Interstitial Dislocation Loops Research Articles

Revealing the critical role of vanadium in radiation damage of tungsten-based alloys

Refractory tungsten-based medium- and high-entropy alloys form a promising class of strong, heat- and radiation-resistant materials for nuclear applications. Here, we systematically explore the radiation tolerance of Mo-Nb-Ta-V-W alloys using molecular dynamics simulations and a machine-learned interatomic potential. Going from pure W to W-based equiatomic binaries, ternaries, quaternary, and the quinary high-entropy alloy, we simulate the radiation damage accumulation up to 0.4–0.8 dpa through cumulative collision cascades. We find that the radiation damage and evolution are controlled by a combination of cascade-induced clustering, the balance in migration of vacancies and interstitials, and defect cluster binding energies. These mechanisms are strongly influenced by atom size, among which vanadium as the smallest atom is decisive. In pure W and alloys without V, the microstructure is characterised by large and growing interstitial dislocation loops. In stark contrast, in alloys containing V, vacancy dislocation loops are formed directly in collision cascades and the balanced migration rate of vacancies and interstitials promotes recombination and prevents growth of interstitial loops. The atom size differences also lead to clear segregation, with small atoms (V) in interstitial clusters and larger atoms (Ta, Nb) segregating around vacancy clusters. Furthermore, the small cluster sizes, formation of vacancy loops, segregation, and balanced migration in V-containing alloys lead to low swelling and an exceptionally efficient defect annealing compared to pure W and V-free alloys. Our results highlight WTaV as a promising low-activation material, where almost 90% of defects are annealed at 2000 K during 5 ns compared to <20% in pure W.

The temperature effect on the formation and growth of helium-vacancy clusters in bcc iron: A comprehensive molecular dynamics simulation

Bubble generation can significantly degrade the performance of fusion reactor structural materials. In this work, the temperature effects on the initial clustering of helium (He) atoms, the formation of helium-vacancy clusters (HenVm) and the subsequent growth of the clusters into bubbles in bcc iron under different conditions were simulated by molecular dynamics. The results show that He behaviors are primarily dominated by coalescence at temperature range from 400 K to 500 K, which leads to a preference for He to bind together. Conversely, when temperature is higher than 500 K, dissociation become more prominent, causing large He clusters to be hard to form. Under the conditions of increasing He concentration and irradiation dose, the higher the temperature, the higher the density of HenVm clusters at He concentration around 100 appm. However, the temperature dependence reversed at high He concentration. The detailed analysis of formation cases revealed the percentage of nucleation sites for He bubbles at defects during He segregation. In addition, there is no significant difference among the He/V ratio of HenVm clusters formed at vacancies, interstitial dislocation loops (IDLs) and interstitial sites during the cluster growth. However, the He/V ratio of He bubbles growing on edge dislocation (ED) is overall lower than those growing at other formation sites, since the growth of HenVm clusters on ED requires a lower pressure to kick out self-interstitial atoms (SIAs) or SIA clusters. The average He/V ratio of He bubbles in bulk decreases as temperature rises from 400 K to 800 K, while that of He bubbles growing on ED is independent of temperature.

Direct formation of nano-scale dislocation loops during displacement cascade in BCC Fe studied by molecular dynamics simulation

In this work, the displacement cascades in bcc iron were simulated by molecular dynamics (MD). The formation of dislocation loops was found in cascade simulations when the energy of PKA (EPKA) was above 20 keV. For the first time, we reported the formation of <100> interstitial dislocation loops in cascade simulation of bcc iron under low EPKA, about 40 keV, much lower than that in Ref. 10. On the other hand, our results suggest that <100> loops all transformed into 1/2<111> loops within 200 ps. This will help us build a microscopic image and mechanism about the formation of interstitial dislocation loop.

In-situ TEM investigation of zirconium alloy under Kr+ single-beam and Kr+-He+ dual-beam synergetic irradiation

The in-situ TEM irradiation experiments of zirconium alloy were conducted at 573 K, 673 K, and 773 K utilizing a 400 keV Kr + single beam and a 400 keV Kr+ and 30 keV He + dual beam. The results show that a large number of dislocation loops have been characterized in the matrix of the zirconium alloy under irradiation. With increasing the irradiation damage dose, some dislocation loops have reacted with one another to form a larger dislocation loop, which has finally formed dislocation lines or other defect structures. In zirconium alloys irradiated with Kr + single beam and Kr+ and He + dual-beam radiation, the proportion of <a> type dislocation loops with different Burgers vectors is essentially the same at low damage doses, but the proportion of interstitial type dislocation loops with the same Burgers vectors is obviously different. The amorphization of the second phase and the dissolution of the small-sized second phase were also pointed out. With the increase in temperature, the density of the dislocation loop in zirconium alloy gradually decreases, and the size of dislocation loop first increases and then decreases. Kr+ and He + dual beam irradiation increases the size of dislocation loops but decreases their density as compared with Kr + single beam irradiation.

The effect of migrating Σ3 {112} incoherent twin boundary on radiation tolerance of nanotwinned copper

Nanotwinned face-centered cubic metals possess superior radiation tolerance; however, the underlying mechanism of alleviating radiation damage is still under debate. In the present study, the effect of Σ3{112} incoherent twin boundary (ITB) on the radiation tolerance of nanotwinned copper was systematically investigated by the molecular dynamic simulation method. The simulation results reveal that both stacking fault tetrahedron and interstitial dislocation loop can be easily absorbed into the migrating ITB. The two types of defects occupy two different types of atom layers of the ITB and move in the opposite direction of each other, resulting in their annihilation, thus enhancing the radiation tolerance of nanotwinned copper. Another interesting finding is that the migration of ITB can be accelerated rather than retarded by both interstitials and vacancies. The above simulation results were further investigated in terms of defect formation and migration energies. The present work provides a deep insight into the radiation tolerance mechanisms of nanotwinned metals.

Planar fault transformation and unfaulting of interstitial dislocation loops in irradiated L12-Ni3Al

The present study has systematically investigated unfaulting mechanism of single-layer interstitial dislocation Loops (IDLs) in irradiated L12-Ni3Al to extend the existing unfaulting theory from pristine FCC lattice to the L12 ordered lattice, employing atomistic simulations combining continuum theory. Two types of variants for the IDLs have been determined with different planar faults enclosed in the Frank loop. The unfaulting routes of IDLs by Shockley partial loops have been unveiled through examining the energy barriers of superlattice planar fault transformation. The symmetry breaking occurring in the unfaulting processes were further elucidated and summarized. A continuum model has been formulated to enable quantitative theoretical analysis of the energetics of IDLs and prediction of the unfaulting threshold. These findings clarify the effects of complex faults and symmetry breaking on unfaulting behaviors of IDLs in irradiated L12-Ni3Al, and could be transferable to other L12 systems, and more generally, other ordered alloy systems.

Phase-field study of the solutes-interstitial loops interaction in Fe–Cr alloys

Interstitial dislocation loops are typically formed in post-irradiated materials, such as Ferritic-Martensitic Fe–Cr alloys. These loops demonstrate solute segregation along their perimeters, effectively pinning the dislocation climb and resulting in high-density, small-sized loops that cause embrittlement. Phase-field simulations were conducted to investigate the behavior of Cr segregation in the stress field of a<100> type and various a/2<111> type interstitial dislocation loops in post-irradiated Fe–10Cr alloy. The study considers the long-range elastic interaction of solute Cr within the stress field of dislocation loops in a cubic elastic anisotropic material. The findings reveal a nonuniform stress distribution along the loop's perimeter depending on the included angle of habit plane normal n and Burgers vector b; the nonuniformity becomes more pronounced as the Burgers vector deviates from the normal direction. Furthermore, this nonuniform stress induces Cr segregation within regions experiencing tensile stress and Cr depletion in compression stress regions. Furthermore, a comprehensive analytical solution has been developed to characterize the diverse stress fields and solute segregation induced by interstitial dislocation loops with varying habit planes and Burgers vectors. The proposed analytical model effectively depicts the stress distribution of various dislocation loops, resulting in a Cr segregation profile that closely approximates those obtained through phase-field simulations. A thorough analysis of solute segregation mechanisms also elucidates the dispersed fine-scale nature and high density observed in experimental investigations of dislocation loops.

Direct Observation of Vacancy-Cluster-Mediated Hydride Nucleation and the Anomalous Precipitation Memory Effect in Zirconium.

Controlling the heterogeneous nucleation of new phases is of importance in tuning the microstructures and properties of materials. However, the role of vacancy-a popular defect in materials that is hard to be resolved under conventional electron microscopy-in the heterogeneous phase nucleation remains intriguing. Here, this work captures direct in situ experimental evidences that vacancy clusters promote the heterogeneous hydride nucleation and cause the anomalous precipitation memory effect in zirconium. Both interstitial and vacancy dislocation loops form after hydride dissolution. Interestingly, hydride reprecipitation only occurs on those vacancy loop decorated sites during cooling. Atomistic simulations reveal that hydrogen atoms are preferentially segregated at individual vacancy and vacancy clusters, which assist hydride nucleation, and stimulate the unusual memory effect during hydride reprecipitation. The finding breaks the traditional view on the sequence of heterogeneous nucleation sites and sheds light on the solid phase transformation related to vacancy-sensitive alloying elements.

Interaction of 1/2〈111〉 interstitial dislocation loop with hydrogen and helium in tungsten: molecular dynamics simulation

The interaction of hydrogen and helium atoms with 1/2 〈111〉 interstitial dislocation loop in tungsten is investigated by molecular dynamics simulation. The binding energies of hydrogen and helium atoms around dislocation loop are calculated by molecular statics method. The results show that the outer region of the loop is attractive to the two atoms and the inner region is repulsive. Notably, the maximum binding energies are located in the core region of the dislocation loop. We have also studied the influence factors of the interaction between the dislocation loop and two atoms: free volume, lattice distortion degree, the radius and shape of the dislocation loop. The results show that large free volume benefits the retention of hydrogen and helium atoms, especially for helium. The less lattice distortion caused by the impurity atom, the more favorable for the dislocation loop to trap it. In addition, the larger dislocation loop with higher defect concentration results in stronger capture ability for the hydrogen and helium atoms. The different dislocation loop shapes lead to different binding energy distribution patterns. And the hydrogen and helium atoms tend to occupy the groove region of the concave dislocation loop. Finally, we employ the nudged elastic band theory and dynamics method to investigate the diffusion pattern of the hydrogen atom in the dislocation loop and find that the hydrogen atom tends to migrate spirally around dislocation line. Based on the obtained results, a reasonable interpretation of the interaction behaviors between the dislocation loop with hydrogen and helium atoms are discussed, which can provide essential parameters for mesoscopic scale simulations.

Displacement cascades database from molecular dynamics simulations in tungsten

In this work, we have carried out a series of displacement cascade simulations in tungsten (W), including seven different primary-knock-on atom (PKA) directions, and PKA energy covers 1 to 300 keV at 363 K; 10, 50, 100, and 200 keV cascades are considered at 600 K, and 10 and 50 keV cascades are simulated at 300 K. Then, a systematic database of W cascade simulations has been established, containing more than 2000 cascades. The number of Frenkel pairs, the type of defect, and its spatial distribution were all determined. The effects of PKA direction, PKA energy and temperature have been discussed in terms of defect generation, defect clustering, and dislocation loops. It is found that, the PKA direction may have an effect on the configuration of defects, such as the formation probability of 〈100〉 interstitial loops, C15 clusters and the sub-cascades, and also has an effect on the number of Frankel pairs when PKA energy larger than 50 keV. Furthermore, from 363 K to 600 K, the number of Frankel pairs is reduced by ∼10%, and high temperature can promote the clustering of interstitial. We further found that temperature may affect the distribution of dislocation loops, and the number of 〈100〉 interstitial dislocation loops is more at 363 K than at 600 K. These results can provide a database of key input parameters for the subsequent larger-scale simulations, and give an important reference for the prediction of defect behaviour, mechanical and thermal properties of W-based materials under fusion neutron irradiation.

Molecular dynamics simulations of displacement cascades in vanadium: Generation and types of dislocation loops

Vanadium(V)-based alloys have received extensive attention as promising first-wall and blanket candidates for fusion reactors. In the present study, molecular dynamics (MD) simulations have been performed to study the displacement cascades in V bulk with the recoil energy up to 80 keV at 300 K. The results indicate that the clustering fraction and cluster size of vacancies are higher than those of interstitials. The nucleation, evolution, and energy of dislocation loops are studied. It’s found that 1/2 〈1 1 1〉 interstitial and vacancy dislocation loops are the primary ones, while low concentration 〈1 0 0〉 vacancy loops are also formed. In addition, the nucleation of 1/2 〈1 1 1〉 interstitial dislocation loops is earlier than that of 1/2 〈1 1 1〉 vacancy ones. The reason for the appearance of 1/2 〈1 1 1〉 vacancy dislocation loops is that the formation energy of the 1/2 〈1 1 1〉 vacancy dislocation loops is close to that of voids containing the same number of vacancies. These results provide a new understanding of applying V-based alloys in fusion reactors.

Molecular dynamics simulation of dynamic migration of &lt;inline-formula&gt;&lt;tex-math id="M3"&gt;\begin{document}$\boldsymbol {\langle 100\rangle} $\end{document}&lt;/tex-math&gt;&lt;alternatives&gt;&lt;graphic xmlns:xlink="http://www.w3.org/1999/xlink" xlink:href="24-20230651_M3.jpg"/&gt;&lt;graphic xmlns:xlink="http://www.w3.org/1999/xlink" xlink:href="24-20230651_M3.png"/&gt;&lt;/alternatives&gt;&lt;/inline-formula&gt; interstitial dislocation loops under (010) surfaces of pure W and W containing helium impurity

In the fusion irradiation environment, dislocation loop defects occurring under plasma-facing tungsten surface affect its mechanical properties and hydrogen/helium retention. This paper studies the dynamic behaviors of a &lt;inline-formula&gt;&lt;tex-math id="M13"&gt;\begin{document}$\langle 100\rangle $\end{document}&lt;/tex-math&gt;&lt;alternatives&gt;&lt;graphic xmlns:xlink="http://www.w3.org/1999/xlink" xlink:href="24-20230651_M13.jpg"/&gt;&lt;graphic xmlns:xlink="http://www.w3.org/1999/xlink" xlink:href="24-20230651_M13.png"/&gt;&lt;/alternatives&gt;&lt;/inline-formula&gt; loop with a radius of 1 nm under the W &lt;inline-formula&gt;&lt;tex-math id="M14"&gt;\begin{document}$(010)$\end{document}&lt;/tex-math&gt;&lt;alternatives&gt;&lt;graphic xmlns:xlink="http://www.w3.org/1999/xlink" xlink:href="24-20230651_M14.jpg"/&gt;&lt;graphic xmlns:xlink="http://www.w3.org/1999/xlink" xlink:href="24-20230651_M14.png"/&gt;&lt;/alternatives&gt;&lt;/inline-formula&gt; surface by using molecular dynamics simulation at the atomic level. It is found that the dislocation loop direction, bulk temperature, depth, and helium atoms can greatly affect the motion of dislocation loops, showing that the &lt;inline-formula&gt;&lt;tex-math id="M15"&gt;\begin{document}$ {\boldsymbol{b}}/ / {\boldsymbol{n}} $\end{document}&lt;/tex-math&gt;&lt;alternatives&gt;&lt;graphic xmlns:xlink="http://www.w3.org/1999/xlink" xlink:href="24-20230651_M15.jpg"/&gt;&lt;graphic xmlns:xlink="http://www.w3.org/1999/xlink" xlink:href="24-20230651_M15.png"/&gt;&lt;/alternatives&gt;&lt;/inline-formula&gt; dislocation loop, where &lt;inline-formula&gt;&lt;tex-math id="M16"&gt;\begin{document}$ {\boldsymbol{b}} $\end{document}&lt;/tex-math&gt;&lt;alternatives&gt;&lt;graphic xmlns:xlink="http://www.w3.org/1999/xlink" xlink:href="24-20230651_M16.jpg"/&gt;&lt;graphic xmlns:xlink="http://www.w3.org/1999/xlink" xlink:href="24-20230651_M16.png"/&gt;&lt;/alternatives&gt;&lt;/inline-formula&gt; is the Burgers vector and &lt;inline-formula&gt;&lt;tex-math id="M17"&gt;\begin{document}$ {\boldsymbol{n}} $\end{document}&lt;/tex-math&gt;&lt;alternatives&gt;&lt;graphic xmlns:xlink="http://www.w3.org/1999/xlink" xlink:href="24-20230651_M17.jpg"/&gt;&lt;graphic xmlns:xlink="http://www.w3.org/1999/xlink" xlink:href="24-20230651_M17.png"/&gt;&lt;/alternatives&gt;&lt;/inline-formula&gt; denotes the surface normal direction, tends to move towards the surface and the &lt;inline-formula&gt;&lt;tex-math id="M18"&gt;\begin{document}$ {\boldsymbol{b}} \bot {\boldsymbol{n}} $\end{document}&lt;/tex-math&gt;&lt;alternatives&gt;&lt;graphic xmlns:xlink="http://www.w3.org/1999/xlink" xlink:href="24-20230651_M18.jpg"/&gt;&lt;graphic xmlns:xlink="http://www.w3.org/1999/xlink" xlink:href="24-20230651_M18.png"/&gt;&lt;/alternatives&gt;&lt;/inline-formula&gt; dislocation loop tends to stay in the material. In the course of its migration, the habit plane of dislocation loop may change and the internal stress decreases gradually. The probability of a &lt;inline-formula&gt;&lt;tex-math id="M19"&gt;\begin{document}$ {\boldsymbol{b}}/ / {\boldsymbol{n}} $\end{document}&lt;/tex-math&gt;&lt;alternatives&gt;&lt;graphic xmlns:xlink="http://www.w3.org/1999/xlink" xlink:href="24-20230651_M19.jpg"/&gt;&lt;graphic xmlns:xlink="http://www.w3.org/1999/xlink" xlink:href="24-20230651_M19.png"/&gt;&lt;/alternatives&gt;&lt;/inline-formula&gt; dislocation loops escaping from the surface is over 90% when the temperature is higher than 800 K and their initial depth is less than 5 nm. The &lt;inline-formula&gt;&lt;tex-math id="M20"&gt;\begin{document}$ {\boldsymbol{b}} \bot {\boldsymbol{n}} $\end{document}&lt;/tex-math&gt;&lt;alternatives&gt;&lt;graphic xmlns:xlink="http://www.w3.org/1999/xlink" xlink:href="24-20230651_M20.jpg"/&gt;&lt;graphic xmlns:xlink="http://www.w3.org/1999/xlink" xlink:href="24-20230651_M20.png"/&gt;&lt;/alternatives&gt;&lt;/inline-formula&gt; dislocation loop can escape from the surface when the temperature is 800 K and the initial depth is less than 2 nm. It is found that &lt;inline-formula&gt;&lt;tex-math id="M21"&gt;\begin{document}$\langle 100\rangle $\end{document}&lt;/tex-math&gt;&lt;alternatives&gt;&lt;graphic xmlns:xlink="http://www.w3.org/1999/xlink" xlink:href="24-20230651_M21.jpg"/&gt;&lt;graphic xmlns:xlink="http://www.w3.org/1999/xlink" xlink:href="24-20230651_M21.png"/&gt;&lt;/alternatives&gt;&lt;/inline-formula&gt; dislocation loops decompose into &lt;inline-formula&gt;&lt;tex-math id="M22"&gt;\begin{document}${1 \mathord{\left/ {\vphantom {1 2}} \right. } 2}\langle 111\rangle $\end{document}&lt;/tex-math&gt;&lt;alternatives&gt;&lt;graphic xmlns:xlink="http://www.w3.org/1999/xlink" xlink:href="24-20230651_M22.jpg"/&gt;&lt;graphic xmlns:xlink="http://www.w3.org/1999/xlink" xlink:href="24-20230651_M22.png"/&gt;&lt;/alternatives&gt;&lt;/inline-formula&gt; dislocations at elevated temperatures. Helium atoms impede the migration of dislocation loops and increase their retention time. The existence of dislocation loops results in the uneven distribution of helium atoms under the W surface, and will potentially affect the surface morphology of tungsten.

Coupled effect of Cr and Al on interactions between a prismatic interstitial dislocation loop and an edge dislocation line in Fe-Cr-Al alloy

Segregation of alloying elements to a prismatic dislocation loop under irradiation is an important phenomenon in understanding the role of loops in radiation effects in alloys. In this work, various segregation sequences of Cr and Al atoms to a 〈100〉 and a 1/2〈111〉prismatic dislocation loop in a Fe-Cr-Al alloy are explored using molecular dynamics method and ab initio energy calculation. The results reveal that Cr can readily segregate to the loops while Al atoms cannot by themselves, but the segregated Cr atoms are able to promote Al segregation at the loops, defined as a coupled segregation of Cr and Al. After interacting with Cr and Al, a previously mobile 1/2〈111〉 loop becomes sessile. Al segregation through the coupled segregation process can further affect the pinning behavior of a dislocation loop to an edge dislocation motion. By varying the Al fraction in the segregated solutes, a nonlinear dependence of the pinning effect on dislocation motion is determined. At lower fractions, the pinning strength increases with the increasing Al fraction until Al fraction reaches a critical value of about 19%, above which the pinning strength decreases with further increase of Al fraction. All these results suggest that, in addition to the effect of Al on the surface resistance to oxidation, Al effect on the dislocation loop behavior is important to optimize the composition of Fe-Cr-Al alloys for their applications in accident tolerant fuel concepts.

Basic study of the relaxation volume of crystalline defects in bcc iron

The relaxation volume (Ωrel), here determined per extra-atom or vacant site, of common crystalline defects in bcc iron (Fe) was calculated from molecular dynamics (MD) simulation cells containing defects of varying size and/or density. To this end, we used both real and reciprocal space data: for the former, the change in the MD cell volume was calculated, while for the latter, we computed X-ray diffraction reciprocal space maps to evaluate the change in the lattice parameter. We show that 〈110〉 dumbbell self-interstitial atoms have the largest Ωrel, ∼1.5atomic volume (∼1.5 Ω0). C15 clusters of size 12 and 48 atoms show Ωrel of ∼ 0.91 Ω0 and ∼ 0.98 Ω0, respectively, and similar values are found for ½〈111〉 and 〈100〉 interstitial dislocation loops, with Ωrel ∼ 0.905 Ω0 and Ωrel ∼ 0.873 Ω0, respectively. Single vacancies are characterized by a negative Ωrel, ∼−0.11 Ω0. For cavities, Ωrel rapidly increases to approach zero as the clusters grow. Using these values, we managed to predict (with an accuracy better than 2 %) the lattice strain in MD cells containing several types of defects, which indicates that the relaxation volumes can be summed up to estimate the microscopic (i.e., lattice) volume change.

Exploring atomic mechanisms of microstructure evolutions in crystals under vacancy super- or undersaturation states by a kinetic amplitude-expanded phase-field-crystal approach

Exploring atomic mechanism of microstructure evolutions at long time still remains a great challenge at present. Amplitude-expanded phase field crystal (APFC) model derived from the classical density functional theory is a promising candidate to access this issue. However, it fails to describe dislocation evolutions in systems under super- or undersaturation states because of the lack of necessary rare events, which hampers its applications in the related realms, such as quick quenching, impacting, irradiating and so on. In this work, we find that the necessary rare events in solids are attributed to the kinetic disturbances due to the motion of local lattice elements instead of the traditional Gaussian noise. The kinetic disturbance is evaluated by the long-time-averaged motions of the local lattice elements as well as a general energy variation principle with respect to the virtual variation of the reciprocal lattice vector. The results by these two approaches are mutually verified. It is demonstrated that the APFC model with the kinetic disturbance converges to the mechanical-equilibrium-condition coupled APFC models at the long time limit and reduces to the original one when the high-energy events are forbidden. Further, the kinetic model is rationalized through theoretical analysis combined with numerical testing on the vacancy-mediated dislocation climb. As a practical application of the kinetic model, we explore the long-time annealing behaviors of dislocation loops in irradiated BCC crystals with different vacancy supersaturations. It is the first time that the vacancy-mediated shrinking, 1-D diffusive motion as well as changing habit plane of interstitial dislocation loops, known in experiments, are correctly predicted at atom scale.

Influence of grain boundaries on the loop-punching mechanism and shape of helium bubbles in tungsten

In the present work, the growth of helium bubble at Σ3(112)[110], Σ5(013)[100] and Σ13(015)[100] three grain boundaries in tungsten is studied using molecular dynamics simulation. The simulation results show that interstitial dislocation loops are pushed out earlier for the case He bubble growing at the Σ3(112)[110] and later for the case bubble growing at the Σ5(013)[100], while no prismatic interstitial dislocation loop is pushed out for the case bubble growing at Σ13(015)[100] by the end of the simulation. This is because the larger the excess volume per area of grain boundary, the later the prismatic interstitial dislocation loop pushed out. The shapes of helium bubbles are monitored by measuring bubbles sizes along three dimensions during their growing process. Helium bubble grows into a shape slightly deviating from the sphere at Σ3(112)[110] twin boundary; while grows into ellipsoid at Σ5(013)[100] and Σ13(015)[100] grain boundaries. Based on a two-dimensional model, the shapes of helium bubbles are then investigated quantitatively in terms of wetting theory. The theoretical shapes of helium bubbles at the three grain boundaries are in agreement with those obtained by molecular dynamics simulation method, indicating that the growth of helium bubble at grain boundaries obeys the wetting theory.

Radiation-induced segregation on dislocation loops in austenitic Fe-Cr-Ni alloys

Radiation-induced segregation of alloying elements to crystallographic defects is commonly observed in irradiated austenitic stainless steels. The interaction between solutes and radiation-induced defects changes the physical distribution of solutes and thus affects the formation and growth of defects. The change of the microstructure consequently affects the mechanical properties of the material. A qualitative and quantitative understanding of the interaction between solutes and defects is desirable to better predict the service lifetime of nuclear materials. We used atom probe tomography to measure the distribution of solutes at dislocation loops in 304L stainless steel, irradiated with 2 MeV protons up to 1.5 displacements per atom at 373 and 633 K. No segregation at dislocation loops was found in samples irradiated at 373 K, whereas Ni and Si enrichment and Cr depletion were detected at dislocation loops irradiated at 633 K. The experimentally observed perfect and faulted dislocation loops in vacancy and interstitial types were reproduced by molecular dynamics (MD). A hybrid MD/Monte Carlo method was used to predict the redistribution of alloying atoms at all possible types of dislocation loops in face-centered cubic Fe-Cr-Ni alloys at the same irradiated temperatures (373 and 633 K). The simulations show that, at both temperatures, Cr clusters were formed and distributed randomly, and Ni atoms enriched or depleted at interstitial or vacancy dislocation loops, respectively. The change of solute concentration reaches the highest at the edge of the loop. Ni profiles exhibit characteristic behavior in terms of the stress field of the loops: tension inside of vacancy loops showing depletion of Ni atoms compared with compression inside of interstitial loops showing enrichment of Ni atoms. In addition, the stress field is reduced after solute redistribution. The absence of alloying segregation observed in experiments at a lower temperature (373 K) is explained by a rate theory model: Low-temperature irradiation requires significantly longer irradiation time to see the same amount of segregation as at high temperatures because of the extremely low diffusion of vacancies at low temperatures.

Atomistic mechanism of phase transformation between topologically close-packed complex intermetallics

Understanding how topologically close-packed phases (TCPs) transform between one another is one of the challenging puzzles in solid-state transformations. Here we use atomic-resolved tools to dissect the transition among TCPs, specifically the μ and P (or σ) phases in nickel-based superalloys. We discover that the P phase originates from intrinsic (110) faulted twin boundaries (FTB), which according to first-principles calculations is of extraordinarily low energy. The FTB sets up a pathway for the diffusional in-flux of the smaller 3d transition metal species, creating a Frank interstitial dislocation loop. The climb of this dislocation, with an unusual Burgers vector that displaces neighboring atoms into the lattice positions of the product phase, accomplishes the structural transformation. Our findings reveal an intrinsic link among these seemingly unrelated TCP configurations, explain the role of internal lattice defects in facilitating the phase transition, and offer useful insight for alloy design that involves different complex phases.

Orientation dependence of shock-induced change of habit plane for the 1/2 dislocation loop and plasticity in tungsten

Tungsten (W) is a promising candidate material for future fusion reactors. The shock waves generated under high-energy neutron radiation can lead to the formation of prismatic interstitial dislocation loops (PIDLs). To understand the details of the mechanisms, the interaction between the shock waves and PIDL with Burgers vector of 1/2<111> was studied by using nonequilibrium molecular dynamics (NEMD) simulation. The shock-induced change of habit plane for the 1/2<111> PIDL and plasticity in W depend strongly on the crystallographic orientations. The driving force for changing the loop's habit plane is derived from the resolved shear stress (RSS). A new rotation mechanism is proposed, which can be used to predict the changing trend of PIDL's habit plane. The rotation angle of the habit plane for the 1/2<111> dislocation loop is proportional to the RSS of the activated slip system. In this work, we also predict a source to induce the plasticity (e.g., deformation twin and dislocation network) observed in the experiment and discuss the nucleation, propagation and interaction of these deformations with different crystallographic orientation. The current results provide significant insights into the evolution of 1/2<111> PIDL depending on the crystallographic orientation under shock loading.

Dislocation loop evolution in Kr‐irradiated ThO2

AbstractThe early stage of microstructural evolution of ThO2, under krypton irradiation at 600, 800, and 1000°C, was investigated using in situ transmission electron microscopy (TEM). Dislocation loops grew faster, whereas their number density decreased with increasing irradiation temperature. Loop density was found to decrease with ion dose. Interstitial dislocation loops, including Frank loops with Burgers vector of a/3〈111〉 and perfect loops with Burgers vector of a/2〈110〉, were determined by traditional TEM and atomic resolution–scanning TEM techniques. Atomistic and mesoscale level modeling are performed to interpret experimental observations. The migration energy barriers of defects in ThO2 were calculated using density‐functional theory. The energetics of different dislocation loop types were studied using molecular dynamics simulations. Loop density and diameter were analyzed using a kinetic rate theory model that considers stoichiometric loop evolution. This analysis reveals that loop growth is governed by the mobility of cation interstitials, whereas loop nucleation is determined by the mobility of anion defects. Lastly, a rate theory model was used to extract the diffusion coefficients of thorium interstitials, oxygen interstitials, and vacancies.