Small Defect Clusters Research Articles

On the capabilities of k-ART over MD for the study of the kinetics of small point defect clusters in α-Fe

Molecular Dynamics simulations, while contributing to the understanding of the mechanisms of diffusion and interactions of point defects and their clusters, are inherently limited in their temporal scope (few nanoseconds). This constraint becomes particularly evident when studying the dynamics of vacancies at low temperatures, where their jump frequency is exceedingly low, posing challenges for accurate reproduction. Additionally, the size of the simulation box imposes constraints, influencing the representation of the system and potentially affecting the accuracy of results. A relatively new kinetic activation-relaxation technique (k-ART) efficiently resolves the limitations of MD simulations, such as computation time and system dimensionality, without the need for a priori knowledge of the simulated system. This technique enables simulations lasting up to several seconds and encompassing systems with higher dimensions. In this paper we check the validity of k-ART to reproduce accurately the migration mechanisms and energies of point defects and small clusters, previously obtained by MD and validated experimentally. We point out the advantages and difficulties of using AKMC with k-ART.

Influence of interstitial cluster families on post-synthesis defect manipulation and purification of oxides using submerged surfaces.

Injection of interstitial atoms by specially prepared surfaces submerged in liquid water near room temperature offers an attractive approach for post-synthesis defect manipulation and isotopic purification in device structures. However, this approach can be limited by trapping reactions that form small defect clusters. The compositions and dissociation barriers of such clusters remain mostly unknown. This communication seeks to address this gap by measuring the dissociation energies of oxygen interstitial traps in rutile TiO2 and wurtzite ZnO exposed to liquid water. Isotopic self-diffusion measurements using 18O, combined with progressive annealing protocols, suggest the traps are small interstitial clusters with dissociation energies ranging from 1.3 to 1.9eV. These clusters may comprise a family incorporating various numbers, compositions, and configurations of O and H atoms; however, in TiO2, native interstitial clusters left over from initial synthesis may also play a role. Families of small clusters are probably common in semiconducting oxides and have several consequences for post-synthesis defect manipulation and purification of semiconductors using submerged surfaces.

Machine learning potential assisted exploration of complex defect potential energy surfaces

Atomic-scale defects generated in materials under both equilibrium and irradiation conditions can significantly impact their physical and mechanical properties. Unraveling the energetically most favorable ground-state configurations of these defects is an important step towards the fundamental understanding of their influence on the performance of materials ranging from photovoltaics to advanced nuclear fuels. Here, using fluorite-structured thorium dioxide (ThO2) as an exemplar, we demonstrate how density functional theory and machine learning interatomic potential can be synergistically combined into a powerful tool that enables exhaustive exploration of the large configuration spaces of small point defect clusters. Our study leads to several unexpected discoveries, including defect polymorphism and ground-state structures that defy our physical intuitions. Possible physical origins of these unexpected findings are elucidated using a local cluster expansion model developed in this work.

An empirical potential for simulating hydrogen isotope retention in highly irradiated tungsten

We describe the parameterization of a tungsten-hydrogen empirical potential designed for use with large-scale molecular dynamics simulations of highly irradiated tungsten containing hydrogen isotope atoms, and report test results. Particular attention has been paid to getting good elastic properties, including the relaxation volumes of small defect clusters, and to the interaction energy between hydrogen isotopes and typical irradiation-induced defects in tungsten. We conclude that the energy ordering of defects changes with the ratio of H atoms to point defects, indicating that this potential is suitable for exploring mechanisms of trap mutation, including vacancy loop to plate-like void transformations.

Defect annealing in heavy-ion irradiated tungsten: Long-time thermal evolution of saturated displacement damage at different temperatures

Understanding the time-dependence of thermal evolution and recovery of displacement damage is of great significance for evaluations of material performance in harsh fusion environments. In this study, the long-time thermal evolution of irradiation-induced defects has been investigated in heavy-ion irradiated tungsten at doses above the saturation limit. Polycrystalline tungsten samples were irradiated with 6 MeV copper ions up to 0.6 dpa at room temperature, followed by a series of isothermal annealing experiments at 653 and 1123 K for 5000, 10000 and 20000 s, respectively. Annealing at both temperatures triggered the coevolution of interstitial and vacancy defects, causing changes in defect morphology, size and population statistics. The damage microstructure remained stable after 5000 s at 653 K. Subtle variations in defect density and size were found, indicating the quasi-thermal-equilibrium state of the damage microstructure. In this temperature regime, small defect clusters were depleted via annihilation and absorption, while large defect clusters and defect-impurity complexes were pinned by traps requiring further thermal activation. Pronounced time-dependence in defect evolution was found at 1123 K. Defect recovery and defect coarsening became more prominent with increasing duration. Almost all dislocation networks dissociated into individual dislocation lines after 20000 s, while nano-sized voids and loops became fewer but gained an increase in size. These findings can be attributed to the collective impact of the increasing mobility of small vacancy clusters, the continual dissociation of large vacancy clusters and small voids, and the thermally activated detrapping of defect clusters from impurities and/or other traps.

Atomic-scale study of He ion irradiation-induced clustering in α-Zirconium

Zircaloy-4 alloy specimens consisting of α-Zr grains were irradiated on a tandem accelerator with 5 MeV He ions to a fluence of 5 × 1021 ions m − 2 at different temperatures. Defect clusters and He bubbles were observed in the TEM micrographs of the irradiated microstructure. The small defect clusters and He bubbles were found to exhibit spherical shapes whereas the large defect clusters showed plate-like shapes. Molecular dynamics simulations and crystallography analyses revealed that the defect clusters are mainly composed of Zr interstitials. During He-ion irradiation, the clustering of Zr dumbbell interstitials results in the formation of small and spherical defect clusters. With further irradiation, the Zr interstitials prefer to occupy the octahedral sites that align along the {12¯14} planes with other site interstitials, forming cluster sections. These initial planar clusters then grow by extending along the <101¯0> directions and by forming more cluster sections along the <12¯16> directions, leading to large plate-like defect clusters on the {1¯21¯1}planes. He bubbles nucleate by the clustering of He atoms and vacancies that are formed by kicking out the lattice atoms, and then grow by the migration and coalescence of helium-vacancy clusters. Small He atoms have many interstitial sites to occupy in the α-Zr HCP crystal, allowing He bubbles to grow to spherical shapes. In addition, grain boundaries induce the accumulation of primary point defects and He atoms and high irradiation temperature accelerates diffusion, and hence larger defect clusters and He bubbles form in the grain boundary regions and at higher irradiation temperature.

Lattice damage in InGaN induced by swift heavy ion irradiation

The microstructural responses of In0.32Ga0.68N and In0.9Ga0.1N films to 2.25 GeV Xe ion irradiation have been investigated using x-ray diffraction, Raman scattering, ion channeling and transmission electron microscopy. It was found that the In-rich In0.9Ga0.1N is more susceptible to irradiation than the Ga-rich In0.32Ga0.68N. Xe ion irradiation with a fluence of 7 × 1011 ions⋅cm−2 leads to little damage in In0.32Ga0.68N but an obvious lattice expansion in In0.9Ga0.1N. The level of lattice disorder in In0.9Ga0.1N increases after irradiation, due to the huge electronic energy deposition of the incident Xe ions. However, no Xe ion tracks were observed to be formed, which is attributed to the very high velocity of 2.25 GeV Xe ions. Point defects and/or small defect clusters are probably the dominant defect type in Xe-irradiated In0.9Ga0.1N.

Atomic modeling assessment of the interaction distance and effective bias for small defect clusters absorption at a void in BCC Fe

Cavity swelling plays a vital role in the microstructural evolution of ferritic-martensitic alloys under irradiation. Recent research reported that the classic thermal criterion of the critical size in cavity growth loses its explanatory power outside the intermediate temperature regimes. In contrast, the newly proposed bias-driven criterion performs nearly equivalently at the entire temperature range by introducing the concept of cavity bias. Molecular statics calculations were performed to determine the interaction energy landscape of single self-interstitial atom (SIA), single vacancy, di-SIA, di-vacancy, 7-SIA cluster, and 7-vacancy cluster interaction with a 59-vacancy void in BCC Fe. Additionally, molecular dynamics simulations were conducted to investigate the rotation and migration behavior of varying size SIA clusters in close proximity to the void. A homogenization method was established to describe the capture volume and mimic the one-dimensional diffusion behavior for large SIA clusters. The interaction distance between the void and defects was determined, and the effective bias was calculated as a function of the void size at 25 ℃, 300 ℃, and 500 ℃. The results provide important atomistic inputs for mesoscale modeling of defect evolution and microstructural changes, such as cluster dynamics simulation.

Insight into defect cluster annihilation at grain boundaries in an irradiated nanocrystalline iron

The design of radiation-tolerant polycrystalline materials has been mainly based on the control and manipulation of grain boundaries (GBs) that leads to annihilation of point defects at grain-boundary neighborhoods thus resulting in the formation of defect denuded zones. Nanocrystalline materials are potential candidates providing large density defect sinks for individual point defects and small highly mobile defect clusters (DCs). Using the in-situ irradiation transmission electron microscopy (TEM) technique, this study not only experimentally revealed the coalescence of small glissile DCs at/near GB and their subsequent annihilation at grain-boundary, but also provide insight into defect cluster dynamics. The small DCs were found to be transported to grain-boundary neighborhoods where they can annihilate by the one-dimensional loop hop or Burgers-vector rotation mechanism to GBs, considered as the main contribution to the long-range flux of interstitials to GB sinks. This process had marked effects on the morphology of the irradiated microstructure in nanocrystalline iron, limiting the length of DC strings and reducing the coalescence of DCs into large clusters (dislocation loops).

Post irradiated microstructure and mechanical properties of pure V

Proton irradiation has been carried out on pure vanadium samples and the changes in the microstructure and mechanical properties have been evaluated as a function of dose. The microstructural parameters such as domain size and microstrain have been assessed along different crystallographic planes using the X-ray diffraction line profile analysis methods such as simplified breadth and the modified Rietveld technique. No significant change in the dislocation density is observed as a function of dose from the XRD results. However, the results of positron annihilation spectroscopy in terms of positron lifetime and S-parameter clearly indicated the formation of small defect clusters with irradiation. Transmission electron microscopy investigation confirmed the presence of small dislocation loops in the irradiated samples. Molecular dynamics simulations carried out to model the formation of irradiation-induced defects clusters with irradiation dose support the experimental observation of small defect clusters and increasing cluster size with dose. The yield strength and tensile strength showed systematic increase with irradiation dose corroborating with the results of microhardness measurement. The fractography of the tensile samples clearly showed the characteristics of brittle fracture in the irradiated samples.

Characterization of in-situ ion irradiated Fe-21Cr-32Ni austenitic model alloy and alloy 800H at low doses

The microstructural evolution of ternary Fe-21Cr-32Ni (21Cr32Ni) model alloy and of alloy 800H was investigated with a series of in-situ ion irradiation experiments performed using the Intermediate Voltage Electron Microscope (IVEM)-Tandem Facility at Argonne National Laboratory (ANL). Samples were irradiated in-situ with 1 MeV Kr++ ions in the temperature range of 50K to 713K to doses up to 2 dpa. The size distribution of defect clusters, average defect cluster diameter, and defect cluster density were measured and compared. Results showed that the evolution of defects (i.e. the average defect cluster size and number density) in 21Cr32Ni model alloy and alloy 800H with dose were similar at irradiation temperatures up to 300K where they initially increased with dose up to 0.1 dpa after which no significant changes in defect size and density were observed with further irradiation. In addition, both alloys exhibited ordered defect structures along the <100> direction at relatively low temperatures, up to 300K, which remained stable throughout post-irradiation in-situ thermal annealing up to a temperature of 773K. During irradiation at 713K, small defect clusters were observed at low doses (<0.1 dpa) in both alloys. However, at this irradiation temperature, the clusters in 21Cr32Ni grew with a faster rate than those formed in alloy 800H, causing the microstructure in the former to be dominated by numerous large dislocation loops having both {111}- and {110}-type habit planes, and in the latter to be dominated by small defect clusters and small {111}-type dislocation loops. This may indicate that defect trapping by the solute atoms in alloy 800H at 713K can slow point defect migration to defect clusters and limit their growth.

Fast recovery of ion-irradiation-induced defects in Ge2Sb2Te5 thin films at room temperature

Phase-change materials serve a broad field of applications ranging from non-volatile electronic memory to optical data storage by providing reversible, repeatable, and rapid switching between amorphous and crystalline states accompanied by large changes in the electrical and optical properties. Here, we demonstrate how ion irradiation can be used to tailor disorder in initially crystalline Ge2Sb2Te5 (GST) thin films via the intentional creation of lattice defects. We found that continuous Ar+-ion irradiation at room temperature of GST films causes complete amorphization of GST when exceeding 0.6 (for rock-salt GST) and 3 (for hexagonal GST) displacements per atom (ndpa). While the transition from rock-salt to amorphous GST is caused by progressive amorphization via the accumulation of lattice defects, several transitions occur in hexagonal GST upon ion irradiation. In hexagonal GST, the creation of point defects and small defect clusters leads to the disordering of intrinsic vacancy layers (van der Waals gaps) that drives the electronic metal–insulator transition. Increasing disorder then induces a structural transition from hexagonal to rock-salt and then leads to amorphization. Furthermore, we observed different annealing behavior of defects for rock-salt and hexagonal GST. The higher amorphization threshold in hexagonal GST compared to rock-salt GST is caused by an increased defect-annealing rate, i.e., a higher resistance against ion-beam-induced disorder. Moreover, we observed that the recovery of defects in GST is on the time scale of seconds or less at room temperature.

Defect accumulation and evolution in refractory multi-principal element alloys

Refractory multiple principal elemental alloys (MPEAs) hold great promise for structural materials in future nuclear energy systems. Compared to the extensively studied face-centered cubic (FCC) MPEAs, the irradiation resistance of body-centered cubic (BCC) refractory MPEAs is relatively less known. In this work, we study defect accumulation and evolution in two BCC VTaTi and VTaW MPEAs comparatively through atomistic simulations. For this purpose, we have parameterized the Embedded Atom Method (EAM) potential parameters for V metals. Combined with available potential parameters for other elements, we construct average atom models for the considered alloys to elucidate the effects of chemical complexities in BCC MPEAs. Our results based on Frenkel pair accumulation simulations suggest that the major influence of chemical fluctuations in BCC MPEAs is on the clustering behavior of defects, which leads to discrete point defects or small defect clusters, in contrast to the large defect clusters observed in the average atom model. We further show that the diffusion of interstitial clusters exhibits different modes due to chemical complexity. While interstitials show either three-dimensional or one-dimensional diffusion in the average atom model, the mean free path of interstitials in the random alloy is strongly suppressed. These results provide fundamental insights into the irradiation response of BCC MPEAs and pinpoint the critical role of chemical complexity on defect evolution, which lay the basis for the future development of irradiation-resistant structural materials based on BCC MPEAs.

In situ TEM investigations of the microstructural changes and radiation tolerance in SiC nanowhiskers irradiated with He ions at high temperatures

Using in-situ transmission electron microscopy (TEM) with ion irradiation, we investigated the microstructural changes in silicon carbide nanowhiskers (SiC NWs) which were used as a model system for nanoporous SiC. Irradiations were carried out using 6 keV He ions at temperatures between 500 and 1000°C and doses up to 20 dpa. These results are compared with the irradiation effects in SiC thin foils under the same conditions to establish differences in their response to radiation damage. The irradiation temperature played a significant role in the evolution of different microstructures; at 500°C, small defect clusters were observed in the NWs together with a segregation of carbon at the surface of the NWs mapped using energy-filtered TEM (EFTEM). At 800°C, small He bubbles (2–4 nm in diameter) were observed in the NW matrix while He platelets and bubble discs formed in the foils. At 1000°C, several changes were observed in the NWs including bubbles at twin boundaries, voids and oxygen-rich precipitates. The large surface area to volume ratio enhances defect recombination supressing the defect density in the SiC NWs compared to the foils indicating high radiation tolerance; however, elemental segregation and precipitation may limit its application in advanced nuclear reactors.

Role of rare events in the pinning problem

The authors show that rare events in the form of small defect clusters dominate the pinning of the vortex lattice in a superconductor in the intermediate regime between weak collective and strong pinning.

Recovery behavior of neutron-irradiated aluminum nitride with and without containing interstitial dislocation loops

Recovery behavior of the same AlN but neutron-irradiated with different conditions was studied by using XRD and dilatometry. Length recovery rates at each isothermal annealing step were analyzed by first to second-order kinetics. The specimen irradiated at lower neutron fluence at low irradiation temperature responded isotopic unit-cell expansion, which indicated random dispersion of point defects of vacancies and/or very small defect clusters. From irradiation temperature up to 523 K, the results of Arrhenius’ plot applied for second-order kinetics gives an activation energy value of 2.9 eV. An intermediate region, 573 to 1023 K, was represented by a slightly lower activation energy of 2.0 eV. For the first lower temperature range, migration of VN may be the mechanism and that of VAl for the second stage. At high-temperature region, 1073 to 1423 K, represents final constant recovery before the saturation of length recovery. In this region, the high activation energy of 5.8 eV was observed. It is supposed that dissociation of the vacancy-antisite cluster, interstitial nitrogen cluster, or formation of a complex containing vacancy and antisite can be candidate mechanisms. The specimen containing interstitial loops after the medium fluence at higher temperatures did not recover up to 723 K, indicating almost free of independent interstitials. Between 773 and 1123 K, recovery required a similar amount of energy to the high-temperature region of the lower fluence specimen (5.8 eV), and then the same mechanism could be applied. In the second region, 1173 to 1423 K, with a very high activation energy of ~18 eV, remarkable large and quick shrinkage was observed. Large volume shrinkage may be the result of diminishing anisotropic lattice swelling, i.e., unit-cell volume expansion by internal strain is relieved.

In situ microstructural evolution in face-centered and body-centered cubic complex concentrated solid-solution alloys under heavy ion irradiation

This study characterizes the microstructural evolution of single-phase complex concentrated solid-solution alloy (CSA) compositions under heavy ion irradiation with the goal of evaluating mechanisms for CSA radiation tolerance in advanced fission systems. Three such alloys, Cr18Fe27Mn27Ni28, Cr15Fe35Mn15Ni35, and equimolar NbTaTiV, along with reference materials (pure Ni and E90 for the CrFeMnNi family and pure V for NbTaTiV) were irradiated at 50 K and 773 K with 1 MeV Kr++ ions to various levels of displacements per atom (dpa) using in-situ transmission electron microscopy. Cryogenic irradiation resulted in small defect clusters and faulted dislocation loops as large as 12 nm in face-centered cubic (FCC) CSAs. With thermal diffusion suppressed at cryogenic temperatures, defect densities were lower in all CSAs than in their less compositionally complex reference materials indicating that point defect production is reduced during the displacement cascade stage. High temperature irradiation of the two FCC CSA resulted in the formation of interstitial dislocation loops which by 2 dpa grew to an average size of 27 nm in Cr18Fe27Mn27Ni28 and 10 nm in Cr15Fe35Mn15Ni35. This difference in loop growth kinetics was attributed to the difference in Mn-content due to its effect on the nucleation rate by increasing vacancy mobility or reducing the stacking-fault energy.

Understanding hydrogen retention in damaged tungsten using experimentally-guided models of complex multispecies evolution

Fuel retention in plasma facing tungsten components is a critical phenomenon affecting the mechanical integrity and radiological safety of fusion reactors. It is known that hydrogen can become trapped in small defect clusters, internal surfaces, dislocations, and/or impurities, and so it is common practice to seed W subsurfaces with irradiation defects in an attempt to precondition the system to absorb hydrogen. The amount of H can later be tallied by performing careful thermal desorption tests where released temperature peaks are mapped to specific binding energies of hydrogen to defect clusters and/or microstructural features of the material. While this provides useful information about the potential trapping processes, modeling can play an important role in elucidating the detailed microscopic mechanisms that lead to hydrogen retention in damaged tungsten. In this paper, we develop a detailed kinetic model of hydrogen penetration and trapping inspired by recent experiments combining ion irradiation, hydrogen plasma exposure, and thermal desorption. We use the stochastic cluster dynamics method to solve the system of coupled partial differential equations representing the mean field description of the multispecies system. The model resolves the spatial distribution of defects and hydrogen clusters during the three processes carried out experimentally and is parameterized with information from atomistic calculations. We find that the calculated thermal desorption spectra are broadly characterized by three H emission regions: (i) a low temperature one where dislocations are the main contributors to the release peaks; (ii) an intermediate one governed by hydrogen release from small overpressurized clusters with multiple overlapping peaks, and (iii) a high temperature one defined by clean isolated emission peaks from large underpressurized bubbles. These three temperature intervals are seen to largely correlate with the depth at which the clusters are found. The relevance of the ‘super abundant’ vacancy mechanism is assessed, finding that its main role is to transfer more clusters from the intermediate to the high temperature regions as its relevance increases. We find this picture to be in very good agreement with the experiments, adding confidence to the predictive potential of the models and their useto understand irradiation damage and plasma exposure effects in plasma facing components.

The role of Ti and TiC nanoprecipitates in radiation resistant austenitic steel: A nanoscale study

This work encompasses an in-depth transmission electron microscopy and atom probe tomography study of Ti-stabilized austenitic steel irradiated with Fe-ions. The focus is on radiation induced segregation and precipitation, and in particular on how Ti and TiC affect these processes. A 15-15Ti steel (grade: DIN 1.4970) in two thermo-mechanical states (cold-worked and aged) was irradiated at different temperatures up to a dose of 40 dpa. At low irradiation temperatures, the cold-worked and aged materials evolved to a similar microstructure dominated by small Si and Ni clusters, corresponding to segregation to small point defect clusters. TiC precipitates, initially present in the aged material, were found to be unstable under these irradiation conditions. Elevated irradiation temperatures resulted in the nucleation of nanometer sized Cr enriched TiC precipitates surrounded by Si and Ni enriched shells. In addition, nanometer sized Ti- and Mn-enriched G-phase (M6Ni16Si7) precipitates formed, often attached to TiC precipitates. Post irradiation, larger number densities of TiC were observed in the cold-worked material compared to the aged material. This was correlated with a lower volume fraction of G-phase. The findings suggest that at elevated irradiation temperatures, the precipitate-matrix interface is an important point defect sink and contributes to the improved radiation resistance of this material. The study is a first of its kind on stabilized steel and demonstrates the significance of the small Ti addition to the evolution of the microstructure under irradiation.

In Situ Microstructural Evolution in Fcc and Bcc Complex Concentrated Solid-Solution Alloys Under Heavy Ion Irradiation

This study characterizes the microstructural evolution of single-phase complex concentrated solid-solution alloy (CSA) compositions under heavy ion irradiation with the goal of evaluating mechanisms for CSA radiation tolerance in advanced fission systems. Three such alloys, Cr18Fe27Mn27Ni28, Cr15Fe35Mn15Ni35, and equimolar NbTaTiV, along with reference materials (pure Ni and E90 for the CrFeMnNi family and pure V for NbTaTiV) were irradiated at 50 K and 773 K with 1 MeV Kr++ ions to various levels of dpa using in-situ TEM. Cryogenic irradiation resulted in small defect clusters and faulted dislocation loops as large as 12 nm in FCC CSAs. With thermal diffusion suppressed at cryogenic temperatures, defect densities were lower in all CSAs than in their less compositionally complex reference materials indicating that point defect accumulation is reduced during the displacement cascade stage. High temperature irradiation of the two FCC CSA resulted in the formation of interstitial dislocation loops which by 2 dpa grew to an average size of 27 nm in Cr18Fe27Mn27Ni28 and 10 nm in Cr15Fe35Mn15Ni35. This difference in loop growth kinetics was attributed to the difference in Mn-content due to its effect on the nucleation rate by increasing vacancy mobility or reducing the stacking-fault energy.