W and W-based high-entropy alloys are promising candidates for plasma-facing materials in fusion reactors. While irradiation studies on W have revealed a tendency for helium (He) bubble formation and radiation-induced defects, investigations of WTaCrV high-entropy alloy (HEA) have demonstrated superior radiation resistance, whether under He⁺ irradiation or heavy ion irradiation. To assess material performance under conditions relevant to fusion reactors -- characterized by fast neutrons and gas production from transmutation reactions -- complex irradiation environments need to be modeled. Using classical molecular dynamics simulations, we examined defect evolution in W and equimolar WTaCrV HEA with and without preexisting He atoms under overlapping displacement cascades up to 0.2 displacements per atom (dpa) at 300 K. In W, dislocation loops and large interstitial clusters formed readily, with increasing He content leading to higher dislocation densities and the formation of polygonal interstitial networks. In contrast, WTaCrV alloy exhibited strong resistance to formation of dislocation loops and large interstitial clusters but was more susceptible to bubble formation at higher He concentrations. Bubble growth was driven by helium trapping at vacancy sites and the coalescence of smaller bubbles. Larger bubbles remained stable against cascade overlap, limiting further growth by coalescence.
The development of advanced materials capable of withstanding the extreme conditions in nuclear fusion reactors represents a critical challenge for achieving sustainable fusion energy. Plasma-facing materials (PFMs), which are directly exposed to intense particle fluxes and extreme heat from high-temperature plasma, must simultaneously resist high-energy neutron irradiation and withstand helium (He) and hydrogen (H) fluxes from the plasma environment.
Tungsten (W) has been recognized as a leading plasma-facing material due to its remarkable properties. Its refractory nature, high melting point, excellent thermal conductivity, low erosion rates, and outstanding thermomechanical stability make it highly suitable for use as the first wall material in ITER and future fusion reactors. However, the formation of helium (He) bubbles during helium irradiation raises significant concerns about the long-term structural integrity of plasma-facing components manufactured from pure tungsten. In recent years, high-entropy alloys (HEAs) have been identified as promising alternatives to traditional materials for use in extreme environments. Among these, W-based HEAs have demonstrated exceptional properties, including high melting points and superior mechanical performance at elevated temperatures, surpassing those of Ni-based superalloys and nanocrystalline tungsten. Additionally, W-based HEAs have exhibited significantly enhanced radiation resistance compared to pure tungsten. For instance, alloys such as MoNbTaVW and MoNbTaTiW have shown greater irradiation resistance than pure tungsten. However, the inclusion of high-activation elements such as Mo and Nb in these alloys raises concerns regarding their long-term applicability as plasma-facing materials (PFMs). To address this issue, W-Ta-Cr-V HEAs have been developed as a potential solution. An experimental study on WTaCrV revealed remarkable irradiation resistance when exposed to 1 MeV Kr ions, even at high irradiation doses of up to 8 displacements per atom (dpa) at 800 °C. Unlike pure tungsten, this alloy exhibited no evidence of dislocation loop formation under these conditions. Furthermore, this HEA exhibited exceptional resistance to helium (He) bubble damage at 1223 K, with small (~ 2-3 nm) bubbles growing uniformly at a slow rate and no preferential formation on grain boundaries. Molecular dynamics (MD) simulations further revealed that, although the number of Frenkel pairs (FPs) created in the primary damage state of WTaCrV is higher than in pure tungsten, the interstitial cluster size and dislocation loop density are significantly lower. These findings highlight the superior radiation resistance of this HEA. Furthermore, ab initio simulations have attributed the radiation resistance of WTaCrV alloy to the slowed interstitial diffusion and the high recombination probability of interstitials and vacancies in the alloy. An experimental study on equimolar WTaCrV HEA has demonstrated the formation of helium (He) bubbles with diameters below 1 nm during helium irradiation, indicating exceptional resistance to He irradiation-induced defect accumulation. These results highlight the alloy's microstructural stability and resistance to irradiation hardening. Additionally, a recent atomistic simulation study demonstrated the exceptional resistance of equimolar WTaCrV HEA to surface modifications caused by energetic helium ions, further highlighting their potential for plasma-facing materials.
The majority of studies on HEAs have focused on their radiation response to heavy ion, neutron, or He irradiation. However, to evaluate the radiation response of materials under fusion reactor-relevant conditions -- characterized by fast neutrons and gas production from transmutation reactions -- a more complex irradiation environment is required. Atwani et al. addressed this challenge by mimicking such conditions through dual-beam irradiation with 1 MeV Kr⁺ and 16 keV He⁺ ions to test the radiation resistance of WTaCrVHf and WTaCrVHf at 1173 K. Their findings revealed no dislocation loops, even after 8.5 displacements per atom (dpa) and 9.13% He implantation. However, cavities were observed under these conditions.
Despite these promising results, no fundamental studies have yet been conducted to investigate the formation and evolution of defects in W-Ta-Cr-V HEAs under complex irradiation environments. Specifically, the combined effects of preexisting helium (introduced through He injection) followed by neutron or ion irradiation remain unexplored. In this work, we present a direct comparison of the radiation resistance of tungsten and equimolar WTaCrV HEA with 0%, 1%, and 2% He, added during the simulation setup. Overlapping displacement cascade simulations were employed to mimic neutron irradiation. These simulations, widely used to model radiation damage in metals and alloys, provide a robust platform for studying the production and evolution of defects. Utilizing a recently developed interatomic potential -- constructed based on density functional theory (DFT) data for WTaCrV-He properties and validated against experimental observations of He bubble growth -- MD simulations provide a reliable platform for evaluating the comparative performance of these materials. The insights obtained from this study will enhance our understanding of the behavior of plasma-facing components (PFCs) under complex irradiation conditions. Furthermore, this work evaluates whether W-Ta-Cr-V HEAs remain strong candidates for advanced plasma-facing materials in the demanding environments of fusion reactors.