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The DIII-D tokamak has elucidated crucial physics and developed projectable solutions for ITER and fusion power plants in the key areas of core performance, boundary heat and particle transport, and integrated scenario operation, with closing the core-edge integration knowledge gap being the overarching mission. New experimental validation of high-fidelity, multi-channel, non-linear gyrokinetic turbulent transport models for ITER provides strong confidence it will achieve Q ⩾ 10 operation. Experiments identify options for easing H-mode access in hydrogen, and give new insight into the isotopic dependence of transport and confinement. Analysis of 2,1 islands in unoptimized low-torque IBS demonstration discharges suggests their onset time occurs randomly in the constant β phase, most often triggered by non-linear 3-wave coupling, thus identifying an NTM seeding mechanism to avoid. Pure deuterium SPI for disruption mitigation is shown to provide favorable slow cooling, but poor core assimilation, suggesting paths for improved SPI on ITER. At the boundary, measured neutral density and ionization source fluxes are strongly poloidally asymmetric, implying a 2D treatment is needed to model pedestal fuelling. Detailed measurements of pedestal and SOL quantities and impurity charge state radiation in detached divertors has validated edge fluid modelling and new self-consistent ‘pedestal-to-divertor’ integrated modeling that can be used to optimize reactors. New feedback adaptive ELM control minimizes confinement reduction, and RMP ELM suppression with sustained high core performance was obtained for the first time with the outer strike point in a W-coated, compact and unpumped small-angle slot divertor. Advances have been made in integrated operational scenarios for ITER and power plants. Wide pedestal intrinsically ELM-free QH-modes are produced with more reactor-relevant conditions, Low torque IBS with W-equivalent radiators can exhibit predator-prey oscillations in T _e and radiation which need control. High- β _P scenarios with q _min > 2, q _95 –7.9, β _N > 4, β _T –3.3% and H _98y2 > 1.5 are sustained with high density ( $\bar n$ = 7E19 m ^−3 , f _G –1) for 6 τ _E , improving confidence in steady-state tokamak reactors. Diverted NT plasmas achieve high core performance with a non-ELMing edge, offering a possible highly attractive core-edge integration solution for reactors.

This work presents a study of plasma transport at low aspect ratio on the National Spherical Torus Experiment tokamak, where the turbulent and neoclassical energy fluxes calculated by the quasilinear Trapped Gyro Landau Fluid (TGLF) model and the multi species drift-kinetic Neoclassical solver (NEO) are validated against experimental data. The turbulent energy transport of two plasma discharges, one in the L-mode confinement regime and another in the H-mode regime, is dominated by electrostatic drift-wave instabilities, while the ion heat transport has a significant neoclassical contribution. The data analysis workflow is described in detail to understand how the variations of mapping and fitting of experimental data affect the power balance solution and subsequent flux-matching plasma profile predictions with the TGYRO solver. On average, the predicted plasma profiles are consistent with experimental data. However, the solutions are sensitive to various input parameters, including boundary conditions, and the electron-ion coupling. Linear gyrokinetic stability analysis demonstrates close agreement of the real frequencies of unstable modes between TGLF and CGYRO gyrokinetic simulations, but higher growth rates are predicted by TGLF, especially for the H-mode case. Estimates of the low- k modes’ contributions to the total flux are consistent with linear stability analysis and the E × B suppression of turbulence in TGLF simulations with the SAT1 saturation model, while the SAT2 saturation model over-predicts the low- k modes’ contribution in the H-mode case. Moreover, the results with SAT1 model are consistent with power balance analysis, which indicates only neoclassical ion energy fluxes inside ρ

On DIII-D, the high scenario has an internal transport barrier (ITB), , , and very high normalized confinement . Recently, plasmas starting with these conditions have been dynamically driven to and , where we find the ITB and high performance persist for five energy confinement times. These conditions are projected to meet the ITER steady-state goal of Q  =  5. The ITB is maintained at lower with a strong reverse shear, consistent with predictions that negative central shear can lower the threshold for the ITB. There are two observed confinement states in the high scenario: H-mode confinement state with a high edge pedestal, and an enhanced confinement state with a low pedestal and an ITB. It has been observed in a scan of external resonant magnetic perturbation amplitude that when there are no large type-I ELMs, there is no transition to enhanced confinement. This is consistent with the proposed mechanism for ITB formation being a type-I ELM. Quasilinear gyro-Landau fluid predictive modeling of ITER suggests that only a modest reverse shear is required to achieve the ITB formation necessary for Q  =  5 when electromagnetic physics including the kinetic ballooning mode (KBM) is incorporated. [ABSTRACT FROM AUTHOR]

The ‘Super H-Mode’ regime is predicted to enable pedestal height and fusion performance substantially higher than standard H-Mode operation. This regime exists due to a bifurcation of the pedestal pressure, as a function of density, that is predicted by the EPED model to occur in strongly shaped plasmas above a critical pedestal density. Experiments on Alcator C-Mod and DIII-D have achieved access to the Super H-Mode (and Near Super H) regime, and obtained very high pedestal pressure, including the highest achieved on a tokamak (p ped ~ 80 kPa) in C-Mod experiments operating near the ITER magnetic field. DIII-D Super H experiments have demonstrated strong performance, including the highest stored energy in the present configuration of DIII-D (W ~ 2.2–3.2 MJ), while utilizing only about half of the available heating power (Pheat ~ 7–12 MW). These DIII-D experiments have obtained the highest value of peak fusion gain, QDT,equiv ~ 0.5, achieved on a medium scale (R  <  2 m) tokamak. Sustained high performance operation (βN ~ 2.9, H98 ~ 1.6) has been achieved utilizing n  =  3 magnetic perturbations for density and impurity control. Pedestal and global confinement has been maintained in the presence of deuterium and nitrogen gas puffing, which enables a more radiative divertor condition. A pair of simple performance metrics is developed to assess and compare regimes. Super H-Mode access is predicted for ITER and expected, based on both theoretical prediction and observed normalized performance, to allow ITER to achieve its goals (Q  =  10) at Ip  <  15 MA, and to potentially enable more compact, cost effective pilot plant and reactor designs. [ABSTRACT FROM AUTHOR]

In this paper we show that the local density gradient in the plasma core depends on the calculated mode-frequency of the most unstable linear mode and reaches a maximum when this frequency is close to zero. Previous theoretical and experimental work on AUG has shown that the ratio of electron to ion temperature, and as such the frequency of the dominant linear gyrokinetic mode, affects the local density gradient close to (Fable et al 2010 Plasma Phys. Control. Fusion52 015007, Angioni et al 2011 Nucl. Fusion51 023006). On DIII-D we find that by adding electron cyclotron heating, we modify the dominant unstable linear gyrokinetic mode from an ion temperature gradient (ITG) mode to a trapped electron mode (TEM), which means that the frequency of the dominant mode changes sign (from the ion to the electron direction). Local density peaking around mid-radius increases by 50% right around the cross-over between the ITG and TEM regimes. By comparing how the particle flux changes, through the derivative of the electron density, ne, with respect to time, , we find that the particle flux also exhibits the same trend versus mode frequency. As a result, we find that the changes in local particle transport are inversely proportional to the changes in electron density, indicating that the changes are driven by a change in thermo-diffusive pinch. [ABSTRACT FROM AUTHOR]

In this paper, we show that the often observed ‘density pump-out’ with electron cyclotron heating (ECH) (Angioni et al 2004 Nucl. Fusion44 827, 2009 Plasma Phys. Control. Fusion51 124017), at low density and/or collisionality is the result of an increase in turbulence drive at the plasma edge (Angioni et al 2004 Nucl. Fusion44 827, 2005 Phys. Plasmas12 040701, Mordijck et al 2015 Nucl. Fusion43 113025). Prior results were limited to comparison of steady-state conditions, before and after the ECH was applied, and thus failed to capture the dynamics of the density pump-out. In this paper, we find, similar to prior results, that when the plasma reaches a new equilibrium after ECH is applied, gyro-kinetic simulations indicate that the plasma has transitioned from the ion temperature gradient (ITG) to a trapped electron mode (TEM) regime around mid-radius. However, this transition from ITG to TEM only occures in the core after 100 ms. The pump-out on the other hand, starts immediately and is strongest around . Linear gyrokinetic simulations with TGLF show that there is an increase in turbulence drive simultaneously with the density pump-out and the doppler backscattering (DBS) measures an instant increase in density fluctuations at the same radial location. On the other hand, around mid-radius the DBS measures no increase in density fluctuations. All these calculations along with experimental measurements show that the density pump-out is not the result of a change in turbulence type (i.e. not caused by a change from ITG to TEM), but the result of a change in turbulence drive (an increase in linear growth rates), which is later followed by the ITG to TEM transition. This highlights the need for studying not just the equilibrium conditions after a transition, but also the time-dependent changes. [ABSTRACT FROM AUTHOR]

Transport modeling of a proposed ITER steady-state scenario based on DIII-D high poloidal-beta () discharges finds that ITB formation can occur with either sufficient rotation or a negative central shear q-profile. The high scenario is characterized by a large bootstrap current fraction (80%) which reduces the demands on the external current drive, and a large radius internal transport barrier which is associated with excellent normalized confinement. Modeling predictions of the electron transport in the high scenario improve as approaches levels similar to typical existing models of ITER steady-state and the ion transport is turbulence dominated. Typical temperature and density profiles from the non-inductive high scenario on DIII-D are scaled according to 0D modeling predictions of the requirements for achieving a steady-state fusion gain in ITER with ‘day one’ heating and current drive capabilities. Then, TGLF turbulence modeling is carried out under systematic variations of the toroidal rotation and the core q-profile. A high bootstrap fraction, high scenario is found to be near an ITB formation threshold, and either strong negative central magnetic shear or rotation in a high bootstrap fraction are found to successfully provide the turbulence suppression required to achieve . [ABSTRACT FROM AUTHOR]

Fusion whole device modeling simulations require comprehensive models that are simultaneously physically accurate, fast, robust, and predictive. In this paper we describe the development of two neural-network (NN) based models as a means to perform a snon-linear multivariate regression of theory-based models for the core turbulent transport fluxes, and the pedestal structure. Specifically, we find that a NN-based approach can be used to consistently reproduce the results of the TGLF and EPED1 theory-based models over a broad range of plasma regimes, and with a computational speedup of several orders of magnitudes. These models are then integrated into a predictive workflow that allows prediction with self-consistent core-pedestal coupling of the kinetic profiles within the last closed flux surface of the plasma. The NN paradigm is capable of breaking the speed-accuracy trade-off that is expected of traditional numerical physics models, and can provide the missing link towards self-consistent coupled core-pedestal whole device modeling simulations that are physically accurate and yet take only seconds to run. [ABSTRACT FROM AUTHOR]

Energy transport analyses of the DIII-D high- EAST-demonstration discharges have been performed using the TGYRO transport package with the TGLF turbulent and NEO neoclassical transport models under the OMFIT integrated modeling framework. Ion energy transport is shown to be dominated by neoclassical transport and ion temperature profiles predicted by TGYRO agree closely with the experimental measured profiles for these high- discharges. Ion energy transport is largely insensitive to reductions in the flow shear stabilization. The Shafranov shift is shown to play a role in the suppression of the ion turbulent energy transport below the neoclassical level. Electron turbulent energy transport is under-predicted by TGLF and a significant shortfall in the electron energy transport over the whole core plasma is found with TGLF predictions for these high- discharges. TGYRO can successfully predict the experimental ion and electron temperature profiles by artificially increasing the saturated turbulence level for ETG driven modes used in TGLF. [ABSTRACT FROM AUTHOR]

In this paper we show how changes in toroidal rotation, by controlling the injected torque, affect particle transport and confinement. The toroidal rotation is altered using the co- and counter neutral beam injection (NBI) in low collisionality H-mode plasmas on DIII-D (Luxon 2002 Nucl. Fusion42 614) with dominant electron cyclotron heating (ECH). We find that there is no correlation between the toroidal rotation shear and the inverse density gradient, which is observed on AUG when is varied using ECH (Angioni et al 2011 Phys. Rev. Lett. 107 215003). In DIII-D, we find that in a discharge with balanced torque injection, the shear is smaller than the linear gyrokinetic growth rate for small for –0.85. This results in lower particle confinement. In the co- and counter- injected discharges the shear is larger or close to the linear growth rate at the plasma edge and both configurations have higher particle confinement. In order to measure particle transport, we use a small periodic perturbative gas puff. This gas puff perturbs the density profiles and allows us to extract the perturbed diffusion and inward pinch coefficients. We observe a strong increase in the inward particle pinch in the counter-torque injected plasma. Finally, the calculated quasi-linear particle flux, nor the linear growth rates using TGLF (Staebler et al 2005 Phys. Plasmas12 102508) agree with experimental observations. [ABSTRACT FROM AUTHOR]

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. Fusion45 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%, , , and . These results bolster the applicability of the high poloidal beta tokamak regime toward the realization of a steady-state fusion reactor. [ABSTRACT FROM AUTHOR]

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 , the conditions for strong electron-Landau damping were found to be independent of launched 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 regimes. Concerning predictive transport simulations, the framework made possible the design and automation of a workflow that enables self-consistent predictions of kinetic profiles and the plasma equilibrium. It is found that the feedback between the transport fluxes and plasma equilibrium can significantly affect the kinetic profiles predictions. Such a rich set of results provide tangible evidence of how bottom-up approaches can potentially provide a fast track to integrated modeling solutions that are functional, cost-effective, and in sync with the research effort of the community. [ABSTRACT FROM AUTHOR]

In a continuing effort to validate turbulent transport models, the electron energy flux has been probed as a function of electron temperature gradient on the DIII-D tokamak. In the scan of gradient, a critical electron temperature gradient has been found in the electron heat fluxes and stiffness at various radii in L-mode plasmas. The TGLF reduced turbulent transport model (Staebler et al 2007 Phys. Plasmas14 055909) and full gyrokinetic GYRO model (Candy and Waltz 2003 J. Comput. Phys. 186 545) recover the general trend of increasing electron energy flux with increasing electron temperature gradient scale length, but they do not predict the absolute level of transport at all radii and gradients. Comparing the experimental observations of incremental (heat pulse) diffusivity and stiffness to the models’ reveals that TGLF reproduces the trends in increasing diffusivity and stiffness with increasing electron temperature gradient scale length with a critical gradient behavior. The critical gradient of TGLF is found to have a dependence on q95, contrary to the independence of the experimental critical gradient from q95. [ABSTRACT FROM AUTHOR]

Magnetic measurements together with kinetic profile and motional Stark effect measurements are used in full kinetic equilibrium reconstructions to test the Sauter and NEO bootstrap current models in a DIII-D high- EAST-demonstration experiment. This aims at developing on DIII-D a high bootstrap current scenario to be extended on EAST for a demonstration of true steady-state at high performance and uses EAST-similar operational conditions: plasma shape, plasma current, toroidal magnetic field, total heating power and current ramp-up rate. It is found that the large edge bootstrap current in these high- plasmas allows the use of magnetic measurements to clearly distinguish the two bootstrap current models. In these high collisionality and high- plasmas, the Sauter model overpredicts the peak of the edge current density by about 30%, while the first-principle kinetic NEO model is in close agreement with the edge current density of the reconstructed equilibrium. These results are consistent with recent work showing that the Sauter model largely overestimates the edge bootstrap current at high collisionality (Belli et al 2014 Plasma Phys. Control. Fusion56 045006). [ABSTRACT FROM AUTHOR]

An Electron Cyclotron Emission Imaging (ECEI) data analysis module has been developed for the OMFIT platform to accommodate the needs of users at the DIII-D tokamak for physics applications. The user can easily access the ECEI spatial observation windows in the plasma that are calculated based on the automatically retrieved hardware setup and available DIII-D equilibria, perform spectral analysis, and obtain 2D electron temperature fluctuation images. The module provides a powerful data post-processing package for extracting important physics parameters from the 2D measurements, including the radial structure and poloidal mode number of Alfven eigenmodes, as well as the frequency-vs-wavenumber dispersion relationship of broadband MHD. The module propagates characterized synthetic fluctuations for the user, so one can perform forward modeling tasks with simple analytical fluctuations.

The Ion Cyclotron Emission (ICE) diagnostic on the DIII-D tokamak consists of two outboard midplane systems. In the first system, straps of an ion cyclotron range of frequencies antenna are configured as receiving antennas. For the second system, dedicated magnetic probes incorporated into the outer wall of carbon tiles have recently been restored. These systems collected a large set of radio frequency measurements in the 2015-2018 experimental campaigns by digitizing signals at 200 MSamples/s for ∼5 s per discharge. Each shot typically yields 32 GB of data; techniques for successful handling and analysis of this challengingly large dataset are discussed. The raw voltage fluctuations (<0.2 V and <1 mW) are analyzed in frequency space via fast Fourier transforms. Signals can be analyzed between 1 and 200 MHz with appropriate filtering and aliasing; this frequency range is limited by DC breaks used to provide 5 kV DC isolation. These high-frequency signals are driven by energetic ions and electrons. In particular, energetic-ion-driven ICE occurs at harmonics of the ion cyclotron frequency, enabling the frequency to be mapped to lab space via equilibrium reconstruction. In many DIII-D plasmas, ICE is emitted from the radial center of the plasma.

An innovative millimeter wave diagnostic is proposed to measure the local magnetic field and edge current as a function of the minor radius in the tokamak pedestal region. The idea is to identify the direction of minimum reflectivity at the O-mode cutoff layer. Correspondingly, the transmissivity due to O-X mode conversion is maximum. That direction, and the angular map of reflectivity around it, contains information on the magnetic field vector B at the cutoff layer. Probing the plasma with different wave frequencies provides the radial profile of B. Full-wave finite-element simulations are presented here in 2D slab geometry. Modeling confirms the existence of a minimum in reflectivity that depends on the magnetic field at the cutoff, as expected from mode conversion physics, giving confidence in the feasibility of the diagnostic. The proposed reflectometric approach is expected to yield superior signal-to-noise ratio and to access wider ranges of density and magnetic field, compared with related radiometric techniques that require the plasma to emit electron Bernstein waves. Due to computational limitations, frequencies of 10-20 GHz were considered in this initial study. Frequencies above the edge electron-cyclotron frequency (f > 28 GHz here) would be preferable for the experiment, because the upper hybrid resonance and right cutoff would lie in the plasma, and would help separate the O-mode of interest from spurious X-waves.