Size Of Defect Clusters Research Articles

Swift Au+9 ion irradiation-induced defects and alloy complexity effect on the mechanical hardness of NiCoCrFePd HEA and NiCoCrFe MEA

The outstanding radiation damage stability of an NiCoCrFePd high entropy alloy (HEA) as compared to conventional alloys poses the question for the mechanism of an ion–matter interaction. The positron annihilation lifetime spectroscopic and TEM (transmission electron microscopic) measurements are implemented to trace different kinds of defects produced by 120 MeV Au+9 ion irradiation and their evolution as a function of ion fluence. The variation of lifetimes and corresponding intensities with the ion fluence indicates the formation of dislocation-type defects at a lower ion fluence and vacancy clusters at a higher ion fluence caused by coalescence or agglomeration of dislocation defects. Formation of different types of defects in turn modulates the strain development inside the crystal. Additionally, the HR-TEM investigation of NiCoCrFePd HEA also exhibits the formation of dislocation and vacancy clusters with the average size of vacancy clusters increases from ∼2.9 ± 0.1 to ∼3.8 ± 0.1 nm with the increases in the ion fluence. Surprisingly, the average defect cluster size in NiCoCrFePd HEA is suppressed compared to NiCoCrFe MEA, thereby showing the enhanced radiation stability on Pd incorporation due to the high defect recombination caused by reduced thermal conductivity and high lattice distortion. Nano-indentation measurement shows that the radiation hardening behavior of the NiCoCrFePd HEA responded slowly owing to its damage suppression property as compared to the NiCoCrFe MEA. Additionally, softening behavior also appeared at an early fluence in NiCoCrFe MEA compared to the NiCoCrFePd HEA signifying its excellent resistance to defect accumulation.

Displacement cascade and defect-driven simulations in V-Ti-Ta-Nb high-entropy alloy

High-entropy alloys (HEA) have attracted considerable attention in the development of nuclear materials due to their excellent properties. In this study, the primary irradiation damage and long-term defect evolution of pure V, V-5Ti-5Ta conventional alloy, V-Ti-Ta MEA and V-Ti-Ta-Nb HEA were simulated by molecular dynamics (MD) to understand the irradiation resistance mechanism of these four systems. The primary irradiation damage simulation results indicate that the V-Ti-Ta MEA and V-Ti-Ta-Nb HEA exhibit a delayed thermal peak, a longer defect recombination time and a slightly higher number of final Frenkel pairs (FPs) than pure V and V-5Ti-5Ta alloy. Both primary irradiation damage and long-time defect evolution results show that V-Ti-Ta MEA and V-Ti-Ta-Nb HEA have lower defect clustering fraction, cluster size and dislocation loop size than pure V and V-5Ti-5Ta. This is because V-Ti-Ta MEA and V-Ti-Ta-Nb HEA exhibit lower dislocation loop binding energy and defect mobility compared to pure V and V-5Ti-5Ta alloy. This study investigates the reasons for the better radiation resistance of V-Ti-Ta MEA and V-Ti-Ta-Nb HEA than pure V and V-5Ti-5Ta conventional alloys.

Defects Act in an “Introverted” Manner in FeNiCrCoCu High-Entropy Alloy under Primary Damage

High-entropy alloys (HEAs) attract much attention as possible radiation-resistant materials due to their several unique properties. In this work, the generation and evolution of the radiation damage response of an FeNiCrCoCu HEA and bulk Ni in the early stages were explored using molecular dynamics (MD). The design, concerned with investigating the irradiation tolerance of the FeNiCrCoCu HEA, encompassed the following: (1) The FeNiCrCoCu HEA structure was obtained through a hybrid method that combined Monte Carlo (MC) and MD vs. the random distribution of atoms. (2) Displacement cascades caused by different primary knock-on atom (PKA) energy levels (500 to 5000 eV) of the FeNiCrCoCu HEA vs. bulk Ni were simulated. There was almost no element segregation in bulk FeNiCrCoCu obtained with the MD/MC method by analyzing the Warren–Cowley short-range order (SRO) parameters. In this case, the atom distribution was similar to the random structure that was selected as a substrate to conduct the damage cascade process. A mass of defects (interstitials and vacancies) was generated primarily by PKA departure. The number of adatoms grew, which slightly roughened the surface, and the defects were distributed deeper as the PKA energy increased for both pure Ni and the FeNiCrCoCu HEA. At the time of thermal spike, one fascinating phenomenon occurred where the number of Frenkel pairs for HEA was more than that for pure Ni. However, we obtained the opposite result, that fewer Frenkel pairs survived in the HEA than in pure Ni in the final state of the damage cascade. The number and size of defect clusters grew with increasing PKA energy levels for both materials. Defects were suppressed in the HEA; that is to say, defects were “cowards”, behaving in an introverted manner according to the anthropomorphic rhetorical method.

Effect of temperature on the anisotropic distribution of defects in self-ion irradiated α-SiC: In-situ TEM investigation

As a structural material suitable for high temperatures and with excellent performance, silicon carbide (SiC) has great potential for application in nuclear systems. However, irradiation-induced anisotropic distribution of defects in hexagonal α-SiC will severely affect its mechanical properties. The effect of temperature on the anisotropic distribution of defects and associated mechanisms is not well understood. In this work, the formation and growth of defects in α-6H–SiC at different temperatures during self-ion irradiation was investigated by in-situ transmission electron microscopy (TEM), and the effect of temperature on the anisotropic distribution of defects was explored. Irradiation temperature significantly affects the average size and number density of defect clusters, the critical irradiation damage at which defect clusters appear, and the growth rate of defect clusters. Moreover, irradiation temperature had a significant effect on the anisotropic distribution of defects in α-SiC and exhibited a critical temperature effect. The possible mechanism dominating the temperature-dependent effects on the number density of defects and anisotropy was discussed, and the effect of recovery rate on the generation of defects was explored.

Local chemical ordering and its impact on radiation damage behavior of multi-principal element alloys

• LCOs promotes the nucleation of dislocations but inhibits their growth in the CrFeNi MPEA under irradiation. • LCOs reduces the number of stair-rod dislocation associated with SFTs, which may help to suppress irradiation swelling. • LCOs is conducive to enhance radiation damage tolerance of MPEAs. Multi-principal element alloys (MPEAs) have attracted much attention as future nuclear materials due to their extraordinary radiation resistances. In this work, we have elucidated the development of local chemical orderings (LCOs) and their influences on radiation damage behavior in the typical CrFeNi MPEA by hybrid-molecular dynamics and Monte Carlo simulations. It was found that considerable LCOs consisting of the Cr-Cr and Ni-Fe short-range orders existed in the ordered configuration with optimized system energy. Through modeling the accumulation cascades up to 1000 recoils, we revealed that the size of defect clusters and dislocation loops is smaller in the ordered configuration than those in the random one, although the former formed more Frenkel pairs (i.e., self-interstitials and vacancies). In addition, the distribution of dislocation loops is relatively more dispersed in the ordered configuration, and the stair-rod dislocations related to irradiation swelling are also smaller, implying that the existence of LCOs is conducive to enhancing radiation damage tolerance. To understand the underlying mechanism, the effects of LCOs on the formation and evolution of defects and radiation resistance were discussed from the aspects of atomic bonding, migration path, and energy of defect diffusion, which provides theoretical guidance for the design of MPEAs with enhanced radiation resistance.

Revealing the governing factors for long-term radiation damage evolution in multi-principal elemental alloys through atomistically-informed cluster dynamics

Multi-principal elemental alloys (MPEAs) exhibit unusual mechanical properties and irradiation resistance. It has been reported that the irradiation tolerance of MPEAs originates from the long-term defect evolution stage rather than the ballistic collision phase. However, understanding the long-term evolution of MPEAs is exceptionally challenging due to intrinsic chemical disorder. In this work, we apply atomistically-informed cluster dynamics to study defect evolution in MPEAs for the first time, which enables us to elucidate the critical features responsible for the irradiation performance of MPEAs. With suitable irradiation parameters, our obtained cluster size distributions are in good agreement with experiments. We further discuss the influences of defect migration energy heterogeneity, cluster geometry, and temperature on the produced defect cluster forms. Our results suggest that the small clusters during the collision phase in MPEAs suppress the formation of large-sized defect clusters in the long term. Contrary to the common belief, a broader migration energy distribution due to chemical disorder cannot reduce the maximal defect cluster size but increase the proportion of small clusters. Our results thus reveal that the major role of chemical complexity in MPEAs is to modify the growth mechanism of clusters by promoting small cluster formation and suppressing overall defect diffusivities.

Insights from an OKMC simulation of dose rate effects on the irradiated microstructure of RPV model alloys

This work studies the defect features in a dilute FeMnNi alloy by an Object Kinetic Monte Carlo (OKMC) model based on the "grey-alloy" method. The dose rate effect is studied at 573 K in a wide range of dose rates from 10−8 to 10−4 displacement per atom (dpa)/s and demonstrates that the density of defect clusters rises while the average size of defect clusters decreases with increasing dose rate. However, the dose-rate effect decreases with increasing irradiation dose. The model considered two realistic mechanisms for producing <100>-type self-interstitial atom (SIA) loops and gave reasonable production ratios compared with experimental results. Our simulation shows that the proportion of <100>-type SIA loops could change obviously with the dose rate, influencing hardening prediction for various dose rates irradiation. We also investigated ways to compensate for the dose rate effect. The simulation results verified that about a 100 K temperature shift at a high dose rate of 1 × 10−4 dpa/s could produce similar irradiation microstructures to a lower dose rate of 1 × 10−7 dpa/s irradiation, including matrix defects and deduced solute migration events. The work brings new insight into the OKMC modeling and the dose rate effect of the Fe-based alloys.

Hardening and ductility loss of RAFM steel CLF-1 irradiated with high-energy heavy ions

Irradiation induced hardening and ductility loss of a Chinese reduced-activation ferritic / martensitic steel CLF-1 is evaluated by utilizing the high-energy heavy ion irradiation and the subsequent miniature specimen test technique. Specimens were irradiated with high energy 14N6+ and 56Fe17+ ions supplied by a cyclotron to generate a quasi uniform atomic displacement damage plateau from surface to a depth about 25 μm. Three successively increasing damage levels (0.05, 0.1 and 0.2 dpa) were approached at a low temperature around 223 K. The small punch test at room temperature showed that the ductility loss can be observed at the lowest damage level and increases with the increasing damage level, similar to the variation trend of yield stress with irradiation dose. The TEM observation revealed that the average size and number density of irradiation-induced defect clusters also increased with displacement damage. Assuming that the hardening is caused by the self-interstitial-type dislocation loops, an estimation by combining the numerical calculation of the number density / mean diameter of dislocation loops with the dispersion barrier hardening model yields a similar dose dependence of the irradiation hardening.

The influence of ultrasonic waves on the characteristics of the radiation detector

The paper presents the results of a study of the field dependences of charge losses, silicon diffusion detectors that allow identifying single atoms and their clusters acting as charge carrier capture centers. In electric fields E &gt; 1470 V/cm, there is an effective decrease in the capture processes, which manifests itself in the form of straightening of the field dependences of charge losses and in the narrowing of spectral lines. Effects - an ultrasonic field with an intensity of I * = 0.4 W / cm2 and a frequency of f = 15 MHz, on samples for a time t £ 45 minutes, leads to an increase in the amplitude of the signal. It is established that ultrasonic exposure leads to a decrease in charge loss, drift time and an improvement in the energy resolution of detectors. It should also be noted that fluctuations of the impurity relief in the Si-p-type volume are smoothed out after low-power ultrasonic waves pass through it for a long time, as evidenced by a significant decrease in the effective size of local defect clusters.

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.

Simulations of primary damage in a High Entropy Alloy: Probing enhanced radiation resistance

High Entropy Alloys (HEA) attract attention as possible radiation resistant materials, a feature observed in some experiments that has been attributed to several unique properties of HEA, in particular to the disorder-induced reduced thermal conductivity and to the peculiar defect properties originating from the chemical complexity. To explore the origin of such behavior we study the early stages (less than 0.1 ns), of radiation damage response of a HEA using molecular dynamics simulations of collision cascades induced by primary knock-on atoms (PKA) with 10, 20 and 40 keV, at room temperature, on an idealized model equiatomic quinary fcc FeNiCrCoCu alloy, the corresponding “Average Atom” (AA) material, and on pure Ni. We include accurate corrections to describe short-range atomic interactions during the cascade. In all cases the average number of defects in the HEA is lower than for pure Ni, which has been previously used to help claiming that HEA is radiation resistant. However, simulated defect evolution during primary damage, including the number of surviving Frenkel Pairs, and the defect cluster size distributions are nearly the same in all cases, within our statistical uncertainty. The number of surviving FP in the alloy is predicted fairly well by analytical models of defect production in pure materials. All of this indicates that the origin of radiation resistance in HEAs as observed in experiments may not be related to a reduction in primary damage due to chemical disorder, but is probably caused by longer-time defect evolution.

Effects of alloying elements Mn, Mo, Ti, Si, P and C on the incubation period of void swelling in austenitic stainless steels

Void swelling, which induces the degradation of the original properties of nuclear materials under high-energy particle-irradiation, is an important problem. The incubation period, a transient stage before the steady void growth, determines the duration of service of nuclear materials. Several experimental studies have been performed on void observations by transmission electron microscopy (TEM), which, however, has a resolution limit for the size of defect clusters. Positron annihilation lifetime spectroscopy (PALS) enables the detection of small vacancy clusters, single vacancies, dislocations and precipitates. The use of these two methods provides complementary information toward detecting defect information in the incubation period. Here, defect structures during the incubation period in austenitic stainless steels, by means of PALS and TEM are reviewed. The role of alloying elements into determining the period is explained. Furthermore, the existing problems and research directions in this field are presented.

Models and regressions to describe primary damage in silicon carbide

Silicon carbide (SiC) and SiC/SiC composites are important candidate materials for use in the nuclear industry. Coarse grain models are the only tools capable of modelling defect accumulation under different irradiation conditions at a realistic time and length scale. The core of any such model is the so-called “source term”, which is described by the primary damage. In the present work, classical molecular dynamics (MD), binary collision approximation (BCA) and NRT model are applied to describe collision cascades in 3C-SiC with primary knock-on atom (PKA) energy in the range 1–100 keV. As such, BCA and NRT are benchmarked against MD. Particular care was taken to account for electronic stopping and the use of a threshold displacement energy consistent with density functional theory and experiment. Models and regressions are developed to characterize the primary damage in terms of number of stable Frenkel pairs and their cluster size distribution, anti-sites, and defect type. As such, an accurate cascade database is developed with simple descriptors. One of the main results shows that the defect cluster size distribution follows the geometric distribution rather than a power law.

A mechanism-based quantitative multi-scale framework for investigating irradiation hardening of tungsten at low temperature

We present the development and validation of a mechanism-based multi-scale modeling framework to quantitatively link the irradiation defect evolution kinetics at the microscopic scale to the macroscopic yield strength and flow stress evolution of tungsten irradiated at low temperature (0.08 < T/Tm < 0.18, where Tm is the melting point and is equal to 3673 K). The mechanism-based strength model, proposed as the superposition of thermal softening and modified dispersed barrier hardening, is developed to understand the underlying strengthening mechanisms. The thermal softening exponent as defined in the Johnson-Cook model is obtained by fitting unirradiated yield strength at different temperatures. A set of irradiation-induced defect kinetics equations with the thermodynamics parameters derived from the atomic calculations reported in the literature is used to determine the densities and sizes of defect clusters at the meso-scale, and the predicted irradiation damage characteristics are in reasonable quantitative agreement with experimental data from literature. The effects of irradiation condition (temperature and irradiation dose) and test temperature on yield strength are quantitatively predicted and compared with experimental measurements from the literature. The predicted irradiation hardening and strain hardening are compared with experimental data as model validation.

An Incremental Model for Defect Production upon Cascade Overlapping**Supported by the National Key Research and Development Program of China (No. 2017YFB0702201) and the National Natural Science Foundation of China (No. 51571129).

An analytic incremental model is proposed to predict the defect production upon cascade overlapping. By resolving the coupled annealing events during cascade overlapping, this model handles cascade overlapping with multiple pre-existing defects of different sizes and number densities. The model is first parameterized and then applied to bcc-Fe. The proposed model satisfyingly reproduces the defect production obtained by molecular dynamics simulations with various radiation damage levels and defect cluster size distributions. The present model provides an essential description of the primary source of radiation damage, especially for high dose irradiation, and could be used in conjunction with reactive diffusion models for better understanding of radiation damage.

Relaxation volumes of microscopic and mesoscopic irradiation-induced defects in tungsten

The low-energy structures of irradiation-induced defects in materials have been studied extensively over several decades, as these determine the available modes by which a defect can diffuse or relax, and how the microstructure of an irradiated material evolves as a function of temperature and time. Consequently, many studies concern the relative energies of possible defect structures, and empirical potentials are commonly fitted to or evaluated with respect to these. But recently [S. L. Dudarev et al., Nucl. Fusion 58, 126002 (2018)], we have shown that other parameters of defects not directly related to defect energies, namely, their elastic dipole tensors and relaxation volumes, determine the stresses, strains, and swelling of reactor components under irradiation. These elastic properties of defects have received comparatively little attention. In this study, we compute relaxation volumes of irradiation-induced defects in tungsten using empirical potentials and compare to density functional theory results. Different empirical potentials give different results, but some clear potential-independent trends can be identified. We show that the relaxation volume of a small defect cluster can be predicted to within 10% from its point-defect count. For larger defect clusters, we provide empirical fits as a function of defect cluster size. We demonstrate that the relaxation volume associated with a single primary-damage cascade can be estimated from the primary knock-on atom energy. We conclude that while annihilation of defects invariably reduces the total relaxation volume of the cascade debris, there is still no conclusive verdict about whether coalescence of defects reduces or increases the total relaxation volume.

Scaling laws and stability of nano-sized defect clusters in niobium via atomistic simulations and statistical analysis

The predictions of scaling laws for the structure and properties of defect clusters are generally limited to small defect clusters in their ground-state configurations. We investigated the size and geometrical configuration dependence of nano-sized defect clusters in niobium (Nb) using molecular dynamics. We studied the structure and stability of large clusters of size up to fifty defects for vacancies and one hundred defects for interstitials, as well as the role of helium and metastable configurations on the stability of these clusters. We compared three different interatomic potentials in order to determine the relative stability of these clusters as a function of their size and geometrical configurations. Additionally, we conducted a statistical analysis to predict the formation and binding energies of interstitial clusters as a function of both their size and configuration. We find that the size dependence of vacancy and interstitial clusters can be approximated by functional forms that account for bulk and surface effects as well as some considerations of elastic interactions. We also find that helium and metastable configurations can make vacancy and interstitial clusters thermally stable depending on the configuration. Our parameterized functional forms for the formation and binding energies are valid for a very broad range of defect sizes and configurations making it possible to be used directly in a coarse-grained modeling strategy such as Monte Carlo, cluster dynamics or dislocation dynamics which look at defect accumulation and evolution in microstructures.

Molecular dynamics study of the material property changes induced by accumulated point defects in graphite

Molecular dynamics simulations were performed on pristine graphite using an adaptive intermolecular reactive empirical bond-order (AIREBO) potential, which introduced various concentrations of point defects and different sizes in defect clusters. Induced by the accumulation of mono-vacancy and self-interstitial atoms (SIA), the changes to the geometry, density, Young’s modulus, fracture strength, and coefficient of thermal expansion (CTE) of graphite were simulated over a range of defect concentrations and cluster sizes. We found that accumulated Frenkel pairs lead to clearly-observed changes to the geometric and elastic properties in graphite crystal, but a relatively negligible effect on CTE. As expected, changing size of defect clusters with a given concentration relates to the variation of material properties. The property changes simulated in the present work are qualitatively consistent with the experimental observations. The results are expected to provide some insights into the link between cascade-induced point defects and experimentally irradiation-assisted material property changes.

Defect evolution in Ni and NiCoCr by in situ 2.8 MeV Au irradiation

The evolution of radiation-induced defects in Ni and the single-phase concentrated solid solution alloy, NiCoCr, were investigated during in situ 2.8 MeV Au ion irradiation and post-irradiation analysis using transmission electron microscopy. Compared to Ni, both the size and area density of defect clusters decreased in NiCoCr under the same irradiation conditions, suggesting that the chemical complexity, i.e., randomness of lattice site occupations, of NiCoCr suppressed radiation-induced damage. One-dimensional glide of defect clusters was observed in Ni but not in the NiCoCr alloy. The structural nature of small defect clusters in NiCoCr were further investigated using high-angle annular dark field scanning transmission electron microscopy.

Defects and solarization in YAG transparent ceramics

Transparent ceramics are emerging as future materials for lasers, scintillation, and illumination. In this paper, an interesting and surprising phenomenon in YAG transparent ceramics is reported. UV light leads to significant changes in the microstructure of open volume defects and nano clusters as well as in the optical properties. Light-induced lattice relaxation is suggested as the mechanism behind this intriguing behavior. The complex F-type color center with broad absorption bands is caused by the aliovalent sintering additives (Ca2+/Mg2+) and Fe ion impurities. Two individual peaks in the thermoluminescence spectra illustrate both shallow and deep level traps. From positron annihilation lifetime data, vacancy clusters and nanovoids are detected and characterized, although these free-volume defects could not be observed by high-resolution transmission electron microscopy. The solarization induced by UV irradiation is associated with a change in the structure and size of defect clusters due to lattice relaxation. Therefore, this work shows how UV irradiation leads not only to a change in the charge state of defects, but also to a permanent change in defect structure and size. It significantly affects the optical properties of YAG ceramics and their performance in lasers and other optical applications. These results are crucial for advancing transparent ceramics technology.