• FUJISAWA Nobol

  Division of Large Tokamak Development, Japan Atomic Energy Research Institute

• UEDA Kojyu

  Division of Large Tokamak Development, Japan Atomic Energy Research Institute

• SAITO Seiji

  Division of Large Tokamak Development, Japan Atomic Energy Research Institute

• HATAYAMA Akiyoshi

  Division of Large Tokamak Development, Japan Atomic Energy Research Institute

関連論文

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• Assessment of feasibility of helium ash exhaust and heat removal by pumped-limiter in tokamak fusion reactor.
• Ignition approach by neutral beam injection heating in impurity contaminated tokamak reactors.

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