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Register a new userNon-destructive study of collision cascade damage in self-ion irradiated tungsten using HR-EBSD and ECCI
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Understanding defect production and evolution under irradiation is a long-standing multi-scale problem. Conventionally, experimental examination of irradiation-induced defects (IIDs) has mainly relied on transmission electron microscopy (TEM), which offers high spatial resolution but requires destructive sample preparation. Furthermore, limited field of view and low strain sensitivity make multi-scale characterisation and quantitative strain measurements difficult. Here we explore the potential of using advanced techniques in the scanning electron microscope (SEM) to non-destructively probe irradiation damage at the surface of bulk materials. Electron channelling contrast imaging (ECCI) is used to image nano-scale irradiation-induced defects in 20 MeV self-ion irradiated tungsten, the main candidate material for fusion reactor armour. The results show an evolution of the damage microstructure from uniformly and randomly distributed nano-scale defects at 0.01 dpa (displacement per atom) to string structures extending over hundreds of nanometres at 1 dpa. Cross-correlation based high-resolution EBSD (HR-EBSD) is used to probe the lattice strain fields associated with IIDs. While there is little strain fluctuation at 0.01 dpa, significant heterogeneity in the lattice strains is observed at 0.1 dpa, increasing with dose until saturation at 0.32 dpa. The characteristic length scale of strain fluctuations is ~500 nm. Together, ECCI and HR-EBSD reveal a transition from a structure where defects are disordered to a structure with long-range order driven by elastic interactions between pre-existing defects and new cascade damage. This study demonstrates that SEM provides an attractive tool for rapid throughput, non-destructive, multi-scale and multi-aspect characterisation of irradiation damage.
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Predicting the dramatic changes in material properties caused by irradiation damage is key for the design of future nuclear fission and fusion reactors. Self-ion implantation is an attractive tool for mimicking the effects of neutron irradiation. However, the damaged layer of implanted samples is only few microns thick, making it difficult to estimate macroscopic properties. Here we address this challenge using a combination of experimental and modelling techniques. We concentrate on self-ion-implanted tungsten, the front-runner for fusion armour components and a prototypical bcc material. To capture dose-dependent evolution of properties, we experimentally characterise samples with damage levels from 0.01 to 1 dpa. Spherical nano-indentation of <001> grains shows hardness increasing up to a dose of 0.032 dpa, beyond which it saturates. AFM measurements show pile-up increasing up to the same dose, beyond which large pile-up and slip-steps are seen. Based on the observations we develop a crystal plasticity (CPFE) model for the irradiated material. It captures irradiation-induced hardening followed by strain-softening through interaction of irradiation-defects and gliding dislocations. Shear resistance of irradiation-defects is derived from TEM observations of similarly irradiated samples. Nano-indentation of pristine and implanted tungsten of doses 0.01, 0.1, 0.32 and 1 dpa is simulated. Two model parameters are fitted to the experimental results of the 0.01 dpa sample and are kept unchanged for all other doses. Peak load, indent surface profiles and damage saturation predicted by the CPFE model closely match experimental observations. Predicted lattice distortions and dislocation distributions around indents agree with corresponding measurements from HR-EBSD. Finally, the CPFE model is used to predict the macroscopic stress-strain response of similarly irradiated bulk tungsten material.
Using _in situ_ transmission electron microscopy (TEM), we have observed nanometre scale dislocation loops formed when an ultra-high-purity tungsten foil is irradiated with a very low fluence of self-ions. Analysis of the TEM images has revealed the largest loops to be predominantly of prismatic 1/2<111> type and of vacancy character. The formation of such dislocation loops is surprising since isolated loops are expected to be highly mobile, and should escape from the foil. In this work we show that the observed size and number density of loops can be explained by the fact that the loops are _not_ isolated - the loops formed in close proximity in the cascades interact with each other and with vacancy clusters, also formed in cascades, through long-range elastic fields, which prevent the escape of loops from the foil. We find that experimental observations are well reproduced by object Kinetic Monte Carlo simulations of evolution of cascades _only_ if elastic interaction between the loops is taken into account. Our analysis highlights the profound effect of elastic interaction between defects on the microstructural evolution of irradiated materials.
Developing a comprehensive understanding of the modification of material properties by neutron irradiation is important for the design of future fission and fusion power reactors. Self-ion implantation is commonly used to mimic neutron irradiation damage, however an interesting question concerns the effect of ion energy on the resulting damage structures. The reduction in the thickness of the implanted layer as the implantation energy is reduced results in the significant quandary: Does one attempt to match the primary knock-on atom energy produced during neutron irradiation or implant at a much higher energy, such that a thicker damage layer is produced? Here we address this question by measuring the full strain tensor for two ion implantation energies, 2 MeV and 20 MeV in self-ion implanted tungsten, a critical material for the first wall and divertor of fusion reactors. A comparison of 2 MeV and 20 MeV implanted samples is shown to result in similar lattice swelling. Multi-reflection Bragg coherent diffractive imaging (MBCDI) shows that implantation induced strain is in fact heterogeneous at the nanoscale, suggesting that there is a non-uniform distribution of defects, an observation that is not fully captured by micro-beam Laue diffraction. At the surface, MBCDI and high-resolution electron back-scattered diffraction (HR-EBSD) strain measurements agree quite well in terms of this clustering/non-uniformity of the strain distribution. However, MBCDI reveals that the heterogeneity at greater depths in the sample is much larger than at the surface. This combination of techniques provides a powerful method for detailed investigation of the microstructural damage caused by ion bombardment, and more generally of strain related phenomena in microvolumes that are inaccessible via any other technique.
Hydrogen isotopes are retained in materials for fusion power applications, changing both hydrogen embrittlement and tritium inventory as the microstructure undergoes irradiation damage. But modelling of highly damaged materials - exposed to over 0.1 displacements per atom (dpa) - where asymptotic saturation is observed, for example tungsten facing the plasma in a fusion tokamak reactor, is difficult because a highly damaged microstructure cannot be treated as weakly interacting isolated defect traps. In this paper we develop computational techniques to find the defect content in highly irradiated materials without adjustable parameters. First we show how to generate converged high dose (>1 dpa) microstructures using a combination of the creation-relaxation algorithm and molecular dynamics simulations of collision cascades. Then we make robust estimates of point defects and void regions with simple developments of the Wigner-Seitz decomposition of lattice sites. We use our estimates of the void surface area to predict the deuterium retention capacity of tungsten as a function of dose. This is then compared to 3He nuclear reaction analysis (NRA) measurements of tungsten samples self-irradiated at 290 K to different damage doses and exposed to deuterium plasma at low energy at 370 K. We show that our simulated microstructures give an excellent match to the experimental data, with both model and experiment showing 1.5-2.0 at.% deuterium retained in tungsten in the limit of high dose.
Understanding the mechanisms of plasticity in structural steels is essential for the operation of next-generation fusion reactors. Elemental composition, particularly the amount of Cr present, and irradiation can have separate and synergistic effects on the mechanical properties of ferritic/martensitic steels. The study of ion-irradiated FeCr alloys is useful for gaining a mechanistic understanding of irradiation damage in steels. Previous studies of ion-irradiated FeCr did not clearly distinguish between the nucleation of dislocations to initiate plasticity, and their propagation through the material as plasticity progresses. In this study, Fe3Cr, Fe5Cr, and Fe10Cr were irradiated with 20 MeV Fe$^{3+}$ ions at room temperature to nominal doses of 0.01 dpa and 0.1 dpa. Nanoindentation was carried out with Berkovich and spherical indenter tips to study the nucleation of dislocations and their subsequent propagation. The presence of irradiation-induced defects reduced the theoretical shear stress and barrier for dislocation nucleation. The presence of Cr further enhanced this effect due to increased retention of irradiation defects. However, this combined effect is still small compared to dislocation nucleation from pre-existing sources such as Frank-Read sources and grain boundaries. The yield strength, an indicator of dislocation mobility, of FeCr increased with irradiation damage and Cr. The increased retention of irradiation defects due to the presence of Cr also further increased the yield strength. Reduced work hardening capacity was also observed following irradiation. The synergistic effects of Cr and irradiation damage in FeCr appear to be more important for the propagation of dislocations, rather than their nucleation.
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