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Dislocation, playing a crucial role in the plastic deformation of metals, can be significantly affected by introducing solute elements. Hydrogen (H) embrittlement is one such example, while the underlying mechanism for H affected dislocation structural stability and mobility remains unclear and the role of H has been controversial. Here, using first-principles calculations, we demonstrate that the effect of H on screw dislocation in bcc metals is H concentration-dependent, signified by a H-induced transition of SD core structure. At low concentrations of H segregation, dislocation maintains the intrinsic easy-core structure, and H atoms are attached to the "periphery" of dislocation to enhance dislocation motion. In contrast, at high H concentrations, dislocation transforms into a hard-core, metal hydride-like structure, as H atoms become the "body" of dislocation to significantly reduce the dislocation mobility. Further, such local easy-to-hard transition is found to be induced by just one solute of other elements, including helium, carbon, nitrogen and oxygen, independent of solute concentration. Our work sheds new light on the H-dislocation interactions in bcc metals, having broad implications in the interstitial solute-related phenomena.

Understanding the evolution of irradiation-induced defects is of critical importance for the performance estimation of nuclear materials under irradiation. Hereby, we systematically investigate the influence of He on the evolution of Frenkel pairs and collision cascades in tungsten (W) via using the object kinetic Monte Carlo (OKMC) method. Our findings suggest that the presence of He has significant effect on the evolution of irradiation-induced defects. On the one hand, the presence of He can facilitate the recombination of vacancies and self-interstitial atoms (SIAs) in W. This can be attributed to the formation of immobile He-SIA complexes, which increases the annihilation probability of vacancies and SIAs. On the other hand, due to the high stability and low mobility of He-vacancy complexes, the growth of large vacancy clusters in W is kinetically suppressed by He addition. Specially, in comparison with the injection of collision cascades and He in sequential way at 1223 K, the average sizes of surviving vacancy clusters in W via simultaneous way are smaller, which is in good agreement with previous experimental observations. These results advocate that the impurity with low concentration has significant effect on the evolution of irradiation-induced defects in materials, and contributes to our understanding of W performance under irradiation.

In this paper, we present an interatomic potential for the ternary W–Ta–He system, which is an extension of our previous W–Ta potential. The new potential parameters for the W–He and Ta–He interactions are determined by fitting the results obtained from first-principles calculations. The formation energies of a single He atom at different sites, the binding energies of He–He and He–Vac (Vac = vacancy) in W/Ta, and the binding energies of a single He atom with a solute Ta atom in W are used as the fitting targets. Then, the binding energies of an additional interstitial He atom to existing HenVacm clusters in W/Ta are calculated, and the results reported correspond with the results from the first-principles. Furthermore, the effects of solute Ta on the diffusion and aggregation of He in bulk W are studied. We observed that small interstitial He atom clusters (NHe ≤ 4) were easy to diffuse in pure W, and their diffusion activation energies were less than 0.3 eV. However, the binding energies between Ta and these clusters were between 0.5 and 0.9 eV, which has a pinning effect on the He cluster diffusion. At high temperature, the solute Ta cannot qualitatively hinder the agglomeration of He atoms, however, due to the pinning effect compared with pure W, solute Ta in W has a certain delay effect on the agglomeration of He atoms in time.

We have investigated the effects of noble gas elements (helium, neon, argon and krypton) on the dissolution and accumulation behavior of hydrogen (H) in tungsten (W) using a first-principles method, as well as the behaviors of themselves. Noble gas atoms energetically prefer to occupy the tetrahedral interstitial site (TIS) in W, as the same of H. The TIS → TIS is their optimal diffusion path, and the diffusion energies increase with the increasing of atomic radius. All of them are energetically favorable clustering by self-trapping. It is found that the presence of the noble gas elements have significant effect on H behavior in W. Both interstitial noble gas atoms and their complex with vacancy can serve as strong trapping centers of H, which can be attributed to the redistribution of electron density and lattice distortion induced by noble gas. Most importantly, we have demonstrated that H trapping capability of noble gas clusters will increase with the increasing of the number of noble gas atoms in both interstitial and vacancy-complex cases. The H trapping energy surrounding the noble gas clusters with four atoms is comparable to that in noble gas-free vacancy. Further, the formation of noble gas clusters is much easier than that of vacancy due to the self-trapping interaction between noble gas atoms (>1 eV for interstitial case and >4 eV for vacancy-complex case), and thus the H trapping centers will rapidly increase after noble gas pre-irradiation/implantation. Consequently, the self-trapping character of noble gas induces their strong trapping effect on H.

The influence of rhenium (Re) on helium (He) evolution in tungsten (W) with different Re distribution is investigated using the first-principles calculations in combination with thermodynamic models. It is found that the He behaviors in W-Re system are closely related to the Re distribution. For the dispersed distribution of Re in W, it is found that a single Re atom only has slight effect on the behaviors of a He atom due to the weak interaction of Re-He1. Interestingly, with the increasing of He numbers, the binding energy between a single Re and He clusters increases rapidly from 0.03 eV for Re-He1 to 0.72 eV for Re-He4. Such high binding energy indicates the strong attraction of Re with He clusters in W. This can be rationalized by the large lattice distortion induced by He clusters, which reduces the electron density of He-occupied position and leads to the transition of neighboring Re/W atom from a substitutional site to an interstitial site. Because of the strong attraction of Re with He clusters, the Re addition effectively reduce the mobility of He clusters in W. As for the aggregated Re distribution, the He trapping capability of vacancy is weakened by Re clusters. These results are well consistent with the experimental observed suppressing effects of Re on the He-induced morphology in W. Further, based on the He behaviors in W-Re σ phase, we suppose that He will segregate from bulk W to Re-rich precipitates, thus enhancing the formation of He clusters. Consequently, our calculations suggest that the distribution of Re plays a key role on the He behaviors in W, which provides a good reference for the design and development of W-based alloys.

Tungsten (W) is regarded as the most promising candidates for plasma-facing materials (PFMs) in future fusion reactors. However, W will be exposed to high fluxes of helium (He) plasma in a nuclear fusion environment, which will degrade the thermal and mechanical properties of W. We have investigated the behavior of He in W using a first-principles method based on density functional theory. It has demonstrated that the quantitative stress indicator and the qualitative hard-sphere model can be used to characterize the effective volume and the stability of interstitial He in W. The hydrostatic tensile (compressive) strain will enhance (suppress) the dissolution and diffusion of He in W. The existing of He will have negative effect on the ideal strength of W. Vacancies and grain boundaries defects in W can serve as trapping centers for He, because the defects can provide large space for He. He will spontaneously form He clusters at the near surface by self-trapping. The pre-existing impurity/alloying element can significantly reduce the solution energy of He in W due to the electron density redistribution induced by them, such as C and Nb. The results will provide good reference for developing W materials as PFMs. This paper reviews our recent findings regarding the behavior of He in W.

Using a first-principles method, we have investigated dissolution, self-trapping and clustering of He at a W(1 1 0) surface. We found that the He atom is not energetically favorable at both the surface and the subsurface, but it becomes stable under the second atomic layer from the surface. The He is easier to be self-trapped to form an He cluster at the near surface in comparison with the bulk due to the larger self-trapping range and the stronger binding energy. With the formation of such He cluster, the vacancy and thus the He-vacancy complex are able to form at the near surface. The results will provide a useful reference for understanding formation of the He bubble at the W surface.

We have investigated the effects of strain on the dissolution and diffusion properties of helium (He) in tungsten (W) using a first-principles method. We show that the solubility and diffusion of He can be enhanced (suppressed) with increasing tensile (compressive) strain. This can be attributed to the variation of the effective volume and the p-projected DOS of He induced by the strain. We demonstrate that the strong electronic interaction between He and W atoms plays a key role in the poor solubility of He in W. The tetrahedral interstitial site is always the preferred site of He under all strains. The same rule is believed to hold for other inert gas elements in strained W.

We perform first-principles calculations based on density functional theory to investigate energetics and site preference of He in a bcc-W Σ = 5 grain boundary (GB). The segregation energy is calculated to be −1.37 eV, indicating that He prefers to segregate in the W GB. The formation energy of He in the W GB is positive and thus He is quite hard to dissolve in the W GB, similar to its behavior in the bulk. Because of its closed-shell electronic structure, He is shown to preferably occupy either interstitial or substitutional site with larger space provided by the GB, changing the GB electronic structure. Moreover, segregation of He gives rise to the W GB expansion. These structure variations can have a large effect on the mechanical properties of the W GB.

Using a first-principles computational tensile test (FPCTT), we have investigated the effect of helium (He) on the structure and bonding properties of tungsten (W), which is a promising plasma-facing material in nuclear fusion Tokamak. Density of states results reveal the underlying reason that the substitutional site for He is the most energetically favorable, while the tetrahedral interstitial site is more favorable than the octahedral interstitial one. The FPCTT shows that the ideal tensile strength is 29.1 GPa at the strain of 14% along the [0 0 1] direction for intrinsic W, while it decreases to 28.2 GPa at the same strain when one impurity He atom is introduced. A local bond-breaking region around He forms in the tensile process due to the presence of He, which suggests He will have a large effect on the bonding properties of W.

Vacancy-type defects created by helium implantation in tungsten and their impact on the nano-hardness characteristics were investigated by correlating the results from the positron annihilation spectroscopy and the nano-indentation technique. Helium implantation was performed at room temperature (RT) and at an elevated temperate of 600 °C. Also, the effect of post-annealing of the RT implanted sample was studied. The S parameter characterizing the open volume in the material was found to increase after helium irradiation and is significantly enhanced for the samples thermally treated at 600 °C either by irradiation at high temperature or by post-annealing. Two types of helium-vacancy defects were detected after helium irradiation; small defects with high helium-to-vacancy ratio (low S parameter) for RT irradiation and large defects with low helium-to-vacancy ratio (high S parameter) for thermally treated tungsten. The hardness of the heat treated tungsten coincides with the S parameter, and hence is controlled by the large helium-vacancy defects. The hardness of tungsten irradiated at RT without thermal treatment is dominated by manufacturing related defects such as dislocation loops and impurity clusters and additionally by trapped He atoms from irradiation effects, which enhance hardness. He-stabilized dislocation loops mainly cause the very high hardness values in RT irradiated samples without post-annealing.

We investigate the effect of carbon (C) on helium (He) trapping in tungsten (W) using a first-principles method. We show C can effectively reduce the solution energy of He in the bulk W originated from the charge density redistribution. This leads to a strong attraction between He and C in W. We demonstrate the C–vacancy (C–V) complex can serve as a trapping center by reducing charge density in its vicinity to induce He nucleation in comparison with the defect-free W. The maximal number of He atoms that can be trapped by such C–V complex is 5, and He diffusion into the C–V complex is kinetically feasible. Further, it is found the binding energy of He to a C–V complex is weaker than that to a C-free vacancy, suggesting C will decrease the He trapping capability of vacancy. We thus propose that C plays a key role in He trapping behavior.

Fission gas retention or release has a critical impact on the function of advanced nuclear materials. Helium trapping in, and release from, radiation defects induced by neutrons and by α decay in YSZ (yttria-stabilized zirconia) is experimentally simulated using synchronized Zr+ and He+ dual ion beam irradiation. The measured damage profiles consist of two peaks which agree well with the calculated profiles of implantation induced excess point defects. This special implantation related effect has to be carefully considered in the evaluation of experimental investigations which simulate isotropic irradiation effects such as α decay. First-principles calculations show that helium is energetically favorable to be trapped by Zr vacancies in YSZ. Implanted helium alone in YSZ is accumulated in undesirable helium bubbles and results in local surface swelling and lift-off. However, under dual beam irradiation helium is released from vacancy defects and is out-diffused at room temperature. Helium is mobilized by a vacancy-assisted trapping/detrapping mechanism induced by the simultaneous Zr+ ion implantation. This behavior avoids the deleterious helium bubble formation and contributes to the suitable application characteristics of YSZ which result in its excellent radiation hardness.

The effects of helium (He) on the sliding and mechanical properties of a vanadium (V)Σ5(310)/[001] grain boundary (GB) have been investigated using a first-principles method. It has been found that He was energetically favorable sitting at the GB region with a segregation energy of −0.27 eV, which was attributed to the special atomic configurations and charge density distributions of the GB. The maximal sliding energy barrier of the He-doped GB was calculated to be 1.73 J/m2, ~ 35% larger than that of the clean GB. This suggested that the presence of He would hinder the V GB mobility. Based on the thermodynamic criterion, the total energy calculations indicated that the embrittlement of V GB would be enhanced by He segregation.

We demonstrate a quantitative stress indicator as well as a qualitative hard-sphere lattice model to characterize the effective volume and stability of interstitial helium in fcc and bcc metals, based on extensive first-principles total-energy and lattice stress calculations in combination with continuum elastic theory analyses. The concept of stress indicator is believed to be generally applicable to quantify the relative stability of other inert gas elements in metals.

We investigate the physical origin of H–He interaction in W in terms of optimal charge density by calculating the energetics and diffusion properties using a first-principles method. On the one hand, we show a strong attraction between H and He in W originated from the charge density redistribution due to the presence of He, driving H segregation towards He. This can block the permeation of H into deeper bulk and thus suppress H blistering. On the other hand, we demonstrate that He, rather than H, energetically prefers to occupy the vacancy centre due to its closed-shell structure, which can block H2 formation at the vacancy centre. This is because He causes a redistribution of charge density inside the vacancy to make it ‘not optimal’ for the formation of H2 molecules, which can be treated as a preliminary nucleation of the H bubbles. We thus propose that H retention and blistering in W can be suppressed by doping the noble gas elements.