Table of contents

The radial diffusion coefficient for confined counter-passing MeV ions is evaluated in a new way, using measurements of escaping D-D fusion products. A class of passing ions near the plasma centre is measured whose orbits could eventually diffuse across their passing/trapped boundary, resulting in unconfined trapped orbits which can be detected at the plasma wall. The analysis indicates that the passing MeV ion diffusion coefficient D is smaller than ≈0.1 m²/s, which is small compared to the thermal diffusivities of ≈1 m²/s for these plasmas. The same experiment and analysis should be applicable to alpha particles in D-T plasmas.

Ion and electron temperatures of up to 5 keV have been obtained within 10 cm of the last closed flux surface (LCFS) in hot ion H-modes with 18 MW of neutral beam heating. The ion temperature profiles appear to have 'pedestals' which scale with power per particle, T_i(a) ⪅ P(a)/n_e(0.9a) [keV, MW/10¹⁹ m⁻³], in H-modes. In L-modes the pedestal temperatures are an order of magnitude lower. This difference can be explained by the one order of magnitude difference in particle residence time in the two confinement modes. In H-modes, the resulting pressure 'pedestals' contribute typically 40% (up to 5 MJ) to the plasma stored energy, accounting for about half of the confinement improvement compared with the L-mode. The remaining improvement occurs across the entire plasma, but most notably in the periphery. There is ample evidence that ions with energies in the keV range escape to and beyond the LCFS and account for most of the loss power in the ion channel. This evidence includes results from energy discriminating probes in the scrape-off layer and the analysis of the characteristics of the heat strike zones in the X-point configuration. The observed boundary ion temperatures have far reaching implications for fusion reactivity, power handling, wall erosion and impurity production in reactor conditions. They suggest that it is possible to bring the entire volume of a reactor plasma to thermonuclear temperatures, with a considerable boost to the overall thermonuclear reactivity.

The performance of deuterium-tritium stellarator reactors is studied with a new, fast, onedimensional (1-D) transport survey code that is based on the spectral collocation method. Two operating modes with different signs of the assumed radial electric field are identified. The operating mode with a positive electric field is characterized by high temperatures and moderate densities, whereas the other mode has lower temperatures and higher densities. Both modes can be applied in possible reactors that can tolerate a large alpha particle energy loss. The sensitivity to device parameters and profile assumptions is examined. Scaling expressions useful for parametric studies are obtained for different quantities of interest, and the 1-D code results are compared with results derived from an empirical scaling relation. Deuterium-helium-3 operation is also feasible but is more demanding. The implications for stellarator reactor design optimization are discussed.

Sawtooth-free Ohmic discharges can serve as a model case of a quiescent tokamak plasma. The properties and global parameters of such discharges as observed in ASDEX are discussed as well as the mechanism apparently responsible for stabilization of the sawtooth instability. Stationary Ohmic discharges were used to investigate particle, impurity and energy transport in the absence of the sawtooth instability. Particular emphasis was placed on comparing these results with the predictions of neoclassical theories. The level of the ion energy transport is found to be that predicted by neoclassical theory. Particle transport and impurity transport are explained in terms of neoclassical inward drift velocities and diffusion coefficients with the same small anomalous contributions. In the central region of the plasma, where the power flux is low, the values of the electron heat conductivity were found to be very low.

The absorption of electron cyclotron waves in L-2 stellarator plasmas has been investigated by adapting the RAYS geometrical optics code developed at Oak Ridge National Laboratory for the parameters of L-2. Two heating schemes were considered: (1) launching of the ordinary wave from the low magnetic field side at the fundamental resonance and (2) launching of the extraordinary wave from the low magnetic field side at the second harmonic resonance of the electron cyclotron gyrofrequency. Significant power absorption (up to 100%) of the extraordinary mode at the second harmonic resonance was obtained. For the plasma parameters of L-2, the single-pass absorption of the extraordinary wave was twice that of the ordinary wave. A multipass absorption model was used to estimate the contribution to plasma heating of the power that remains after the first pass and is subsequently reflected from the vessel walls. Finally, results obtained with the RAYS code and with a code developed at the Institute of Automation and Electrometry were compared and found to be in good agreement.

The experimental lower hybrid current drive efficiency at zero electric field is found to agree with theory if the spectrum of the lower hybrid waves is fully accessible to the central region of the plasma. For broad or very low parallel wave number spectra, N_∥, for which the plane wave accessibility condition is not fully satisfied, the observed efficiency is higher than predicted. At non-zero electric field, the measurements agree with theories for an electric field dependent efficiency. The model can also be applied to experiments in which a symmetric spectrum of waves is launched, or to cases where the current is driven opposite to the DC electric field if runaway electrons can be neglected. A shift of the deposition zone by superposition of a low N_∥ current drive spectrum and a high N_∥ symmetric spectrum does not change the current drive efficiency.

Three distinct types of high toroidal mode number instabilities are obtained from a comprehensive kinetic calculation, using as input transport code results from the analysis of a recent design for the Burning Plasma Experiment (BPX). These instabilities are: the collisionless trapped electron-ion temperature gradient mode, the magnetohydrodynamic ballooning mode, and a high toroidal mode number version of the 'toroidicity induced Alfvén eigenmode' or 'gap' mode. The dependence of the instability linear eigenfrequencies on minor radius, beta and toroidal mode number is investigated, together with the effects of hot alpha particles. Relative quasi-linear fluxes of particles and energy for each species are also obtained. In addition, the beta dependence of the magnetohydrodynamic ballooning mode is investigated for a case using as input the results of a transport code calculation for the Tokamak Fusion Test Reactor (TFTR) in an extrapolation to a deuterium-tritium mixture. The effects of alpha particles and the relative quasi-linear fluxes are also investigated for this case.

The optimum confinement of plasmas produced by electron cyclotron heating (ECH) in Heliotron E and some other significant transport characteristics of ECH plasmas have been studied experimentally by changing the magnetic field configuration. It is found that the plasma confinement can be improved by minor changes of the magnetic field configuration. The main results are as follows: (1) Plasmas with a minor radius reduced (by means of a limiter) to two thirds of the value in the original (non-limited) configuration have optimum confinement properties when the magnetic axis is shifted inward by 20 mm. (2) The electron pressure in an optimized plasma configuration (confirmed by neutral beam injection) is higher than that in a standard ECH plasma. (3) The insertion of a carbon limiter causes the plasma line density to increase, with no significant degradation of the core temperature or density. (4) Magnetic field perturbations hardly affect the temperature and density profiles, unless the magnetic axis is shifted outward by a large (50 mm) distance. (5) Transport analysis indicates that significant improvement in the confinement properties can be achieved by slightly manipulating the magnetic field configuration.

Edge fluctuations are characterized and their associated transport is determined from Langmuir probe measurements in the ZT-40M reversed field pinch. It is found that the fluctuations have high normalized amplitudes and |ñ|/n = 0.4). There are significant contributions from magnetic perturbations acting on the equilibrium gradients. Compared to the global estimates, the fluctuation driven particle flux is large, whereas the corresponding electron energy flux is not. In the limiter shadow, the equilibrium density and electron temperature scale lengths are shorter and the fluctuation levels are higher. The fluctuation driven particle flux in the limiter shadow is 60% less than that in the plasma edge; most of the reduction is in the low frequency spectral region, which is where global MHD magnetic fluctuations are strongest.

Direct current (DC) driven and turbulence driven cross-field particle fluxes have been deduced from Langmuir probe measurements during discharges with ion cyclotron resonance heating (ICRH) in the TEXTOR tokamak. Spectral analysis of fluctuations of the floating potential at spatially separated probe pins has been used to determine the velocity associated with the poloidal rotation of the boundary plasma. Characteristics of the turbulence spectra of density and potential fluctuations at the edge during ICRH have been measured and compared with the corresponding spectra during standard ohmically heated discharges. It has been observed that both DC driven and turbulence driven fluxes increase significantly during ICRH. The measured high particle fluxes indicate motion in DC convective cells. The DC convective flux and the turbulence driven flux are partly counteracting.

A new concept of burn control in a tokamak reactor is discussed where the poloidal divertor is used as a feedback suppressor. The sensor signal of temperature fluctuations from the operating point is amplified, phase shifted and applied to the poloidal divertor plates as a modulating voltage. The resulting electrical current of the correct phase and amplitude can modulate the heat losses in a manner that renders the closed loop system thermally stable. Its completely electrical nature, simplicity and non-interference with the vital systems for heating and fuelling of the tokamak make this scheme technologically attractive.