Table of contents

An experimental study of neutral particle dynamics has been carried out over a wide range of operating conditions in the Alcator C Modified (Alcator C-Mod) tokamak using special purpose pressure gauges. Both divertor and midplane pressures are found to depend primarily on the edge plasma regimes defined by the scrape-off layer heat transport. While the maximum divertor neutral pressure (30-60 mtorr) is achieved at high core plasma densities (n_e=(2-4)*10²⁰ m^-3), corresponding to the detached divertor state, the maximum ratio of divertor to midplane pressure (~70) is achieved in the high recycling transport regime. Variations in the divertor geometry have a weaker effect on the neutral pressures. For otherwise similar plasmas, the divertor pressure and compression are maximized when the strike point is located at the bottom of the vertical target plate. Modelling shows that the high pressure sustained during detached divertor operation (despite a considerable drop in the recycling source) can be explained by scattering of neutrals off the cold plasma plugging the divertor throat

The dependence of the ignition threshold on the velocity v_imp and compressibility of an imploding fuel mass is central to establishing the driver requirements and implosion strategy for inertial confinement fusion (ICF). Using a series of LASNEX calculations, it is found that ke_imp varies as nu _imp^{- alpha} beta ^a, where ke_imp is the kinetic energy in the imploding fuel at the ignition threshold, alpha =5.5+or-0.5, a=1.7+or-0.2 and v_imp is the implosion velocity. Here, the compressibility parameter beta is related to the pressure P and density rho of the DT fuel by the relation P= beta p⁵3/. These results are obtained by starting at the peak implosion velocity for a fuel shell of a high gain ICF capsule and scaling the isentrope, mass and velocity of the fuel shell. In the presence of a mix of hot and cold material at the edge of the central hot spot, it is also found that the results can be fitted by assuming that the reduced clean fuel radius for a mixed capsule requires a velocity increase of the same magnitude as that which would be required if the entire capsule had been rescaled in size by the same ratio

The analyses of the neutral energy distribution measured by the time of flight (TOF) method in the JAERI Fusion Torus 2M (JFT-2M) tokamak are summarized. The TOF technique is also reviewed. Rapid changes of the neutral energy distribution at L-H, H-L transitions, edge localized mode (ELM) and sawtooth crash are observed and these occur earlier than the change of H_alpha intensity. This means that the ion energy distribution changes and is playing a very important role at L-H, H-L transitions and ELM. The change of the energy distribution is observed as an increase or decrease of the high energy outflux above an energy of 200 eV. An energy of greater than 200 eV for hydrogen corresponds to the collisionless condition nu _*i<1 just inside the separatrix. The observations during an L-H transition from a sawtooth crash to a fall in the H_alpha intensity are discussed sequentially

The longitudinal losses of charged particles from a mirror device with an arbitrary mirror ratio are calculated in a systematic way by using a linearized Fokker-Planck equation

A transport analysis is reported of a series of discharges in the Rijnhuizen Tokamak Project (RTP) tokamak, in which Z_eff was varied by using mixtures of helium and neon as the filling gas. Z_eff was scanned over the range 5.1 to 9.5. As a reference, pure deuterium plasmas with Z_eff ≲ 2 were used. Only a very weak dependence of χ_e on Z_eff is observed, both in ohmic and in ECH discharges. The scaling of χ_e with n_e and Z_eff is consistent with the scaling of the global τ_Ee, both showing an absence of scaling with Z_eff. Scaling with the collisionality shows χ_e ~ ν_e^*0.4 in ohmic discharges. A comparison is made with the predictions of transport models. Some models correlate with either the ohmic or the ECH data set. None of the tested models, however, predicts the observed Z_eff dependence of χ_e in both ohmic plasmas and during ECH. It is noted that the ohmic discharges are in the plateau regime, while the ECH discharges are largely in the banana regime

Measurements of radiofrequency (RF) loading on the Doubler III D shape (DIII-D) tokamak using a fast wave ion cyclotron antenna with and without a Faraday shield present indicate that the loading resistance seen by the antenna is significantly higher at low values of RF power (P ⩽ 20 kW with no Faraday shield) than is predicted by standard fast wave antenna coupling theory. Measurements of the density profile in the plasma edge region allow a direct comparison of calculated and measured loading. At high power, the calculated fast wave loading agrees with the measurements. The high loading observed at low power can be explained by direct losses in an RF enhanced sheath. The density value at the current strap of approximately=4*10¹⁰ cm^-3 needed to give the observed loading behaviour is in agreement with the measured density values in the plasma scrape-off region, extrapolated to the current strap location using a simple Bohm diffusion model

Radiative collapses in the Advanced Toroidal Facility (ATF) torsatron have been studied using the predictive transport code Proctr for realistic shots with moderate radiation levels (P_rad/P_in approximately=30%). Simulations to study the influence of density and of different impurity species on the occurrence of thermal collapses have been carried out and it has been found that the impurity cooling rate profiles play a decisive role. A local breaking of the power balance is suggested as the possible origin of the thermal collapse and, once the collapse has started, the rate of increase of the local energy lost by radiation seems to be the important parameter. The lack of efficiency in the radial energy transfer to sustain the local energy balance in the regions with enhanced losses leads to plasma collapse

The results of electron cyclotron heating (ECH) experiments in the L-2M stellarator are presented. The main goal of the experiments is to investigate the physics of ECH and of plasma confinement at high values of the volume heating power density. A current free plasma is produced and heated by extraordinary waves at the second harmonic of the electron cyclotron frequency ( omega ₀=2 omega _ce). The experimental results are compared with the empirical LHD scaling and with numerical simulations of plasma confinement and heating processes based on neoclassical theory using the full matrix of transport coefficients, including some additional anomalous corrections

A five chord, high resolution X-ray spectrometer array has been used to measure vertical brightness profiles of helium-like argon emission from Alcator C-Mod plasmas, out to the last closed flux surface. During standard Alcator C-Mod operation, with the X point and the ion B* Del |B| drift downward, the helium-like argon brightness is a factor of ~8 larger at the top of the plasma than at the bottom, near the plasma edge. In these edge regions, where the electron temperature is low, the upper levels of observed transitions are populated by radiative recombination of hydrogen-like argon. This up-down brightness asymmetry can be explained by a factor of ~8 enhancement of hydrogen-like argon at the top of the machine at r/a=0.9. This implies a vertical impurity drift from inside the plasma, since the hydrogen-like argon is born near the plasma centre. With the ion B* Del |B| drift upward, the enhancement switched to the bottom of the machine, but did not change when the X point was moved up. This asymmetry agrees qualitatively with the predictions of neoclassical parallel impurity transport

Employing International Thermonuclear Experimental Reactor (ITER) scaling laws for the energy confinement time tau _E, burn conditions for an ignited deuterium-tritium (DT) plasma with self-consistent inclusion of helium ash are transformed from the beta , T plane into the n_e, T and the beta , T planes. The linear and non-linear stabilities of these burn equilibria are studied, taking into account not only the usual temperature and density fluctuations, but also fluctuations of the fuel concentrations. It is shown that extending the application of ITER scaling for tau _E to the dynamics of unstable plasma states is reasonable and leads to an appreciable stabilization. Furthermore, methods of stabilizing unstable burn equilibria dynamically and through feedback control are discussed. Finally, conclusions are drawn concerning the optimal burn regime, including the requirements of burn stability, the restrictions by operational limits ( beta and density limits) due to magnetohydrodynamic (MHD) stability requirements, and performance demands, such as the attainable fusion power P_fus and the fraction gamma =P_rad/P_alpha of alpha particle heating power that can be radiated away

The pellet charge exchange (PCX) diagnostic on the Tokamak Fusion Test Reactor (TFTR) presently measures trapped alpha distribution functions with very small pitch angles ( nu _||/ nu ~0.05) at the midplane. The measured PCX alpha signal exhibits a depletion region near the outboard region. Results of the alpha energy spectra and radial profile suggest that stochastic ripple diffusion is the cause of the depletion. Comparison of the ripple stochastization boundary with Goldston-White-Boozer theory also shows the correct functional dependence on alpha energy and q profile