In preparation for high field side lower hybrid current drive (HFS LHCD) experiments in DIII-D, the HFS LHCD was to be commissioned and physics experiments commence once the system operated up to 300 kW for 0.5 s. The initial physics experiments sought to characterize coupling, wave propagation, and driven current measurements. In HFS LHCD first campaign, the maximum power was limited to <200 kW due to waveguide pressure leaks limiting the number of available modules and power per module. Here we summarize commissioning progress and initial physics observations. The HFS LHCD coupler was optimized for high qmin, DIII-D discharges where efficient off-axis current at r/a~0.6-0.8 is desired. The coupler n|| spectrum is peaked at 2.7 and is predicted to generate ~0.14 MA/MW coupled for 1.6 MW injected. In preparation, the HFS scrape-off layer density profile was characterized and found to have steeper profiles and lower fluctuation levels than the low field side. Furthermore, the HFS SOL density profile can be accurately predicted using global plasma quantities using machine learning. Thus far, one module has injected ~100 kW for 0.5 s with <5% reflected power. Nonthermal electrons have been observed on lower frequency channels of the electron cyclotron emission radiometer correlated with the LH power indicating core wave absorption. To avoid 30R neutral beam heat flux, a split launcher is proposed to avoid the high heat flux region while maintaining power spectrum and directivity.

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.

To address the needs for a fusion pilot plan design, DIII-D/EAST joint experiments on DIII-D have demonstrated high normalized beta β N ∼ 4.2, toroidal beta β T ∼ 3.3% with q min > 2, q 95 ⩽ 8 sustained for more than six energy confinement times in high poloidal beta regime. The excellent energy confinement quality ( H 98y2 ∼ 1.8) is achieved with an internal transport barrier at high line-averaged Greenwald density fraction f Gr > 0.9. The trapped gyro-Landau fluid (TGLF) modeling of the transport characteristics shows that the beam-driven rotation does not play an important role in the high confinement quality. The modeling also captures very well several transport features, giving us confidence in using integrated modeling to project these experimental results to future machines. The high-performance phase is terminated by fast-growing modes triggered near the n = 1 ideal-wall kink stability limit. New radio frequency (RF) capabilities for off-axis current drive could remove the residual ohmic current to achieve a fully non-inductive state, and improve the mode–wall coupling to increase the ideal-wall β N limit, enabling sustainment of the fully non-inductive high performance plasma in stationary conditions.

Helicon waves are thought to be promising in various tokamaks, such as DIII-D, because they can penetrate reactor-grade high-density cores and drive the off-axis current with higher efficiency. In the frequency regime similar to 476 MHz, both slow electrostatic and fast electromagnetic helicon waves can coexist in DIII-D. If the antenna parasitically excites the slow mode, these waves can propagate along the magnetic field line into the scrape-off layer (SOL). Although the importance of the misalignment of the Faraday screen and the electron density in the SOL on the excitation and propagation of slow modes is well known, the conditions for minimizing slow mode excitation have yet to be optimized. Using the Petra-M simulation code in the 2D domain, we analyze the effects of the misalignment of the antenna in the poloidal direction, the misalignment of the Faraday screen in the toroidal direction, and the density in front of the antenna on slow mode generation. Our results suggest that the misalignment of the Faraday screen is a critical factor in reducing the slow mode and that the misalignment angle should be below similar to 5 degrees to minimize the slow wave excitation. When the electron density is higher than 3.5x10(18 )m(-3) in the SOL, the generation of the slow mode from the antenna is minimized and unaffected by the misalignment of the Faraday screen.

More than 0.6 MW of rf power at 476 MHz has been coupled to DIII-D plasmas by launching helicon (whistler) waves with a traveling-wave antenna (comb-line) in the fast-wave polarization (Van Compernolle et al 2021 Nucl. Fusion 61 116034) which resulted in the observation of electron heating of the core plasma with single-pass absorption based on ray-tracing in L-mode discharges. The coupling performance of the 1.5 m wide 30-element comb-line traveling-wave antenna has been consistent with expectations based on the 2015-2016 experiments on DIII-D with a low-power 12-element prototype (Pinsker et al 2018 Nucl. Fusion 58 106007). The conditioning process that was necessary to carry out high-power experiments is discussed; rf-specific impurities have not been observed. Parametric decay instabilities have been observed and are being investigated as a potential edge absorption mechanism (Porkolab et al 2023 AIP Conf. Proc. 2984 070004).

A high-level design study for a new experimental tokamak shows that advances in fusion science and engineering can be leveraged to narrow the gaps in energy confinement and exhaust power handling that remain between present devices and a future fusion pilot plant (FPP). This potential new U.S. facility, an Exhaust and Confinement Integration Tokamak Experiment (EXCITE), will access an operational space close to the projected FPP performance regime via a compact, high-field, high-power-density approach that utilizes advanced tokamak scenarios and high-temperature superconductor magnets. Full-device optimization via system code calculations, physics-based core-edge modeling, plasma control simulations, and finite element structural and thermal analysis has converged on a BT=6 T, IP=5 MA, R0=1.5 m, A=3, D-D tokamak with strong plasma shaping, long-legged divertors, and 50 MW of auxiliary power. Such a device will match several absolute FPP parameters: plasma pressure, exhaust heat flux, and toroidal magnetic field. It will also narrow or close the gap in key dimensionless parameters: toroidal beta, bootstrap fraction, collisionality, and edge neutral opacity. Integrated neutron shielding preserves personnel access by limiting nuclear activation and maximizes experimental run time by reducing site radiation. In addition to design study results and optimization details, parameter sensitivities and uncertainties are also discussed.

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.

In this paper, we present an overview of the development of two high power traveling wave antennas (TWAs) termed the “comb-line” and the “finline” for launching fast and slow waves in plasmas. The comb-line antenna which has been deployed at the DIII-D National Fusion Facility tokamak can couple megawatt (MW) level RF power into the tokamak plasma with $\sim$ 2% reflected power and $\sim$ 2% dissipated power per array element at its design frequency, 476 MHz. A prototype of the finline antenna which has been designed to launch lower hybrid slow waves at 4.6 GHz, has reflection coefficient below 10% with 1.8 dB insertion loss.

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.

For the first time, experiments on the DIII-D tokamak have demonstrated electron cyclotron current drive with more than double the conventional efficiency by tailoring the wave–particle interactions in velocity space using a novel ‘top launch’ geometry. Steering the EC waves to propagate nearly parallel to the resonance drives current more efficiently by (1) selective damping on electrons with higher parallel velocity v ||, and (2) longer absorption path to compensate for inherently weaker absorption at higher v ||. Experiments using a fixed-injection top launch system find an optimal velocity space interaction for maximum current drive efficiency at ρ ∼ 0.5 where the ease of drawing out a high v || electron tail is balanced by sufficient absorption.

Novel internal measurements and analysis of ion cyclotron frequency range fast-ion driven modes in DIII-D are presented. Observations, including internal density fluctuation ( n~ ) measurements obtained via Doppler backscattering, are presented for modes at low harmonics of the ion cyclotron frequency localized in the edge. The measurements indicate that these waves, identified as coherent ion cyclotron emission (ICE), have high wave number, k ⊥ ρ fast ≳ 1, consistent with the cyclotron harmonic wave branch of the magnetoacoustic cyclotron instability, or electrostatic instability mechanisms. Measurements show extended spatial structure (at least ∼1/6 the minor radius). These edge ICE modes undergo amplitude modulation correlated with edge localized modes (ELM) that is qualitatively consistent with expectations for ELM-induced fast-ion transport.

We report mode structure measurements of coherent ion cyclotron emission (ICE) and sub-cyclotron modes on DIII-D. Through a dedicated experiment, we aimed to characterize a variety of modes via the upgraded ICE diagnostic in both L- and H-mode plasmas. In the L-mode plasmas, autopower spectrum peaks at harmonics of the ion cyclotron frequency f ci were observed and are localized in the core of the plasma. Sub-cyclotron modes ( f ∼ 0.5 f ci ) were also observed in L-mode plasmas when the toroidal magnetic field strength was lowered from B T = 2.17 T. In H-mode plasmas, many ICE harmonics localized to the plasma edge were observed, with some exceeding the nominal ICE diagnostic bandwidth of f ∈ [0, 100] MHz. Polarization estimates made using an orthogonal pair of vertical and horizontal magnetic pickup loops on the outer wall of the machine indicate that ICE and sub-cyclotron modes have compressional polarization at the plasma edge, the latter being consistent with simulation efforts for comparable DIII-D plasmas. For all modes, the same harmonics are observed on both the centerpost and outer wall loops, indicating that ICE is poloidally extended. Finally, toroidal mode numbers were calculated using three outer wall loops for both core ICE and sub-cyclotron modes in L-mode plasmas. The sub-cyclotron case served as a benchmark for our calculation method, with measured numbers of n ∈ [−24, −18] roughly agreeing with heuristic estimates of n ∈ [−20, −13]. Core ICE mode numbers were measured to be n ∈ [−10, 5], which is comparable to measurements made on other machines.

The ion cyclotron emission diagnostic on the DIII-D tokamak comprises seven single-turn loops that measure high-frequency (1-100 MHz) magnetic field fluctuations that are often excited by energetic particles in the plasma. The raw voltage signals induced in the loops in response to these fluctuations travel through a series of cables, isolation transformer DC blocks, low-pass filters, and finally a digitizer before being analyzed in frequency space. The diagnostic has been recently upgraded, most notably to include four additional graphite tile loops and a new eight-channel digitizer. The previous three loops are all on the low-field side of the tokamak. The measurement capabilities of the system have been expanded by the addition of a new horizontally oriented loop on the low-field side, an additional toroidal loop on the low-field side, and two toroidal loops on the high-field side. These loops will be used to provide approximate mode polarization, improved toroidal mode number calculations, and information on modes in inward-shifted plasmas, respectively.

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.

Fast-ion driven Alfvén waves with frequency close to the ion cyclotron frequency (f=0.58f_{ci}) excited by energetic ions from a neutral beam are stabilized via a controlled energetic ion density ramp for the first time in a fusion research plasma. The scaling of wave amplitude with injection rate is consistent with theory for single mode collisional saturation near marginal stability. The wave is identified as a shear-polarized global Alfvén eigenmode excited by Doppler-shifted cyclotron resonance with fast ions with sub-Alfvénic energetic ions, a first in fusion research plasmas.