Formation Of Dislocation Loops Research Articles

Microstructure evolution under He2+ ions irradiation in a high entropy alloy designed by constructing nanostructure engineering

This study aims to reduce the size and density of He bubbles and dislocation loops during the irradiation process by constructing nanostructure engineering through precipitation. The Ni35(CoFe)55V5Nb5 high-entropy alloy (HEA) with solid solution and aging states underwent He2+ ions irradiation (500 keV at a fluence of 1017 ions∙cm−2) at 500 °C. The results indicate that the preexisting nanoprecipitates increased the surface area for He atoms adsorption, and improved the ability of the alloy to suppress the coarsening of He bubbles. In the aged HEA, He bubble size decreased (∼3.3 ± 0.4 nm), and density significantly reduced (0.58 ± 0.14 × 1023 m−3) compared to the solid solution HEA (∼6.1 ± 0.3 nm, 2.42 ± 0.15 × 1023 m−3). Moreover, the preexisting nanoprecipitates coordinate the movement and recombination of irradiation defects, which overwhelms the energy driving force that improves the stability and growth of dislocation loops due to the reduction of stacking fault energy, thereby suppressing the formation and growth of dislocation loops. The aged HEA exhibited reduced size and density of dislocation loops (∼6.3 ± 0.5 nm and 0.78 ± 0.15 × 1020 m−3) compared to the solid solution HEA (∼8.5 ± 0.6 nm, 3.83 ± 0.13 × 1020 m−3).

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.

Effect of dislocation defects on the nano-scratching process of 4H–SiC

To investigate the influence of defects on the damage mechanism of 4H–SiC, this study primarily conducted molecular dynamics simulations of a single diamond abrasive indentation process and carried out subsequent nanoindentation experiments on 4H–SiC samples. The simulation and experimental results were compared, and the force–displacement curves showed consistent trends. On this basis, the scratching simulation of an ideal material and a single diamond abrasive particle containing 4H–SiC with dislocation defects was carried out to study the influence of internal dislocation (screw dislocation and edge dislocation) in 4H–SiC and its position on the material damage mechanism during scratching, and the development and evolution of internal dislocation were further analyzed. The results reveal that the presence of internal dislocations in the material is one of the causes of the formation of dislocation loops.

Vacancy‐Type Defects and Their Trapping/Detrapping of Charge Carriers in Ion‐Implanted GaN Studied by Positron Annihilation

Annealing behaviors of vacancy‐type defects in ion‐implanted GaN are studied by positron annihilation. Mg+ and N+ ions are implanted to obtain 700 nm deep box profiles with Mg and N concentrations of 1 × 1018 cm−3, and the samples are annealed using an ultrahigh‐pressure annealing system. For as‐implanted samples, the major defect species is identified as Ga‐vacancy (VGa)‐type defects. For N‐implanted GaN, the size of the vacancies increases as the annealing temperature increases up to 1100 °C and then shrank above 1200 °C. This behavior is attributed to recombinations between N‐vacancy (VN)‐type defects and excess N. For Mg‐implanted GaN, the major defect species after annealing above 1000 °C is vacancy clusters such as (VGaVN)3. Some of them act as nonradiative recombination centers for blue and ultraviolet luminescence. Their energy levels corresponding to the transition from positive to neutral are located between 2.6 eV above the valence band maximum and the conduction band minimum. The thermal activation process of electron detrapping from the vacancy clusters is also studied. For Mg‐implanted GaN, one of the major secondary defects is collapsed vacancy disks forming dislocation loops. They are eliminated by additional N implantation, which is associated with vacancy agglomerations under the annealing in VGa‐rich condition.

Prediction of radiation embrittlement of VVER-1000 RPV welded seam material at lifetime extension up to 60 years and more

The correctness of the model assumption about additivity of the contributions of hardening (formation of dislocation loops and radiation-induced precipitates) and non-hardening (formation and accumulation of grain boundary segregations) mechanisms to the final radiation embrittlement of VVER-1000 RPV weld materials has been verified. The paper presents the results of the experiment to obtain predicted values of the critical embrittlement temperature of weld metal with nickel content of 1.59–1.88% using accelerated irradiation in the IR-8 research reactor from states characterized by different starting level of grain boundary segregation. The procedure of accounting for the flux effect for the values of critical embrittlement temperature obtained using accelerated irradiation has been proposed. The conservativity of the weld metal radiation embrittlement dependence has been verified based on the modal assumption of additivity of the contributions of the strengthening and non-strengthening mechanisms up to the values of fast neutron fluence ~75–1022 neutrons/m2.

TEM Investigation on the Dislocation Loops in Zirconium Alloy by Neutron and Kr+ Ion Irradiation Tests.

Neutron irradiation (E ≥ 1 MeV) and 400 keV Kr+ ion irradiation tests were carried out and compared to study the formation and growth mechanism of dislocation loops in zirconium alloys. The results show that, as the irradiation reached 1.9 × 1025 n/m2 and 1.0 × 1020 Kr+/m2, a large number of <a>-type dislocation loops and a small amount of large-sized <c>-type dislocations hundreds of nanometers or even micrometers in size are observed in both samples. EDS results confirmed that the concentrations of Nb and Fe in the <c>-type dislocation loops are higher than that in the matrix. With the increases of the dislocation loop size, part of the loop structure will decompose into the twisted, small-sized dislocation lines. Between two of the small-sized dislocation lines, a cubic structure was observed.

Free-surface proximity induced bistability and semi-semicoherency in precipitation systems

In systems without a ‘high misfit’ and in the early stage of growth, the precipitates tend to remain coherent with the parent matrix. On further growth beyond a critical size (r*), the precipitates undergo a coherent to semi-coherent transition by formation of interfacial misfit dislocation loops to minimize the total strain energy of the system. The variation of r* with domain size has been studied and the concept of re-entrant interface transition to the coherent state was demonstrated in finite crystals. Dislocations can exist in a state of mechanical equilibrium; but are thermodynamically metastable. In the proximity of ‘interfaces’ dislocations experience a configurational force. The current work pertains to study the stability of an interfacial misfit dislocation loop in the proximity of a free-surface; wherein we evaluate configurational effects in a semi-infinite body. The Cu-2 wt%γFe alloy is used as a model system in this study. In this work we: (i) demonstrate the stabilization of the coherent state (divergence of r*), induced by the proximity of the free-surface; (ii) discover the existence of bistability in materials at the microstructural scale, resulting from the destabilization of the equatorial position of the dislocation loop and arising from symmetry breaking in climb of the interfacial loop due to the vicinity of a free-surface. (iii) identify regimes of ‘equi-bistablity’ and regions with asymmetry in energy wells, (iv) establish the existence of a ‘semi’-semicoherent state. Here, one specific system has been chosen to illustrate the effects; however, we expect the conclusions to have a broader applicability.

Study of Vacancy Clusters and Antisite Defects in Titanium Beryllide Be12Ti Using Positron Annihilation Spectroscopy

Beryllium (Be) intermetallic compounds (beryllides) are promising materials for neutron multiplication in fusion reactors. In particular, titanium beryllide (Be12Ti) exhibits low swelling and good corrosion resistance compared to other beryllides. Herein, the defects in neutron‐irradiated Be12Ti are investigated using positron annihilation spectroscopy, and the results are analyzed based on positron annihilation calculations. Calculation results indicate that the positron lifetime in a perfect Be12Ti crystal is 134.2 ps, while that of a Be monovacancy is found to be within the 177.3–195.0 ps range, depending on the configuration of the monovacancy. On the other hand, the positron lifetime of the Ti monovacancy is 222.7 ps. In addition, the lifetimes of the Be and Ti vacancy clusters increase with an increase in the number of vacancies. Finally, the positron lifetimes of the antisite defects in different configurations are found to be within the 137.9–141.7 ps range. In contrast to the calculated results, the positron lifetime of unirradiated Be12Ti is 141.1 ps. Furthermore, in the neutron‐irradiated Be12Ti, no vacancies are detected and only Be dislocation loop formation occur based on the calculation results. Although the irradiation dose of Be12Ti is low, its irradiation resistance is satisfactory.

Post-irradiation examination of commercial tantalum alloys following neutron irradiation

Post irradiation examination of two commercial tantalum alloys (based on Ta-8 %W) revealed significant irradiation hardening following neutron irradiation in the High Flux Isotope Reactor at Oak Ridge National Laboratory. Tensile samples of commercial Ta alloys T-111 and ASTAR-811C were irradiated to a total neutron fluence (E > 0.1 MeV) in the range (2–4)x1021 n/cm2 (0.39–0.75 dpa) at an irradiation temperature near 900 °C. Tensile testing conducted at room temperature and at 800 °C revealed significant irradiation hardening under all test conditions. Samples that were tensile tested at room temperature exhibited brittle failure, whereas ductility was maintained for elevated temperature testing. The irradiation hardening and embrittlement noted for both alloys was attributed to irradiation-induced dislocation loop formation as confirmed through scanning transmission electron microscopy analysis. Corresponding energy dispersive x-ray spectroscopy revealed radiation-induced segregation of Hf in these loop populations. This Hf enrichment is shown to result in eventual co-precipitation of elongated (Hf,O)-rich precipitates along the dislocation loops. These results show that irradiation hardening is enhanced via irradiation-enhanced precipitation at elevated irradiation temperatures. Consequently, the minimum recommended operating temperature window for Ta alloys should be increased to account for observed hardening/embrittlement at irradiation temperatures as high as 900 °C (1,173 K).

Gradual degradation in InAs quantum dot lasers on Si and GaAs.

Reliable quantum dot lasers on silicon are a key remaining challenge to successful integrated silicon photonics. In this work, quantum dot (QD) lasers on silicon with and without misfit dislocation trapping layers are aged for 12 000 hours and are compared to QD lasers on native GaAs aged for 8400 hours. The non-trapping-layer (TL) laser on silicon degrades heavily during this time, but much more modest gradual degradation is observed for the other two devices. Electroluminescence imaging reveals relatively uniform gradual dimming for the aged TL laser on silicon. At the same time, we find nanoscale dislocation loop defects throughout the quantum dot-based active region of all three aged lasers via electron microscopy. The Burgers vector of these loops is consistent with . We suggest that the primary source of degradation, however, is the generation and migration of point defects that substantially enhance non-radiative recombination in the active region, the visible symptom of which is the formation of dislocation loops. To prevent this, we propose that laser fabrication should be switched from deeply etched to shallow etch ridges where the active region remains intact near the mesa. Additionally, post-growth annealing and altered growth conditions in the active region should be explored to minimize the grown-in point defect density.

Theoretical strength of austenite in the presence of a pore or vacancies in the crystal: Molecular dynamics study

The molecular dynamics method was used to study the influence of pores of different diameters, as well as the corresponding concentration of individual vacancies, on the theoretical strength of austenite at different temperatures. The deformation in the model was carried out by shear at a cons­tant rate of 20 m/s. We considered a shear along two directions: [ \(\bar 1\ \bar 1\) 2] and [111]. The computational austenite cell had the shape of a rectangular parallelepiped 14.0 nm long, 14.0 nm high, and 5.1 nm wide. To describe interatomic interactions, the Lau EAM potential was used, which reproduces well the structural, energy, and elastic characteristics of austenite. The stress-strain curves obtained for both considered shear directions had a similar form. In the absence of dislocation sources, plastic deformation was carried out by the formation of dislocation dipoles (dislocations with opposite Burgers vectors). The presence of a pore significantly reduced the yield strength of austenite. In this case, it was found that single vacancies randomly scattered over the volume of the computational cell also lead to a decrease in the yield strength, but, of course, not as much as the pore. The emission of dislocations during deformation occurred by the formation of dislocation loops, as a rule, in two slip planes at once. The effect of pores and vacancies on the yield strength was stronger at low temperatures. As the temperature increased, the effect of defects on the critical stress at which dislocations were formed decreased. With an increase in the pore size, as well as the concentration of vacancies, the yield strength decreased. In this case, the strongest dependence was observed for pores up to 1 nm in diameter. The influence of the concentration of vacancies in the considered range on the yield strength turned out to be comparatively smoother and almost linear.

Defect-specific strength factors and superposition model for predicting strengthening of ion irradiated Fe18Cr alloy

A high-purity binary alloy Fe18Cr was subjected to heavy-ion irradiation in order to provide improved understanding of the irradiation effect on radiation hardening associated with dislocation loop and network formation, α’ precipitation. The specimens were irradiated with 8 MeV Fe ions (∼2 μm ion range) to midrange doses of 0.37 and 3.7 dpa at 300, 350 and 450 °C using dose rates of ∼10−5–10−3 dpa/s. Nanoindentation testing was performed to extract the bulk equivalent hardness at low depth from these ion irradiated specimens. High-quality transmission electron microscopy (TEM) images were acquired by utilizing a flash electropolishing method on the Focused Ion Beam (FIB) prepared liftouts. An accurate calculation of the strength factor was provided based on detailed TEM characterization of the irradiated microstructures. A newly refined hardening superposition method was applied to combine the strengthening components from each microstructure. The good agreement between microstructure predicted strength and measured strength further demonstrated the fidelity to use hardening model to quantify the mechanical properties of ion irradiated materials.

Effect of temperature, pressure, crystal defect types, and densities on the mechanical behavior of tungsten under tensile deformation: A molecular dynamics simulation study

The mechanical behavior of tungsten under tensile deformation was investigated using molecular dynamics simulations, with a focus on the effects of temperature, pressure, crystal defect types, and densities. Single vacancies, spherical voids, and self-interstitials were studied at concentrations ranging from 0.1% to 2%. The Common neighbor analysis reveals that the crystal structure underwent a phase transition from the body-centered cubic to the face-centered cubic structure at increasing levels of applied strain, followed by the formation of a disordered phase. Increasing temperature reduced the mechanical properties of tungsten, including yield strength, yield strain, modulus of elasticity, and ultimate tensile stress, possibly due to the activation of dislocation nucleation and propagation, resulting in the formation of dislocation loops. Increasing pressure enhanced the mechanical properties of tungsten, with the formation of dislocation loops occurring at much lower strains at higher pressures. All mechanical parameters decreased with increasing defect density, with self-interstitial defects having a more severe effect than single vacancies and spherical voids. The number and length of dislocation loops also increased with increasing defect density. Temperature and pressure had varying effects on the stress–strain curve of tungsten with different defect types and densities, with increasing pressure generally enhancing its mechanical behavior and increasing temperature decreasing it. These findings provide valuable insights into the design and development of tungsten-based materials with enhanced mechanical properties.

Study on the mechanism of He effect on dislocation loop formation in three types of dual ion beam irradiations

A comprehensive investigation was conducted to explore the impact of helium on dislocation loops in RAFM steel. This was achieved by employing three types of dual-beam irradiations, which include simultaneous dual-beam irradiation with Ar++He+, sequential dual-beam irradiation with Ar+/He+ and He+/Ar+. In addition, single-beam irradiation with Ar+ and He+ were also performed for reference. The irradiation temperature is 723 ± 5 K. The evolutions of dislocation loop were characterized by transmission electron microscopy (TEM). After He+ irradiated to 0.005 dpa, high-density loops with an average size of 4.3 nm were observed. The main mechanism for the formation of the loops at such a low damage dose should be attributed to the process of the ejection of self-interstitial atoms triggered by He clusters. Significantly, in the RAFM steel exposed to Ar++He+ irradiation, the dislocation loops exhibit a significantly larger average size compared to those resulting from Ar+/He+ and He+/Ar+ irradiation. The loop density produced by He+/Ar+is the largest, and the loop density produced by Ar+/He+ is the smallest. The order of the amount of self-interstitial atoms (loops) caused by three types of dual ion beam irradiation is Ar++He+ > He+/Ar+ > Ar+/He+.

Orientation selected micro-pillar compression behavior of heavy-ions irradiated RAFMs

The present study aimed to investigate the mechanical property of heavy ions irradiated Reduced Activation Ferrite/Martensitic (RAFM) steels. The CLF-1 steel was irradiated by 33.5 MeV Fe ions at room temperature, resulting a peak displacement damage of 20 displacements per atom (dpa). Experimental analysis was conducted using Electron Backscatter Diffraction (EBSD) mapping-assisted Focused Ion Beam (FIB) technique to prepare micro-pillars oriented along crystal directions <0 0 1>. Subsequently, these fabricated micro-pillars underwent compression tests utilizing a flat indenter within the Nano-indentation setup. Engineering Stress-strain curves were generated for specimens both un-irradiated and irradiated. The yield strength of the un-irradiated and irradiated specimens in the <0 0 1> directions was determined by using the engineering stress-strain curves. In addition, the critical resolved shear stress (CRSS) of the un-irradiated and irradiated specimens was calculated. Microstructural analysis revealed that the formation of dense and fine dislocation loops were appeared in the material after irradiation. By combining the results of microstructure analysis with the dispersion barrier hardening model, the changes in CRSS caused by irradiation were calculated. Furthermore, the number of dislocation lines participating in deformation at a strain of 20 % was quantified, with the irradiated specimen exhibiting an increase of 33 % in comparison to the un-irradiated specimen.

Enhanced irradiation tolerance of a medium entropy alloy via precipitation and dissolution of nanoprecipitates

High/medium entropy alloys (HEAs/MEAs) are good candidates for nuclear applications due to the excellent mechanical properties, good corrosion resistance and radiation resistance. In this work, a novel cobalt-free MEA was developed by introducing L12 nanoprecipitates. The microstructure evolution and radiation tolerance were evaluated after bombarded using 3 MeV Fe11+ ions at 500 °C. The evolution of nanoprecipitates was closely related to the irradiation dose, and dominated by irradiation-enhanced diffusion and ballistic dissolution mechanism. For the solid-solution MEA (without L12 nanoprecipitates), irradiation hardening occurred due to the irradiation-induced formation of precipitates, voids and dislocation loops. However, in MEA with L12 nanoprecipitates after aging, irradiation induced dissolution and reprecipitation of nanoprecipitates were observed. Different from the solid-solution MEA, the hardness kept almost unchanged in the aging sample after irradiation. The swelling rate of the solid-solution sample after irradiation is 9.4 × 10−6 %/dpa, while no swelling occurs in the aging sample under the same irradiation condition. Besides, the average size and number density of dislocation loops in the aging sample decreased by ∼ 40% and ∼ 28%, compared with the solid-solution sample. The precipitation and dissolution of nanoprecipitates substantially improved the radiation tolerance of the cobalt-free MEA.

Radiation-induced hardening of ion-irradiated Zircaloy-2 at 573 K under applied stress

The microstructural evolution and hardening of Zircaloy-2 under low-dose irradiation (<1.0 dpa) with 3.2 MeV Ni3+ ions under applied stress at 573 K were investigated using transmission electron microscopy (TEM) examination and nano-indentation measurements in this study. The irradiation doses ranged from 0.05 to 1 dpa. The formation of < a > loops was observed at doses below 0.05 dpa, and the density was found to be saturated at about 0.1 dpa. The < a > loops then grew and intersected with adjacent loops to become entangled at doses above 0.4 dpa. Radiation-induced hardening was confirmed in Zircaloy-2 at the early stage of irradiation below 0.1 dpa, and the hardening effect was attributed to the nucleation and growth of < a > loops. The Orowan equation was used to estimate the contribution of dislocation loops to the hardening, and the estimated values were in good agreement with the experimental results, with the obstacle strength factor of < a > loops as 0.25. Furthermore, the effect of applied stress of 1 dpa irradiated tensile samples on radiation-induced hardening was also discussed. This study provides dislocation loop formation and hardening behavior of Zircaloy-2 under low-dose irradiation at 573 K with stress conditions.

Interactive effect of thermal aging and proton irradiation on microstructural evolution and hardening of δ-ferrite in 308L stainless steel weld metal

In the harsh service environment of high temperature and intense neutron irradiation in water-cooled nuclear reactors, the austenitic stainless steel weld overlay cladding on the inner surface of the reactor pressure vessel suffers from thermal aging and irradiation damage simultaneously, which can induce microstructural evolution and hardening of the material. Since it is quite difficult to achieve this simultaneous process out of the pile, two kinds of combined experiments, i.e., post-irradiation thermal aging and post-aging irradiation were performed on 308 L stainless steel weld metals in this work. The interactive effect of thermal aging and proton irradiation on microstructural evolution and hardening of δ-ferrite in 308 L weld metal was investigated by combining atom probe tomography, transmission electron microscopy and nanoindentation tests. The results revealed that thermal aging could eliminate the dislocation loops induced by irradiation and affect the phase transition process by accelerating spinodal decomposition and G-phase precipitation, thus enhancing hardening of irradiated δ-ferrite. For the effect of irradiation on the microstructure and hardening of thermally aged δ-ferrite, however, intensive collision cascades can intensify G-phase precipitation and dislocation loop formation but decrease spinodal decomposition, leading to a limited effect on hardening of thermally aged δ-ferrite. Furthermore, the interaction of thermal aging and irradiation can promote G-phase precipitation. Meanwhile, the interaction can cause δ-ferrite hardening, which is mainly influenced by spinodal decomposition, followed by G-phase and dislocation loops, where spinodal decomposition and G-phase cause hardening by inducing strain fields.

Primary radiation damage in tungsten-based high-entropy alloy: Interatomic potential and collision cascade simulations

Tungsten (W)-based high-entropy alloys (HEAs) have shown promising properties as nuclear fusion materials. Characterizing the primary damage is a critical step in describing and revealing the irradiation-induced damage and radiation resistance mechanism. However, there are a limited number of studies on collision cascades and primary radiation damage in W-based HEAs, owing to the large amount of calculations involved and a lack of appropriate interatomic interaction potentials. In this work, we developed a semi-empirical interatomic potential for W–Ta–Cr–V. By using the developed potential, molecular dynamics simulations of collision cascades in W-based HEAs were performed to assess the primary damage due to irradiation. Based on experimental samples, we reported defect production in W38Ta36Cr15V11 and compared it to pure W for primary knock-on atom with energies ranging from 1 keV to 100 keV. Our findings showed that the number of FPs at the thermal spike and the number of surviving FPs at the end of the cascade in W38Ta36Cr15V11 are more than those in pure W, mainly due to the lower threshold displacement energy, melting temperature, and formation energy of point defects. Collision cascades in the W38Ta36Cr15V11 are less likely to result in the formation of dislocation loops compared to pure W. After collision cascade, in W38Ta36Cr15V11, the concentrations of Cr and V atoms in defects is significantly higher than their corresponding concentrations in the system, showing an aggregation tendency. The current collision cascade results provide insights into the primary damage of W-based HEA system under irradiation and should provide reliable guidance for describing the primary damage source terms needed in the kinetic models used to simulate radiation-induced microstructural evolution.

Diffusion of small anti-Schottky clusters in UO2

The aggregation of irradiation-induced defect clusters in UO2 leads to the formation of dislocation loops and cavities which contribute swelling and ultimately produce deleterious effects on the fuel. Despite their fundamental role in fuel evolution, the kinetics of these defect clusters are not currently well understood. In this work, we investigate the diffusive behavior of interstitial clusters via Molecular Dynamics (MD) simulations. Our investigation considers a range of defect cluster sizes, each composed of N UO2 units or anti-Schottky defects. We report a complex cluster size - mobility relation; increasing cluster size corresponds to an increase in mobility up until a critical size (N=4) where diffusivity reaches a maximum, after which the trend reverts and further increasing size corresponds to a decrease in mobility. The rapid migration observed correlates well with the very low barriers reported in a few historic experimental studies. Further analysis of the cluster shape and orientation reveals a strong structural preference for near-planar configurations oriented normal to the 〈100〉 direction. Rotational energy barriers are found to be similar in both magnitude and trend to the observed diffusion barriers. Lastly, connections are drawn to the behavior of large defect clusters including the formation of (1/3)〈111〉 Frank dislocation loops.