One modeling framework for integrated tasks (OMFIT) is a comprehensive integrated modeling framework which has been developed to enable physics codes to interact in complicated workflows, and support scientists at all stages of the modeling cycle. The OMFIT development follows a unique bottom-up approach, where the framework design and capabilities organically evolve to support progressive integration of the components that are required to accomplish physics goals of increasing complexity. OMFIT provides a workflow for easily generating full kinetic equilibrium reconstructions that are constrained by magnetic and motional Stark effect measurements, and kinetic profile information that includes fast-ion pressure modeled by a transport code. It was found that magnetic measurements can be used to quantify the amount of anomalous fast-ion diffusion that is present in DIII-D discharges, and provide an estimate that is consistent with what would be needed for transport simulations to match the measured neutron rates. OMFIT was used to streamline edge-stability analyses, and evaluate the effect of resonant magnetic perturbation (RMP) on the pedestal stability, which have been found to be consistent with the experimental observations. The development of a five-dimensional numerical fluid model for estimating the effects of the interaction between magnetohydrodynamic (MHD) and microturbulence, and its systematic verification against analytic models was also supported by the framework. OMFIT was used for optimizing an innovative high-harmonic fast wave system proposed for DIII-D. For a parallel refractive index > ∥ n 3, the conditions for strong electron-Landau damping were found to be independent of launched ∥ n and poloidal angle. OMFIT has been the platform of choice for developing a neural-network based approach to efficiently perform a non-linear multivariate regression of local transport fluxes as a function of local dimensionless parameters. Transport predictions for thousands of DIII-D discharges showed excellent agreement with the power balance calculations across the whole plasma radius and over a broad range of operating Nuclear Fusion

The goal of the Lower Hybrid Current Drive (LHCD) system on the Alcator C-Mod tokamak is to investigate current profile control under plasma conditions relevant to future devices such as ITER and DEMO. Experimental observations of an LHCD "density limit" for C-Mod are presented in this paper. Bremsstrahlung emission from relativistic fast electrons in the core plasma drops suddenly above line averaged densities of 10 20 m −3 (ω/ω LH ∼3-4), well below the density limit previously observed on other experiments (ω/ω LH ∼ 2). Electric currents flowing through the scrape off layer (SOL) between the inner and outer divertors increase dramatically across the same density range that the core bremsstrahlung emission drops precipitously. These experimental x-ray data are compared to both conventional modeling, which gives poor agreement with experiment above the density limit, and a model including collisional absorption in the SOL, which dramatically improves agreement with experiment above the observed density limit. These results show that strong absorption of LH waves in the SOL is possible on a high density tokamak and the SOL must be included in simulations of LHCD at high density.

This paper reports on disruption prediction using a shallow machine learning method known as a random forest, trained on large databases containing only plasma parameters that are available in real-time on Alcator C-Mod, DIII-D, and EAST. The database for each tokamak contains parameters sampled ∼10 6 times throughout ∼10 4 discharges (disruptive and nondisruptive) over the last four years of operation. It is found that a number of parameters (e.g. P rad /P input , i , n/n G , B n=1 /B 0 ) exhibit changes in aggregate as a disruption is approached on one or more of these tokamaks. However, for each machine, the most useful parameters, as well as the details of their precursor behaviors, are markedly different. When the prediction problem is framed using a binary classification scheme to discriminate between time slices 'close to disruption' and 'far from disruption', it is found that the prediction algorithms differ substantially in performance among the three machines on a time slice-by-time slice basis, but have similar disruption detection rates (∼80%-90%) on a shot-by-shot basis after appropriate optimisation. This could have important implications for disruption prediction and avoidance on ITER, for which development of a training database of disruptions may be infeasible. The algorithm's output is interpretable using a method that identifies the most strongly contributing input signals, which may have implications for avoiding disruptive scenarios. To further support its real-time capability, successful applications in inter-shot and real-time environments on EAST and DIII-D are also discussed.

Recent EAST/DIII-D joint experiments on the high poloidal beta tokamak regime in DIII-D have demonstrated fully noninductive operation with an internal transport barrier (ITB) at large minor radius, at normalized fusion performance increased by ⩾30% relative to earlier work (Politzer et al 2005 Nucl. Fusion 45 417). The advancement was enabled by improved understanding of the 'relaxation oscillations', previously attributed to repetitive ITB collapses, and of the fast ion behavior in this regime. It was found that the 'relaxation oscillations' are coupled core-edge modes amenable to wall-stabilization, and that fast ion losses which previously dictated a large plasma-wall separation to avoid wall over-heating, can be reduced to classical levels with sufficient plasma density. By using optimized waveforms of the plasma-wall separation and plasma density, fully noninductive plasmas have been sustained for long durations with excellent energy confinement quality, bootstrap fraction ⩾80%, β ⩽ 4 N , β ⩾ 3 P , and β ⩾ % 2 T . These results bolster the applicability of the high poloidal beta tokamak regime toward the realization of a steady-state fusion reactor.

Fast waves at frequencies far above the ion cyclotron frequency and approaching the lower hybrid frequency (also called 'helicons' or ‘whistlers’) have application to off-axis current drive in tokamaks with high electron beta. The high frequency causes the whistler-like behaviour of the wave power nearly following field lines, but with a small radial component, so the waves spiral slowly towards the plasma centre. The high frequency also contributes to strong damping. Modelling predicts robust off-axis current drive with good efficiency compared to alternatives in high performance discharges in DIII-D and Fusion Nuclear Science Facility (FNSF) when the electron beta is above about 1.8%. Detailed analysis of ray behaviour shows that ray trajectories and damping are deterministic (that is, not strongly affected by plasma profiles or initial ray conditions), unlike the chaotic ray behaviour in lower frequency fast wave experiments. Current drive was found to not be sensitive to the launched value of the parallel index of refraction n‖, so wave accessibility issues can be reduced. Use of a travelling wave antenna provides a very narrow n‖spectrum, which also helps avoid accessibility problems.