The question of heating a tokamak plasma by means of electromagnetic waves in the ion cyclotron range of frequencies (ICRF) is considered in the perspective of large rf powers and in the low collisionality regime. In such a case, the quasilinear theory (QLT) is validated by the Hamiltonian dynamics of the wave–particle interaction which exceeds the threshold of the intrinsic stochasticity. The Hamiltonian dynamics is represented by the evolution of a set of three canonical action angle variables well adapted to the tokamak magnetic configuration. This approach allows derivation of the rf diffusion coefficient with very few assumptions. The distribution function of the resonant ions is written as a Fokker–Planck equation but the emphasis is put on the QL diffusion instead of on the usual diffusion induced by collisions. The Fokker–Planck equation is then given a variational form from which a solution is derived in the form of a semianalytical trial function of three parameters: the percentage of resonant particles contained in the tail, an isotropic width ΔT, and an anisotropic width ΔP. This solution is successfully tested against real experimental observations. It is shown that in the case of the JET tokamak [Plasma Phys. Controlled Fusion 30, 1467 (1988)] the distribution function is influenced by adiabatic barriers which in turn limit the Hamiltonian stochasticity domain within energy values typically in the MeV range. Consequently and for a given ICRF power, the tail energy excursion is lower and its concentration higher than that from a bounce-averaged prediction. This may actually be an advantage for machines like JET [Plasma Phys. Controlled Fusion 30, 1467 (1988)] considering the energy range required to simulate the α-particle behavior in a relevant fusion reactor.

The electromagnetic perturbation produced through a set of antennas in a tokamak plasma is studied near the ion cyclotron resonances and the two-ion hybrid resonance. Using an appropriate variational principle, it is possible to derive ' the perturbed field (thecompressional wave and the converted torsional wave) with the help of relatively smooth test functions of the compressional type. The method produces conditions under which the variational functional L, an extremum with respect to the perturbation, can be written in a suitable manner for computer handling. Moreover, this functional provides an expression for the power damped by the thermal motion of particles, and in the two-ion hybrid regime it provides the power coupled to the converted wave. The ion cyclotron resonance leads to different regimes for the ratio of the parallel scalelength, X, of the field and the intrinsic parallel scale of the resonance, X £ , with X £ % l v |/V||W c l 1/2 . The usual WKB formulas used by previous authors appear to be limited to the case X < X c only, and new formulas are given for X > X . In the mode conversion case, the method generalizes Budden's results, with restrictions taking into account the non-uniformity effects along the resonance surface.

The power deposition profile in the ion cyclotron range of frequencies (ICRF) has been investigated experimentally in JET by means of a square wave modulated RF perturbation. The study has been conducted in D(H) and D(3He) plasmas for two heating scenarios. In D(3He) plasmas and for central heating in a scenario where mode conversion to Bernstein waves is accessible, the direct power deposition profile on electrons has been derived. It accounts for 15% of the total coupled power and extends over 25% of the minor radius. Outside the RF power deposition zone, the electron thermal diffusivity χe inside the inversion radius surface (ri) can be estimated through observation of the diffusive electronic transport. In discharges without monster sawteeth and for a low central temperature gradient (∇Te(r ≤ ri) ≤ ∇Te(r ≥ ri) ≈ 5 keV·m−1) the value obtained is small (≈ 0.24 ± 0.05 m2·s−1), typically ten times ower than χe values deduced from heat pulse propagation in similar discharges at radii larger than the inversion radius. For the D(H) minority heating scheme, a large fraction of the ICRF modulated power is absorbed by minority ions, and the minority tail is modulated with a characteristic ion-electron (i–e) slowing-down time. In this scheme, electron heating occurs only through collisions with the minority ion tail and no modulation of the electron temperature is observed in sawtoothing discharges. This is interpreted as a consequence of the long i–e equipartition time, acting as an integrator for the modulated ICRF signal. Finally, a correlation between the time of the sawtooth crash and the periodic turn-off of the ICRF power is found and its consequence for modulation experiments is reviewed.

Near breakeven conditions have been attained in the JET tokamak [Fusion Technol. 11, 13 ( 1987) 1, with beryllium as the first-wall material. A fusion triple product (n,r, Ti ) of 8-9 x 10" m -3 set keV has been reached (within a factor of 8 of that required in a fusion reactor). However, this has only been achieved transiently. At high heating powers, an influx of impurities still limits the achievement of better performance and steady-state operation. In parallel, an improved quantitative understanding of fusion plasmas has emerged from the development of a particular plasma model. Good quantitative agreement is obtained between the model and JET data. The main predictions are also consistent with statistical scaling laws. With such a model, a predictive capability begins to emerge to define the parameters and operating conditions of a DEMO, including impurity levels. Present experimental results and model predictions suggest that impurity dilution is a major threat to a reactor. A divertor concept must be developed further to ensure impurity control before a DEMO can be constructed. A New Phase for JET is planned in which an axisymmetric pumped divertor configuration will be used to address the problems of impurity control, plasma fueling, and helium ash exhaust. It should demonstrate a concept of impurity control and the operational domain for such a device. A single Next Step facility (ITER) is a high risk strategy in terms of physics, technology, and management, since it does not provide a sufficiently sound foundation for a DEMO. A Next Step program is proposed, which could comprise several complementary facilities, each optimized with respect to specific clear objectives. In a minimum program, there could be two Next Step tokamaks, and a Materials Test Facility. Such a program would allow division of effort and sharing of risk across the various scientific and technical problems, permit cross comparison, and ensure continuity of results. rt could even be accomplished without a significant increase in world funding.