Establishment of an efficient central solenoid (CS) free tokamak plasma start-up method may lead to an economical fusion reactor. CS-free start-up using lower hybrid (LH) waves has been studied on the TST-2 spherical tokamak. Plasma current of about a quarter of CS-driven discharges has been obtained fully non-inductively using the outer-midplane and top LH launchers. Recently, an outer-off-midplane LH launcher was developed to achieve higher plasma current by optimizing for core absorption and minimal fast electron losses. Using the (outer-)off-midplane launcher, fully non-inductive plasma current start-up up to about 8 kA was achieved. Coupled ray-tracing and Fokker–Planck simulation was performed on equilibria reconstructed with an extended MHD model. It was found that the experimentally observed plasma current was in reasonable agreement with the numerical simulation. The simulation predicted appreciable orbit losses for the off-midplane launcher driven discharge at the present parameters, which was consistent with the experimentally observed x-ray radiation characteristics. The simulation showed that the current density was saturated for the present off-midplane launcher discharges and higher density and higher LH power was necessary to achieve higher plasma current.

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.

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).

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.

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.

This work describes a novel optical heterodyne detection scheme that significantly extends the frequency response of the phase contrast imaging method to detect electron density fluctuations in tens of megahertz frequency range. The system employs a variable frequency electro-optic modulator to allow operation at any frequency in the range 10-40 MHz without the need to realign the system. The frequency coverage of the system makes it suitable to measure the radial structure of the electron density component of ion cyclotron emission on devices having confining magnetic field of a few tesla, thus extending the purely temporal measurements provided so far by magnetic probes.

A novel “finline” antenna was designed to excite lower-hybrid waves (LHWs) for plasma current drive at 2.45 GHz in the TST-2 spherical tokamak. A periodic fin array acts as a transmission line for electromagnetic surface waves. Cold tests confirmed propagation of the electromagnetic surface wave on the fin array which was dominantly Transverse Magnetic (TM). Unlike the conventional waveguide array “grill” antenna, the finline antenna requires only two ports to feed all the elements, and calculations show that there is little reflection regardless of the plasma conditions in front of the antenna. Characteristics of the microwave transmission on the fin array were investigated using a three-dimensional full-wave simulation and the fin geometry was tuned to achieve the desired refractive index. Coupling to LHWs under typical TST-2 scrape-off layer conditions was confirmed by the simulation. To demonstrate LHW excitation, the finline antenna was energized in the presence of a plasma generated by another 200 MHz LH antenna. The x-ray radiation showed strong enhancement during the finline antenna pulse which indicated successful generation of fast electrons by LHWs.

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.

Presence of a large population of fast particles may qualitatively modify the tokamak equilibrium from the one that can be described by standard magneto-hydrodynamics (MHD). The kinetic modification of the MHD equilibrium was studied for a lower-hybrid (LH) wave driven plasma on the TST-2 spherical tokamak. The analysis was performed using an equilibrium reconstruction method based on an extended MHD model that considers a two-component plasma of bulk MHD fluid and kinetic fast electrons. Scrape-off-layer current appeared naturally in the extended MHD analysis because of the finite electron orbit excursion from the flux surfaces. This reduced the bulk pressure contribution to the toroidal current substantially. The resulting poloidal flux profile was more consistent with that inferred from the Thomson scattering measurement.

Plasma start-up experiments using the lower hybrid wave (LHW) were performed on the TST-2 spherical tokamak. Two capacitively-coupled combline antennas (outboard-launch and top-launch) are presently being used. Bottom-launch can be simulated using the top-launch antenna with reversed toroidal field. The unexpectedly good result of the simulated bottom-launch case was investigated by a combined ray-tracing and Fokker-Planck analysis. In the case of simulated bottom-launch the parallel index of refraction (n||) downshifts initially, but because of the plateau formed by the upshifted n|| LHW after reflecting from the cut-off layer at the bottom of the plasma, strong absorption of the low n|| LHW occurs during the first pass through the plasma, generating a distribution function which extends to very high energies.

The establishment of reactor-relevant radiofrequency heating and current drive techniques is a focus of work on DIII-D in the next five-year period. This paper gives an overview of the planned experimental work in the areas of (1) nearly vertically launched ECCD, (2) ‘helicon’ (whistlers or fast waves in the lower hybrid range of frequencies) current drive, and (3) high-field-side-launch (HFS) lower hybrid (slow wave) current drive. Each of these techniques addresses the need for efficient off-axis current drive for a steady-state tokamak reactor to supplement the bootstrap current and to provide current profile control, and each will be experimentally assessed at a coupled power level of ~1 MW on DIII-D in the next few years.

The capacitively-coupled combline (CCC) antenna has been developed for current drive by the lower hybrid wave (LHW) on the TST-2 spherical tokamak. The combline antenna was developed to satisfy the requirements of high directionality, low reflectivity, and simple feeding. Since the combline antenna makes use of mutual coupling between neighboring elements, only the first and the last elements are connected to external feedlines. RF powers and power densities of the order of 100 kW and 1 MW m(-2) can be achieved easily in small antennas of the order of 0.1 m(2). The two CCC antennas installed in TST-2 (outboard-launch and top-launch) excite toroidal refractive index (n(phi)) spectra peaked around five with the frequency of 200 MHz. Wave excitation calculation using a finite element code shows that the excited power of the n(phi) = 5 LHW component increases rapidly when the plasma cutoff density layer (where n(e) = 5 x 10(14) m(-3)) becomes closer than 27 mm from the antenna surface, in agreement with experiment. Experimentally, the density profile in front of the antenna can be controlled by adjusting the side limiter location or antenna-plasma distance, and should be optimized for antenna-plasma coupling, since too high coupling results in a broadened and less directional n(phi) spectrum, and too low coupling results in a less efficient power coupling. Using these antennas, successful ST plasma start-up and I-p ramp-up to over 25 kA (about 1/4 of the nominal I-p for OH operation) have been achieved with RF power of less than 100 kW in about 40 ms.

DIII-D research is addressing critical challenges in preparation for ITER and the next generation of fusion devices through focusing on plasma physics fundamentals that underpin key fusion goals, understanding the interaction of disparate core and boundary plasma physics, and developing integrated scenarios for achieving high performance fusion regimes. Fundamental investigations into fusion energy science find that anomalous dissipation of runaway electrons (RE) that arise following a disruption is likely due to interactions with RE-driven kinetic instabilities, some of which have been directly observed, opening a new avenue for RE energy dissipation using naturally excited waves. Dimensionless parameter scaling of intrinsic rotation and gyrokinetic simulations give a predicted ITER rotation profile with significant turbulence stabilization. Coherence imaging spectroscopy confirms near sonic flow throughout the divertor towards the target, which may account for the convection-dominated parallel heat flux. Core-boundary integration studies show that the small angle slot divertor achieves detachment at lower density and extends plasma cooling across the divertor target plate, which is essential for controlling heat flux and erosion. The Super H-mode regime has been extended to high plasma current (2.0 MA) and density to achieve very high pedestal pressures (similar to 30 kPa) and stored energy (3.2 MJ) with H-98y2 approximate to 1.6-2.4. In scenario work, the ITER baseline Q = 10 scenario with zero injected torque is found to have a fusion gain metric beta(TE) independent of current between q(95) = 2.8-3.7, and a lower limit of pedestal rotation for RMP ELM suppression has been found. In the wide pedestal QH-mode regime that exhibits improved performance and no ELMs, the start-up counter torque has been eliminated so that the entire discharge uses approximate to 0 injected torque and the operating space is more ITER-relevant. Finally, the high-beta(N) (<= 3.8) hybrid scenario has been extended to the high-density levels necessary for radiating divertor operation, achieving similar to 40% divertor heat flux reduction using either argon or neon with P-tot up to 15 MW.

A comb-line antenna to demonstrate efficient off-axis non-inductive current drive from the absorption of toroidally directed very high harmonic fast waves is being designed and built for DIII-D. The antenna consists of a toroidal array of 30 modules, spanning 1.7 m on the outer vessel wall just above the tokamak midplane. This antenna will be fed with 1 MW of RF power at 476 MHz through a stripline (SL) feed on both sides inside the vacuum vessel. COMSOL Multiphysics, a commercial finite element analysis software, was used to perform the RF analysis and the induced force analysis due to plasma disruption events on the whole antenna system. Results on the RF performance of the SL and the RF coupling into and between the modules, and the SL and modules' RF losses will be presented. In addition, the disruption induced current and forces are shown, and a method to mitigate them is presented.

The DIII-D National Fusion Facility is advancing the science and technology of steady-state fusion plasma sustainment through the implementation of two first-of-a-kind radio frequency current drive systems: the “helicon” or fast wave in the lower hybrid range of frequencies (LHRF), and high field side (HFS) launch of the LHRF slow wave. This paper focuses on the HFS LHRF system, with a companion paper on the helicon system [O’Neill FEC 2018]. Using existing DIII-D discharges, we have identified high performance scenarios that are predicted to have excellent wave penetration, strong single pass absorption and high current drive efficiency. Simulations predict this will raise ideal βN limits in DIII-D and permit access to higher density advanced tokamak regimes. The higher B-field on the HFS improves wave accessibility and allows for use of lower n||, resulting in higher current drive efficiency for LHRF slow waves and damping at r/a ∼ 0.6−0.8 on the first pass. Calculations show that HFS launch of slow waves in the LHRF can lead to a physics current drive efficiency of 0.17×1020 A·W−1m−2 at r/a ∼ 0.6−0.8 in DIII-D and 0.4 × 1020 A·W−1m−2 in a high B-field reactor. HFS LHRF represents an integrated solution that both improves core wave physics and mitigates PMI/coupling issues. Experimental results from a mockup antenna structure located on the HFS of DIII-D support theoretical predictions that HFS antennas will be able to survive long-term in a tokamak environment. An innovative, compact HFS LHRF antenna design has been developed combining an H-plane T-junction poloidal splitter (used on C-Mod and tested recently on COMPASS) and multi-junction toroidal splitter (used on Tore Supra, EAST). Models show good coupling properties for predicted edge density profiles.