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Plasma-facing materials in future large-scale fusion reactors must be designed to withstand high heat fluxes from extreme off-normal events such as edge localized modes and unmitigated plasma disruptions. The erosion rates of possible tungsten-alternative materials are tested under high heat flux conditions at the DIII-D National Fusion Facility. High-purity β-3C CVD silicon carbide was exposed alongside MAX phase ceramic Ti3SiC2 to both L- and H-mode plasma discharges in the DIII-D divertor. Samples survived average heat fluxes ranging from 2–10 MW/m2 over 16 s. A new micro-trench erosion measurement technique was successfully implemented and measured Ti3SiC2 and SiC erosion rates of 0–9 nm/s and 27–73 nm/s, respectively. Additionally, average ion impact angle estimates for an incident B-field angle of ∼1.5° from surface parallel were made using micro-trench impact patterns. Measurements ranged from θ = 24o–34o with respect to Bt and ϕ = 51.5o–55o below the surface normal.

The plasma-surface interactions from samples of high-conductivity graphitic foam biased to 120 V and placed in 6–8 eV deuterium plasmas with densities as high as 1019 m−3 were investigated at the PSI-2 linear plasma device in Jülich. Graphitic foam-plasma interactions were also studied at the Wendelstein 7-X (W7-X) stellarator in Greifswald by exposure to hydrogen and helium plasmas using the Jülich multi-purpose manipulator. The purpose was to explore the possibility of using the material in a plasma facing component, and initial results were encouraging. In W7-X, no measurable erosion or cracking was observed. The PSI-2 samples received a deuterium fluence of 5 × 1025 m−2 resulting in an average erosion of 43 µm or about 5 mg per sample. Residual gas analysis (RGA) data were acquired to monitor sample outgassing. Laser-induced Breakdown Spectroscopy (LIBS) was used to measure deuterium retention in the porous foam. After exposure, the surfaces were characterized with scanning electron microscopy, energy dispersive x-ray analysis and time-of-flight secondary ion mass spectrometry (ToF-SIMS). The graphitic foam has a thermal conductivity as high as 287 W/mK and is considered as a replacement to more exotic carbon fiber composites such as SepCarb NB31 or isotropic graphites like ATJ that are no longer produced but used in present-day tokamak experiments. Actively cooled monoblocks were made from the foam and underwent extensive materials characterization including infrared response studies at Oak Ridge National Laboratory. This material is under consideration for the proposed actively-cooled W7-X divertor scraper element. Keywords: Graphitic foam, Graphite, Plasma facing component, Plasma exposure, Monoblock

We report measurements of He-ion-beam induced tungsten nanofuzz formation for normal and non-normal incidence angles in the energy range 218 eV–10 keV. At 218 eV, the fuzz tendrils are fine and grow randomly away from the interface in the direction of the surface normal. Above 480 eV, the fuzz tendrils become increasingly coarser, and their growth direction is in the direction of the incident beam. This change is attributed to the ion-induced displacement damage which becomes effective once the displacement damage threshold energy is exceeded, and produces additional near-surface trapping sites in those portions of the surface that are in direct line of sight of the incident beam which can nucleate He clusters and initiate bubble growth. Once the surface morphology roughens sufficiently for shadowing to occur, the subsequent fuzz growth occurs preferentially toward the incident ion beam. Molecular dynamics (MD) simulations were carried out to determine the displacement damage threshold energies in the near-surface region along the three major crystallographic directions. It was found that the tungsten bulk values are established within the first 2–4 atomic layers below the tungsten surface. SRIM simulations based on the MD energy thresholds indicate that vacancy damage production in the near-surface region quickly dominates over sputtering in near-surface lattice modification effects as the energy above the damage threshold increases.

The microstructural changes of a tempered F82H ferritic–martensitic steel following He implantation at 60 and 500 °C have been examined by transmission electron microscopy (TEM) and atom probe tomography (APT). After irradiation at 500 °C, numerous He bubbles were formed throughout the matrix, whereas after irradiation at 60 °C, no bubbles were seen to form in the matrix. In both irradiations, He bubbles were observed to have formed within large carbide precipitates, determined by APT compositional analysis to be M23C6. The observed preferential He bubble formation in carbides during low temperature He irradiation occurs as a result of the diffusing He being trapped in the carbide due to the strong He–C bond. As the He concentration increases in the carbide due to trapping, He bubbles are formed.