Table of contents

The physics basis of a simulation model for lower hybrid current drive (LHCD) is discussed. Issues associated with LH power deposition-wave propagation, mode conversion and cut-offs in toroidal geometry, as well as linear and quasi-linear Landau damping-are analysed. A simulation model (ACCOME) is applied to the LHCD experiment now operating on the Tokamak de Varennes (TdeV). The profiles and values of density and temperature needed as inputs to ACCOME are taken from the experiment. The predictions of current density and loop voltage from ACCOME are then compared with experimental LHCD results. Possible LH current profile control experiments are also analysed for TdeV using composite LH spectra to control the location of RF power deposition. Finally the relevance of these current profile control results to future devices is discussed and an example is shown for ITER-like parameters

A regime of improved H mode energy confinement, the VH mode, was obtained following boronization of the DIII-D Tokamak vacuum vessel. The gradual confinement improvement in VH mode is associated with expansion of the H mode pedestal and reduction of thermal diffusivity in a region just inside the pedestal. Disappearance of density fluctuation bursts is associated with a confinement improvement in the latter region. The VH mode is terminated in a rapid global energy loss which is initiated by a mode with toroidal mode number, n approximately 5, consistent with calculated instability to an edge localized ideal kink. The improved confinement in VH mode is consistent with the extension of the region of the E*B velocity shear turbulence suppression zone further in from the plasma boundary. Expansion of the ideal ballooning second stability region and of the region with drift reversal stabilization of trapped particle modes are also considered as possible explanations for the confinement improvement

The confinement in the bulk and scrape-off layer plasmas of the ITER EDA and CDA is investigated with special versions of the 1.5-D BALDUR predictive transport code for the case of peaked density profiles (C_u=1.0). The code self-consistently computes 2-D equilibria and solves 1-D transport equations with empirical transport coefficients for the ohmic, L and ELMy H mode regimes. Self-sustained steady state thermonuclear burn is demonstrated for up to 500 s. It is shown to be compatible with the strong radiation losses for divertor heat load reduction caused by the seeded impurities iron, neon and argon. The corresponding global and local energy and particle transport are presented. The required radiation corrected energy confinement times of the EDA and CDA are found to be close to 4 s, which is attainable according to the ITER ELMy H mode scalings. In the reference cases, the steady state helium fraction is 7%, which already causes significant dilution of the DT fuel. The fractions of iron, neon and argon needed for the prescribed radiative power loss are given. It is shown that high radiative losses from the confinement zone, mainly by bremsstrahlung, cannot be avoided. The radiation profiles of iron and argon are found to be the same, with two thirds of the total radiation being emitted from closed flux surfaces. Fuel dilution due to iron and argon is small. The neon radiation is more peripheral, since only half of the total radiative power is lost within the separatrix. But neon is found to cause high fuel. Dilution. The combined dilution effect by helium and neon conflicts with burn control, self-sustained burn and divertor power reduction. Raising the helium fraction above 10% leads to the same difficulties owing to fuel dilution. The high helium levels of the present EDA design are thus unacceptable. For the reference EDA case, the self-consistent electron density and temperature at the separatrix are 5.6*10¹⁹ m^-3 and 130 eV, respectively. The bootstrap current of 2.4 MA is much lower tha

An ergodic magnetic structure is generated in the TBR-1 Tokamak boundary to handle heat flux and minimize impurity generation. In this work, the effect of this structure on the plasma edge turbulence is investigated. Thus, spectral alterations driven by perturbing magnetic fields created by resonant helical windings are observed in the small Tokamak TBR-1. The investigated oscillations are detected with an array of Langmuir and magnetic probes. Spectral and bispectral analyses revealed significant changes induced by the resonant helical windings, mainly in the MHD frequency band. Thus, whereas the known Mimov oscillation amplitude reduction is observed, the density and potential oscillation spectra are slightly affected. However, the bispectral analysis shows a clear reduction of the wave-wave mode coupling in the MHD frequency range, while the high frequency turbulent coupling is steadily reduced

The effect of the adiabatic invariant diffusion caused by repeated separatrix crossings, when the ion emerges step by step from and is entrapped by the local magnetic well, on the transport of locally trapped ions in Tokamaks with a rippled magnetic field is considered. The corresponding alpha particle loss rate is comparable with that induced by the stochastic ripple diffusion of banana tips

A new scheme of plasma production and ion heating in the central cell of the tandem mirror is demonstrated. Helicon mode plasma production in combination with two ion species plasma heating by mode conversion of a fast Alfven wave into a slow Alfven wave realizes ion heating of a high density plasma, and could remove the density limitation caused by the conventional scheme using slow ion cyclotron waves. The achieved value of products of electron density and average ion temperature is 9*10²⁰ eV.m^-3 for an RF power of 165 kW. Analysis of wave damping indicates that matching the radiation spectrum of the antenna with the damping spectrum results in efficient plasma production. The phased dual double-half-turn antenna is suitable for helicon wave spectrum control

A new gain model based on an adiabatic self-similar solution of the hydrodynamic equations is proposed for inertial confinement fusion (ICF) targets ignited by means of a thermonuclear spark at the fuel centre. The model is applied to analyse gain curves corresponding to fixed values of the implosion velocity U_im. It is shown that the adequate ignition criterion, allowing for the inertia of the cold fuel, implies ρ_sR_sT_s varies as U_im at the time of ignition, as contrasted to fixed values of the spark areal density, ρ_sR_s, and the temperature, T_s, assumed in many of the earlier publications. The modified ignition condition leads to the scaling E_min varies as α³U_im^-7 and G_f* varies as (E/α³)^0.4 for the ignition energy threshold, E_min, and the limiting fuel gain, G_f*, of ICF capsules; α is the isentrope parameter of the cold fuel, E is the energy invested in the DT fuel. Stability and symmetry constraints do not affect this scaling when the initial aspect ratio of the fusion capsule A₀ >> 1; in the opposite case of initially thick capsule shells, the scaling of E_min and G_f* with U_m, and α becomes ill defined

A numerical study of the dynamics of the L-H transition in JET was done using models of anomalous transport which depend either on local or on global plasma parameters. Comparison with experimental results shows that the best agreement is reached when using a global model for reducing χ_e, χ_i and D in the region where shear becomes large (e.g., outside the q=1 surface). Thus the L-H transition manifests itself as a sudden drop of the transport coefficients everywhere outside the q=1 surface, not only in a narrow layer near the plasma edge. The slow improvement of plasma confinement which follows such a fast transition could then be attributed to the dependence on the local plasma parameters of the transport coefficients in the plasma core

Local atomic hydrogen densities inside plasmas of the Compact Helical System (CHS) were measured using laser induced fluorescence tuned to H_α, and the results were compared with those calculated using a neutral particle simulation code, in order to determine the penetration energy of hydrogen atoms from the chamber wall into the plasma. The results of these and supplementary measurements of H_α emission intensities using CCD cameras and photomultipliers at various toroidal positions yielded local atomic hydrogen density distributions throughout the whole plasma. The distribution of the particle source rate was evaluated from the atomic hydrogen distribution, and then the radial particle flux profile was obtained from a particle continuity equation. Estimated errors for the latter were ±30% at an absolute level and several per cent in the profile shape. The radial particle flux thus obtained is valuable for evaluations of particle diffusion coefficient and convection velocity