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Environmental protection treatment process relies on the robustness, durability, and performance of catalysts to drive the development of cutting-edge sustainable technologies for the elimination of refractory contaminants. Herein, prepared SrCoO2.5 were successfully prepared through citric acid by sol-gel method, and further utilized as super catalysts to degrade Rhodamine B (RhB) by coupling with peroxymonosulfate (PMS) in environmental protection. Three optimized parameters (SrCoO2.5 dosage of 0.2 g L−1, PMS concentration of 0.93 g L−1, and initial pH of 2) were obtained via Conditional test method in environmental protection. Benefiting from the larger specific surface area, pore-volume, and existence of abundant hydroxyl groups, SrCoO2.5 prepared through citric acid by sol-gel method with more available active sites exhibited an super efficiency of 99.4% toward catalytic degradation of RhB within 10 min under the optimal conditions. Moreover, SrCoO2.5 demonstrated durability and long-term stability even during the seven consecutive cycle, it almost regenerates in 500 °C for 1 h. The scavenging experiments and electron paramagnetic resonance technologies revealed that non-radical singlet oxygen (1O2), sulfate radicals(SO4•−) were associated as active species in the SrCoO2.5/PMS system. Besides, the reaction mechanism on the SrCoO2.5 degradation pathways toward RhB was speculated under PMS activation. The results indicated that the synergistic effects between Co–Sr structures not only significantly boosted the removal efficiency and long-term stability of SrCoO2.5. But also facilitated the redox cycles of Co3+/Co2+ and Sr3+/Sr2+, which produce 1O2. This proof-of-concept approach to develop such high-efficient Co–Sr structures produced Co3+/Co2+ and Sr3+/Sr2+ will open up novel avenues for wastewater decontamination via PMS activation.

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.

Lithium–sulfur batteries are very promising for next-generation energy storage. However, most studies use flooded electrolytes to achieve a high specific capacity at the expense of lowering the spe...

Fission product lanthanides in metallic fuels are known to cause adverse fuel-cladding chemical interaction (FCCI). Palladium (Pd) is being explored as a potential additive to reduce or mitigate FCCI by forming stable Pd-Ln compounds. The current study is an investigation of diffusion behaviors between UZrPd, UZrPdLn, UPuZrPd, and UPuZrPdLn alloys (Ln = 53Nd-25Ce-16Pr-6La wt. %) and iron (Fe). Diffusion couple tests were performed followed by microstructural analysis using scanning electron microscopy (SEM) and transmission electron microscopy (TEM). It is found that the diffusion of Fe into the fuel is reduced by PdZr2 precipitates randomly dispersed throughout the fuel matrix. Ln-Pd compounds form in UZrPdLn alloys, and (Ln,Pu)Pd forms in the UPuZrPdLn alloy. In both diffusion couples with Ln, small amounts of lanthanides are diffusing into the Fe. This could be due to small amounts of dissolved lanthanides in the precipitates, or could be due to the Ln-Pd compounds decomposing. In either case, diffusion caused by lanthanides is greatly reduced as compared to alloys without Pd.

Proton (H+) irradiation effects in polycrystalline UO2 have been studied. The irradiation was carried out using three ion energies and two different ion fluxes at 600 °C. Scanning electron microscopy (SEM) investigations showed that significant surface flaking took place. Focused ion beam (FIB) milling in SEM was successfully applied for extracting lamellas from uneven blistered surfaces for transmission electron microscopy (TEM) investigations allowing detailed investigations for the degradation mechanisms. High-resolution TEM for the flaked UO2 surfaces revealed that the implanted H+ formed sharp two-dimensional cavities at the peak ion-stopping region instead of diffusing to the matrix. The resulting lateral stress likely caused UO2 surface deterioration in good agreement with previous blistering and flaking studies on crystalline materials. Graphical abstract

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.

Sodium-cooled Fast Reactors (SFR) are one of the advanced nuclear reactor concepts to be commercially applied for electricity production. The benefits of SFR are well-known and include: the possibility of a closed fuel cycle, proliferation resistance, nuclear waste minimization via actinides burning, and fissile breeding capabilities. Metallic fuel used in SFR has well demonstrated irradiation performance. However, more studies are necessary to optimize and extend operational and safety limits for their commercialization and licensing. This could be achieved through a better understanding of fuel behaviors during transient and of fuel failure thresholds. This paper describes the experimental Research and Development (R&D) program aimed at providing the necessary data to support the development of SFR-optimized safety limits. This program integrates separate effects testing and integral effects testing, combined with advanced Modeling and Simulation (M&S). This R&D program, finally, focuses on delivering the science-based information necessary for supporting the licensing and utilization of SFR based on metallic fuel. In this paper we will describe the three research areas centered on fuel development and focused on separate effect testing, namely: (1) microstructural, chemistry, and material properties; (2) thermo-mechanical behavior; and (3) source term and fission product behavior. Preliminary results from these Separate Effect Tests (SET) studies and the current instruments and experimental plan are also presented.

In this study, we investigated the type of dislocation loops formed in the proton-irradiated uranium-nitrogen-oxygen (U-N-O) system, which involves uranium mononitride (UN), uranium sesquinitride (α-U2N3), and uranium dioxide (UO2) phases. The dislocation loop formation is examined using specimens irradiated at 400°C and 710°C. Based on the detailed transmission-based electron microscopy characterization with i) the morphology-based on-zone and ii) the invisibility-criterion based two-beam condition imaging techniques, only a single type of dislocation loop in each phase is found: a/2⟨110⟩, a/2⟨111⟩, or a/3⟨111⟩ dislocation loops in UN, α-U2N3, and UO2 phases, respectively. Molecular statics calculations for the formation energy of perfect and faulted dislocation loops in the UN phase indicate a critical loop size of ∼ 6 nm, above which perfect loops are thermodynamically favorable. This could explain the absence of faulted loops in the experimental observation of the irradiated UN phase at two temperatures. This work will enhance the understanding of irradiation induced microstructural evolution for uranium mononitride as an advanced nuclear fuel for the next-generation nuclear reactors.

In situ monitoring of corrosion processes is important to fundamentally understand the kinetics and evolution of materials in harsh environments. A quasi in situ transmission electron microscopy technique was utilized to study microstructural and chemical evolution of a Ni-20Cr disc sample exposed to molten KCl-MgCl2 salt for 60 s in consecutive 20 s iterations. In situ synchrotron X-ray nanotomography was performed to characterize the morphological evolution of a Ni-20Cr microwire exposed to molten KCl-MgCl2. Both techniques captured key corrosion events and revealed mechanisms at different time and length scales, potentially bringing greater insights and deeper understanding beyond conventional analysis.

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

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 dpa using in-situ TEM. Cryogenic irradiation resulted in small defect clusters and faulted dislocation loops as large as 12 nm in 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 accumulation 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.

As an alternative fuel form, annular metallic fuel design eliminated 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 (FIMA). Compared to sodium-bonded solid fuels, negligible amounts of lanthanides were found in the FCCI layer in the investigated helium bonded annular fuel. Instead, most of the lanthanides were retained in the newly formed UZr2 phase at the fuel center region. The interdiffusion of iron and uranium resulted in tetragonal (U,Zr)6Fe phase (space group I4/mcm) and cubic (U,Mo)(Fe,Cr)2 phase (space group Fd3 m). The (U,Mo)(Fe,Cr)2 phase contains a high density of voids and intergranular uranium monocarbides of NaCl-type crystal structure (space group Fm3 m). At the interface of the interdiffusion zone and inner cladding, 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 (BCC) 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.

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.

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 performed at temperatures ranging from 500 °C to 1000 °C for 30 hours and 100 hours. 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. While no interaction was observed at 500 °C, interaction was observed at and above 600 °C. The amount of interaction between the two materials enhanced progressively at elevated temperatures.

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.

Structure of the pellet/cladding-interaction (PCI) layer of a high-burnup nuclear fuel with niobium-bearing cladding was investigated using modern transmission electron microscopy (TEM) techniques. Morphology of the PCI consisted of uniform zirconia layer on the cladding side and finger-like features towards to the ceramic fuel. The PCI layer showed a complex microstructure that constitutes three distinct zirconia layers formed by two zirconia phases. From cladding to fuel, monoclinic, tetragonal, and another tetragonal phase were determined via TEM diffraction patterns and precession electron diffraction (PED) using ASTAR microscopy platform (AMP). Characterizations showed that monoclinic and tetragonal phases close to the cladding were crack-free, but the tetragonal layer adjacent to the fuel contained cracks and pores. Presence of the monoclinic layer suggested that its formation involved the gaseous diffusion of oxygen below threshold dose of fission product damage before the fuel/cladding contact occurred. Furthermore, high-resolution TEM revealed compound phase of (U,Zr)O2 at fringes of zirconia fingers on the fuel side.

Ferritic Martensitic (F/M) steel HT9 specimens were irradiated in the Advanced Test Reactor up to 4.16 dpa in three temperature ranges (roughly from 300 to 600 °C). The post-irradiation microstructure, including dislocation structure, precipitation and radiation-induced segregation (RIS) was characterized using analytical scanning / transmission electron microscopy (S/TEM), and atom probe tomography (APT). Irradiation hardening was measured using nanoindentation. The results reveal a distinctive pattern of dislocation and precipitate evolution at high temperature, around 600 °C, where various defects and precipitates formed in the low dose regime followed by a recovering process with increasing dose. Dislocation loops formed in all temperature ranges, and the growth of dislocation loops is unconstrained above certain critical temperature, contributing to the increasing dislocation density even prior to doses of 0.5 dpa at 600 °C. Ni/Mn/Si clusters were identified in all temperature ranges and the compositions of these clusters converged to G phase stoichiometrically. Significant coarsening of G phase particles was observed at 600 °C, accompanied by the formation of G phase on grain boundaries. α’ precipitates were only found in the medium and low temperature ranges (below 500 °C). The number density and volume fraction were higher in the low temperature specimens, while larger particles were observed in the medium temperature range. RIS of Cr, Ni, Mn, Si, P was identified at dislocation lines, grain boundaries and phase boundaries, and the temperature dependence is consistent with previous studies. The RIS of Cr to the existing VN particles was confirmed by APT and may accelerate the transition of VN to Cr-rich nitrides. The irradiation hardening contribution from dislocation loops, dislocation lines, G phase and α’ phase was parsed based on a linear dispersed barrier hardening model. The results suggest that most irradiation hardening at high temperature is due to increasing dislocation density with dose.

A Fe-9Cr ferritic martensitic (F/M) steel T91 was neutron-irradiated in the Advanced Test Reactor up to 3.96 dpa in two temperature ranges, 466 °C to 534 °C and 571 °C to 632 °C. The microstructure evolution including dislocation loops, precipitation, segregation of elements and phase stability were studied using analytical scanning-transmission electron microscopy and atom probe tomography. The hardening induced by irradiation was measured by nanoindentation. Ni/Si/Mn clusters were identified in all conditions except the one irradiated around 600 °C to 3.23 dpa. The compositions of Ni/Si/Mn clusters were found to be converging to G phase stoichiometrically with increasing dose, with Mn partially substituted by Cu. Significant coarsening of this phase was observed in high temperature cases, with total dissolution of intragranular G phase after prolonged irradiation. A similar trend was also identified for dislocation loops. The results obtained in this experiment provide evidences that the absence of dislocation loops in some specimens irradiated to high dose level under high irradiation temperature range (typically above 500 °C) is not due to the suppression of nucleation and growth by high point defect recombination rate, but rather fast coalescence of dislocation loops. Hardness measurements show different dose dependences for two temperature ranges. For specimen irradiated around 600 °C significant hardening before 0.5 dpa followed by softening process was observed, while in lower temperature range (450 °C ~ 500 °C) the normal hardening pattern with increasing dose was observed.

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.

Crack growth behaviour of additively manufactured (AM) Alloy 800H were studied under cyclic and static load in high-temperature water, focusing on heat treatment effects. Under cyclic loading, heat treatment gently affected corrosion fatigue. The growth rate is comparable to wrought 800H. As the load cycle frequency reduced to below 0.001 Hz, AM 800H exhibited inconsistent stress corrosion cracking response as compared to wrought 800H. Depending on heat treatment, inhomogeneous crack propagation and localized pinning were observed in stress-relieved and recrystallized AM 800H, attributed to carbides, microstructural inhomogeneity, and twin boundaries. Grain boundary sensitization at ˜10 wt% Cr caused fast crack growth.

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.