RESUMEN
Intermetallic alloy ZrCo is a good material for storing tritium (T). However, ZrCo is prone to disproportionation reactions during the process of charging and discharging T. Alloying atoms are often added in ZrCo, occupying the Zr or Co site, in order to restrain disproportionation reactions. Meanwhile, T often decays into helium (He), and the purity of T seriously decreases once He escapes from ZrCo. Therefore, it is necessary to understand the influence of alloying atoms on the basic stability property of He. In this work, we perform systematical ab initio calculations to study the stability property of He in ZrCoH3 (ZrCo adsorbs the H isotope, forming ZrCoH3). The results suggest that the He atom will undergo displacements of 0.31 and 0.12 Å when it substitutes for Co and Zr, respectively. In contrast, the displacements are very large, at 0.67-1.09 Å, for He replacing H. Then, we introduce more than 20 alloying atoms in ZrCo to replace Co and Zr in order to examine the influence of alloying atoms on the stability of He at H sites. It is found that Ti, V, Cr, Mn, Fe, Zn, Nb, Mo, Tc, Ru, Ta, W, Re, and Os replacing Co can increase the substitution energy of H by the He closest to the alloying atom, whereas only Cr, Mn, Fe, Mo, Tc, Ru, Ta, W, Re, and Os replacing Co can increase the substitution energy of H by the He next closest to the alloying atom. The influence of the alloying atom substituting Zr site on the substitution energies is inconspicuous, and only Nb, Mo, Ru, Ta, and W increase the substitution energies of H by the He closest to the alloying atom. The increase in the substitution energy may suggest that these alloy atoms are conducive to fix the He atom in ZrCo and avoid the reduction in tritium purity.
RESUMEN
Oxidation corrosion of steel is a universal problem in various industries and severely accelerated in nuclear reactors. First-principles calculations are performed to explore the dissolution and diffusion properties of interstitial oxygen in the body-centered-cubic iron grain boundaries Σ3ã110ã(111) and Σ5ã001ã(310). Solution energies indicate that interstitial oxygen atoms prefer to dissolve in body-centered-cubic iron, and energetically segregate to grain boundaries. Energy barriers show that oxygen atoms would segregate towards Σ3ã110ã(111) with a low energy barrier. However, they concentrate to the transition region of Σ5ã001ã(310) due to the high-energy barrier in the transition zone. When O atoms arrive at grain boundaries, they would stay there due to the larger solution energy and diffusion energy barrier in grain boundaries compared to that in the defect-free Fe bulk. These results indicate that O atoms would prefer to diffuse through the bulk, and oxidize grain boundaries. This study provides insight into oxidation phenomena in experiments and necessary parameters for future studies on the oxidation of steel under irradiation in nuclear reactors.
RESUMEN
The interplay between hydrogen and nanovoids, despite long being recognized as a central factor in hydrogen-induced damage in structural materials, remains poorly understood. Here, focusing on tungsten as a model body-centred cubic system, we explicitly demonstrate sequential adsorption of hydrogen adatoms on Wigner-Seitz squares of nanovoids with distinct energy levels. Interaction between hydrogen adatoms on nanovoid surfaces is shown to be dominated by pairwise power-law repulsion. We establish a predictive model for quantitative determination of the configurations and energetics of hydrogen adatoms in nanovoids. This model, combined with the equation of states of hydrogen gas, enables the prediction of hydrogen molecule formation in nanovoids. Multiscale simulations, performed based on our model, show good agreement with recent thermal desorption experiments. This work clarifies fundamental physics and provides a full-scale predictive model for hydrogen trapping and bubbling in nanovoids, offering long-sought mechanistic insights that are crucial for understanding hydrogen-induced damage in structural materials.