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At least 235 records · Page 13

Single-Crystallization of O3-Type Layered Oxide Cathode for Na-Ion Battery

The development of high-energy-density Na-ion batteries places significant demands on single-crystal layered oxide cathodes, especially for further high-voltage, solid-state battery scenarios. In the O3-type structure, due to the original sluggish Na ion diffusion kinetics (approximately 1 order of magnitude lower than that of Li-ion), and further hindrance against diffusion kinetics caused by single-crystal architecture, these inherent defects lead to the decline in the electrochemical performance. Herein, we demonstrated that the single crystallization of O3-type NaNi 1/3 Fe 1/3 Mn 1/3 O 2 cathode (d 50 = 5.04 μm) aggravates surface-to-bulk phase inhomogeneity distribution, which is attributed to the uneven Na ions extraction. Moreover, the Na-depletion of the surface/shell region not only aggravates Na ion diffusion resistance but also leads to a higher valence state of transition-metal elements (e.g., Ni/Fe) near the surface of the single-crystal particle, which further compromises the cathode-electrolyte interface stability. Furthermore, not limited to revealing the challenges, tuning the particle size and moderating quasi-single-crystal strategies have been proven to effectively mitigate the negative uneven distributions of Na ions, phases, and valence/oxidative states, resulting in efficient modification for single crystallization of Na-layered oxide cathodes.

36 MATERIALS SCIENCE↗

Oxygen Vacancy Evolution at Li x V 2 O 5 /LiPON Solid State Electrochemical Interfaces Using Depth Resolved Cathodoluminescence Spectroscopy

The formation of oxygen vacancies at buried LiPON/ Li x V 2 O 5 interfaces has been observed on a near-nanometer scale and nondestructively using depth-resolved cathodoluminescence spectroscopy (DRCLS) and interfacial markers. Before electrochemical cycling, as-deposited LiPON/Li x V 2 O 5 exhibits a 1.6 eV defect optical emission, which density functional theory calculations identify as originating from oxygen vacancies. This defect appears first within a few nanometers of the buried LiPON/Li x V 2 O 5 interface without cycling, indicating that spontaneous O diffusion from the Li x V 2 O 5 lattice into LiPON may have caused these interface-localized oxygen vacancy defects. DRCLS measured the intensity and spatial distribution of this oxygen vacancy signal as a function of electrochemical cycling in a LiPON/Li x V 2 O 5 half-cell, showing oxygen vacancy signal increasing and moving deeper into the electrode with increased cycle number. Significant electrochemical irreversibility was also observed, with poor Coulombic efficiency and a 15% drop in capacity over 50 cycles. Theoretical simulations predict that the presence of oxygen vacancies increases the energy barrier for lithium diffusion significantly, indicating that this aggregation of oxygen vacancies could be another battery degradation mechanism accompanying lithiation induced phase changes.

36 MATERIALS SCIENCE↗

Atomistic and cluster dynamics modeling of fission gas (Xe) diffusivity in TRISO fuel kernels

TRISO fuel particles are candidates for use in next generation reactors including gas reactors, fluoride salt-cooled high temperature reactors, and micro-reactors. The UCO fuel kernel consists of a uranium dioxide (UO) and uranium carbide mixture. The addition of UC helps suppress the formation of carbon monoxide gas, which led to failures during initial TRISO development. The addition of uranium carbide alters the chemistry of the UO kernel, which is known to influence performance parameters such as fission gas diffusivity, although the impact has not been quantified and no models exist that take the change in chemistry into account. Therefore, better understanding and more accurate models of the impact of chemistry on fuel performance are of high priority. In this paper, a first-principles density functional theory (DFT) and empirical potential based multi-scale study has been carried out to model the diffusivity of fission gas xenon (Xe) in UCO TRISO fuel kernels. The focus is on the UO component in the UCO fuel kernels, as that represents the largest volume fraction of the fuel kernels. The study relies on DFT and empirical potential calculations to determine Xe and point defect properties, which are then used in thermodynamic and kinetic models to predict diffusion for intrinsic conditions. In addition, the information is utilized in cluster dynamics simulations using the Centipede code to estimate the impact of irradiation on defect transport. Additionally, the presence of UC or UC in the UCO fuel kernels is shown to have a substantial impact on the UO non-stoichiometry by inducing oxygen vacancies and driving UO sub-stoichiometric, which causes much slower Xe diffusion in UCO compared to light water reactor UO fuel. The application of this model in fuel performance simulations using the Bison code is also demonstrated.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Differential multi-probe thermal transport measurements of multi-walled carbon nanotubes grown by chemical vapor deposition

Carbon nanotubes (CNTs) are quasi-one dimensional nanostructures that display both high thermal conductivity for potential thermal management applications and intriguing low-dimensional phonon transport phenomena. In comparison to the advances made in the theoretical calculation of the lattice thermal conductivity of CNTs, thermal transport measurements of CNTs have been limited by either the poor temperature sensitivity of Raman thermometry technique or the presence of contact thermal resistance errors in sensitive two-probe resistance thermometry measurements. Here we report advances in a multi-probe measurement of the intrinsic thermal conductivity of individual multi-walled CNT samples that are transferred from the growth substrate onto the measurement device. The sample-thermometer thermal interface resistance is directly measured by this multi-probe method and used to model the temperature distribution along the contacted sample segment. The detailed temperature profile helps to eliminate the contact thermal resistance error in the obtained thermal conductivity of the suspended sample segment. A differential electro-thermal bridge measurement method is established to enhance the signal-to-noise ratio and reduce the measurement uncertainty by over 40%. The obtained thermal resistances of multiple suspended segments of the same MWCNT samples increase nearly linearly with increasing length, revealing diffusive phonon transport as a result of phonon-defect scattering in these MWCNT samples. The measured thermal conductivity increases with temperature and reaches up to 390 ± 20 W m –1 K –1 at room temperature for a 9-walled MWCNT. In conclusion, theoretical analysis of the measurement results suggests submicron phonon mean free paths due to extrinsic phonon scattering by extended defects such as grain boundaries. The obtained thermal conductivity is decreased by a factor of 3 upon electron beam damage and surface contamination of the CNT sample.

42 ENGINEERING↗

Emergence of Subsurface Oxygen on Rh(111)

Oxygen atoms on transition metal surfaces are highly mobile under the demanding pressures and temperatures typically employed for heterogeneously catalyzed oxidation reactions. This mobility allows for rapid surface diffusion of oxygen atoms, as well as absorption into the subsurface and reemergence to the surface, resulting in variable reactivity. Subsurface oxygen atoms play a unique role in the chemistry of oxidized metal catalysts, yet little is known about how subsurface oxygen is formed or returns to the surface. Furthermore, if oxygen diffusion between the surface and subsurface is mediated by defects, there will be localized changes in the surface chemistry due to the elevated oxygen concentration near the emergence sites. We observed that oxygen atoms emerge preferentially along the boundary between surface phases and that subsurface oxygen is depleted before the surface oxide decomposes.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗

Coherent Erbium Spin Defects in Colloidal Nanocrystal Hosts

We demonstrate nearly a microsecond of spin coherence in Er 3+ ions doped in cerium dioxide nanocrystal hosts, despite a large gyromagnetic ratio and nanometric proximity of the spin defect to the nanocrystal surface. The long spin coherence is enabled by reducing the dopant density below the instantaneous diffusion limit in a nuclear spin-free host material, reaching the limit of a single erbium spin defect per nanocrystal. We observe a large Orbach energy in a highly symmetric cubic site, further protecting the coherence in a qubit that would otherwise rapidly decohere. Spatially correlated electron spectroscopy measurements reveal the presence of Ce 3+ at the nanocrystal surface, which likely acts as extraneous paramagnetic spin noise. Even with these factors, defect-embedded nanocrystal hosts show tremendous promise for quantum sensing and quantum communication applications, with multiple avenues, including core-shell fabrication, redox tuning of oxygen vacancies, and organic surfactant modification, available to further enhance their spin coherence and functionality in the future.

77 NANOSCIENCE AND NANOTECHNOLOGY↗

Development of Corrosion Resistant Coatings for Structural Materials for Liquid Fueled Molten Salts Reactors Applications (Final Report)

Current structural alloys, code certified for the MSRs temperature ranges, contain high levels of chromium, making them highly susceptible to molten halide salt corrosion. One potential solution to circumvent the need for code certification of novel alloys, which is expensive and time consuming, is to design claddings that will protect the underlying code certified materials from corrosion damage during operation. In this work, we examined the corrosion of Ni and Cu electroplated, Ni and CuNi weld overlay, Mo-laser clad, and carburized claddings on SS316H for use in molten salt reactor environments. After characterization of the cladded materials (Task 1), static corrosion tests were used to assess corrosion resistance of the claddings in typical molten fluoride salts (Task 2). The corrosion tests were performed in molten FLiNaK at 700°C up to 1000 hours. Pre- and post-corrosion Scanning Electron Microscopy (SEM), energy-dispersive-spectroscopy (EDS), Transmission Electron Microscopy (TEM), X-ray Diffraction (XRD) and glow-discharge-optical-emission-spectroscopy (GDOES) were performed on cladding cross-sections and surfaces to evaluate degradation. The Cu and Ni electroplated samples, as well as the carburized samples, showed excellent corrosion resistance relative to the bare SS316H. To assess high temperature cladding stability, ageing experiments were performed at temperatures up to 900 C in inert atmosphere for the electroplated and carburized samples and a diffusion model was developed to predict long term cladding behavior (Task 3). It was found that the Cu cladding was basically insensitive to the high temperature ageing, except for small secondary phases forming at the interface. On the other hand, the Ni electroplated cladding experienced significant interdiffusion. Nevertheless, the gain in corrosion resistance for a 100μm Ni electroplated cladding is phenomenal, with more than 50% reduction in chromium dissolution from the substrate material for the first 15 years of salt exposure at 700°C. To assess radiation resistance and phase stability, high-temperature 4MeV Ni heavy ion irradiation was performed across the cladding/substrate interface up to 50 displacement-per-atom (DPA) at 500°C and 700°C to assess the phase stability and irradiation behaviors of the electroplated systems (Task 4). The Ni and Cu electroplated systems did not experience void swelling at the contrary to the SS316H substrate due to their nanocrystalline nature. Some level of recrystallization was observed in the cladding, as well as radiation induced segregation and enhanced diffusion. The interface acted as a potent sink for point defects with the presence of a void denuded zone. The mechanical properties of the claddings were assessed using thermal shock, micro-indentation, nano-indentation, and four-point bend testing experiments (Task 5). These experiments were performed on the claddings in as-received, corroded, irradiated, and thermally aged states to determine the effects of typical molten salt reactor environments on cladding integrity. No significant mass loss was observed after repeated thermal shocks. While the electroplated samples softened after high-temperature ageing, irradiation hardening compensate this effect, such that there is little different with the as-received materials. The results of these experiments suggest that the electroplated (copper/nickel) show the greatest promise for application in molten salt reactor development. These claddings prevented any chromium dissolution from occurring during the static corrosion experiments. Additionally, they demonstrated favorable properties for high temperature diffusion, phase stability, interfacial mechanical properties, and irradiation resistance. Weld-overlay cladding are also of great interest and could reach properties similar to the electroplated samples upon optimization (especially with multiple weld passes). The carburized SS316H showed excellent behavior as well, but more work needs to be done to assess their long-term stability. Finally, the Mo-laser clad system was not pursued further due to the manufacturing challenges. It is also worth noting that the Ni cladding systems should behave relatively well in terms of weldability since the Ni-weld overlay microstructure and associated corrosion rate are sound. While the NiCu weld-overlay has even lower corrosion rates than the Ni-weld overlay, more studies on welding of Cu-electroplated systems are necessary since Cu clusters are known to embrittle steels.

36 MATERIALS SCIENCE↗

Dislocation loop evolution in Kr-irradiated ThO 2

Here, the early stage of microstructural evolution of ThO 2 , under krypton irradiation at 600, 800, and 1000°C, was investigated using in situ transmission electron microscopy (TEM). Dislocation loops grew faster, whereas their number density decreased with increasing irradiation temperature. Loop density was found to decrease with ion dose. Interstitial dislocation loops, including Frank loops with Burgers vector of $\textit{a}/3\langle111\rangle$ and perfect loops with Burgers vector of $\textit{a}/2\langle110\rangle$, were determined by traditional TEM and atomic resolution–scanning TEM techniques. Atomistic and mesoscale level modeling are performed to interpret experimental observations. The migration energy barriers of defects in ThO 2 were calculated using density-functional theory. The energetics of different dislocation loop types were studied using molecular dynamics simulations. Loop density and diameter were analyzed using a kinetic rate theory model that considers stoichiometric loop evolution. This analysis reveals that loop growth is governed by the mobility of cation interstitials, whereas loop nucleation is determined by the mobility of anion defects. Lastly, a rate theory model was used to extract the diffusion coefficients of thorium interstitials, oxygen interstitials, and vacancies.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Trailblazing Kr/Xe Separation: The Birth of the First Kr-Selective Material

Efficient separation of Kr from Kr/Xe mixtures is pivotal in nuclear waste management and dark matter research. Thus far, scientists have encountered a formidable challenge: the absence of a material with the ability to selectively adsorb Kr over Xe at room temperature. This study presents a groundbreaking transformation of the renowned metal–organic framework (MOF) CuBTC, previously acknowledged for its Xe adsorption affinity, into an unparalleled Kr-selective adsorbent. This achievement stems from an innovative densification approach involving systematic compression of the MOF, where the crystal size, interparticle interaction, defects, and evacuation conditions are synergistically modulated. The resultant densified CuBTC phase exhibits exceptional mechanical resilience, radiation tolerance, and notably an unprecedented selectivity for Kr over Xe at room temperature. Simulation and experimental kinetic diffusion studies confirm reduced gas diffusion in the densified MOF, attributed to its small pore window and minimal interparticle voids. The lighter Kr element demonstrates facile surface passage and higher diffusivity within the material, while the heavier Xe encounters increased difficulty entering the material and lower diffusivity. This Kr-selective MOF not only represents a significant breakthrough in Kr separation but also demonstrates remarkable processability and scalability to kilogram levels. Furthermore, the findings presented herein underscore the transformative potential of engineered MOFs in addressing complex challenges, heralding a new era of Kr separation technologies.

36 MATERIALS SCIENCE↗

First-principles study of intrinsic point defects and Xe impurities in uranium monocarbide

Based on density functional theory (DFT) calculations, we perform an extensive investigation of intrinsic point defects and Xe impurities in uranium monocarbide (UC). The DFT calculations involve both the conventional generalized gradient approximation (GGA) and the GGA+U approach with the Hubbard parametric term (U), using up to 5×5×5 supercells. GGA calculations for the formation energy of intrinsic defects demonstrate the significant effect of using larger supercells than in previous studies. Results confirm that the ⟨111⟩ and ⟨100⟩ dumbbell interstitials are the most stable interstitial configurations for U and C, respectively. The interstitial mechanisms are favored for self-diffusion of both uranium and carbon and diffusion of Xe under equilibrium conditions. Calculations also reveal that the Xe substitutional defect at the C lattice site tends to adopt an off-site configuration, which can be interpreted as a Xe interstitial–C vacancy complex. We also utilize GGA+U to assess the impact of effective U parameter (Ueff) on the results. Moreover, we introduce a method to estimate the carbon chemical potential by fitting the phase diagram composition data and propose a selection Ueff=1.25 eV based on the experimental Xe diffusion activation energy. With this approach, GGA+U calculations reproduce the available experimental data for the formation energy of the carbon Frenkel pair and can explain the stepwise recovery of intrinsic properties and burst Xe release behavior in UC observed in annealing experiments.

Huang, Gui-Yang (ORCID:0000000301447376)↗

Molecular dynamics study of grain boundary and radiation effects on tritium population and diffusion in zirconium

Tritium population thermodynamics and transport kinetics critically define the tritium storage performance of zirconium tritides that can be used for a variety of nuclear applications including tritium-producing burnable absorber rods. Both thermodynamic and kinetic properties can be sensitive to grain sizes of materials and can be significantly altered by irradiated defects during operation under the reactor environments. A thorough experimental characterization of how these properties evolve under different reactor conditions and different initial grain structures is extremely challenging. Here molecular dynamics simulations are used to investigate tritium population and diffusion in zirconium with and without different planar symmetric and asymmetric tilt grain boundaries and irradiated defects. Here, we found that in addition to trapping tritium, the most significant effect of planar grain boundaries is to increase tritium diffusivity on the boundary plane. Furthermore, fine grain structures are found to mitigate the change of tritium diffusivity due to irradiated point defects as these point defects are likely to migrate to and sink at grain boundaries.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Antiphase boundary migration as a diffusion mechanism in a P3 sodium layered oxide

The electrode materials of Na-ion batteries must exhibit fast ion transport, a property that is often negatively impacted when mobile cations order. Many layered Na (and K) intercalation compounds adopt the P3 crystal structure, which tends to stabilize cation-vacancy orderings over a wide range of Na concentrations. The ordered phases have been predicted to consist of domains that are periodically separated by antiphase boundaries (APBs) to accommodate variations in Na concentration. The mechanisms and rates with which Na diffuses within such ordered phases of P3 are not understood. We have conducted a first-principles study of Na diffusion in the P3 structure, using Na x CoO 2 as a model system. We find that most simple nearest-neighbor hops in Na-vacancy ordered phases in P3 result in unstable configurations and are therefore forbidden. We identify an alternative mechanism of Na transport that is mediated by APB migration via the formation and expansion of kinks along domain boundaries. The limiting kinetic barriers for this mechanism are quite low (0.03 to 0.30 eV). Vacancy defects are not required, and furthermore have formation energies that are higher than those of the APB kink defects. Furthermore, our results suggest that APB migration is a fundamental diffusion mechanism in the P3 structure.

36 MATERIALS SCIENCE↗

Simulations of self- and Xe diffusivity in uranium mononitride including chemistry and irradiation effects

A combination of density functional theory and empirical potential atomic scale simulations have been used to determine a model for defect stability and mobility in uranium mononitride (UN), as a function of temperature (T) and N 2 partial pressure (p N 2 ). Using the model, predictions of hypo-stoichiometry under U-rich conditions compare favorably to CALPHAD calculations using the TAF-ID database. Furthermore, our predictions of U and N self-diffusivity are in good agreement with experiments carried out as a function of T at specific partial pressures under thermal equilibrium. The validated atomic scale data have then been implemented within a cluster dynamics method to simulate irradiation-enhanced defect concentrations. All defects and clusters studied have significantly enhanced concentrations, with respect to thermal equilibrium, as T is lowered. The irradiation-enhanced Xe diffusivity is compared to post-irradiation annealing and in-pile experiments. In conclusion, the contributions of various defects and clusters to non-stoichiometry, self-diffusivity, and Xe diffusivity are discussed.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗

Influence of pre-existing defects on thermal transport in nuclear graphite

Nuclear graphite is a critical material in high-temperature nuclear reactors due to its superior thermal and mechanical properties. The manufacturing process leaves multi-scale ‘pre-existing’ defects that can affect thermal transport characteristics. Because these defects are remnant of graphitization temperature, they cannot be thermally annealed. This study employs a non-thermal electron wind force (EWF) annealing technique to avoid this obstacle. 2 min of EWF treatment of the as-received graphite IG-110 at temperatures below 100 °C led up to 67% increase in thermal diffusivity and ~ 35% decrease in electrical resistivity in average. Differential scanning calorimetry also showed similar outcome for specific heat. X-ray diffraction characterization was performed by fitting a bi-modal crystallite size distribution model to reveal the enhancement in crystallinity after the EWF treatment. The findings emphasize the potential of EWF annealing for optimizing thermal performance in nuclear graphite and its implications for reactor efficiency and safety.

Annealing↗

Influence of grain size on α' Cr precipitation in an isothermally aged Fe-21Cr-5Al alloy

Cr-rich α' precipitation during aging typically leads to hardening and accordingly embrittlement of FeCrAl alloys, which needs to be suppressed. The influence of grain size on α' precipitation was studied by aging coarse-grained (CG), ultra-fine grained (UFG), and nanocrystalline (NC) ferritic Kanthal-D [KD; Fe-21Cr-5Al (wt.%) alloy] at 450, 500 and 550 °C for 500 h. After aging at 450 and 500 °C, less hardening was observed in the UFG KD than in CG KD. Atom probe tomography indicated a lower number density and larger sized intragranular α' in the UFG versus the CG alloy. The smaller grain size and higher defect (vacancy and dislocation) density in the UFG KD facilitated diffusion and accordingly enhanced precipitation kinetics, leading to coarsening of precipitates, as well as saturation of precipitation at lower temperatures, as compared to those in CG KD. No hardening occurred in UFG and CG KD after aging at 550 °C, indicating that the miscibility gap is between 500 and 550 °C. NC KD exhibited softening after aging owing to grain growth. α' precipitation occurred in NC KD aged at 450 °C but not at 500 °C, indicating that miscibility gap is between 450 and 500 °C. Thus, the significantly smaller grain size in NC KD decreased the miscibility gap, as compared to that in CG and UFG KD. Finally, this is attributed to the absorption of vacancies by migrating grain boundaries during aging, suppressing α' nucleation and enhancing Cr solubility.

36 MATERIALS SCIENCE↗

Sluggish, chemical bias and percolation phenomena in atomic transport by vacancy and interstitial diffusion in Ni-Fe alloys

In this paper, we report the results of extended atomistic modeling of intrinsic mobility of point defects and associated atomic transport in Ni-Fe model binary alloys. We consider the effects of composition and temperature and present evidence of the sluggish and chemically biased diffusion, and percolation effects occurring in atomic transport via the vacancy and interstitial migration mechanisms. The results are analyzed and discussed in the light of previous studies and some experimental observations. It is demonstrated that the sluggish diffusion, the chemically biased diffusion, and the percolation are interlinked phenomena that are defined by the chemical complexity of particular alloys. Methods for predicting these phenomena in multicomponent alloys are discussed.We report a fundamental understanding of sluggish diffusion, chemically-biased diffusion, as well as percolation phenomena, in Ni-Fe random alloys for vacancy and interstitial atom migration mechanisms.

36 MATERIALS SCIENCE↗

Radiation-induced vacancy injection in heterogeneous multiphase materials

Understanding the synergy between corrosion and defect dynamics is a key consideration in the development of advanced materials for extreme environments. Here, we reveal a surprising phenomenon for the transport of point defects induced by irradiation in a heterogeneous multiphase structure of a metal and an oxide, similar to that formed under metal corrosion that takes place in most environments. Despite the confinement of the produced damage within the oxide, vacancies were injected into the unirradiated metal and coarsened with dose. Furthermore, the results show that the nature of the oxide layer dictates the defect evolution in the metal layer. This work reveals an interesting mechanism for point defect interactions in multiphase materials, with broad implications in many fields, while also emphasizing the complex coupling between corrosion and irradiation. Corrosion leads to the formation of multiphase materials, while irradiation enhances diffusion within the heterogeneous phases, which can impact corrosion rates.

36 MATERIALS SCIENCE↗

Diffusion Behavior of Oversized Fission Products in bcc Fe Cladding: A First-Principles Study

Fuel-Cladding Chemical Interaction (FCCI) poses significant challenges in nuclear reactors, where fission products from nuclear fuel interact with Fe-based cladding materials, potentially compromising their structural integrity. This study investigates the diffusion behavior of oversized fission products, Pr, Nd, Ce, and La, within bcc Fe cladding using density functional theory (DFT), nudged elastic band (NEB) method, and self-consistent mean field (SCMF) theory. Our results reveal significant long-range vacancy binding energies, particularly up to the 6th nearest neighbor, with La exhibiting the strongest binding affinity, followed by Nd, Ce, and Pr. The NEB calculations indicate significant high barriers for the dissociation of 1nn vacancy-solute pairs for all fission products. The tracer diffusion coefficients of these fission products was derived in Arrhenius form. The significant trapping effect of vacancies by a very dilute amount of fission products reduces vacancy mobility, leading to an oversaturation of point defects, void nucleation, and swelling. These are critical issues for irradiated cladding materials. The tracer diffusion coefficients indicate that Nd diffuses the fastest, followed by La, Ce, and Pr. This study provides essential insights for developing advanced cladding materials and design strategies to mitigate FCCI, ultimately enhancing nuclear reactor safety and performance.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗