Theoretical studies have suggested that nonlinear effects can lead to 'radio frequency (RF) condensation', where an initially broad current profile can coalesce in islands when they reach sufficient width. In suitable conditions, RF condensation can 'self-focus' the driven current to the center of an island, improving stabilization efficiency and reducing control complexity. In unsuitable conditions, the effect can prematurely deplete the RF energy before it reaches the island center, impairing stabilization. It is predicted that the RF condensation effect can significantly impact reactor-scale tokamaks. This paper presents a set of simulations investigating the conditions under which RF condensation might be encountered in present-day tokamaks. For concreteness, the calculations use equilibrium reconstructions for two shots from DIII-D and AUG. The Current Condensation Amid Magnetic Islands (OCCAMI) simulation code has been used for this investigation. The code takes as its input a numerically specified axisymmetric EFIT equilibrium solution, and it perturbatively constructs a 3D field with an island embedded at the appropriate rational surface. In the OCCAMI code, the GENRAY code is used for ray tracing and for calculating the power deposition along a ray trajectory, and GENRAY is coupled self-consistently to a solution of the thermal diffusion equation in the island. The simulation results described in the paper illuminate the conditions required for experimental validation of the theory of RF condensation. The simulations also provide an explanation of why the effect was not noticed in experiments prior to the publication of theoretical papers on the subject.

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 beta 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-beta(P) scenarios with q(min) > 2, q(95)-7.9, beta(N) > 4, beta(T)-3.3% and H-98y2 > 1.5 are sustained with high density ((n) over bar = 7E19 m(-3), f(G)-1) for 6 tau(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.

We have demonstrated for the first time that turbulent plasma density fluctuations in the edge of the DIII-D tokamak are responsible for substantial broadening of an injected microwave beam by successful quantitative comparison between experimental observations and first principles 2D full-wave simulations. The broadening of the beam has important implications for control of tokamak discharges through localized electron cyclotron deposition needed for eliminating magnetohydrodynamic instabilities. This new predictive capability is mandatory to design & operate present & future tokamaks in such a way that microwave heating schemes achieve their intended objectives.

The injection of electron cyclotron (EC) waves fulfills a number of important tasks in nuclear fusion devices for which detailed knowledge of the spatial power deposition profile is critical. This deposition profile is commonly determined using forward models such as beam or ray tracing. Recent numerical and experimental studies have shown that small-angle scattering of the EC beam as it passes through the turbulent plasma edge can cause significant broadening of the effective deposition profile, leading to considerable underestimation of the deposition width by forward methods. However, traditional inverse methods to determine the deposition profile from measurements overestimate the deposition profile width due to transport broadening. In this work, we implement three novel methods to resolve the EC power deposition profile from measurements that counteract transport broadening by simultaneously resolving transport and power deposition. We validate their assumptions and compare the results from these methods to the traditional break-in-slope method as well as to the TORAY ray-tracing code in a set of DIII-D discharges spanning five different confinement modes. We show that the four different inverse methods, novel and established, paint a consistent picture of deposition broadening. Specifically, we show that the measured power deposition profile is between 1.6 and 3.6 times wider than the TORAY profiles. Moreover, we show the considerable consequences that this level of broadening can have for ITER.

A novel quadrature Doppler Backscattering (DBS) system has been developed and optimized for the E-band (60-90 GHz) frequency range using either O-mode or X-mode polarization in DIII-D plasmas. In general, DBS measures the amplitude of density fluctuations and their velocity in the lab frame. The system can simultaneously monitor both low-frequency turbulence (f < 10 MHz) and radiofrequency plasma density fluctuations over a selectable frequency range (20-500 MHz). Detection of high-frequency fluctuations has been demonstrated for low harmonics of the ion cyclotron frequency (e.g., 2fci ∼ 23 MHz) and externally driven high-frequency helicon waves (f = 476 MHz) using an adjustable frequency down conversion system. Importantly, this extends the application of DBS to a high-frequency spectral domain while maintaining important turbulence and flow measurement capabilities. This unique system has low phase noise, good temporal resolution (sub-millisecond), and excellent wavenumber coverage (kθ ∼ 1-20 cm-1 and kr ≲ 30 cm-1). As a demonstration, localized internal DIII-D plasma measurements are presented from turbulence (f ≤ 5 MHz), Alfvenic waves (f ∼ 6.5 MHz), ion cyclotron waves (f ≥ 20 MHz), as well as fluctuations around 476 MHz driven by an external high-power 476 MHz helicon wave antenna. In the future, helicon measurements will be used to validate GENRAY and AORSA modeling tools for prediction of helicon wave propagation, absorption, and current drive location for the newly installed helicon current drive system on DIII-D.

A fast wave interferometer (FWI), which can measure ion mass density, has been developed on DIII-D for its use on future fusion reactors, as well as for the study of ion behavior in current plasma devices. The frequency of the fast waves used for the FWI is around 60 MHz, and require antennas and coaxial cables or waveguides, which, unlike traditional mirror-based optical interferometers, are less susceptible to neutron/gamma-ray radiation and are relatively immune to impurity deposition and erosion as well as alignment issues. The bulk ion density evaluated using FWI show good agreement with that derived from CO2 interferometry within about 15%. When the ion mass density measurement by FWI is combined with an electron density measurement from CO2 interferometry, Z(eff) measurements are also enabled and are in agreement with those from visible Bremsstrahlung measurements. Additionally, large-bandwidth FWI measurements clearly resolve 10-100 kHz coherent modes and demonstrate its potential as a core fluctuation diagnostic, sensitive to both magnetic and ion density perturbations.

DIII-D physics research addresses critical challenges for the operation of ITER and the next generation of fusion energy devices. This is done through a focus on innovations to provide solutions for high performance long pulse operation, coupled with fundamental plasma physics understanding and model validation, to drive scenario development by integrating high performance core and boundary plasmas. Substantial increases in off-axis current drive efficiency from an innovative top launch system for EC power, and in pressure broadening for Alfven eigenmode control from a co-/counter- I p steerable off-axis neutral beam, all improve the prospects for optimization of future long pulse/steady state high performance tokamak operation. Fundamental studies into the modes that drive the evolution of the pedestal pressure profile and electron vs ion heat flux validate predictive models of pedestal recovery after ELMs. Understanding the physics mechanisms of ELM control and density pumpout by 3D magnetic perturbation fields leads to confident predictions for ITER and future devices. Validated modeling of high- Z shattered pellet injection for disruption mitigation, runaway electron dissipation, and techniques for disruption prediction and avoidance including machine learning, give confidence in handling disruptivity for future devices. For the non-nuclear phase of ITER, two actuators are identified to lower the L–H threshold power in hydrogen plasmas. With this physics understanding and suite of capabilities, a high poloidal beta optimized-core scenario with an internal transport barrier that projects nearly to Q = 10 in ITER at ∼ 8 MA was coupled to a detached divertor, and a near super H-mode optimized-pedestal scenario with co- I p beam injection was coupled to a radiative divertor. The hybrid core scenario was achieved directly, without the need for anomalous current diffusion, using off-axis current drive actuators. Also, a controller to assess proximity to stability limits and regulate β N in the ITER baseline scenario, based on plasma response to probing 3D fields, was demonstrated. Finally, innovative tokamak operation using a negative triangularity shape showed many attractive features for future pilot plant operation.

The flux fit method is used to self-consistently estimate the power deposition profile and heat transport profiles from temperature measurements originating from perturbative experiments with a modulated source. This Letter improves on this method by addressing the limitations and assumptions. The most crucial improvement is the additional freedom in the source deposition profile. Allowing for a variable central deposition location and height and including a skewness parameter produces deposition profiles more consistent with the measurement data, but still wider than equilibrium ray tracing in two different DIII-D discharges. Moreover, we show that the quality of the estimated deposition profile is key to the accuracy of diffusivity and convectivity estimates, but inversely, the estimated transport parameters hardly affect the quality of the power deposition estimate. Using this method, we show that the power deposition profile estimate is broadened with respect to ray-tracing by about 1.7–1.8 times in two DIII-D discharges.

Helicon current drive (CD), also called fast wave CD in the lower hybrid range of frequencies, has long been regarded as a promising CD tool for reactor grade plasmas.A newly installed MW-level system at DIII-D will be the first test of this technology in reactor-relevant plasmas, in the sense that full single-pass absorption is expected.A 30-module traveling wave antenna has been installed and optimized in-vessel in early 2020.The linear electromagnetic characteristics of the unloaded module array have been extensively tested both on the bench and in the vessel at instrumentation power levels.Excellent performance has been achieved, ∼2% reflected power and ∼1.5% dissipated power per module in air, in a 10 MHz band around 476 MHz.Stripline feeds on both ends of the antenna allow either co or counter CD.The installation of a 1.2 MW klystron and associated high-power electronics was completed in Fall 2020.Commissioning of the antenna is ongoing.An important goal of this experiment is to validate the helicon CD physics basis using an extensive set of new and upgraded diagnostics.

Microwave heat pulse propagation experiments have demonstrated a correlation between millimeter-scale turbulence and deposition profile broadening of electron cyclotron (EC) waves on the DIII-D tokamak. In a set of discharges in DIII-D, a variation in edge density fluctuations on the mm-scale is associated with 40%–150% broader deposition profiles, expressed in terms of normalized minor radius, as compared with equilibrium ray tracing. The 1D power profile is determined from transport analysis of the electron temperature response to EC power modulation using perturbative analysis with a square wave power modulation at 20–70 Hz, producing a series of Fourier harmonics that are fit collectively to resolve transport. Fitting an integrated heat flux expressed in the Fourier basis of the modulation to diffusive, convective, and coupled transport terms in a linear model can resolve the broadened EC deposition width from the power perturbation to resolve a broadening in each case. The best fit degree of beam broadening observed scales approximately linearly with the Doppler backscattering measured fluctuation level in the steep gradient region. Quantifying the effect of edge fluctuation broadening on EC current drive power needs of future devices will require 3D full-wave codes that can be validated on the current generation of machines. These DIII-D experiments provide a quantitative measure of fluctuation effects and a dataset to benchmark full-wave simulations that can model and eventually predict nonlinear effects neglected by 1D equilibrium beam and ray tracing.

Helicon waves have been recently proposed as an off-axis current drive actuator due to their expected high current drive efficiency in the mid-radius region in high beta tokamaks. This paper focuses on a numerical study to better understand effects of scrape-off-layer (SOL) turbulence on helicon wave propagation and absorption on the DIII-D tokamak using a recently developed helicon full-wave model with turbulent density inputs from synthetic single wavelength SOL turbulence and first-principles HERMES multi-wavelength turbulence models. With both input turbulence models, three key effects are observed: the helicon wave can scatter to undesirable locations in the plasma, large helicon wave electric fields can form in localized regions near the SOL turbulence, and the helicon wave can mode convert to slow waves in the SOL. This is shown to cause helicon wave refraction to undesirable locations and strong helicon wave absorption in the SOL resulting in significantly less helicon wave power in the core plasma. Using synthetic SOL turbulence, the simulations additionally show that high amplitudes and long wavelengths greater than a few cm on average have the largest effect on modifying the helicon wave propagation and absorption; the modeling predicts, for example, that approximately 60% of helicon power can be absorbed in the SOL for ñ/n ∼ 0.8 and lambda_perp ∼ 0.05 m. Several potential physical mechanisms that may explain the interaction of helicon waves with SOL turbulence in these simulations are discussed.

Experiments in DIII-D have measured the width of the power deposition and current drive profiles for electron cyclotron (EC) waves by modulating the gyrotrons and fitting the periodic response of the electron temperature profile or motional Stark effect signals to a linearized diffusion equation. The measured EC power deposition widths are scattered around the predicted values, including both narrower and wider cases. The average width for single gyrotron modulation is ≈88% of the theoretical value from ray tracing, while the average width for multiple gyrotron modulation is ≈127% of theory; aiming misalignment between the different EC launchers may explain why multi-gyrotron measurements are wider than for single gyrotrons. The measured EC current drive widths have nearly the same amount of profile broadening (≈131% of theory, on average) as does multi-gyrotron power deposition. Finally, while the measured locations of EC current drive are well centered on the expected values, the measured heating locations are skewed to larger radius. This difference may be due to rapid heat pulse propagation causing an inherent outward bias to the deduced power deposition profile, whereas magnetic flux diffusion has less effect on current drive measurements.