Charles Lasnier

Dispersion strengthened tungsten (DSW) alloys are promising materials for plasma facing components, displaying improved mechanical properties and microstructure stability compared to pure tungsten. In this work, the performance of DSW reinforced with 1 wt% transition-metal carbide (TaC, TiC, or ZrC) has been examined under H-mode plasma in the DIII-D tokamak. DSW samples were mounted both flush with the divertor floor and at a 10° angle towards the incoming plasma fluxes to further increase the heat and particle loading. To simulate the impact of prolonged tokamak operation, similar plasma exposures were repeated on two additional sets of DSW samples: one set pre-implanted with 10 ²² m ^-2 helium, and another set heated in-situ to an additional 500°C. Flush mounted samples exhibited good surface stability, with no evidence of melting or cracking regardless of composition or treatment. Carbon redeposition was observed on most samples, likely due to the graphite environment of DIII-D. Under coupled thermal loading and helium implantation, surface nanostructuring developed in some DSW samples; however, DSW containing tantalum dispersion phases showed strong resistance to such surface modification. In contrast, obvious surface melting was observed on 10° angled samples containing TaC and ZrC dispersoids, likely associated with dispersoid oxidation during fabrication, which reduced thermal conductivity and promoted elevated surface temperatures during plasma exposure. The grain morphology of DSW containing TiC remained largely unchanged, although near-surface dispersoids on the angled TiC sample were missing after plasma exposure. The loss is associated with oxidation of the dispersoids during fabrication, followed by sublimation during plasma exposure. Through systematic comparison of DSW performance under H-mode plasma conditions, this study identifies key degradation mechanisms and highlights strategies to improve the performance of DSW as plasma facing materials.

A first-of-its-kind experiment was conducted in the DIII-D tokamak under WEST similarity constraints on plasma shape and core parameters. This work presents a detailed transport study comparing a reference regime dominated by intrinsic carbon radiation and a high-radiation regime resulting from controlled tungsten (W) injection using the laser blow-off system, with a core tungsten concentration n W / n e ∼ 3 × 10 − 4 and a radiated-power fraction f rad > 0.5 . The W-induced radiative cooling lowered the electron temperature, thereby decreasing T e / T i and stabilizing trapped-electron-mode (TEM) turbulence. This transition in turbulence regime reduced momentum and ion thermal diffusivities, yielding ion temperature peaking and a factor-of-two increase in toroidal rotation. At the outer plasma region, enhanced E × B shear and increased collisionality further suppressed ion-scale turbulence, causing a sharp drop in ion heat flux. Consequently, impurity transport, predominantly turbulent in the low-radiation regime, acquired a strong neoclassical inward W convection during radiative cooling, bootstrapping the cooling cycle. Despite f rad > 0.5 , radiative collapse was not observed, likely owing to collisional ion-to-electron energy exchange acting as an electron-energy reservoir, together with 1 / 1 MHD activity modulating the radiated power through core impurity neoclassical T i -screening. These results support preparation for a tungsten wall change in DIII-D by elucidating tungsten-induced turbulence stabilization. They also provide key insights for interpreting plasma performance in WEST and are relevant to future reactors expected to operate with radiating tungsten-walled plasmas.

Runaway electrons (REs) generated during tokamak disruptions are a major concern for the safe operation of future fusion devices. The interaction of REs with waves has been proposed as a potential mechanism for their mitigation. This study investigates the effect of electron–cyclotron heating (ECH) on post-disruption plasmas containing REs. O- and X-mode EC waves are routinely used for plasma heating and current drive. However, these modes do not interact directly with relativistic electrons and cannot be injected into plasmas with densities exceeding their respective cutoff densities. In contrast, the internal slow X-mode (sX) can resonate with relativistic electrons and may therefore provide a pathway for their mitigation. We report DIII-D experiments designed to access the internal slow X-mode through O-sX conversion during post-disruption RE plateaus. ECH increased the background-plasma density, doubled the loop voltage, and strongly enhanced the RE synchrotron signal, demonstrating substantial ECH-plasma coupling and increased effective dissipation of the RE channel. Although ECH produced a pronounced response consistent with enhanced RE dissipation, we could not isolate conclusive evidence of resonant sX-RE interaction. Strong heating, ionization, and impurity-redistribution effects appear to dominate the response and may mask a direct wave-particle signature. Nevertheless, the results demonstrate the potential of ECH for RE control through background-plasma modification, with O-sX access offering an additional possibility of direct resonant interaction.

Instrumented sacrificial limiter heads, both domed (proud) and flat (flush) are used in DIII-D runaway electron (RE) wall strikes to study the wall impact dynamics with high spatial and time resolution. The approximate structure of the RE wetted area and heating depth on the domed limiter heads were predicted qualitatively using orbit-tracking simulations, although a strong left-right asymmetry (about the magnetic field direction) was not captured well by the simulations. It is hypothesized that this difference is perhaps due to the local 3D magnetic field perturbation of the dome limiter head. The average kinetic energy K and pitch angle  of REs striking the limiter head were estimated from the spatial distribution of local HXR emission and were estimated to be roughly K ≈ 4 MeV and  ≈ 0.2 rad. These values are roughly consistent with in-plasma values estimated before the loss event, indicating that RE kinetic energy and pitch angle are not drastically altered when transporting to the wall. Large shot-shot variations (1 - 10 kJ) in energy deposition into the limiter head were observed and were explained by shot-shot variations in locked MHD mode toroidal phase. For the largest deposited energies (10 kJ), graphite material failure and explosive dust release was observed, and the depth of material failure at higher energy deposition was successfully reproduced using modelling of volumetric energy deposition and brittle failure. The presence of energetic (keV level) ion impact during the RE wall strike was confirmed by three different surface analysis techniques. The ratio of energetic ion to RE flux appears to be larger on flat surfaces within the RE wetted area, although the total energy flux due to energetic ions has yet not been quantified.

It is demonstrated that the peak toroidal impact phase of the post-disruption runaway electrons (REs) can be varied shot to shot by means of applied static nonaxisymmetric (3D) magnetic fields, or resonant magnetic perturbations (RMPs). In the experiments, high-current (500 kA), purged RE plateaus (i.e., low impurity postdisruption plasmas with current carried completely by REs) were terminated on the DIII-D tokamak center post. The global toroidal impact peak of the REs appears to be dominantly set by the toroidal phase of a wall-locked tearing mode, typically (m/n)=(2/1), but sometimes (m/n)=(3/2). In the absence of external RMP, the locked mode reliably locks close to a preferred toroidal phase ϕ≃75° probably set by intrinsic error fields. With applied RMP, the locked mode and RE impact phase is pushed to a new location. When the tearing mode is (m/n)=(2/1), the phase is well controlled and tracks the applied RMP phase, but in the case of an (m/n)=(3/2) tearing mode, the new phase location is not well controlled. These results could have an application in international thermonuclear experimental reactor and other fusion power plant-sized tokamaks as a method for spreading disruption first wall damage from REs.

A major challenge in tokamak fusion research is first-wall erosion caused by steady heat loads and sudden energy bursts known as edge-localized modes. Divertor detachment reduces steady-state heat flux, while resonant magnetic perturbations can suppress these instabilities. However, integrating the two has been difficult because they require conflicting operating conditions. Here we demonstrate simultaneous achievement of resonant magnetic perturbations mitigated small edge-localized modes and impurity seeded partial divertor detachment in plasmas with an ITER-similar shape on the DIII-D tokamak. Experiments and simulations show that resonant magnetic perturbations facilitate detachment by redistributing particles, lowering the core density and increasing the scrape-off layer density, thereby reducing the amount of injected gas required. Cooling-gas injection eliminates the secondary heat-flux peak created by three-dimensional magnetic lobes, while edge cooling weakens the plasma response to the applied magnetic fields. These advances illustrate a viable pathway for integrating edge stability control with power exhaust in future fusion reactors.

An experimental discharge from the DIII-D tokamak is modeled using the SOLPS-ITER code suite and compared against measurements in the pumped and relatively closed upper divertor. Uncertainties of boundary plasma simulations are identified by attempting to match code inputs to experimental conditions, including iteratively solving transport coefficients to match upstream experimental profiles using varying quantities of core particle flux, different pumping models, and various assumptions of ion thermal transport. Simulated boundary conditions for particle injection at the core interface are shown to be relevant to the plasma solution at the divertor targets, even if upstream transport is modified so that plasma profiles are comparatively similar, although seperatrix density is not held constant. When upstream plasma profiles are matched to experimental measurements by varying diffusive transport coefficients, using either poloidally symmetric or ballooning structure, the model finds a majority of injected energy being transported radially off the computational domain, in conflict with experimental radiated power measurements and heat flux measurements at the divertor target. Imposing a maximum thermal diffusivity or radially shifting the experimental separatrix location of the fitted profiles to increase power conducted to the targets by increasing the upstream electron temperature does not significantly modify this result. Including a thermalizing plenum volume in the simulation domain is shown to maintain the experimental volumetric pumping rate without knowing the neutral energy distribution incident on the pump duct a priori . By modifying transport parameters to match different assumptions for ion temperature, downstream neutral pressure changes by more than a factor of two, suggesting that attention to ion thermal transport may be a critical parameter for simulations to accurately resolve recycling and neutral transport, particularly in a closed divertor geometry. In addition to quantifying various modeling uncertainties, this work motivates both further experimental study and modeling improvements to improve predictive capabilities.

Dedicated H-mode hydrogen experiments has been conducted at DIII-D and are compared to a similar deuterium dataset under a range of divertor conditions, revealing significant effects of the isotope mass on divertor conditions and pedestal behavior. H-mode confinement is found to be systematically lower in hydrogen and experiences a more pronounced degradation with increased fueling. Hydrogen is found to have a higher n _e,sep at similar fueling rates. When comparing hydrogen and deuterium discharges under similar divertor regimes, the hydrogen cases are shown to have a higher electron density and lower electron temperature in the outer divertor. Hydrogen and deuterium density ramps demonstrate a 20% higher ion current and of a 17% higher upstream density at detachment onset in hydrogen. The trends of divertor conditions with isotope mass through detachment are studied through a 2 point model analysis and detailed 2D interpretive SOLPS-ITER modeling. Both approaches are able to reproduce key experimental trends. Highlighted in the modeling are a significant role of the ion sound speed, the momentum balance, and carbon sputtering and radiation. An analysis of the particle source in the 2D modeling confirms a higher core particle source for hydrogen, consistent with the experimental hydrogen pedestal structure. However, deuterium is shown to have a lower ionization mean free path at the target than in hydrogen at the same upstream density, contrary to expected scaling with isotope mass. An experimental analysis of the pedestal stability and neutral fueling demonstrates that neither of these effects are sufficient to explain the pedestal differences between isotopes nor the hydrogen pedestal structure respectively. These results indicate that increasing ion mass is likely beneficial for divertor performance in future devices, reducing heat and particle fluxes as well as the n _e,sep required for detachment.

Reduced radial transport, short midplane-to-target parallel connection lengths, and a strong effect of cross-field drifts were responsible for the high densities required for detachment in strong negative triangularity (NT) configurations in DIII-D. Dissipative divertor conditions were achieved in NT discharges at different triangularity, injected power, plasma current, and toroidal field direction. Differences between negative and positive triangularity (PT) discharges are analyzed in this paper to understand the requirements for access to detached divertor conditions: power balance, geometry, radial transport and effect of cross-field particle drifts. Parametric dependencies of access to detachment on plasma current and power flowing into the scrape-off layer remained similar in negative and PT and impurity seeding was observed to reduce the density needed to detach by up to 30% at the expense of core impurity dilution. The impact of triangularity on core-edge integration was tested varying bottom triangularity at fixed top triangularity. The high density needed to detach was not intrinsic to the NT edge as shapes with positive lower triangularity and negative upper triangularity were able to detach at lower upstream densities while maintaining an ELM-free NT edge. Confinement degradation at deeper detachment levels was however observed in all NT shapes, often associated with radiation instabilities.