Helical dislocations: Observation of vacancy defect bias of screw dislocations in neutron irradiated Fe–9Cr

J.C Haley,F Liu,E Tarleton,A.C.F Cocks,G.R Odette,S Lozano-Perez,S.G Roberts

doi:10.1016/j.actamat.2019.09.031

Abstract

We have analysed the microstructure of a model alloy of Fe9Cr irradiated with neutrons to a dose of 1.6 dpa at 325 °C. Helical dislocations comprise a major part of the damage; these formed from the interaction of pre-existing screw dislocations with irradiation-induced defects. We have investigated the process behind how these helices form, and how they cause local clustering of dislocation loops. Specifically, we have shown experimentally that the interaction of vacancy defects with pre-existing screw dislocations causes the formation of mixed screw-edge helical dislocations. Interstitials and vacancies were generated in equal numbers, which shows that the screw dislocations must have acted as vacancy-biased sinks.Helical dislocations in general were analysed from a theoretical perspective, and three Dimensional Discrete Dislocation Dynamics (3D-DDD) was used to develop a model for the formation and growth of a vacancy-fed helical dislocation.Since the helical dislocations cause the removal of vacancies from the local microstructure, this leaves a higher supersaturation of interstitials close to the dislocations. We argue that this supersaturation is responsible for enhanced interstitial loop coarsening, leading to a higher proportion of visible interstitial clusters in the vicinity of helical dislocations. These findings offer a new perspective on how dislocations affect the spatial homogeneity of radiation damage.

Highlights

Ferritic-martensitic (FM) steels with chromium content close to 9 at% are prime candidates for use as radiation-tolerant structural materials in both Gen-IV fission and fusion reactors
Irradiation at temperatures below 500 °C leads to hardening, loss of ductility, and an increase in the brittle to ductile transition temperature [3,4]. These macroscale mechanical changes are caused by the defects that form during irradiation; neutron irradiation inside a fission or fusion reactor can change the microstructure of FM steels through the production of dislocation loops, alpha-prime precipitates, voids and gas bubbles, and segregation of alloying elements, including solute defect cluster complexes [1,2,3,4,5,6,7]
Some black-dot damage was visible outside the helix, this was mostly indistinguishable from Focussed Ion Beam (FIB) damage or oxide

Summary

Introduction

Ferritic-martensitic (FM) steels with chromium content close to 9 at% are prime candidates for use as radiation-tolerant structural materials in both Gen-IV fission and fusion reactors. Irradiation at temperatures below 500 °C leads to hardening, loss of ductility, and an increase in the brittle to ductile transition temperature [3,4] These macroscale mechanical changes are caused by the defects that form during irradiation; neutron irradiation inside a fission or fusion reactor can change the microstructure of FM steels through the production of dislocation loops, alpha-prime precipitates, voids and gas bubbles, and segregation of alloying elements, including solute defect cluster complexes [1,2,3,4,5,6,7]. While the phenomenon appears to be dose-dependent, the residual dislocations and grain structure were considered by English [13] to be pre-requisites for such heterogeneous distributions

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