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Understanding microstructural and strain evolutions induced by noble gas production in the nuclear fuel matrix or plasma-facing materials is crucial for designing next generation nuclear reactors, as they are responsible for volumetric swelling and catastrophic failure. We describe a multimodal approach combining synchrotron-based nanoscale X-ray imaging techniques with atomic-scale electron microscopy techniques for mapping chemical composition, morphology and lattice distortion in a single crystal W induced by Kr irradiation. We report that Kr-irradiated single crystal W undergoes surface deformation, forming Kr containing cavities. Furthermore, positive strain fields are observed in Kr-irradiated regions, which lead to compression of underlying W matrix.

This study characterizes the microstructural evolution of single-phase complex concentrated solid-solution alloy (CSA) compositions under heavy ion irradiation with the goal of evaluating mechanisms for CSA radiation tolerance in advanced fission systems. Three such alloys, Cr18Fe27Mn27Ni28, Cr15Fe35Mn15Ni35, and equimolar NbTaTiV, along with reference materials (pure Ni and E90 for the CrFeMnNi family and pure V for NbTaTiV) were irradiated at 50 K and 773 K with 1 MeV Kr++ ions to various levels of displacements per atom (dpa) using in-situ transmission electron microscopy. Cryogenic irradiation resulted in small defect clusters and faulted dislocation loops as large as 12 nm in face-centered cubic (FCC) CSAs. With thermal diffusion suppressed at cryogenic temperatures, defect densities were lower in all CSAs than in their less compositionally complex reference materials indicating that point defect production is reduced during the displacement cascade stage. High temperature irradiation of the two FCC CSA resulted in the formation of interstitial dislocation loops which by 2 dpa grew to an average size of 27 nm in Cr18Fe27Mn27Ni28 and 10 nm in Cr15Fe35Mn15Ni35. This difference in loop growth kinetics was attributed to the difference in Mn-content due to its effect on the nucleation rate by increasing vacancy mobility or reducing the stacking-fault energy.

Irradiation-induced defect evolution in graphite is particularly important for its application in graphite-moderated nuclear reactors. The evolution of defects directly influences macroscopically observed property changes in irradiated nuclear graphite which, in turn, can govern the lifetime of graphite components. This article reports novel defect structures and the irradiation response of microstructural features occurring in high-temperature irradiated nuclear graphite IG-110. High resolution transmission electron microscopy (HRTEM) was used to characterize specimens neutron-irradiated at a high temperature (≥800 °C) at doses of 1.73 and 3.56 atomic displacements per atom (dpa). Concentric shelled and fullerene-like defects were found to result in swelling along the c-axis and contraction along the a/b-axis of crystallites. Furthermore, such defects are shown to occur within, and partially fill, Mrozowski cracks prior to turnaround dose. In addition, in situ TEM under similar irradiation conditions was used to capture the real-time dynamic evolution of defects, providing unambiguous analysis of the evolution of the graphite structures during irradiation. Results suggest the mainstream theory for radiation damage in nuclear graphite (which assumes additional basal plane formation as the sole reason) to be an incorrect interpretation of defect evolution contributing to irradiation-induced property changes at higher temperatures.

The microstructure and nanomechanical properties of three samples of the modified stainless steel (referred as the alloy D9) were systematically characterized after neutron irradiation in the Advanced Test Reactor. The samples were irradiated to 5.0, 8.2, and 9.2 displacements per atom at 448, 430, and 683°C, respectively. The evolutions of dislocation loops, cavities, and radiation-induced precipitates were quantitatively studied to reveal their dose and temperature dependencies. Nanohardness and nanoindentation creep tests were conducted at room temperature on the irradiated samples. Unexpected radiation hardening was observed in the highest-temperature-irradiated sample due to the formation of an unknown type of Ni- and Si-rich precipitates whose contributions to the radiation and mechanical performances of the alloy were discussed. We provide the radiation-microstructure-property correlations of alloy D9 with new insights, which can benefit the development and optimization of advanced austenitic alloys for future nuclear applications.

In situ characterization of defects, microstructure, and properties will provide new perspectives regarding the structure-property relationship of materials in extreme environments. In this communication, we investigate the utility of laser-based thermal transport measurements in combination with X-ray diffraction as a means to characterize the early-stage evolution of irradiation-induced defects in ceramics. Uranium dioxide is used as a model system to analyze the impact of irradiation-induced defects with 2.6 MeV H and 3.9 MeV He ions up to a dose of 0.1 displacement per atom (dpa) at low temperature. For these radiation regimes, the formation of extended defects such as loops and voids is limited as compared to point defects. Lattice expansion was determined from X-ray diffraction analysis. Modulated thermoreflectance was used to measure the thermal conductivity of the ion damaged region. Both H and He irradiation leads to an expansion of the crystal lattice and a reduction in thermal conductivity. For the same dpa, the lattice expansion and conductivity reduction were notably different for H and He irradiated samples. The results were analyzed using simple models for lattice expansion and thermal conductivity reduction, informed by atomistic simulation from the literature. The modeling results suggest that the difference in the defect kinetics between two conditions can be attributed to ionization induced enhanced defect mobility and the stability of Schottky defects. These results demonstrate the utility of thermal conductivity measurements as a tool for characterization of microstructure under irradiation.

The current mainstream theory for radiation-induced dimensional change of nuclear graphite is based on the notion that interstitially displaced carbon atoms will coalesce into dislocation loops (i.e., additional basal planes). This standard atomic-displacement model has been challenged by theories based on first principles calculations. The so-called ‘ruck and tuck’ of basal planes has been proposed as an alternative mechanism to explain the observed c-axis expansion under irradiation; however, no such defects have been observed experimentally so far. In this study, the first experimental evidence for the presence of a ‘ruck and tuck’ defect in high-temperature neutron-irradiated nuclear graphite is presented.

Proton irradiation induced Nb redistribution in Zr-xNb alloys (x = 0.4, 0.5, 1.0 wt%) has been investigated using scanning transmission electron microscopy/energy dispersive X-ray spectroscopy (STEM/EDS). Zr-xNb alloys are mainly composed of Zr matrix, native Zr–Nb–Fe phases, and β-Nb precipitates. After 2 MeV proton irradiation at 350 °C, a decrease of Nb content in native precipitates, as well as irradiation-induced precipitation of Nb-rich platelets (135 ± 69 nm long and 27 ± 12 nm wide) were found. Nb-rich platelets and Zr matrix form the Burgers orientation relationship, [ 1 1 ¯ 1 ]//[ 2 1 ¯ 1 ¯ 0 ] and (011)//(0002). The platelets were found to be mostly coherent with the matrix with a few dislocations near the ends of the precipitate. The coherent strain field has been measured in the matrix and platelets by the 4D-STEM technique. The growth of Nb-rich platelets is mainly driven by coherency and dislocation-induced strain fields. Irradiation may both enhance the diffusion and induce segregation of interstitial Nb to the ends of the irradiation induced platelets, further facilitating their growth.

Environmental barrier coatings (EBCs) are necessary to protect SiC/SiC ceramic components against oxidation and hot corrosion in high-temperature applications. The volatilization of SiO2 in SiC-reinforced materials is a major obstacle for the implementation of these self-crack-healing ceramics. The Yb2Si2O7-Yb2SiO5-SiC composite is known as a self-healing material that can help to avoid this SiC recession. In this research, the crack-healing behavior of this composite is investigated by using pre-cracking followed by annealing in an oxidizing environment. The crack-healing mechanism is explored and elucidated as a function of the filler morphology, crack size, annealing time, and annealing temperature. The two main crack-healing mechanisms are the filling of cracks with SiO2 glass and the volume expansion of Yb2Si2O7 induced by the reaction between SiO2 and Yb2SiO5. Full crack recovery is achieved with only 10 vol% SiC, with evidence from XRD and EDS analyses. SiC nanoparticulates are more efficient fillers than nanofibers and nanowhiskers.

In the framework of the SELENIUM project, ZrN coated U-Mo fuel kernels were irradiated in the Belgium Reactor 2 of SCK•CEN to a plate average burnup of 48% 235U and a local maximum burnup just below 70% 235U. The microstructure and chemical composition of ZrN coating before and after neutron irradiation have been analysed using scanning transmission electron microscopy (STEM) equipped with energy dispersive X-ray spectroscopy (EDS) and electron energy loss spectroscopy (EELS), and atom probe tomography (APT). The atomic ratio of N/Zr and fission product distribution determined by three techniques were compared and the combination of three techniques shows the advantages in characterization of chemical information for nuclear materials.

Growing interest in molten salts as effective high-temperature heat-transfer fluids for sustainable energy systems drives a critical need to fundamentally understand the interactions between metals and molten salts. This work utilizes the multimodal microscopy methods of synchrotron X-ray nanotomography and electron microscopy to investigate the 3D morphological and chemical evolution of two-model systems, pure nickel metal and Ni-20Cr binary alloy, in a representative molten salt (KCl-MgCl

Palladium is being investigated as a potential additive to metallic fuel to bind fission product lanthanides, with the goal of reducing or preventing fuel-cladding chemical interactions (FCCI). A primary cause of FCCI is the lanthanide fission products moving to the fuel periphery and interacting with the cladding. This interaction will lead to wastage of the cladding and eventually to a cladding breach. The current study is the microstructural characterization of annealed U-20Pu-10Zr-3.86Pd and U-20Pu-10Zr-3.86Pd-4.3Ln in wt. %, where Ln = 53Nd-25Ce-16Pr-6La, using scanning electron microscopy (SEM), transmission electron microscopy (TEM), and X-ray diffraction (XRD). In both alloys, the matrix is comprised of ζ-U0.4Pu0.6 and δ-(U,Pu)Zr2. Based on the matrix compositions, modifications to the room temperature extrapolated U-Pu-Zr ternary phase diagram are suggested. In U-20Pu-10Zr-3.86Pd, there is very little δ phase, due to formation of PdZr2. In U-20Pu-10Zr-3.86Pd-4.3Ln, a lanthanide-rich phase is present, although it does not have the crystal structure for 53Nd-25Ce-16Pr-6La. This phase cannot be identified based on known compounds. (Ln,Pu)Pd is the primary lanthanide phase, with Pu substituting into the crystal structure in place of Nd.

Although the temperature window of helium ion irradiation for gas bubble superlattice (GBS) formation was found to be in the range of approximately 0.15–0.35 melting point in literature, the thermal stability of He GBS has not been fully investigated. This work reports the experiment using an in-situ heating holder in a transmission electron microscope (TEM). A 3.0 mm TEM disc sample of Mo (99.95% pure) was irradiated with 40 keV He ions at 300 °C to a fluence of 1.0E+17 ions/cm2, corresponding to a peak He concentration of approximately 10 at.%, in order to introduce He GBS. In-situ heating was conducted with a ramp rate of ∼25 °C/min, hold time of ∼30 min, and temperature step of ∼100 °C up to 850 °C (0.39Tm homologous temperature). The result shows good thermal stability of He GBS in Mo with no noticeable change on GBS lattice constant and ordering. The implication of this unique and stable ordered microstructure on mechanistic understanding of GBS and its advanced application are discussed.

We report the formation of tetragonal gas bubble superlattice in bulk molybdenum under helium ion implantation at 573 K. The transmission electron microscopy study shows that the helium bubble lattice constant measured from the in-plane d-spacing is ~4.5 nm, while it is ~3.9 nm from the out-of-plane measurement. The results of synchrotron-based small-angle x-ray scattering agree well with the transmission electron microscopy results in terms of the measurement of bubble lattice constant and bubble size. The coupling of transmission electron microscopy and synchrotron high-energy X-ray scattering provides an effective approach to study defect superlattices in irradiated materials.

By successive impregnation method, the Ce-modified Cu-O/γ-Al2O3 catalyst was prepared and characterized using nitrogen adsorption-desorption, scanning electron microscopy energy dispersive X-ray analysis (SEM-EDS), high-resolution transmission electron microscopy (HRTEM), X-ray diffraction (XRD), X-ray photoelectron spectroscopy (XPS), Raman, and H2-Temperature programming reduction (H2-TPR). In catalytic wet-air oxidation (CWAO) process for the printing and dyeing wastewater (PDW), the effects of Ce addition on performance, mechanism and kinetics of the catalyst were investigated. The Ce addition increases the Brunauer-Emmett-Teller (BET) surface area and pore volume of the catalyst and makes the active components uniformly distributed on the catalyst surface. Formation of a stable CuAl2O4 solid solution by anchoring Cu onto the γ-Al2O3 crystal lattice leads to a significant decrease in metal leaching of the Ce-modified catalyst. The proportion of lattice oxygen in the catalyst substantially increases and the apparent activation energy of Cu-O/γ-Al2O3 catalyst decreases owing to Ce addition. Therefore, the catalytic activity and stability of the Ce-modified catalyst are considerably improved. The scavengers experiments identify the active species existed in the CWAO reaction system, with the order of reactivity: h+ > O2•− > H2O2 > HO•. This novel Cu-Ce-O/γ-Al2O3 catalyst has great potential in applications for treatment of concentrated organic wastewater due to its superior catalytic activity and improved stability.

Uranium silicide (U3Si2) is a candidate to replace uranium oxide (UO2) as light water reactor (LWR) fuel because of its higher thermal conductivity and higher fissile density relative to the current standard, UO2. A class of Fe, Cr, Al alloys collectively known as FeCrAl alloys that have superior mechanical and oxidation resistance are being considered as an alternative to the standard Zirconium based LWR cladding. The interdiffusion behavior between FeCrAl and U3Si2 is investigated in this study. Commercially available FeCrAl, along with U3Si2 pellets were placed in diffusion couples. Individual tests were ran at temperatures ranging from 500 °C to 1000 °C for 30 h and 100 h. The interdiffusion was analyzed with an optical microscope, scanning electron microscope, and transmission electron microscope. Uniform and planar interdiffusion layers along the material interface were illustrated with backscatter electron micrographs and energy-dispersive X-ray spectroscopy. Electron diffraction was used to validate phases present in the system, including distinct U2Fe3Si/UFe2 and UFeSi layers at the material interface. U and Fe diffused far into the FeCrAl and U3Si2 matrix, respectively, in the higher temperature tests. No interaction was observed at 500 °C for 30 h.

Palladium is being investigated as a potential additive to metallic fuel to control fuel-cladding chemical interaction (FCCI). A primary cause of FCCI is the lanthanide fission products moving to the fuel periphery and interacting with the cladding. This interaction will lead to wastage of the cladding and, given enough time or burn-up, eventually to a cladding breach. The current study is a scanning electron microscopy (SEM) and transmission electron microscopy (TEM) characterization of annealed U-12Zr-4Pd and U-12Zr-4Pd-5Ln, where Ln = 53Nd-25Ce-16Pr-6La. The present study shows that Pd preferentially binds the lanthanides over other fuel constituents, which may prevent lanthanide migration and interaction with the cladding during irradiation. The SEM analysis indicates the 1:1 Pd-Ln compound is being formed, while the TEM analysis, due to higher resolution, found the 1:1 compound, as well as Pd-rich compounds Pd2Ln and Pd3Ln2.

A series of novel Pt-modified Cu–Fe–La/γ-Al2O3 catalysts were prepared by an incipient-wetness impregnation method, and their performances were evaluated in catalytic wet air oxidation (CWAO) of printing and dyeing wastewater (PDW). The as-prepared catalysts were characterized by X-ray photoelectron spectroscopy (XPS), nitrogen adsorption–desorption, X-ray diffraction (XRD), field emission scanning electron microscopy (FESEM), transmission electron microscopy (TEM), Raman spectroscopy, and temperature programmed reduction by hydrogen (H2-TPR). The effects of catalyst composition and calcination temperature on the catalytic performance were investigated. The chemical oxygen demand (COD) removal efficiency and decolorization rate are satisfactory over Cu1.5Fe1.5La3/γ-Al2O3 catalyst, with poor stability under CWAO conditions. In contrast, catalytic performance and stability of all Pt-modified catalysts are significantly improved. With increasing calcination temperatures from 350 °C to 750 °C, the activity of the catalysts significantly decreases from 97.3% to 84.6%, and the size of the particles varies from 4.71–6.95 nm. The XRD, HRTEM, and XPS results indicate that the Pt1Cu1Fe1La3/γ-Al2O3 catalyst mostly comprises Pt, CuO, Fe3O4, and La2O3, and the incorporation of Pt increases the binding energies (BEs) of the metal elements in the catalyst, owing to synergistic interactions between the Pt and the Cu, Fe, and La elements. After five cycles, the COD removal efficiency and decolorization rate for PDW over the Pt1Cu1Fe1La3/γ-Al2O3 catalyst calcined at 650 °C remain comparable with the fresh catalyst, indicating that Pt-modified Cu–Fe–La/γ-Al2O3 ternary oxide is a promising catalyst with high activity and stability for CWAO treatment.

U3Si2, an advanced fuel form proposed for light water reactors (LWRs), has excellent thermal conductivity and a high fissile element density. However, limited understanding of the radiation performance and fission gas behavior of U3Si2 is available at LWR conditions. This study explores the irradiation behavior of U3Si2 by 300 keV Xe+ ion beam bombardment combining with in-situ transmission electron microscopy (TEM) observation. The crystal structure of U3Si2 is stable against radiation-induced amorphization at 350 °C even up to a very high dose of 64 displacements per atom (dpa). Grain subdivision of U3Si2 occurs at a relatively low dose of 0.8 dpa and continues to above 48 dpa, leading to the formation of high-density nanoparticles. Nano-sized Xe gas bubbles prevail at a dose of 24 dpa, and Xe bubble coalescence was identified with the increase of irradiation dose. The volumetric swelling resulting from Xe gas bubble formation and coalescence was estimated with respect to radiation dose, and a 2.2% volumetric swelling was observed for U3Si2 irradiated at 64 dpa. Due to extremely high susceptibility to oxidation, the nano-sized U3Si2 grains upon radiation-induced grain subdivision were oxidized to nanocrystalline UO2 in a high vacuum chamber for TEM observation, eventually leading to the formation of UO2 nanocrystallites stable up to 80 dpa.

Techniques for training artificial neural networks (ANNs) and convolutional neural networks (CNNs) using simulated dynamical electron diffraction patterns are described. The premise is based on the following facts. First, given a suitable crystal structure model and scattering potential, electron diffraction patterns can be simulated accurately using dynamical diffraction theory. Secondly, using simulated diffraction patterns as input, ANNs can be trained for the determination of crystal structural properties, such as crystal orientation and local strain. Further, by applying the trained ANNs to four-dimensional diffraction datasets (4D-DD) collected using the scanning electron nanodiffraction (SEND) or 4D scanning transmission electron microscopy (4D-STEM) techniques, the crystal structural properties can be mapped at high spatial resolution. Here, we demonstrate the ANN-enabled possibilities for the analysis of crystal orientation and strain at high precision and benchmark the performance of ANNs and CNNs by comparing with previous methods. A factor of thirty improvement in angular resolution at 0.009 degrees (0.16 mrad) for orientation mapping, sensitivity at 0.04% or less for strain mapping, and improvements in computational performance are demonstrated., 19 pages, 12 figures, for PICO 2021 special issue in Ultramicroscopy, the Sixth Conference on Frontiers of Aberration Corrected Electron Microscopy

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.

A facile method to synthesize silicon oxide clad uranium oxide nanowires is presented. U3Si2, used as a precursor, was oxidized to produce uranium oxide nanocrystallites and amorphous silicon oxide under hydrothermal conditions at 300°C and a pressure of 7.8 × 106 Pa. The growth of uranium oxide nanowires was assisted by silicon oxide via assembling the uranium oxide nanocrystallites in an amorphous silicon oxide matrix. The microstructure and composition of silicon oxide clad uranium oxide nanowires were characterized by XRD, SEM, TEM, and EDS. The uranium oxide in the nanowires was determined as UO2.34 with a fluoride cubic structure.

Thermal annealing or radiation induced segregation of solute and impurity elements to grain boundaries (GBs) in metallic alloys changes GB chemistry and thus can alter the GB cohesive strength. In this work, first principles based density functional theory calculations are conducted to study how the segregation of substitutional solute and impurity elements (Al, C, Cr, Cu, P, Si, Ti, Fe, which are present in Ni-based X-750 alloys) influences the cohesive strength of Σ 3 111 , Σ 3 112 , Σ 5 210 and Σ 5 310 GBs in Ni. It is found that C and P show strong embrittlement potencies while Cr and Ti can strengthen GBs in most cases. Other solute elements, including Si, have mixed but insignificant effects on GB strength. In terms of GB character effect, these solute and impurity elements affect the GB strength of the Σ 5 210 GB most and that of the Σ 3 111 least. Detailed analyses of solute-GB chemical interactions are conducted using electron localization function, charge density map, partial density of states, and Bader charge analysis. The results suggest that the bond type and charge transfer between solutes and Ni atoms at GBs may play important roles on affecting the GB strength. For non-metallic solute elements (C, P, Si), their interstitial forms are also studied but their effects on GB strength are weaker than their substitutional counterparts.

Laser-generated ultrasound was used to monitor microstructure evolution during thermal processing of as-cast, polycrystalline binary uranium-zirconium metallic fuel alloys with compositions of U-20wt.%Zr (U-20Zr), U-50wt.%Zr (U-50Zr) and U-80wt.%Zr (U-80Zr). Ultrasonic waveforms were recorded during heating and cooling the samples from room temperature to >973 K and back. A phase transition temperature for all three compositions was estimated from the temperature at which an abrupt and rapid reduction in ultrasonic velocities was observed. Microstructural features on the length scale of tens of micrometers were inferred from the observation of scattering of ultrasonic waves by elastic heterogeneities above ~823 K in U-20Zr, while a hysteresis in the ultrasonic velocities of U-80Zr upon cooling was attributed to a partial retention of the high temperature phase following thermal annealing. The U-50Zr alloy exhibited a reversible viscoelastic response above 933 K, as evidenced by the observation of high frequency attenuation of the shear component of the waveforms at high temperature. Ultrasonic measurements were supplemented by in situ transmission electron microscopy (TEM). The TEM images revealed that the δ-U-Zr matrix in the three compositions underwent a spinodal decomposition above ~823 K into nanoscale regions. The ultrasonic measurements revealed larger, micron-scale structure evolution in the U-20Zr alloy at the same temperature. This large-scale structure is associated with heterogeneous regions having different Zr content. Our findings show the potential heating rate dependence of microstructural evolution in U-Zr alloys and highlight differences in the thermomechanical response and associated length scales during thermal annealing between single- and dual-phase compositions. These results demonstrate the utility of laser ultrasonics to rapidly and efficiently scan phase boundaries and monitor micrometer-scale structure evolution in metallic fuel alloys.

A comprehensive characterization study was carried out on 1 and 4 cycles X2®, aiming to reveal effect of irradiation-induced alloying element redistribution on the in-reactor corrosion kinetics. Using a combination of (scanning) transmission electron microscopy ((S)/TEM) and atom probe tomography (APT), the results strongly evidenced the existence of Nb-rich native precipitates and irradiation-induced platelets (IIPs)/nanoclusters in both metal and oxide. In addition, the suboxide may reject Nb content back into the metal matrix to achieve a thermodynamic equilibrium state. On the other hand, Fe-rich transition cracks were observed for the first time, along with Fe-rich oxide grain boundaries.

This work focuses on the full characterization of dislocation loops induced by proton irradiation in single-crystal ThO2. Irradiation was performed using 2 MeV H+ ions with sample temperature at 600oC and a dose of up to 0.47 displacements per atom (dpa). Transmission electron microscopy (TEM) characterization was performed on a large number of dislocation loops. Burgers vector ( b → ) analysis using standard g → . b → = 0 invisibility criterion revealed different variants of 1 / 3 〈 111 〉 type dislocation loops. TEM analysis of edge-on dislocation loops was used to determine habit planes as { 111 } type. The nature of dislocation loops was revealed using the inside-outside contrast method as interstitial type. Rel-rod dark field images were obtained by selecting streak at g → = 1 / 2 [ 31 1 ¯ ] in diffraction pattern to identify the Frank loops present in the microstructure. Subsequent analysis of a single dislocation loop using atomic resolution scanning transmission electron microscopy (STEM) confirmed the interstitial nature of the loop with { 111 } habit plane.

Advanced nuclear reactor concepts aim to use fuels that must withstand unprecedented temperature and radiation extremes. In these fuels, thermal energy transport under irradiation is directly related to fuel longevity, reactor safety, and is arguably one of the most important performance metrics. Here we provide a comprehensive, first-principles-informed treatment of phonon mediated thermal transport in a defect-bearing actinide oxide with direct comparison to experimental measurements. Pristine and proton irradiated thorium dioxide was chosen as a model system to treat the complexity of thermal transport in the presence of lattice defects. A thermal transport model is implemented using the linearized Boltzmann transport equation (LBTE) with input from first principles calculations and defect evolution models. Density functional theory is used to calculate phonon dispersion in thorium dioxide and used as an input to calculate both intrinsic and extrinsic, defect-induced relaxation times. In addition, a defect evolution model is benchmarked using microstructure characterization of as-irradiated thorium dioxide using a combination of electron microscopy and optical spectroscopy. The output of the LBTE is compared directly to mesoscopic measurements of thermal conductivity on length scales commensurate with defect accumulation. Parametric measurements of conductivity with irradiation dose and temperature suggest a saturation in the reduction of thermal conductivity with increasing defect generation, which is partially captured in our defect evolution model and LBTE framework. This comprehensive, atomistic- to meso-scale treatment provides the necessary basis to investigate thermal transport under irradiation in more complex systems that exhibit strong electron correlation.

Engineered spinodal decomposition and spinodal precursors to precipitation hardening are effective approaches for tailoring material's properties and performance. These approaches leverage the spinodal and binodal nature of miscibility gaps. However, little is known about the chemical and crystallographic mechanisms controlling phase evolution across the spinodal curve and into the binodal regime. This study aims to reveal spinodal-binodal phase transformation in a binary U-50 wt.% Zr model alloy through coupled X-ray diffraction (XRD), in-situ transmission electron microscopy (TEM), and atom probe tomography (APT). The hexagonal ⍵-UZr2+x phase initially undergoes a spinodal decomposition into interconnected nanosized hexagonal Zr-rich and hexagonal U-rich domains at ~575 °C, near the hexagonal-bcc phase transformation. Between 600 and 800 °C, the microstructure evolution is dominated by the coarsening and chemical purification of decomposed bcc domains and a transition from spinodal to binodal nucleation. At 800 °C, the metastable Zr domains transfer to stable α-Zr phase and stay the same phase up to 1000 °C. On the contrary, metastable U domains transfer to orthorhombic α’-U firstly at 800 °C and then to tetragonal β-U at 1000 °C. Spinodal decomposed domain size is found to be linearly related to extrinsic specimen geometry. The effect of oxygen on spinodal decomposition is also discussed. In-situ TEM observations successfully captured the spinodal decomposition and the subsequent transition to binodal phase separation with temperature, providing critical evidence for an expanded miscibility gap, concerning both composition and temperature, on the U-Zr phase diagram.

This work utilizes electron energy loss spectroscopy (EELS) to identify oxidation state of alloying elements in Ni-based alloys after exposure to molten chloride salt systems. Pure Ni and Ni-20Cr model alloy were corroded in molten ZnCl2 and KCl-MgCl2 under argon atmosphere at various temperatures. Oxidation states of Cr (Cr3+) and Ni (Ni2+) in the molten salt after corrosion were determined by monitoring changes in the L2,3 edges of corresponding EELS spectra. Oxidation state mapping technique using principal component analysis and multiple linear least squares fitting in HyperSpy Python package was developed.

A two-step processing was developed to prepare Yb2Si2O7-SiC nanocomposites. Yb2Si2O7-Yb2SiO5-SiC composites were first fabricated by a solid-state reaction/hot-pressing method. The composites were then annealed at 1250°C in air for 2 hours to activate the oxidation of SiC, which effectively transformed the Yb2SiO5 into Yb2Si2O7. The surface cracks purposely induced can be fully healed during the oxidation treatment. The treated composites have improved flexural strength compared to their pristine composites. The mechanism for crack healing and silicate transformation have been proposed and discussed in detail.

Fuel swelling during normal reactor operations could lead to unfavorable chemical interactions when in contact with its cladding. As new fuel types are developed, it is crucial to understand the interaction behavior between fuel and its cladding. Diffusion experiments between U3Si2 and Zricaloy-4 (Zry-4) were conducted at 800 and 1000 °C up to 100 h. The microstructure of pristine U3Si2 and U3Si2/Zry-4 interdiffusion products were examined using scanning electron microscopy (SEM) and transmission electron microscopy (TEM) equipped with an energy dispersive X-ray spectroscopy (EDS) system. The primary interdiffusion product observed at 800 °C is ZrSi2, with secondary phases of U-Zr in the Zry-4, and Fe-Cr-W-Zr-Si phases at Zry-4/ZrSi2 interface and Fe-Cr-U-Si phases at ZrSi2/U-Si interface. The primary interdiffusion products at 1000 °C were Zr2Si, U-Zr-Fe-Ni, U, U-Zr, and a low melting point phase U6Fe.

Herein we report on the formation of defects in response to neutron irradiation of polycrystalline Ti3SiC2 and Ti3AlC2 samples exposed to total fluences of ≈6 × 1020 n/m2, 5 × 1021 n/m2 and 1.7 × 1022 n/m2 at irradiation temperatures of 121(12), 735(6) and 1085(68)°C. These fluences correspond to 0.14, 1.6 and 3.4 dpa, respectively. After irradiation to 0.14 dpa at 121 °C and 735 °C, black spots are observed via transmission electron microscopy in both Ti3SiC2 and Ti3AlC2. After irradiation to 1.6 and 3.4 dpa at 735 °C, basal dislocation loops, with a Burgers vector of b = ½ [0001] are observed in Ti3SiC2, with loop diameters of 21(6) and 30(8) nm after 1.6 dpa and 3.4 dpa, respectively. In Ti3AlC2, larger dislocation loops, 75(34) nm in diameter are observed after 3.4 dpa at 735 °C, in addition to stacking faults. Impurity particles of TiC, as well as stacking fault TiC platelets in the MAX phases, are seen to form extensive dislocation loops under all conditions. Cavities were observed at grain boundaries and within stacking faults after 3.4 dpa irradiation, with extensive cavity formation in the TiC regions at 1085 °C. Remarkably, denuded zones on the order of 1 μm are observed in Ti3SiC2 after irradiation to 3.4 dpa at 735 °C. Small grains, 3–5 μm in diameter, are damage free after irradiation at 1085 °C at this dose. The results shown herein confirm once again that the presence of the A-layers in the MAX phases considerably enhance their irradiation tolerance. Based on these results, and up to 3.4 dpa, Ti3SiC2 remains a promising candidate for high temperature nuclear applications as long as the temperature remains >700 °C.

Uranium mononitride (UN) with 5 wt.% uranium dioxide (UO2) is used as a model system to study the phase and defect evolution under proton irradiation in nitride-oxide composite. Phase composition, crystallographic orientation relationships (ORs) and dislocation loops were characterized using X-ray diffraction, transmission electron microscopy, and energy dispersive X-ray spectroscopy techniques. Proton-irradiation at elevated temperatures promoted the transformation of UN into uranium sesquinitride (U2N3) and UO2 phases. U2N3 and UO2 formed a fully coherent structure with two ORs: {002}U2N3 parallel to{002}UO2 and [001]U2N3 parallel to[001]UO2; U2N3{101}parallel to UO2{101} and U2N3[101]parallel to UO2[101] due to low lattice misfit (2.3%) and low interfacial energy (127 mJ/m(2)). Observed oxidation of UN and coherent interface are consistent with density-functional theory calculations which suggest lower energy for oxidized configuration and low energy of the interface. The dislocation loops grew while their number density decreased with the temperature and dose. The loop size was over three times larger in two nitride phases than that in UO2, while the number density was one order of magnitude higher in UO2 than in nitride phases. Loop density and diameter were analyzed using a kinetic rate theory that considers stoichiometric loop evolution. This analysis led to the conclusion in all compounds loop growth is governed by mobility of uranium interstitials, and enabled measurement of diffusion coefficients of uranium interstitials and non-metal interstitials and vacancies. This analysis provided a comparative study of early stage of microstructure evolution under irradiation which has implications for use of this mixture as advanced fuel in nuclear energy systems.

A rate theory model informed by multimodal characterization is used to evaluate the concentration of point defects in irradiated materials. Cerium dioxide (CeO2) is used as a model ionic compound, whose cation and anion sublattice point defects evolve independent of each other, but extended defects in the form of dislocation loops retain stoichiometry of the compound. To demonstrate this, we performed extensive measurement of defect evolution in CeO2 exposed to energetic protons at elevated temperature. Two sintered polycrystalline CeO2 samples were irradiated with protons having energies up to 2.5 MeV. Both samples were irradiated at 600°C to a dose of 0.14 dpa, but with different dose rates. These irradiation conditions produced a rich microstructure with resolvable extended defects and a significant concentration of point defects. Dislocation loop density revealed by electron microscopy and lattice constant changes measured by X-ray diffraction (XRD), and mesoscale thermal conductivity measurements were used to parameterize the rate theory model. The model, which includes point defect generation, recombination, and clustering into stoichiometric interstitial loops, suggests a large concentration of cerium vacancies and interstitials is present under these irradiation conditions. This work lays the foundation for expanded multimodal characterization of microstructure, including more direct characterization of point defects using optical spectroscopies.

As an alternative fuel form, the annular metallic fuel design eliminates the liquid sodium bond between the fuel and the cladding, providing back-end fuel cycle and other benefits. The fuel-cladding chemical interaction (FCCI) of annular fuel also presents new features. Here, state-of-the-art electron microscopy and spectroscopy techniques were used to study the FCCI of a prototype annular U-10wt%Zr (U-10Zr) fuel with ferritic/martensitic HT-9 cladding irradiated to 3.3% fission per initial heavy atom. Compared with sodium-bonded solid fuels, negligible amounts of lanthanides were found in the FCCI layer in the investigated helium-bonded annular fuel. Instead, most lanthanides were retained in the newly formed UZr2 phase in the fuel center region. The interdiffusion of iron and uranium resulted in tetragonal (U,Zr)6Fe phase (space group I4/mcm) and cubic (U,Zr)(Fe,Cr)2 phase (space group Fd 3 ¯ m). The (U,Zr)(Fe,Cr)2phase contains a high density of voids and intergranular uranium monocarbides of NaCl-type crystal structure (space group Fm 3 ¯ m). At the interdiffusion zone and inner cladding interface, a porous lamellar structure composed of alternating Cr-rich layers and U-rich layers was observed. Next to the lamellar region, the unexpected phase transformation from body-centered cubic ferrite (α-Fe) to tetragonal binary Fe-Cr σ phase (space group P42/mnm) occurred, and tetragonal Fe-Cr-U-Si phase (space group I4/mmm) was identified. Due to the diffusion of carbon into the interdiffusion zone, carbon depletion inside the HT-9 led to the disappearance of the martensite lath structure, and intergranular U-rich carbides formed as a result of the diffusion of uranium into the cladding. These detailed new findings reveal the unique features of the FCCI behavior of annular U-Zr fuels, which could be a promising alternative fuel form for high burnup fast reactor applications.

Corrosion of Ni and Ni-20Cr in molten ZnCl2 under argon atmosphere at 593 K was investigated using cyclic polarization in combination with various electron microscopy techniques. Cyclic polarization measurements yielded significantly lower corrosion rates for pure Ni than for Ni-20Cr. Corrosion of the Ni-20Cr alloy is attributed to selective anodic dissolution of Cr, which forms a chromium chloride film at the metal-salt interface. The corrosion in Ni and Ni-20Cr is accelerated by grain boundary attack through intergranular corrosion, followed by the formation of pore-salt network regions. The related corrosion mechanisms and dissolution kinetics are discussed.

To develop metallic fuel with ultra-high burnup of 30%-40%, an annular U-10Zr fuel with 55% smear density was fabricated through a casting route and irradiated at the Advanced Test Reactor at Idaho National Laboratory. The annular fuel design also serves as a demonstration of the feasibility to replace the sodium bond with a helium bond to benefit the geological disposal of irradiated fuel, cut the cost of fuel fabrication, and boosts the overall metallic fuel economy. This paper reports the results from transmission electron microscopy (TEM) based post-irradiation examination of this fuel type irradiated to a burnup of 3.3% fissions per initial heavy metal atoms. The low burnup was planned for initial screening of this fuel design. After irradiation, the initial U-10Zr microstructure separated into an α-U annular region and an UZr2+x center region with a nanoscale spinodal decomposed microstructure. Because of the large amount of interface areas created in this microstructure, the fission gas atoms and vacancies generated in the UZr2+x phase are possibly pinned at the interface areas, leading to ~20 times smaller fission gas bubbles than those in the neighboring α-U. The large bubbles in α-U become connected and merged into large pores that provide fast paths for fission gas release into capsule plenum which prevents further swelling of fuel slug. The fuel slug center still has open space to accommodate further fuel swelling from solid fission products at higher burnup. Other neutron irradiation induced phase and microstructure change are also characterized and compared with traditional solid fuel designs.

Thermal transport is a key performance metric for thorium dioxide in many applications where defect-generating radiation fields are present. An understanding of the effect of nanoscale lattice defects on thermal transport in this material is currently unavailable due to the lack of a single crystal material from which unit processes may be investigated. In this work, a series of high-quality thorium dioxide single crystals are exposed to 2 MeV proton irradiation at room temperature and 600 °C to create microscale regions with varying densities and types of point and extended defects. Defected regions are investigated using spatial domain thermoreflectance to quantify the change in thermal conductivity as a function of ion fluence as well as transmission electron microscopy and Raman spectroscopy to interrogate the structure of the generated defects. Together, this combination of methods provides important initial insight into defect formation, recombination, and clustering in thorium dioxide and the effect of those defects on thermal transport. These methods also provide a promising pathway for the quantification of the smallest-scale defects that cannot be captured using traditional microscopy techniques and play an outsized role in degrading thermal performance.

We report on imaging subsurface grain microstructure using picosecond ultrasonics. This approach relies on elastic anisotropy of crystalline materials where ultrasonic velocity depends on propagation direction relative to the crystal axes. Picosecond duration ultrasonic pulses are generated and detected using ultrashort light pulses. In materials that are transparent or semitransparent to the probe wavelength, the probe monitors gigahertz frequency Brillouin oscillations. The frequency of these oscillations is related to the ultrasonic velocity and the optical index of refraction. Ultrasonic waves propagating across a grain boundary experience a change in velocity due to a change in crystallographic orientation relative to the ultrasonic propagation direction. This change in velocity is manifested as a change in the Brillouin oscillation frequency. Using the ultrasonic propagation velocity, the depth of the interface can be determined from the location in time of the transition in oscillation frequency. A subsurface image of the grain boundary is obtained by scanning the beam along the surface. We demonstrate this subsurface imaging capability using a polycrystalline UO2 sample. Cross section liftout analysis of the grain boundary using electron microscopy was used to verify our imaging results.

Herein we report on the characterization of defects formed in polycrystalline Ti 3 SiC 2 and Ti 2 AlC samples exposed to neutron irradiation – up to 0.1 displacements per atom (dpa) at 350 ± 40 °C or 695 ± 25 °C, and up to 0.4 dpa at 350 ± 40 °C. Black spots are observed in both Ti 3 SiC 2 and Ti 2 AlC after irradiation to both 0.1 and 0.4 dpa at 350 °C. After irradiation to 0.1 dpa at 695 °C, small basal dislocation loops, with a Burgers vector of b = 1/2 [0001] are observed in both materials. At 9 ± 3 and 10 ± 5 nm, the loop diameters in the Ti 3 SiC 2 and Ti 2 AlC samples, respectively, were comparable. At 1 × 10 23 loops/m 3 , the dislocation loop density in Ti 2 AlC was ≈1.5 orders of magnitude greater than in Ti 3 SiC 2 , at 3 × 10 21 loops/m 3 . After irradiation at 350 °C, extensive microcracking was observed in Ti 2 AlC, but not in Ti 3 SiC 2 . The room temperature electrical resistivities increased as a function of neutron dose for all samples tested, and appear to saturate in the case of Ti 3 SiC 2 . The MAX phases are unequivocally more neutron radiation tolerant than the impurity phases TiC and Al 2 O 3 . Based on these results, Ti 3 SiC 2 appears to be a more promising MAX phase candidate for high temperature nuclear applications than Ti 2 AlC.

Irradiation induced phase transformation of γ-austenite to α-ferrite has been observed in austenitic steels for the past several decades. This transformation can be detrimental to structural materials in a nuclear reactor environment as the increased fraction of the ferritic phase can increase corrosion and embrittlement and lead to stress corrosion cracking. This transformation is caused by both strain induced martensite transformation as well as radiation induced segregation and precipitation. In this study, two radiation tolerant nanostructured 304L austenitic steels (one ultrafine grained and one nanocrystalline) were manufactured using severe plastic deformation. These nanostructured 304L steels were compared to conventional coarse-grained 304L, after self-ion irradiation at 500°C up to a peak damage of 50 displacements per atom. Phase fraction after irradiation was analyzed using grazing incidence x-ray diffraction, precession electron diffraction, and electron backscatter diffraction. Nanostructured 304L steels showed significant resistance to irradiation induced austenite to ferrite transformation. This resistance was shown to be due to a decrease in defect formation, as well as a reduction in radiation induced segregation and precipitation.

The effect of proton irradiation on corrosion rate of α-annealed and β-quenched Zr-0.5Nb alloys is investigated. The major focuses of this study are to understand i) if the nucleation of irradiation-induced platelets (IIPs)/nanoclusters requires dissolution of Nb-rich native precipitates, ii) if the irradiated native precipitates and interlaths are stable in the oxide, and iii) how much Nb content in the solid solution is suitable to lower the corrosion rate for Zr-Nb alloys. To answer these questions, the major characterization techniques used in this study are APT and (S)TEM/EDS to study the microstructure and microchemistry evolution following irradiation and oxidation.

Degradation mechanism of lead-vanado-iodoapatite (IAPT) in NaCl solution at 90 °C was investigated through systematic characterization of surface morphology, microstructure and microchemistry evolution of the alteration layer. Nano-scale characterization indicated that IAPT crystals degraded from grain boundaries toward inner grains, forming ultrafine Cl-bearing crystallites. A coupled interface dissolution-reprecipitation replacement mechanism and the ion exchange between Cl− and I− across the degradation zone may together play roles in determining the nano-scale degradation behavior of IAPT. This work highlights the degradation through multiple reactions and elemental transport across the liquid-solid and solid-solid interfaces at length scales from sub-millimeter to nanoscale.

A thermoreflectance method was used to measure intragranular thermal diffusivity and conductivity of samples having a composition of U-50 wt%Zr (U–50Zr). Three phases at this composition were investigated: the high temperature γ phase, the low temperature δ phase, and the metastable ω phase. This approach uses a tightly focused laser to inject micron scale thermal waves and a second tightly focused laser to monitor the temperature distribution. The thermal properties are extracted by comparing experimental temperature profiles to an analytical heat diffusion model. The probe laser can monitor the temperature field in orthogonal directions along the surface of the sample and is well suited to measure thermal anisotropy. We show that the δ phase exhibits significant thermal anisotropy. The γ phase has the highest thermal conductivity. The higher conductivity of the γ phase is thought to be due to the presence of Zr precipitates that slightly change the stoichiometry of the γ matrix. The highly disordered ω phase appears to be thermally isotropic and has a lower conductivity than the δ phase. Both observations are likely due to the presence of γ domains that reside between ω domains. This supposition is supported by the presence of thermal heterogeneities that appear as noise in the measured signals.

The purpose of this work is to develop a model for normal grain growth in U3Si2. The average grain boundary energy was determined from previously published molecular dynamics simulations. The grain growth kinetics were quantified at various temperatures by annealing nanocrystalline samples. The mobility was determined by comparing phase field grain growth simulations to the experimental data. From these various methods, we found that the average grain size D in U3Si2 can be estimated over time t using the equation D 2 − D 0 2 = 2 α M γ t , where D 0 is the initial average grain size, the geometry factor α = 0.96 , the average grain boundary mobility M = 6.30 × 10 − 18 e − 0.33 [ eV ] k b T m 4 / ( Js ) with the Boltzmann constant k b and temperature T, and the average grain boundary energy has been found as a function of temperature, e.g. γ ¯ = 0.83 J/m2 at 673 K.

The composition and crystal structure of the “Joint Oxyde Gaine” (JOG) has been investigated by means of electron microscopy. Microstructural characterization reveals a highly heterogeneous porous structure with inclusions containing both fission products and cladding components. Major fission products detected, other than Cs and Mo, are Te, I, Zr and Ba. The layer is composed by sub-micrometric crystallites. The diffraction data refinement, together with chemical mapping, confirms the presence of Cs2MoO4, which is the major component of the JOG. However, combinatorial analyses reveal that other non-stoichiometric phases are possible, highlighting the complex nature of the crystalline structure of the JOG. Fe is found in metallic Pd-rich precipitates with structure compatible with the tetragonal structure of FePd alloy. Cr is found in different locations of the JOG, in oxide form, but no structural data could be obtained due to local beam sensitization of the sample in those areas.

Finely dispersed two phase microstructures resulting from a spinodal decomposition are of interest as they are associated with enhanced mechanical properties and excessive interfaces to mitigate defect related behavior. This study reports a spinodal-like phase decomposition in a U–50Zr alloy by thermal annealing at 620 °C and ion irradiation at 550 °C, with the latter temperature too low to initiate pure thermal phase transformation. The results hold broad impact for U–Zr alloy systems and its application as advanced nuclear fuel.

The microstructure of EBR-II neutron-irradiated austenitic stainless steel 304 and nickel-base alloy X-750 was investigated. Both alloys were irradiated at low dose rates (∼2 × 10−8 dpa/s) to a neutron fluence of 6.9 × 1022 n/cm2 (E > 0.1 MeV) at 371–389 °C. Different types of defects, including Frank loops, cavities, and precipitates were characterized. The Frank loops in Type 304 stainless steel (SS) are larger in size (∼50 nm in diameter) and lower in number density (2.58 × 1021 m−3), compared to most previous higher dose rate neutron irradiation studies. The Frank loops in X-750 have an average size 26.0 nm of and a number density of 9.44 × 1021 m−3. In 304 SS and X-750, cavities are of ∼20 nm and ∼14 nm in diameter, respectively. The swelling of both alloys was found to be insignificant. In 304 SS, Ni and Si were found enriched at the cavity surfaces and Ni,Si-rich precipitates were also found. Multivariate statistical analysis using non-negative matrix factorization reveals that these Ni,Si-rich precipitates contain only ∼5.7 at.% Si, differing from the Ni3Si γ ’ precipitates found in several previous studies. In X-750, L12-structured γ ’ precipitates were found, and multivariate statistical analysis confirmed the 3:1 stoichiometry (Ni3(Ti,Al)) of the γ ’ precipitates and the superlattice reflections confirmed the stability of the crystal structure of these γ ’ precipitates, indicating higher-than-expected precipitate stability under high-dose neutron irradiation.

Generation IV reactor structural materials will be exposed to high doses and temperatures during reactor operation that may lead to irradiation-induced degradation. This degradation will differ from that seen in light water reactors and therefore understanding mechanisms controlling material performance during irradiation is critical for evaluating the viability of Generation IV nuclear reactor concepts. This chapter discusses irradiation effects and microstructural changes that affect mechanical properties and dimensional stability of Generation IV reactor materials.

Copper-based heterogeneous catalyst is a kind of catalyst which is created by the effect that copper, ferrum, palladium, lanthanum together with their corresponding oxides have on the carriers such asγ-Al2O3. With the development of related research in recent years, its applications and theories are becoming more and more abundant. The properties, structures and designs of experiments of copper-based heterogeneous catalysts are reviewed and the establishment of applied kinetics equations is briefly introduced in this paper.