Table of contents

JET is investigating beryllium as material for walls and limiters. Studies were carried out with initially a thin layer of beryllium evaporated onto the walls of the machine, later in addition the graphite material of the belt limiter was exchanged against beryllium. The use of this material was generally beneficial for the plasma behaviour. Combined with a reduction in the oxygen content, strong pumping of hydrogen isotopes was found which allowed JET to widen considerably the operational space with respect to ion temperatures, densities and plasma purity. In this paper a comparison of the plasma performance with graphite and beryllium will be presented. The author discusses especially the impurity behaviour with respect to fluxes, concentrations, effective charge and dilution, and reports on density limits, disruption behaviour, wall pumping, hydrogen isotope retention, and the power handling capability of the beryllium limiter in the present design. Examples of improved plasma performance are given.

Recent lower hybrid current drive (LHCD), and heating (LHH) experiments on JT-60 are reported. The current drive product of n_eR_pI_RF approximately 12.5*10¹⁹ m^-2 MA was achieved at the LH power of approximately 4.5 MW, and the CD efficiency, the energy confinement, the global power balance and the heat load on divertor plates were investigated in high power LHCD plasmas. Nearly steady state H-mode discharges were found during LHCD with two different frequency injections. Sawtooth suppression in NB heated plasmas by LHCD have shown an improvement in confinement near the plasma center. Parametric instabilities in LH heating experiments were significantly reduced by increasing the plasma current, and the stored energy increased linearly with heating power of up to approximately 9 MW at n_e approximately 7*10¹⁹ m^-3 and I_p=2.75 MA. Parametric instabilities near the plasma edge in the ion heating regime were also reduced in peaked density plasmas produced by pellet injection and LH waves increased the central plasma pressure at n_e(0) > 1.4*10²⁰ m^-3.

Recent results from the DIII-D tokamak have provided significant contributions to the understanding of many of the elements of tokamak physics and the application of this understanding to the design of next generation devices including ITER and CIT. The limitations of magnetohydrodynamic stability on the values of plasma beta (the ratio of kinetic pressure to the containing pressure of the magnetic field) that can be attained has been experimentally demonstrated and found to be described by existing theory. Values of beta (10.7%) well in excess of those required for proposed devices (ITER and CIT) have been demonstrated. Regimes of confinement (H-mode) have been established that scale favorably to proposed next generation devices, and experiments demonstrating the dependence of the energy confinement on plasma size have been completed. Understanding of confinement is rapidly developing especially in the areas of bulk transport and the role of turbulence in the plasma edge. Key experimental results in areas of plasma transport and edge plasma phenomena are in agreement with theories based on short wavelength turbulence. Control of the divertor heat loads and impurity influx has been demonstrated, and new progress has been made in the understanding of plasma edge phenomena. Experiments with ion Bernstein wave heating have not found regimes in which these waves can produce effective central ion heating. Electron cyclotron current drive experiments have demonstrated 70 kA of driven current in 400 kA discharges.

Neutral beam injection (NBI-co and NBI-counter), ICRH and their combinations have been compared on TEXTOR for what concerns the heating and their effect on the energy anisotropy and the confinement. In most of the cases, stationary heated plasma conditions have been obtained with boronized wall. The main results are: (i) Production of a hot ion mode with NBI-co. It is characterized by a large energy anisotropy, by a large noninductively driven toroidal current and by the stabilization of the sawteeth (of the monster type) due to the hot ion tail. (ii) Significant enhancement of these effects when ICRH is added to NBI-co. The neutron yield is also increased by the addition of ICRH. Peak electron and ion temperature around 3 keV have been obtained and 70% of the total current has been non inductively driven. (iii) Large excess of the total energy content (up to 2.3) with respect to the L-mode scaling predictions has been obtained with the combinations NBI-co+ICRH or NBI-co+NBI-counter. (iv) Up to 6 MW of additional power has been coupled to the plasma leading to beta _p=1.5 and beta _t equal to 70% of the Troyon limit.

In the reversed field pinch various turbulent transport mechanisms might be simultaneously active, including MHD flow effects, free-streaming in a stochastic magnetic field, direct transfer of fluctuating energy to ions, and electrostatic fluctuation driven transport. By comparing fluctuations and transport in the RFP to that in the related toroidal configurations of the tokamak and stellarator, a greater understanding of toroidal confinement in general might be achieved. In addition to particle and energy transport, current transport sustains the reversed field magnetic configuration and constitutes the dynamo effect. Recent results are available from RFP experiments on electrostatic fluctuations, anomalous ion heating, the presence of free-streaming fast electrons, and the extension of these effects to plasmas of large size. The reactor promise of the RFP is based upon the conjecture that the beta value will remain high and the resistance scaling classical, as has been the case in past experiments. One might expect beta to be limited by resistive interchange turbulence. The MST, RFX, and ZTH experiments will form a sequence of devices to test this scaling conjecture to large size and large current values.

The self consistency conditions for magnetic turbulence are reviewed. The principal features of a magnetic topology involving stochastic flux lines-transport, ambipolarity, basic equations-are summarized. Two driving sources are considered: thermal effects which require large scale residual islands and electron diamagnetism which involves fluctuation scales smaller than the ion Larmor radius and a poloidal beta threshold of order one. Detailed stability criteria and transport coefficients are given in each case.

ECR heating at B₀=2.5T has been extensively used in the 1990 experimental period of the W VII-AS stellarator. As it is a low-shear experiment the magnetic configuration (especially details of the rotational transform profile) depends sensitively on plasma currents (pressure driven, ohmic, EC driven, Ohkawa current) which in turn have a strong influence on energy and particle confinement properties. For the stationary phase a transport analysis has been performed, yielding the profiles of the electron heat conduction and the ion particle diffusion coefficients. The former was subjected to a statistical analysis resulting in phenomenological expressions for chi _e and tau _E. First experiments using neutral beam injection (ECRH target plasma) as well as combined heating (NBI+ECRH) are also discussed.

The versatility of the JET device allows a wide range of tokamak operating regimes to be explored and plasmas bounded both by material limiters and by a magnetic separatrix have been investigated extensively. This has permitted the confinement and MHD stability properties of plasmas heated to temperatures above 10 keV by neutral beam injections or ion cyclotron resonance heating to be studied in detail. The results of analyses of transport and confinement in the L- and H-mode regimes in JET are discussed and the properties of H-mode plasmas produced by both major forms of heating are compared. Several aspects of the MHD stability of such plasmas, particularly at high toroidal beta beta _psi , and at the density limit, are reviewed.

There exist two classes of discharges on ASDEX which overcome the severe confinement deterioration at high densities or heating power. The first class is characterized by unusually peaked density profiles and the second one by exceptionally broad ones. These profiles and the concomitant confinement enhancement imply a change in both the particle and heat transport. Whereas the changes in particle transport are not fully understood, transport analyses point out that the improved heat transport in the first profile class can be explained by reduced ion contribution losses coming close to the neoclassical ones. The different results for the ion transport with flat and peaked density profiles are qualitatively consistent with that expected from eta _i-modes. This conclusion is supported by momentum transport studies. The analyses, however, cannot yet explain the electron heat transport which is anomalously high in all regimes.

Studies of the transport properties of tokamak plasmas using perturbation analysis are discussed. The focus is on experiments with not too large perturbations, such as sawtooth induced heat and density pulse propagation, power modulation and oscillatory gas-puff experiments. The approximations made in the standard analysis of such experiments are made explicit and are discussed. References are given to papers that deal with specific aspects of the theory. Points of agreement as well as discrepancies between different experiments and gaps in the experimental data base are highlighted. The analysis of cross-coupling between electron thermal and particle transport using simultaneous measurements of heat and density pulses in JET is discussed, as an illustration of the potentiality to measure off-diagonal elements of the transport matrix in perturbative experiments.

The paper considers the prospectives of the neutron sources for material and component testing for the future fusion reactors. The present state of mirror physics gives good chances for development of a compact and relatively inexpensive (5-10% of ITER cost) neutron source that will allow one to obtain the flux of 14 MeV neutrons in the range 3-5 MW/m² in a test volume exceeding 100 l. The possible role of such a source in the fusion program is discussed.

TORE SUPRA is a large superconducting tokamak designed for sustaining long inductive pulses (t approximately 30 s). In particular, all the first wall components have been designed for steady-state heat and particle exhaust, particle injection, and additional heating. In addition to these technological assets, a strict control of the plasma-wall interactions is required. This has been done at low power: experiments with ohmic heating have been mainly devoted to the pump limiter, ergodic divertor and pellet injection experiments. Some specific problems arising in large tokamaks are encountered; the pump limiter and the ergodic divertor yield the expected effects on the plasma edge. The effects on the bulk are discussed.

A high density implosion of a hollow shell pellet by a GEKKO XII glass laser has been carefully diagnosed to clarify the relation between the achievable density and possible nonuniformities. The density achieved, 600 times solid density, has been well interpreted with the experimental conditions of pellet and laser intensity uniformity.

Stellarators with significant magnetic shear in the United States, the Soviet Union, and Japan are described, and recent results are discussed in terms of their contributions to the physics understanding relevant for stellarator optimization and to toroidal confinement understanding in general. The areas discussed are the properties of stellarators with significant shear, magnetic surfaces, trapped-particle losses, magnetohydrodynamic (MHD) stability, global confinement scaling, local transport, fluctuations, and particle and impurity control.

Marfes are toroidally symmetric bands of radiating plasma that form at the smaller major radius side of tokamaks. A model of marfes and density limit is studied. Though very simple, the model provides alternative explanation of the marfe onset beyond the density limit, explains the marfe poloidal location, states that a marfe can onset only when the criterion ( in /q)² chi /sub // approximately /^< (T_o/T_L)² chi _{perpendicular to} is satisfied, where T_o is the central temperature and T_L approximately 10eV is the temperature of intensive impurity radiation. Density fluctuations observed in marfe experiments are explained as the manifestation of sound waves which are intermediate between the isothermal and adiabatic.

The Los Alamos Dense Z-Pinch program has completed construction of a new, high current (1.2 MA) pulsed power device for ohmically heating solid deuterium fibers and studying the dynamics of these dense, magnetically confined plasmas. Experiments to date have been performed at half maximum bank energy (100 kJ) and half current (650 kA) with fiber diameters of approximately 20 mu m. Peak current is achieved in 100 ns in this three stage machine. A variety of optical X-ray, neutron, and electrical diagnostics are employed to examine the plasma parameters. Motivation for these experiments, experimental results, and the program for future work are presented.

An overview is given of fusion relevant performance in a number of JET operational regimes. In most areas JET provides a suitable platform on which to base the design and predict the performance of a next step device. The main outstanding problem area is that of controlling impurity influxes and plasma exhaust, these are to be studied in a new phase programme under consideration for JET. The fusion performance already obtained in JET is described with 100 kW of fusion power being generated in charged particles from the D-He³ reaction and more than 10 MW of fusion power predicted from present discharges if operated in D-T.

The extrapolation of energy confinement to the conditions of a Next Step tokamak, to operate with an ignited DT plasma, is an essential ingredient of the data base needed for such a device. The present knowledge on energy confinement and the approaches used for providing a basis for the extrapolation are concisely described. The lines of further research are indicated which have to be followed to be able, in a few years, to extrapolate energy confinement to a Next Step tokamak with improved accuracy.