The preferential trapping of interstitials at dislocations

Heald, P. T.

doi:10.1080/14786437508226537

Cited by 123 publications

(23 citation statements)

References 17 publications

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“…The intend was to obtain rough estimates; nevertheless, Foreman concluded that the bias was larger than the empirical estimate. Shortly thereafter, Heald [19] employed the embedding approach and used the solution of Ham [20] in the form presented by Margvelashvili and Saralidze [21]. We shall return to this solution below, as it is the only analytical one known.…”

Section: Introductionmentioning

confidence: 99%

The Dislocation Bias

Wolfer

2007

J Computer-Aided Mater Des

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Bias factors for dislocations quantify the preferential diffusion of self-interstitials relative to the diffusion of vacancies to dislocations. These parameters are essential for rate theory computer codes that model nucleation and growth of voids, helium bubbles, dislocation loops, and the evolution of microstructures in materials subject to high dose irradiations. Compact formulae for these factors are derived and the accuracy of several approximations is explored.

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Section: Introductionmentioning

confidence: 99%

The Dislocation Bias

Wolfer

2007

J Computer-Aided Mater Des

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“…There is extensive literature on the dynamics of accumulation of defects in microstructure and particularly on the evaluation of dislocation bias factors [6,[12][13][14][15][16][17], which are then used as input parameters for meanfield rate theory models describing the dynamics of growth of voids in materials exposed to irradiation [9][10][11]18]. These mean-field models do not require, and do not evaluate, the macroscopic stresses and strains resulting from the accumulation of defects.…”

Section: Introductionmentioning

confidence: 99%

A multi-scale model for stresses, strains and swelling of reactor components under irradiation

et al. 2018

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Predicting strains, stresses and swelling in nuclear power plant components exposed to irradiation directly from the observed or computed defect and dislocation microstructure is a fundamental problem of fusion power plant design that has so far eluded a practical solution. We develop a model, free from parameters not accessible to direct evaluation or observation, that is able to provide estimates for irradiation-induced stresses and strains on a macroscopic scale, using information about the distribution of radiation defects produced by high-energy neutrons in the microstructure of materials. The model exploits the fact that elasticity equations involve no characteristic spatial scale, and hence admit a mathematical treatment that is an extension to that developed for the evaluation of elastic fields of defects on the nanoscale. In the analysis given below we use, as input, the radiation defect structure data derived from ab initio density functional calculations and large-scale molecular dynamics simulations of high-energy collision cascades. We show that strains, stresses and swelling can be evaluated using either integral equations, where the source function is given by the density of relaxation volumes of defects, or they can be computed from heterogeneous partial differential equations for the components of the stress tensor, where the density of body forces is proportional to the gradient of the density of relaxation volumes of defects. We perform a case study where strains and stresses are evaluated analytically and exactly, and develop a general finite element method implementation of the method, applicable to a broad range of predictive simulations of strains and stresses induced by irradiation in materials and components of any geometry in fission or fusion nuclear power plants.

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“…A greater number of interstitials than vacancies are expected to migrate to dislocations [3,4]. However, the lack of evidence for climb of mixed and edge dislocations in the ternary alloys, or for the nucleation of interstitial loops along dislocations, may suggest that irradiation-generated interstitials are core (or "pipe") diffusing after the annihilation of pre-existing jogs.…”

Section: Quaternary Alloymentioning

confidence: 95%

“…A striking example of loop nucleation, after low dose irradiation, on individual partials is shown In figure 1. At higher doses subsequent interaction (described fully in the paper of Clicrns et ai [2]) of the decorated dislocation with the osmotic climb force, which arises from an expected high super saturation of intcrstitials at dislocations [3,4], was believed responsible for the creation of extremely complex post-irradiation configurations.…”

Section: Copper-aluminium Alloysmentioning

confidence: 99%

The Interaction of HVEM-Generated Point-Defects with Dissociated Dislocations

King¹,

Jenkins²,

Kirk³

et al. 1992

MSF

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This paper describes experiments we have performed to investigate the mechanisms of interaction of IIVEM-gcncralcd point-defects with dissociated dislocations in a scries of austcniticFc-Ni-Cr alloys and reviews earlier work in Cu-AI alloys and in Ag. In the Fc-Ni-Cr alloys interstitial climb was observed only at favourable sites such as pre-existing jogs, whilst vacancies clustered near dislocations to form stacking-fault tetrahedra. These observations arc similar to these in Ag; the complex climb mechanisms seen in Cu-AI alloys were not found. The differences bet ween materials is believed to be due to differences in the case of interstitial pipe diffusion. INTRODUCTIONThe interaction of radiation-produced point defects witli dislocations is central to the understanding of phenomena which give rise to dimensional or mechanical-property changes in nuclear reactors. As part of a more general effort to understand microslructural evolution under irradiation, studies of the mechanisms of interaction of IIVEM-gcncralcd point-defects with dissociated dislocations have been carried out in Oxford, Harwell and Argonnc over the past 11 years. Various model materials have been chosen, starling with very low slacking-faull energy (SFE) CuAl alloys with wide partial separations and progressing towards the structural steels of practical importance. These studies have demonstrated that dislocation climb in these materials may be far from straightforward. In this paper we briefly review early studies in CuAl alloys and in Ag and present some of our more recent results in Fc-Ni-Cr alloys, The observations in the different materials will be compared and contrasted and possible origins for differences and similarities between materials will be discussed.

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The preferential trapping of interstitials at dislocations

Cited by 123 publications

References 17 publications

The Dislocation Bias

The Dislocation Bias

A multi-scale model for stresses, strains and swelling of reactor components under irradiation

The Interaction of HVEM-Generated Point-Defects with Dissociated Dislocations

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