Effects of the plasma boundary can have a substantial influence on the behaviour of the entire plasma in tokamaks. Progress in the field, particularly that over the last decade, is reviewed, with emphasis on experimental observation. Simple modelling for interpretation is also included.

A series of experiments, examining the confinement properties of ICRF heated H-mode plasmas, has been carried out on the C-Mod tokamak. C-Mod is a compact tokamak which operates at high particle, power, and current densities at toroidal fields up to 8T. Under these conditions the plasma is essentially thermal with very little contribution to the stored energy from energetic ions (typically no more than 5%) and with Ti~Te. Most of the data were taken with the machine in a single-null "closed" divertor configuration with the plasma facing components clad in molybdenum tiles. The data include those taken both before and after the first wall surfaces were coated with boron, with emphasis on the latter. H-modes obtained from plasmas run on boronized walls typically had lower impurity content and radiated power and attained higher stored energy than those run on bare molybdenum. Confinement enhancement, the energy confinement time normalized to L-mode scaling, for discharges with boronized walls, ranged from 1.6 to 2.4. The unique operating regime of the C-Mod device provided a means for extending the 1 tests of global scaling laws to parameter ranges not previously accessible. For example, the C-Mod ELMfree data was found to be 1.1-1.6 times the ITERH93 scaling and the ELMy data almost 2.0-2.8 times the ITERH92 ELMy scaling law, suggesting that the size scaling in both scalings may be too strong. While both ELMfree and ELMy discharges were produced, the ELM characteristics were not easily compared to observations on other devices. No large, low frequency ELMs were seen despite the very high edge pressure and temperature gradients that were attained. For all of our H-mode discharges, a clear linear relationship between the edge temperature pedestal and the temperature gradient in the core plasma was observed; the discharges with the "best" transport barriers also showing the greatest improvement in core confinement.

I IntroductionAlcator C-MOD', the third high-field compact tokamak in the Alcator line, has been operating tokamak plasmas since May 1993. Its design capability includes toroidal field, BT = 9 T, plasma current I, up to 3 MA, in plasmas with major radius R = 0.67 m, minor radius a = 0.21 m, with elongation up to n = 1.8. Divertor operation can be either into its closed, baffled, divertor chamber or to open flat plates. The magnetic configuration is rather similar to that presently envisaged for the International Thermonuclear Experimental Reactor, ITER, except that it is about a factor of ten smaller.The high particle-, current-and power-densities characteristic of such compact tokamaks lead to edge conditions that are in many respects comparable to those expected in ITER, and offer the opportunity to investigate so-called dissipative divertor operation, in which the power scraped off into the divertor is exhausted through a combination of neutral and radiative processes rather than through plasma conduction direct to the divertor plates.Alcator C-MOD offers excellent port access to the plasma for diagnostic and heating purposes. Its present complement of diagnostics includes full magnetics for equilibrium reconstruction, electron temperature profiles from electron cyclotron emission (ECE), density profiles from a ten-channel CO 2 laser interferometer, ion temperature profiles from high-resolution x-ray doppler measurements, neutron emission, and fast neutral particle analysis, various spectroscopic measurements such as visible bremsstrahlung, H. arrays, and vacuum ultraviolet impurity measurements, bolometer arrays, and x-ray and UV tomography. In addition, detailed edge, scrape-off-layer and divertor diagnosis based on probes and spectroscopy is available.The primary auxiliary heating method in the short term is ICRF, and two transmitters are available, providing a total 4 MW at 80 MHz. Thus far, experiments have concentrated on plasma coupling studies using a movable monopole antenna. Good power coupling into high density plasmas has been obtained, with loading resistance in the range of 5 to 15 Q, 2 in reasonable agreement with the theoretical calculations.So far the magnetic field has been limited to about 5.3 T awaiting power systems upgrades that will enable full-field operation next year. Even so, plasma currents up to 1 MA have been obtained, and durations over 1 second. Peak electron densities up to 9 x 1020 m-3, and temperatures up to T = 2.6, Ti = 1.6 keV have been achieved. Energy confinement is observed to exceed Neo-Alcator scaling.In section II we review some MHD and operational characteristics of the plasma.Section III discusses divertor experiments, section IV the confinement results, and section V the first ICRF coupling studies. II MHD and OperationA unique feature of the design of Alcator C-MOD is its thick stainless-steel vacuum vessel and structure. For reasons of mechanical strength, these have no insulating breaks and thus constitute 'shorted turns' on the ohmic transformer and the eddy ...

A summary is given of the present status of research on plasma-surface interactions in tokamaks. Three groups of important interactions are considered: recycling of the principal ion species, usually hydrogen or deuterium; the release and effect of low-Z contaminants; and the release and effect of high-Z contaminants. In each case the basic physical processes are reviewed and the relevant data from particlebeam measurements are summarized. Emphasis is given to discussing the effect of the various surface interactions in present-day tokamak discharges and in future fusion reactors. Surface studies in tokamaks are reviewed and methods of controlling the surface interactions and their effects are considered.

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JET carbon screening experiments were performed using methane gas injection. L-Mode experiments scanned parameters influencing the JET scrape-off-layer (SOL) and/or intrinsic impurity level. Scaling relations are derived to describe methane injected into L-Mode plasmas from the JET horizontal mid-plane. L-Mode screening was 3–20 times better for plasmas connected to the divertor than for similar limited plasmas. The screening was worse for methane injection from the mid-plane and best for injection from the divertor. The screening was 1.5–2 times worse for H-Mode than L-Mode. Both ELM-averaged and inter-ELM H-Mode screening was documented. The screening results were used to understand the intrinsic impurity levels. Zeff reduced at higher densities partly due to better carbon screening at the higher density, and partly due to decreased carbon influxes. Diverted L-Mode intrinsic carbon levels arose from both main chamber and divertor sources, while H-mode carbon primarily originated from the divertor. DIVIMP and EDGE2D were used to model the observed screening. The modelling indicated that carbon removal to the divertor required lower temperatures for Coulomb collisions to couple the impurity ions to the SOL deuterium flows. The carbon removal occurred primarily in the outer SOL regions.

Imbalances between the inboard and outboard legs of the single null divertor in tokamak Alcator C-Mod are observed to reverse when the direction of the toroidal field is reversed. These imbalances are measured by embedded probes in the target plates, tomographic reconstructions of bolometry and line radiation, and visible imaging. Density imbalances of about a factor of ten at the targets are observed at moderate density, decreasing as the density is raised until they are almost balanced. The data indicate that the electron pressure is not imbalanced, thus arguing against momentum imbalance as the cause of these drift-induced effects. Instead, power flux imbalance caused by E, A B convection, and enhanced by radiation, is suggested as the underlying cause.

The object of this review is to summarize the achievements of research on the Alcator C-Mod tokamak [Hutchinson et al., Phys. Plasmas 1, 1511 and Marmar, Fusion Sci. Technol. 51, 261 (2007)] and to place that research in the context of the quest for practical fusion energy. C-Mod is a compact, high-field tokamak, whose unique design and operating parameters have produced a wealth of new and important results since it began operation in 1993, contributing data that extends tests of critical physical models into new parameter ranges and into new regimes. Using only highpower radio frequency (RF) waves for heating and current drive with innovative launching structures, C-Mod operates routinely at reactor level power densities and achieves plasma pressures higher than any other toroidal confinement device. C-Mod spearheaded the development of the vertical-target divertor and has always operated with high-Z metal plasma facing componentsapproaches subsequently adopted for ITER. C-Mod has made ground-breaking discoveries in divertor physics and plasma-material interactions at reactor-like power and particle fluxes and elucidated a) Paper AR1 1, Bull. Am. Phys. Soc. 58, 21 (2013). b) Invited speaker.