Clarifying the formation pathways of a0/2⟨110⟩ perfect loops (PLs) is pivotal for understanding radiation damage evolution in face-centered cubic (FCC) alloys. While traditional models emphasize the transformation of Frank loops (FLs), direct nucleation of PLs under irradiation remains underexplored. Here, using Pd-25Ag as a model alloy, we employ in-situ ion irradiation to trace dislocation loop evolution. We identify and characterize three distinct formation mechanisms for PLs: direct nucleation, FL self-unfaulting, and loop reaction. A morphology-mechanism correlation of dislocation loops is established based on in-situ TEM observation and detailed crystallographic analysis, providing a diagnostic framework for microstructural analysis. Crucially, the proportion of directly nucleated PLs is significantly higher in Pd-25Ag than in pure Pd. This enhancement suggests the dual role of the alloying element Ag: It thermodynamically promotes nucleation by lowering the stacking fault energy, and kinetically aids loop stabilization by suppressing point-defect migration. This work decouples the solute effects on defect kinetics and offers a morphology-based paradigm for interpreting irradiation-induced defects in FCC alloys.

The interplay between displacement defects governs the evolution of irradiation damage in materials and is of great fundamental interests with important practical implications, from microelectronics industry to advanced nuclear system. Hydrogen, a ubiquitous impurity, is known to segregate to vacancies, but its role in altering vacancy-interstitial recombination-the key process underlying defect annihilation-has not been established. Here, using tungsten as a model system, we show that hydrogen adsorption on the inner surfaces of vacancy clusters significantly suppresses recombination with self-interstitial atoms, thereby inhibiting defect annihilation. We identify a stress-mediated mechanism in which hydrogen adsorption transforms the local stress field of vacancy clusters, weakening their long-range attraction to self-interstitial atoms. Based on this mechanism, we develop a predictive model that quantitatively relates the relative reduction of recombination radius to the hydrogen inner surface density, independent of cluster size. By integrating atomistic parametrization with multiscale simulations, we investigate the co-evolution of hydrogen and displacement defects, which show quantitative agreement with recent experiments, including the hydrogen isotope retention, distribution and desorption. Our results establish a direct link between impurity-defect interactions and defect-defect recombination, providing a physically grounded framework for understanding and controlling irradiation damage in structural materials.

As a technologically important class of materials, group-IIIA oxides M2O3 (M = Al, Ga, In) have been extensively applied in various areas, yet their polaronic nature of self-trapped excess charge remains controversial. Here, utilizing the recently developed first-principles theory of polarons, we investigate the formation of hole polarons in M2O3 to clarify whether they are small or large. We consider the most common rhombohedral and monoclinic phases of M2O3 and reveal that an excess hole forms a highly anisotropic large polaron in rhombohedral Al2O3, while it also forms a large polaron in monoclinic Al2O3 but with much weaker anisotropy. For Ga2O3, our calculations confirm the formation of a small polaron in monoclinic phase, but a large polaron is more favorable in rhombohedral phase. In rhombohedral In2O3, the hole is found to exhibit as a large polaron with an ellipsoid feature. The drastically different polaronic characteristics in M2O3 are explained in terms of the distinct electronic band structures and phonon spectra. This work not only clarifies the existing controversies between the available theoretical calculations and experimental observations, but it also sheds light on the potential applications of oxide materials with diverse functionalities.

Tungsten is a candidate for plasma-facing components in fusion reactors, where it must withstand extreme heat and neutron irradiation. Nanocrystalline tungsten (NC-W) shows improved radiation tolerance due to grain boundaries acting as defect sinks, but its plasticity degrades after irradiation. Although the mechanical response depends on the interplay between grain size and defect microstructure, the underlying mechanisms remain unclear. Using molecular dynamics, we investigate the combined effects of grain size and vacancy defects on the plastic deformation of NC-W. In defect-free samples, decreasing grain size shifts deformation from dislocation glide to twinning, leading to strengthening followed by softening, with a Hall-Petch to inverse Hall-Petch transition at approximately 25.0 nm. Vacancy defects alter this behavior: randomly distributed single vacancies cause only minor softening, whereas vacancy clusters suppress twinning and promote dislocation-mediated and grain-boundary-mediated deformation. Consequently, vacancy clustering weakens the grain-size dependence of flow stress and promotes vacancy-assisted softening, accompanied by pronounced strain localization. These results indicate that grain refinement alone cannot ensure mechanical stability in irradiated W. Instead, optimal performance requires coordinated control of grain size and vacancy defects, particularly through microstructures that suppress vacancy-cluster accumulation.

Since W/Cu laminates have great potential application in the plasma-facing components, understanding their response to high-energy neutron is therefore an essential requirement. Recent experiments have reported unexpected brittle fracture in W/Cu laminates following low-dose neutron irradiation, yet the underlying mechanism remains unclear. Here, we systematically investigate the behaviors of point defect in W(110)/Cu(111) interface and its influence on the evolution of displacement damage as well as the hardness increase in W using the multi-scale simulations. It is found that the W/Cu interface can serve as the effective trapping center for point defects in both W and Cu. Intriguingly, the defect segregation into the interface substitutes the Cu atoms, which further traverse the interface and enter the W side. Due to the different atomic radius and/or electron structure, Cu atoms interact strongly with self-interstitial atoms (SIAs) and also bind with vacancies in W, thereby forming Cu-SIA and Cu-Vac complexes. Especially for the interstitial Cu, its diffusion energy barrier is only 0.34 eV, corresponding to the dominant method for Cu diffusion in irradiated W. Combining the atomistic parameters with object kinetic Monte Carlo (OKMC) simulations, we demonstrate that the Cu addition increases the number of surviving defects in W under neutron irradiation. This can be attributed to the strong attraction between Cu and SIAs, resulting in the formation of immobile Cu-SIA complexes and suppressing the vacancy-interstitial recombination. Such effect largely depends on the Cu insert rates and temperatures, and also leads to an extra contribution on the irradiation hardening of W. Our findings provide a plausible explanation for experimental observations and highlight the crucial role of interfaces in governing material responses to neutron irradiation.

Tungsten is a leading plasma-facing material for fusion reactors, but prolonged high-heat loads can induce recovery, recrystallization, and grain growth that degrade its performance. In this study, the microstructural evolution, dislocation dynamics, and texture changes of ITER-like tungsten (IG-W) annealed at 1000–1600 ℃ were investigated by electron backscatter diffraction (EBSD) and X-ray diffraction (XRD). A two-stage thermal evolution was identified: a recovery-dominated regime below 1400 ℃, followed by a regime of primary recrystallization and anisotropic grain growth at 1500 ℃ and above. During recovery, the XRD-derived dislocation density exhibited a transient peak at 1400 ℃, while the EBSD-derived geometrically necessary dislocation (GND) density remained nearly constant at . This divergence indicates that recovery mainly involves dislocation rearrangement and subgrain formation rather than dislocation multiplication. At 1500 ℃ and above, recrystallization led to a sharp decrease in GND density and a pronounced increase in the high-angle grain boundary fraction. Orientation analysis further revealed that {211}-oriented grains dominated up to 1400 ℃ but were subsequently consumed by {321}- and {310}-oriented grains, resulting in a non-monotonic texture evolution. These findings clarify the thermal stability thresholds and defect elimination kinetics of tungsten for fusion applications.

As ion irradiation techniques have increasingly extended from three-dimensional (3D) bulk targets to low-dimensional nanomaterials, there is a growing need for developing an efficient tool capable of simulating ion irradiation for materials of different dimensionalities. Here, we present a general framework ASIM (Atomic-scale Simulation on Ion Irradiation of Matter) to achieve full-cascade ion irradiation simulations of multi-dimensional systems, ranging from micrometer-scale 3D bulk to nanoscale two-dimensional (2D) one-dimensional (1D) and zero-dimensional (0D) structures, as well as their combined geometries. Based on binary collision approximation (BCA), ASIM describes both nuclear and electronic stopping, and simultaneous collision is also explicitly included. Importantly, ASIM offers a high degree of flexibility with a modular design, allowing variable displacement thresholds and customized irradiation conditions. The predictive accuracy is demonstrated by systematic comparisons with available experimental results and molecular dynamics calculations for irradiation of silicon, free-standing/supported graphene, carbon nanotube and fullerene with various incident ions, showing excellent agreement. We expect the developed framework to provide an attractive and versatile platform for interpreting, optimizing and designing irradiation-driven defect engineering strategies in emerging nanoscale materials and devices.

Helium (He) irradiation-induced degradation, particularly through the formation and growth of bubbles, poses a critical challenge for tungsten plasma-facing materials (W-PFMs). In service, these materials often experience tensile strain fields due to thermal cycling and bubble accumulation, yet the underlying physical mechanisms of external strain on the fundamental processes of He bubbles, from self-trapping, nucleation, growth, to long-term evolution, remain unclear. In this work, we systematically investigate these processes under isotropic tensile/compressive strain using multiscale simulations. We reveal that tensile strain induces a counterintuitive behavior: although the He solution energy is lowered under tensile strain, the He-He binding energy is significantly enhanced, a trend opposite to the positive correlation observed across different metals. This originates from the synergistic effect of strain and He, causing the variation in the electronic structure. Furthermore, tensile strain promotes bubble nucleation and growth by facilitating Frenkel pair nucleation and trap mutation. Extending these atomic insights to long-term evolution, we demonstrate how tensile strain accelerates bubble coalescence, rupture, and the early-stage formation of fuzz, consistent with experimental observations of strain-dependent bubble evolution. Crucially, a positive feedback mechanism integrating these findings is suggested: dense bubbles induce a subsurface tensile strain field in PFMs, which in turn further promotes bubble nucleation and growth. The proposed strain-enhanced mechanism informs the development of predictive models for He-induced damage in PFMs under fusion-relevant conditions.

The potential of tungsten (W)-based high-entropy alloys (HEAs) as plasma-facing materials (PFMs) for future fusion reactors has been evaluated through an investigation of fuzz growth. W-based HEAs (WTaCrV, WTaCrVTi) and pure W fabricated via mechanical alloying and high-pressure sintering have been exposed to helium (He) plasma at 1073 K with an ion energy of 60 eV. After He plasma exposure, all samples exhibit fuzz nanostructures. Detailed analyses of fuzz length, He bubble size, and elemental distribution within the fuzz nanostructures are conducted. The two W-based HEAs demonstrate distinct advantages, exhibiting fuzz lengths reduced by 37.1% (WTaCrV) and 34.3% (WTaCrVTi), alongside smaller He bubble sizes compared to pure W. However, no significant differences in fuzz length or He bubble size are observed between WTaCrV and WTaCrVTi. Significantly, scanning transmission electron microscopy-energy dispersive X-ray spectroscopy (STEM-EDS) analysis reveals that the fuzz nanostructures predominantly comprise W and Ta, attributed to the preferential nucleation and growth of He bubbles in regions enriched with these elements. These findings provide critical insights into the underlying mechanisms governing fuzz formation in W-based HEAs. The superior resistance to fuzz growth in these W-based HEAs guides the design and development of PFMs in future fusion reactors. (c) 2026 Published by Elsevier Ltd on behalf of The editorial office of Journal of Materials Science & Technology.

Irradiation creep can alter core geometry and potentially relax torque in screws or mechanical springs, thereby affecting overall system performance. A major contributor to the increased creep rate is enhanced dislocation multiplication and migration. Traditionally, irradiation-driven dislocation multiplication has been attributed either to irradiation-induced point defects that enable dislocation climb through the Bardeen–Herring mechanism, or to interactions between irradiation-induced dislocation loops that generate mobile dipoles. In this study, molecular dynamics simulations of body-centered cubic (BCC) tungsten reveal a distinct mechanism of cascade-enhanced Frank–Read dislocation multiplication. Specially, the coalescence of irradiation induced loops establishes an edge-dominated dislocation network that form the structural basis for multiplication, whereas cascade-induced transient stresses facilitate the glide and bowing of pinned edge dislocations within this network, thereby promoting dislocation multiplication and contributing to the increase in creep strain. Since these cascade-induced transient stresses occur only during irradiation, the present results suggest that cascade-enhanced dislocation multiplication may be one of the factors contributing to the higher creep rates observed under irradiation.

Temperature is a crucial factor affecting the behavior of hydrogen isotopes in plasma-facing materials. Our previous work demonstrated that the behavior of deuterium was significantly affected by temperature changes during plasma exposure. The present work investigated deuterium retention in pristine tungsten (W) and pre-damaged tungsten (Pre-W) exposed to deuterium plasma at two declining-temperature conditions: the high initial-temperature group of declining from 720 K to 420 K (HTD), and the low initial-temperature group of declining from 570 K to 370 K (LTD). Surface observations reveal severe blisters on both W and Pre-W after LTD exposure, and these blisters are substantially reduced under HTD exposure. It is indicated that HTD conditions are less favorable for deuterium accumulation than LTD conditions. The thermal desorption spectrum (TDS) reveals that HTD is beneficial for reducing deuterium retention while mitigating the pre-damage-induced enhancement of deuterium. Further analysis of desorption peaks indicates that the de-trapping energies associated with each type of trap are comparable in HTD- and LTD-W, but different in HTD- and LTD-Pre-W. The de-trapping energies of deuterium traps in HTD-Pre-W are lower than those in LTD-Pre-W. This difference likely arises because the higher initial temperature in HTD promotes greater decomposition of dislocation-type defects and enhances the level of deuterium occupation of monovacancies. Furthermore, the amount of deuterium captured by all types of traps is lower under HTD than under LTD, indicating that the high initial temperature of HTD impedes effective trapping and retention of deuterium. These results highlight the influence of different temperature ranges on the trapping and accumulation of deuterium in both W and Pre-W during the temperature-declining exposures.

Rhenium ( Re ) alloying is generally considered as an effective method to improve the performance of body-centered cubic (bcc) refractory metals, while the transmutation Re may adversely affect the thermomechanical property of bcc tungsten (W) under neutron irradiation. This highlights the importance of the Re introduction method in determining its effects on bcc metals, a factor that has yet to be fully clarified. In this study, we systematically investigate the co-evolution of Re and irradiation defects in W using the object kinetic Monte Carlo (OKMC) method, considering different Re introduction methods and transmutation rates. It is found that the extent of Re aggregation in neutron-irradiated pure W (with continuous Re introduction via nuclear transmutation) is significantly greater than in ion-irradiated W-Re alloys (where Re is introduced only at the initial stage in the solid solution state), even with identical Re concentrations. These differences align well with experimental observations and can be explained by the Re-to-defect ratio and the mobility of Re atoms, both of which strongly depend on the Re introduction method. Moreover, we quantify the volume fraction of irradiation defects and the average Re concentration in Re clusters within W-Re system across various transmutation rates, identifying critical conditions for Re's transition from beneficial to detrimental. Our findings provide valuable insights for assessing Re effects in neutron and ion irradiated W-Re systems and support the application of bcc refractory metals in nuclear environments. (c) 2025 Published by Elsevier Ltd on behalf of The editorial office of Journal of Materials Science & Technology.

Tungsten has demonstrated a competitive figure of merit in its application to plasma-facing components (PFCs) of fusion reactors. During service, the material is exposed to high temperatures and high-level displacement damage. A common interest is fostered in the nuclear materials community to address the issue of defect evolution at operating temperatures, and how they recover throughout service. During maintenance, the application of in situ thermal repair technologies is tempting, featuring attractive efficiency in defect removal via an optimal selection of post-irradiation annealing (PIA) parameters. In previous studies, we examined the role of PIA temperature, PIA duration, and initial defect concentration on defect evolution, and redefined the damage recovery stages for tungsten, but this was done from a room-temperature heavy-ion irradiation perspective; see Wang et al (2023 J. Nucl. Mater.581 154454), Wang et al (2024 Acta Mater.273 119942). In this study, the scope is expanded to displacement damage saturation induced by heavy-ions at high temperatures, relevant to the service conditions of tungsten-based PFCs. The damage microstructure evolution in response to varied irradiation temperatures (T-Irr) and PIA temperatures (T-PIA) was assessed via transmission electron microscopy and Doppler broadening positron annihilation spectroscopy. Irradiation hardening was evaluated via nano-indentation. A scientific framework is proposed to guide thermal healing of displacement damage in tungsten via PIA treatment. It was ineffective when T-PIA <= T-Irr. An adverse effect of PIA-induced secondary hardening occurred when T-PIA (stage III) > T-Irr (stage III). The optimal PIA scheme was confirmed when T-PIA (stage IV) > T-Irr (stages III-IV), eluding PIA-induced secondary hardening and minimizing PIA-enhanced recrystallization.

Temperature is an important factor affecting the behavior of deuterium in plasma-facing material. This paper focuses on the effect of variable temperature on deuterium behavior in tungsten. Deuterium plasma exposures were conducted for the pristine and pre-damaged tungsten at two variable temperatures: one is rising from 323 K to 723 K (named TR exposure), and the other is declining from 723 K to 323 K (named TD exposure). Statistical results of surface blisters show that the TD exposure favors the nucleation of small intergranular and intragranular blisters, while the TR exposure favors the formation of large intergranular and intragranular blisters. Thermal desorption spectra results demonstrate that total deuterium retention is reduced, and the defects could evolve into higher-energy deuterium traps in the TR exposure compared to the TD exposure. Comparing the pristine and damaged tungsten revealed that the pre-damage suppressed the formation of intragranular blisters, accelerated the intergranular blisters, and enhanced deuterium retention in both the TR and TD exposures. Nevertheless, the aggravating effect of pre-damage is more pronounced in the TR exposure, while the suppressive effect on intragranular blisters is alleviated in the TD exposure. Furthermore, the evolution result of pre-damage defects in the TR exposure is more pronounced than that of the pristine one. The above results reveal significant differences in the deuterium behavior and defect evolution in tungsten under the TR and TD exposures, demonstrating the inconstant temperature effect on deuterium behavior.

Tungsten (W), as the most promising candidate for plasma-facing materials, will experience significant irradiation hardening in nuclear fusion environment, which is originated from the formation of displacement damages, such as voids and dislocation loops. Hydrogen (H) can further exacerbate the hardening effect, but the underlying physical mechanisms remain unclear. Using molecular dynamics simulations, we investigate the impact of H aggregation within voids and 1/2<111> dislocation loops on obstructing the glide of 1/2<111> edge dislocations. On the one hand, the pinning effect of H-void complexes is closely related to the ratio of H to vacancy (H:Vac). When the H:Vac ratio is high, H atoms will overflow from the H-void complexes along the dislocation, enhancing the attractive interaction of complexes with dislocation and thereby causing a significant increase in the critical resolved shear stress (CRSS). On the other hand, the accumulation of H around dislocation loops can increase the CRSS by an order of magnitude. This is mainly because the binding of H to the dislocation loop hinders its movement along with the edge dislocation. Our findings advocate that the presence of interstitial impurities can dramatically modify the mechanical properties of materials under irradiation, and provide an important reference for the prediction of W performance and the development of advanced nuclear materials.

Sink strength, as a fundamental parameter in mean-field approaches, describes the ability of sinks (e.g., dislocation lines) to capture migrating defects and is crucial for simulating the microstructure evolution of irradiation damage in nuclear materials. Here, taking body centered cubic tungsten (W) as an example, we systematically investigate the sink strength of dislocation lines using the object Kinetic Monte Carlo (OKMC) method. It is found that there are noteworthy discrepancies of sink strength between the traditional theoretical expression and OKMC simulations. This should be attributed to two factors, namely temperature and probability density distribution. The former can be derived from a master curve that has already been proposed for 1D to 3D diffusion–reaction kinetics, while the latter can be well described by a modified analytical expression of sink strength for dislocation lines. By incorporating these factors, the discrepancy between theoretical results and OKMC simulations is eliminated. Notably, the results of defect evolution in irradiated W, obtained using the modified sink strength expression, exhibit a greater consistency with experimental observations than those derived from the conventional model. These results provide a better insight into the sink strength model, and have broad implications for understanding and reproducing the microstructure evolution of irradiation defects in materials.

Pure steps such as Σ3{112} incoherent twin boundary (ITB) steps often present along the coherent twin boundary regardless of deformation or annealing twins in metals with a face-centered cubic (FCC) structure. Corresponding to the defect structure of Σ3{112} ITB step, this pure step is composed of multiple Shockley partial dislocations with a net zero Burgers vector. It is generally believed that such pure steps do not move under mechanical loading because of the zero Peach–Koehler force. Here, we demonstrate the migration of pure steps by atomistic simulations under uniaxial tension and reveal the driving force associated with elastic anisotropy, i.e., the driving force originates from an imbalance in the elastic strain energy density across the TBs. Therefore, we introduce a compliance anisotropy factor, calculated by density functional theory, to characterize the migration capability of the Σ3{112} ITB step in FCC metals.

This study reveals that thermal fatigue loading (transient thermal shock), similar to that in fusion environments, can serve as a surface processing technique for BCC metals. Regions with a {110} grain orientation can be selectively achieved in varying sizes and locations on the sample surface. Furthermore, our experiments confirm that the specific localized orientation transformation obtained through this method exhibits certain high-temperature stability at 1573 K (above the recrystallization temperature of tungsten). The experiment employed a 0.25 GW/m(2) high-energy pulsed electron beam for 1 ms to cyclically load the tungsten surface, simulating edge localized mode events in fusion conditions. It was found that tungsten exhibited significant surface grain orientation transformation (distinct {110} grain orientation) under low strain (similar to 1 %) after transient thermal shocks, a phenomenon rarely mentioned in studies of thermal shock on fusion reactor divertor materials. Microstructure characterization results suggest that this localized orientation transformation, induced by minor surface damage, primarily results from the generation, movement, and evolution of dislocations into subgrain and low-angle grain boundaries. The cyclic accumulation of the migration of kink-like subgrain/low-angle grain boundaries under transient thermal stress at high temperatures drives this process. Subsequently, crystal plasticity finite element method simulations based on dislocation slip were conducted to study the surface grain orientation transformation of tungsten under compressive thermal stress. This predictive capability provides valuable guidance for understanding the service conditions of fusion reactor divertor materials. Furthermore, we propose that cyclic transient thermal shocks can serve as an effective surface processing technique for metals, enabling the formation of specific localized grain orientations.

Understanding the behavior of tungsten (W) surface damage under the synergistic effects of high heat flux (HHF) 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 like HHF He neutral beam pulse irradiation on polycrystalline W samples at the test facility Garching LArge DIvertor Sample. 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) (Zhou et al 2010 Nucl. Fusion 50 115010; Becquart and Domain 2006 Phys. Rev. Lett. 97 1-4; Shu et al 2013 Nucl. Instrum. Methods Phys. Res. B 303 84-6; Fu et al 2021 J. Nucl. Mater. 543 152599). 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.