Uranium mononitride (UN) is a promising candidate fuel for next-generation fast reactors due to its high fissile density, superior thermal conductivity, and high melting point compared to conventional oxide fuels. However, scarce experimental data on UN and its thermophysical behaviour under fission product incorporation limits its performance assessment. Zirconium nitride (ZrN) is an efficient thermal conductor and a candidate material for inert matrix fuels. Given its high thermal conductivity, ZrN addition at sufficient concentrations should, in principle, induce percolation conduction and increase thermal conductivity in UN. To decouple chemistry from irradiation-induced porosity, known to dominate thermal degradation at high burnup, this study isolates the intrinsic chemical contribution of Zr incorporation under dense, low-porosity conditions. (U,Zr)N pellets with 6.5 and 20 at. % Zr were fabricated by spark plasma sintering (SPS), using powders produced from arc-melted alloy via the hydride-nitride-denitride route. Synchrotron powder X-ray diffraction confirmed the formation of solid solutions and enhanced Zr solubility after sintering, resulting in improved microstructural homogeneity. Thermal diffusivity was measured between 300 and 1500 K using light flash analysis, and thermal conductivity was derived using heat capacity and density correlations with porosity correction. Despite the intrinsically higher thermal conductivity of ZrN, the incorporation of 6.5 at. % Zr reduced the thermal conductivity relative to UN, consistent with impurity scattering. The 20 at. % Zr composition further decreased conductivity, indicating the microstructure does not meet the conditions required for percolation conduction. Differences in the temperature dependence of thermal diffusivity between UN and Zr-bearing samples highlight a compositional influence on heat transport. The results provide benchmark data for (U,Zr)N and insights into chemical and thermophysical interactions in nitride ceramics.

The diffusion of two volatile fission products, xenon (Xe) and krypton (Kr), in zirconia (ZrO₂) is investigated. Samples of Yttria (Y₂O₃)-stabilised tetragonal ZrO₂ were implanted with either Xe or Kr, at 300 keV, with a fluence of 10¹⁷ at./cm², and subsequently analysed with time-of-flight elastic recoil detection analysis (ToF-ERDA) to obtain elemental composition depth profiles. Samples were then annealed at 1200 °C for 9 h, and the effect of the annealing was assessed by ToF-ERDA measurements. From these measurements, first-order approximations of diffusion coefficients for Xe and Kr in ZrO<inf>2</inf> were derived, using a model based on Fick's second law, these being (1.36±0.87)×10⁻¹⁹ m²/s and (2.94±1.96)×10⁻¹⁹ m²/s at 1200 °C for Kr and Xe respectively. It was shown that ToF-ERDA can provide data to analyse the diffusion of elements in solid sample matrices and that a model based on Fick's Law can predict the diffusion of the implanted ions.

U(X)N-based SIMFUEL samples, where X represents Zr, Nb, Mo, and Ru, were fabricated using spark plasma sintering. These samples were characterized by neutron diffraction and scanning electron microscopy to gain insights into fission product solubility and speciation at high burnup levels. The fabricated samples included pseudo-binary and higher-order compositions, allowing for the decomposition of individual fission product effects. The characterization revealed the presence of U1-xZrxN, Zr1-xUxN, ZrN, Nb1-xUx, UxNb1-x, Nb2N, URu3, Mo, and (U,Mo)Ru3 as distinct fission-product-containing phases. Notably, only Zr was found to be soluble within the primary UN fuel matrix. Significant agglomeration and formation of a (Nb-rich core)–(Nb-poor shell) microstructure was observed for the Nb-containing samples. Mo was the only fission product to form metallic inclusions and the presence of Ru led to the formation of URu3 in the pseudo-binary system (UN-10at.%Ru), or (U,Mo)Ru3 in the higher-order samples containing 1, 1.5, and 2 at.% each of all of fission product elements i.e. UN-1at.%(ZrN, Nb, Mo, Ru). No complex nitride precipitates were found to form. The phases identified in the pseudo-binary compositions were analyzed using the Thermodynamics of Advanced Fuels-International Database (TAF-ID) and showed good agreement to experimental data, except for a possible miscibility gap in the UN-ZrN tie line and absence of the (U,Mo)Ru3 phase.

Direct damage evolution simulations based on electronic structure physics show a significant correlation between Cr concentration and polymorphism in the form of localized formation of C15 Laves phase structures in Fe-Cr alloys under irradiation. We elucidate the role of Cr content in the formation and stabilization of the C15 Laves phase structure, which is crucial to understanding the behavior of materials under extreme conditions. This study also reveals a connection between non-linear magnetic behavior and irradiation-induced swelling in Fe-Cr alloys. These results advance the comprehension of radiation-induced changes in magnetization and suggest a novel experimental approach for detecting C15 clusters in irradiated Fe-Cr alloys.

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.