Interstitial-type Dislocation Loops Research Articles

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.

Micron-scale 1D migration of interstitial-type dislocation loops in aluminum

Micron-scale one-dimensional (1D) migration of dislocation loops was discovered by in-situ transmission electron microscope (TEM) observations. Interstitial-type dislocation loops were initially implanted in pure aluminum foils with 30 keV H+. The dynamic behaviour of these loops was investigated in a conventional TEM, operated at 200 kV. During electron irradiation, dislocation loops exhibited super long-range (1.5 μm) 1D migration, featured average speeds over 1 nm/s, and left behind long-lasting tracks up to 60 s. The increase in loop migration speed gave rise to increased track length but showed a subtle impact on the existing time-scale of tracks. Deliberate investigations by varying electron beam positions revealed that the migration distance could be boosted up to micron-scale when exposed to high concentration gradients of self-interstitial atoms built during electron irradiation. Current findings may shed light on the understanding of irradiation damage evolution and bring about new insights into the development of irradiation-resistant materials.

Delayed plasticity during spherical nanoindentation of an iron-chromium-aluminum alloy: Effect of ferric ion irradiation

Spherical nanoindentation of an iron-chromium-aluminum alloy was conducted to study the effect of ferric-ion (Fe3+) irradiation on the time-dependent plasticity behavior in the surface layers of this alloy. It was observed that the initiation of plasticity by the appearance of displacement burst or “pop-in” event occurred after a period of waiting time in the apparent elastic regime and that Fe3+ irradiation at 360 °C and up to ∼0.5 displacements per atom could make it happen under the lower applied loads but with a reduced magnitude. Through the experimental data, an activation volume and activation energy were extracted for the delayed plasticity. The results show that Fe3+-irradiation significantly reduced its activation volume from ∼3.05b3 to ∼1.75b3 (where b = Burgers vector), but slightly increased its activation energy from ∼0.65 to ∼0.71 eV. On the other hand, high-resolution scanning transmission electron microscopy observations reveal that the irradiation at the elevated temperature created interstitial atom pair onto the (100) habit plane that can serve as the nucleation site of a 〈100〉 dislocation loop while eliminating the pre-existing dislocations. Consequently, it is indicated that heterogeneous nucleation of the dislocation loop was predominant in the delayed plasticity initiation of this alloy and that the nucleation of the interstitial-type dislocation loop was involved due to Fe3+-irradiation.

Atomistic simulations of defect accumulation and evolution in heavily irradiated titanium for nuclear-powered spacecraft

Titanium alloys are expected to become one of the candidate materials for nuclear-powered spacecraft due to their excellent overall performance. Nevertheless, atomistic mechanisms of the defect accumulation and evolution of the materials due to long-term exposure to irradiation remain scarcely understood by far. Here we investigate the heavy irradiation damage in α-titanium with a dose as high as 4.0 canonical displacements per atom (cDPA) using atomistic simulations of Frenkel pair accumulation. Results show that the content of surviving defects increases sharply before 0.04 cDPA and then decreases slowly to stabilize, exhibiting a strong correlation with the system energy. Under the current simulation conditions, the defect clustering fraction may be not directly dependent on the irradiation dose. Compared to vacancies, interstitials are more likely to form clusters, which may further cause the formation of 1/3<1¯210> interstitial-type dislocation loops extended along the (¯1010) plane. This study provides an important insight into the understanding of the irradiation damage behaviors for titanium.

Direct Observation of Cu Clusters and Dislocation Loops by Cs-Corrected STEM in Fe-0.6wt%Cu Alloy Irradiated in BR2

The neutron irradiation of Fe-based fusion and fission reactor materials leads to an increase in ductile-to-brittle transition temperature with a decrease in upper shelf energy. It is well known that Cu content has a strong influence on the embrittlement phenomenon, as Cu-rich clusters (CRPs) are thought to be directly responsible for embrittlement. In contrast, mechanical property studies for steels with different Cu levels exhibit dominant matrix defects in the embrittlement of both low-Cu steels and high-Cu steels at high fluences. To determine the effects of dislocation loops and CRPs on radiation hardening in those steels, neutron irradiation was conducted on Fe-0.6wt%Cu alloy. The neutron irradiation was performed in BR2 at 290 °C up to a dose of 4.1 × 1024 n/m2. After irradiation, the microstructure was observed and analyzed by spherical aberration-corrected transmission electron microscopy (TEM) and scanning transmission electron microscopy (STEM) combined with X-ray energy-dispersive spectroscopy, using a JEOL ARM200FC. This technique enabled simultaneous observation of ~10 nm CRPs and dislocation loops. Additional high-voltage electron irradiation was performed at room temperature, and the dislocation loops were identified as interstitial-type dislocation loops. Radiation-induced hardening due to neutron irradiation was estimated by measuring the size and density of the dislocation loops and the CRPs. These results suggest that simultaneous observation of dislocation loops and CRPs using the Cs-corrected STEM with EDS analysis is essential for the study of radiation-induced hardening in Fe-based alloys.

Atomistic structure and thermal stability of dislocation loops, stacking fault tetrahedra, and voids in face-centered cubic Fe

Irradiation-induced hardening and changes to mechanical properties can often lead to material degradation and can limit the operation of reactor components. In addition, changes to mechanical properties can be a precursor to other material environmental degradation. Dislocation loops, stacking fault tetrahedra (SFT), and voids are major irradiation-induced lattice defects that occur in structural materials and are often observed post-irradiation. However, these defects are often considered only as obstacles to dislocation motion, with hardening assumed to be dependent on the habit plane and Burgers vector of the defect. The stability and nature of dislocation loops and SFTs with temperature are often overlooked, and this may play a crucial role in the mechanical properties of structural materials. We combine high-resolution transmission electron microscopy (HR-TEM) and atomistic simulations to determine the atomistic configurations and thermal stability of dislocation loops and SFTs. Dislocation dissociations are observed in both vacancy and interstitial-type perfect dislocation loops as well as in vacancy-type Frank loops, whereas interstitial-type Frank loops remain stable and are the least influenced by temperature. Triangle-shaped intrinsic stacking faults are formed at the edges of Frank vacancy loops, and a direct transformation from triangle-shaped Frank vacancy loops to SFTs is observed in atomistic simulations. SFTs are energetically more stable than Frank vacancy loops based on formation energy calculations in pure face-centered cubic Fe, which may explain why vacancy-type Frank loops are infrequently observed experimentally. The motivation for this study is to develop a framework for the stability and nature of irradiation defects as a function of temperature, which will be subsequently used for more complex (alloy) systems.

Evolution of dislocation loops and effect of annealing temperature on hydrogen-ion-implanted Fe-based binary alloys

Reduced activation ferritic/martensitic (RAFM) steels have been considered as a family of prime candidate structural materials for fusion reactors due to low radioactivity and good resistance to irradiation swelling. Various types of defects such as dislocation loops can form in these materials during irradiation. Effects of alloying elements in iron on the formation and migration of dislocation loops have been widely investigated. However, most studies dealt with interstitial-type dislocation loops in iron alloys, while very few focused on vacancy-type dislocation loops. Previous high voltage electron microscope (HVEM) studies from the authors' group have shown that interstitial loops are fully eliminated in hydrogen-ion-implanted α-Fe at 500 ℃, only vacancy loops remain and can achieve up to 100 nm in size. The addition of Ni in α-Fe can reduce the formation temperature of vacancy-type dislocation loops (&lt;i&gt;T&lt;/i&gt;&lt;sub&gt;c&lt;/sub&gt;) to ~450 ℃, while the addition of Cr can increase the temperature to above 600 ℃. However, these experiments are usually difficult to perform due to the scarce resource of HVEM facilities. In this work, in-situ observations by conventional transmission electron microscope (CTEM, 200 kV) are systematically carried out on the hydrogen-ion-implanted α-Fe and Fe-based binary alloys (Fe-3wt.%Cr, Fe-1.4wt.%Ni and Fe-1.4wt.%Mn). The evolutions of morphology and average size of dislocation loops under different annealing temperatures are investigated. The formation temperatures of vacancy-type dislocation loops are determined from the change of average loop size with annealing temperature. The results are consistent with previous studies by HVEM. The effect of Mn atoms in α-Fe is similar to that of Cr atoms, which leads to &lt;i&gt;T&lt;/i&gt;&lt;sub&gt;c&lt;/sub&gt; increase, and the addition of Ni in α-Fe can reduce &lt;i&gt;T&lt;/i&gt;&lt;sub&gt;c&lt;/sub&gt;. Furthermore, the results of D thermal desorption spectrum analysis show that &lt;i&gt;T&lt;/i&gt;&lt;sub&gt;c&lt;/sub&gt; is affected by the binding and release process of hydrogen isotopes to vacancies in α-Fe. Alloying element Ni promotes the binding and release of hydrogen isotopes to vacancies, which leads to&lt;i&gt; T&lt;/i&gt;&lt;sub&gt;c&lt;/sub&gt; decrease. Cr and Mn inhibit the binding and release of hydrogen isotopes to vacancies, causing &lt;i&gt; T&lt;/i&gt;&lt;sub&gt;c&lt;/sub&gt; to increase.

Atomistic study on the dynamic response of helium bubbles to high-energy displacement cascades in tungsten

In a fusion reactor, plasma-facing materials simultaneously suffer high-flux low-energy helium/hydrogen irradiation and high-energy neutron bombardment. Numerous modeling studies have focused on radiation damage caused by individual particles, but ignored their synergistic effects. In this study, molecular dynamics simulations were performed to investigate the cascade damage near a pre-existing helium bubble in tungsten. The existence of helium bubbles generally inhibits defect generation but enhances interstitial-type dislocation loop formation. Furthermore, the distance between the primary knock-on atom and the bubble plays an important role in defect evolution. This corresponds to the defect distribution characteristics of the cascade damage in the perfect crystal. In addition, the change in morphology of helium bubbles during cascade damage was considered. Cascade-induced ballistic dissolution of helium from bubbles dominates helium ejection, but damage-assisted thermal dissolution is also observed. This work can contribute to understanding complex damage mechanisms in multiple irradiation fields, such as the fusion environment.

Iron nitride, α″-Fe16N2, around interstitial type dislocation loops in neutron-irradiated iron

ABSTRACT Atom probe tomography (APT) and selected area electron diffraction (SAED) by transmission electron microscopy (TEM) were applied to nano-scale precipitates formed in a neutron-irradiated iron (99.99% pure) at high temperatures. In the conventional TEM, we discovered donut-type precipitates on the {100} planes that formed at the dilatational side of interstitial type dislocation loops with the Burgers vector of <100> on the habit planes of {100} in the bcc iron lattice. The precipitates were identified as nitride, bct α″-Fe16N2, by chemical composition and lattice structural analyses using APT and SAED. Image contrasts of the α″-Fe16N2 in a weak beam dark-field electron microscopy study were carefully analysed with the diffraction vector and the sign of deviation parameter from the Bragg condition, and it was concluded to be a selective visualisation of α″-Fe16N2 on one of the {100} planes. These results were in agreement with the crystallographic orientation between the α″-Fe16N2 and matrix iron, thereby bridging a knowledge gap in α″-Fe16N2 formation during neutron irradiation using a sodium-cooled experimental fast reactor, JOYO, or a 14 MeV D-T neutron source, RTNS-II.

Defect dynamics in [formula omitted]-U, Mo, and their alloys

Defect dynamics constitutes the foundation for describing microstructural evolution in any material systems for nuclear applications, including body-centered cubic γ-U, Mo, and their alloys. However, defect properties and evolution, and the impact of a large atomic size mismatch between U and Mo atoms on defect dynamics have not been elucidated. In this work, we use molecular dynamics to extensively examine composition-dependent defect behavior in U-Mo alloys and the pure metals. It has been found that point defect migration is strongly correlated and mediated by minor atoms via preferential paths in alloys. Interstitial dumbbells migrate three-dimensionally through the major atoms with a preferred 〈110〉 configuration. Vacancies are less mobile than interstitials, but become comparable (one order of magnitude difference in diffusivity) in U-rich systems. Overall, compared with the pure metals, defect diffusivity can be tuned up or down based on the alloy composition. Finally, interstitial clustering is found to be unfavorable in U-rich systems, as opposed to Mo which exhibits an efficient formation of interstitial-type dislocation loop with 1D diffusion mode. These findings not only provide necessary input to high-fidelity meso-scale simulations of microstructural evolution in these systems, but also have important implications towards explaining radiation effects influenced by the dimensionality and rates of defect diffusion.

An in-situ study of one-dimensional motion of interstitial-type dislocation loops in hydrogen-ion-implanted aluminum

核聚变堆材料在高能粒子辐照过程中会产生大量点缺陷，导致辐照脆性和辐照肿胀等现象。因而，研究点缺陷在辐照过程中的演变过程至关重要。点缺陷团簇的一维迁移现象是这种演变过程的主要研究内容之一。本文采用普通低压（200 kV）透射电镜，在室温条件下对注氢纯铝中的间隙型位错环在电子辐照下的一维迁移现象进行了观察和分析。在200 keV电子辐照下，注氢纯铝中的位错环可多个、同时发生一维迁移运动，也可单个、独立进行一维迁移运动。位错环沿柏氏矢量?111的方向可进行微米尺度的一维长程迁移，沿柏氏矢量?110的方向一维迁移也可达数百纳米。电子束辐照时产生的间隙原子浓度梯度是引起位错环一维迁移并决定其迁移方向的原因。位错环发生快速一维迁移时，其后会留下一条运动轨迹；位错环一维迁移的速率越快，运动的轨迹则越长，在完成迁移过后的几十秒内这些运动轨迹会逐渐消失。

Coupled cluster-dislocation dynamics of irradiation-induced defects

We present a new formulation for modeling concurrent radiation damage and plasticity of irradiated materials during service. We consider the case of irradiated zirconium as an example, where the dominant microstructure features are vacancy and interstitial type dislocation loops. We first develop Cluster Dynamics (CD) equations that describe the temporal evolution of both types of loops, taking into account their nucleation and growth. The size distribution is described, where interstitial loops agglomerate mainly by single-step absorption of a few mobile defect species. Vacancy loops are nucleated heterogeneously at collision cascades sites by collapse of loose vacancy clusters into Stacking Fault Tetrahedra (SFT) that transform into sessile vacancy loops on the basal planes within a dose of a few displacements per atom (dpa). A quasi steady-state for fast mobile point defects is quickly reached, while the nucleation and further growth of dislocation loops continues in a slow transient to high radiation dose. The results are consistent with experimental observations of dislocation loop density, mean size and size distribution. CD equations are then used to establish a discrete representation of the dislocation microstructure and point defect radiation-induced super-saturation conditions in finite grains of Zr as function of irradiation dose through statistical sampling of the size distribution. Discrete Dislocation Dynamics (DDD) simulations involving glide motion of dislocations are performed to determine the mechanical state of deformation and the collective behavior of dislocation loops in post-irradiated materials. Results of DDD simulations show that spatial correlation of loops, their reorientation, and their dose-dependent evolution are in qualitative agreement with experimental observations.

Effect of pulse irradiation on the evolution of damage structure

Charged particle beams from accelerators driven by radiofrequency or with beam scanners possess time structures (i.e., pulse beam operations). However, it is not clear whether the irradiation effects of these pulse beam operations are the same as those of continuous beams, although the average damage rate is the same in both cases. In this study, the difference between continuous beams and pulse beams, and the effect of the repetition frequency and pulse duration on the defect accumulation are evaluated by the rate equation analysis for Al and Fe. The growth rate of interstitial type dislocation loops in Al is simulated with fixed duty ratios (constant pulse duration over the irradiation period) at 453 K. The effect of pulse duration on a constant period of 1 s is simulated in Fe, changing the pulse duration from 1 to 10−6 s, while the damage (dpa) caused by one pulse remains the same. The results are analyzed and the effects of pulse duration and repetition rate on the damage structure evolution are demonstrated in relation to materials parameters.

Neutron irradiation-induced microstructure damage in ultra-high temperature ceramic TiC

Titanium carbide (TiC) is an ultra-high temperature ceramic with potential as a structural material candidate for advanced reactor concepts. However, the irradiation tolerance of TiC is not well understood. Here, we reveal the key irradiation damage microstructure degradation processes in TiC using mixed spectrum neutron irradiations at dose of ∼2 displacements per atom (dpa) at temperatures of ∼220, 620, and 1115 °C, combined with state-of-art microstructure characterization using transmission electron microscopy (TEM) and high-resolution TEM (HRTEM). The helium injection (∼65 atomic parts per million) produced by neutron transmutation also occurred in the sample. TiC was observed to form irradiation-induced interstitial-type dislocation loops and He-stabilized cavities. At 220 and 620 °C, the analysis of the electron diffraction patterns, rel-rod imaging and HRTEM revealed that the dislocation loops were faulted Frank loops with Burgers vector bFrank = 1/3<111> lying on {111} planes. A detailed Burgers vector identification performed by the g.b technique revealed that the dislocation loops forming at 1115 °C were unfaulted, edge-type, with Burgers vectors b = a<100> and a/2<110> with corresponding {100} and {110} habit planes. Using continuum mechanics, we estimated the critical radius at which a faulted dislocation loop transitions to a perfect loop to be 9 nm. Further, no amorphization occurred in TiC under-investigated irradiation conditions while macroscopic swelling under point defect swelling regime was observed. He-stabilized cavities were detected at 1115 °C, but not at lower temperature irradiation. These observations indicate the onset of long-range vacancy migration in TiC occurs between 620 and 1115 °C.

Lattice disorder and N elemental segregation in ion implanted GaN epilayer

The lattice disorder and elemental distribution in He-implanted GaN epilayer at a fluence of 4 × 1016 He+/cm2 at 450 °C have been investigated by a combination of conventional, high resolution transmission electron microscopy and electron energy loss spectroscopy. High resolution transmission electron microscopy shows interstitial-type dislocation loops attached to He bubbles. The nature of dislocation loops and stacking faults are carefully investigated by high resolution transmission electron microscopy. Electron energy loss spectroscopy shows the deficit of N atoms and enriched Ga nanocrystals formed in the damage region, which is comparable with the previous reports. In addition, the chemical environment of nitrogen upon He+ implantation is first investigated in the present experimental results. The research results will be used for understanding ion irradiation-induced damage in GaN.

Evolution of microstructures and hardening property of initial irradiated, post-irradiation annealed and re-irradiated Chinese-type low-Cu reactor pressure vessel steel

Investigation of microstructures and hardening property of initial irradiated, post-irradiation annealed, and re-irradiated Chinese-type low-Cu (0.01 wt%) reactor pressure vessel (RPV) steel were conducted by slow positron beam (SPB), TEM, and nanoindentation. Results of the SPB measurements indicated that a large number density of open volume defects such as vacancies, vacancy clusters, vacancy-solute/H complexes, and dislocation loops were introduced in both initial-irradiated and re-irradiated specimens. The TEM results indicated that interstitial-type dislocation loops with a number density of ∼1022 m−3 were formed and the mean size (∼3 nm) of the dislocation loops was almost unchanged in both initial-irradiated and re-irradiated specimens. The open volume defects such as vacancy-type defects and dislocation loops were almost annealed out and some stable defects still existed in the post-irradiation annealed specimen. The nanoindentation results identified that the obvious hardening phenomena were found in the initial irradiated, post-irradiation annealed and re-irradiated specimens. The quantitative analysis suggested that irradiation hardening of Chinese-type low-Cu RPV steel were attributed to matrix defects including large-size vacancy clusters, vacancy-solute/H complexes, and dislocation loops produced during initial irradiation and re-irradiation. On the other hand, the existed stable defects such as solute-atom (Mn, Ni, and Si) complexes/precipitates, which cannot be distinguished by SPB or TEM, might be responsible for the rest fraction of irradiation hardening of highly initial irradiated and re-irradiated Chinese-type low-Cu RPV steel.

Irradiation damage in xenon-irradiated α-Al2O3 before and after annealing

The irradiation damage build-up of α-Al2O3 under Xe20+ ion irradiation has been investigated by a combination of Raman spectroscopy and transmission electron microscopy. α-Al2O3 crystalline was irradiated with 5 MeV Xe20+ ions to fluences of 1 × 1014 cm−2, 5 × 1014 cm−2, 1 × 1015 cm−2 and 5 × 1015 cm−2 at room temperature. No amorphous phase was formed under the experimental condition. The Raman intensities of feature peaks of Al2O3 decrease after Xe ion irradiation. The interstitial-type dislocation loops with Burgers vectors of b = 1/3 < 10-11> on the {10-10} and (0001) habit planes were found. The formation of basal and prism dislocation loops is related to the lattice damage and position. After annealing, the Raman intensities of feature peaks of Al2O3 increases with annealing temperature. With annealing at 1500℃ for 30 min, lattice defects were completely annealed out in the near surface region. Meanwhile, long dislocations and facet cavities on long dislocations were found in the Xe deposition region. Some lattice defects beyond the projected region were found due to the diffusion toward deep region during thermal annealing.

Effect of irradiation and irradiation defects on the mobility of [formula omitted] symmetric tilt grain boundaries in iron: An atomistic study

Body center cubic iron materials are commonly used in nuclear power plants. In bcc-iron, symmetric tilt grain boundaries (STGBs), which are believed to play important roles on self-healing, are susceptible to configuration and structural change under irradiation. The effect of such changes on mechanical properties of such GBs is still unknown. In this work, using molecular dynamics simulations, we find that the critical shear stress τc required for Σ5 STGB migration is greatly reduced by either displacement cascade nearby or absorption of defect clusters. Moreover, we find that the trapping of radiation-induced interstitial-type dislocation loops could also decrease τc, indicating that displacement cascades could also exert a long-range effect on GBs due to the fast diffusion of interstitial clusters.

Characterization of defect clusters in ion-irradiated tungsten by X-Ray diffuse scattering

Defect clusters in ion-irradiated tungsten are characterized by X-ray diffuse scattering. Single crystal tungsten samples are irradiated at 300 K by 0.5–5 MeV copper ions with total doses of 0.2, 0.6, 2, and 5 displacements per atom (DPA). Analysis of the diffuse scattering intensities allows the determination of the concentration of vacancy- and interstitial-type dislocation loops as a function of loop radius for each sample. More vacancies than interstitials are observed in each sample, and the radius of the vacancy loops increases with irradiation dose. For the 0.2DPA sample the average vacancy loop radius is 10.1±3.0Å with a defect density (counting all atoms in the loops) of (10.2±1.5)×10−4 per tungsten atom, increasing to an average loop radius of 16.1±3.1Å at a density of (8.5±1.0)×10−4 per tungsten atom for the 5DPA sample.

Evolution of irradiation defects in Ti2AlC ceramics during heavy ion irradiation

The evolution of irradiation defects in Ti2AlC under Kr ion irradiation at room and elevated temperatures has been investigated. The dependence of defect size and density on the irradiation dose and temperature was characterized by in-situ transmission electron microscopy (TEM) irradiation experiments. At 25–800 °C, interstitial-type dislocation loops with several nanometers in diameter form on the basal plane of Ti2AlC mostly with a Burgers vector of 1/2[0 0 0 1], and no voids have been observed. The number density of dislocation loops approaches saturation at a level of about 2 dpa at 800 °C. Ti2AlC samples remain crystalline in structure without amorphization or phase transformation. However, the localized disorder of atomic positions on the basal plane of the irradiated Ti2AlC was revealed by high-resolution transmission electron microscopy (HRTEM) studies, which increases with irradiation dose while decreases with elevated temperatures.