Fuel retention in plasma facing tungsten components is a critical phenomenon affecting the mechanical integrity and radiological safety of fusion reactors. It is known that hydrogen can become trapped in small defect clusters, internal surfaces, dislocations, and/or impurities, and so it is common practice to seed W subsurfaces with irradiation defects in an attempt to precondition the system to absorb hydrogen. The amount of H can later be tallied by performing careful thermal desorption tests where released temperature peaks are mapped to specific binding energies of hydrogen to defect clusters and/or microstructural features of the material. While this provides useful information about the potential trapping processes, modeling can play an important role in elucidating the detailed microscopic mechanisms that lead to hydrogen retention in damaged tungsten. In this paper, we develop a detailed kinetic model of hydrogen penetration and trapping inspired by recent experiments combining ion irradiation, hydrogen plasma exposure, and thermal desorption. We use the stochastic cluster dynamics method to solve the system of coupled partial differential equations representing the mean field description of the multispecies system. The model resolves the spatial distribution of defects and hydrogen clusters during the three processes carried out experimentally and is parameterized with information from atomistic calculations. We find that the calculated thermal desorption spectra are broadly characterized by three H emission regions: (i) a low temperature one where dislocations are the main contributors to the release peaks; (ii) an intermediate one governed by hydrogen release from small overpressurized clusters with multiple overlapping peaks, and (iii) a high temperature one defined by clean isolated emission peaks from large underpressurized bubbles. These three temperature intervals are seen to largely correlate with the depth at which the clusters are found. The relevance of the ‘super abundant’ vacancy mechanism is assessed, finding that its main role is to transfer more clusters from the intermediate to the high temperature regions as its relevance increases. We find this picture to be in very good agreement with the experiments, adding confidence to the predictive potential of the models and their useto understand irradiation damage and plasma exposure effects in plasma facing components.
Irradiation damage is known to alter a material’s microstructure due to the accumulation of high densities of defect clusters. Such irradiated microstructures change the mechanical response of the material due to dislocation-defect interactions, which leads to a host of issues such as hardening, swelling, irradiation creep, embrittlement, etc. Traditionally, the effect of irradiation on the mechanical response of materials is evaluated via tensile tests of pre-irradiated specimens at different doses and temperatures. From a modeling perspective, methods exist that simulate irradiation and deformation as separate processes, with the former based on kinetic transport theory and the latter on crystal plasticity (CP). Generally, these are connected by a state variable, usually in the form of a characteristic length scale, that represents the defect concentration and strength and its effect on dislocation-mediated slip. However, cases where deformation takes place during irradiation are also important despite being less common. In this paper, we develop a coupled CP and stochastic cluster dynamics (SCD) approach capable of treating all instances of irradiation/deformation in irradiated materials. We apply the methodology to tungsten crystals due to its importance as a high-temperature candidate structural material and to its extensive defect and mechanical data base. SCD evolves the defect microstructure stochastically, providing a statistically-averaged defect cluster spacing parameter that informs CP calculations of the material’s mechanical deformation. The coupling is bi-directional in the sense that the SCD method updates the obstacle density and furnishes a resistance stress to the CP model, while CP feeds updated dislocation densities that act as defect sinks in the SCD calculation cycles. The coupling can be done sequentially, as in standard tensile tests of pre-irradiated materials, or concurrently, as in in situ straining tests during irradiation. We carry out simulations of realistic irradiation/deformation scenarios and highlight the differences between the present method and past works considering similar situations.
The formation of elongated zirconium hydride platelets during corrosion of nuclear fuel clad is linked to its premature failure due to embrittlement and delayed hydride cracking. Despite their importance, however, most existing models of hydride nucleation and growth in Zr alloys are phenomenological and lack sufficient physical detail to become predictive under the variety of conditions found in nuclear reactors during operation. Moreover, most models ignore the dynamic nature of clad oxidation, which requires that hydrogen transport and precipitation be considered in a scenario where the oxide layer is continuously growing at the expense of the metal substrate. In this paper, we perform simulations of hydride formation in Zr clads with a moving oxide/metal boundary using a stochastic kinetic diffusion/reaction model parameterized with state-of-the-art defect and solute energetics. Our model uses the solutions of the hydrogen diffusion problem across an increasingly-coarse oxide layer to define boundary conditions for the kinetic simulations of hydrogen penetration, precipitation, and dissolution in the metal clad. Our method captures the spatial dependence of the problem by discretizing all spatial derivatives using a stochastic finite difference scheme. Our results include hydride number densities and size distributions along the radial coordinate of the clad for the first 1.6 h of evolution, providing a quantitative picture of hydride incipient nucleation and growth under clad service conditions.
Face-centered-cubic Ni single crystal was compressed along [011] direction at a strain rate of 10 3 s -1 , and microscopic characterization of slip bands was analyzed to investigate dynamical deformation mechanism of the Ni single crystal. By analyzing Schmid Factor and orientation of slip bands, the specific slip planes were revealed. Furthermore, by analysis of Schmid Factors and dimensions of the deformed sample, the slip directions are revealed, and the ratio of the slipping-induced displacement (D-ratio) related to each slip system were determined. . Moreover, the quasi-static compression experimental results show the same slip systems, and the calculated slipping-induced displacement related to above 4 slip system have a ratio of nearly 1:3.8:3.8:1 under quasi-static compression. Thus, in a considerable strain rate range from 10 -3 to 10 3 s -1 , the activated slip systems and the ratio of their corresponding slipping-induced displacement are the same. However, experimental results show that, the slip bands on side surface of Ni single crystal subjected to dynamic compression is denser than that of the quasi-static compressed sample, indicating that the slip bands are formed more easily under dynamic loading condition.
Irradiation creep is known to be an important process for structural materials in nuclear environments, potentially leading to creep failure at temperatures where thermal creep is generally negligible. While there is a great deal of data for irradiation creep in steels and zirconium alloys in light water reactor conditions, much less is known for first wall materials under fusion energy conditions. Lacking suitable fusion neutron sources for detailed experimentation, modeling, and simulation can help bridge the dose-rate and spectral-effects gap and produce quantifiable expectations for creep deformation of first wall materials under standard fusion conditions. In this paper, we develop a comprehensive model for irradiation creep created from merging a crystal plasticity representation of the dislocation microstructure and a defect evolution simulator that accounts for the entire cluster dimensionality space. Both approaches are linked by way of a climb velocity that captures dislocation-biased defect absorption and a dislocation strengthening term that reflects the accumulation of defect clusters in the system. We carry out our study in Fe under first wall fusion reactor conditions, characterized by a fusion neutron spectrum with average recoil energies of 20 keV and a damage dose rate of [Formula: see text] dpa/s at temperatures between 300 and 800 K.
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