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.

Direction reversals of intrinsic toroidal rotation have been observed in Alcator CMod Ohmic L-mode plasmas following modest electron density or toroidal magnetic field ramps. The reversal process occurs in the plasma interior, inside of the q = 3/2 surface. For low density plasmas, the rotation is in the co-current direction, and can reverse to the counter-current direction following an increase in the electron density above a certain threshold. Reversals from the co-to counter-current direction are correlated with a sharp decrease in density fluctuations with k R ≥2 cm −1 and with frequencies above 70 kHz. The density at which the rotation reverses increases linearly with plasma current, and decreases with increasing magnetic field. There is a strong correlation between the reversal density and the density at which the global Ohmic L-mode energy confinement changes from the linear to the saturated regime.

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 ...

Nonlinear gyrokinetic simulations of trapped electron mode (TEM) turbulence, within an internal particle transport barrier, are performed and compared with experimental data. The results provide a mechanism for transport barrier control with on-axis radio frequency heating, as demonstrated in Alcator C-Mod experiments [S. J. Wukitch et al., Phys. Plasmas 9(5) 2149 (2002)]. Off-axis heating produces an internal particle and energy transport barrier after the transition to enhanced Dα high confinement mode. The barrier foot reaches the half-radius, with a peak density 2.5 times the edge density. While the density profile peaks, the temperature profile remains relatively unaffected. The peaking and concomitant impurity accumulation are controlled by applying modest central heating power late in the discharge. Gyrokinetic turbulence simulations of the barrier formation phase, using the GS2 code [W. Dorland et al., Phys. Rev. Lett. 85, 5579 (2000)] show that toroidal ion temperature gradient driven modes are suppressed inside the barrier foot, but continue to dominate in the outer half-radius. As the density gradient steepens further, trapped electron modes are driven unstable. The onset of TEM turbulence produces an outflow that strongly increases with the density gradient, upon exceeding a new nonlinear critical density gradient, which significantly exceeds the linear critical density gradient. The TEM turbulent outflow ultimately balances the inward Ware pinch, leading to steady state. Moreover, the simulated turbulent particle diffusivity matches that inferred from particle balance using measured density profile data and the calculated Ware pinch. This turbulent diffusivity exhibits a strong unfavorable temperature dependence that allows control with central heating.

Double transport barrier plasmas comprised of an edge enhanced D α (EDA) H-mode pedestal and an internal transport barrier (ITB) have been observed in Alcator C-Mod. The ITB can be routinely produced in ICRF heated plasmas by locating the wave resonance off-axis near |r/a| ∼ 0.5, provided the target plasma average density is above ∼1.4 × 10 20 /m 3 , and can develop spontaneously in some Ohmic H-mode discharges. The formation of the barrier appears in conjunction with a decrease or reversal in the central (impurity) toroidal rotation velocity. The ITB foot is located near r/a = 0.5, regardless of how the barrier was produced. The ITBs can persist for ∼15 energy confinement times (τ E), but exhibit a continuous increase of the central electron density, up to values near 1×10 21 /m 3 (in the absence of an internal particle source), followed by collapse of the barrier. This barrier is also evident in the ion temperature profiles, and a significant drop of the core thermal conductivity, χ eff , when the barrier forms is confirmed by modeling. Application of additional on-axis ICRF heating arrests the density and impurity peaking, which occurs along with an increase (co-current) in the core rotation velocity. Steady state double barrier plasmas have been maintained for 10 τ E or longer, with n/n GW ∼ 0.75 and with a bootstrap fraction of 0.13 near the ITB foot. The trigger mechanism for the ITB formation is presently not understood.