[en] A 12-processor Linux PC cluster has been installed to perform between-pulse magnetic equilibrium reconstructions during tokamak operations using the EFIT code written in FORTRAN. The MPICH package implementing message passing interface is employed by EFIT for data distribution and communication. The new system calculates equilibria eight times faster than the previous system yielding a complete equilibrium time history on a 25 ms time scale 4 min after the pulse ends. A graphical interface is provided for users to control the time resolution and the type of EFITs. The next analysis to benefit from the cluster is CERQUICK written in IDL for ion temperature profile analysis. The plan is to expand the cluster so that a full profile analysis (Te, Ti, ne, Vr, Zeff) can be made available between pulses, which lays the ground work for Kinetic EFIT and/or ONETWO power balance analyses

[en] The tokamak plasmas in most of the present experiments and those considered for the future reactors are up-down asymmetric in nature. This asymmetry arises due to external coils and conducting structures which surround the plasma. The analytical description of these equilibria using mathematically simpler methods is useful for the theoretical study of stability and transport. Such a tokamak equilibrium has been constructed analytically for arbitrary aspect ratios. The asymmetric nature arises through the homogeneous part of the exact solution of the Grad-Shafranov equation with a choice of pressure and toroidal function as linear in poloidal flux. These solutions can describe both single- and double-null divertor plasmas with an appropriate choice of plasma boundary. This has been used to construct DIII-D and ITER-like equilibrium and compared with those computed numerically.

[en] Experimentally, during fast wave (FW) radio frequency (rf) heating in DIII-D L-mode discharges, strong acceleration of neutral beam (NB) deuterium beam ions has been observed. Significant effects on the n/m = 1/1 sawtooth stability are also seen. Simulations using the Monte-Carlo Hamiltonian code ORBIT-RF, coupled to the TORIC full wave code, predict beam ion tails up to a few hundred keV, in agreement with the experiment. The simulations and experiment both clearly show a much greater efficiency for 4th harmonic FW heating than for 8th harmonic heating. Simple analyses of the kinetic contribution to the ideal magnetohydrodynamic (MHD) potential energy from energetic beam ions generated by FW heating yields reasonable consistency with the observations. A more detailed analysis shows a more complicated picture, however. Other physics effects such as geometry, plasma rotation, and the presence of a free boundary, play a significant role

[en] Understanding the stability physics of the H-mode pedestal in tokamak devices requires an accurate measurement of plasma current in the pedestal region with good spatial resolution. Theoretically, the high pressure gradients achieved in the edge of H-mode plasmas should lead to generation of a significant edge current density peak through bootstrap and Pfirsh-Schl(umlt u)ter effects. This edge current is important for the achievement of second stability in the context of coupled magneto hydrodynamic (MHD) modes which are both pressure (ballooning) and current (peeling) driven. Many aspects of edge localized mode (ELM) behavior can be accounted for in terms of an edge current density peak, with the identification of Type 1 ELMs as intermediate-n toroidal mode number MHD modes being a natural feature of this model. The development of a edge localized instabilities in tokamak experiments code (ELITE) based on this model allows one to efficiently calculate the stability and growth of the relevant modes for a broad range of plasma parameters and thus provides a framework for understanding the limits on pedestal height. This however requires an accurate assessment of the edge current. While estimates of j_edge can be made based on specific bootstrap models, their validity may be limited in the edge (gradient scalelengths comparable to orbit size, large changes in collisionality, etc.). Therefore it is highly desirable to have an actual measurement. Such measurements have been made on the DIII-D tokamak using combined polarimetry and spectroscopy of an injected lithium beam. By analyzing one of the Zeeman-split 2S-2P lithium resonance line components, one can obtain direct information on the local magnetic field components. These values allow one to infer details of the edge current density. Because of the negligible Stark mixing of the relevant atomic levels in lithium, this method of determining j(r) is insensitive to the large local electric fields typically found in enhanced confinement (H-mode) edges, and thus avoids an ambiguity common to MSE measurements of B_pol

[en] A new sustained high-performance regime, combining discrete edge and core transport barriers, has been discovered in the DIII-D tokamak. Edge localized modes (ELMs) are replaced by a steady oscillation that increases edge particle transport, thereby allowing particle control with no ELM-induced pulsed divertor heat load. The core barrier resembles those usually seen with a low (L) mode edge, without the degradation often associated with ELMs. The barriers are separated by a narrow region of high transport associated with a zero crossing in the ExB shearing rate

[en] The extension of the EFIT equilibrium reconstruction code to fine spatial-grid resolutions is discussed. The residue in the force-balance relation of the Grad-Shafranov (G-S) equation and the convergence property of these fine spatial-grid EFIT equilibria are studied in detail. The results suggest that fine spatial-grid equilibria generally better satisfy the force-balance constraint described by the G-S equation. Finer spatial-grid equilibria have typically smaller average error in satisfying the force-balance equation than coarse-grid equilibria and those extrapolated from coarse-grid results. Analysis of the equilibrium iteration algorithm employed in EFIT reveals that the iteration process is related to the spatial feedback stabilization of the plasma with flux control at various specified locations. Thus, for a converged equilibrium, axisymmetric stability is generally expected with feedback. The iteration error decreases self-similarly in the final stage of the iteration process and is related to the least stable axisymmetric mode in the feedback-stabilized equilibrium.

[en] The possibility of kinetic Alfven wave current drive at the center of tokamaks is proposed. The amount of driven current could be substantial and strongly affect the dynamics of the plasma central region. The relevance of this mechanism to present day and future tokamaks operating in the hybrid regime is discussed

[en] Fast shutdowns of DIII-D and ITER plasmas are simulated in 3D with the NIMROD code. The simulations assume uniform deposition of deuterium, raising the total electron density by factors of 100-150. Under these conditions, the plasma is found to be stable to n = 1 and n = 2 instabilities for the duration of the shutdown. However, 2D effects are significant, particularly in the evolution of the electron density, which tends to drop as recombination occurs in the cold edge plasma. The DIII-D simulation finds that a factor of 100 densification will not enable collisional suppression of runaway electron avalanching. Although the ITER simulation marginally exceeds the threshold for avalanche suppression, ITER is a more promising candidate to remain in the avalanche-free regime given carefully tailored initial conditions.

[en] The quiescent H (QH) mode, an edge localized mode (ELM)-free, high-confinement mode, combines well with an internal transport barrier to form quiescent double barrier (QDB) stationary state, high performance plasmas. The QH-mode edge pedestal pressure is similar to that seen in ELMing phases of the same discharge, with similar global energy confinement. The pedestal density in early ELMing phases of strongly pumped counter injection discharges drops and a transition to QH-mode occurs, leading to lower calculated edge bootstrap current. Plasmas current ramp experiment and ELITE code modeling of edge stability suggest that QH-modes lie near an edge current stability boundary. At high triangularity, QH-mode discharges operate at higher pedestal density and pressure, and have achieved ITER level values of β_PED and ν*. The QDB achieves performance of α_NH₈₉ ∼ 7 in quasi-stationary conditions for a duration of 10 tE, limited by hardware. Recently we demonstrated stationary state QDB discharges with little change in kinetic and q profiles (q₀ > 1) for 2 s, comparable to ELMing ''hybrid scenarios'', yet without the debilitating effects of ELMs. Plasma profile control tools, including electron cyclotron heating and current drive and neutral beam heating, have been demonstrated to control simultaneously the q profile development, the density peaking, impurity accumulation and plasma beta

[en] Free-boundary 3D tokamak equilibria and resistive wall instabilities are calculated using a new resistive wall model in the two-fluid M3D-C1 code. In this model, the resistive wall and surrounding vacuum region are included within the computational domain. This implementation contrasts with the method typically used in fluid codes in which the resistive wall is treated as a boundary condition on the computational domain boundary and has the advantage of maintaining purely local coupling of mesh elements. This new capability is used to simulate perturbed, free-boundary non-axisymmetric equilibria; the linear evolution of resistive wall modes; and the linear and nonlinear evolution of axisymmetric vertical displacement events (VDEs). Calculated growth rates for a resistive wall mode with arbitrary wall thickness are shown to agree well with the analytic theory. Equilibrium and VDE calculations are performed in diverted tokamak geometry, at physically realistic values of dissipation, and with resistive walls of finite width. Simulations of a VDE disruption extend into the current-quench phase, in which the plasma becomes limited by the first wall, and strong currents are observed to flow in the wall, in the SOL, and from the plasma to the wall.