Comparative study on radiation resistance of WTaCrV high-entropy alloy and tungsten in helium-containing conditions

Helium accumulation and bubble formation in FeCoNiCr alloy under high fluence He+ implantation

Effect of H+ pre-irradiation on irradiation hardening and microstructure evolution of FeCoCrNiAl0.3 alloy with He2+ irradiation

Interstitial strengthening is one of the main approaches to improving the mechanical properties of high-entropy alloys (HEAs), but its effects on the irradiation resistance of HEAs need further study. Here, we investigated the irradiation-induced defects and swelling of Fe38Mn40Ni11Al4Cr7 HEAs with different carbon contents under 5 MeV Xe23+ irradiation at 300 °C and 500 °C. Results show that the irradiation-induced swelling was significantly suppressed as the carbon content increased. Under the observation of TEM, the size of irradiation-induced dislocation loops also decreases with increasing carbon content. By comparing the effects of carbon content at different temperatures on the evolution of defects, the pinning effect of interstitial carbon on irradiation-induced defects of HEAs was proposed and analyzed. Carbon atoms, which are stabilized in the octahedron clearance of HEAs with FCC structure, not only promote the recombination of point defects by enhancing the sluggish diffusion effect of HEAs, but also pin the common 1/3 faulted loops caused by irradiation. This pinning effect is the main mechanism of interstitial carbon for improving the irradiation resistance of HEAs below 300 °C. In summary, this study provides an essential experimental basis for the irradiation effects of carbon-doped HEAs and strives to reveal the effect of interstitial carbon on irradiation-induced defects at different temperatures.

Helium bubble formation in nickel under in-situ krypton and helium ions dual-beam irradiation

Among the currently developed multi-principal element alloys (MPEAs), Ti-V-Ta MPEA stands out because of its good high-temperature strength, good room-temperature plasticity, stable organizational structure, and low neutron activation, and becomes a first option for cladding material of special power reactors. The radiation resistance of Ti-V-Ta MPEA is the focus of current research. Dislocation loops are the main irradiation defects in Ti-V-Ta MPEA, which can significantly affect the mechanical properties. Therefore, clarifying the formation mechanism of dislocation loops in Ti-V-Ta HEA can help to understand its radiation resistance. The formation behavior of dislocation loops in Ti-V-Ta MPEA is studied based on molecular dynamics method in this work. Cascade overlap simulations with vacancy clusters and interstitial clusters are carried out. The cascade overlap formation mechanism of dislocation loops is analyzed and discussed. In Ti-V-Ta MPEA, the cascade overlap with defect clusters can directly produce different types of dislocation structures. The defect configuration after cascade overlap is determined by the primary knock-on atom (PKA) energy and the type and size of the preset defect clusters. Cascade overlap can improve the formation probability of <inline-formula><tex-math id="M6">\begin{document}$ \left\langle {100} \right\rangle $\end{document}</tex-math></inline-formula> dislocation loops in Ti-V-Ta MPEA. Cascade overlap with vacancy clusters is an important mechanism for the formation of <inline-formula><tex-math id="M7">\begin{document}$ \left\langle {100} \right\rangle $\end{document}</tex-math></inline-formula> vacancy dislocation loops, and the size of vacancy clusters is the dominant factor for the formation of <inline-formula><tex-math id="M8">\begin{document}$ \left\langle {100} \right\rangle $\end{document}</tex-math></inline-formula> vacancy dislocation loops. When the PKA energy is enough to dissolve the defect clusters, <inline-formula><tex-math id="M9">\begin{document}$ \left\langle {100} \right\rangle $\end{document}</tex-math></inline-formula> vacancy dislocation loops are more likely to form. Furthermore, cascade overlap with interstitial clusters in Ti-V-Ta MPEA is a possible mechanism for the formation of <inline-formula><tex-math id="M10">\begin{document}$ \left\langle {100} \right\rangle $\end{document}</tex-math></inline-formula> interstitial dislocation loops. This study can contribute to understanding the evolution behavior of irradiation defects in Ti-V-Ta MPEA, and provide theoretical support for designing the composition and optimizing the high-entropy alloys.

Diffusion controlled helium bubble formation resistance of FeCoNiCr high-entropy alloy in the half-melting temperature regime

Effects of single and simultaneous Fe, He ions irradiations on CoCrFeMnNi high entropy alloy

Besides high dose radiation and extreme thermal loads, a major concern for materials deployed in novel nuclear fusion reactors is the formation and growth of helium bubbles. This work investigates the swelling and mechanical property degradation after helium implantation of ultrafine-grained W and nanocrystalline W-Cu, possible candidates for divertor and heat-sink materials in fusion reactors, respectively. It is found that ultrafine-grained W and single crystalline W experience similar volumetric swelling after helium implantation but show different blistering behavior. The W-Cu nanocomposite, however, shows a reduced swelling compared to a coarse-grained composite due to the effective annihilation of radiation-induced vacancies through interfaces. Furthermore, the helium-filled cavity structures lead to considerable softening of the composite.

In fusion reactors, plasma facing components (PFC) and, in particular, the divertor will be irradiated with high fluxes of low-energy (∼100 eV) helium and hydrogen ions. Tungsten is one of the leading candidate divertor materials for ITER and DEMO fusion reactors. However, the behaviour of tungsten under high dose, coupled helium/hydrogen exposure remains to be fully understood. The PFC response and performance changes are intimately related to microstructural changes, such as the formation of point defect clusters, helium and hydrogen bubbles or dislocation loops. Computational materials' modelling results are described here that investigate the mechanisms controlling microstructural evolution in tungsten. The aim of this study is to understand and predict sub-surface helium bubble growth under high flux helium ion implantation (∼1022 m−2 s−1) at high temperatures (>1000 K). We report results from a spatially dependent cluster dynamics model based on reaction–diffusion rate theory to describe the evolution of the microstructure under these conditions. The key input parameters to the model (diffusion coefficients, migration and binding energies, initial defect production) are determined from a combination of atomistic modelling and available experimental data. The results are in good agreement with results of an analytical model that is presented in a separate paper. In particular, it is found that the sub-surface evolution with respect to bubble size and concentration of the helium bubbles strongly depends on the flux and temperature.

Effects of carbon doping on irradiation resistance of Fe38Mn40Ni11Al4Cr7 high entropy alloys

We review recent results on the mechanisms of atomic mixing, radiation-induced disordering, and defect production and clustering induced by displacement cascades in Cu, Ag, Cu3Au and Ni3Al. We employ molecular dynamics computer simulation (MD) methods with isotropic many body potentials and recoil energies near the subcascade formation regime (up to 30 keV). Atomic mixing is shown to be dominated by diffusion in the locally molten cascade core in both the pure metals and the intermetallic compounds. Disordering of the intermetallics was found to take place in the core of the cascades. However, because of the short lifetime of the displacement cascade, chemical short range order is preserved in the molten zone. Our results reveal the appearance of very large vacancy and interstitial type defect clusters at high recoil energy and cascade energy density. Vacancies agglomerate and collapse into Frank dislocation loops as the result of the quenching of the cascade molten core. Large interstitial clusters are directly produced in cascades and form prismatic dislocation loops. The fraction of defects in clusters for low temperature cascades increases with recoil energy and approaches ∼70% and 60% for interstitials and vacancies, respectively, at recoil energies near the threshold for subcascade formation. In the case of intermetallic (A3B) compounds, the large energy required to produce and transport a superdislocation appears to inhibit the interstitial prismatic loop punching process and interstitials appear as isolated (100) dumbbells formed by pairs of majority-type atoms.

The effect of interstitial carbon atoms on defect evolution in high entropy alloys under helium irradiation

The formation of helium bubbles and subsequent property degradation poses a significant challenge to tungsten as a plasma-facing material in future long-pulse plasma-burning fusion reactors. In this study, we investigated helium bubble formation in dispersion-strengthened tungsten doped with transition metal carbides, including TaC, ZrC, and TiC. Of the three dispersoids, TaC exhibited the highest resistance to helium bubble formation, possibly due to the low vacancy mobility in the Group VB metal carbide and oxide phases. Under identical irradiation conditions, large helium bubbles formed at grain boundaries in tungsten, while no bubbles were observed at the interfaces between the carbide dispersoid and tungsten matrix. Moreover, our results showed the interfaces could suppress helium bubble formation in the nearby tungsten matrix, suggesting that the interfaces are more effective in trapping helium as tiny clusters. Our research provided new insights into optimizing the microstructure of dispersion-strengthened tungsten alloys to enhance their performance.

In fusion reactors, plasma facing components (PFC) and in particular the divertor will be irradiated with high fluxes of low energy (∼100 eV) helium and hydrogen ions. Tungsten is one of the leading candidate divertor materials for ITER and DEMO fusion reactors. However, the behavior of tungsten under high dose, coupled helium/hydrogen exposure remains to be fully understood. The PFC response and performance changes are intimately related to microstructural changes, such as the formation of point defect clusters, helium and hydrogen bubbles or dislocation loops. Computational materials modeling has been used to investigate the mechanisms controlling microstructural evolution in tungsten following high dose, high temperature helium exposure. The aim of this study is to understand and predict helium implantation, primary defect production and defect diffusion, helium-defect clustering and interactions below a tungsten surface exposed to low energy helium irradiation. The important defects include interstitial clusters, vacancy clusters, helium interstitials and helium-vacancy clusters. We report results from a one-dimensional, spatially dependent cluster dynamics model based on the continuum reaction–diffusion rate theory to describe the evolution in space and time of all these defects. The key parameter inputs to the model (diffusion coefficients, migration and binding energies, initial defect production) are determined from a combination of atomistic materials modeling and available experimental data.

The sluggish diffusion and severe lattice distortion effects in high-entropy alloys (HEAs) theoretically impede the movement of radiation-induced point defects, thereby effectively suppressing the formation of larger defect clusters and ultimately enhancing the radiation resistance of materials. Current research on the radiation resistance of HEAs primarily concentrates on the qualitative analysis of the migration behaviors of radiation-induced defects, while the quantitative research on the energy barriers of the migration behavior of point defects is still limited. As a representative HEA system, FeCoCrNiAl-based alloys exhibit exceptional properties, including enhanced ductility, remarkable shear resistance, high tensile yield strength and excellent oxidation resistance. In this study, we selected the FeCoCrNiAl<sub>0.3</sub> alloy as the model material and conducted in-situ observations using a 1.25 MV high voltage electron microscope (HVEM) to systematically investigate the temporal evolution of irradiation-induced defects and precipitates under different temperatures. Based on the statistical data of saturated defect number density and defect growth rates under three irradiation temperatures, two intrinsic parameters of the alloy - the interstitial atom migration energy and vacancy migration energy - were determined respectively, which is 1.09 eV and 1.47 eV, respectively. The higher interstitial atomic migration energy may be related to the incorporation of Al, which has a smaller threshold energy and exhibits a larger atomic radius difference compared to the other elements in the alloy. In addition, the morphology and distribution of dislocation loops formed under 723 K high energy electron irradiation were characterized in detail, revealing the coexistence of perfect dislocation loops and Frank dislocation loops, both of which grow along different crystal planes. No systematic difference is observed between the growth process of the two types of loops.