Stabilizing modes that limit plasma beta and reducing their deleterious effects on plasma rotation are key goals for efficient operation of a fusion reactor. Passive stabilization and active control of global kink/ballooning modes and resistive wall modes (RWM) have been demonstrated on NSTX and research now advances to understanding the stabilization physics and reliably maintaining the high beta plasma for confident extrapolation to ITER and CTF. Active n = 1 control experiments with an expanded sensor set, combined with low levels of n = 3 field phased to reduce error fields, reduced resonant field amplification and maintained plasma rotation, exceeded normalized beta = 6, and produced record discharge durations limited by magnet system constraints. Details of RWM active control show the mode being converted to a rotating kink that decays, or saturates leading to tearing modes. Discharges with rotation reduced by n = 3 magnetic braking suffer beta collapse at normalized beta = 4.2 approaching the no-wall limit, while normalized beta greater than 5.5 has been reached in these plasmas with n = 1 active control, in agreement with single-mode RWM theory. Advanced state-space control algorithms proposed for RWM control in ITER theoretically yield significant stabilization improvements. Values of relative phase between the measured n = 1 mode and the applied correction field that experimentally produce stability/instability agree with theory. Experimental mode destabilization occurs over a large range of plasma rotation, challenging the notion of a simple scalar critical rotation speed defining marginal stability. Stability calculations including kinetic modifications to ideal theory are applied to marginally stable experimental equilibria. Plasma rotation and collisionality variations are examined in the calculations. Intermediate rotation levels are less stable, consistent with experimental observations. Trapped ion resonances play a key role in this result. Recent experiments have demonstrated magnetic braking by non-resonant n = 2 fields. The observed rotation damping profile is broader than found for n = 3 fields. Increased ion temperature in the region of maximum braking torque increases the observed rate of rotation damping, consistent with theory.

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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.

The plasma response to external resonant magnetic perturbations is measured as a function of stability of the resistive wall mode ͑RWM͒. The magnetic perturbations are produced with a flexible, high-speed waveform generator that is preprogrammed to drive an in-vessel array of 30 independent control coils and to produce an m/nϭ3/1 helical field. Both quasi-static and ''phase-flip'' magnetic perturbations are applied to time-evolving discharges in order to observe the dynamical response of the plasma as a function of RWM stability. The evolving stability of the RWM is estimated using equilibrium reconstructions and ideal stability computations, facilitating comparison with theory. The plasma resonant response depends upon the evolution of the edge safety factor, q*, and the plasma rotation. For discharges adjusted to maintain relatively constant edge safety factor, q* Ͻ3, the amplitude of the plasma response to a quasistatic field perturbation does not vary strongly near marginal stability and is consistent with the Fitzpatrick-Aydemir equations with high viscous dissipation. Applying ''phase-flip'' magnetic perturbations that rapidly change toroidal phase by 180°allows observation of the time scale for the plasma response to realign with the applied perturbation. This phase realignment time increases at marginal stability, as predicted by theory. This effect is easily measured and suggests that the response to time-varying external field perturbations may be used to detect the approach to RWM instability.

Research on the National Spherical Torus Experiment, NSTX, targets physics understanding needed for extrapolation to a steady-state ST Fusion Nuclear Science Facility, pilot plant, or DEMO. The unique ST operational space is leveraged to test physics theories for next-step tokamak operation, including ITER. Present research also examines implications for the coming device upgrade, NSTX-U. An energy confinement time, τ E , scaling unified for varied wall conditions exhibits a strong improvement of B T τ E with decreased electron collisionality, accentuated by lithium (Li) wall conditioning. This result is consistent with nonlinear microtearing simulations that match the experimental electron diffusivity quantitatively and predict reduced electron heat transport at lower collisionality. Beam-emission spectroscopy measurements in the steep gradient region of the pedestal indicate the poloidal correlation length of turbulence of about ten ion gyroradii increases at higher electron density gradient and lower T i gradient, consistent with turbulence caused by trapped electron instabilities. Density fluctuations in the pedestal top region indicate ion-scale microturbulence compatible with ion temperature gradient and/or kinetic ballooning mode instabilities. Plasma characteristics change nearly continuously with increasing Li evaporation and edge localized modes (ELMs) stabilize due to edge density gradient alteration. Global mode stability studies show stabilizing resonant kinetic effects are enhanced at lower collisionality, but in stark contrast have almost no dependence on collisionality when the plasma is off-resonance. Combined resistive wall mode radial and poloidal field sensor feedback was used to control n = 1 perturbations and improve stability. The disruption probability due to unstable resistive wall modes (RWMs) was surprisingly reduced at very high β N /l i > 10 consistent with low frequency magnetohydrodynamic spectroscopy measurements of mode stability. Greater instability seen at intermediate β N is consistent with decreased kinetic RWM stabilization. A model-based RWM state-space controller produced long-pulse discharges exceeding β N = 6.4 and β N /l i = 13. Precursor analysis shows 96.3% of disruptions can be predicted with 10 ms warning and a false positive rate of only 2.8%. Disruption halo currents rotate toroidally and can have significant toroidal asymmetry. of this phenomenon in designing future RF systems. The snowflake divertor configuration enhanced by radiative detachment showed large reductions in both steady-state and ELM heat fluxes (ELMing peak values down from 19 MW m −2 to less than 1.5 MW m −2 ). Toroidal asymmetry of heat deposition was observed during ELMs or by 3D fields. The heating power required for accessing H-mode decreased by 30% as the triangularity was decreased by moving the X-point to larger radius, consistent with calculations of the dependence of E × B shear in the edge region on ion heat flux and X-point radius. Co-axial helicity injection reduced the induct...

In order to achieve the highest plasma pressure limits in ITER, resistive wall kink mode stabilization is required. A novel resistive wall mode linear observer and feedback controller designed using model reduction and optimal control theory and employing only proportional gain are described here that allow operation of ITER up to C β = 86% of the ideal wall limit using the present design external control coils. The full VALEN finite element ITER model containing ∼3000 modes was reduced to a minimum of 8 modes making real-time controller implementation possible. We find an order of magnitude reduction of the required control coil current and voltage in the presence of white noise from the no-wall limit to the optimal feedback system performance limit as compared with a traditional, classical controller.

Fundamental theory, experimental observations, and modeling of resistive wall mode (RWM) dynamics and active feedback control are reported. In the RWM, the plasma responds to and interacts with external current-carrying conductors. Although this response is complex, it is still possible to construct simple but accurate models for kink dynamics by combining separate determinations for the external currents, using the VALEN code, and for the plasma's inductance matrix, using an MHD code such as DCON. These computations have been performed for wall-stabilized kink modes in the HBT-EP device, and they illustrate a remarkable feature of the theory: when the plasma's inductance matrix is dominated by a single eigenmode and when the surrounding current-carrying structures are properly characterized, then the resonant kink response is represented by a small number of parameters. In HBT-EP, RWM dynamics are studied by programming quasi-static and rapid "phase-flip" changes of the external magnetic perturbation and directly measuring the plasma response as a function of kink stability and plasma rotation. The response evolves in time, is easily measured, and involves excitation of both the wall-stabilized kink and the RWM. High-speed, active feedback control of the RWM using VALEN-optimized mode control techniques and high-throughput digital processors is also reported. Using newly-installed control coils that directly couple to the plasma surface, experiments demonstrate feedback mode suppression in rapidly rotating plasmas near the ideal wall stability limit.

The mission of the National Spherical Torus Experiment (NSTX) is the demonstration of the physics basis required to extrapolate to the next steps for the spherical torus (ST), such as a plasma facing component test facility (NHTX) or an ST based component test facility (ST-CTF), and to support ITER. Key issues for the ST are transport, and steady state high β operation. To better understand electron transport, a new high-k scattering diagnostic was used extensively to investigate electron gyro-scale fluctuations with varying electron temperature gradient scale length. Results from n = 3 braking studies are consistent with the flow shear dependence of ion transport. New results from electron Bernstein wave emission measurements from plasmas with lithium wall coating applied indicate transmission efficiencies near 70% in H-mode as a result of reduced collisionality. Improved coupling of high harmonic fast-waves has been achieved by reducing the edge density relative to the critical density for surface wave coupling. In order to achieve high bootstrap current fraction, future ST designs envision running at very high elongation. Plasmas have been maintained on NSTX at very low internal inductance l i ∼ 0.4 with strong shaping (κ ∼ 2.7, δ ∼ 0.8) with β N approaching the with-wall β-limit for several energy confinement times. By operating at lower collisionality in this regime, NSTX has achieved record non-inductive current drive fraction f NI ∼ 71%. Instabilities driven by super-Alfvénic ions will be an important issue for all burning plasmas, including ITER. Fast ions from NBI on NSTX are super-Alfvénic. Linear toroidal Alfvén eigenmode thresholds and appreciable fast ion loss during multi-mode bursts are measured and these results are compared with theory. The impact of n > 1 error fields on stability is an important result for ITER. Resistive wall mode/resonant field amplification feedback combined with n = 3 error field control was used on NSTX to maintain plasma rotation with β above the no-wall limit. Other highlights are results of lithium coating experiments, momentum confinement studies, scrape-off layer width scaling, demonstration of divertor heat load mitigation in strongly shaped plasmas and coupling of coaxial helicity injection plasmas to ohmic heating ramp-up. These results advance the ST towards next step fusion energy devices such as NHTX and ST-CTF.