Table of contents

The paper provides a brief introduction to the main aims, overall design philosophy and the planned parameter range of the large tokamak device (major radius R = 2.96 m; horizontal and vertical minor radii, respectively: a= 1.25 m, b = 2.10 m; plasma current, I_p= 4.8 MA), the Joint European Torus (JET), situated on the Culham Laboratory site, UK, whose main objective is to obtain and study plasmas in conditions and with dimensions approaching those needed in a fusion reactor. The main emphasis is on initial operation in the Ohmic-heating phase, in which results are presented covering a wide range of parameters: plasma currents I_p < 3.7 MA; toroidal magnetic fields B_T= 1.3–3.4 T; elongation ratios b/a= 1.2–1.7; and safety factor values, q = 2.3–10. Average electron densities n_e = (1–4) X 10¹⁹m⁻³, with high central electron temperatures (T_e up to 5 keV) and ion temperatures (T_i up to 3 keV) have been achieved, although Z_eff was in the range of 2.4–10. Energy confinement times (τ_E) of up to 0.8 s have been obtained. Some problems with metallic and low-Z impurities are noted, causing high radiation levels. Initial experiments, with ion cyclotron resonance frequency (ICRF) heating of H and ³He minorities in deuterium plasmas at MW levels, are reported. The paper concludes with a description of planned future experiments on impurity control, additional heating (ICRF ≈ 15 MW, and neutral injection ≈ 10 MW), and preparations for tritium operation.

A historical survey is given of the experimental fusion-oriented research programme carried out in the People's Republic of China since 1958. China's research programme covers all major magnetic confinement systems, heating methods and diagnostics as well as laser development for inertial confinement. Experimental results obtained in the different plasma physics institutes in Hefei, Leshan, Beijing,and Shanghai are briefly described.

Twelve years of life of the TFR tokamak are briefly reviewed.

The feasibility of two RF heating schemes, Transit Time Magnetic Pumping and Lower Hybrid heating, has been tested at the Grenoble Laboratory. TTMP results on PETULA were accurately consistent with theory; power absorption was only limited by the unfavourable scaling for small devices. – Launching of lower hybrid waves was made convenient by the grill concept. WEGA and PETULA B experiments showed that ions can be efficiently heated and interaction with electrons leads to non-inductive current drive. Tokamak discharges for which the current is entirely or mainly supported by waves have good confinement and stability properties.

The first few years of exploitation of the TCA tokamak have been mainly devoted to the proof of RF principle studies of Alfvén wave heating. The major findings are briefly described.

ASDEX is a large tokamak (R = 1.65 m, a = 0.4 m, I_p ≤ 500 kA) that started operation in 1980. Its distinctive features are a double- (or single-) null poloidal divertor, the capability for long-pulse operation (up to 10 s) and high-power neutral-beam (4.3 MW), ICRH (3 MW) and LH (2.4 MW) heating systems. Several highly significant experimental findings have been obtained, facilitated by the large flexibility of the machine. The high-recycling divertor regime is considered the most viable solution for handling the large power fluxes envisaged for the next generation of tokamak devices. Studies of impurity transport in the bulk plasma led to the postulation of an additional inward particle drift, while gas puffing experiments revealed the importance of thermoelectric forces along field lines for impurity retainment by the divertor. The high-confinement regime (H-mode) of neutral-beam-heated plasmas, discovered on ASDEX, is also intimately connected with the divertor configuration. The H-mode.confinement, in turn, enabled the investigation of β-limits resulting in the experimental scaling ≅ 2.8 I_p/(a B_T) [%; MA, m, T], in excellent agreement with theoretical predictions. The most important results from recent experiments with RF heating and current drive include the recharging of the OH transformer at constant plasma current, and the observed high heating efficiency obtained with combined second-harmonic ICRH and neutral-injection heating.

The paper reviews the previous Garching Belt Pinch experiments. According to these pulsed, tokamak-type experiments with elongated plasma cross-section, a significant improvement of the maximum beta in a tokamak configuration seems possible. The basic conditions are: increased vertical plasma elongation (e.g. b/a > 1.5) in combination with D-shaping, increased central q-values (e.g. q_o ⪆ 1.5 – 2), a rather flat toroidal current distribution and operation at high poloidal beta (e.g. β_pol ⪆ A/2)

Pulsator was operated by IPP Garching from 1973 to 1979. It made two important contributions to international tokamak research. 1. In Pulsator, for the first time a tokamak plasma was influenced locally by an external, resonant, helical field. The accompanying formation of islands on magnetic surfaces, which affects transport and stability, provided the starting point for experimental investigation of the disruptive instability and its theoretical description. 2. By gas puffing during the discharge it was possible to attain a high-density regime which dramatically raised the then prevailing nτ_E values by almost two orders of magnitude. In this high-density regime an inward directed particle drift velocity and, as a consequence, under certain conditions an accumulation of impurities in the plasma core was found.

The performance of the TEXTOR tokamak, as manifested by its density limit, confinement time, behaviour of density and current density profiles, and impurity level, is described. By means of ICRH heating, powers significantly above the Ohmic one are obtained. The deposition of thin carbon layers onto all inner surfaces is shown to be an effective method for achieving low impurity concentrations in the plasma. Results of a systematic study of plasma-wall interaction, with particular emphasis on a quantitative description of the particle sources and of the boundary layer, are presented. Pump limiter experiments resulting in high particle removal rates are described.

The results obtained in the ohmically heated and RF-heated discharges of the Frascati tokamak FT are reviewed.

The paper summarizes the major experimental results obtained with the tokamak devices JFT-2, JFT-2a/DIVA, JFT-2M and Doublet-III. Reference is also briefly made to major achievements pertaining to the JT-60 programme as well as to fusion reactor design.

A summary is given of the objectives, design features, construction and future programme of the JT-60 project. Experimental results obtained during the first period of operation of JT-60 are briefly described.

In the WT-2 tokamak at Kyoto University,lower hybrid and electron cyclotron waves have been applied separately or in combination to initiate or to maintain a toroidal plasma current, and a tokamak plasma was formed by RF only without OH power. Electron cyclotron waves have also been applied to suppress an anisotropy-driven instability that appears in plasmas with LH-driven current.

The contribution gives a brief summary of experimental research carried out at Tokyo University on TNT-A, a small non-circular tokamak studied in the Department of Physics.

The contribution gives a brief summary on research carried out at Tokyo University on low-q operation in the TORIUT experiments in the Department of Nuclear Engineering.

The major experimental results on low-q, high-beta plasmas obtained in the circular and non-circular versions of SPICA and its predecessors are summarized.

The achievements of the CLEO and DITE tokamaks since 1972 are reviewed. The areas covered include: bundle divertor physics, neutral-beam injection heating and current drive, charge-exchange spectroscopy, scaling of confinement, beta and density limits, plasma-surface interactions and diagnostic development.

The TOSCA device was built to investigate non-circular plasma cross-sections, minor radius compression and limiting values of beta. A wide range of investigations on minor radius and minor/major radius compression have been conducted which show improved confinement on detachment from the limiter. The shaping studies have explored elliptic, D-shaped and triangular plasmas. Detailed comparisons between theory and experiment for positional instabilities in a variety of shapes have been made. The investigations also demonstrated the failure of shaping to significantly decrease the MHD activity on a tokamak. Investigations of limiting beta both ohmically and using powerful electron cyclotron resonant heating have shown that the critical beta is proportional to the plasma current and reaches values very close to the predicted theoretical limits. A variety of studies on MHD instabilities have been made and attempts to control them using small resonant helical fields. Basic investigations of the influence of fluctuations on confinement have been made by a variety of techniques.

The Torus II high-beta tokamak has studied plasmas with ⟨β⟩ in the range 9–13%. Both stable and unstable equilibria were observed; in the case of unstable equilibria, the instability had the characteristics of high-toroidal-mode-number ballooning. The experimental equilibria were modelled computationally and tested for stability. Predicted growth rates and critical mode numbers are in reasonable agreement with observation.

GA Technologies Inc. (formerly General Atomic Company), as one of the oldest industrial participants in fusion research, has had a most profound influence on worldwide fusion progress. The principal theme, spanning nearly three decades of research, has been the tailoring of the magnetic geometry in toroidal configurations to improve plasma stability and confinement properties. Multipole configurations, pioneered at GA, gave some of the earliest evidence that confinement need not be limited by Bohm scaling. The Doublet programme, a direct outgrowth of the multipole work, was the beginning of an international effort, still bearing fruit, to maximize the achievable values of β in thetokamak. OHTE, a helically symmetric toroidal pinch with pitch reversal, is another GA concept, developed recently but harking back to very early work, that has already produced an extensive and promising database.

The major results and accomplishments of the MIT tokamak programme are surveyed. These are considered to be 1) discovery of an Ohmic-heating confinement law in which τ_E ∝ ;aR²; 2) reduction of anomalous ion conduction to the neoclassical value by use of pellet fuelling; 3) formulation of an empirical model for confinement of impurities in ohmically heated tokamaks; 4) seminal experiments on current drive by lower hybrid waves and production of quasi-stationary driven current discharges with n ≈10²⁰ m^–3; and 5) heating of electrons by Landau damping of lower hybrid waves with ΔT_e≈1 keV. The advance of n₀τ_E is also traced from values of about 10^l8 s·m⁻³ which were typical of tokamaks at the beginning of the Alcator programme to values achieved on Alcator C in excess of 6 X 10¹⁹ s·m⁻³, which is required for thermalized energy breakeven at higher temperature.

Results obtained with the ORMAK, ISX-A, and ISX-B tokamaks at Oak Ridge National Laboratory (1972-1984) are summarized. Topics considered are plasma heating, confinement, MHD studies, impurity effects, pellet fuelling, and limiters and surface physics.

The Adiabatic Toroidal Compressor (ATC) successfully demonstrated major-radius compression of Ohmic and auxiliary heated tokamak plasmas. Plasma densities well above 10¹⁴ cm⁻³ were reached, together with central ion and electron temperatures above 1 keV.

A review of high-power neutral-beam heating, impurity studies, and high-beta research on the PDX/PBX tokamak is presented.

In the PLT experimental programme, Ohmic, neutral-injection, and ion-cyclotron heating techniques have been studied and improved, resulting in record ion temperatures. In addition, non-inductive current drive with lower-hybrid waves has been demonstrated. These achievements were made possible by the continual development and deployment of advanced plasma diagnostics, high-power RF systems, and specially tailored materials and structures. The intermediate size of PLT has permitted flexible experimental operation and the separation of plasma core from plasma surface effects.

The confinement time on the ST Tokamak was proportional to the plasma current and density and reached values of 10–20 ms near the disruption limit. The confinement showed the effect of neoclassical, single-particle behaviour for non-axisymmetry. The Spitzer conductivity was found by considering Z_eff from measured impurities. Runaway electron production was classical, the cross-field transport anomalous. The MHD modes, internal (sawtooth) and major disruptions, and island structures for the m = 2 modes were analysed by soft-X-ray imaging techniques. Ion-cyclotron heating at second-harmonic operation increased the plasma temperature by 100 eV.

The Tokamak Fusion Test Reactor (TFTR) is intended to achieve approximate energy breakeven in D-T plasmas. Construction approval was received in March 1976, and the first plasma was produced in December 1982. Three major experimental run periods, the last ending April 1985, have yielded experimental results on confinement and heating that extend the scaling laws of smaller machines. The plasma parameters in TFTR are now approaching those required for breakeven in unthermalized, two-component regimes. They span the range from high-density operation with n_e(0)τ_E ≈ 4 × 10¹⁹ m⁻³·s to a low-density regime with T_i(0) ≈ (9±2)keV. The maximum product n_e(0)τ_E T_i(0) of about 9 × 10¹⁹m⁻³·s·keV was obtained at a central density, 1.1 × 10²⁰ m⁻³, by means of pellet injection.

The fusion research result in Macrotor has been focused on the basic issues of impurity generation (O, C), the role of the radial electric field in particle and energy transport, magnetic turbulence, particle pumping with a limiter, and the development of a large-area ICRF coupler for phased arrays for current drive and plasma heating. The total impurity line radiation has been reduced below the value contributed by the hydrogen edge radiation in ohmically heated plasmas. In RF-heated discharges, edge electron heating and the corresponding influx of impurities have been reduced by k-parallel shaping using large-area phased ICRF couplers. The Faraday shields have been used as limiters, and are used to remove the plasma heat. Current drive due to the lower hybrid fast wave has been observed with these couplers.

A historical survey is given of the experimental multipole and tokamak fusion research programme carried out at the University of Wisconsin since 1962. The programme has concentrated on axisymmetric toroidal confinement through the utilization of a series of different multipole devices and of a poloidal divertor tokamak. The gross plasma stability and good confinement properties of the multipoles led to a number of experimental results not easily achievable in other toroidal devices. The paper describes the major results obtained in multipoles with and without Ohmic heating current and toroidal field, and their influence on the development of plasma theory. In the poloidal divertor tokamak, stable discharges have been found at very low values of the safety factor.

The paper presents a brief survey of experiments on high-frequency plasma heating in the FT-1 and FT-2 tokamaks and of experiments in tokamaks with plasma compression: Tuman-2, Tuman-2A and Tuman-3.

25 years of tokamak research at the I.V. Kurchatov Institute in Moscow, spanning the evolution from T-l to T-15, are reviewed.

The paper describes a series of experiments carried out at Nagoya during the past 18 years to investigate the viability of RF plugging of cusp or mirror-cusp configurations for future reactor applications. An important result of these investigations is the reduction of end losses obtained by RF plugs in the ion cyclotron frequency range.

A survey is given of the experimental tandem mirror results obtained at the University of Tsukuba since 1978. The first tandem mirror, GAMMA 6, demonstrated the tandem mirror potential configuration and the validity of the Pastukhov confinement scaling law. GAMMA 10 demonstrated the plug/thermal-barrier potential distribution in axisymmetric end mirrors. The Pastukhov scaling for axial confinement was confirmed and improved radial confinement was observed, indicating the benefit of an axisymmetrized magnetic-field configuration.

The paper summarizes results from the-TMX and TMX-U tandem mirror experiments at the Lawrence Livermore National Laboratory.

The Phaedrus Tandem Mirror Programme has demonstrated that RF alone can provide ion and electron heating, MHD stability, start-up, fuelling, plasma potential control, and enhanced ion confinement. Major results have been the demonstration of RF-sustained operation, the determination of MHD stability limits and the demonstration of axisymmetric operation with stability provided by radial ponderomotive force.

A survey of Soviet research on open traps over the past 25 years is given.

The experimental High-Beta Stellarator programme at Garching (1969-1978) is reviewed, emphasis being placed on the key results. These results may be useful as benchmarks for some of the new stellarator concepts under discussion at present.

Experiments on plasma confinement in stellarators started at the Max-Planck-Institute of Physics and Astrophysics immediately after the 1958 Geneva Conference and were continued at IPP Garching. The first devices of the Wendelstein line were W I–A and W I–B, both racetrack-type stellarators. In the circular ℓ= 2 stellarator W II–A, steady-state barium plasmas were investigated and were found to be governed by collisional diffusion. In W II–B, mainly Ohmic heating was applied. The main stellarator experiment at IPP Garching is W VII–A; after a period of Ohmic heating, the first net-current-free plasma was achieved in 1980. Plasma parameters of up to T_i = 1 keV and n(0) = 10¹⁴ cm⁻³ could be reached by this method. A further heating method is ECRH, which yields maximum electron temperatures of T_e ≥ 2 keV. The paper describes the major results of the devices mentioned and gives a short survey of further planning.

The experimental activities of the Plasma Physics Laboratory at Kyoto University are reported. The laboratory has been conducting the Heliotron project which is a research project on plasma confinement and heating comprising the heliotron machines Heliotron A, B, C, D, DM, DR and E. The heliotron field is one of the basic plasma-confining field configurations investigated since 1959 in the laboratory. In the paper, the results of experimental studies on the Heliotron D, DM and E machines are reported.

A survey is made of experiments in different versions of the Nagoya Bumpy Torus (NBT). Hot-electron-plasmas with β-values up to 5% were routinely produced by electron-cyclotron heating. Plasma confinement studies of the toroidal plasma have revealed that the confinement times fall short of neoclassical predictions, but global stability can be achieved. A density of 10¹³cm⁻³ and an ion temperature of 100 eV have been obtained simultaneously on NBT-1M by the use of hot-electron rings in combination with ICRF heating. The electrons are still in the collisional regime.

Research activities on the JIPP Ia and Ib stellarators, the JIPP T-IIU stellarator/tokamak and the JIPP T-IIU tokamak are surveyed. In the ℓ = 3 JIPP Ia stellarator, convective transport and enhanced fluctuations were found to be responsible for the confinement behaviour. The dependence of plasma transport on the magnetic field configuration was studied on the ℓ = 2 + ℓ = 3 JIPP Ib stellarator. In the tokamak operation of JIPP T-II, a stable high-density plasma was obtained, free from major disruptions by programmed gas puffing and a second plasma current rise. Scaling laws fortthe ohmically heated ℓ = 2 stellarator plasma were obtained and compared to those for tokamaks. Major disruptions were suppressed when ι/2π>0.14. Heating characteristics of NBI, LHH, ECH and ICH were compared in both configurations. On JIPP T-IIU, ICRF heating experiments were carried out. The radial profile control of plasma parameters by programming gas puffing and plasma current waveform was shown to be effective in reducing the impurities and suppressing the growth of an m = 2 mode. Ion Bernstein wave heating was found to be promising. A current start-up experiment is also carried out on JIPP T-IIU, giving a physical picture of the start-up and the return current heating of the bulk plasma.

The stellarator programme at Culham is reviewed. Results from the CLASP, PROTO-CLEO, TWIST, TORSO and CLEO devices are presented. The CLASP experiment showed that stellarator fields have very good potential for confining singly charged particles. Results from PROTO-CLEO showed that stellarator confinement could be better than that given by the Bohm formula; ohmically heated plasmas on CLEO showed that parameters similar to those of tokamaks could be obtained. Neutral injection into CLEO successfully heated the plasma to beta values close to the stability limit. The first ECRH results on CLEO indicated the significant potential that stellarators posses for currentless operation.

Theta-pinch devices have produced fusion plasmas of high beta, temperature and density, with the first observation of thermonuclear neutrons recorded in 1958. Linear theta-pinch plasmas are neutrally stable and display good confinement properties before the onset of end effects. An end-shortening-induced instability, and axial particle and thermal losses dominate these plasmas. In Scylla IV-P, the plasma stability was enhanced and axial particle losses were eliminated by the use of material end plugs, but thermal losses remained unabated. Scyllac was a large-aspect-ratio, toroidal theta-pinch proposed to circumvent the end-loss problem. The high-beta-stellarator plasma produced in Scyllac was found to be unstable to a poloidal M 1 mode (sideways displacement). Experiments demonstrated feedback stabilization of the mode. Although experimental lifetimes were too short for transport issues to be addressed, Scyllac did demonstrate the existence of stabilized high-beta stellarator equilibria.

The ELMO Bumpy Torus experiment was designed to test the idea of stabilizing a multiple mirror plasma by the use of energetic electron rings, while eliminating end losses by closing the device into a circular toroid. The experiment did provide a quiescent mode of operation; however, there were major radial losses of particles and energy. These were attributed to effects associated with the uniformly distributed toroidal curvature; an improved device, the ELMO Bumpy Square, was proposed to eliminate this problem.

The Model C Stellarator, constructed and operated at the Plasma Physics Laboratory of the Princeton University, was the last and most powerful of the stellarator devices operated at Princeton during the period from 1953 to 1969. Experiments on Model C treated the problem of anomalous plasma loss and tested major innovative concepts including the divertor and radiofrequency heating. This review focuses on portions of Model C work that have strongly influenced present-day research and understanding of toroidal confinement in stellarators and tokamaks.

Experimental investigations on the three stellarators at the University of Wisconsin are summarized. Heating experiments on the Proto-Cleo stellarator included Alfvén wave heating, ion cyclotron heating and inductive heating. The plasma injection process was also investigated and measurements of Pfirsch-Schlüter and bootstrap currents were performed. Measurements on the Proto-Cleo torsatron focused on fluctuation studies, confinement characteristics and on examination of the runaway instability. On the newest device, the interchangeable modular stellarator, studies have concentrated on magnetic-surface mapping, modular divertors and electron cyclotron heating.

The main theoretical and experimental results obtained at the Plasma Physics Laboratory of the General Physics Institute of the Academy of Sciences of the USSR during the period from 1960 to 1984 are briefly surveyed.

The Padua Reversed-Field Pinch research is reviewed with reference to three toroidal devices: ETA-BETA I which operated during 1974–1977, ETA-BETA II which is currently operating, and RFX which is being constructed. The paper summarizes in particular studies on impurity radiation losses, plasma density limits, and the properties of the mean and fluctuating magnetic field. The range of I/N values, limited by excessive radiation at high density and by excessive electron streaming and resistance anomaly at low density, is identified. Plasma confinement is discussed in comparison with the classical ion value and studies on magnetic field fluctuations indicate as a possible underlying process the diffusion in stochastic magnetic fields created by resistive MHD instabilities. The dependence of the plasma parameters on current and size is briefly discussed, together with the main motivations of future RFP experiments such as RFX.

The paper summarizes fourteen years of toroidal pinch research in the Electrotechnical Laboratory at Ibaraki. The experimental programme started using a screw pinch configuration and continued with a series of reversed-field pinches in quartz and metal tori. The experimental results confirm the existence of, and the possibility to maintain, a quiescent period in the reversed-field pinches. High temperatures (up to 600 eV) found to be proportional to the plasma current have been obtained and maintained, and scaling laws for the plasma resistance and for the energy and particle confinement times have been established.

Experiments carried out at Nagoya since 1973 have shown that the injection of relativistic electron beams (REB) into toroidal devices can lead to a variety of magnetic field configurations having q-values ranging from q = 0 to q = 7. Among them, the REB ring-core configuration is a spherator-like configuration characterized by the presence of a REB and a currentless plasma confinement region outside the core.

Research on the symmetric toroidal pinch (STP) has been carried out at Nagoya during the past 15 years. From 1970 to 1982, this research concentrated on heating and confinement of screw pinch plasmas with force-free currents. Reversed-field experiments in metal tori started in 1983. The main object of these experiments is to study transport processes in reversed-field pinches with high current density (5–10 MA· m⁻²) At present, plasma currents up to 15 MA (6.2 MA· m⁻²) can be driven during 1.6 ms. Toroidal flux enhancement and related phenomena are being investigated.

The contribution gives a brief summary of experimental research carried out at Tokyo University on the reversed-field pinch REPUTE-1, a joint project of the Department of Nuclear Engineering and the Department of Physics.

The foundations of present-day RFP research are to be found in the early results from ZETA in 1967, in which a period of improved stability was observed in discharges with a spontaneously generated reversed field, during which the magnetic field configuration was in accord with ideal-MHD stability requirements. Theoretical work at Culham subsequently established that the reversed-field configuration is a minimum magnetic energy state to which a plasma has a strong propensity to relax naturally. – HBTX1, designed to study fastprogrammed RFPs in a quartz torus, became operational during 1972; spontaneous field reversal was again demonstrated. By operating with current risetimes shorter or longer than the natural relaxation time, both reversal by strong helical instability and relaxation to reversed field more gently along a minimum energy trajectory in F-θ space were demonstrated. However, although much important information was obtained from HBTX1, including checking the predictions of relaxation and resistive instability theory, high temperatures could not be sustained,and ZETA-like quiescence was not found; this was attributed to the use of an insulating vessel. – A new assembly, HBTX1A, with a stainless-steel liner, was commissioned in 1981. By using a steady vertical field to assist in providing equilibrium, plasma lifetimes of up to 15 ms were observed with electron temperatures of up to 350 eV (at 200 kA) at densities of (1–2) × 10¹⁹ m⁻³. Magnetic field fluctuations with /B ≥ 1–2%, seen throughout the discharge, are dominated by poloidal m = 0 and m = 1 global modes of resistive MHD origin, and localized activity in the core at higher frequencies. The m = 1 modes can contribute to anomalous energy transport, and to reversed-field penetration. The value of β_θ≈ 10% varies relatively little with conditions in HBTX1 A, which is consistent with theoretical predictions of transport due to g-modes; if β_θ remains constant at about 10% as the current and machine size are increased reactor conditions will be reached.

Experiments and theory at Los Alamos have contributed to advances and increased understanding of spheromak physics. Application of the relaxation principle and the concept of helicity injection has led to new, improved formation methods and to the ability to sustain spheromaks for long times against resistive decay. Use of oblate flux conservers has provided gross stability of the spheromak, even in the presence of bias magnetic fields. Magnetic diagnostics have seen oscillations caused by rotating non-resonant internal kink modes. The stability thresholds of these modes agree with the measured equilibrium of the spheromak, confirming that those equilibria depart significantly from the minimum-energy state. Reduction of impurities and use of background filling gas have created resistively decaying spheromaks with non-radiation-dominated confinement.

Exploratory field-reversed-configuration (FRC) experiments, initiated at Los Alamos in the midseventies, demonstrated FRC lifetimes substantially longer than predicted from MHD stability theory. Subsequent experimental and theoretical advances have provided considerable understanding of FRC stability physics, the characteristics of the configuration loss processes, and the particle confinement scaling with size. The critical FRC physics issues, which directly relate to the development of an FRC fusion reactor and need to be addressed in a new generation of experiments, have been clearly identified.

The development of the present Los Alamos reversed-field pinch experimental programme from its early Z-pinch beginnings is reviewed.

The S-l device inductively generates spheromak plasmas with major radii of 40 to 65 cm and toroidal plasma currents up to 500 kA. The major objective is the investigation of MHD stability and transport characteristics of spheromaks that are stabilized in externally applied magnetic fields by loose-fitting conductors and coils.

The laser-matter work done at Centre d'Etudes de Limeil-Valenton during the past twenty-two years is summarized. The main physics results on interaction, particle transport and hydrodynamics are underlined. The development of high-power lasers is described.

The main line of laser plasma research at Garching is described. It includes the development of a terawatt iodine laser, the first description of phase conjugation by stimulated Brillouin backscattering, the investigation of absorption and energy transport mechanisms and theoretical work on the hydrodynamics of laser plasmas and ICF fusion pellets.

The Institute of Laser Engineering, Osaka University, has performed inertial fusion experiments using various kinds of energy drivers, such as the GEKKO series of glass lasers, the LEKKO series of CO₂ lasers, and the REIDEN series of light ion beams. – As for the fuel pellet design, the well-known Cannonball target was invented to warrant a uniform compression under laser irradiation. The super-computer SX-2 was introduced to pursue the implosion physics and to optimize the pellet design. – A conceptual scheme for a laser fusion reactor was also investigated. For inertial confinement fusion reactors the SENRI series was proposed.

The paper presents highlights of basic research work related to laser fusion carried out by United Kingdom universities at the SERC Central Laser Facility since its establishment in 1976.

Los Alamos has constructed a high-power CO₂ laser and is completing a thorough technical evaluation of the CO₂ laser as an inertial confinement fusion (ICF) driver. Recent experimental work has shown unambiguously that 0.25 μm is near the wavelength for optimum coupling of laser radiation into a fusion target. Since KrF lases at 0.248 μm, has the potential for 10% wall plug efficiency and for low capital cost, Los Alamos is continuing to investigate the feasibility of KrF as a future ICF driver.

Recent experiments at the Lawrence Livermore National Laboratory using the 10 TW Novette laser demonstrated increased absorption and conversion to X-rays and decreased production of suprathermal electrons with short-wavelength (<0.5 μm) light. Stimulated Raman scattering was identified as the primary source of suprathermal electrons. The collisionality of the laser/target coupling can be controlled by the proper selection of laser wavelength and target material. This very favourable laser/target coupling has led to achieving high inertial fusion target compressions and to the unambiguous demonstration of the first laboratory X-ray laser.

The paper reviews the contributions of the Laboratory for Laser Energetics of the University of Rochester to inertial fusion research, including investigation of the direct-drive approach to laser fusion, the development of short-wavelength laser systems and the development of high-density diagnostic techniques.

Inertial confinement fusion requires the generation and focusing of several megajoules of energy at > 100 TW power and > 100 TW·cm⁻² power density onto a target for approximately 10 ns. Lasers and particle beam drivers have been developed for this purpose. Lightion beams offer the potential for a cost-effective, efficient and versatile driver with excellent energy deposition and no significant preheat. The research and development to date has emphasized technology development of the driver. Advances in pulsed power technology, magnetically insulated power flow, intense ion beam generation, focusing and transport, and ion beam deposition have led to the Particle Beam Fusion Accelerator II at Sandia National Laboratories which will begin operation in January 1986. This accelerator has the potential for achieving ignition of thermonuclear fuel in the laboratory.