Bart Van Compernolle

Helicon current drive has been observed experimentally for the first time in dedicated DIII-D tokamak experiments, providing a proof-of-principle and an attractive source of efficient current drive in reactor-grade tokamak plasmas.

Resonant wave–particle interactions between externally launched helicon waves (also known as whistler waves) and runaway electrons (REs) have been demonstrated on the DIII-D tokamak. In this work we extend the initial results reported in (Choudhury H. et al 2026 Phys. Rev. Lett. 136 025101) by exploring the effects of antenna alignment with the edge magnetic field, toroidal wave propagation direction, and coupled power on RE scattering in the quiescent RE experimental scenario. Two distinct experimental configurations have been investigated: one in which the antenna’s Faraday cage aligns well with the edge background magnetic field, known as the ideal antenna configuration, and one with misalignment, known as the non-ideal case. Previously, it had been found that helicon power in the ideal antenna configuration prevented RE growth despite the normalized toroidal electric field remaining high enough to drive exponential RE growth in the absence of helicon power. In this paper, we show that scattering via the normal Doppler resonance ( n = 1 ) effectively limits the growth of the RE population in both the ideal and non-ideal antenna configurations, with the effect of average power and energy deposited investigated. In contrast, launching waves that favor the anomalous Doppler resonance ( n = − 1 ) is observed to enhance rather than reduce the RE population. In addition, fast magnetic measurements reveal rising-tones in the 30–60 MHz range during helicon-off periods, which are not observed prior to helicon power. Finally, the challenges of using launched helicon waves to scatter post-disruption RE beams are discussed. Collisional damping and a large vacuum gap between the plasma and antenna on the outboard side present significant obstacles to helicon waves propagating into the plasma core.

Experiments on DIII-D have demonstrated a density-confinement synergy that enables sustainment of high performance in a previously unattained parameter regime of simultaneous very high energy confinement quality (H _98y2 ⩾ 1.5), very high line-average density Greenwald fraction ( ƒ _Gr = πa ² < n >/ I _P ⩾ 1.4), and high toroidal beta ( β T ⩾ 3%). Tokamak operation in this regime is essential for a compact steady-state FPP, as well as for Q = 10 with 500 MW of fusion power in ITER at I _P ≪ 15 MA. These experiments leveraged the knowledge that, in the high-poloidal-beta ( β P ) regime, density gradients can enhance turbulence stabilization caused by high α MHD ( α MHD ∼ d β P / d r ). This was described by theoretical predictions and gyrokinetic transport simulations (Kotschenreuther et al 2024 Nucl. Fusion 64 076033), and later confirmed by experiments on DIII-D (Ding et al 2024 Nature 629 555). To increase both β P and β T , the new experiments increased the ideal-wall stability β N -limit by using a smaller plasma-outer wall distance and higher triangularity in the plasma cross section (top/bottom average δ ∼ 0.9), enabled by the recent ‘shape & volume rise’ (SVR) modification to the DIII-D divertor. The higher triangularity also contributed to achieving higher ƒ _Gr by enabling higher pedestal density. At high density, the pedestal is ballooning limited and exhibits small and frequent ELMs, while the divertor is near detachment even without any impurity seeding. Highplasma performance was attained and sustained reproducibly, with the eventual terminations brought about by an MHD mode destabilized as the current profile slowly continued to evolve. A path to stationary fully noninductive operation might include ECH injection to reduce both core impurity accumulation and the electron collisionality, thus increasing the bootstrap current. These experiments provide the first experimental demonstration of the ƒ _Gr , H _{98 y 2} , and β T values required simultaneously for ITER Q = 10 at I _P < 10 MA, pointing to practical ways to improve the energy confinement in a fusion reactor.

The high-power helicon wave system in the DIII-D tokamak could potentially introduce new plasma-material interaction (PMI) challenges owing to rectified radio-frequency (RF) sheath potentials that develop near the antenna and surrounding plasma-facing components. We present the first application of the STRIPE (Simulated Transport of RF Impurity Production and Emission) framework to helicon-induced PMIs, extending previous STRIPE studies of ICRH antennas by incorporating net erosion, local re-deposition, and three-dimensional (3D) global impurity transport. The integrated workflow couples SOLPS, COMSOL, RustBCA, GITR, and GITRm to simulate carbon erosion, re-deposition, and impurity transport for two experimentally constrained DIII-D H-mode helicon operating scenarios with different antenna–plasma gaps, coupled RF powers, and edge plasma conditions. COMSOL predicts rectified RF sheath potentials of 1–5 kV localized near the lower portion of the antenna, where the magnetic field intersects the surface at grazing incidence. Carbon self-sputtering dominates the erosion source, whereas RF-accelerated D + ions contribute approximately 1% of the total gross erosion. The smaller-gap operating scenario exhibits substantially stronger gross erosion, enhanced local re-deposition ( ∼ 12%), and a larger confined carbon inventory owing to increased plasma accessibility and broader RF sheath coverage. Comparison with available DIII-D measurements shows no distinct change in the global carbon signal that correlates with the helicon RF pulse, consistent with the simulations indicating that the helicon-generated carbon source remains small compared with the existing background carbon inventory under the present graphite-wall operating conditions. These results demonstrate the capability of STRIPE to integrate RF sheath modeling, plasma transport, surface interaction physics, and 3D impurity transport for the interpretation of helicon-induced PMIs. The study further identifies the principal sources of modeling uncertainty, including grazing-angle RF sheath physics, slow-wave resolution, plasma-background extrapolation, and trace-impurity transport assumptions, providing a framework for future validation and model development.

The National Spherical Torus Experiment Upgrade (NSTX-U) is preparing to resume operation, representing a crucial step toward realizing compact, cost-effective fusion pilot plants (FPPs). In advance of this, extensive modeling and data analysis have been conducted to advance the physics basis for low-aspect-ratio, high-performance plasma regimes, focusing on three core objectives: confinement and stability, power and particle handling, and steady-state operation. Significant progress has been made in understanding the electron temperature flattening in high-β plasmas, which is shown to be driven by a complex interplay of MHD instabilities (e.g., non-resonant infernal modes), fast-ion-driven Alfvén eigenmodes, and electron and ion-scale micro-instabilities, particularly Kinetic Ballooning Modes (KBMs), whose destabilization is strongly dependent on parallel magnetic field fluctuations (δB _‖ ). Furthermore, a new Gyrokinetic Critical Pedestal model was developed, accurately predicting pedestal structure by identifying KBMs as the primary stability limit, offering a critical constraint for future high-confinement scenarios. To address the challenge of high heat flux, novel liquid lithium plasma-facing components were modeled. The analysis confirmed that lithium vapor shielding is a self-regulating mechanism for heat mitigation, while also emphasizing that strong main ion parallel flow is essential to minimize core lithium contamination. Finally, progress toward steady-state operation was anchored by developing the required physics basis and control tools. This includes predictive modeling for Reversed Magnetic Shear sustainment, demonstrating that magnetic island-induced bootstrap current reduction is negligible in STs, and advancing real-time control and disruption avoidance capabilities. The development of high-speed surrogate models (e.g., MMMNet) provides computationally efficient tools vital for non-inductive scenario optimization and integrated, low-disruptivity operations planned for NSTX-U.

Helicon waves (also known as whistler waves) satisfying the normal wave-particle cyclotron resonance are observed to limit the growth and maximum energy of relativistic electrons (REs) in low-density Ohmic DIII-D tokamak plasmas. Following the application of helicon waves, pitch-angle scattering of high-energy REs causes an increase in both synchrotron and electron-cyclotron emissions. The hard x-ray emission, a proxy for the RE population, ceases to grow. Energy-resolved hard x-ray measurements also show a striking decrease in the number of high-energy REs (above the resonance at approximately 8 MeV) to below the noise floor and an increase in low-energy (∼4 MeV) REs. This occurs despite the toroidal electric field remaining high enough to drive exponential RE growth in the absence of helicon waves. These results open new directions for limiting the maximum energy of RE populations in laboratory and fusion plasmas.

A set of experiments were conducted on the LArge Plasma Device (LAPD) at UCLA to test the operational principles of a traveling wave antenna of the comb-line type. This antenna was designed to launch helicon waves (fast waves in the lower hybrid range of frequencies) on DIII-D. With the order-of-magnitude lower static magnetic field on LAPD, the antenna excites waves in a different regime. Whenever fast waves can propagate in LAPD, slow waves are also supported by the plasma so it is necessary to distinguish between the two cold-plasma branches in evaluating the effectiveness of the launcher. The results show that the launcher couples well to fast waves when the plasma supports fast-wave propagation; control of the principal imposed parallel wavenumber can be achieved through varying the launch frequency on the antenna within its bandwidth of operation; and that the launched waves exhibit strong directionality. We also investigate the role of the plasma profile and wave mode on the loading characteristics. Additionally, a comparison with full-wave modeling of the propagating waves is shown using both a cold-plasma model in COMSOL and a hot-plasma model in RFPisa, which obtain similar results in the present regime.

High harmonic fast waves (HHFW) are a versatile heating and current drive tool for scenario development. Extensive modeling scans were performed to find optimal parameters for different uses of HHFW in National Spherical Tokamak Experiment (NSTX-U). Scans of plasma density, temperature, magnetic field, and antenna phasing were performed both with and without neutral beam injection. For speed of calculation, the ray-tracing code GENRAY coupled to the quasilinear Fokker–Planck code CQL3D was used. CQL3D allows for a more accurate description of the fast ion population, as well as for quasilinear effects such as HHFW-induced modifications of the distribution function. Best current drive results are obtained at elevated electron temperatures and with the lowest k _φ phasing. Adding neutral beams however typically strongly reduces the HHFW current drive efficiency at the low density cases due to HHFW absorption on beam ions. Results of this parametric study will feed into scenario development and predict-first whole-shot modeling of NSTX-U discharges.

Ion cyclotron emission (ICE) is widely observed from spatially localised minority energetic ion populations in toroidal magnetically confined fusion (MCF) plasmas, both tokamaks and stellarators. Its spectral structure is typically regular with narrow suprathermal peaks, whose frequency separation matches a local energetic ion cyclotron frequency. Here we report the first computational (fully nonlinear self-consistent kinetic particle-in-cell code) and analytical (linear magnetoacoustic cyclotron instability (MCI)) studies of ICE observations from cylindrical plasmas contained in the Large Plasma Device (LAPD). Because LAPD is cylindrical, the plasma physics giving rise to the observed ICE spectrum necessarily excludes toroidal effects. Our approach, previously successful for toroidal plasma ICE, assumes slab geometry and hence is well adapted to LAPD. ICE from LAPD is strongly electrostatic, as distinct from electromagnetic, and is driven by 15 keV beam ions for which the ratio of perpendicular speed to the local Alfven speed, v ⊥ / v A , is 0.15, lower than in MCF plasmas from which beam-driven ICE has previously been observed. Our results are in good agreement with these observations. There is congruence between simulated ICE spectra, obtained in the saturated nonlinear regime of our computations, and observed ICE spectra; the underlying physics is essentially the same as in toroidal plasmas; and there is alignment with linear analytical theory where appropriate. The present work establishes a mapping from the cylindrical LAPD ICE observations to toroidal MCF ICE observations. The LAPD spectra are instances of beam-driven MCI-type ICE in its sub-Alfvenic, predominantly electrostatic manifestation, which has precedents in MCF stretching back to the 1990s. An interesting corollary is that, for many purposes, ICE in toroidal MCF plasmas ‘might as well’ be occurring in a cylinder.

A new frequency-stepped Doppler backscattering (DBS) system has been integrated into a real-time steerable electron cyclotron heating launcher system to simultaneously probe local background turbulence (f < 10 MHz) and high-frequency (20–550 MHz) density fluctuations in the DIII-D tokamak. The launcher allows for 2D steering (horizontally and vertically) over wide angular ranges to optimize probe location and wavenumber response. The vertical steering can be optimized during a discharge in real time. The new DBS system employs a programmable frequency synthesizer with adjustable dwell time as a source to launch either O or X-mode polarized millimeter waves. This system can step in real-time over the entire E-band frequency range (60–90 GHz). This combination of capabilities allows for the diagnosis of the complex internal spatial structure of high power (>200 kW) helicon waves (476 MHz) injected from an external antenna during helicon current drive experiments in DIII-D. Broadband density fluctuations around the helicon frequency are observed during real-time scans of measurement location and wavenumber during these experiments. Analysis indicates that these broadband high-frequency fluctuations are a result of backscattering of the DBS millimeter-wave probe beam from plasma turbulence modulated by the helicon wave. It is observed that background turbulence is effectively locally “tagged” with the helicon wave electric field, forming images of the turbulent spectrum in the overall density fluctuation spectrum that appear as high-frequency sidebands of the turbulence. These observations of background turbulence and high-frequency fluctuations open up the possibility of monitoring local helicon wave amplitude by comparing the high-frequency signal amplitude to the simultaneously measured background turbulence. In combination with the real-time measurement location and wavenumber scanning capabilities (offered by real-time frequency-stepping and steering), this allows rapid determination of the spatial distribution of the helicon wave power during steady-state plasma operation. In the long term, such measurements may be used to validate predictive modeling (GENRAY [Smirnov and Harvey, Bull. Am. Phys. Soc. 40, 1837 (1995)] or AORSA [Lau et al., Nucl. Fusion 58 066004 (2018)]) of helicon current drive in DIII-D plasmas.