Table of contents

A fifteen chord far infrared polarimeter has been used to measure internal poloidal field profiles in the Microwave Tokamak Experiment (MTX) during sawteeth, disruptions and electron cyclotron heating (ECH). A brief overview is presented of the instrument and the techniques used to analyse the profile data, followed by experimental results. Polarimeter current density (J) measurements are shown to be in good agreement with the expected resistivity profile based on T_e measurements. Measurements of safety factor (q) profiles during sawtoothing ohmic discharges are presented and are compared with various sawtooth models. It is found that q₀ approximately 0.7-0.85 during sawteeth and q₀>1 without sawteeth. The change in q₀ during the sawtooth ramp is small, at delta q₀ approximately 0.01. Finally, data are presented for ohmic discharges with higher core impurity radiation levels. The large radiated power produces hollow T_e and J profiles, and often leads to a disruption. This type of disruption has been stabilized on MTX by peaking the J profile through ECH in the core

The information provided by different diagnostics has been combined in order to characterize the fast electron distribution function in lower hybrid current drive experiments on JET. In particular, X-ray and electron cyclotron emission data are complementary in defining the fast electron energy contents in the directions parallel and perpendicular to the magnetic field. A numerical analysis chain has been developed which identifies the main moments of the distribution function of the current carrying fast electrons and allows simulations based on these moments to be compared with X-ray emission, electron cyclotron emission and magnetic data. The method of analysis and the associated diagnostics are described, and results are presented which have been obtained during the lower hybrid current drive campaign on JET

The ablation process of hydrogen fuel pellets in hot magnetized plasmas is investigated using a newly developed two dimensional time dependent hydrodynamics code, P2D. Surface evaporation and ablated fluid flow are evolved, coupled with a kinetic calculation of the non-local heating due to slowing down of hot Maxwellian background electrons in the cold ablated plasma. Results of the neutral gas shielding model of pellet ablation are extended to include effects due to non-spherical ablation flows, proper treatment of scattering in the kinetic calculation of the hot electron heat deposition, and atomic physics processes in the ablation cloud and at the pellet surface. These effects have been considered previously, though never as a comprehensive whole. The non-spherical nature of the ablation flow reduces the ablation rate by about a factor of two, while the use of a full Maxwellian distribution for the hot electrons increases ablation by a factor of about four. With Maxwellian hot electrons, atomic physics effects (principally due to dissociation and sublimation) can reduce the ablation rate by up to a factor of two. These effects are systematically quantified using simple physical arguments and code results. For comparison with experiment, we give a simple fit to the code ablation rate results for spherical deuterium pellets: G_2DGS approximately=1.65*10¹⁵R_p(cm)^1.26n_e(cm^-3)^0.35T_e(eV)^1.87 atoms/s, where R_p is the pellet radius and n_e (T_e) is the plasma density (temperature). This fit is compared with measured local ablation rates in Ohmic discharges on TFTR and JET. Predicted ablation rate profiles agree well with those on TFTR, but poorly with those on JET; pellet penetration depths agree well on both machines. Probable causes of the observed discrepancies (both experimental and theoretical) are discussed; in particular, transport processes in the post injection plasmas are shown to be able to account for some of the differences

Detailed consideration is given to the problem of current drive by ion Bernstein waves (IBWs). An analytical investigation of ray tracing equations is used to prove that a significant evolution of k_|| for the IBWs in tokamaks can ensure the possibility of current drive by these waves, if certain conditions are fulfilled. The main condition is the displacement of the antenna out of the midplane by a proper distance. Several scenarios for current drive are discussed, including direct excitation of the IBWs and the conversion of fast waves into IBWs. In all of these a loop antenna with a symmetrical spectrum in k_|| can be used, the electron distribution function also being symmetrical in V_||. The development of asymmetry in k_|| for the IBWs during their propagation is caused by the gradient of the toroidal magnetic field along the major radius. Numerical calculations are also performed - for the cases of ITER and the T-10 tokamak. It is concluded that IBWs can be used as an inexpensive and effective current drive and current density profile control in tokamaks

The global energy confinement of combined ohmic and lower hybrid driven TORE SUPRA plasmas has been analysed at various densities. In contradiction to the L mode ITER scaling law, this analysis indicates that the global energy confinement time depends strongly on the plasma density. Furthermore, the thermal electron energy content of steady state discharges is found to be in good agreement with the global Rebut-Lallia-Watkins (RLW) scaling law. Current ramp experiments show an enhancement of the global energy confinement with the internal inductance, l_i. These results have been extended to steady state regimes with lower hybrid current drive. Improved confinement has been obtained in a high l_i steady state plasma (l_i=1.7), where the modification of the current profile by lower hybrid waves leads to an increase in the central value of the safety factor (qψ(0) ≈ 2). In this case, the global confinement time is shown to exceed the value predicted by the RLW scaling law by 40%

Existing theories predict that in ohmic and/or L mode discharges the drift wave is stable in a sheared slab geometry. The representative experimental values for the plasma parameters show, however, that because of the very steep density gradient the shear stabilization criteria for collisionless drift waves are not likely to be satisfied in the H mode edge plasma. Magnetic shear is, therefore, not a candidate for stabilizing the slab branch of drift waves in the H mode. However, a radially sheared transverse (to the equilibrium magnetic field) velocity field is found to play a key role in stability. Velocity profiles corresponding to those observed in a region of a few centimetres inside the separatrix of an H mode plasma stabilize drift waves. On the other hand, velocity profiles corresponding to the L mode render drift waves unstable in the experimental range of parameters, even though the magnetic shear continues to play its stabilizing role

The relationships between plasma parameters that are dependent on the density profile (fusion rate, core and edge densities and temperatures, density and temperature profile peaking), fuel particle pumping in the divertor (expressed as particle throughput) and pellet injection parameters (size, velocity and injection frequency) are examined for tokamak reactor plasmas. These relationships are illustrated with the results of transport simulations of steady state conditions for a nominal set of reactor parameters representative of the International Thermonuclear Experimental Reactor (ITER) using pellet parameters characteristic of both present and advanced injectors. The radial separation between the pellet source and the pumping in the divertor region, even with very modest pellet velocities, allows for the possibility of density profile control in the plasma edge. When matching both fusion power and edge density constraints, fuel particle throughput decreases with increasing pellet penetration. The sensitivity of the results to the details of the transport model and to the limit on the edge density is also examined

A non-rotating helical magnetic field perturbation grows in the JET plasma when the electron density falls below a minimum value. The growth of this perturbation usually leads to a disruption. The helical phase of the perturbation is reproducible for a given configuration of the applied poloidal and toroidal magnetic fields. Small asymmetries in the applied magnetic fields are shown to cause the growth of the perturbation. Various sources of such field errors in JET have been identified. The existence of a minimum density for a stable plasma can place a severe constraint on the operation of JET. The magnitude of the minimum density depends upon various plasma parameters. Neutral beam injection can be used to reduce the minimum density

Inward transport of energy is observed on the DIII-D tokamak during off-axis electron cyclotron heating (ECH) experiments under a variety of operating conditions. Power balance analysis of steady state discharges shows that the electron heat flux can be negative (inward) interior to the ECH resonance location with a certainty of up to three times the experimental error bars. The ion heat flux is always positive (outward), as is the total heat flux. The electron heat flux becomes increasingly negative as the plasma density is increased. A detailed error analysis of the energy transport is performed as well as an independent verification of the measured plasma profiles. These results contradict purely diffusive models of radial heat transport, since energy in the electron channel is observed to flow in the direction of increasing temperature

This paper describes an update of the H mode confinement database that has been assembled for the ITER project. Data were collected from six machines of different sizes and shapes: ASDEX, DIII-D, JET, JFT-2M, PBX-M and PDX. The updated database contains better estimates of fast ion energy content and thermal energy confinement times, discharges with RF heating, data using boronization, beryllium and pellets, more systematic parameter scans, and other features. The list of variables in the database has been expanded, and the selection criteria for the standard dataset have been modified. We also present simple scalings of the total and thermal energy confinement time to the new dataset