Table of contents

A soft X-ray imaging system consisting of three arrays of silicon surface barrier diodes is applied to the tomographic analysis of internal plasma perturbations in the T-l0 tokamak (R₀=1.5 m, a=0.32 m). It is found that density limit disruptions in a plasma with a high safety factor at the edge (q_a=3.5-4.5) are associated with joint rotation of the m=2, n=1 and m=1, n=1 modes, overlapping at the energy quench stage. Electron cyclotron resonance heating is used to prevent density limit disruptions or to recover stable operation of the discharge after the energy quench at the density limit

During pellet injection in the TORE SUPRA tokamak, two distinct phases of electron temperature response, a gradual decrease phase followed by a sudden drop phase, have been observed by means of a fast acquisition ECE Fabry-Perot interferometer system. The time delay of the temperature drop between the plasma edge and centre is less than 20 mu s, corresponding to a propagation velocity of the order of 25 km/s, much larger than both the pellet velocity and the ordinary diffusion velocity. A neutral atom diffusion model, together with a modified plasmoid ablatant description, is proposed to explain such phenomena

A new method for obtaining a transient ('pulse') electron heat diffusivity (χ_e^p) in the radial region 0.38 < r/a < 0.56 in TFTR L mode discharges is presented. Small electron temperature perturbations were caused by single bursts of injected impurities which radiated and cooled the plasma edge. A case of iron injection by laser ablation was found to be more definitive than a supporting helium gas puff case. In this new 'cold pulse' method, the authors concentrate on modelling just the electron temperature perturbations, tracked with electron cyclotron emission diagnostics, and on being able to justify separation of the perturbations in space and time from the cooling source. This χ_e^p is obtained for these two cases to be χ_e^p = (6.0 m²/s ± 35%) ~ 4χ_e (power balance), which is consistent with but more definitive than results from other studies that are more susceptible to ambiguities in the source profile

In simulations of lower hybrid heating and current drive in tokamaks, an important part of the calculation is the determination of ray paths from the antenna to the central region of the plasma where they are absorbed. The role of the parallel refractive index spectrum and the need to launch a set of rays that cover it adequately are well known. However, the antenna also has a finite poloidal extent and a corresponding poloidal wavenumber spectrum. Here a method of estimating the effect of this poloidal structure on the spatial and spectral widths of the beam is discussed. The objective is to take this into account without the need to launch a set of rays covering the range of poloidal positions and wavenumbers as well as that of parallel wavenumbers

This paper describes an investigation of the three dimensional (3-D) magnetic fields produced by currents flowing in plasmas that are confined in stellarators and in tokamaks with magnetic field ripples. A simple semianalytical procedure is proposed and a linear 3-D equation is derived that makes it possible to investigate the influence of finite pressure effects and plasma currents on the ripples of the longitudinal magnetic field. Analytical estimations are presented that can be useful for the analysis of the transport properties of finite beta plasmas. It is shown that the accuracy of the experimental identification of the magnetohydrodynamic (MHD) equilibria can be improved using external measurements of the 3-D magnetic fields. This method appears to have considerable promise for the investigation of fairly weak effects, as for example, for distinguishing bootstrap current profiles in stellarators. Analytical expressions for the 3-D external magnetic fields produced by plasma currents are derived that can be used for a rough interpretation of the experimental data

Many different models for energy transport in a tokamak have been tested against subsets of the wealth of data accumulated from tokamak experiments and claimed to describe adequately the data used in the tests. Yet no single transport model has emerged as an obvious candidate for accurate extrapolations from today's large tokamak to a reactor. The principles involved in testing a model against data are examined. It is shown how: (i) collinearities in the data, (ii) confinement degradation with power and (iii) measurement errors in the data, can make tests insensitive to particular details of transport models. This lack of sensitivity is illustrated via tests of models against global confinement data from several tokamaks (ITER L mode database) and via tests of different models against local confinement data from one single tokamak (JET database). The latter tests demonstrate that the testing of transport models must include data from two or more tokamaks of different size

The Large Helical Device (LHD) now under construction is a heliotron/torsatron device with a closed divertor system. The edge LHD magnetic structure has been studied in detail. A peculiar feature of the configuration is the existence of edge surface layers, a complicated three dimensional magnetic structure which does not, however, seem to hamper the expected divertor functions. Two divertor operational modes are being considered for the LHD experiment-high density, cold radiative divertor operation as a safe heat removal scheme and high temperature divertor plasma operation. In the latter operation, a divertor plasma with a temperature of a few keV, generated by efficient pumping, is expected to lead to a significant improvement in core plasma confinement. Conceptual designs of the LHD divertor components are under way

Active feedback stabilization of the vertical instability is studied for highly elongated tokamak plasmas (2 ⩽ κ ⩽ 3) and evaluated in particular for the TCV configuration. It is shown that the feedback can strongly affect the form of the eigenfunction for these highly elongated equilibria, and this can have detrimental effects on the ability of the feedback system to detect properly and stabilize the plasma. A calculation of the vertical displacement that uses poloidal flux measurements, poloidal magnetic field measurements, and corrections for the vessel eddy currents and active feedback currents was found to be effective even in the cases with the worst deformations of the eigenfunction. These deformations are also examined to determine how they affect equilibria with various degrees of shaping, and it is seen that the magnitude of the deformation of the eigenfunction is a strong function of the plasma elongation

A theory is formulated to describe the dynamics of the high density, partially ionized region of the ablation cloud in the vicinity of a spherical light-atom refractory pellet exposed to a tokamak plasma. Owing to the finite sublimation energy of refractory pellets, the surface boundary conditions require a more detailed treatment than in the case of frozen hydrogen. With the exception of conditions that lead to relatively cold ablation clouds (low plasma density, or small pellet radius), transonic flow solutions were found to exist for most parameters of interest. The shielding properties of the ablation cloud, degree of ionization at the M=1 surface, and the surface flow Mach number were found to be in fairly good agreement with the heuristic model of Parks, Leffler and Fisher

Experiments have been performed on the high aspect ratio, high current density Extrap-T1 reversed field pinch to study the response in confinement properties to variations in plasma current and pinch parameter. The study includes measurements of energy and particle confinement times as well as radiated power, impurity concentrations and magnetic fluctuations. The ratio of Spitzer to total input power in these experiments is varied over the wide range 0.4 to 0.8. The operational β₀ value is found to be primarily a function of plasma current and pinch parameter and scales only weakly with density since an increase in density is found to be accompanied by a decrease in temperature. Changes in T_e/n_e, however, directly affect the dynamo activity. At high T_e/n_e, the scaling of the energy confinement time is governed by the enhancement of the dynamo activity, which offsets the decrease in Spitzer input power with temperature. On the other hand, the particle confinement time is not deteriorated by enhanced dynamo activity. Instead, particle confinement is degraded at a high fraction of Spitzer input power. In this limit, convection of thermal particles can account for a major part of the total energy loss

Coil placement errors break the axisymmetry of tokamak magnetic fields. The non-axisymmetries cause magnetic islands, which can enhance locked modes, decrease confinement and lead to disruptions. An algorithm is presented for calculating the currents in small correction coils so that specified islands caused by known magnetic field errors will be eliminated. Surface of section plots indicate that the calculated correction coil currents will substantially eliminate islands. A slight modification of the algorithm allows one to calculate correction coil currents which give rise to isolated islands of prescribed size and toroidal phase. Changing the toroidal phase would give rise to a rotating island, and rotating magnetic islands may be useful for plasma edge control

Toroidal momentum transport is examined experimentally by using on- and off-axis tangential neutral beam injections on the JT-60U tokamak. From a steady state momentum balance analysis-on the assumption that momentum flux is diffusive-it is found that the profiles of the momentum diffusivity are quite different in the two cases of on- and off-axis beam injections. In addition, transient toroidal momentum transport was examined by using a momentum source modulation experiment. On the assumption that the toroidal momentum flux consists of a diffusive and a convective flow term, it is found that there is non-diffusive inward flux of toroidal momentum whose absolute value is comparable to that of the diffusive flux

The presence of an extended region of open flux surfaces (halo), during the current quench phase of the disruption of elongated plasmas, is supported by measurements of halo currents and by numerical simulations. The halo, in addition to providing a poloidal current path between the plasma and the first-wall components, allows rapid conduction and convection of energy along field lines, and therefore a mechanism for the localized deposition of energy onto the wall. The heat load to the region of the plasma-first-wall interaction is higher than in the scenario in which the magnetic energy is mostly dissipated by radiative processes

It is shown that interchange type plasma instabilities can be stabilized by inverted flow velocity profiles. The stabilization criterion is that the gradient in flow velocity be opposite to that of the plasma density and the velocity gradient scale length be sufficiently smaller than that of the density. This technique is applied to ballooning modes and curvature driven trapped particle instabilities in tokamaks. The necessary inversion of the toroidal flow velocity profile can he achieved by appropriate injection of neutral beams