M. Kikuchi

Steady state tokamak reactor based on the bootstrap current

A concept of a steady state tokamak fusion reactor based on the bootstrap current is presented. Operation at high poloidal beta (?p ? 2.0) and high q (4?5) with a relatively small limit on ??p (< 0.5) makes it possible to drive a bootstrap current constituting up to 70% of the total plasma current without exceeding the Troyon beta limit. The rest of the plasma current can be driven by the high energy neutral beam with an energy multiplication factor Q of 30. Energy confinement scaling laws predict that the reactor condition is attainable by increasing the major radius up to 9 m in such a high ?p and high q plasma at a relatively low plasma current (12 MA) with a confinement enhancement factor of 2 compared with the L-mode scaling. This reactor has a reasonable size (Vp = 1500 m3) and fusion output power (2.5 GW) and is consistent with present knowledge regarding tokamak plasma physics, namely the Troyon limit, the energy confinement scalings, the bootstrap current and the current drive efficiency (neutral beam current drive with a total power of 70 MW and a beam energy of 1 MeV) with good prospects for the formation of a cold and dense divertor plasma.

Bootstrap current during perpendicular neutral injection in JT-60

The neoclassical bootstrap current effect is investigated in the JT-60 tokamak. The experimental resistive loop voltages are compared with the calculations, using the neoclassical resistivity, with and without the bootstrap current, and the Spitzer resistivity for a wide range of plasma current (Ip = 0.5-2 MA) and poloidal beta (βp = 0.1-3.2). The neoclassical bootstrap current is calculated directly with the force balance equations for viscous and friction forces according to the Hirshman–Sigmar theory. The bootstrap current driven by the fast ion component is also included. The calculated resistive loop voltage is consistent with the neoclassical prediction including the bootstrap current. It is shown that up to 80% of total plasma current is driven by the bootstrap current in the regime with an extremely high poloidal beta value (βp = 3.2) while the beam driven current is negligibly small.

Ripple induced fast ion loss and related effects in JT-60U

Experiments have been carried out in JT-60U to verify the modelling of fast ion ripple transport. The ripple induced loss was estimated from the neutron decay following neutral beam pulse injection and the loss related heat load on the first wall. Comparison of the lost fraction and the hot spot positions between measurements and orbit following Monte Carlo calculations exhibited good agreement, indicating that the ripple transport governing fast ion losses is explained within the framework of existing theory. Neutral beam heating experiments in JT-60U also indicate that H modes free of ELMs are still obtainable for ripple amplitudes of up to 2.2%

Reduced transport and E r shearing in improved confinement regimes in JT-60U

The global confinement and the local transport properties of improved core confinement plasmas in JT-60U were studied in connection with Er shear formation. In the improved core confinement mode with internal transport barriers (ITBs), these are roughly classified into `parabolic type ITBs and `box type ITBs. The parabolic type ITB has a reduced thermal diffusivity χ in the core region; however, the Er shear, dEr/dr, is not as strong. The box type ITB has a very strong Er shear at the thin ITB layer and χ decreases to the level of neoclassical transport there. The estimated E × B shearing rate, ωE × B, becomes almost the same as the linear growth rate of the drift microinstability, γL, at the ITB layer in the box type ITB. Experiments with hot ion mode plasmas during the repetitive L-H-L transition showed that the thermal diffusivity clearly depends on the Er shear and the strong Er shear contributes to the reduced thermal diffusivity.

Internal transport barrier with improved confinement in the JT-60U tokamak

Characteristics of the internal transport barrier (ITB) were studied. The region of steep and , i.e. the ITB front, propagated from the core outwards. The thickness of the ITB front was about 3 cm. The ITB worked as a particle transport barrier as well as a thermal transport barrier for ions. The threshold heating power for ITB formation strongly increased with electron density and was independent of the toroidal magnetic field. ITB with was sustained for twice the global energy confinement time . A repetitive relaxation phenomenon at ITB was observed, which induced spikes like ELMs but had a different poloidal distribution.

Blanket and divertor design for the Steady State Tokamak Reactor (SSTR)

Abstract The tritium breeding blanket and plasma facing components (first wall and divertor) have been designed for the Steady State Tokamak Reactor (SSTR). Low activation ferritic steel (F82H) was chosen for the structural material of the first wall and the blanket. The ceramic breeder pebble bed concept with beryllium neutron multiplier was adopted. The blanket is divided into two parts; a replaceable blanket and a permanent blanket. Electrical insulation made of functionally gradient material (FGM) was introduced in the blanket box structure to drastically reduce electromagnetic loads during a plasma disruption and to simplify the attaching mechanism of the blanket. Low-temperature and high-density divertor plasma conditions and a radiative cooling concept lower the heat flux to the divertor plate to below 3 MW/m2 and enable us to adopt the same cooling conditions as the blanket coolant. Puffing mechanisms of gas (deuterium) and iron are installed for the radiative cooling of the divertor plasma.

Recent directions in plasma physics and its impact on tokamak magnetic fusion design

Abstract Steady progress has been made in recent years towards achieving fusion breakeven and reactor-relevant plasma conditions in the largest tokamak experiments. In particular, the experimental observation of a large self-sustaining bootstrap current in the plasma permits development of steady-state reactor concepts with modest current drive power and relatively high plasma energy gain ( Q ). The SSTR (Japan) and ARIES-I (USA) designs reported in this paper are first-stability, steady-state tokamak reactors based on “modest” extrapolations from the present tokamak physics database. The plasma is characterized by a high edge safety factor ( q a ) and a high poloidal beta ( β p ) in order to increase bootstrap current fraction (∼70%). This mode of operation is achieved by selecting high values of both aspect ratio ( A = R / a = 4−4.5) and toroidal magnetic field on axis (9–11 T) in these reactors. Both reactor studies suggest that the tokamak system can be a steady-state power reactor with net electricity of ∼1 GW and with plant efficiency 30–40%. The SSTR is characterized by its technical feasibility in the near future. On the other hand, the ARIES-I focuses on better safety and environmental aspects and a longer time frame. The SSTR and ARIES-I studies show that, with proper R&D programs, tokamak fusion reactors can be developed that will have acceptable cost of electricity.

Objectives and design of the JT-60 superconducting tokamak

A fully superconducting tokamak named JT-60SC is designed for the modification programme of JT-60 to enhance economical and environmental attractiveness in tokamak fusion reactors. JT-60SC aims at realizing high-β steady-state operation in the use of low radio-activation ferritic steel in a low ν* and ρ* regime relevant to the reactor plasmas. Objectives, research issues, plasma control schemes and a conceptual design for JT-60SC are presented.

Non-inductively current driven H mode with high beta N and high beta p values in JT-60U

Attainable βp and βN values were widely improved in JT-60U by peaking of the current profile and broadening of the pressure profile. In a quasi-steady state of ELMyy H mode, βp approximately 2.5-3, βN approximately 2.5-3.1 and H factor approximately 1.8-2.2 were sustained for approximately 1 s simultaneously under the full current drive condition (bootstrap current 60%, beam driven current 48% at Ip = 0.5 MA). In a transient case, the maximum values of βp, βp and βN reached 4.7, 1.2 and 4.2, respectively. The achievable βp value increases systematically with safety factor at the edge. To obtain these improved high β plasmas, it is essential to control MHD activities; suppression of βp collapse and medium m/n (poloidal mode number/toroidal mode number) modes by the modification in current and pressure profiles and control of ELM activity by suitable selection of the heating power

Advanced tokamak research on JT-60

The Japanese large tokamak JT-60 has been focusing its research emphasis on the development of high performance plasmas, with high confinement, high temperature and high density and non-inductive sustainment for a long time with possible minimization of external power input. The first demonstration of high bootstrap current discharges up to 80% of the total plasma current in a high-poloidal-beta (high-βp) mode and the concept development of a steady-state tokamak reactor (SSTR) based on this experimental achievement initiated the so-called advanced tokamak research. The internal transport barriers (ITBs), discovered in the JT-60 high-βp mode, have been followed by worldwide explorations of reversed shear discharges associated with ITBs. The highest DT equivalent energy gain of was achieved in the JT-60 reversed shear H-mode discharges. The highest ion temperature of Ti = 45u2009keV and the highest fusion triple product of ni(0)τETi(0) = 1.5 × 1021u2009m−3u2009su2009keV were obtained in high-βp discharges. Advanced tokamak research is now the major trend of the current tokamak development. A new concept of compact ITER has been developed and proposed in the context of this advanced tokamak approach pursued on JT-60. Prospects for burning plasma physics have been investigated along the progress made in these modern tokamak experiments on JT-60 and related computer simulation analyses.