Nanoscale analysis of ion irradiated ODS 14YWT ferritic alloy

In this work, the nanoscale microstructure of an advanced oxide dispersion strengthened (ODS) 14YWT ferritic alloy (SM13 heat) with nominal composition Fe–14Cr–3W-0.4Ti-0.3Y2O3 (wt. %) has been characterized by atom probe tomography (APT) before and after ion irradiation with 70 MeV Fe9+ ions at 450 °C to a total dose of 21 dpa. A detailed solute cluster analysis of APT data reveals that, in the manufacturing process, larger nanoparticles form in or close to the grain boundaries respective to those inside grains. The evolution of the nanoparticles after irradiation seems to be related to their location, as a higher increase in the number density and in the Y:Ti ratio is observed for the nanoparticles in or close to grain boundaries. APT analysis also shows Cr, W and C segregation to grain boundaries enhanced by the irradiation. A previous study of this same alloy before and after irradiation reports that the mechanical properties do not seem to be affected, but the microstructure was not investigated to confirm. The present work confirms little microstructural evolution after irradiation in this 14YWT alloy, indicating tolerance at the given irradiation conditions

TEM of neutron, proton and self-ion irradiation damage in FeCr alloys

In the absence of a high-flux fusion-neutron irradiation source, the microstructural and mechanical changes expected within materials exposed to a nuclear-fusion environment must be replicated by fission-neutron and other surrogate-particle irradiations. This study uses transmission electron microscopy (TEM) to compare the microstructural defects produced in FeCr alloys during exposure to neutrons, protons, and self-ions. Alloys of Fe6Cr and Fe9Cr were irradiated using fission-neutrons, 2.0MeV Fe+ ions and 1.2MeV protons at similar temperatures (~300C) and similar doses (~2.0dpa). The neutron-irradiated alloys contained a population of interstitial dislocation loops with b= (>70%) and b=. The visible dislocation loops were on average ~5nm in size, and the density varied from 2±1 x1014cm-3 in the matrix to 1.2±0.3 x1017cm-3 close to helical dislocation lines. Dislocations loops were mostly clustered around sub-grain boundaries and helical-dislocations. Helical-dislocations formed from initially straight screw dislocations experiencing radial-climb in response to a vacancy-biased defect flux. Small chromium clusters were identified in the neutron-irradiated Fe6Cr, and chromium α’-phase precipitates were identified in the Fe9Cr. Self-ion irradiation produced mostly homogeneously distributed dislocation loops (6-7nm on average), but with a greater fraction of loops (~40%) than was seen in the neutron-irradiated alloys. The self-ion irradiated Fe6Cr and Fe9Cr contained only vacancy-type loops, unlike the neutron or proton irradiated sample which contained only interstitial loops. Chromium remained in solution in both ion-irradiated samples. Proton-irradiated Fe9Cr contained dislocation loops close to helical-dislocation segments, similar to the neutron-irradiated sample. Chromium α’-phases were also identified. The proton-irradiated Fe6Cr contained much larger loops (~13nm on average) than the neutron or ion-irradiated alloys, and chromium was shown to have segregated on and around these loops. Both proton-irradiated alloys contained large voids (>4nm and up to 12nm) at a density greater than 1016cm-3. In the neutron and ion-irradiated alloys, voids were mostly <2nm

Correlations of Microstructural Features between Neutron and Self-ion Irradiated MA957

Nuclear energy is a sustainable, efficient, carbon-neutral energy source to meet the ever-increasing global energy demands. Some current advanced reactor concepts require reactor core components, such as fuel cladding and ducts, to withstand up to 300–500 dpa neutron irradiation. Achieving such a dose level in test reactors would require years or even decades of radiation, which makes reactor irradiation impractical for materials development and testing. As an alternative approach to study high-dose irradiation effects in materials, ion irradiation techniques are regaining popularity due to their high damage production rate. To date, the correspondence between neutron and ion irradiation is still being investigated, and correlations of microstructural features between neutron and ion irradiated materials are still not well understood. As a means to understand these correlations, the microstructures of neutron- and ion-irradiated steel alloy MA957 were examined and compared. This material was selected due to its high relevance to high-dose applications. The characterization relies heavily on atom probe tomography (APT). The results show that YTiO oxide particles in MA957 followed similar trends with irradiation temperature after neutron and ion irradiation but size and number density were not identical. The presence of alpha-prime phase has been confirmed at lower irradiation temperatures in both neutron and ion irradiations. Assessing effects on grain boundary segregation after ion irradiation is more challenging. Thus, only qualitative comparisons of segregation of Cr, Ti, and TiO were conducted between neutron and ion irradiation. Rate theory calculations were carried out to investigate precipitate stability. The defect balance equation and precipitate stability models were reanalyzed using materials-specific and experiment-specific parameters. Precipitate nucleation was not included in the analysis, and so size change is more readily predicted with this model than number density change. Predicted results of particle stability were found to be in good agreement with our experimental results. The ratio of diffusivity to dose rate was identified as a probable measure to determine whether application of temperature shift is necessary for reproducing microstructural evolution in ion irradiation. The dose rate dependence has been reanalyzed and temperature ranges where precipitate growth is independent of dose rate have been identified for future investigations

Heavy ion irradiation as a proxy to neutron irradiation in impure Fe and Fe-Cr alloys

The realisation of nuclear fusion energy will require materials that can withstand high doses of neutron irradiation, and one class of candidate materials for structural applications is reduced-activation ferritic-martensitic (RAFM) steels containing around 9% Cr. The present lack of a fusion neutron source has meant that much of the research into radiation damage has been conducted using specimens irradiated in fission test reactors, or by using heavy ion irradiation as a simulation method. This work explores details of the use of this emulation technique, alongside the effect of composition and impurities on radiation damage in these materials. Beam rastering is often employed during heavy ion irradiation in order to deliver a more uniform dose across the target area in comparison to continuous irradiation with a defocused beam, but this practice has been reported to alter the radiation-induced microstructural changes. This work investigated this in Fe and Fe-Cr alloys, using nanoindentation. No difference in post-irradiation hardness was observed between samples irradiated with a rastered or defocussed beam. A direct comparison of the microstructure and mechanical properties of heavy ion and neutron irradiated Fe-9Cr samples was made, and it was found that the segregation to dislocation loops and clustering of impurity elements Si and P reported in the neutron irradiated specimen was also present in the ion irradiated specimen. The α' precipitation reported after neutron irradiation did not occur after ion irradiation. The post-irradiation increase in hardness and yield stress of both samples were found to be the same, measured by nanoindentation and microcantilever testing. An increase in hardness of 0:88 ± 0:21 and 0:81 ± 0:26 GPa was found in the ion and neutron irradiated specimens respectively. The increase in hardness for the neutron irradiated specimen was found to be less then the 1:5 ± 0:28 GPa reported for an identically irradiated Fe-6Cr sample. The impact of carbon concentration on irradiation hardening in Fe was investigated by ion irradiating a high purity Fe sample containing regions of varying carbon concentration. Two samples were irradiated at two different dose rates, but no dose rate effect was observed. The increase in hardness however was found to increase with increasing carbon concentration, with a maximum increase of around 3 times that observed in pure Fe. This work has identified some limitations of and future challenges for the use of heavy ion irradiation as a surrogate to neutron irradiation in the absence of a dedicated fusion materials testing facility

Microstructural evolution and transmutation in tungsten under ion and neutron irradiation

This study aims to compare the effects of neutron and self-ion irradiation on the mechanical properties and microstructural evolution in W. Neutron irradiation at the HFR reactor to 1.67 dpa at 800 °C resulted in the formation of large Re and Os rich clusters and voids. The post-irradiation composition was measured using APT and verfified against FISPACT modelling. The measured Re and Os concentration was used to create alloys with equivalent concentrations of Re and Os. These alloys were exposed to self-ion irradiation to a peak dose of 1.7 dpa at 800 °C. APT showed that self-ion irradiation leads to the formation of small Os clusters, wheras under neutron irradiation large Re/Os clusters form. Voids are formed by both ion and neutron irradiation, but the voids formed by neutron irradiation are larger. By comparing the behaviour of W-1.4Re and W-1.4Re-0.1Os, suppression of Re cluster formation was observed. Irradiation hardening was measured using nanoindentation and was found to be 2.7 GPa, after neutron irradiation and 1.6 GPa and 0.6 GPa for the self-ion irradiated W-1.4Re and W-1.4Re-0.1Os. The higher hardening is attributed to the barrier strength of large voids and Re/Os clusters that are observed after neutron irradiation

Cluster dynamics modeling of Mn-Ni-Si precipitates in ferritic-martensitic steel under irradiation

Mn-Ni-Si precipitates (MNSPs) are known to be responsible for irradiation-induced hardening and embrittlement in structural alloys used in nuclear reactors. Studies have shown that precipitation of the MNSPs in 9-Cr ferritic-martensitic (F-M) alloys, such as T91, is strongly associated with heterogeneous nucleation on dislocations, coupled with radiation-induced solute segregation to these sinks. Therefore it is important to develop advanced predictive models for Mn-Ni-Si precipitation in F-M alloys under irradiation based on an understanding of the underlying mechanisms. Here we use a cluster dynamics model, which includes multiple effects of dislocations, to study the evolution of MNSPs in a commercial F-M alloy T91. The model predictions are calibrated by data from proton irradiation experiments at 400 {\deg}C. Radiation induced solute segregation at dislocations is evaluated by a continuum model that is integrated into the cluster dynamics simulations, including the effects of dislocations as heterogeneous nucleation sites. The result shows that MNSPs in T91 are primarily irradiation-induced and, in particular, both heterogeneous nucleation and radiation-induced segregation at dislocations are necessary to rationalize the experimental observations

Microstructural evolution and transmutation in tungsten under ion and neutron irradiation

This study aims to compare the effects of neutron and self-ion irradiation on the mechanical properties and microstructural evolution in W. Neutron irradiation at the HFR reactor to 1.67 dpa at 800 °C resulted in the formation of large Re and Os rich clusters and voids. The post-irradiation composition was measured using APT and verfified against FISPACT modelling. The measured Re and Os concentration was used to create alloys with equivalent concentrations of Re and Os. These alloys were exposed to self-ion irradiation to a peak dose of 1.7 dpa at 800 °C. APT showed that self-ion irradiation leads to the formation of small Os clusters, wheras under neutron irradiation large Re/Os clusters form. Voids are formed by both ion and neutron irradiation, but the voids formed by neutron irradiation are larger. By comparing the behaviour of W-1.4Re and W-1.4Re-0.1Os, suppression of Re cluster formation was observed. Irradiation hardening was measured using nanoindentation and was found to be 2.7 GPa, after neutron irradiation and 1.6 GPa and 0.6 GPa for the self-ion irradiated W-1.4Re and W-1.4Re-0.1Os. The higher hardening is attributed to the barrier strength of large voids and Re/Os clusters that are observed after neutron irradiation

Atom probe field ion microscopy and related topics: A bibliography 1989

This bibliography includes references related to the following topics: atom probe field ion microscopy (APFIM), field ion spectroscopy (FIM), field emission microscopy (FEM), liquid metal ion sources (LMIS), scanning tunneling microscopy (STM), and theory. Technique-orientated studies and applications are included. This bibliography covers the period 1989. The references contained in this document were compiled from a variety of sources including computer searches and personal lists of publications

\u3cem\u3eIn Situ\u3c/em\u3e TEM Micropillar Compression Testing in Irradiated Oxide Dispersion Strengthened Alloys

The objective of this study is to determine the validity of in situ transmission electron microscopy (TEM) micro-compression of pillars in as received and ion-irradiated Fe-9%Cr oxide dispersion strengthened (ODS) alloy. The growing role of charged particle irradiation in the evaluation of nuclear reactor candidate materials requires the development of novel methods to assess mechanical properties in near-surface irradiation damage layers just a few micrometers thick. In situ TEM mechanical testing is one such promising method, yet size effects must be understood to validate the technique. In this work, a micro-compression pillar fabrication method is developed. Yield strengths measured directly from TEM in situ compression tests are within expected values, and are consistent with predictions based on the irradiated microstructure. Measured elastic modulus values, once adjusted for deformation and deflection in the base material, are also within the expected range. A pillar size effect is only observed in samples with minimum dimension ≤ 100 nm due to the low inter-obstacle spacing in the as received and irradiated material. By comparing the microstructural obstacle spacing with specimen dimensions, size effects can be understood and TEM in situ micropillar compression tests can be used to quantitatively determine mechanical properties of shallow ion-irradiated layers