This review intends to illustrate the physics processes that govern the various kinds of catastrophic events in toroidal plasmas. Varieties of abrupt changes of plasma parameters, e.g., disruptions, sawtooth, beta-collapse, etc., have been known in the history of plasma confinement research. They are called collapse events, and considered to be key issues because they might impose the achievable boundary of operation. Recent progresses of experiments (either in the achieved plasma parameters or in the accuracy of measurement) have illuminated the general view, that a sudden onset of the growth of the symmetry-breaking perturbations is the key to understand the collapse events. The growth of deformation does not proceed gradually associated with the variation of the plasma parameters; instead, the growth is “triggered”. In this review, overview of the collapse events is given first. Next it is illustrated that the trigger phenomena and accompanied large-scale deformation are the generic process in collapse events. Theoretical studies to understand the physics of collapse are then explained. It is also stressed that the studies of the physics of collapse events are exploring the frontiers of the modern physics.

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Heliotron J is an L=1 helical-axis heliotron being constructed at Kyoto University, and the first experimental test of the “helical-axis heliotron” line. The design feature of Heliotron J is, as compared with that of Heliotron E, the reduced neoclassical transport and enhanced beta limit with small bootstrap current, which carries the potential for developing the currentless “quasi-isodynamic” configuration in a helical-axis heliotron. To refine its design principle, Heliotron J is planned to offer a high experimental flexibility by using (i) the auxiliary poloidal and toroidal coils to control the bumpiness and the magnetic well and (ii) the existing heating systems such as ECH, NBI and ICRF. The first plama will be produced in FY1999.

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High power negative-ion-based neutral beam injectors for the LHD were constructed and started in operation in September 1998. Two beam lines with a pair of ion sources are installed tangentially in opposing direction. The maximum energy of the beam is 180 keV for hydrogen, and the total injection power is designed to be 15 MW with the pulse length of 10 seconds. One of the specific features of the negative-ion-based NBI system is the requirement on the vacuum pressure in order to avoid the stripping loss of negative ions as well as to realize the optimum gas line density for neutralization, which results in the long beam line. The system was then designed carefully by considering the focal length of the ion sources, neutralizing efficiency, and geometrical and re-ionizing loss of the beam. It is also considered the error magnetic field produced by the ferromagnetic materials that are indispensable for NBI for the magnetic shield over the long neutralizing region.

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Data bases of reversed field pinch (RFP) plasma have been gradually accumulated by recent experiments of several RFP devices. New confinement scalings τ^X(X=RFP_s1)_E=0.024Aa²I_P/P^1/2_heat, τ^X(X=RFP_s2)_E=0.04s(I_N)Aa²I^1.25_P/P^1/2_heat which are consistent to the recent data are presented, where units are in [s], [m], [MA] and [MW] respectively and s(I_N) is a correction function of I_N≡I_P/πa²‹n_e›₂₀). From the standpoint of new scalings, dependences among parameters of possible RFP reactors are analyzed to find the conditions for RFP reactors. H_s1 H_s2 are defined by the ratios of necessary energy confinement time for RFP reactors for burning against τ^X(X=RFP_s1) and τ^X(X=RFP_s2) respectively. When confinement time follows τ^X(X=RFP_s1)_E scaling, confinement enhancement factor of at least H_s1=23 is necessary for RFP reactors to be realistic. When confinement time follows τ^X(X=RFP_s2)_E scaling, data points in I_P-a space of RFP reactors are within the region of target.

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In the CHS heliotron/torsatron two types of MHD instabilities excited by the presence of energetic ions are observed during neutral beam injection heating. One is Fishbone-like burst modes (FBs) and the other is toroidal Alfven eigenmodes (TAEs). FBS of m=3/n=2 (m, n: poloidal and toroidal mode numbers) induce pulsed increase in energetic ion loss flux. FBs of m=2/n=1 often induce the saw-tooth oscillations. TAEs with n=1 and n=2 are excited only when the beam velocity exceeds about half of the central Alfven velocity and the net plasma current is induced by NBI to be in the required range. The observed TAEs are localized in the plasma core region where the magnetic shear is appreciably reduced by the net plasma current.

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Mechanisms for fast/driven magnetic reconnection have been investigated by measuring 2.5-dimensional current sheet structures of laboratory merging plasmas. The reconnection speed is observed to increase with decreasing magnetic field component parallel to the “X-point” line and with increasing external force to compress the current sheet. The measured resistivity of the current sheet is found to increase significantly right after the current sheet is compressed shorter than its ion Larmor radius. The anomalous resistivity of the current sheet caused by the finite Larmor radius effect is concluded essential to the fast/driven-type reconnection whose speed depends strongly on the X-line field and the external force.

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The first results on wall conditioning in the first and second campaign of LHD experiments are described. By confirming device integrity, operational reliability, and environmental safety, LHD has been successfully started up with intensive discharge cleaning without doing high-temperature baking. Evacuation of about 100 molecular layers of surface contaminants is found to be the key to start up the target plasma by ECH for NBI injection. Importance of combination between mild baking, GDC, ECR-DC and main shots is experimentally confirmed.

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When a high magnetic field tokamak reactor is designed with reasonable sizes, the force-reduction technique is necessary. We have proposed Force-Balanced Coils (FBCs) for the purpose of vanishing of Lorentz forces exerted on the conductors and simplifying the construction of coil systems. The force-balanced winding provides not only the toroidal magnetic fields but also the poloidal magnetic flux to induce the plasma current. The net radial force is reduced by a factor of 10 from that exerted on the toroidal magnetic field coils. We have manufactured a small tokamak device with FBCs (R_X(X=major)/a_X(X=minor) 0.3/0.1 m) to demonstrate plasma production and confinement. The toroidal magnetic field of 0.9 T at the vessel center and the plasma current up to 10 kA were achieved by two-step FBC excitation. The plasma column with T_e of 50 eV and n_e of 2×10¹⁹ m^-3 was well centered in the vacuum vessel.

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The technique reviewed in this article can be traced back to the work first demonstrated in 1984 in which an electric field in a glow discharge plasma was measured using laser optogalvanic (LOG) spectroscopy against the Stark splitting of Rydberg levels of helium atoms. Subsequently, the technique of laser-induced fluorescence (LIF) was shown to be possible for this scheme in 1993, and electric fields of more than a few tens of V/mm in helium and argon for pressures above a few hundreds of mTorr have been measured and sheath structures in DC and RF discharges have been studied. Because these limitations have been very restrictive, it has so far failed to become a standard and general technique of electric field measurements in plasmas. However, during the last couple of years, new proposals have been made in which the technique has been extended by using two tunable lasers. This technique looks very promising in providing one of the “dreams” in plasma diagnostics, namely the measurement of electric field of “the order of or less than 1 V/mm”. In this article, the developments of the above laser spectroscopy are first reviewed and the main results of the discharge studies are summarized, New proposals are then described, together with results of computer simulations of LIF signals and the first experimental observation.

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On the superconducting high-field tokamak TRIAM-1M, the high performance plasmas has been demonstrated. A high ion temperature (HIT) mode with T_i2.5 keV was sustained for longer than 70 s in the limiter configuration. In the single null divertor configuration, the ion temperature has been achieved at 5 keV as the results of the improvement of plasma-wall interaction. By improvement of the time response of position control, the single null divertor configuration with κ=1.5 can be also sustained successfully for I min. These high performance plasmas have been obtained using 2.45 GHz lower hybrid current drive (LHCD). The current density profile can be controlled sufficiently longer than the current diffusion time using both 2.45 GHz and 8.2 GHz LHCD. The effects on the plasma wall of high energy neutral produced via charge exchange process has been also made clear in the HIT plasma.

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Final Design Report of the ITER Engineering Design Activities has been approved by the ITER Council with completion of the initially planned six years period. One of the main activities was the implementation of the technology R&D to provide data-base for, and to demonstrate technological feasibility of the ITER design. Most of the initially planned R&D tasks have been performed. In this review paper, we report on the achievements of the ITER technology R&D carried out during the Engineering Design Activities by the Japanese Home Team. The three Parties except US are developing necessary framework for three years extension to implement site adaptation studies as well as residual system tests of the ITER technology R&D.

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