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A plasma fluid model which takes into account ion drifts has been constructed and applied to the scrape-off layer of a tokamak with a poloidal divertor. The model equations have been solved numerically in a realistic two-dimensional geometry. The solutions are characterized by near-sonic toroidal velocities and large poloidal flows in most of the scrape-off layer, together with steep gradients in the pressure and electrostatic potential along the magnetic field near the X-point, contrary to the predictions of the standard model. The potential step at the X-point should reduce parallel heat transport and could act as an H-mode trigger.

During the 1987 run, TFTR reached record values of QDD, neutron source strength, and Ti(0). Good confinement together with intense auxiliary heating has resulted in a plasma pressure greater than 3*105 Pascals on axis, which is at the ballooning stability boundary. At the same time improved diagnostics, especially ion temperature profile measurements, have led to increased understanding of tokamak confinement physics. Ion temperature profiles are much more peaked than previously thought, implying that ion thermal diffusivity, even in high ion temperature supershot plasmas, is greater than electron thermal diffusivity. Based on studies of the effect of beam orientation on plasma performance, one of the four neutral beamlines has been re-oriented from injecting co-parallel to counter parallel, which will increase the available balanced neutral injection power from 14 MW to 27 MW. With this increase in balanced beam power, and the addition of 7 MW of ICRF power it is planned to increase the present equivalant QDT of 0.25 to close to break-even conditions in the coming run.

A fluid model which includes ion drifts has been applied to the SOL of a tokamak with a poloidal divertor. Large parallel velocities and steep gradients in the pressure and electrostatic potential are found near the X-point, in contrast with the solutions of the standard model.

The Compact Ignition Tokamak (CIT) is a proposed modest-size ignition experiment designed to study the physics of alpha particle heating. The basic concept is to achieve ignition in a modest-size minimum cost experiment by using a high plasma density to achieve nT{sub E} {approx} 2 x 10{sup 20}s/m{sup 3} required for ignition. The high density requires a high toroidal field (10 T). The high toroidal field allows a large plasma current (10 MA) which provides a high level of ohmic heating, improves the energy confinement, and allows a relatively high beta ({approx} 6%). The present CIT design also has a high degree of elongation (k {approx} 1.8) to aid in producing the large plasma current. A double null poloidal divertor and pellet injection are part of the design to provide impurity and particle control, improve the confinement, and provide flexibility for improving the plasma profiles. Auxiliary heating is expected to be necessary to achieve ignition, and 10-20 MW of ICRF is to be provided.

Two dimensional calculations of impurtiy transport in a high recycling divertor scrape-off region have been made with an updated version of the ZTRANS Monte Carlo computer code. The calculations use plasma parameters for the Doublet 3 divertor, as computed by the Planet Fluid Transport Code. The effects of electric field, particle drift velocities, and thermal forces are included in the calculations. For all impurity species studied, it is found that impurity transport is dominated by frictional forces, over most of the scrape-off region. Light impurities, however, impinge substantially closer to the divertor plate center than do heavy impurities, which tend to impinge at the outer plate boundary. 8 refs., 4 figs., 2 tabs.

The poloidal magnetic divertor operating in the cool-high density plasma mode represents a good solution to the tokamak first wall erosion and impurity transport problems. It also, in general, improves the energy confinement time in the beam-heated discharges. In the existing experiments, the neutralizer plates are oriented close to normal to the flux surfaces in the poloidal plane, but the planned high-power experiments require that the plate be sharply angled to reduce the power flux density. We have studied the effect of changing the plate angle on the radial plasma temperature and density profiles at the plate. We find that some plate orientations shift the electron flux profile relative to the power flux profile, resulting in unfavorable temperature profiles. The question of existence of a steady-state solution was also addressed. For a steady state to exist, the radial neutral fluxes have to exactly cancel the radial diffusive ion fluxes. Depending on the diffusion coefficients, the plasma density may have to increase with radius to satisfy this condition. The implications of these conclusions for the INTOR, and CIT experiments is also discussed.

A simple empirical particle transport model has been constructed with the purpose of gaining insight into the L- to H-mode transition in tokamaks. The aim was to construct the simplest possible model which would reproduce the measured density profiles in the L-regime, and also produce a qualitatively correct transition to the H-regime without having to assume a completely different transport mode for the bulk of the plasma. Rather than using completely ad hoc constructions for the particle diffusion coefficient, we assume D = 1/5 chi/sub total/, where chi/sub total/ approx. = chi/sub e/ is the thermal diffusivity, and then use the kappa/sub e/ = n/sub e/chi/sub e/ values derived from experiments. The observed temperature profiles are then automatically reproduced, but nontrivially, the correct density profiles are also obtained, for realistic fueling rates and profiles. Our conclusion is that it is sufficient to reduce the transport coefficients within a few centimeters of the surface to produce the H-mode behavior. An additional simple assumption, concerning the particle mean-free path, leads to a convective transport term which reverses sign a few centimeters inside the surface, as required by the H-mode density profiles.