Size Of Defect Clusters Research Articles

Investigation of interactions between defect clusters in stainless steels by in situ irradiation at elevated temperatures

In recent years, modeling studies on the interactions between defect clusters have been extensively conducted to predict the behavior of stainless steels in reactors. However, comparable experimental results are desired to validate existing results and provide guidelines for future modeling. The size of defect clusters is expected to be a key factor influencing cluster interactions. Thus, in this work, the effects of growing clusters on their surrounding microstructure were quantitatively examined by in situ transmission electron microscopy during ion irradiation. To this end, a high-purity 316 L stainless steel model alloy was irradiated by 2 MeV Fe2+ to 0.2 dpa at 400 °C and 300 °C separately. At 300 °C, the number density of defect clusters monotonically increased with increasing fluence, whereas at 400 °C, the number density decreased almost immediately after the rapid nucleation regime. This decrease could be explained by the distinct growth of some clusters, which suppressed the nucleation of clusters around them as well as the lifetime of neighboring tiny clusters. Large interstitial-type clusters approximately 6–11 nm in size could be absorbed by neighboring interstitial-type clusters of similar sizes via interstitial crowdion diffusion. The absorption would not occur until both clusters grew large enough to permit a <110> diffusion path between them.

Ion irradiation effects on FeCrAl-ODS ferritic steel

The correlation between microstructure and mechanical property of ion irradiated 12Cr-6Al ODS ferritic steel was studied. Ion irradiation experiments were performed using 10.5 MeV Fe ions up to the nominal displacement damage of 20 dpa with the damage rate of 1 × 10−4 dpa/s, while the irradiation temperature was 300 °C. Oxide nanoparticles showed stable size distribution and mean size under ion irradiation up to 20 dpa. The irradiation microstructure examined by TEM revealed that the mean size and number densities of irradiation-induced defect clusters increased with the displacement damage. The correlation between irradiation microstructure and radiation hardening was theoretically calculated using the dispersed barrier hardening model. The results showed a good agreement with the experimentally measured hardness data up to irradiation at 5 dpa, while a slight discrepancy was found between theoretical and experimental hardness values under irradiation at 20 dpa. Radiation hardening in 12Cr-6Al ODS ferritic steel was mainly caused by irradiation-induced defect clusters below the irradiation dose of 5 dpa. As the irradiation dose increased toward 20 dpa, an additional influence of the radiation appeared, which was assumed to be induced by α’ phase transformation.

Molecular dynamics simulations of the coupled effects of strain and temperature on displacement cascades in α-zirconium

Abstract In this article, we conducted molecular dynamics simulations to investigate the effect of applied strain and temperature on irradiation-induced damage in alpha-zirconium. Cascade simulations were performed with primary knock-on atom energies ranging between 1 and 20 KeV, hydrostatic and uniaxial strain values ranging from −2% (compression) to 2% (tensile), and temperatures ranging from 100 to 1000 K. Results demonstrated that the number of defects increased when the displacement cascade proceeded under tensile uniaxial hydrostatic strain. In contrast, compressive strain states tended to decrease the defect production rate as compared with the reference no-strain condition. The proportions of vacancy and interstitial clustering increased by approximately 45% and 55% and 25% and 32% for 2% hydrostatic and uniaxial strain systems, respectively, as compared with the unstrained system, whereas both strain fields resulted in a 15–30% decrease in vacancy and interstitial clustering under compressive conditions. Tensile strains, specifically hydrostatic strain, tended to produce larger sized vacancy and interstitial clusters, whereas compressive strain systems did not significantly affect the size of defect clusters as compared with the reference no-strain condition. The influence of the strain system on radiation damage became more significant at lower temperatures because of less annealing than in higher temperature systems.

Radiation response of nanotwinned Cu under multiple-collision cascades

In this paper, multiple collision cascades (MCC) of nanotwinned (nt) Cu with three different twin spacings are performed to model the response of nt Cu upon a radiation dose of 1 displacements per atom (dpa). Considering the defects developed with high randomness in the material during a MCC process, each MCC in a nt Cu is conducted for eight times. This enables us to analyze some average properties of defect clusters in the radiated nt Cu with different twin spacings at the different radiation doses. We also analyze the microstructural evolution in the nt Cu during the MCC. Smaller size of defect clusters and lower defect density are seen in the nt Cu with smaller twin spacing. In addition, a number of defect clusters could be removed via their frequent interactions with the coherent twin boundaries (CTBs) during the MCC. This induces either the migration of CTBs or the healing of CTBs. Moreover, the potential formation and elimination mechanisms of stacking fault are found to be due to the climb of Frank partial dislocation and glide of Shockley partial dislocations. This study provides further evidence on the irradiation tolerance of CTBs and the self-healing capability of CTBs in response to radiation.

Effect of alloying elements on defect evolution in Ni-20X binary alloys

The effect of alloying elements on radiation-induced microstructural evolution in Ni and Ni-20X (X = Fe, Cr, Mn and Pd) binary alloys was investigated using ion irradiation and cross-sectional transmission electron microscopy. The three-dimensional migration mode is identified to be the dominating migration mechanism for interstitial clusters in these binary alloys, contrary to the one-dimensional mode that dominates in the well-studied dilute alloys. The results reveal that: (1) the average size of defect clusters decreases as the solute atomic volume size factor increases. Smaller void size in Ni-20Cr is attributed to faster vacancy mobility in the near surface region, and weaker vacancy binding energy beyond the irradiation peak than Ni-20Fe. The smaller voids observed in Ni-20Mn and Ni-20Pd beyond the damage peak are due to the stronger Mn/Pd-vacancy binding effect of largely oversized solute atoms. (2) Oversized solutes can act as strong trapping sites for interstitials. The larger the solute atomic volume factor, the stronger the trapping force. This leads to a more significantly sluggish interstitial migration and smaller dislocation loop size. The average dislocation loop size in Ni-20Fe was four times larger than Ni-20Pd (atomic volume factor being 10.6% and 41.3%) but an order of magnitude lower in density. The smaller dislocation loop size in Ni-20Cr is attributed to stronger interstitial binding of Cr-Ni. Overall, the alloying effect on defects is more significant in concentrated binary alloys than in dilute binary alloys, due to the concentration difference of alloying atoms and the interstitial dominant migration mechanisms in the main irradiated region.

Rate Theory for Dislocation Loops Evolution in AL-6XN Austenitic Stainless Steel under Proton Irradiation

The average size and density evolution of dislocation loops in AL-6XN austenitic stainless steel, a candidate fuel cladding material for supercritical water-cooled reactor, under proton irradiation were simulated through a rate theory model. The simulation results exhibit relatively good agreement with the experimental results at 563 K. The size and density of defect clusters are calculated under irradiation temperature between 550 K and 900 K and irradiation doses up to 15 dpa which satisfies the working condition in supercritical water-cooled reactor. The fast nucleation between self-interstitials happens at the initial stage of irradiation. The average size of dislocation loops increases while the average density of these loops reduces with the increasing temperature, and the average density approaches to a constant when irradiated at higher irradiation doses. The mechanism is discussed based on the variation of rate constants of defect reactions and the variation of the diffusion coefficients of interstitials and dislocation loops with dose and temperature.

Characterization of Microstructure and Stability of Precipitation in SIMP Steel Irradiated with Energetic Fe Ions * *Supported by the Young Scientists Fund of the National Natural Science Foundation of China under Grant No 11505246, and the Major Research Plan of the National Natural Science Foundation of China under Grant No 91426301.

A type of home-made reduced activation martensitic steel, high silicon (SIMP) steel, is homogeneously irradiated with energetic Fe ions to the doses of 0.1, 0.25 and 1 displacement per atom (dpa), respectively, at 300°C and 1 dpa, at 400°C. Microstructural changes are investigated in detail by transmission electron microscopy with cross-section technique. Interstitial defects and defect clusters induced by Fe-ion irradiation are observed in all the specimens under different conditions. It is found that with increasing irradiation temperature, size of defect clusters increases while the density drops quickly. The results of element chemical mapping from the STEM images indicate that the Si element enrichment and Ta element depletion occur inside the precipitates in the matrix of SIMP steel irradiated to a dose of 1 dpa at 300°C. Correlations between the microstructure and irradiation conditions are briefly discussed.

Achieving Radiation Tolerance through Non-Equilibrium Grain Boundary Structures

Many methods used to produce nanocrystalline (NC) materials leave behind non-equilibrium grain boundaries (GBs) containing excess free volume and higher energy than their equilibrium counterparts with identical 5 degrees of freedom. Since non-equilibrium GBs have increased amounts of both strain and free volume, these boundaries may act as more efficient sinks for the excess interstitials and vacancies produced in a material under irradiation as compared to equilibrium GBs. The relative sink strengths of equilibrium and non-equilibrium GBs were explored by comparing the behavior of annealed (equilibrium) and as-deposited (non-equilibrium) NC iron films on irradiation. These results were coupled with atomistic simulations to better reveal the underlying processes occurring on timescales too short to capture using in situ TEM. After irradiation, NC iron with non-equilibrium GBs contains both a smaller number density of defect clusters and a smaller average defect cluster size. Simulations showed that excess free volume contribute to a decreased survival rate of point defects in cascades occurring adjacent to the GB and that these boundaries undergo less dramatic changes in structure upon irradiation. These results suggest that non-equilibrium GBs act as more efficient sinks for defects and could be utilized to create more radiation tolerant materials in future.

Dose-rate dependence of damage buildup in 3C-SiC

The influence of the defect generation rate on radiation damage processes in SiC remains poorly understood. Here, we use a combination of ion channeling and transmission electron microscopy to systematically study the dose-rate dependence of damage buildup in 3C-SiC bombarded in the temperature range of 25–200 °C with 500 keV Ar ions. The results reveal a pronounced dose-rate effect, whose magnitude increases close-to-linearly with temperature. When ion dose and temperature are held constant, the dose-rate dependence of the damage level is nonlinear, with saturation at high dose rates. Electron microscopy reveals that the average size of stable defect clusters increases with increasing dose rate. These findings have important implications for understanding and predicting radiation damage in SiC.

Subcascade formation and defect cluster size scaling in high-energy collision events in metals

It has been recently established that the size of the defects created under ion irradiation follows a scaling law (Sand A. E. et al., EPL, 103 (2013) 46003; Yi X. et al., EPL, 110 (2015) 36001). A critical constraint associated with its application to phenomena occurring over a broad range of irradiation conditions is the limitation on the energy of incident particles. Incident neutrons or ions, with energies exceeding a certain energy threshold, produce a complex hierarchy of collision subcascade events, which impedes the use of the defect cluster size scaling law derived for an individual low-energy cascade. By analyzing the statistics of subcascade sizes and energies, we show that defect clustering above threshold energies can be described by a product of two scaling laws, one for the sizes of subcascades and the other for the sizes of defect clusters formed in subcascades. The statistics of subcascade sizes exhibits a transition at a threshold energy, where the subcascade morphology changes from a single domain below the energy threshold, to several or many sub-domains above the threshold. The number of sub-domains then increases in proportion to the primary knock-on atom energy. The model has been validated against direct molecular-dynamics simulations and applied to W, Fe, Be, Zr and sixteen other metals, enabling the prediction of full statistics of defect cluster sizes with no limitation on the energy of cascade events. We find that populations of defect clusters produced by the fragmented high-energy cascades are dominated by individual Frenkel pairs and relatively small defect clusters, whereas the lower-energy non-fragmented cascades produce a greater proportion of large defect clusters.

Microstructural evolution in NF616 (P92) and Fe–9Cr–0.1C-model alloy under heavy ion irradiation

In this comparative study, in situ investigations of the microstructure evolution in a Fe–9Cr ferritic–martensitic steel, NF616, and a Fe–9Cr–0.1C-model alloy with a similar ferritic–martensitic microstructure have been performed. NF616 and Fe–9Cr–0.1C-model alloy were irradiated to high doses (up to ∼10 dpa) with 1 MeV Kr ions between 50 and 673 K. Defect cluster density increased with dose and saturated in both alloys. The average size of defect clusters in NF616 was constant between 50 and 573 K, on the other hand average defect size increased with dose in Fe–9Cr–0.1C-model alloy around ∼1 dpa. At low temperatures (50–298 K), alignment of small defect clusters resulted in the formation of extensive defects in Fe–9Cr–0.1C-model alloy around ∼2–3 dpa, while similar large defects in NF616 started to form at a high temperature of 673 K around ∼5 dpa. Interaction of defect clusters with the lath boundaries were found to be much more noticeable in Fe–9Cr–0.1C-model alloy. Differences in the microstructural evolution of NF616 and Fe–9Cr–0.1C-model alloy are explained by means of the defect cluster trapping by solute atoms which depends on the solute atom concentrations in the alloys.

Transport of charge carriers through the thin base of a heterobipolar transistor under the impact of radiation

The transport of electrons in heterobipolar transistors with radiation defects is studied under conditions where the characteristic sizes of defect clusters and the distances between them can be comparable or can even exceed the sizes of the device base. It is shown that, under some levels of irradiation, neutron radiation can bring about a decrease in the time of flight of hot electrons through the base, which retards the degradation of the transistor parameters.

Long term simulation of point defect cluster size distributions from atomic displacement cascades in Fe70Cr20Ni10

We have used an Object Kinetic Monte Carlo (OKMC) model to simulate the long term evolution of the primary damage in Fe70Cr20Ni10 alloys. The mean number of Frenkel pairs created by different Primary Knocked on Atoms (PKA) was estimated by Molecular Dynamics using a ternary EAM potential developed in the framework of the PERFORM-60 European project. This number was then used to obtain the vacancy–interstitial recombination distance required in the calculation of displacement cascades in the Binary Collision Approximation (BCA) with code MARLOWE (Robinson, 1989). The BCA cascades have been generated in the 10–100keV range with the MARLOWE code and two different screened Coulomb potentials, namely, the Molière approximation to the Thomas–Fermi potential and the so-called “Universal” potential by Ziegler, Biersack and Littmark (ZBL). These cascades have been used as input to the OKMC code LAKIMOCA (Domain et al., 2004), with a set of parameters for describing the mobility of point defect clusters based on ab initio calculations and experimental data. The cluster size distributions have been estimated for irradiation doses of 0.1 and 1dpa, and a dose rate of 10−7dpa/s at 600K. We demonstrate that, like in the case of BCC iron, cluster size distributions in the long term are independent of the cascade energy and that the recursive cascade model suggested for BCC iron in Souidi et al. (2011) also applies to FCC Fe70Cr20Ni10. The results also show that the influence of the BCA potential is sizeable but the qualitative correspondence in the predicted long term evolution is excellent.

FP7 Project LONGLIFE: Overview of results and implications

LONGLIFE (“Treatment of long term irradiation embrittlement effects in RPV safety assessment”) was a collaborative project of the 7th Framework Programme of EURATOM under the umbrella of NULIFE/NUGENIA, aiming at an improved understanding of irradiation effects in reactor pressure vessel steels under conditions representative of long term operation. The LONGLIFE project was completed by the end of January 2014. The paper gives an overview of the main project results and their implications for future research, as discussed at the final project workshop. The microstructural database for neutron-irradiated RPV steels was extended considerably and existing gaps on mechanical property data were closed. Indications of late blooming effects (LBE) were found in some cases, but clear criteria for the occurrence/exclusion in terms of irradiation conditions and chemical composition have still to be developed. The commonly accepted trend, that low flux and low irradiation temperature promotes LBE, is supported. A significant flux effect on the size of defect clusters was observed in two high Cu weld materials, while the changes of mechanical properties are not affected by the neutron flux. The database requires completion in particular for low-Cu RPV steels. The shift of reference temperature T0 over the thickness location of a VVER-440 welding seam does not follow the prediction Russian code, because of the strong variation of the intrinsic weld bead structure. Therefore, the effect of the initial microstructure and of the heterogeneity on the radiation behaviour has to be addressed in future works. Existing embrittlement trend curves models were applied to the LONGLIFE data base. None of the trend curves could predict the behaviour of the entirety of the LONGLIFE materials sufficiently. A guideline for monitoring radiation embrittlement during life extension periods was developed.

Microstructural Evolution of an Ion Irradiated Ni&amp;ndash;Mo&amp;ndash;Cr&amp;ndash;Fe Alloy at Elevated Temperatures

The irradiation behavior of a Ni­Mo­Cr­Fe alloy, of the type currently being considered for use in future molten salt cooled reactors, has been investigated in situ using 1MeV Kr ions at temperatures of 723 and 973K. When irradiated to 5 dpa, experimental observations reveal the instantaneous formation and annihilation of point defect clusters, with such processes attributed to the long range elastic interactions that occur between defects through multiple intra-cascade overlap. Corresponding differences in the defect cluster density and size distribution suggest that changes to the microstructure were dependent upon temperature and dose, affecting the growth, accumulation and mobility of irradiationinduced defect clusters under these conditions. [doi:10.2320/matertrans.MD201311]

Simulation of the nanostructure evolution under irradiation in Fe–C alloys

Neutron irradiation induces in steels nanostructural changes, which are at the origin of the mechanical degradation that these materials experience during operation in nuclear power plants. Some of these effects can be studied by using as model alloy the iron–carbon system.The Object Kinetic Monte Carlo technique has proven capable of simulating in a realistic and quantitatively reliable way a whole irradiation process. We have developed a model for simulating Fe–C systems using a physical description of the properties of vacancy and self-interstitial atom (SIA) clusters, based on a selection of the latest data from atomistic studies and other available experimental and theoretical work from the literature. Based on these data, the effect of carbon on radiation defect evolution has been largely understood in terms of formation of immobile complexes with vacancies that in turn act as traps for SIA clusters. It is found that this effect can be introduced using generic traps for SIA and vacancy clusters, with a binding energy that depends on the size of the clusters, also chosen on the basis on previously performed atomistic studies.The model proved suitable to reproduce the results of low (<350K) temperature neutron irradiation experiments, as well as the corresponding post-irradiation annealing up to 700K, in terms of defect cluster densities and size distribution, when compared to available experimental data from the literature. The use of traps proved instrumental for our model.

In situ Evidence of Defect Cluster Absorption by Grain Boundaries in Kr Ion Irradiated Nanocrystalline Ni

Significant microstructural damage, in the form of defect clusters, typically occurs in metals subjected to heavy ion irradiation. High angle grain boundaries (GBs) have long been postulated as sinks for defect clusters, like dislocation loops. Here, we provide direct evidence, via in situ Kr ion irradiation within a transmission electron microscope, that high angle GBs in nanocrystalline (NC) Ni, with an average grain size of ~55 nm, can effectively absorb irradiation-induced dislocation loops and segments. These high angle GBs significantly reduce the density and size of irradiation-induced defect clusters in NC Ni compared to their bulk counterparts, and thus NC Ni achieves significant enhancement of irradiation tolerance.

Molecular dynamics simulations of irradiation cascades in alpha-zirconium under macroscopic strain

Numerous computer simulation studies have been performed on the radiation damage of zirconium. In contrast to most of the work in the literature which has focused on the effects of temperature and recoil energy on defect production and defect clustering, we have developed a computational model to consider the influence of elastic strain field on the formation of defects and their clusters, as strain is commonly present in a real reactor environment. In this work, irradiation induced displacement cascades in alpha-zirconium experiencing a macroscopic strain have been studied by molecular dynamics (MD) simulations using a many-body interatomic potential. The external strain mainly affects the size of defect clusters rather than the total number of defects. The sizes of interstitial and vacancy clusters respond differently to the external strain conditions.

In situ study of heavy ion induced radiation damage in NF616 (P92) alloy

NF616 is a nominal 9Cr ferritic–martensitic steel that is amongst the primary candidates for cladding and duct applications in the Sodium-Cooled Fast Reactor, one of the Generation IV nuclear energy systems. In this study, an in situ investigation of the microstructure evolution in NF616 under heavy ion irradiation has been conducted. NF616 was irradiated to 8.4dpa at 50K and to 7.6dpa at 473K with 1MeV Kr ions. Nano-sized defects first appeared as white dots in dark-field TEM images and their areal density increased until saturation (∼6dpa). Dynamic observations at 50K and 473K showed appearance and disappearance of TEM-visible defect clusters under irradiation that continued above saturation dose. Quantitative analysis showed no significant change in the average size (∼3–4nm) and distribution of defect clusters with increasing dose at 50K and 473K. These results indicate a cascade-driven process of microstructure evolution under irradiation in these alloys that involves both the formation of TEM-visible defect clusters by various degrees of cascade overlap and cascade induced defect cluster elimination. According to this mechanism, saturation of defect cluster density is reached when the rate of defect cluster formation by overlap is equal to the rate of cluster elimination during irradiation.

On the correlation between primary damage and long-term nanostructural evolution in iron under irradiation

Atomic displacement cascades in solids are complex phenomena, the outcome of which can be statistically characterised by properties such as their spatial extent, morphology and the spatial correlation of defects. Some properties scale in a simple way with parameters such as the cascade energy, others have limited variability with energy, for example point defect cluster size distributions. Taking advantage of the latter invariance, we use object kinetic Monte Carlo simulations to demonstrate that most properties of displacement cascade play no significant role in the evolution of point defect cluster size distributions after long enough time. It is suggested that reliable long-term predictions are possible, when using only the self-interstitial and vacancy cluster size distributions from low energy displacement cascades as building blocks to represent the complete spectrum of cascade energies obtained under neutron irradiation conditions. This is shown on the basis of recursive properties of displacement cascades evidenced for the first time and taking only approximately into account the average volumes in which vacancies and self-interstitial atoms are confined. The model has been successfully used to simulate the evolution of point defect clusters in iron for displacement rates in the range of 10 −6 dpa/s and doses of the order of 0.1 dpa. The applicability beyond this range and to more complex materials is discussed.