Atomistic simulation of interaction of collision cascade with different types of grain boundaries in α-Fe (original) (raw)
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Modelling and Simulation in Materials Science and Engineering, 2013
He atoms introduced to materials may lead them to intergranular fracture, and understanding such an effect is a key issue for the design of future fusion reactors. In the present study, we investigated the decrease of grain boundary (GB) strength caused by He segregation at several kinds of α-Fe GBs by exploiting the first principles calculations and a set of empirical potentials. We found enough evidence to support the notion that the GB cohesive energy, a critical measure of GB strength, approximately scales with the He concentration at the GB surface, regardless of the GB type.
Journal of Nuclear Materials, 2006
The primary state of damage obtained in molecular dynamics (MD) simulations of displacement cascades in α-Fe, particularly the fraction of point-defects in clusters, depends on the interatomic potential used to describe the atomic interactions. The differences may influence the microstructural evolution predicted in damage accumulation models which use results from MD cascade simulations as input. In this work, a number of displacement cascades of energy ranging from 5 to 40 keV have been simulated using the same procedure with four different interatomic potentials for α-Fe, each of them providing, among other things, varying descriptions of self-interstitial atoms (SIA) in this metal. The behaviour of the cascades at their different phases and the final surviving defect population have been studied and compared applying the same cascade analysis criteria for all potentials. The outcome is discussed trying to identify the characteristics of the potential that have the largest influence on the predicted primary state of damage.
Nuclear Materials and Energy, 2021
This paper is the first of three that overview the main mechanisms that drive the microstructure evolution in Fe alloys under irradiation. It focuses on pure α-Fe and compiles the parameters that describe quantitatively the mobility and stability of point-defects and especially their clusters, including possible reactions and criteria to decide when they should react. These parameters are the result of several years of calculations and application in microstructure evolution models. They are mainly collected from the literature and the parameter choice tries to reconcile different sets of values that, while being in general qualitatively similar, are often quantitatively not coincident. A few calculation results are presented here for the first time to support specific approximations concerning defect properties or features. Since calculations cannot cover all possible defect configurations, the definition of these parameters often requires educated guesses to fill knowledge gaps. These guesses are here listed and discussed whenever relevant. This is therefore a "hands-on" paper that: (i) collects in a single report most microstructure evolution parameters that are found in the literature for irradiated α-Fe, including a discussion of the most important mechanisms at play based on current knowledge; (ii) selects a ready-to-use set that can be employed in microstructure evolution models, such as those based on object kinetic Monte Carlo (OKMC) methods. This work also identifies parameters that are needed, but not known, hopefully prompting corresponding calculations in the future.
Atomistic Simulations of Nanoindentation Response of Irradiation Defects in Iron
Sains Malaysiana, 2019
Radiation response of a material is a consequence of defects' evolution in any radiation damage event. The radiationinduced defects can significantly alter the mechanical properties of a material. Radiation damage initiates from incident neutron by bombardment on solid material causing production and evolution of Frenkel defects. Since voids are formed due to aggregation of a large number of vacancies that cause dimensional changes and hence irradiation-induced swelling. In order to characterize the effect of irradiation defects, we have performed molecular dynamics (MD) simulations to investigate nanoindentation response of point defects and voids in Fe and their effects on mechanical parameters. The radial effect of voids and their interaction mechanism is also explored by nanoindentation simulation. It has been found that most of the dislocation produced are <111> and <100> during nanoindentation in all simulated models. There will be an increase in dislocation density which will harden the material and reduce its toughness. The mechanical parameters such as hardness H and reduced elastic modulus E r of irradiation defects are calculated from P-h curves. It
Molecular dynamics simulation of displacement cascades in α-Fe: A critical review
Journal of Nuclear Materials, 2006
An embedded atom method (EAM) empirical potential recently fitted and validated for Fe-Cr systems is used to simulate displacement cascades up to 15 keV in Fe and Fe-10%Cr. The evolution of these cascades up to thermalisation of the primary damage state is followed and quantitatively analysed. Particular attention is devoted to assessing the effect of Cr atoms on the defect distribution versus pure Fe. Using the Wigner-Seitz cell criterion to identify point defects, first results show that the main effect of the presence of Cr in the system is the preferential formation of mixed Fe-Cr dumbbells and mixed interstitial clusters, with expected lower mobility than in pure Fe.
Journal of Nuclear Materials, 2002
During irradiation, mobile defects, defect clusters and impurity atoms segregate on dislocations. When an external stress is applied, plastic flow is initiated when dislocations are unlocked from segregated defects. Sustained plasticity is achieved by continuation of dislocation motion, overcoming local forces due to dispersed defects and impurities. The phenomena of flow localization, post-yield hardening or softening and jerky flow are controlled by dislocation-defect interactions. We review here computational methods for investigations of the dynamics of dislocation-defect interactions. The influence of dislocations on the motion of glissile self-interstitial atoms (SIAs) and their clusters is explored by a combination of kinetic Monte Carlo and dislocation dynamics. We show that dislocation decoration by SIAs is a result of their 1-D motion and rotation as they approach dislocation cores. The interaction between dislocations and immobilized SIA clusters indicates that the unlocking mechanism is dictated by shape instabilities. Finally, computer simulations for the interaction between freed dislocations and stacking fault tetrahedra in irradiated Cu, and between dislocations and microvoids in irradiated iron are presented, and the results show good agreement with experimental observations.
MD modeling of defects in Fe and their interactions
Journal of Nuclear Materials, 2003
Ferritic/martensitic steels considered as candidate first-wall materials for fusion reactors experience significant radiation hardening at temperatures below $400°C. A number of experimental studies in ferritic alloys, performed at higher temperatures, have shown the existence of large interstitial loops with Burgers vector 1 2 h1 1 1i and h1 0 0i in the bulk, which may provide a significant contribution to the hardening caused during irradiation at lower temperatures. Hardening arises from a high number density of loops, voids and small precipitates, which pin system dislocations, impeding their free glide. In this work, we review the nature of the different interstitial dislocation loops observed in a-Fe and ferritic materials, assess the effect of substitutional impurities on migrating 1 2 h1 1 1i clusters, and apply atomistic modeling to investigate the mechanisms of formation and growth of h1 0 0i loops from smaller cascade-produced 1 2 h1 1 1i clusters. The proposed mechanism reconciles experimental observations with continuum elasticity theory and recent MD modeling of defect production in displacement cascades. In addition, the interaction of screw dislocations, known to control the low-temperature plastic response of b.c.c. materials to external stress, with h1 0 0i dislocation loops is investigated with MD, where the main physical mechanisms are identified, cutting angles estimated and a first-order estimation of the induced hardening is provided.
Communications Materials
Dislocation loops are ubiquitous in irradiated materials, and dislocation loop bias plays a critical role in void swelling. However, due to complicated interactions between dislocation loops and point defects, it is challenging to evaluate the bias factors of dislocation loops. Here, we determine the bias of sessile < 100 > loops in α-iron using a recently developed atomistic approach based on the lifetime of point defects. We establish a mechanistic understanding of the loop interaction based on the diffusion tendency of point defects near the loop core region. Mobile self-interstitial atoms tend to be absorbed from the edge of the loop, and a trapping region perpendicular to the habit plane of the loop exists. The dislocation loop bias is found to be substantially lower than those of straight dislocations in α-iron and should be included in swelling rate estimates. With the obtained sink strength and bias values, agreement is achieved with experimental results for both absol...
MD simulation of atomic displacement cascades in Fe–10at.%Cr binary alloy
Journal of Nuclear Materials, 2009
The present paper is devoted to radiation damage simulation of Fe-9at.%Cr binary alloy with twin grain boundaries (GBs) by the molecular dynamics method. Evaluations of specific energy of five GBs and sizes of corresponding GB regions have been obtained for iron and FeCr alloy at temperatures of 0 and 300 K. The binding energies of the vacancy, self-interstitial atom (SIA) and substitutional Cr atom to the GB in pure Fe have been estimated. The results showed that GB regions are energetically preferable for the point defects. Interaction of 10 keV displacement cascades with the GBs has been studied. The tendency to accumulate at the GB region has been shown for produced defects. Some quantitative results which describe features of radiation damage nearby the GB have been obtained. It is revealed that Cr fraction in SIAs inside the GB region is slightly lower than that in the initial alloy matrix. Cr fraction in interstitial configurations outside the GB region is almost three times as high. However, no remarkable chromium redistribution nearby the GB has been detected.
Modeling dislocation evolution in irradiated alloys
Metallurgical Transactions A, 1990
Neutron irradiation of structural materials leads to such observable changes as creep and void swelling. These effects are due to differential partitioning of point defects. Although most radiationproduced point defects recombine with an antidefect, a very small fraction of the defects survives. The surviving defect fraction is directly related to the density and type of extended defects that act as point defect sinks. Defect partitioning requires the presence of more than one type of sink and that at least one of the sinks has a capture efficiency for either vacancies or interstitials that is different from that of the other sink(s). For example, dislocations provide the interstitial "bias" that drives swelling, and the ratio of the dislocation to cavity sink strength determines the swelling rate. These sink strengths change during irradiation, and an explicit model of their evolution is required to simulate swelling or creep. Such a model has been developed; the influences of various model assumptions and parameters are discussed. The model simulates the evolution of Frank faulted interstitial loops, providing a dislocation source and the glide/climb of the dislocation network leading to annihilation of dislocation segments. Good agreement is found between model predictions and experimental data. Swelling simulations are shown to be quite sensitive to the dislocation model.