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Optimal error field correction (EFC) is thought to be achieved when coupling to the least-stable "dominant" mode of the plasma is nulled at each toroidal mode number (n). The limit of this picture is tested in the DIII-D tokamak by applying superpositions of in- and ex-vessel coil set n = 1 fields calculated to be fully orthogonal to the n = 1 dominant mode. In co-rotating H-mode and low-density Ohmic scenarios, the plasma is found to be, respectively, 7x and 20 x less sensitive to the orthogonal field as compared to the in-vessel coil set field. For the scenarios investigated, any geometry of EFC coil can thus recover a strong majority of the detrimental effect introduced by the n = 1 error field. Despite low sensitivity to the orthogonal field, its optimization in H-mode is shown to be consistent with minimizing the neoclassical toroidal viscosity torque and not the higher-order n = 1 mode coupling. [ABSTRACT FROM AUTHOR]

Recent DIII-D and RFX-mod experiments have demonstrated stable tokamak operation at very low values of the edge safety factor q(a) near and below 2. The onset of n=1 resistive wall mode (RWM) kink instabilities leads to a disruptive stability limit, encountered at q(a)=2 (limiter plasmas) and q95=2 (divertor plasmas). However, passively stable operation can be attained for q(a) and q95 values as low as 2.2. RWM damping in the q(a)=2 regime was measured using active MHD spectroscopy. Although consistent with theoretical predictions, the amplitude of the damped response does not increase significantly as the q(a)=2 limit is approached, in contrast with damping measurements made approaching the pressure-driven RWM limit. Applying proportional gain magnetic feedback control of the n=1 modes has resulted in stabilized operation with q95 values reaching as low as 1.9 in DIII-D and q(a) reaching 1.55 in RFX-mod. In addition to being consistent with the q(a)=2 external kink mode stability limit, the unstable modes have growth rates on the order of the characteristic wall eddy-current decay timescale in both devices, and a dominant m=2 poloidal structure that is consistent with ideal MHD predictions. The experiments contribute to validating MHD stability theory and demonstrate that a key tokamak stability limit can be overcome with feedback. [ABSTRACT FROM AUTHOR]

The system for controlling neoclassical tearing modes (NTMs) in DIII-D now catches the NTM the moment it becomes unstable by turning on the stabilizing electron cyclotron current drive (ECCD) and promptly bringing it back to stable before it has grown to a large size. Between NTMs, the ECCD can be turned off to save power, which will improve the fusion gain, Q, when used in ITER. This technique, named ‘catch and subdue’ (C&S), has been made possible by several advancements over the years at DIII-D. Firstly, ECCD must be very accurately aligned to the NTM; this is achieved by algorithms that probe how the NTM responds to changes in the alignment. Secondly, the alignment must be maintained even when the NTM is gone so that the ECCD will immediately stabilize when turned on in response to a new NTM. This is made possible by real-time equilibrium reconstructions that include measurements of the motional Stark effect and by a refraction estimator. Thirdly, real-time steerable mirrors are now fast and accurate actuators for the alignment adjustments. Fourthly, early NTM detection is made possible by a real-time mode analysis that filters noise to minimize false positives. These various control elements will be described and followed by a discussion of the further development needed for NTM control on ITER. [ABSTRACT FROM AUTHOR]

The δf kinetic-magnetohydrodynamic (MHD) model in the 3D extended-MHD code NIMROD is used to perform a linear simulation study of energetic particle effects on the n = 1 mode in a DIII-D hybrid discharge. The hybrid discharge has a long steady state with low qmin ≥ 1 at high confinement, is useful for numerous physics studies, and is a candidate operational scenario for burning plasma experiments. But hybrid discharges are limited to moderate βN by the m/n = 2/1 instability, which onsets with small increases in βN. Using realistic equilibria based on experimental reconstructions from DIII-D, the stability of the n = 1 mode during the steady state of a hybrid discharge is computed over a (qmin, βN) space. MHD stability analyses do not indicate instability to the n = 1 for small increases in βN above that of the experimental discharge. Our results show that energetic particles significantly change the stability map in (qmin, βN) parameter space from the MHD-only result. Unstable modes are driven by energetic particles far into the MHD stable region in (qmin, βN) space. Three different unstable regions are identified, each being defined by the fastest growing mode and distinctly different frequencies. We examine sample eigenmodes from these three regions. At low qmin ∼ 1 the drive is associated with the fishbone mode, while the higher qmin ≥ 1.12 the drive is associated with the Beta induced Alfven eigenmode. Overlaying the experimental trajectory in the same (qmin, βN) parameter space shows that all three regions are in proximity to the trajectory. In the higher qmin region a mode with a broad m/n = 2/1 dominant structure is most unstable, while in the region of the trajectory the most unstable mode has a m/n = 1/1 component localized near the axis. Experimentally, a nonlinearly saturated m/n = 1/1 structure is observed localized near the magnetic axis. This agreement suggests that the m/n = 2/1 mode may be triggered by energetic particles in these discharges. [ABSTRACT FROM AUTHOR]

Steady-state operation of the advanced tokamak reactor relies on maintaining plasma stability with respect to the resistive wall mode (RWM). Active magnetic feedback and plasma rotation are the two methods proposed and demonstrated for this purpose. A comprehensive modeling effort including both magnetic feedback and plasma rotation is needed for understanding the physical mechanisms of the stabilization and to project to future devices. For plasma with low rotation, a complete solution for the feedback issue is obtained by assuming the plasma obeys ideal magnetohydrodynamics (MHDs) and utilizing a normal mode approach (NMA) [M. S. Chu et al., Nucl. Fusion 43, 441 (2003)]. It is found that poloidal sensors are more effective than radial sensors and coils inside of the vacuum vessel more effective than outside. For plasmas with non-negligible rotation, a comprehensive linear nonideal MHD code, the MARS-F has been found to be suitable. MARS-F [Y. Q. Liu et al., Phys. Plasmas 7, 3681 (2000)] has been benchmarked in the ideal MHD limit against the NMA. The effect of rotation stabilization of the plasma depends on the plasma dissipation model. Broad qualitative features of the experiment are reproduced. Rotation reduces the feedback gain required for RWM stabilization. Reduction is significant when rotation is near the critical rotation speed needed for stabilization. The International Thermonuclear Experimental Reactor (ITER) [R. Aymar et al., Plasma Phys. Controlled Fusion 44, 519 (2002)] (scenario IV for advanced tokamak operation) may be feedback stabilized with β; above the no wall limit and up to an increment of ∼50% towards the ideal limit. Rotation further improves the stability. [ABSTRACT FROM AUTHOR]

A set of twelve coils for stability control has recently been installed inside the DIII-D [J. L. Luxon, Nucl. Fusion 42, 614 (2002)] vacuum vessel, offering faster time response and a wider range of applied mode spectra than the previous external coils. Stabilization of the n=1 ideal kink mode is crucial to many high beta, steady-state tokamak scenarios. A resistive wall converts the kink to a slowly growing resistive wall mode (RWM). With feedback-controlled error field correction, rotational stabilization of the RWM has been sustained for more than 2.5 s. Using the internal coils, the required correction field is smaller than with the external coils, consistent with a better match to the mode spectrum of the error field. Initial experiments in direct feedback control have stabilized the RWMs at higher beta and lower rotation than could be achieved by the external coils in similar plasmas, in qualitative agreement with numerical modeling. The new coils have also allowed wall stabilization in plasmas with toroidal beta up to 6%, almost 50% greater than the no wall limit. [ABSTRACT FROM AUTHOR]

Determining the scaling of transport with beta (β), the ratio of the plasma kinetic pressure to the magnetic pressure, helps to differentiate between various proposed theories of turbulent transport since mechanisms that are primarily electrostatic show little change in transport with increasing β, while primarily electromagnetic mechanisms generally have a strong unfavorable β scaling. Experiments on the DIII-D tokamak [J.L. Luxon, Nucl. Fusion 42, 614 (2002)] have measured the β scaling of heat transport with all of the other dimensionless parameters held constant in high confinement mode (H-mode) plasmas with edge localized modes. A four point scan varied β from 30% to 85% of the ideal ballooning stability limit (normalized beta from 1.0 to 2.8) and found no change in the normalized confinement time, i.e., B τth∞ β0.01-0.09. The measured thermal diffusivities, normalized to the Bohm diffusion coefficient, also did not vary during the bscan to within the experimental uncertainties, whereas the normalized helium particle transport decreased with increasing β. The H-mode pedestal β varied in concert with the core band showed no signs of saturation. This weak, possibly nonexistent, β scaling of transport favors primarily electrostatic mechanisms such as E×B transport, and is in marked disagreement with the strong unfavorable β dependence contained in empirical scaling relations derived from multimachine H-mode confinement databases. [ABSTRACT FROM AUTHOR]

Time-series analysis of magnetics data in tokamaks is typically done using block-based fast Fourier transform methods. This work presents the development and deployment of a new set of algorithms for magnetic probe array analysis. The method is based on an estimation technique known as stochastic subspace identification (SSI). Compared with the standard coherence approach or the direct singular value decomposition approach, the new technique exhibits several beneficial properties. For example, the SSI method does not require that frequencies are orthogonal with respect to the timeframe used in the analysis. Frequencies are obtained directly as parameters of localized time-series models. The parameters are extracted by solving small-scale eigenvalue problems. Applications include maximum-likelihood regularized eigenmode pattern estimation, detection of neoclassical tearing modes, including locked mode precursors, and automatic clustering of modes, and magnetics-pattern characterization of sawtooth pre- and postcursors, edge harmonic oscillations and fishbones. [ABSTRACT FROM AUTHOR]