A heuristic criterion for the full suppression of an NTM was formulated as ηNTM ≡ j CD,max/j BS ⩾ 1.2 (Zohm et al 2005 J. Phys. Conf. Ser. 25 234), where j CD,max is the maximum in the driven current density profile applied to stabilize the mode and j BS is the local bootstrap current density. In this work we subject this criterion to a systematic theoretical analysis on the basis of the generalized Rutherford equation. Taking into account only the effect of j CD inside the island, a new criterion for full suppression by a minimum applied total current is obtained in the form of a maximum allowed value for the width of the driven current, w dep, combined with a required minimum for the total driven current in the form of w depηNTM, where both limits depend on the marginal and saturated island sizes. These requirements can be relaxed when additional effects are taken into account, such as a change in the stability parameter Δ′ from the current driven outside the island, power modulation, the accompanying heating inside the island or when the current drive is applied preemptively. When applied to ITER scenario 2, the requirement for full suppression of either the 3/2 or 2/1 NTM becomes w dep ≲ 5 cm and w depηNTM ≳ 5 cm in agreement with (Sauter et al 2010 Plasma Phys. Control. Fusion 52 025002). Optimization of the ITER ECRH Upper Port Launcher design towards minimum required power for full NTM suppression requires an increase in the toroidal injection angle of the lower steering mirror of several degrees compared with its present design value, while for the upper steering mirror the present design value is close to the optimum.

Full wave simulations of fusion plasmas show a direct correlation between the location of the fast-wave cut-off, radiofrequency (RF) field amplitude in the scrape-off layer (SOL) and the RF power losses in the SOL observed in the National Spherical Torus eXperiment (NSTX). In particular, the RF power losses in the SOL increase significantly when the launched waves transition from evanescent to propagating in that region. Subsequently, a large amplitude electric field occurs in the SOL, driving RF power losses when a proxy collisional loss term is added. A 3D reconstruction of absorbed power in the SOL is presented showing agreement with the RF experiments in NSTX. Loss predictions for the future experiment NSTX-Upgrade (NSTX-U) are also obtained and discussed.

The paraxial WKB code TORBEAM [E. Poli et al., Comp. Phys. Comm. 136 (2001), 90] is widely used for the description of electron-cyclotron waves in fusion plasmas, retaining diffraction effects through the solution of a set of ordinary differential equations. With respect to its original form, the code has undergone significant transformations and extensions, in terms of both the physical model and the spectrum of applications. The code has been rewritten in Fortran 90 and transformed into a library, which can be called from within different (not necessarily Fortran-based) workflows. The models for both absorption and current drive have been extended, including e.g. fully-relativistic calculation of the absorption coefficient, momentum conservation in electron-electron collisions and the contribution of more than one harmonic to current drive. The code can be run also for reflectometry applications, with relativistic corrections for the electron mass. Formulas that provide the coupling between the reflected beam and the receiver have been developed. Accelerated versions of the code are available, with the reduced physics goal of inferring the location of maximum absorption (including or not the total driven current) for a given setting of the launcher mirrors. Optionally, plasma volumes within given flux surfaces and corresponding values of minimum and maximum magnetic field can be provided externally to speed up the calculation of full driven-current profiles. These can be employed in real-time control algorithms or for fast data analysis.

The National Spherical Torus Experiment (NSTX) has undergone a major upgrade, and the NSTX Upgrade (NSTX-U) Project was completed in the summer of 2015. NSTX-U first plasma was subsequently achieved, diagnostic and control systems have been commissioned, the H-mode accessed, magnetic error fields identified and mitigated, and the first physics research campaign carried out. During ten run weeks of operation, NSTX-U surpassed NSTX record pulse-durations and toroidal fields (TF), and high-performance ~1 MA H-mode plasmas comparable to the best of NSTX have been sustained near and slightly above the n  =  1 no-wall stability limit and with H-mode confinement multiplier H98y,2 above 1. Transport and turbulence studies in L-mode plasmas have identified the coexistence of at least two ion-gyro-scale turbulent micro-instabilities near the same radial location but propagating in opposite (i.e. ion and electron diamagnetic) directions. These modes have the characteristics of ion-temperature gradient and micro-tearing modes, respectively, and the role of these modes in contributing to thermal transport is under active investigation. The new second more tangential neutral beam injection was observed to significantly modify the stability of two types of Alfven eigenmodes. Improvements in offline disruption forecasting were made in the areas of identification of rotating MHD modes and other macroscopic instabilities using the disruption event characterization and forecasting code. Lastly, the materials analysis and particle probe was utilized on NSTX-U for the first time and enabled assessments of the correlation between boronized wall conditions and plasma performance. These and other highlights from the first run campaign of NSTX-U are described.

Time-dependent simulations are used to evolve plasma discharges in combination with a modified Rutherford equation for calculation of neoclassical tearing mode (NTM) stability in response to electron cyclotron (EC) feedback control in ITER. The main application of this integrated approach is to support the development of control algorithms by analyzing the plasma response with physics-based models and to assess how uncertainties in the detection of the magnetic island and in the EC alignment affect the ability of the ITER EC system to fulfill its purpose. Simulations indicate that it is critical to detect the island as soon as possible, before its size exceeds the EC deposition width, and that maintaining alignment with the rational surface within half of the EC deposition width is needed for stabilization and suppression of the modes, especially in the case of modes with helicity (2, 1). A broadening of the deposition profile, for example due to wave scattering by turbulence fluctuations or not well aligned beams, could even be favorable in the case of the (2, 1)−NTM, by relaxing an over-focussing of the EC beam and improving the stabilization at the mode onset. Pre-emptive control reduces the power needed for suppression and stabilization in the ITER baseline discharge to a maximum of 5 MW, which should be reserved and available to the upper launcher during the entire flattop phase. Assuming continuous triggering of NTMs, with preemptive control ITER would be still able to demonstrate a fusion gain of Q = 10.

Scattering effects induced by edge density fluctuations on lower hybrid (LH) wave propagation are investigated. The scattering model used here is based on the work of Bonoli and Ott (1982 Phys. Fluids 25 361). It utilizes an electromagnetic wave kinetic equation solved by a Monte Carlo technique. This scattering model has been implemented in GENRAY, a ray-tracing code which explicitly simulates wave propagation, as well as collisionless and collisional damping processes, over the entire plasma discharge, including the scrape-off layer that extends from the separatrix to the vessel wall. A numerical analysis of the LH wave trajectories and the power deposition profile with and without scattering is presented for Alcator C-Mod discharges. Comparisons between the measured hard x-ray emission on Alcator C-Mod and simulations of the data obtained from the synthetic diagnostic included in the GENRAY/CQL3D package are shown, with and without the combination of scattering and collisional damping. Implications of these results on LH current drive are discussed.

One of the main aims of the ITER electron cyclotron resonance heating (ECRH) system, in particular of the Upper Port Launcher, is the control of magnetohydrodynamics instabilities. This control typically requires non-inductively driven currents with a high degree of localization, i.e. with a very narrow profile. A numerical analysis of the effect of the radial diffusion of the EC driven current carrying electrons has been performed in order to estimate the effectiveness of electron cyclotron current drive (ECCD) for neoclassical tearing mode (NTM) stabilization. In particular, Fokker–Planck calculations including radial diffusion for the case of the ITER ECRH Upper Port Launcher are presented. These show a significant decrease in the local current density when radial diffusion at a rate of only 1 m2 s−1 is included and consequently a broadening of the profile with a drop in the predicted efficiency for NTM control. Furthermore, it is shown that a simple formula combining the effect of the radial diffusion and the width of the EC power deposition profile reproduces quite accurately the maximum EC driven current density, which is the more relevant number in determining the NTM suppression figure of merit, for typical ITER parameters.

In this paper, a reduced model of quasilinear velocity diffusion by a small Larmor radius approximation is derived to couple the Maxwell's equations and the Fokker planck equation self-consistently for ion cyclotron range of frequency waves in a tokamak. The reduced model ensures the important properties of the full model by Kennel-Engelmann diffusion, such as diffusion directions, wave polarizations, and H-theorem. The kinetic energy change (Ẇ ) is used to derive the reduced model diffusion coefficients for the fundamental damping (n=1) and the second harmonic damping (n=2) to the lowest order of the finite Larmor radius expansion. The quasilinear diffusion coefficients are implemented in a coupled code (TORIC-CQL3D) with the equivalent reduced model of dielectric tensor. We also present the simulations of ITER minority heating scenario, in which the reduced model is verified within the allowable errors from the full model results.