Table of contents

A method is developed for studying MHD equilibrium and stability of plasmas confined in toroidal systems with a large constant magnetic field and a small field which varies rapidly in space. The method is used to obtain a system of magnetohydrodynamic equations averaged over the period of the rapidly varying field. In this form, the problem is greatly simplified, and the equations become similar to those used for axisymmetric tokamaks, for which methods have now been developed in some detail. – The problem of the limiting plasma pressure for equilibrium and stability in a stellarator is analysed, and it is shown that, within the framework of an ideal MHD model, β = 2P/B² can be larger than 10% in systems with large shear and partial compensation of the magnetic fields produced by the diamagnetic currents. Also the question of toroidal drift compensation and, as a result, of the decrease in the neoclassical diffusion coefficient is discussed.

Scrape-off plasma parameters have been measured in ASDEX both in the divertor chamber and in the mid-plane, for neutral-injection powers of up to 2.5 MW. Measurements demonstrate the existence of strong inhomogeneities parallel to the field lines and show the dependence of divertor density and temperature on bulk plasma density, heating power and mode of divertor operation (gettered/ungettered, different divertor slit widths, additional gas puffing in the divertor). Three recycling regimes, characterized by different dependences of the divertor density on the bulk plasma density, are identified in agreement with a numerical fluid model of the scrape-off plasma. In the particularly attractive high-recycling regime, divertor temperatures could be kept around 7 eV for the whole range of heating powers tested.

Finite-ion-Larmor-radius (FLR) effects provide a band of additional stability for values of β(≡ plasma pressure/magnetic pressure) exceeding the limit predicted for ideal magnetohydrodynamic (MHD) ballooning modes. The effect of particle collisions on the stable modes of the FLR-modified ideal theory that exist in this range of β values is examined.

The paper presents a parameter study of implosion, ignition conditions, burn and gain of a single-shell, multi-layered, heavy-ion-beam driven ICF target as a function of input pulse parameters including input power, input energy, incident ion range and pulse shape. It is shown that a prepulse is necessary to achieve central ignition in the fuel. Moreover, the target gain is very sensitive to the level of the prepulse power. The minimum power in the main pulse required to ignite this target is 500 TW. Some calculations are also done by using different ignitor masses to help in understanding the process of spark formation under different conditions. The effect of the shortening of the range of the incident ions on the target gain and the ignition conditions is also considered. These simulation results are explained in terms of the theory of isentropic compression and the theory of shock waves. In this way, it is shown how ignition in such a target is obtained and how its performance is optimized in order to achieve maximum fractional burn and gain.

The d(d,p)t and d(d,n) ³He fusion reactions produce 1 MeV tritons and 0.8 MeV ³He ions which can subsequently undergo d(t, n) α and d(³He,p) α fusion reactions. The magnitude of this triton and ³He ion 'burn-up' was measured on the PLT and PDX tokamaks by detection of the 14 MeV neutron and 15 MeV proton emission. In discharges with B_φ > 2 T, the measured ³He and triton burn-up is consistent (within a factor of three) with predictions based on classical theories of ion confinement and slowing down. In discharges with weaker toroidal fields but constant plasma current, the burn-up of both ions fell by more than a factor of ten so that the observed burn-up was significantly less than expected classically.

A density clamp is observed in DITE tokamak with neutral beam injection heating. Measurements in the scrape-off plasma are described which show an increase in the particle flux to the limiters during injection. This is sufficient to account for the density clamp without requiring a change in the recycling coefficient.

Mechanisms for equilibrium poloidal electric field generation in neutral-beam heated tokamaks are identified and analysed, and predictions are made for the PDX, Dili, and ISX-B tokamaks. A toroidal outward shift of particle density is found to agree qualitatively with recent experimental observations in the PDX tokamak.

A new feedback scheme for the suppression of tearing modes which are considered to be a principal cause of major disruptions in tokamak devices is proposed. Attention is focused on the suppression of the 2/1 mode as this mode strongly destabilizes the other tearing modes. Fourier components of the perturbed radial field are fed back to the toroidal field. Numerical experiments show that the feedback system is worth being discussed as a means of stabilization.

Scientists and engineers, engaged in design, construction and operation, as well as programme planners, met last year for the '12th SOFT' (Symposium on Fusion Technology) at the Jülich Nuclear Research Centre (KFA Jülich) from September 13 to 17. Just under 500 participants from twenty countries in North America, Asia, Australia and Europe were registered.

The latest research results in the areas of tritium breeding materials properties, coolants, hydrogen isotope interactions with structural materials, tritium recovery cycles and plasma interactions with the first wall are reviewed. Areas where required information is currently lacking are identified, and specific recommendations are made for work which should be given priority in order to most rapidly advance the international effort in fusion technology. These areas include: liquid-metal/structural-alloy interactions (including magnetic-field effects), analytical techniques for low levels of tritium and impurities in breeder materials, tritium interactions with breeder and structural materials, thermochemistry and therm ophysical properties of solid breeder materials, corrosion product transport and deposition in magnetic fields, hot-water reactions with Li₁₇Pb₈₃ and ceramic breeding materials, tritium handling and recovery cycles, chemical reactions at the plasma/wall interface and impurity getters, and in-pile of pilot-scale tests of breeder materials, tritium interactions with structural materials and discharge cleaning of advanced first-wall materials.