Table of contents

Since the tokamak concept gained recognition in 1968, steady progress has been made towards its development into a viable fusion core for a thermonuclear reactor. The paper summarizes the present status of tokamak research, concentrating on physics aspects relating to the operational domain in current, density and pressure; plasma activity and relaxation phenomena; the principles, strengths and weaknesses of plasma heating techniques; global and local plasma confinement and the various regimes of enhanced confinement identified so far; the potential and limitations of fuelling, impurity control and exhaust systems; and, finally, the performance of non-inductive current drive schemes.

The present status of stellarator experiments and recent progress in stellarator research (both experimental and theoretical) are reported by groups in the USA, the USSR, Japan, Australia and the European Community (Federal Republic of Germany and Spain). Experiments under construction and studies of large, next-generation stellarators are also described.

The paper summarizes the theoretical basis for the RFP and reviews the status of research in this field. The RFP is a relaxed state system well described by Taylor's theory which explains many observations. The RFP is of interest because its study will increase the understanding of toroidal confinement in general, which might lead to better reactor designs, and the RFP itself has potential as a reactor, for example the improved, high energy density, compact RFP reactor. In the last five or ten years, RFP research has expanded, with some 15-20 machines operating or under construction, and the plasma parameters have improved substantially, with a confinement time of 0.5-1 ms and temperatures approaching 1 keV; values of β_θ ∼ 5-15% are reached routinely. Following an overview of recent results, three key problems are discussed in more detail: (1) resistivity, edge physics and ion heating (the ions are heated by fluctuations which drive the RFP dynamo); (2) operation of the RFP with an (electrically) thin shell which permits the growth of new unstable modes which degrade the confinement; and (3) scaling. Over the current range of 0.5 MA, favourable scaling trends of temperature and confinement with current are identified, but experiments at much higher currents on the two mega-ampere machines – RFX at Padua and CPRF at Los Alamos – due to operate in the early 1990s, are needed. A brief account of the compact RFP reactor is given, followed by a summary with an indication of future trends.

The status of controlled nuclear fusion research is reviewed for two major compact toroidal confinement concepts: the field reversed configuration (FRC) and the spheromak. The FRC is an inherently high beta concept offering the advantages of a cylindrical geometry, a natural divertor and the possibility of axial movement of the confined plasma during the formation/burn cycle. The essential techniques of FRC formation, translation and magnetic compression heating have been successfully demonstrated, and gross instabilities driven by plasma rotation have been controlled. Emphasis of present FRC research is on formation symmetry, and on the effects of reducing the large gyroradii of the ions on stability and confinement. The spheromak, like the reversed field pinch, embodies a nearly force-free equilibrium whose evolution is governed significantly by the principle that magnetic helicity is conserved during relaxation processes that minimize magnetic energy. Recent spheromak research has shown how improved edge conditions can enhance global confinement by reducing relaxation and turbulence. Attention to this issue has led to major gains in spheromak energy confinement and plasma beta, and to the discovery of new pressure driven effects.

Recent developments in mirror physics and prospects for mirror based neutron sources and a D-³He mirror reactor are presented.

The paper reviews the status of research on inertial confinement fusion (ICF) with a view to its potential as a power generation source for the future. In recent investigations on implosion physics the necessary conditions for ignition and break-even as well as the required high gain for a fusion reactor have been established quantitatively. Recent progress in the approach to the required confinement parameters and technical achievements have given confidence in the eventual realization of an ICF reactor for energy production.

The fraction of fusion reaction energy that is released in energetic charged ions, such as the alpha particles of the D-T reaction, can be thermalized within the reacting plasma and used to maintain its temperature. This mechanism facilitates the achievement of very high energy multiplication factors Q, but also raises a number of new issues of confinement physics. To ensure satisfactory reactor operation, three areas of energetic ion interaction need to be addressed: (1) single ion transport in imperfectly symmetric magnetic fields or turbulent background plasmas; (2) energetic ion driven (or stabilized) collective phenomena; (3) fusion heat driven collective phenomena. The first of these topics is already being explored in a number of tokamak experiments, and the second will begin to be addressed in the D-T burning phase of TFTR and JET. Exploration of the third topic calls for high-Q operation, which is a goal of proposed next generation plasma burning projects. Planning for future experiments must take into consideration the full range of plasma physics and engineering R&D areas that need to be addressed on the way to fusion power demonstration.

The paper reviews the major engineering components of magnetic confinement devices, i.e. the magnet systems, the vacuum system, fuelling systems, components for heat removal and particle exhaust, and heating and current drive systems. Most technological developments have been achieved for tokamak systems, and the main focus is now on the ITER design actitivities. In addition, some information is given on stellarators and reversed field pinches. The field to be reviewed is very large so that it was not possible to give a comprehensive review. Reference is made to other review papers in the different areas and only specific examples are given for solutions representing the current state of technology.

The paper reviews the present status of the available key technologies for implosion experiments as regards the energy driver, pellet fabrication, diagnostics of dense and hot plasmas, and numerical simulations. It is shown that the development of the necessary conditions is well advanced so that it is now possible to proceed to the next step in the achievement of ignition and break-even. Various design studies for inertial confinement fusion reactors are also reviewed; from these studies it has become clear that inertial confinement fusion and magnetic confinement fusion are technologically complementary systems for future fusion reactors.

The fusion reactor represents the most difficult challenge of any energy system from the viewpoint of developing materials which will allow fusion to be realized as an economic, safe and environmentally acceptable energy source. Substantial progress has been made over the past decade towards an understanding of the effects of the fusion environment on the properties of alloys, in identifying alloy systems which have the greatest potential for development as useful structural alloys, and in formulating and testing a metallurgical approach for developing irradiation resistant alloys with acceptable chemical and mechanical properties. Results to date give confidence that the long term objective – materials that will allow fusion to be an economical, safe and environmentally acceptable energy source – can be achieved. The next experimental fusion reactor, which will be either the International Thermonuclear Experimental Reactor (ITER), the Fusion Experimental Reactor (FER) or the Next European Torus (NET), will be the first device in which the effects of irradiation on the properties of materials will be a critical and limiting factor. At present, fission reactors are the only source of neutrons for the investigation of neutron irradiation damage and the development of radiation resistant materials. Spectral tailoring experiments in mixed spectrum fission reactors in which the neutron spectrum is adjusted to control the balance of transmutation produced helium and atomic displacements provide the closest approximation of fusion irradiation damage that can presently be achieved. These experiments suggest that for ITER conditions the most significant irradiation effects are in the areas of irradiation creep, loss of fracture toughness and possibly irradiation assisted stress corrosion cracking. A major expansion of the experimental programme will be required to develop a database for supporting the engineering design of ITER.

The paper reviews the interaction of plasmas with materials and presents a status summary based on experience in large fusion experiments, laboratory investigations and design studies. The phenomena that are discussed limit the power densities and confinement properties attainable in plasmas; their control is an essential element of the design of future fusion devices.

An overview of international efforts in the area of nuclear engineering and technology is presented in this paper. It includes technological and design developments for the three major tokamak in-vessel components – blanket, first wall and divertor plates – as well as the tritium fuel cycle. Most of the results and analyses are based on the ITER activity and the supporting programmes.

Controlled fusion energy is one of the long term, non-fossil energy sources available to mankind. It has the potential of significant advantages over fission nuclear power in that the consequences of severe accidents are predicted to be less and the radioactive waste burden is calculated to be smaller. Fusion can be an important ingredient in the future world energy mix and can be part of an 'insurance policy' energy strategy to develop new sources as a hedge against environmental, supply or political difficulties connected with the use of fossil fuel and present-day nuclear power. Progress in fusion reactor technology and design is described for both magnetic and inertial fusion energy systems. The projected economic prospects show that fusion will be capital intensive, and the historical trend is towards greater mass utilization efficiency and more competitive costs. Recent studies emphasizing safety and environmental advantages show that the competitive potential of fusion can be further enhanced by specific choices of materials and design. The safety and environmental prospects of fusion appear to exceed substantially those of advanced fission and coal. For example, the level of radioactivity in a low activation fusion reactor at 1 year and at 100 years after shutdown is calculated to be about one-millionth of the radioactivity in a fission reactor of the same power. Likewise, the maximum plausible dose predicted at the site boundary in the case of a low activation fusion reactor is estimated to be between 100 and 500 times smaller than that estimated for a fission power plant. Clearly, a significant and directed technology effort is necessary to achieve these advantages. Typical parameters have been established for magnetic fusion energy reactors, and a tokamak at moderately high magnetic field (about 7 T on axis) in the first regime of MHD stability (β ≤ 3.5 I/aB) is closest to present experimental achievement. Further improvements of the economic and technological performance of the tokamak are possible through the following achievements: higher magnetic fields to lower the required plasma current and reactor size; higher values of the plasma beta, including reaching the second stable MHD regime, to lower the requirements on field and plasma current; and more efficient techniques to drive the plasma current. In addition, alternative, non-tokamak magnetic fusion approaches may offer substantive economic and operational benefits, although at present these concepts must be projected from a less developed physics base. For inertial fusion energy, reactor studies are at an earlier stage, but the essential requirements are a high efficiency (≥ 10%) repetitively pulsed pellet driver capable of delivering up to 10 MJ of energy on target, targets capable of an energy gain (ratio of energy produced to energy on target) of about 100, reactor chambers capable of absorbing the energy released per shot at conditions consistent with power generation, and effective means of isolating the target chamber and driver system.

The paper discusses two main items: (1) technology transfers between fusion research and development (R&D) and other science and technology areas, and (2) the possible utilization of fusion energy in other science and technology areas. The first item is supported by systematic studies of fusion spin-offs carried out by the United States Department of Energy, the Commission of the European Communities, the Japan Atomic Industrial Forum, and the USSR. Fusion R&D has stimulated, and contributed to, the technological development in such mutually distant fields as plasma physics, plasma processing, vacuum technology, technology of high power densities, cryogenic technology, magnets and coil construction, radiofrequency technology and materials science. Continuous training of numerous young scientists and engineers through international fusion activities is also under way. Regarding the second item, a vast amount of applications of fusion energy, in addition to electricity generation, are discussed. Typical fields of applications, excluding military applications, are as follows: burning of nuclear waste of fission products and transuranium isotopes, production of ⁶⁰Co and other isotopes, production of fission fuel and synthetic fuel gas, new materials development and processing, a D-³He space propulsion system and tritium production.

Within the framework of the INTOR activities, a collection of experimental data from major tokamaks in the world was started, to provide the tokamak community with quantitative information on a choice of well documented discharges that can be used as a starting point for further analysis of the physics of tokamak plasmas. The response was excellent. It was therefore decided to publish these data in Nuclear Fusion as a Special Topic, after a check and updating by the contributors. Since the responsibility for the correctness of the data is with the persons who provided them, the data are published under their authorship. It is hoped that collecting and publishing representative tokamak data will become a continuing effort so that, after some time, there will be a data collection that is significantly wider than what can be found in published articles and reports and that can be easily accessed. The data will also be kept in a special database at the Max-Planck-Institut fur Plasmaphysik (IPP), Garching, using the ORACLE system; they can be made available upon request on a data carrier.

(1) The main impurities in FT are oxygen and iron. Iron impurities are negligible at high density (n_Fe < 10⁻⁴ n_e). Radiation losses account for 20-40% of the total input power in all conditions [1, 2].

(2) FT is provided with a complete poloidal limiter made of stainless steel. The same material is used for the vacuum chamber. To a good approximation, the volume inside the outermost closed flux surface equals 2π²R_geoab.

(1) Three types of discharges have been chosen for JET. Within each discharge type, an approximate 'Greek-Latin square' with respect to the variables B, n, P and q has been constructed. The following features have been chosen to characterize the discharge type: material limiter ('limiter') or magnetic limiter ('X-point'); additional heating by RF waves ('RF') or by neutral beams ('NBI'); discharge periods in L-mode or in H-mode; discharge periods with normal sawtooth oscillations or without sawtooth oscillations. The discharge types presented are:

(a) X-point – H-mode – NBI – normal sawteeth;

(b) Limiter – L-mode – NBI – normal sawteeth;

(c) Limiter – L-mode – RF – normal sawteeth.

(1) H-mode discharges under various conditions have been chosen: single-null divertor H-mode or limiter H-mode; H-mode with pellet injection or with gas puff; hydrogen or deuterium plasma; low or high plasma current. All discharges are clear H-modes with a T_e pedestal at the plasma edge.

(1) For shots 18528 and 20345 the limiter is believed to have had a high recycling coefficient; shot 21790 occurred after modest limiter 'conditioning', which produced some reduction in recycling, though much less than for 'supershots'; shot 22871 is a 'supershot'. These plasma regimes have been described previously [1, 2]. The older work does not include the 1987 corrections to the TVTS [2] electron temperatures (and T_e dependent quantities) and so those data cannot be directly compared with the corrected data in the present tables.

(1) Gyrotron set-up: up to 11 gyrotrons in four ports of T-10 (4-3-2-2); low magnetic field-side launch λ₁ = 3.69 mm (f₁ = 81 GHz): six gyrotrons; λ₂ = 4.0 mm (f₂ = 75 GHz): five gyrotrons.