The use of molten lead as a heat exchange fluid poses important critical issues, both in terms of corrosion resistance and creep resistance, due to the temperatures and structural stresses reached during operation. The objective of this work has been the investigation of the corrosion resistance and mechanical properties of new experimental compositions of alumina-forming stainless-steel candidates for these applications. The exposures to stagnant liquid lead were carried out for 500 and 1, 000 hours, at temperatures of 550 and 650 °C, with controlled amounts of oxygen dissolved in the liquid lead. In comparison with the AISI 316L and T91 both tested as reference materials, the studied alloys showed highly promising corrosion behavior and mechanical properties. According to these results, the proposed steels are appropriate for components that will operate in liquid lead at elevated temperatures without corrosion, while maintaining good mechanical properties.

We here use density functional theory and the creation-relaxation algorithm to investigate the appearance of polymorphism in α-Fe, driven by irradiation-induced microstructural changes. Local constriction leads to magnetic instability and provides excess energy required for structural phase transformation. Under extreme conditions, α-Fe undergoes local transformations into icosahedral C15 Laves phase with highly close-packed stacking and internal short-range ferromagnetic ordering, antiparallel to the bulk magnetisation. Analysing local magnetic moments and atomic volumes, in conjunction with the magneto-volume relations of different Fe structures, suggests two other alternatives for local phase transformation under irradiation conditions: the double-layer antiferromagnetic γ-Fe and non-magnetic ϵ-Fe.

This study investigated the irradiation-induced microstructure evolution in Fe, employing the Creation-Relaxation Algorithm and different interatomic potentials. The influence of self-interstitial atoms (SIAs), which were either locally or uniformly being distributed during the creation of the Frenkel pairs, was investigated on the evolving microstructure. The spatially localized distribution of SIAs, mimicking the low-energy transfer irradiation conditions, moderated the microstructure development, compared to uniform distribution of SIAs, delaying the nucleation of dislocation for higher irradiation doses. Introducing multiple Frenkel pairs facilitated a cumulative irradiation dose of 5 dpa in large supercells. In small supercells, the accumulation of SIAs led to the formation of an artificially stabilized self-interacting planar interstitial cluster, suggesting a minimum cell dimension of 10 nm for an accurate modeling of microstructure evolution when the development of the dislocation network is of interest. The formation and evolution of the C15 Laves phase structure were explored. The evolving C15 structure developed larger clusters with uniformly distributed SIAs, and their sizes depended on the interatomic potential employed. Finally, a comparison with experimental measurements demonstrated that the density and the average size of interstitial dislocation loops aligned well with those observed in experimentally irradiated ultra-high purity Fe at low and room temperatures.

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 temperature-dependent effective thermal conductivity of UN-X-UO2 (X = Mo, W) nuclear fuel composite was estimated. Following the experimental design, the thermal conductivity was calculated using Finite Element Modeling (FEM), and compared with analytical models for 10%, 30%, 50%, and 70% (in mass) uncoated/coated UN microspheres in a UO2 matrix. The FEM results show an increase in the fuel thermal conductivity as the mass fraction of the UN microspheres increases from 1.2 to 4.6 times the UO2 reference at 2,000 K. The results from analytical models agree with the thermal conductivity estimated by FEM. The results also show that Mo and W coatings have similar thermal behaviors, and the coating thickness influences the thermal conductivity of the composite. At higher weight fractions, the thermal conductivity of the fuel composite at room temperature is substantially influenced by the high thermal conductivity coatings approaching that of UN. Thereafter, the thermal conductivity from FEM was used in the fuel thermal performance evaluation during LWR normal operation to calculate the maximum centerline temperature. The results show a significant decrease in the fuel maximum centerline temperature ranging from -94 K for 10% UN to -414 K for 70% (in mass) UN compared to UO2 under the same operating conditions.

In this study, the temperature-dependent effective thermal conductivity of the innovative UN-X-UO₂ (X=Mo, W) nuclear fuel composite has been estimated in the temperature range from room temperature to 2000 K. This composite fuel concept is considered as a promising accident tolerant fuel for light water reactors (LWRs). Following the previously reported experimental composite design, the composite fuel thermal conductivity was calculated using Finite Element modeling (FEM), and it is compared with analytical models of thermal conductivity for 10, 30, 50, and 70 wt.% uncoated/coated UN microspheres in a UO₂ matrix. The FEM results show an expected increase in the fuel thermal conductivity as the wt.% of the coated/uncoated UN microspheres increases – from 1.5 to 5.7 times the UO₂ reference at 2000 K. However, the analytical models show an overestimation of the fuel thermal conductivity as the wt.% increases. The results also show that Mo and W coatings have similar thermal behaviors and the coating thickness varying from 1-5 μm has an insignificant effect on the thermal behavior of the composite. However, at higher weight fractions, the thermal conductivity of the fuel composite at room temperature is substantially influenced by the high thermal conductivity coatings exceeding that of UN. Thereafter, the thermal conductivity profiles from FEM were used in the fuel thermal performance evaluation during LWR normal operation to calculate the maximum centerline temperature of the fuel composites. The results show a significant decrease in the fuel maximum centerline temperature ranging from −72 K for 10 wt.% UN to −438 K for 70 wt.% UN compared to the UO₂ under the same irradiation conditions, providing an enhanced safety margin and thermal and neutronic advantages.

Understanding the formation and evolution of Cu precipitates in Fe-based alloys is crucial as they are key factors responsible for hardening and embrittlement. Dilute FeCu alloys are model materials for structural components in various application areas, including that of reactor pressure vessel steels for light water nuclear reactors. Positron annihilation spectroscopy (PAS) is a powerful tool to study the nucleation stage of both homogeneous and heterogeneous Cu precipitation, which are beyond the reach of most other experimental techniques. In this work, we present a first-principles study of positron annihilation in Fe -Cu systems. The positron annihilation characteristics (positron lifetimes, Doppler broadening spectra) are calculated for both homogeneous vacancy-free Cu clusters and heterogeneous vacancy-Cu complexes us-ing two-component density functional theory. The theoretical results excellently agree with the available reference PAS experimental results. Our calculations show that the types of Cu precipitates can be clearly distinguished by positron annihilation, and the sizes of Cu precipitates can also be reasonably well es-timated with our calculations. Moreover, we also successfully analyze the evolution of the experimental signals during isochronal annealing where the small Cu clusters change character. This work improves the understanding of the early-stage Cu precipitation in Fe matrix.

Modelling of UO2 mechanical behavior requires detailed knowledge of the local stresses and strains during the fuel's operation in normal and accident conditions. Therefore, a crystal plasticity formulation is proposed for polycrystalline UO2. The model contains a dislocation-density-based formulation including three slip families and their interactions. The model is parametrized with single crystal and polycrystal experimental data using an optimization scheme. The model's capability to represent yield point, strain hardening behavior, temperature and strain rate dependencies are evaluated. Finally, different approaches to include porosity at the polycrystal are analyzed to assess the effect of porosity on homogenized macro-scopic stress-strain behavior, and stress/strain localization at the grain level.

We here model the radiation-induced microstructure evolution in bcc iron using the creation-relaxation algorithm directly driven by density functional theory calculations and compare to interatomic potential simulations. The method is in essence a relatively simplified model without thermally driven diffusion. The microstructure evolution in this model is driven mostly by the stress field introduced by sequential direct damage insertions. The defect populations and the corresponding defect cluster characteristics have been analyzed as a function of irradiation dose. Different distribution functions have been investigated for the self-interstitial atom implantation, to make model predictions as close as possible to actual irradiation conditions under which defects are produced. The stability and magnetic characteristics of the defect structures that are formed are studied. Our first-principles simulations revealed that the C15 structure can be dynamically formed during the irradiation of the material and that the constituent atoms align antiferromagnetically to the lattice. For doses on the order of a fraction of 1 displacement per atom, the model material Fe experiences mechanical changes caused by continuous irradiation and approaches a saturation state of about 2% swelling. The radiation-induced change in the local magnetic moments as well as the charge density differences for some isolated and clustered defects have been investigated. The results revealed a net reduction in total magnetization per atom and a tendency for interstitial sites to have a spin polarization opposing the intrinsic atomic site spins when the coordination number was increased compared to that of the initial lattice structure. It is suggested that radiation-induced damage could be traced using nondestructive measures of bulk magnetization changes.