Table of contents

At present approximately 85% of energy consumption is met by burning fossil fuels. The world population is at present 5.5 billion. United Nations projections for 2060 give a median prediction of population doubling, a 'worst case' prediction of tripling and a 'best possible' 50% increase. Clearly the demand for energy is likely to increase by factors of 1.5 to 10 depending on the actual population growth and the achieved improvements, if any, in the mean standard of living. Fossil fuels are a finite resource and estimated reserves correspond at present energy consumption rates to approximately 50 years for oil and gas and several hundred years for coal. The only major alternatives to fossil fuels are solar, fission and fusion power. These are all at different states of development and have very different environmental effects and perceived safety aspects. Given the magnitude of the long-term energy problem it is clearly important to aim at diversity of supply and to develop each system to its full potential. The purpose of fusion research is therefore to explore the related science and technology and to develop prototype power-generation systems.

A description of the design of ITER as it has progressed during the first year of the Engineering Design Activities is given. The approach is guided by a conservative extrapolation of present-day tokamak physics and the intent to produce a device relevant to development of an attractive fusion reactor. The main parameters are determined by ensuring adequate confinement margin, and the beta -limit then permits a fusion power in excess of 3 GW. A major problem confronting the design is management of the exhaust of thermal power in the divertor. Progress on modeling plasma behavior in the novel divertor design proposed by Rebut is presented. A conceptual design of a first-wall and blanket using an advanced material (vanadium) and liquid lithium for cooling and breeding is also described.

Based on recent progress in world tokamak research, the prospects of continuous operation of a tokamak fusion power reactor are discussed. Efficient steady state operation (Q>30) becomes feasible when a large fraction of the plasma current (>70%) is driven by the bootstrap current. Moderate current (I_p approximately 10-20 MA) and an enhanced confinement mode of operation are the key for efficient steady state operation. High magnetic fields (B_t=8-11T) and high Troyon factors ( beta _N>3.5) are required to attain high beta _p operation and a reasonably high fusion power density. Control of the current and edge pressure profiles is necessary for high beta _N operation with a high bootstrap fraction. It is proposed that the Next Step Tokamak should be optimized for continuous operation as well as high current pulsed operation.

Regimes of tokamak operation more favourable to either inductive or non-inductive current drive are discussed. Steady-state operation, while desirable, appears to be limited to 'advanced' operation regimes. Non-inductive and inductive current drive are shown for 'conventional', high current, and 'advanced', low current, concepts. Thermal energy storage and augmentation systems suitable for pulsed systems are shown, and the issues of stress and thermal cycling are introduced. The advances in physics that may lead to attractive steady-state concepts are shown also to lead to attractive pulsed concepts.

Solution of the ITER divertor problem will require techniques which redistribute the SOL power onto a much larger surface than conventional target plate design permits, while maintaining good control of impurities. Progress made in developing such techniques is reviewed. These include gas target divertors, impurity-seeded radiating divertors, and glancing incidence targets.

Some key physical aspects of the inertial confinement fusion (ICF) are discussed. The minimum scale of ICF microexplosions is determined by the ability to implode spherical shells with high radial convergence ratios C_R and high initial aspect ratios A_R0. The attainable values of C_R are limited by large-scale drive asymmetries, while the values of A_R0 are constrained by the Rayleigh-Taylor instability. Under the indirect drive approach to ICF, it is easier to achieve the required uniformity of the drive pressure, but the penalty is a factor 4-5 reduction of the target energy gain as compared to the direct drive option. Spark ignition is a crucial issue for indirect drive targets (at least for those to be used in power reactors), while the targets driven directly by heavy ion beams could, in principle, utilize a less demanding volume ignition scheme.

Small tokamaks have provided close support to their bigger brothers since the beginning of tokamak research, leading to improvements in operation and understanding through programmes of work aimed at elucidating the basic plasma physics underlying tokamak phenomena. They have proved flexible and quick to respond to testing new ideas and have pioneered a number of advanced concepts now being pursued on the major tokamaks. Many of the experimentalists, operators and diagnosticians of tomorrow's large machines will have trained on the small tokamaks of today. These devices have an important role to play in providing input to the ITER physics R&D tasks and to the development and implementation of new diagnostics. About 50 'small' tokamaks are currently in operation, focusing on a wide range of physics issues. Assessment of the output from these small devices shows that there need to be stronger links between the small and large tokamaks.

Despite its limitations, magnetohydrodynamic theory remains the best possibility for a predictive framework for the large-scale dynamics of a magnetized plasma. The mathematical structure is very similar to that for Navier-Stokes fluids, and indeed contains fluid dynamics as a special case. Fluid dynamics would not have gotten very far without understanding the crucial role played by dimensionless numbers (such as the Reynolds number) in classifying its regimes of different kinds of behavior. The situation seems to be much the same in magnetohydrodynamics. In particular, the Hartmann number, familiar in the theory of MHD power generation, seems to be the crucial number describing the onset of MHD activity in voltage-driven dissipative equilibria that model such confinement devices as tokamaks. Stability thresholds are calculable and the supercritical behavior above those thresholds may be quantitatively compared with the numerical computations of Shan et al. (1991, 1993).

Development and results of the MHD description of stellarators over the past half-decade are reviewed. Major events were the acceleration of convergence of those equilibrium codes which assume nested surfaces, the advent of equilibrium codes which do not a priori assume nested surfaces, and the advent of stability codes for nonlocal modes. Major results having these developments as prerequisites were that stellarators can definitely be optimized, can have nicely nested surfaces at finite beta at least in a limited sense, and can be completely ideal-MHD stable at elevated beta values.

Most of the recent studies on fast particles deal with the stabilization-destabilization of magnetohydrodynamics modes. Here the authors present another aspect of fast particle resonances in tokamaks and review some recent results on the dynamics of energetic electrons and alpha particles. The interaction between the runaway population and the magnetic ripple is analyzed. The occurrence of an induced radiative slowing down is demonstrated and its potential is assessed. Then, the resonant channeling of alpha particle free energy, to either fast electrons or fast ions, by means of lower hybrid waves is investigated, and the efficiency of this process is analyzed.

An overview of reflectometry measurements of density profiles and density fluctuations in fusion devices, is presented. The results are analysed in the light of the expectations from recent theoretical and numerical studies. The spatial resolution of the diagnostic and the wave number sensitivity are discussed. The diagnostic capabilities are considered in view of its application in next generation devices.

The ultimate aim of pellet ablation studies is to predict what the plasma profiles are just after a pellet injection. This requires description of the pellet ablation process, the parallel expansion of the ablatant and the fast outward motion of the deposited material since these three phenomena successively occur from the time of pellet injection to the moment when new axisymmetric profiles are reached. Only the first two points have been quantitatively modelled. If the most important processes of ablation physics are identified and although current models reproduce both measured penetrations and averaged characteristics of ablation clouds, some debatable points remain, mainly bearing on the drifts associated with the pellet motion and, consequently, on the effective shielding efficiency of the ionized part of the ablation cloud. During its parallel expansion, the ablated material experiences a strong poloidal rotation which depends on the ratio of the pellet and plasma masses and is due to the total kinetic momentum conservation on each magnetic surface. The fact that this rotation occurs on the same timescale as the outward motion suggests that both phenomena can be linked and that a comprehensive model of the whole fuelling process may emerge from considering the pellet and the plasma as a unique system.

Experimental results from TEXTOR are presented to provide strong evidence for the feasibility of the 'cold radiative plasma mantle', a concept which might be a possible solution for the energy exhaust problem in a fusion reactor. The concept is compared with the high density divertor. The compatibility to other constraints, limitations and open problems are discussed, in particular the issues of stationarity (feed-back control, thermal instabilities, q=2), energy confinement, He-exhaust and fuel dilution.

An overview is presented of recent results on the creation of electric fields in the edge of limiter or divertor tokamaks. The practical implementation and theoretical basis of several schemes, generating radial and/or poloidal fields, are outlined. The manipulation of edge and scrape-off-layer profiles and the control of particle exhaust is discussed. Contributions of biasing experiments to H-mode physics are highlighted. Some prospects for biasing in next generation tokamaks are finally given.

This review examines results from all non-circular tokamaks with a distinct emphasis on investigations in ASDEX-Upgrade. There a major fraction of the experimental time has been dedicated studying vertical displacement events of single null plasmas over a large range of q-values in an attempt to obtain the scaling of both the displacement dynamics and the splitting of forces between those associated with poloidal and toroidal plasma currents as a function of q and B_t. These studies on different tokamaks are accompanied by simulations with-among other codes-the tokamak simulation code TSC, in a version where halo currents flowing in the plasma scrape-off layer (SOL) evolve self-consistently. The technical consequences of VDEs for the machine design, measures taken and first predictions are discussed. Safety setups that have been developed and possible avoidance strategies are briefly described.

ASDEX Upgrade is a poloidal divertor experiment very similar to ITER with respect to magnetic field properties and especially to the plasma boundary geometry. A large part of the programme on ASDEX Upgrade is therefore dedicated to investigating and optimizing reactor-relevant plasma boundary issues. This contribution is concerned with H-mode studies and edge physics in general. The L/H-power threshold, H-mode confinement, details of ELM dynamic and 'dithering' L/H-transitions will be discussed. The edge physics investigations are concerned with model calculations and validations, the characteristic 'approach' to the density limit (DL), target plate sputtering and the influence of hydrocarbons, asymmetries and drifts depending on the direction of the toroidal magnetic field, and finally first results of high-Z material experiments.

Recent results from JT-60U on high performance experiments and studies on steady state plasma are presented. High fusion performance of n_D(0) tau _ET_i(0)=1.1*10²¹m^-3s keV, S_n=5.6*10¹⁶s^-1, Q_DT=0.6, T_i(0)=40 keV has been obtained in the regime names 'high- beta _p H-mode'. By intensive survey on pressure/current profile control, beta limit has been substantially expanded: beta _p<or=4.3, beta _N<or=4.0%mT/MA. As a result, up to 10 seconds of quasi steady state high confinement discharges, with a substantial bootstrap current fraction have been experimentally demonstrated. Up to 3.6 MA of plasma current has been driven by LHCD. In low current (<or=1.2 MA), full non-inductive current drive has also been achieved by NB and NB+LHCD.

The Uragan-2M (U-2M) (IPP, Kharkov) concept is the result of a search of torsatron system with a reduced helical ripple but still having a moderate shear and magnetic well. In this report the authors present the description of concept of the torsatron with an additional toroidal field, results of studies of confining properties of magnetic system of U-2M, design concept and its realization and first results of magnetic configuration studies for U-2M.

Electron cyclotron current drive experiments have been carried out in Compass-D using a power of approximately 500 kW at a frequency of 60 GHz. The fundamental resonance was used and the waves were launched from the high field side of the torus through four mirror-antennae. Significant asymmetry in the loop voltage behaviour is observed when comparing co and counter current drive cases, suggesting a driven current approximately 15 kA (in a plasma current of 130 kA). The BANDIT-3D Fokker Planck code has been used to model these discharges revealing a similar value for the predicted driven current at zero loop voltage but strong synergistic effects with the toroidal electric field at the approximately 0.4 Volts/turn typical of the experiments. This synergy appears to be sufficient to explain the inferred net 'co' current drive in most of the 'counter' current drive shots. In H-mode discharges featuring a strongly asymmetric single null separatrix configuration, discrete ELM's sometimes produce sufficient relaxation of the current profile to cause loss of vertical control, leading to vertical displacement events. Application of resonant magnetic perturbations (RMPs) of predominantly n=1 (with various low m values) is seen reliably to increase the frequency of the ELM's, although the exact mechanism underlying this effect has still to be elucidated. A real-time electronic neural network has been used for the first time for feedback control of the equilibrium in a tokamak. The neural network was used for feedback control of the plasma elongation throughout the discharge, while simultaneously monitoring the plasma vertical and horizontal position.

The study sets out to investigate the parts of the FTU operating space over which pellet injection effects performance improvement and to elucidate the phenomena which hinder it. In this respect the behaviour of the q=1 resonance is found to play the most relevant role.

A review of fast electron dynamic studies in the tokamaks TS, JET, and ASDEX during lower hybrid current drive is presented. In almost all experiments, the collisional slowing down is predominant, so a straightforward assessment of the fast electron radial transport is difficult, especially for large size tokamaks. From recent LH power modulation experiments performed on TS at low density, the fast electron diffusion coefficient, D_st, is found between 0.1-0.3m²s^-1, close to the value determined on ASDEX in similar plasma conditions. An estimate of D_st is also obtained in TS and JET by modeling the plasma dynamics, using ray-tracing+Fokker-Planck codes. For JET, the fast electron transport is fully taken into account in the Fokker-Planck equation. Despite some quantitative discrepancies between D_st values determined from different methods, the diffusion rate of the fast electrons is weak enough for current profile control scenarii to be relevant on large tokamaks. A determination of the type of the turbulence - electrostatic or magnetic - that could drive the radial transport of fast electrons requires finer LH experiments.

The VH-mode regime of high confinement has been observed in both DIII-D and JET. VH-mode is characterized by thermal confinement twice that seen in H-mode, with the edge transport barrier penetrating deeper into the plasma. Two mechanisms have been identified as important in achieving this high level of confinement. Expansion of the E*B velocity shear turbulence suppression zone is important in allowing reductions in local transport, while access to the second ballooning stability regime in the edge allows avoidance or elimination of ELMs which impede the confinement improvement. The high performance phase of these discharges is usually terminated by an MHD event which removes energy from a large portion of the plasma cross-section, and is followed by an H-mode phase.

Limitations of present knowledge of plasma equilibrium profiles hamper a proper identification of MHD modes. However, through continued improvement of both numerical calculations and experimental observations the authors may witness the birth of a new kind of spectroscopy, properly called MHD spectroscopy, in the coming decade. This is illustrated by studies of edge localised modes and external excitation of toroidal Alfven eigenmodes.

Anomalous transport in tokamaks is usually explained in terms of turbulent fluctuations. Predictions for the confinement properties of tokamaks would be placed on a more sound basis if the level of fluctuations and the resulting transport could be calculated. At present there is no clear agreement on whether the responsible fluctuations are electrostatic or magnetic. This paper discusses the existing evidence and arguments, both direct and circumstantial, that might help to resolve this question.

A discussion of linear drive and damping mechanisms on shear Alfven waves in hot, reactor-relevant plasmas is presented. According to whether equilibrium geometry, kinetic effects, or both are important in determining the mode structure, different branches of Alfven waves may exist: TAE, KAW and KTAE, respectively. The existence of open problems, as the necessity of non-perturbative calculations, is also emphasized.

The local electron and ion heat transport as well as the particle and impurity transport properties in stellarators are reviewed. In this context, neoclassical theory is used as a guideline for the comparison of the experimental results of the quite different confinement concepts. At sufficiently high temperatures depending on the specific magnetic configuration, neoclassical predictions are confirmed by experimental findings. The confinement properties in the LMFP collisionality regime are discussed with respect to the next stellarator generation, for which at higher temperatures the neoclassical transport is expected to become more important.

The results of the first year of operation of the experiment RFX are reported. Profiles of electron density, electron and ion temperature and impurity emission have been measured at plasma current I<0.7 MA. The energy confinement parameters at different density are reported, the best values ( tau _E approximately 1ms, beta _theta approximately 8%) being obtained operating at higher density. The role of the impurity content in determining the present performance of the experiment is discussed.

The gyrotron is one of the most powerful sources of short wavelength microwaves. It is used in ECH and ECCD plasma experiments, and is intended to be used for plasma diagnostics and is proposed to be used for LHCD.