Table of contents

Internal transport barriers have been observed in several tokamak operating regimes and the reduced transport has been shown to be qualitatively consistent with E*B rotational shear suppression of turbulence. The interaction of the toroidal rotation with the diamagnetic and poloidal rotation contributions of the electric field is explored. Toroidal momentum injection counter to the current is predicted to lower the power threshold for improved confinement if the thermal ion density is peaked on-axis. Peaking of the density or power deposition profile near the magnetic axis is also found to lower the power threshold, in agreement with experimental trends

An escaping alpha collector probe has been developed for the deuterium-tritium (DT) phase of the Tokamak Fusion Test Reactor (TFTR) to complement the results of the lost alpha scintillator detectors, which have been operating on TFTR since 1988. Measurements of the energy distribution of escaping alphas have been made by measuring the range of alphas implanted into nickel foils located within the alpha collector. Exposed samples have been analysed for four DT plasma discharges at plasma currents of 1.0 and 1.8 MA. The results at 1.0 MA are in good agreement with predictions for first orbit alpha loss at 3.5 MeV. The 1.8 MA results, however, indicate a large anomalous loss of partially thermalized alphas at an energy ~30% below the birth energy and at a total fluence nearly an order of magnitude above the expected first orbit loss. This anomalous loss is not observed with the lost alpha scintillator detectors in DT plasmas but does resemble the anomalous `delayed' loss seen in DD plasmas. Several potential explanations for this loss process are examined. None of the candidate explanations proposed thus far are fully consistent with the anomalous loss observations

Deuterium gas injected into ELMing H mode divertor discharges in the DIII-D tokamak typically reduced the total power at the divertor target ~2 times and the peak heat flux ~3 to 5 times with modest (<10%) degradation in plasma energy confinement. The parameter range for the discharges investigated was: I_p=1.0-2.0 MA, q₉₅ approximately= 2.4-6.0 and total input power (≲20 MW. Most of this reduction in heat flux occurred at the sudden formation of a high density, highly radiating region located between the outboard divertor separatrix strike point and the X point. This divertor behaviour is associated with a `partially detached' divertor plasma condition, which is referred to in this paper as the partially detached divertor (PDD) regime. With the onset of the PDD, typically at a line averaged density of 0.6 to 0.7 times the Greenwald density limit, an abrupt reduction in plasma electron pressure (≳4 times) was observed at the outboard divertor separatrix strike point; at the same time, however, only a modest (≲30%) change in the electron pressure was observed upstream near the outboard midplane separatrix. The data suggest that significant plasma momentum loss occurred between the high density, highly radiative region and the (downstream) divertor separatrix target. Plasma performance showed little degradation with the onset of the PDD regime. Deuterium injection made only modest changes in the temperature and density profile shapes near the midplane separatrix of the main plasma. The PDD approach is shown to be compatible with discharges operating at low safety factor (i.e. q₉₅ equivalent to 2.9) and to be effective in significantly reducing toroidal asymmetry in observed divertor plasma properties (e.g., heat flux). The potential for operating in a steady state has been demonstrated using feedback control of the neutral pressure outside the main plasma

Experiments on m=2, n=1 tearing mode suppression and on avoidance of density limit disruptions by electron cyclotron resonance heating (ECRH) were performed on the T-10 tokamak. Partial suppression of the m=2, n=1 mode by the high frequency (HF) power deposition in the vicinity of the q=2 surface was observed. Development of external kink modes with HF power injection can result in m=2, n=1 mode destabilization under specific operating conditions. ECRH suppresses m=2, n=1 mode activity at extremely high values of electron densities and prevents the density limit disruptions practically independently of EC resonance position. Complete compensation of the additional peripheral heat losses near the density limit by ECRH should be responsible for this result. No effect of electron cyclotron current drive (ECCD) on m=2, n=1 mode stability has been observed because of insufficient values of HF driven current in the vicinity of the q=2 surface under the operating conditions of the experiment

The spectrum of n=0 Alfven modes is calculated analytically and numerically in cylindrical and toroidal geometries. It includes global Alfven eigenmodes (GAEs) and surface modes (SMs) of the fast magnetoacoustic wave. These modes are not induced by toroidicity. The n=0 GAEs owe their existence to shear. The frequency spacing between different radial and poloidal modes and the correlation of eigenfrequencies with changes in the edge density are examined and found to be in complete agreement with experimental observations of what have been named the `Alfven frequency modes' (AFMs) so far (Chang, Z., et al., Nucl. Fusion 35 (1995) 1469). Although the eigenfrequency is related to the edge density, the n=0 GAEs (AFMs) are not necessarily edge localized

The non-linear evolution of low mode number tearing modes is analysed, emphasizing the need for a threshold condition, to account for observations in tokamaks. The discussion is illustrated by two models recently introduced in the literature. The models can be compared with the available data and/or serve as a basis for planning experiments in order either to test theory (by means of beta -limit scaling laws, as proposed in this paper) or to attempt to control undesirable tearing modes. Introducing a threshold condition in the tearing mode stability analysis is found to reveal some bifurcation points and thus domains of intrinsic stability in the island dynamics operational space

Previously it has been shown that there is no lack of plasma induced magnetic field components that are strongly dependent on the plasma profiles. Here, a new method for collecting and interpreting magnetic data is presented. The properties of the magnetic probe system needed for the determination of both the plasma pressure profile and the current distribution are discussed. Considerable attention has been given to the evaluation of the required accuracy of magnetic measurements. Possible ways to simplify the general technique for the solution of particular problems are also discussed. As an example the Large Helical Device (LHD) configuration is referred to. A brief description of the peculiarities of plasma equilibria in this system relevant to the problem under consideration is presented. Emphasis is given to the cases in which plasma induced destruction of the plasma boundary is important

Over the last few years, a great deal of effort has been devoted to solving the problem of power and particle handling in divertors, which has been recognized as a critical issue for the operation of a magnetic fusion reactor. In particular, the choice of materials for plasma facing components has been examined with a view to developing heat and erosion resistant materials for divertor target plates. A large database on the behaviour of low-Z (carbon or beryllium) materials in tokamaks is available, while for high-Z materials there is little experience in the present generation of magnetic fusion devices. Frascati Tokamak Upgrade (FTU), a high field compact tokamak, has devoted part of its experimental campaign to studying the plasma characteristics when its limiter material is changed from the usual Inconel (nickel) to molybdenum and tungsten. Siliconization of the machine has also allowed the comparison of plasma performance when a relatively low-Z (silicon) ion is the dominant impurity. In this article, results are reported concerning the plasma operation, the differences in plasma characteristics and radiation losses, the impurity generation mechanisms and the relative impurity concentrations in the core plasma. A simulation of the experimental results, made with a self-consistent edge-core coupled model is presented, in order to provide evidence of the main physics mechanisms responsible for the observed behaviour

The maximum operational high beta in single-null divertor (SND) long pulse tokamak discharges in the DIII-D tokamak with a cross-sectional shape similar to the proposed International Thermonuclear Experimental Reactor (ITER) device is found to be limited by the onset of resistive instabilities that have the characteristics of neoclassically destabilized tearing modes. There is a soft limit due to the onset of an m/n=3/2 rotating tearing mode that saturates at low amplitude and a hard limit at slightly higher beta due to the onset of an m/n=2/1 rotating tearing mode that grows, slows down and locks. By operating at higher density and thus collisionality, the practical beta limit due to resistive tearing modes approaches the ideal magnetohydrodynamic (MHD) limit

Friction due to charge exchange with cold neutral atoms in the edge is investigated as a candidate to govern the poloidal rotation in the edge of a tokamak plasma. The Hirshman and Sigmar neoclassical moment approach is used to determine the rotation velocities of the main plasma ions and of one impurity species, when charge exchange friction is included. It is found that the poloidal rotation of the main plasma ions is controlled by charge exchange friction with neutrals. The impurity ion poloidal rotation is governed by the balance between the impurity viscous force and the main-ion-impurity-ion friction force. The results of the calculation are compared with the measurements obtained in the edge of a DIII-D high (H) mode plasma, using charge exchange recombination (CER) spectroscopy. It is found that the measured main ion poloidal rotation can be accurately predicted by the neoclassical theory including the effect of neutrals, assuming a neutral density ⟨n_n⟩ = 3 × 10¹⁷ m^-3 at the separatrix, decreasing exponentially inside the plasma with an e-folding length of 0.012 m, and peaking near the X point region with a poloidal peaking parameter γ ≡ ⟨n_n⟩⟨B²⟩/⟨n_nB²⟩ = 1.5. However, for the impurity ions, the neoclassical theory including a single impurity charge state, and regardless of the effect of the neutrals, gives a prediction that has the correct sign, but whose value is a factor of 5 or 6 different from the experimental value

Predictive transport simulations of dimensionally similar low confinement mode discharges have been performed using experimental data from Joint European Torus (JET) databases. The simulations, based on a gyro-Bohm transport model for temperature gradient driven drift waves, are found to reproduce the experimental profiles of density and temperatures well. In particular, apparent Bohm- and Goldston-like scalings of the local diffusivities are obtained as the gyroradius is varied. The results are due to small systematic deviations in the similarity conditions combined with marginal stability effects

Impurity toroidal rotation has been observed in the centre of ohmic plasmas in the Alcator C-Mod tokamak from the Doppler shifts of argon and molybdenum X-ray lines. The rotation is highest (~6 × 10⁶ cm/s) in the early portion of the discharges, when the loop voltage is highest and the electron density is lowest, and then typically settles to values ⩽ 2 × 10⁶ cm/s during the steady state period. The impurity rotation is in the same direction as the electron toroidal drift, opposite to the plasma current, and reverses direction when the plasma current direction is reversed. Molybdenum and argon ions rotate with the same velocity. These observations are in qualitative agreement with neoclassical theory

Report on the 12th International Conference held at St Raphael, France, 20-24 May 1996.