Table of contents

A magnetic analysis to determine the plasma cross-sectional shape and equilibrium parameters in the superconducting tokamak TRIAM-1M is discussed using magnetic data obtained in ohmic discharges. First, it is shown that the eddy currents in the tokamak structures should be taken into account. Two magnetic analysis codes are applied. One is the SHP code in which the plasma current is replaced by several filamental loop currents and the other is the FITEQ code in which the plasma current density is consistent with the MHD equilibrium. The magnetic fields due to eddy currents are expanded into multipole fields. The central safety factor q(0) is fixed at 1.0. The eddy currents reduce the vertical field by 10%. Taking account of them, the accuracy of the magnetic analysis is improved and estimation of the stored thermal energy in a plasma becomes possible. The SHP code has the difficulty that the local plasma surface position can be distorted by the eddy current magnetic field. In addition, it is also noted that the pressure profile analysed by the FITEQ code is consistent with the Thomson scattering measurement and that the energy confinement characteristics are consistent with the neo-Alcator scaling

A new type of antenna for launching waves in the ion cyclotron frequency range in toroidal plasma devices has been developed and tested in the Phaedrus-T tokamak. This antenna, a stacked stripline (SSL), generates a radiofrequency (RF) magnetic field similar to that of a conventional strap antenna but develops voltages 2 orders of magnitude less, and can operate without a Faraday shield. In comparison tests with a conventional strap antenna at a power level of 69 kW, the RF induced density rise and the impurity production (as indicated by soft X ray and loop voltage increases) were reduced by a factor of 3. In addition, it was demonstrated that capacitive loading effects were reduced or eliminated in comparison with those observed on a conventional strap antenna operated without a Faraday shield, and in the operation of a pair of SSL antennas at 90 degrees phase, small current drive effects were observed

The edge region of the reversed field pinch experiment RFX has been investigated with Langmuir and calorimetric probes. The energy flux measurements reveal a spatial structure that is consistent with the presence of a superthermal tail in the energy distribution function of the electrons, as expected according to the kinetic dynamo theory (KDT). In the framework of this model, the value of the magnetic field line diffusion coefficient in the edge region has been derived. The radial electric field obtained from the plasma potential gradient is opposite in sign to the ambipolar electric field expected in a stochastic magnetic field. The discrepancy is discussed in terms of particle recycling at the wall

Ignition and burn of deuterium-tritium (DT) fuel is investigated for inertial confinement fusion (ICF). Collisional kinetic equations describing the interaction of the high energy products of the fusion reactions with the plasma are solved by the particle-in-cell (PIC) method. Results are compared with simpler models, such as local alpha deposition and one-group alpha particle diffusion. Significant differences are found in temperature and density distributions as they evolve during burn. The total fraction of burned fuel is similar for the different models as long as ρR >>1 g/cm² and T⩾10 keV; for ρR<1 g/cm² and T<10 keV, however, the kinetic simulation gives considerably lower burn. Uniform as well as spark ignition configurations are simulated for initial temperatures and ρR values of practical interest and for fuel masses between 0.1 and 10 mg. In addition, optically thick configurations igniting at temperatures below 5 keV are considered

Accessibility and damping of the slow wave in a reversed field pinch (RFP) plasma is investigated theoretically, using projected Reversed Field Experiment (RFX) plasma parameters. By numerically solving the hot plasma dispersion relation, regions of propagation are found and the possibility of mode conversion is analysed. If the parallel index of refraction of the wave is chosen judiciously at the edge of the plasma, the slow wave is accessible to a target region located just inside the reversal surface without mode conversion. Landau damping is also optimized in this region. A representative fast electron population is then added in order to determine its effect on accessibility and damping. The presence of these electrons, whose parameters were estimated by extrapolation of Madison Symmetric Torus (MST) data, does not affect the accessibility of the wave. However, the initial phase velocity of the wave needs to be increased somewhat in order to maintain optimal damping in the target zone

In ion cyclotron range of frequency (ICRF) heating experiments in Heliotron-E, high energy particle production and confinement were studied. The high energy tail (up to 100 keV) formation of minority protons was clearly observed in a low density plasma experiment. The high energy tail below several tens of kiloelectronvolts was formed within 1 ms. The higher energy flux above this range developed within about 10 ms. The flux decay time after termination of the ICRF pulse was short compared with the values expected from the collisional process. The orbit loss of the accelerated particles is important for estimating the ICRF heating, particularly in helical confinement devices. To analyse the heating and orbit loss mechanism, a Monte Carlo code was used with an ICRF wave acceleration term. The time evolution of the tail formation and its decay resulting from the calculations agreed with the experimental ones. From this analysis, it is clear that the high energy particle production in ICRF experiments in helical confinement systems can be described by RF acceleration, collisional relaxation and orbit loss processes, and also that the orbit loss is not negligible in the ICRF heating of the low density plasmas of Heliotron-E

The timing of alpha losses, with respect to the various phases of a disruption, and the impact location of the losses are characterized during high fusion power operation of TFTR with deuterium and tritium fuels. Characterization of alpha losses is important for the design of future fusion devices such as ITER. In addition, characterization of the alpha losses with respect to the disruption evolution helps in the understanding of the dynamics of the disruption process and related MHD events such as sawtooth crashes. Disruptions are characterized as having several standard phases, applicable to most disruptions in all tokamaks: precursor, thermal quench(es) and current quench. Most of the losses are observed to occur during the thermal quench phase(s). In high beta disruptions, alpha losses start abruptly during the growth of MHD precursors, just before the onset of the thermal quench. The initial burst of losses, lasting as little as 100 mu s, can release a significant fraction (one third) of the total disruption induced losses during the thermal quench. An inventory of alpha particles suggests that the alpha loss distribution during disruptions might be quite different from that expected during non-disruptive discharges. There are no obvious differences between fast fusion product losses during deuterium-deuterium (DD) and deuterium-tritium (DT) disruptions, aside from the large alpha loss component in DT discharges

The particle exhaust fate of the DIII-D tokamak divertor cryopump system was investigated as a function of gas-puff-regulated core plasma line-averaged density, divertor electrical bias potential and position of the divertor strike point with respect to the pump entrance aperture. The experiment was done with rapidly ELMing H mode plasmas. Divertor bias approximately doubled the maximum achievable exhaust rate and also reduced its sensitivity to the strike point position. The achievable exhaust rate varied roughly as the third power of the core plasma density

In recent experiments carried out on the Phaedrus tokamak, the conducting plates that protect both sides of a poloidal antenna set have been replaced by non-conducting boron nitride plates. With this modification, direct heating and current drive by Alfven waves were demonstrated. An analysis proves that elimination of poloidal conducting protectors plays a crucial role in antenna performance improvement. The effect of the conducting side limiters on the wave spectrum and antenna impedance is calculated numerically for a cylindrical plasma column and model antenna current. The results demonstrate that conducting side limiters cause the excitation of additional modes that are deleterious for the Alfven heating and current drive scheme

A strong interaction of fast ions with ion Bernstein waves has been observed in TFTR. It results in a large increase in the fast ion loss rate and heats the lost particles to several megaelectronvolts. The lost ions are observed at the passing-trapped boundary and appear to be either DD fusion produced tritons or accelerated deuterium neutral beam ions. Under some conditions, enhanced loss of DT alpha particles is also seen. The losses provide experimental support for some of the elements required for alpha energy channelling

Fast changes of electric potential with different polarities are observed during sawtooth oscillations in the core region of a tokamak plasma using a heavy ion beam probe. Near the inversion radius the polarity of the observed change of the potential is found to be dependent on the swift movement of the hot core at the crash and is consistent with the prediction of one-fluid MHD theory. Near the magnetic axis the change of the potentials is positive and outside the inversion radius the change is negative. This is in contradiction with the MHD prediction

Partial sawtooth crashes in ohmic plasmas on TFTR are analysed using the high resolution electron cyclotron emission (ECE) image reconstruction technique. During the partial sawtooth crash, (m,n)=(2,2) or (3,3) modes are dominant. The (1,1) mode occurs after the partial crash. The partial sawtooth crash can be explained by the flattening of the (2,2) or (3,3) islands due to the localized reconnection

Predictive code simulations of high density scenarios of ITER and other reactor-grade devices require an experimentally validated scaling relation for the one-fluid effective heat diffusivity chi in the ELMy H mode regime. A comprehensive empirical chi scaling compatible with the ITERH92-P ELMy H mode scaling of the thermal energy confinement time is presented. It follows from a power law ansatz for chi and integration of the single-fluid energy equation and recovers all the exponents of the global confinement law. The numerical factor of the global scaling is used to calibrate the heat diffusivity. The dependences of chi on the temperature and temperature gradient, connected with the power degradation of confinement, are inferred from profile information of a high density H mode discharge. The scaling law obtained is successfully tested against JET, ASDEX and ASDEX Upgrade H mode discharges covering a wide parameter range. It is found to predict the strong rise of the experimental chi profiles with increasing effective radius

Corrections to Nuclear Fusion36 (1996) 159.