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We report the first proof of principle of an efficient and cost-effective bentchtop alternatives to synchrotron radiation beamlines to perform at laboratory scale Xray Absorption Spectroscopy (XAS) at the U L3-edge in transmission mode. We find excellent agreement with synchrotron-based studies for concentrated samples, in reasonable acquisition time, for UO2, KUO3 and b-UO3 samples. The approach described here already constitutes an inexpensive answer to the XAS beamline over-subscription in the field of actinide’s research for state of the art experiments. Moreover, our results opens the door to many future applications in the field of actinide’s research, including f-electron chemistry, environmental chemistry and nuclear energy physico-chemistry such as advanced nuclear fuel and long term nuclear waste disposal.

Transmission electron microscopy observation of Kr bubble evolution in polycrystalline UO 2 annealed at high temperature was conducted in order to understand the inert gas behavior in oxide nuclear fuel. The average diameter of intragranular bubbles increased gradually from 0.8 nm in as-irradiated sample at room temperature to 2.6 nm at 1600 °C and the bubble size distribution changed from a uniform distribution to a bimodal distribution above 1300 °C. The size of intergranular bubbles increased more rapidly than intragranular ones and bubble denuded zones near grain boundaries formed in all the annealed samples. It was found that high-angle grain boundaries held bigger bubbles than low-angle grain boundaries. Complementary atomistic modeling was conducted to interpret the effects of grain boundary character on the Kr segregation. The area density of strong segregation sites in the high-angle grain boundaries is much higher than that in the low angle grain boundaries.

The charge compensation mechanisms in U 1-x Gd x O 2 and Th 1-x Gd x O 2-x/2 have been systematically studied using X-ray Absorption Spectroscopy (XAS) upon gradually increasing the Gd content. Gd doped nuclear fuels are widely used for optimizing the fresh core neutronics, yet when Gd 3+ is substituted into U 4+ or Th 4+ lattice position in UO 2 or ThO 2 , respectively, charge must be compensated for charge neutrality. In U 1-x Gd x O 2 the general hypothesis has been that the U 4+ will oxidise to U 5+ /U 6+ while in Th 1-x Gd x O 2-x/2 the fixed Th 4+ valence requires generation of O vacancies. Our XAS results for a series of technologically relevant Gd contents (x = 0.04 to 0.14) in U 1-x Gd x O 2 clearly demonstrate that upon increasing the Gd content U 5+ is formed inducing slight increase in the U coordination number and contraction for the U-O distances while the Gd local environment remains virtually unchanged. For the Th 1-x Gd x O 2-x/2 larger Gd fractions were applied (x = 0.10 to 0.34). Nonetheless, both Gd and Th local environments show changes upon increasing the Gd content; the average Gd-O and Th-O distances decrease gradually and the Gd and Th coordination numbers follow the expected trend considering the O vacancy formation to obtain charge neutrality. Finally, comparison to Gd 2 O 3 allowed us to propose that one of the Gd L 3 -edge XANES resonance features is directly connected to the generation of O vacancies.

Two binary Ni-Cr model alloys with 5 wt% Cr and 18 wt% Cr were irradiated using 2 MeV protons at 400 and 500 °C and 20 MeV Ni4+ ions at 500 °C to investigate microstructural evolution as a function of composition, irradiation temperature, and irradiating ion species. Transmission electron microscopy (TEM) was applied to study irradiation-induced void and faulted Frank loops microstructures. Irradiations at 500 °C were shown to generate decreased densities of larger defects, likely due to increased barriers to defect nucleation as compared to 400 °C irradiations. Heavy ion irradiation resulted in a larger density of smaller voids when compared to proton irradiations, indicating in-cascade clustering of point defects. Cluster dynamics simulations were in good agreement with the experimental findings, suggesting that increases in Cr content lead to an increase in interstitial binding energy, leading to higher densities of smaller dislocation loops in the Ni-18Cr alloy as compared to the Ni-5Cr alloy.

The lattice deformations in two doped fluorite systems, (U1−xGdx)O2.00 and (Th1−xGdx)O2−x/2, have been reassessed by precise X-ray and electron diffraction investigations and the results were interpreted using the Bond Valence Sum (BVS) approach. For the (U1−xGdx)O2.00 system, the experimental findings and theoretical assessment confirm earlier work: the lattice keeps its fluorite structure with a unit cell parameter that contracts linearly with dopant concentration. The lattice contraction in the (Th1−xGdx)O2−x/2 system has for the first time been analyzed up to the solubility limit. Similar as for (U1−xGdx)O2.00, the (Th1−xGdx)O2−x/2 solid solution contracts linearly as a function of dopant concentration but additionally, it develops a superstructure which is closely related to the parent fluorite structure. An excess anion bixbyite trial model is proposed to describe this superstructure.

Thorium is most commonly precipitated as oxalate, because of the high efficiency regardless of precipitation conditions. Thorium oxalate readily decomposes into fine-grained ThO2 during heating, but keeps the macrostructure of the oxalate platelets. To assess the effects of precipitate macrostructure on the sintering behavior, different platelet morphologies were prepared and sintered. There are two factors which influence the sintering: the presence of holes within the platelets and the size of the platelets. Small platelets or precipitates with holes generally sinter to thorium dioxide pellets of a higher density and smaller grain size.

To enable the life extension of Light Water Reactors (LWRs) beyond 60 years, it is critical to gain adequate knowledge for making conclusive predictions to assure the integrity of duplex stainless steel reactor components, e.g. primary pressure boundary and reactor vessel internal. Microstructural changes in the ferrite of thermally aged, neutron irradiated only, and neutron irradiated after being thermally aged cast austenitic stainless steels (CASS) were investigated using atom probe tomography. The thermal aging was performed at 400 °C for 10,000 h and the irradiation was conducted in the Halden reactor at ~315 °C to 0.08 dpa (5.6 × 1019 n/cm2 E > 1 MeV). Low dose neutron irradiation at a dose rate of 5 × 10-9 dpa/s was found to induce spinod,al decomposition in the ferrite of as-cast microstructure, and further to enhance the spinodal decomposition in the thermally aged cast alloys. Regarding the G-phase precipitates, the neutron irradiation dramatically increases the precipitate size, and alters the composition of the precipitates with increased, Mn, Ni, Si and Mo and reduced Fe and Cr contents. Lastly, The results have shown that low dose neutron irradiation can further accelerate the degradation of ferrite in a duplex stainless steel at the LWRmore » relevant condition.« less

In situ and ex situ transmission electron microscopy observation of small Kr bubbles in both single-crystal and polycrystalline UO2 were conducted to understand the inert gas bubble behavior in oxide nuclear fuel. The bubble size and volume swelling are shown as weak functions of ion dose but strongly depend on the temperature. The Kr bubble formation at room temperature was observed for the first time. The depth profiles of implanted Kr determined by atom probe tomography are in good agreement with the calculated profiles by SRIM, but the measured concentration of Kr is about 1/4 of the calculated concentration. This difference is mainly due to low solubility of Kr in UO2 matrix and high release of Kr from sample surface under irradiation.

The microstructural changes and associated effects on thermal conductivity were examined in UO2 after irradiation using 3.9 MeV He2+ ions. Lattice expansion of UO2 was observed in x-ray diffraction after ion irradiation up to 5×1016 He2+/cm2 at low-temperature (< 200 °C). Transmission electron microscopy (TEM) showed homogenous irradiation damage across an 8 µm thick plateau region, which consisted of small dislocation loops accompanied by dislocation segments. Dome-shaped blisters were observed at the peak damage region (depth around 8.5 µm) in the sample subjected to 5×1016 He2+/cm2, the highest fluence reached, while similar features were not detected at 9×1015 He2+/cm2. Laser-based thermo-reflectance measurements showed that the thermal conductivity for the irradiated layer decreased about 55 % for the high fluence sample and 35% for the low fluence sample as compared to an un-irradiated reference sample. Detailed analysis for the thermal conductivity indicated that the conductivity reduction was caused by the irradiation induced point defects.

A novel phase field model has been developed to investigate grain growth in porous polycrystalline UO2. Based on a system of Cahn–Hilliard and Allen–Cahn equations, the model takes into consideration both the curvature driven grain boundary motion and pore migration by surface diffusion. As such, the model accounts for the interaction between pore and grain boundary kinetics, which tends to retard the growth process. The phase field model parameters are found in terms of measurable material properties. Hence, quantitative results that can be compared with experiments were obtained. The model has been used to investigate the effect of porosity on the kinetics of grain growth in UO2. It is found that, as the amount of porosity increases, grain growth in UO2 gradually changes from boundary controlled growth to pore controlled growth. For high porosity levels, the grain growth completely stops after a short evolution time. It is also found that the inhomogeneous distribution of pores leads to abnormal grain growth even without taking into account the anisotropy in grain boundary energy and mobility. The effects of porosity, temperature and initial microstructure on grain growth were thoroughly investigated. The model predictions are in good agreement with published experimental results of grain growth in UO2.

In situ transmission electron microscopy (TEM) observation of UO2 single crystal irradiated with Kr ions at high temperatures was conducted to understand the dislocation evolution due to high-energy radiation. The dislocation evolution in UO2 single crystal is shown to occur as nucleation and growth of dislocation loops at low-irradiation doses, followed by transformation to extended dislocation segments and networks at high doses, as well as shrinkage and annihilation of some loops and dislocations due to high temperature annealing. Generally the trends of dislocation evolution in UO2 were similar under Kr irradiation at different ion energies and temperatures (150 keV at 600 °C and 1 MeV at 800 °C) used in this work. Interstitial-type dislocation loops with Burgers vector along 〈1 1 0〉 were observed in the Kr-irradiated UO2. The irradiated specimens were denuded of dislocation loops near the surface.