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As shown on DIII-D edge localized modes (ELMs) can be either completely eliminated or mitigated with resonant magnetic perturbation (RMP) fields. Two infrared cameras, separated 105° toroidally, were used to make simultaneous measurements of ELM heat loads with high frame rates. Without the RMP fields ELMs display a variety of different heat load dynamics and a range of toroidal variability that is characteristic of their 3D structure. Comparing radial averages there is no asymmetry between two toroidal locations. With RMP-mitigated ELMs, the variability in the radially averaged power loads is significantly reduced and toroidal asymmetries in power loads are introduced. In addition to RMP ELM suppression scenarios an RMP scenario with only very small ELMs and very good confinement has been achieved.

Stochastic boundaries in fusion devices have been investigated in tokamaks, stellarators and reversed field pinch experiments for many years. However, since edge localized modes (ELMs) have been successfully eliminated in H-mode plasmas at the DIII-D tokamak [1,2] with small, edge resonant magnetic perturbations, they have become a widely investigated topic in tokamaks. In DIII-D stochastic boundaries are produced by coils external to the plasma. The magnetic field there consists of field lines with very different connection lengths, which produces a three dimensional, heterogenous structure of stochastic volume. The most obvious manifestation of the perturbed plasma edge is the strike line splitting observable in heat and particle fluxes, which changes with collisionality. The interaction of the magnetic perturbation and the magnetic equilibrium is of a resonant nature and the structure of the stochastic volume is a strong function of q(95). This is observed as a modulation of e.g. electron temperature as measured by ECE or Thomson scattering. In this work we summarize recent experimental findings on properties of the stochastic boundary in DIII-D. (C) 2010 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim

This paper reports on the interchange stability of a realistic analytical equilibrium solution of a field-reversed configuration (FRC) contained within a tight fitting cylindrical flux conserver with end walls, no external flux, and arbitrary elongation E. For E≫1, interchange stability is possible with ps/p0>7/12, where p0 is the peak pressure at the magnetic axis, and ps is the pressure at the separatrix/wall, a condition that appears to be compatible with high-pressure wall-confined, pulsed FRC implosions. Oblate configurations require somewhat less ps/p0 for stability. Analytical equilibria have been found for FRCs with two magnetic axes (doublets) with an internal figure-of-eight separatrix. A striking improvement in stability with respect to interchange modes was found in doublets: the private flux region (core) can be completely stabilized with zero separatrix pressure. This new finding suggests that doublet FRCs may be more relevant for steady-state fusion applications, which for good performance ...

The field-reversed configuration (FRC) is a high-beta compact toroidal plasma confinement scheme in which the external poloidal field is reversed on the geometric axis by azimuthal (toroidal) plasma current. A quasineutral, hybrid, particle-in-cell (PIC) approach [Y. A. Omelchenko and R. N. Sudan, Phys. Plasmas 2, 2773 (1995)] is applied to study long-term nonlinear stability of computational FRC equilibria to a number of toroidal modes, including the most disruptive tilt mode. In particular, a self-generated toroidal magnetic field is found to be an important factor in mitigating the instability and preventing the confinement disruption. This is shown to be a unique FRC property resulting from the Hall effect in the regions of vanishing poloidal magnetic field. The instability-driven toroidal field stabilizes kink formation by increasing the magnetic field energy without destabilizing curvature-driven plasma motion. Finally, the tilt instability saturates due to nonlinear, finite Larmor radius (FLR) effects and plasma relaxation to a quasisteady kinetic state. During this transition the FRC is shown to dissipate a substantial amount of initially trapped flux and plasma energy. These effects are demonstrated for kinetic and fluid-like, spherical and prolate FRCs.

Coaxial helicity injection has been investigated on the National Spherical Torus Experiment (NSTX). Initial experiments produced 130 kA of toroidal current without the use of the central solenoid. The corresponding injector current was 20 kA. Discharges with pulse lengths up to 130 ms have been produced.

A study of three-dimensional (3D) perturbed magnetic field structures and transport for edge localized mode control experiments with resonant magnetic perturbations at DIII-D is presented. We focus on ITER-Similar Shape plasmas at ITER relevant electron pedestal collisionalities . This study is performed in comparison with results from TEXTOR-Dynamic Ergodic Divertor circular limiter plasmas. For both experiments the magnetic field structure is analyzed in the vacuum paradigm—superimposing the external RMP field on the unperturbed equilibrium. For TEXTOR L-mode plasmas this description holds for normalized poloidal flux ΨN > 0.7 without tearing modes driven by the RMP field. For DIII-D H-mode plasmas the validity of this approach still needs to be established. In this paper a method is discussed to diagnose the degree of edge stochastization based on a comparison between modeled magnetic footprints on the divertor targets and experimental data. Clear evidence is presented for the existence of a generic separatrix perturbation causing striation of target particle fluxes. However, heat fluxes into these striations are small. This observation can be explained by accounting for the different heat and particle source locations and the 3D trajectories of the open, perturbed field lines toward the divertor target. Analysis of the transport characteristics filling the perturbed separatrix lobes based on initial EMC3/EIRENE modeling suggests the existence of open field lines connecting the stochastic edge to the target pattern. However, the width and inward most extent of the actual stochastic layer cannot yet be quantified.

The effect of resonant magnetic perturbations on heat transport in DIII-D H-mode plasmas has been calculated by combining the TRIP3D field line tracing code with the E3D two-fluid transport code. Simulations show that the divertor heat flux distribution becomes non-axisymmetric because heat flux is efficiently guided to the divertor along the three-dimensional invariant manifolds of the magnetic field. Calculations demonstrate that heat flux is spread over a wider area of the divertor target, thereby reducing the peak heat flux delivered during steady-state operation. Filtered optical cameras have observed non-axisymmetric particle fluxes at the strike point and Langmuir probes have observed non-axisymmetric floating potentials. On the other hand, the predicted magnitude of stochastic thermal transport is too large to match the pedestal plasma profiles measured by Thomson scattering and charge exchange recombination spectroscopy. The Braginskii thermal conductivity overestimates the experimental heat transport in the pedestal because the mean free paths of both species are longer than estimates of the parallel thermal correlation lengths, and collisionless transport models are probably required for accurate description. However, even the collisionless estimates for electron thermal transport are too large by one to two orders of magnitude. Thus, it is likely that another mechanism such as rotational screening of resonant perturbations limits the stochastic region and reduces transport inside of the pedestal.

Large Type-I edge localized modes (ELMs) are completely eliminated with small n = 3 resonant magnetic perturbations (RMP) in low average triangularity, , plasmas and in ITER similar shaped (ISS) plasmas, , with ITER relevant collisionalities . Significant differences in the RMP requirements and in the properties of the ELM suppressed plasmas are found when comparing the two triangularities. In ISS plasmas, the current required to suppress ELMs is approximately 25% higher than in low average triangularity plasmas. It is also found that the width of the resonant q95 window required for ELM suppression is smaller in ISS plasmas than in low average triangularity plasmas. An analysis of the positions and widths of resonant magnetic islands across the pedestal region, in the absence of resonant field screening or a self-consistent plasma response, indicates that differences in the shape of the q profile may explain the need for higher RMP coil currents during ELM suppression in ISS plasmas. Changes in the pedestal profiles are compared for each plasma shape as well as with changes in the injected neutral beam power and the RMP amplitude. Implications of these results are discussed in terms of requirements for optimal ELM control coil designs and for establishing the physics basis needed in order to scale this approach to future burning plasma devices such as ITER.

A new concept is proposed to measure a dc magnetic induction vector B whose magnitude is sufficiently large to saturate a small ferro- or ferrimagnetic sphere and align its magnetization vector M parallel to B. The direction of M is measured by modulating B locally at a convenient ac frequency by one or more small exciter coils near the sphere and using separate detector coils to measure components of the modulated induction due to the ac perturbed orientation of M. Equations are derived to relate the perturbed quantities to the original induction B. Several combinations of exciter and detector coils are possible, especially if distinct frequencies are used for each direction component. All three vector components of B can be calculated from the raw data. Because the proposed sensor has neither moving parts nor semiconductor components, it should be suitable for measurements in ionizing radiation environments, such as long pulse or steady state magnetically confined fusion power reactors.

Coaxial Helicity Injection (CHI) on the National Spherical Torus Experiment (NSTX) has produced 240 kA of toroidal current without the use of the central solenoid. Values of the current multiplication ratio (CHI produced toroidal current/injector current) up to 10 were obtained, in agreement with predictions. The discharges which lasted for up to 200 ms, limited only by the programmed waveform, are more than an order of magnitude longer in duration that any CHI discharges previously produced in a Spheromak or a Spherical Torus (ST).

Results from recent experiments demonstrate that quiescent H-mode (QH-mode) sustained by magnetic torque from non-axisymmetric magnetic fields is a promising operating mode for future burning plasmas. Using magnetic torque from n=3 fields to replace counter-Ip torque from neutral beam injection (NBI), we have achieved long duration, counter-rotating QH-mode operation with NBI torque ranging from counter-Ip to up to co-Ip values of 1-1.3 Nm. This co-Ip torque is 3 to 4 times the scaled torque that ITER will have. These experiments utilized an ITER-relevant lower single-null plasma shape and were done with ITER-relevant values of νped* and βNped. These discharges exhibited confinement quality H98y2=1.3, in the range required for ITER. In preliminary experiments using n=3 fields only from a coil outside the toroidal coil, QH-mode plasmas with low q95=3.4 have reached fusion gain values of G=βNH89/q952=0.4, which is the desired value for ITER. Shots with the same coil configuration also operated with net zero...

The authors' data indicate that the L-mode to H-mode transition in the DIII-D tokamak is associated with the sudden reduction in anomalous, fluctuation-connected transport across the outer midplane of the plasma. In addition to the reduction in edge density and magnetic fluctuations observed at the transition, the edge radial electric field becomes more negative after the transition. They have determined the scaling of the H-mode power threshold with various plasma parameters; the roughly linear increase with plasma density and toroidal field are particularly significant. Control of the ELM frequency and duration by adjusting neutral beam input power has allowed us to produce H-mode plasmas with constant impurity levels and durations up to 5 s. Energy confinement time in ohmic H-mode plasmas and in deuterium H-mode plasmas with deuterium beam injection can exceed saturated ohmic confinement times by at least a factor of two. Energy confinement times above 0.3 s have been achieved in these beam-heated plasmas with plasma currents in the range of 2.0 to 2.5 MA. Local transport studies have shown that electron and ion thermal diffusivities and angular momentum diffusivity are comparable in magnitude and all decrease with increasing plasma current.