Journal of Nuclear Materials最新文献

In order to model correctly the corrosion of spent nuclear fuel under disposal conditions, it is important to understand its behavior in the presence of oxidants. To advance in this direction, we consider the oxidation of UO₂. We investigate computationally the adsorption of various species on its three most stable surfaces: (111), (110), and (100), with emphasis on incorporating a full non-collinear PBE+U approach. Various species, namely O, O₂, H₂O and H₂O₂ are considered due to their relevance for the oxidation of UO₂. The dissociation energy and an estimate for the dissociation barrier for O₂ were obtained, using the preferred adsorption configurations of O and O₂. The adsorption configurations for H₂O in our study compare well with previous studies that used collinear approximations, both in terms of relative stability of configurations and bond lengths. Differences in adsorption energies were found, which may be important for reaction kinetics. Dissociative reactions in which the water molecule splits in hydrogen and hydroxyl occur only on one of the three surfaces. The hydrogen further reacts with a surface oxygen to also form a hydroxyl group. Not surprisingly, we find that H₂O₂ binds more strongly to the three surfaces than water (lower formation energy), and similar to H₂O adsorption, dissociative reactions may occur. The dissociated hydrogen reacts with a surface oxygen to form a hydroxyl group and the hydroperoxyl molecule binds with a surface uranium. Our study, which includes a detailed study of electron transfer, magnetic structure and the preferred adsorption configurations, gives insight into the uranium oxidation states and the influence of surface geometry on adsorption. The findings contribute to a more comprehensive understanding of the early stages of UO₂ oxidation.

High-entropy alloys (HEA) have attracted considerable attention in the development of nuclear materials due to their excellent properties. In this study, the primary irradiation damage and long-term defect evolution of pure V, V-5Ti-5Ta conventional alloy, V-Ti-Ta MEA and V-Ti-Ta-Nb HEA were simulated by molecular dynamics (MD) to understand the irradiation resistance mechanism of these four systems. The primary irradiation damage simulation results indicate that the V-Ti-Ta MEA and V-Ti-Ta-Nb HEA exhibit a delayed thermal peak, a longer defect recombination time and a slightly higher number of final Frenkel pairs (FPs) than pure V and V-5Ti-5Ta alloy. Both primary irradiation damage and long-time defect evolution results show that V-Ti-Ta MEA and V-Ti-Ta-Nb HEA have lower defect clustering fraction, cluster size and dislocation loop size than pure V and V-5Ti-5Ta. This is because V-Ti-Ta MEA and V-Ti-Ta-Nb HEA exhibit lower dislocation loop binding energy and defect mobility compared to pure V and V-5Ti-5Ta alloy. This study investigates the reasons for the better radiation resistance of V-Ti-Ta MEA and V-Ti-Ta-Nb HEA than pure V and V-5Ti-5Ta conventional alloys.

We investigated the stability of oxide nano particles in oxide dispersion-strengthened (ODS) steel, which is a promising candidate material for next-generation reactors, under neutron irradiation at high temperature to high doses. MA957, a 14Cr-ODS steel, was irradiated with Joyo in Japan Atomic Energy Agency under irradiation conditions of 130 dpa at 502 ºC, 154 dpa at 589 ºC, and 158 dpa at 709 ºC. Three-dimensional atom probe (3D-AP) and transmission electron microscope (TEM) observation were performed to characterize the oxide particles in the ODS steels. A high number density of Y-Ti-O particle was observed in the unirradiated and irradiated samples. Almost no change in the morphology of the oxide particles, i.e. average diameter, number density, and chemical composition, has been observed in the samples irradiated to 130 dpa at 502 ºC and to 154 dpa at 589 ºC. A slight decrease in number density was observed in the sample irradiated to 158 dpa at 709 ºC. The hardness of any of the irradiated samples was almost unchanged from that of the unirradiated sample. It was revealed that the oxide particles existed stable, and the strength of the material was sufficiently maintained even after being neutron irradiated to high dose of ∼160 dpa at high temperature up to 700 ºC.

Master curve (MC) testing of VVER-440 RPV surveillance specimens treated for 27 years (∼200,000 h) in a surveillance channel of Metsamor-NPP was performed to investigate the influence of long term high fluence irradiation on RPV embrittlement. The surveillance chain consisted of both thermal aged specimens (above the core level) and irradiated specimens (inside the core). The reference temperature (T_o) values obtained from irradiated specimens are compared with the results from thermal aged specimens to characterize irradiation induced shifts in T_o values for both base and weld metal specimens. It was found that the high fluence irradiation up to a nominal fluence of 3.2 × 10²⁵ n.m^-2 at E > 0.5 MeV resulted large embrittlement with T_o shift values greater than 300 °C for both base and weld metal specimens. The obtained shifts in T_o values at these high fluence values were used to compare with the predictions from PNAE procedure in the Russian regulatory guide outside its validity range. It was found that the measured shift in reference temperature for the weld metal was well below the predicted value while, the shift in reference temperature for base metal was largely under-predicted at these high fluence values.

In this work, we investigated the impact of different concentrations of helium (He) impurities on the electronic thermal transport properties of tungsten plasma-facing materials (W-PFMs) at finite temperatures using the W-He tight-binding (TB) potential model. We found that the electronic transport performance decreases with increasing He atom concentration at different sites, where the greatest reduction in the electrical conductivity of the system is caused by the introduction of He atoms at neighboring tetrahedral sites. As the temperature increases, the electrical conductivity decreases, while the electronic thermal conductivity increases. Importantly, the higher the temperature is, the weaker the response of the electrical conductivity and electronic thermal conductivity to the He atom concentration. We suggest that this behavior is attributed to the diverse contributions of scattering mechanisms within various temperature ranges. Furthermore, as the temperature increases, the electron scattering mechanism gradually transitions from electron-impurity scattering to electron-electron scattering. Additionally, our calculated atomic resolved electrical conductivity data indicate that at lower temperatures, the electrical conductivity is predominantly contributed by W atoms around the He cluster.

Based on a critical analysis of the existing model for the effect of excess vacancies and interstitials on the nucleation of coherent particles under irradiation, a new nucleation model is developed that uses the Brailsford-Bullough model for the steady state occupation probabilities of vacancies and interstitials at the particle interface, recently refined by the author. It is shown that, depending on the surface tension $\gamma$ of a coherent particle, the nucleation mechanism can be qualitatively different. In the case of a relatively small $\gamma$, an instability can arise in the irradiated solid solution leading to the barrier-free nucleation of coherent precipitates. In the opposite case of a relatively large $\gamma$, the classical one-dimensional theory of homogeneous nucleation is applicable. In order to eliminate the uncertainty in the magnitude of surface tension (from the literature) and to better understand the underlying mechanisms of coherent particle nucleation in irradiation tests, additional atomistic studies are recommended.

The mechanical behaviors and deformation mechanisms of Chinese low activation martensitic (CLAM) steel under extreme loading conditions were systematically studied. The mechanical experiments were performed at a wide range of strain rate (from 0.001 to 3500 s^-1) and temperature (from 300 to 1073 K). The main results show that the strength of the CLAM steel shows an apparent positive strain rate and temperature softening effect. In particular, at quasi-static loading conditions, the elongation of CLAM steel first decreases (300–673 K) and then increases (673–1073 K) with the temperature rising. Under dynamic conditions, the elongation of the CLAM steel is positively correlated with temperature rising and is larger than that under quasi-static loading conditions. The microstructure characterization results indicate that grain refinement during deformation and the positive strain rate effect on elongation are primarily governed by changes in grain size, especially at high temperatures. The relationship between the plasticity capability, precipitates and grain refinement are also analyzed. The obvious competitive mechanisms under different loading conditions in the recrystallization process of the CLAM steel. In summary, precipitates contribute to grain refinement in martensitic structures by providing nucleation sites for new grains and by obstructing dislocation movement, thereby raising the local strain and promoting dynamic recrystallization (DRX). Both of these mechanisms result in a finer and more uniform grain structure, which enhances the mechanical properties of the material, such as strength and toughness.

Irradiation induced growth is a constant volume shape change that occurs without externally applied stress that is observed in some materials under irradiation damage. Proton irradiation was carried out on a pure Zr sample to many dpa, to enable investigation of microscale aspects of the irradiation growth phenomenon. Irradiation induced a significant surface morphology change in the irradiated area, which is believed to be the result of the anisotropic growth behavior of the hcp structured Zr. The localized strain that developed was characterized by Electron back scatter diffraction (EBSD), on both the irradiated surface and on a cross sectional through-thickness plane. It was found that there is a correlation between the amount of local deformation and level of misorientation existing between two adjacent grains. The irradiation induced defect microstructure was characterized by transmission electron microscopy (TEM), showing 〈a〉 and 〈c〉 component loops similar to that generated by neutron irradiation in literature. Lastly, site specific focused ion beam (FIB) TEM lift-outs were prepared on local grain boundaries to investigate the origin of the localised deformation.

Liquid plasma facing walls allow for increased neutron-wall loading expanding the design space of fusion power-plants and experimental devices towards compact high-field reactors. This study presents the design of a compact radial build blanket for fusion devices composed of variable quantities of Lead (Pb) and Lithium-Lithium Hydride (Li-LiH). A tank-like cylindrical neutronic model of the early design of the stellarator reactor proposed by Renaissance Fusion is implemented in OpenMC (neutron transport and dose rate analyses). The reactor's radial build composition and blanket layer thicknesses are varied to fulfill the requirements on tritium breeding ratio (TBR), nuclear heat extraction, radiation shielding (for the coils, internal structures and external environment) for a stellarator-based power-plant. The analyses suggest that a radial build lower than a meter thick between the plasma and coils would be sufficient to allow for a TBR ∼ 1.60, an energy multiplication factor of ∼ 1.07, to capture ≥ 90% of the nuclear heat, limit the neutron fluence at the coils below 10¹⁹ n/cm², and limit the structural damage on the liquid metal vessel and magnet structure. In particular, a blanket composed of 32 cm of Pb and Li-LiH, 54 cm of a heavy metal hydride such as vanadium hydride (VH₂), along with a 1.3 m of concrete bioshield, would minimize the radial build of the stellarator reactor while fulfilling tritium breeding, shielding and heat extraction requirements.