Table of contents

During H-mode experiments with ion cyclotron resonance frequency heating on JET it was observed that the influxes of deuterium and of impurities can substantially modify the quality of both energy and particle confinement. In particular, a low particle confinement H-mode (LPCH-mode) was triggered either by deuterium input into the X-point region or as a result of a moderate impurity influx into the plasma boundary. In the LPCH-mode the particle confinement is at least three times less than in a normal H-mode. The electron density, the deuterium density and the radiated power are also less than in a normal H-mode. However, the deuterium dilution reaches a steady state. The approximately 20% decrease in the plasma energy content is due to the drop in density. The H-mode duration is extended to 2.8 s by the LPCH-phase and is limited only by the length of the heating pulse. The LPCH-mode particle confinement is linked to L-mode-like particle diffusion within the radius 0.5 < ρ < 0.8 and to the H-mode-like barrier due to the large ratio of the inward convective velocity to diffusion coefficient in the plasma boundary. This mode has the potential for substantially extending the H-mode phase to make it more reactor relevant

Many experiments have proved the effectiveness of lower hybrid waves for driving toroidal current in tokamaks. However, the use of these waves to provide all the current in a reactor is thought to be uncertain because it may happen that the waves do not penetrate into the centre of the more energetic reactor plasma and, if they do, the wave power may be absorbed by alpha particles rather than by electrons. The authors address mathematically the interaction between lower hybrid waves and alpha particles

Symmetrization of the radiation field inside hohlraum targets for indirectly driven heavy ion beam inertial confinement fusion (ICF) is investigated numerically. The targets considered consist of a casing, enclosing the spherical fuel capsule, and a few cylindrical radiators, schematically representing ion beam irradiated converters. Radiation absorption and re-emission are dealt with as in the paper by Murakami and Meyer-ter-Vehn (Nucl. Fusion 31 (1991) 1333), but with the geometry extended to three dimensions and with the finite size of the radiators taken into account. It is found that, for a practical casing to capsule area ratio (of the order of ten) and practical converter aspect ratios, two converters (allowing for two-side axisymmetric irradiation) cannot provide the uniformity required for ICF. However, with a spherical casing with six converters (placed in couples along the axes of a Cartesian co-ordinate system) it would be possible to illuminate a capsule with a non-uniformity well below 2%, which could satisfy the ICF requirements. The effects of changing the area ratio, the size and position of the converters and the geometry of the hohlraum are also discussed

The authors presents a plasma density perturbation analysis which closely couples measurements of perturbations in density, temperature and turbulent fluctuation levels with a theoretical model in order to calculate particle fluxes. Non-linearities in the plasma model and the modulation of plasma parameters such as the electron temperature caused by the density perturbation are found to be significant. Some necessary details of the modification of the density fluctuation spectrum, and the plasma potential and Z_eff profile remain to be measured. However, the available data allow steady state and perturbed particle fluxes in the low collisionality region of TEXT (5 cm < r < 20 cm) to be calculated using a simplified dissipative trapped electron model. These calculated fluxes do not agree with the measured fluxes

The design of a radiation driven capsule suitable for use in a heavy ion beam driven inertial confinement fusion reactor system is discussed. Numerical simulations of compression, ignition and burn of the capsule are presented. These simulations have been carried out using the one-dimensional, three-temperature, Lagrangian radiation hydrodynamic computer code MEDUSA-KAT. The capsule consists of a 4 mg cryogenic spherical DT shell surrounded by an ablator shell made of a mixture of 90% Li and 10% Au atoms. Inside the fuel shell there is a low density DT gas. It has been found that an input radiation energy of 5.00 MJ is needed to implode the capsule, and the implosion yields about 660 MJ output thermonuclear energy. The capsule energy gain is thus about 130, which may be sufficient to operate a heavy ion beam driven reactor system because of the relatively high efficiency of the heavy ion driver. A study of the capsule gain as a function of the input parameters has also been carried out. It is found that the gain is not sensitive to a variation in the prepulse temperature and prepulse duration of up to 50%

The author discusses a method of calculating particle and heat fluxes (including impurity ions) in the Pfirsch-Schluter regime in stellarators by integrating along the magnetic field lines. A computer is used to calculate the diffusion fluxes through the magnetic surfaces; these are determined by numerical methods, and full account is taken of the influence of three-dimensional inhomogeneity in a known confining magnetic field

For interpreting fusion rate measurements in present fusion experiments and predicting the fusion performance of future devices or of d-t experiments in present devices, it is important to know the fusion cross-sections as precisely as possible. Usually, it is not measured data that are used, but parametrizations of the cross-section as a function of the ion energy and parametrizations of the Maxwellian reactivity as a function of the ion-temperature. Since the publication of the parametrizations now in use, new measurements have been made and evaluations of the measured data have been improved by applying R-matrix theory. The authors show that the old parametrizations no longer adequately represent the experimental data and present new parametrizations based on R-matrix calculations for fusion cross-sections and Maxwellian reactivities for the reactions D(d,n)³He, D(d,p)T, T(d,n)⁴He and ³He(d,p)⁴He

For a detailed prediction of the plasma parameters in the Large Helical Device (LHD), 3-D equilibrium/1-D transport simulations including empirical or drift wave turbulence models are performed which suggest that the global confinement time of the LHD is determined mainly by the electron anomalous transport in the plasma edge region rather than by the helical ripple transport in the core region. Even if the ripple loss can be eliminated, the increase in global confinement is 10%. However, the rise in the central ion temperature is more than 20%. If the anomalous loss can be reduced to half of the value used in the present scaling, as is the case in the H-mode of tokamak discharges, the neoclassical ripple loss through the ion channel becomes important even in the plasma core. The 5% radial inward shift of the plasma column with respect to the major radius improves the plasma confinement and increases the fusion product by more than 50% by reducing the neoclassical asymmetric ion transport loss and increasing the plasma radius (10%)

Density limits and the evolution of disruptions were studied on TEXTOR for different wall materials facing the plasma surface. A systematic extension of the density limit was found over a large current range when the metallic wall was first exchanged by a carbon coated wall and later by a boron-carbon coated wall. The maximum attainable density nearly doubled when metallic impurities were replaced by low Z impurities such as boron and carbon. With a neutral beam additional heating power of P_NI=3.2 MW, average densities exceeding n_e=10²⁰ m^-3 were obtained. The maximum density observed scales like the square root of the auxiliary heating power. The sequence of processes leading to a density limit disruption was experimentally analysed in detail for ohmically heated plasmas

The Thomson scattering spectrum represents the projection of the three-dimensional electron velocity distribution on the scattering vector. From this the local electron temperature and density can be derived. To determine the three-dimensional electron velocity distribution it is necessary to have several viewing directions, assuming axial symmetry in velocity space perpendicular to the magnetic field. Thomson scattering experiments in the TORTUR tokamak demonstrated the first experimental determination of f(v_⊥, v_||) from radially and tangentially observed spectra. The latter made it possible to determine the toroidal current density. The scattering spectra, averaged over 30 laser pulses observed in both the radial and the tangential direction, show clearly two symmetrical dips near the top of the Gaussian spectrum

Mixing experiments have been performed using Si/Al/Au trilayer targets accelerated with laser beams smoothed by random phase plates. The Al-Au mixing is detected by heating the rear of the target with a probe beam and observing the Al line emissions. A null-mix experiment has been performed with Si/Al/Si, which is stable according to Rayleigh and Taylor, in order to check that the probe beam does not induce a specific mixing. Simulations with a hydrocode have been made to characterize the main features of the target when the instabilities occur. In the case of Si/Al/Au targets, the Al emission indicates a mixing of Al and Au for a driver irradiance as low as 10¹³ W/cm² and that most of the Au is mixed with Al

Recent experiments in DIII-D have obtained low density, hot ion plasmas. It was found that the ion thermal diffusivity was substantially smaller than the electron thermal diffusivity in the central plasma region. Previously it was suggested that this result is a consequence of hot ion trapped ion mode turbulence. This effect has been incorporated into a steady state transport code. Simulations of L-mode and H-mode DIII-D hot ion plasmas have been performed. The simulation results for the temperature profiles are found to be in reasonably good quantitative agreement with the experiment

The effect of the magnetic flux geometry of a poloidal divertor on the profiles and parameters at the divertor target is described by means of a simple model. Because of the magnetic flux expansion near the X-point, exponential profiles at the midplane of the tokamak are transformed into Gaussian profiles at the target. When the angle of incidence of the field line onto the target is considered, the possible existence of a maximum in the flux onto the target away from the magnetic separatrix appears naturally, in agreement with experimental results

Using a waveguide array designed to excite only fast-wave polarization, both slow and fast waves were observed in a toroidal plasma. The relative energy in the two modes was found to be a function of the plasma density, indicating that fast-wave current drive experiments should be performed at densities well above the calculated fast-wave cut-off density

Off-axis heating experiments have been conducted in DIII-D using 75 keV neutral beam injection as the auxiliary power source. The application of 7.5 MW of deuterium neutral beam power resulted in the plasma transiting into the high density H-mode confinement regime with equal electron and ion temperatures. The global thermal energy confinement was not affected by changing the heating location. A power balance analysis employing the 1¹/₂-D ONETWO transport code and assuming purely diffusive heat transport found that the one-fluid effective diffusivity changed dramatically between the two cases. These results indicate the possibility that the local diffusivity is a function of some quantity other than the density, temperature and their gradients. An alternative explanation is that the diffusivity remains unchanged and that there is an inward non-diffusive heat flow to balance the power flow

Low aspect ratio tokamaks may provide a low cost route to developing magnetic fusion energy and may offer simpler and more economic fusion reactors. The first test of this concept with hot plasmas has been conducted with the Small Tight Aspect Ratio Tokamak (START) experiment now in operation at Culham Laboratory. The novel method of plasma formation is described and observations of plasmas at aspect ratios as low as 1.3 are presented, indicating high temperature discharges with good global stability properties and typical tokamak characteristics. This represents a new and significant extension of the parameter range of the conventional tokamak

Summary report on the IAEA Technical Committee Meeting on Fast Wave Current Drive in Reactor Scale Tokamaks (Synergy and Complementarity with LHCD and ECRH) held at Arles, France, 23-25 September 1991.