Modeling tungsten response under helium plasma irradiation: a review

Helium (He)-induced damages are sensitive concerns for the performance of tungsten plasma facing materials (W-PFMs). Recent experiments have revealed that trace impurities in He plasma can effectively prevent the formation of He bubbles and fuzz on W surfaces. To explore its plausibility and underlying mechanism, we performed a multiscale computational study that combines density functional theory calculations and object kinetic Monte Carlo simulations to investigate the effects of a small quantity of beryllium (Be) on the evolution of He bubbles. It is found that there is strong attractive interaction between He and Be, which can be attributed to the decrease of electron density and the lattice distortion induced by embedded Be atom. Therefore, the co-implantation of Be continuously introduces trapping centers for He. Due to low implantation depth and high migration energy of Be, the Be atoms are located close to the surface, leading to the trapping of the majority of He within the near-surface region and the development of a shielding layer for He permeation. The presence of Be facilitates the dispersion of the trapped He, skewing the He clusters into smaller size. More importantly, the Be trapping centers bring the He clusters closer to the surface, significantly increasing the probability of bubble bursting and the release of He back to the vacuum. These ultimately leads to a lower retention of He in the case of He+Be co-irradiation, compared with the case of He-only irradiation. Consequently, our findings elucidate the suppressive effect of a low flux of Be atoms on the growth of He bubbles, highlighting the need to focus on the synergetic effects between plasma species.

Exploring the suppression methods of helium-induced damage in tungsten by investigating the interaction between beryllium and helium

Zhou,

Li,

Lin

et al. 2024

The impact of helium on plasma-driven hydrogen permeation and implications for direct internal recycling in the fusion fuel cycle

Li,

Way,

Fuerst

et al. 2024

Metal foil pumps (MFPs) are the leading technology for direct internal recycling (DIR) of hydrogen isotopes from the plasma exhaust in future fusion plants. MFPs rely on the concept of superpermeation, where superthermal H atoms directly absorb into the metal foil, rapidly diffuse, and desorb downstream. To date, studies of superpermeation have predominantly employed either pure hydrogen or in some cases trace levels of impurities. The plasma exhaust is expected to contain just ~1% helium, but in DIR the source gas would be enriched in helium as hydrogen isotopes are extracted. In this work, we explore the impact of helium on hydrogen superpermeation at low temperature (75-200 ºC) using Pd-based foils. To first order, the flux scaled linearly with the hydrogen mole fraction. Stable permeation was observed until the helium fraction reached ~80%, where the flux began to decline slowly with time. In addition, short term (1-5 minute) exposure to pure helium plasma significantly attenuated subsequent hydrogen plasma permeation, and the degree was more dramatic at elevated temperature. This attenuation was correlated with He retention in the foils, which was detected by time-of-flight secondary ion mass spectrometry at low levels (< 0.1 at. %) and limited to the near surface (< 10 nm). Similar trends were observed among all alloys (Pd, PdAg, PdCu), and the foils were restored to full performance with an Ar+ sputter clean. The potential for helium plasma exposure to impact MFP performance under these conditions has not been previously reported, and these findings have significant implications to the design and implementation of practical DIR systems.

Surface damage in tungsten induced by high heat flux helium irradiation at high temperatures across melting point

Wang,

Yuan,

et al. 2024

Understanding the behavior of tungsten (W) surface damage under the synergistic effects of high heat flux loading and helium (He) irradiation is essential for predicting material performance during off-normal operations in ITER. In this study, surface modifications occurring at high temperatures (> 2200 K) up to the melting point were investigated by conducting experiments involving two campaigns of Vertical Displacement events (VDEs) like high heat flux He neutral beam pulse irradiation on polycrystalline W samples at the test facility GLADIS. As the surface temperature of W increased due to irradiation (2253 - 3683 K), pinholes appeared on the surface, showing a trend of increasing size and decreasing number density, indicating severe lattice damage. Accordingly, we proposed a model for pinhole growth under high-temperature He irradiation based on thermal activation diffusion of He. The calculated activation energy for He diffusion in this process was found to be 0.51 eV, which is considerably higher than the results obtained from previous simulations (0.021 – 0.157 eV) [1–4]. This suggests that extensive defects in the matrix have a significant impact on the diffusion of He in high-temperature environments, which is distinct from diffusion behavior at lower temperatures. However, as the surface temperature further increased beyond the melting point, the melting and re-solidification process nearly completely repaired almost all defects induced by He ion irradiation. The re-solidified grains were characterized by being intact, damage-free, and having lower residual stress. This study establishes a foundation for the quantitative analysis of helium migration mechanisms under high-temperature helium irradiation, which lays the foundation for understanding material structural damage behavior under off-normal operations for ITER.