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Register a new userMHD modeling of a DIII-D low-torque QH-mode discharge and comparison to observations
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Extended-MHD modeling of DIII-D tokamak [J. L. Luxon, Nucl. Fusion 42, 614 (2002)] quiescent H-mode (QH-mode) discharges with nonlinear NIMROD [C. R. Sovinec et al., J. Comput. Phys. 195, 355 (2004)] simulations saturates into a turbulent state but does not saturate when the steady-state flow inferred from measurements is not included. This is consistent with the experimental observations of the quiescent regime on DIII-D. The simulation with flow develops into a saturated turbulent state where the n=1 and 2 toroidal modes become dominant through an inverse cascade. Each mode in the range of n=1-5 is dominant at a different time. Consistent with experimental observations during QH-mode, the simulated state leads to large particle transport relative to the thermal transport. Analysis shows that the amplitude and phase of the density and temperature perturbations differ resulting in greater fluctuation-induced convective particle transport relative to the convective thermal transport. Comparison to magnetic-coil measurements shows that rotation frequencies differ between the simulation and experiment, which indicates that more sophisticated extended-MHD two-fluid modeling is required.
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The calculation presented in A neoclassical calculation of rotation profiles and comparison with DIII-D measurements by Stacey, Johnson, and Mandrekas, [Physics of Plasmas, 13, (2006)], contains several errors, including the neglect of the toroidal electric field, an unphysical expression for the electrostatic potential, and an unevaluated relation among its parameters. An alternative formulation is discussed.
The physical processes taking place at the edge region are crucial for the operation of tokamaks as they govern the interaction of hot plasma with the vessel walls. Numerical modeling of the edge with state-of-the-art codes attempts to elucidate interactions between neoclassical drifts, turbulence, poloidal and parallel flows that control the physical set-up of the SOL region. Here, we present post-processing analysis of simulations from the gyrokinetic code XGC1, comparing edge turbulence characteristics from a simulation of DIII-D against one of C-Mod. We find that the equilibrium $E times B$ flux across the separatrix has a similar poloidal pattern in both discharges which can be explained by magnetic drifts and trapped ion excursions. However, collisionality is noted to play a major role in that it prevents local charge accumulations from having global effects in C-Mod. In both cases, turbulent electron heat flux is higher than the ion one. This seems to be a universal characteristic of the tokamak edge. We identify turbulent frequencies and growth rates of the dominant mode in both simulations. In C-Mod, these numbers point to the presence of a drift wave. In DIII-D, linear simulations with Gene reveal a trapped electron mode. Furthermore, we present the amplitude and size distributions of the blobs from both simulations. Amplitude distributions are in qualitative agreement with experimental observations while size distributions are consistent with the fact that most blobs are not connecting to the divertor plates and suggest that they are generated by the shearing of the turbulent modes.
The toroidal phase and rotation of otherwise locked magnetic islands of toroidal mode number n=1 are controlled in the DIII-D tokamak by means of applied magnetic perturbations of n=1. Pre-emptive perturbations were applied in feedforward to catch the mode as it slowed down and entrain it to the rotating field before complete locking, thus avoiding the associated major confinement degradation. Additionally, for the first time, the phase of the perturbation was optimized in real-time, in feedback with magnetic measurements, in order for the modes phase to closely match a prescribed phase, as a function of time. Experimental results confirm the capability to hold the mode in a given fixed-phase or to rotate it at up to 20 Hz with good uniformity. The control coil currents utilized in the experiments agree with the requirements estimated by an electromechanical model. Moreover, controlled rotation at 20 Hz was combined with Electron Cyclotron Current Drive (ECCD) modulated at the same frequency. This is simpler than regulating the ECCD modulation in feedback with spontaneous mode rotation, and enables repetitive, reproducible ECCD deposition at or near the island O-point, X-point and locations in between, for careful studies of how this affects the island stability. Current drive was found to be radially misaligned relative to the island, and resulting growth and shrinkage of islands matched expectations of the Modified Rutherford Equation for some discharges presented here. Finally, simulations predict the as designed ITER 3D coils can entrain a small island at sub-10 Hz frequencies.
The guiding-center kinetic neoclassical transport code, XGC0, [C.S. Chang et. al, Phys. Plasmas 11, 2649 (2004)] is used to compute the heat fluxes and the heat-load width in the outer divertor plates of Alcator C-Mod and DIII-D tokamaks. The dependence of the width of heat-load fluxes on neoclassical effects, neutral collisions and anomalous transport is investigated using the XGC0 code. The XGC0 code includes realistic X-point geometry, a neutral source model, the effects of collisions, and a diffusion model for anomalous transport. It is observed that width of the XGC0 neoclassical heat-load is approximately inversely proportional to the total plasma current $I_{rm p}$. The scaling of the width of the divertor heat-load with plasma current is examined for an Alcator C-Mod discharge and four DIII-D discharges. The scaling of the divertor heat-load width with plasma current is found to be weaker in the Alcator C-Mod discharge compared to scaling found in the DIII-D discharges. The effect of neutral collisions on the $1/I_{rm p}$ scaling of heat-load width is shown not to be significant. Although inclusion of poloidally uniform anomalous transport results in a deviation from the $1/I_{rm p}$ scaling, the inclusion of the anomalous transport that is driven by ballooning-type instabilities results in recovering the neoclassical $1/I_{rm p}$ scaling. The Bohm or Gyro-Bohm scalings of anomalous transport does not strongly affect the dependence of the heat-load width on plasma current. The inclusion of anomalous transport, in general, results in widening the width of neoclassical divertor heat-load and enhances the neoclassical heat-load fluxes on the divertor plates. Understanding heat transport in the tokamak scrape-off layer plasmas is important for strengthening the basis for predicting divertor conditions in ITER.
Alfven Eigenmodes and magneto-hydrodynamic modes are destabilized in DIII-D reverse magnetic shear configurations and may limit the performance of the device. We use the reduced MHD equations in a full 3D system, coupled with equations of density and parallel velocity moments for the energetic particles (with gyro-fluid closures) as well as the geodesic acoustic wave dynamics. The aim of the study consists in finding ways to avoid or minimize MHD and AE activity for different magnetic field configurations and neutral beam injection operational regimes. The simulations show at the beginning of the discharge, before the reverse shear region is formed, a plasma that is AE unstable and marginally MHD stable. As soon as the reverse shear region appears, ideal MHD modes are destabilized with a larger growth rate than the AEs. Both MHD modes and AEs coexist during the discharge, although the MHD modes are more unstable as the reverse shear region deepens. The simulations indicate the destabilization of Beta induced AE, Toroidal AE, Elliptical AE and Reverse Shear AE at different phases of the discharges. A further analysis of the NBI operational regime indicates that the AE stability can be improved if the NBI injection is off axis, because on-axis injection leads to AEs with larger growth rate and frequency. In addition, decreasing the beam energy or increasing the NBI relative density leads to AEs with larger growth rate and frequency, so an NBI operation in the weakly resonant regime requires higher beam energies than in the experiment. The MHD linear stability can be also improved if the reverse shear region and the q profile near the magnetic axis are in between the rational surfaces q=2 and q=1, particularly if there is a region in the core with negative shear, avoiding a flat q profile near the magnetic axis.
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