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Register a new userParameter-free quantitative simulation of high dose microstructure and hydrogen retention in ion-irradiated tungsten
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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.
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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.
The changing thermal conductivity of an irradiated material is among the principal design considerations for any nuclear reactor, but at present few models are capable of predicting these changes starting from an arbitrary atomistic model. Here we present a simple model for computing the thermal diffusivity of tungsten, based on the conductivity of the perfect crystal and resistivity per Frenkel pair, and dividing a simulation into perfect and athermal regions statistically. This is applied to highly irradiated microstructures simulated with Molecular Dynamics. A comparison to experiment shows that simulations closely track observed thermal diffusivity over a range of doses from the dilute limit of a few Frenkel pairs to the high dose saturation limit at 3 displacements per atom (dpa).
Deuterium(D) retention behavior in tungsten(W) exposed to deuterium plasma and gas was studied by means of thermal desorption spectroscopy (TDS): deuterium plasma exposure in which W was exposed to D plamsa with 35 eV/D at 393 K to the fluence of 3.8E24 D/m2; D2 gas charging in which W was exposed to D2 gas of 500 kPa at 773 K for 4 hours. TDS shows that the total D retention in plasma exposure W is 1.00E22 D/m2, one order of magnitude higher than that of gas charging W; however, the D2 desorption peak of gas charging W is 952 K, much higher than 691 K of plasma exposure W. The detrapping energies of deuterium were determined experimentally from the measured peak temperatures at different heating rates and were found to be 2.17 eV for gas charging W and 1.04 eV for plasma exposure W, respectively.
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.
Tungsten is the main candidate material for plasma-facing armour components in future fusion reactors. In-service, fusion neutron irradiation creates lattice defects through collision cascades. Helium, injected from plasma, aggravates damage by increasing defect retention. Both can be mimicked using helium-ion-implantation. In a recent study on 3000 appm helium-implanted tungsten (W-3000He), we hypothesized helium-induced irradiation hardening, followed by softening during deformation. The hypothesis was founded on observations of large increase in hardness, substantial pile-up and slip-step formation around nano-indents and Laue diffraction measurements of localised deformation underlying indents. Here we test this hypothesis by implementing it in a crystal plasticity finite element (CPFE) formulation, simulating nano-indentation in W-3000He at 300 K. The model considers thermally-activated dislocation glide through helium-defect obstacles, whose barrier strength is derived as a function of defect concentration and morphology. Only one fitting parameter is used for the simulated helium-implanted tungsten; defect removal rate. The simulation captures the localised large pile-up remarkably well and predicts confined fields of lattice distortions and geometrically necessary dislocation underlying indents which agree quantitatively with previous Laue measurements. Strain localisation is further confirmed through high resolution electron backscatter diffraction and transmission electron microscopy measurements on cross-section lift-outs from centre of nano-indents in W-3000He.
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