Journal of Nuclear Materials最新文献

The evolution of Frenkel pairs has been studied experimentally and theoretically in tungsten, a Body-Centered Cubic metal. We used positron annihilation spectroscopy to characterize vacancy defects induced by electron irradiation in two sets of polycrystalline tungsten samples at room temperature. Doppler Broadening spectrometry showed that some positrons were trapped at pure single vacancies with a lower concentration than expected. At the same time, positron annihilation lifetime spectroscopy revealed that positrons are annihilated in unexpected states with a lifetime 1.44–1.64 times shorter than that of single vacancy (200 ps), namely unidentified (X) defects. Secondary ions mass spectrometry detected a significant concentration of oxygen in these samples, of the same order of magnitude as electron-induced single vacancy. In addition, Cluster dynamics simulated defect behaviors under experimental conditions, and Two-component density functional theory was used to calculate defect annihilation characteristics that are difficult to obtain in experiments. Finally, by combining the theoretical data, we simulated the positron signals and compared them with the experimental data. This enabled us to elucidate the interactions between oxygen and Frenkel Pairs. The X defects were identified as oxygen-vacancy complexes formed during irradiation, as oxygen is mobile in tungsten at room temperature, and can be trapped in a vacancy, while its binding to self-ion atoms leads to their immobilization thus reducing defect recombination. Therefore, we anticipate oxygen to play an important role in the evolution of tungsten microstructure under irradiation.

The thermal conductivity of U-Mo alloys for research reactor fuel was studied. Thermal conductivity is one of the most important properties concerning design, performance analysis, and safety evaluation of these alloys as research reactor fuel. Thermal conductivity data of unirradiated U-Mo alloy collected from literature were examined, analyzed, and used to develop a correlation as a function of Mo content and temperature. For irradiated U-10Mo alloy, a thermal conductivity correlation was developed as a function of fission density and temperature. This model considers the effects of porosity growth by fission gas bubbles, solid fission product accumulation, and buildup of grain boundaries due to recrystallization during irradiation.

The introduction of Hydrogen (H) into Zirconium (Zr) influences many mechanical properties, especially due to low H solubility and easy formation of Zirconium hydride phases. Understanding the various effects of H requires studies with atomistic resolution but at scales that incorporate defects such as cracks, interfaces, and dislocations. Such studies thus demand accurate interatomic potentials. Here, a neural network potential (NNP) for the Zr-H system is developed within the Behler-Parrinello framework. The Zr-H NNP retains the accuracy of a recent NNP for hcp Zr and exhibits excellent agreement with first-principles density functional theory (DFT) for (i) H interstitials and their diffusion in hcp Zr, (ii) formation energies, elastic constants, and surface energies of relevant Zr hydrides, and (iii) energetics of a common Zr/Zr-H interface. The Zr-H NNP shows physical behavior for many different crack orientations in the most-stable ε-hydride, and structures and reasonable relative energetics for the $〈a〉$ screw dislocation in pure Zr. This Zr-H NNP should thus be very powerful for future study of many phenomena driving H degradation in Zr that require atomistic detail at scales far above those accessible by first-principles.

Novel 1Al-Sc-Y ODS Eurofer steel for use in high temperature heavy metal cooling technologies has been developed by a mechanical alloying of feedstock powders using spark plasma sintering technique and a two-step heat treatment. The 1 wt.% Al has been added to increase the resistance in the environments at high temperatures. This paper reports the microstructure and mechanical properties of the material in the as-manufactured state, and after exposure in liquid lead. Emphasis is placed on the determination of environmental resistance and namely susceptibility to cracking in contact with liquid lead. The material specimens were exposed to static liquid lead with an average concentration of oxygen 1 × 10⁻⁶ wt.% at 600 °C for 1000 h and then, tensile, impact, fracture toughness and slow three-point bend testing was performed. After the exposure, some Pb penetrated into a subsurface layer which in fact caused wetting but testing with the remains of liquid lead at the surface of corroded specimens at 350 °C showed that the steel stayed very low susceptible to the cracking.

The development and application of robust tritium permeation barriers (TPBs) are crucial for a safe and economical fusion reactor operation because tritium permeation could cause serious safety problems, such as the brittleness of the structural material, fuel loss, and radioactive contamination. Coatings may effectively inhibit the permeation through the structural materials of tritium fuel generated inside the breeding blanket (BB) reducing the loss of lithium inventory contamination of the ancillary systems and the exposure of workers and the population. Al₂O₃ has been preliminary selected as the EU candidate for tritium permeation reduction of structural steels in the Water-cooled lithium–lead breeding blank (WCLL) concept, because of its very high permeation reduced factor (PRF). However, because of the chemical reactivity of lithium, several authors report the formation of Lithium aluminates after exposure to PbLi baths, based on different experimental techniques. In this contribution, we will show recent results about the quantification of the extent of the reaction and the thickness of the reacted layer and discuss the possible effects, that may happen under neutron irradiation based on the species produced (T and He) in the nuclear reactions (n, T) and (n, α) with ⁶Li and ⁷Li, with the Li atoms in the reacted layer To do that, the atomic density of produced T and He per second out of the transmutation reaction and the He/T ratio are calculated for diverse Li contents. Subsequently, we implant He into the coatings and characterize the desorption rate and the morphology. Finally, discussion and recommendations for further coating development will be presented.

General corrosion of nuclear reactor in core material like Zircaloy-2 and breakaway corrosion in particular are of great importance in ensuring its smooth long term operation. Metal ions like Mg²⁺ added to the coolant, to mitigate corrosion of other structural materials like Carbon steel, can get incorporated in the corrosion product oxide on Zircaloy-2 and alter its corrosion behavior. Plasma Electrolytic Oxidation (PEO) method was employed to investigate the effect of Mg²⁺ ions during breakaway corrosion of Zircaloy-2. The main objective was to understand the morphology, protectiveness, semiconducting properties of the Mg modified oxide films on Zircaloy-2 surface. DC potentials 300, 400 and 500 V were used to accelerate corrosion kinetics and form oxide mimicking the breakaway corrosion regime. Borate buffer (pH 9.8) was used as the electrolyte, and Mg acetate was added as Mg source for probing the effect of Mg²⁺ ions. Oxide morphology depended largely on the formation potentials. The presence of magnesium resulted in thinner oxides with lower defect densities. GIXRD of the oxide showed stabilization of tetragonal phase in presence of Mg at 300 and 400 V. Oxide resistance and charge transfer resistances measured by EIS were observed to be higher due to Mg incorporation at these potentials. Mg addition facilitated formation of t-ZrO₂. The electrochemical measurements suggested that the presence of Mg in ZrO₂ could reduce the connected porosity in the oxide thereby stifling the diffusion paths. These factors lead to improved corrosion protectiveness of the oxide formed in the presence of Mg.

The impact of welding-induced mechanics on the oxidation and stress corrosion cracking (SCC) behavior of Alloy 600 in a pressurized water reactor (PWR) primary water was investigated using an Alloy 600-Alloy 152 M weldment. The microstructural analysis found a 0.05 mm-wide composition transition zone from welding. The Alloy 600 heat-affected zone (HAZ) grain size matches the Alloy 600 base (600B). Deformation hardening initially decreases and stabilizes from the fusion boundary (FB) to the 600B Comparing the microstructural characteristics of the specimen from the HAZ at about 0.5 mm from the FB with specimen 600B shows identical composition, but the former exhibits higher Vickers hardness (HV), kernel average misorientation (KAM), dislocation density, and residual stresses. Results from both short- and long-term oxidation tests indicate that the specimen from the HAZ exhibits a greater thickness of the inner oxide film, a higher number of local oxidation sites, and an increased maximum depth of intergranular (IG) oxidation compared to the 600B specimen. Multiple sets of SCC test results demonstrate that the crack growth rate (CGR) in the specimens from the HAZ is higher than that in the base, with the former showing longer IG cracks and a greater proportion of IG cracking.

The bonding properties at the interface between the first wall oxide dispersion strengthened tungsten (ODS-W) and the heat sink copper (Cu) are vital for their application in future nuclear fusion reactors. However, the immiscibility and significant differences in the coefficient of thermal expansion between ODS-W and Cu pose considerable challenges for bonding these two materials. This study utilizes anodization and hydrogen reduction processes to form a nanoporous structure on the surface of ODS-W, thereby achieving effective bonding between ODS-W and Cu by spark plasma sintering (SPS). The effects of the anodization parameters on the surface characteristics of ODS-W, and the influence of the bonding temperature on the interfacial microstructure and mechanical properties of the ODS-W/Cu joints were investigated. The results demonstrate that under an anodization condition of 50 V for 40 min, a nanoporous structure with an average pore size of about 90 nm can forms on the surfaces of ODS-W. This significantly increases the surface roughness, thereby enhancing the interfacial bonding of ODS-W with Cu during SPS. As the bonding temperature increases, the interfacial microstructure changes from linear to serrated shape, effectively suppressing crack propagation and forming a stronger mechanical interlock between ODS-W and Cu. This significantly enhances the mechanical properties of the joints. Typically, at a bonding temperature of 1000 °C, the tensile strength of the anodized ODS-W/Cu joints reaches 245.2 MPa, which is 157.9 % higher than that of directly bonded joints. Additionally, the tensile fracture primarily occurs within the Cu substrate, transitioning from brittle to ductile fracture modes.

Next-generation accident-tolerant fuels for light-water reactors (LWRs) are being developed with high thermal conductivity additives. To reduce the high centerline temperatures that occur in present LWR fuel, complex fuels with improved heat dissipation capabilities are the goal. Depending on the manufacturing process, these fuels can use a variety of additives, and their microstructures vary. The thermal conductivity of UO₂-SiC complex fuel is predicted using the finite element method (FEM) and machine learning (ML), as shown in this research. The method accurately predicts the corresponding complex fuel thermal conductivity and captures the characteristic value successfully. Novel anisotropic fuel pellets with higher thermal conductivity in a desired direction can be designed through the application of FEM and ML to material design. Data from published experiments with UO₂-SiC complex fuel are used to validate the model and methodology. The model finally predicts the ideal characteristic parameters of the UO₂-SiC based on the expected thermal conductivity.

Due to the excellent resistance to radiation damage, high-entropy alloys (HEAs) have been considered as important candidates for nuclear materials. However, to realize the great potential of HEAs in fusion energy, it is important to consider the hydrogen (H) behaviors, which remain to be elucidated. Here, we systematically investigate the dissolution, diffusion and desorption of H in equiatomic WTaVCr using the first-principles calculations. It is found that the H solution energy in WTaVCr is lower than that in pure W, which can be clarified by the H affinity of metallic elements and the available volume of interstitial sites. Accordingly, a possible diffusion path of interstitial H is selected, containing 26 metastable sites, and the corresponding energy barrier is 0.46 eV. Besides, the maximum trapping energy and accommodation number of H at a mono-vacancy in WTaVCr is only 0.30 eV and 3 ∼ 5 H, respectively, which is much lower than that in pure W (1.28 eV and 12 H). More importantly, because of the low trapping energy and moderate diffusion energy barrier, the desorption temperature of H from a mono-vacancy is lower than the room temperature, implying that vacancies are not efficient trapping centers for interstitial H atoms in WTaVCr. Therefore, although the equilibrium interstitial H concentration in WTaVCr is higher than that in pure W, H retention at high temperatures and H-induced blistering are suppressed due to the weak H-defect interactions. Our results will provide a good reference for understanding the H behaviors in HEAs.