Table of contents

Quasi-stationary, high β tokamak plasmas (β_N^dia ≈ 2) have been developed in the COMPASS-D tokamak using ECRH. These regimes are sustained for 20 energy confinement times and are limited by the duration of the heating pulse. For the first time, fast particle instabilities are observed with strong electron heating only. The plasmas are sawtooth free and the most dangerous (m = 2, n = 1) neoclassical tearing modes are avoided. The scenario is a result of the optimization of the timing of the heating pulse and application of LHCD.

MHD induced beam ion losses were observed on the CHS heliotron/torsatron with an energy and pitch angle resolved scintillator detector. Beam ions escaping into the probe were characterized using orbit calculation codes. It was found that two types of ions, namely `passing boundary ions' and `trapped ions', were lost periodically during an m/n = 3/2 fishbone-like instability. There are thresholds in the mode amplitude for the ejection of fast ions. Above this threshold, losses are enhanced together with an increase of the mode amplitude. The excitation of the fishbone-like instability and the enhancement of MHD induced losses strongly depend on the degree of accumulation of beam ions in the plasma. Orbit calculations show that passing boundary ions, which have large pitch angle and v_ϕ close to zero (or close to the propagating velocity of the magnetic fluctuation) near the q = 3/2 surface, are ejected by the instability.

The neoclassical return current and parallel viscosity due to the radial electric field (non-equilibrium poloidal rotation) in tokamaks are calculated using the 5-D (3-D in configuration space and 2-D in velocity space) Monte Carlo code ASCOT. Results are compared with the known neoclassical non-ambipolar return current expressions by Shaing [Phys. Fluids B 2 (1990) 2847] and Stringer [Nucl. Fusion 33 (1993) 1249] over a wide range of poloidal Mach numbers M_p = E_r/(v_tiB_θ) and for various collisionalities. Here, v_ti is the ion thermal velocity, E_r is the radial electric field and B_θ is the poloidal magnetic field. Quantitative differences exist due to a simpler collision operator and the various other approximations used in deriving the analytic expressions.

Graphical methods are developed by which the guiding centre orbit topology of fast ions in tokamaks is efficiently obtained and displayed. For constant energy we characterize the orbits by a new set of constant of motion variables (S,χ) describing the radial position and pitch angle of the orbits as they intersect with the stagnation surface. This surface is a generalization of the equatorial plane for non-top-bottom symmetric magnetic profiles. The orbit topology is obtained simply by studying the contours of the magnitude of the magnetic moment μ and canonical angular momentum p_ϕ in the (S,χ) plane. The methods, suitable for routine analysis, can also be applied to evaluate the orbit topology in advanced tokamak scenarios where the presence of special orbit types (co-pinch orbits) causes first orbit loss of fusion products.

The effect of magnetic field on the velocity distribution function of tritons in D-³He fuelled/field reversed configuration (FRC) plasmas is investigated. Some of the tritons produced by D(d,p)T reactions in the FRC are immediately and asymmetrically lost from the device without any interaction with background charged particles, and then non-symmetric loss causes a distortion of the velocity distribution function as well as a decrease in the number of trapped tritons. Using the distorted triton distribution, the 14 MeV neutron generation rate is estimated and compared with the values for Maxwellian plasmas (the effect of the magnetic field is neglected). It is found that in a typical ignited D-³He/FRC plasma, for example, T_i = T_e = 80 keV, n_D = 2n_³He = 0.5n_e = 3 × 10²⁰m^-3, τ_E = 0.5τ_P = 3 s, r_s = 1.6 m and B_W = 6 T, the reduction in 14 MeV neutron generation is about 20%.

In the tokamak Tore Supra, when a poloidally and toroidally localized limiter, called a modular limiter, is introduced into the lower part of the scrape-off layer, density fluctuations located in the vicinity of this limiter present a specific feature with the appearance of a new spectral pattern in the associated frequency spectrum. This leads to a strong up-down asymmetry observed in both the plasma edge and the plasma outer core, with a maximum level of turbulence at the bottom of the plasma. The observed asymmetry characteristics show that magnetic connection lengths play a critical role here and that the limiter configuration has some effect on the outer core turbulence.

Pfirsch-Schlüter current (PSC) is an inherent property of a finite pressure toroidal plasma equilibrium in tokamaks and stellarators. However, it has been pointed out (V.D. Pustovitov, Nucl. Fusion 36 (1996) 583) that the PSC in a conventional stellarator could be suppressed completely by adjusting the external vertical magnetic field so that it satisfies the condition that the quantity characterizing the inhomogeneity of the magnetic field strength on a flux surface is a flux function. In this article an example is presented of such an l = 3 stellarator equilibrium. Complete suppression of PSC in a finite beta plasma is shown numerically by using the fixed boundary version of the VMEC code. The calculated equilibrium is also analysed from the viewpoint of the Boozer magnetic spectrum. It is shown that a reduction of the poloidal inhomogeneity of the magnetic field strength is essential for reducing the PSC.

Modelling of LHCD with transport calculations is performed with the JETTO transport code, which has been upgraded by implementing the Fast Ray Tracing Code to calculate self-consistent LH power deposition profiles. Heat and particle transport models that are able to reproduce the experimental JET temperature and density profiles are used in JETTO for predictive high performance modelling. Application of 3.5 MW LHCD power provides an inverted q profile across 50-70% of the plasma radius whereas, without LHCD, the q profile is monotonic during the flat-top phase. The results predict that the fusion power is about 60% higher for high performance DT plasmas in the optimized shear scenario with 3.5 MW LHCD applied during the high performance phase than without LHCD at B_t = 3.4 T and I_p = 3.9 MA on JET. In addition, the width of the internal transport barrier (ITB) is 0.25-0.30 m larger and the ITB can be sustained for a longer time with LHCD.

For controlling the edge particle flux, a proposed method of pumping is to capture backscattered neutrals due to atomic processes with a dedicated vented structure. This solution has been tested with the ergodic divertor (ED) of Tore Supra. The investigated ranges of volume averaged density ⟨n_e⟩ and total power P_tot are (2.5-5.5) × 10¹⁹m^-3 and 1-6.5 MW, respectively, covering both the high recycling and semidetached plasma regimes. In this domain of parameters, the neutral pressure in the ED plenum always remains proportional to the incident ion flux. It follows that the high recycling regime is the most efficient for density control. This regime is then studied in more detail and the behaviour of the edge temperature and density with ⟨n_e⟩ and P_tot characterized, together with the location of the transition towards the semidetached regime in the [⟨n_e⟩ × P_tot] plane. The particle balance around the neutralizers is analysed with the code EDCOLL (Ergodic Divertor COLLection), the simulations of which show a satisfactory agreement with both the pressure and D_α emission measurements, demonstrating that both neutral recirculation and particle collection with the vented structure are correctly understood. Finally, possible improvements of particle control with the ED are discussed and the code EDCOLL is used to estimate the collection efficiency of an optimized structure. It is shown that the present performance could be multiplied by ≈ 3, allowing efficient density control in the semidetached plasma regime at a moderate level of power.

A spheromak type compact torus (CT) plasma is investigated by studies of the gun type plasma formation mechanism and from the characteristics of the magnetic decay following the establishment of a CT equilibrium in the flux conserver. Radiofrequency heating in the lower hybrid range (432 MHz) is used to study the electron heat confinement using a 20 MW, 100 μs power pulse. While electron temperatures up to 200 eV have been measured with RF heating, the impact on magnetic decay from the RF heating is relatively weak. A simple model based upon parallel electron heat transport along stochastic magnetic field lines gives a scaling law for magnetic decay which fits the experimental data. Results of two dimensional MHD code calculations are given which match many details of the experimentally observed formation physics.