[en] The INTOR impurity control system studies have been focused on the development of an impurity control system which would be able to provide the necessary heat removal and He pumping while satisfying the requirements for (1) minimum plasma contamination by impurities, (2) reasonable component lifetime (approx. 1 year), and (3) minimum size and cost. The major systems examined were poloidal divertors and pumped limiters. The poloidal divertor was chosen as the reference option since it offered the possibility of low sputtering rates due to the formation of a cool, dense plasma near the collector plates. Estimates of the sputtering rates associated with pumped limiters indicated that they would be too high for a reasonable system. Development of an engineering design concept was done for both the poloidal divertor and the pumped limiter

[en] This paper is divided into two sections: the first is a discussion of the interactions of neutral beams with confined plasmas, the second is concerned with the production and diagnosis of the neutral beams. In general we are dealing with atoms, molecules, and ions of the isotopes of hydrogen, but some heavier elements (for example, oxygen) will be mentioned. The emphasis will be on single-particle collisions; selected atomic processes on surfaces will be included

[en] Transport simulations of the present designs for the INTOR and TIBER ignition devices predict that broad sawtooth oscillations will appear in these experiments. As was noted previously in studies of the Compact Ignition Tokamak, the primary effect of the oscillations is to reduce fusion power production on the average through profile flattening. Due to the disparate time scales for energy and current diffusion between sawtooth crashes, the simulations also produce peaked pressure profiles over a large low shear region inside the q = 1 surface (q is the safety factor). Pressure-driven modes will likely be unstable in this case. 5 figs., 2 tabs

[en] Operational limits for density and β play a role in determining whether ITER can meet its goal of about 1 MW/m² neutron flux onto the first wall. This goal corresponds to roughly 1500 MW of fusion power and is equivalent to <β> approx. 0.032. Given the optimum temperature < t> approx. 10 keV for achieving ignition power balance, a line-average density of n-bar = 1.5n_GR is required, where n_GR = I_p/πa² denotes the empirical Greenwald limit. Recent experiments have obtained steady state ELMing H modes with line average densities exceeding the Greenwald limit, thereby documenting that the limit is empirical, not fundamental. Models for the occurrence of a density limit are discussed. Even though they are safely below the ideal MHD β limits, long pulse ITER-like tokamak discharges exhibit β limits arising from spontaneously generated magnetic island structures, which have been interpreted as arising from neoclassical boostrap current. Prospects for ECH stabilization of these islands appear favourable, theoretically. (author). 30 refs

[en] As part of the documentation of the ITER Conceptual Design Activities, this report describes the operations and research program of the ITER project. The project is divided into two phases: physics, and technology. Descriptions of the operations during the two phases, including experimental goals, are described. Included is a description of the three phases involving different activation resulting from fusion reactions. The operational flexibility and technological testing programs are described. Refs, figs and tabs

[en] 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 the condition of ntau/sub E/ ∼ 2 x 10²⁰ sec m^-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 improves the energy confinement, and provides a high level of ohmic heating. The present CIT design also has a gigh degree of elongation (k ∼ 1.8) to aid in producing the large plasma current. A double null poloidal divertor and a pellet injector are part of the design to provide impurity and particle control, improve the confinement, and provide flexibility for impurity and particle control, improve the confinement, and provide flexibility for improving the plasma profiles. Since auxiliary heating is expected to be necessary to achieve ignition, 10 to 20 MW of Ion Cyclotron Radio Frequency (ICRF) is to be provided

[en] Computational simulations using experimentally validated models are important for the development of the ITER divertor concept. The results of the simulations indicate that operation of a 'detached' plasma in ITER will require significant amounts of impurity radiation in the scrape-off layer and good confinement of the neutrals in the divertor chamber. (author). 17 refs, 1 fig

[en] We have investigated the theoretical feasibility of using fast ³He⁺⁺ ions, produced by ICRF minority heating, to simulate the behavior of fusion-produced ⁴He⁺⁺ ions in large fusion experiments. This proposed technique would allow the study of alpha-particle-like heating in near-term experiments, such as TFTR, JET, and JT-60, without the necessity of high levels of neutron production and the subsequent activation. The results of our calculations study show that ICRF minority heating should be able to produce confined fast ³He⁺⁺ distributions in TFTR-sized experiments with an energy distribution and density similar to those expected for fast alpha particles in fusion-reactor experiments, such as INTOR/FED/NET

[en] Calculations have been performed which demonstrate the possibility of operating poloidal divertors at high densities and low temperatures. This operating regime is caused primarily by ionization of recycling neutral gas near the divertor neutralizer plate which amplifies the input particle flux thereby raising the plasma density and lowering the plasma temperature. Low temperature, high density operation of poloidal divertors would ease the design requirements for future large tokamaks such as INTOR or FED by reducing the erosion rate in the divertor and reducing the neutral density and the associated charge exchange erosion near the main plasma. This regime may have already been observed on several divertor and limiter experiments

[en] The transport of neutral atoms and molecules in the edge and divertor regions of fusion experiments has been calculated using Monte-Carlo techniques. The deuterium, tritium, and helium atoms are produced by recombination in the plasma and at the walls. The relevant collision processes of charge exchange, ionization, and dissociation between the neutrals and the flowing plasma electrons and ions are included, along with wall reflection models. General two-dimensional wall and plasma geometries are treated in a flexible manner so that varied configurations can be easily studied. The algorithm uses a pseudo-collision method. Splitting with Russian roulette, suppression of absorption, and efficient scoring techniques are used to reduce the variance. The resulting code is sufficiently fast and compact to be incorporated into iterative treatments of plasma dynamics requiring numerous neutral profiles. The calculation yields the neutral gas densities, pressures, fluxes, ionization rates, momentum transfer rates, energy transfer rates, and wall sputtering rates. Applications have included modeling of proposed INTOR/FED poloidal divertor designs and other experimental devices