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At least 397 records · Page 22
All sourcesNASA NTRSDOE OSTI
DOE OSTI2022

AGR-5/6/7 Final Release-to-Birth Ratio Data Analysis

AGR-5/6/7 is the last of a series of Advanced Gas Reactor (AGR) experiments conducted in the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL) in support of development and qualification of tristructural isotropic (TRISO) low-enriched fuel for use in the high-temperature gas cooled reactor (HTGR). AGR configuration and irradiation conditions are based on prismatic HTGR technology that is distinguished primarily through the use of helium coolant, a low-power-density ceramic core capable of withstanding very high temperatures, and TRISO-coated particle fuel. The AGR tests provide valuable irradiation-performance data to support fuel process development, qualify fuel for normal operation and accident conditions, and support development and validation of fuel performance and fission-product (FP) transport models and codes. Each AGR test consists of multiple independently controlled and monitored capsules containing fuel compacts placed in a graphite cylinder shrouded by a steel shell. Release-to-birth ratios (R/B) for fission-gas isotopes released from each capsule are calculated from release rates, measured by germanium detectors in the Fission Product Monitoring System (FPMS) installed downstream from each capsule, and birth rates calculated using numerical models of FG generation. The R/Bs are a critical measure of the ability of the fuel kernel, the particle coating layers, and the compact matrix to retain fission-gas atoms, preventing their release into the sweep-gas flow, and the impact of initially defective particles and/or particle-coating failures that occur during irradiation. For fission-gas isotopes, particle failure is defined as failure of all coating layers, allowing gaseous fission atoms to escape from a particle. During the first five cycles (162B ? 165A), R/Bs were stable in the 10-8?10-6 range, and no in-pile particle failures were observed, based on the gross gamma counts. The maximum R/B value of around 2 ? 10-6 for Kr-85m resulted from the presence of as-fabricated exposed kernels (based on the high exposed kernel fraction), the dispersed uranium, and high fuel particle temperatures in Capsule 1. Comparison of capsule-measured R/Bs from these early cycles to predictions using the previously developed AGR R/B model demonstrated FG release from the AGR-5/6/7 TRISO fuel was comparable to that of previous experiments. In addition, the Kr-85m R/B per-exposed-kernel values are comparable to R/B values obtained in AGR-3/4 irradiation experiment and four irradiation experiments performed during 1980s: (1) HRB-17/18, (2) COMEDIE-BD1, (3) HFR B1, and (4) HRB-21. In contrast, all measured R/B values are lower than predictions by the commonly used Richards and German models, which are intentionally conservative. A large number of in-pile particle failures occurred in Capsule 1 by the end of Cycle 166A. During the final four cycles (166A ? 168A), apparent damage to the Capsule 1 gas line appeared to cause FG leakage from that capsule into the other four capsules, resulting in an increase in fission gas (FG) detected in the effluent for all capsules. Isolation of the Capsule 1 gas line during the last three cycles also prevented measurement of its FG release. Thus, R/Bs in all capsules after Cycle 166A are highly uncertain because of undefined amount of leakage from Capsule 1, especially for long-lived isotopes. A few hundred in-pile particle failures were estimated for Capsule 1 before the end of Cycle 166A, but the total number of failures is unknown due to the lack of FG release data in the later cycles. Based primarily on evidence from the gross gamma counts during Cycle 168A, approximately 15 particles failed in Capsule 3 and four particles failed in Capsule 2. In-pile failures in Capsule 3 were anticipated because this capsule was designed to operate beyond the HTGR normal operating temperature range. In contrast, no in-pile failures were identified in the top two capsules (4 and 5) based on the absence of the typical spikes in gross gamma counts and low failure estimates using the AGR model, developed in INL/EXT-14-32970, for R/B of the short-lived isotopes (Kr-89 and Xe-137) with minimal leakage from Capsule 1.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Microstructural evolution of a silicon carbide-carbon coated nanostructured ferritic alloy composite during in-situ Kr ion irradiation at 300°C 450°C

This work focuses on irradiation behaviors of a novel silicon carbide and carbon coated nanostructured ferritic alloy (SiC-C@NFA) composite for potential applications as a cladding and structural material for next generation nuclear reactors. The SiC-C@NFA samples were irradiated with 1 MeV Kr ions up to 10 dpa at 300 and 450 °C. Microstructures and defect evolution were studied in-situ at the IVEM-Tandem facility at Argonne National Laboratory. The effects of ion irradiation on various phases such as α-ferrite matrix, (Fe,Cr) 7 C 3 , and (Ti,W)C precipitates were evaluated. The α-ferrite matrix showed a continuous increase in dislocation density along with spatial ordering of dislocation loops (or loop strings) at >5 dpa. The size of the dislocation loops at 450 °C was larger than that at 300 °C. The nucleation and growth of new (Ti,W)C precipitates in α-ferrite grains were enhanced with the ion dose at 450 °C. This study provides new insight into the irradiation resistance of the SiC-C@NFA system.

(FeCr)(7)C-3↗
DOE OSTI2022

Grain boundary strain as a determinant of localized sink efficiency

The opportunity to achieve radiation tolerance in crystalline materials hinges on understanding the structure and response of grain boundary sinks to irradiation. A common descriptor of grain boundary efficiency as a defect sink is the denuded zone, which is a defect free zone adjacent to the grain boundary dictated by its ability to absorb radiation induced defects. This descriptor is often used at the mesoscale, which requires an averaging of absorption events. In this paper, we resolve the defect sink efficiency as a function of interfacial strain with respect to grain boundary character, and correlate high levels of grain boundary strain to an enhanced absorption efficiency. Here, we also introduce a key relationship between localized strain in proximity with the grain boundary sink and the variation absorption efficiency associated with these regions, revealing the pitfalls of averaging absorption events along a grain boundary, and presenting a path forward toward improved models for denuded zones and localized grain boundary absorption phenomena.

36 MATERIALS SCIENCE↗
DOE OSTI2022

Ab initio calculations for void swelling bias in α - and δ -plutonium

Void swelling can develop in materials under persistent irradiation when nonequilibrium vacancy and self-interstitial populations migrate under sufficiently asymmetric interaction biases. In conventional metals, the propensity is determined to the first approximation by comparing point-defect relaxation strains. We thus present DFT-based calculations of structures and formation energies and volumes of point defects in the α and the δ phases of plutonium. Here we discuss the pros and cons of various levels of electronic structure theory: spin polarization, spin-orbit coupling, and orbital polarization. Our results show that lattice defects in δ-Pu, in contrast to most fcc metals, have surprisingly small formation volumes. Equally unexpected are the large defect formation volumes found in the low-symmetry α-Pu phase. Both these unusual properties can be satisfactorily explained from defect-induced spin/orbital moment formation and destruction in the Pu phases. Surprisingly, the point defects in α-Pu are found to induce far larger transformation of the local electronic structure than in δ-Pu. When we use the calculated defect properties to estimate the classic void swelling bias in each of the phases, we find it to be unusually small in δ-Pu but likely much larger in α-Pu. Hence, swelling rates and mechanisms can diverge dramatically between the different phases of Pu. Especially in the transient regime before the formation of large defect clusters, the swelling rate of α-Pu can reliably be expected to be much larger than δ-Pu. However, accurate forecasts over longer times will require the conventional void-swelling theory to be modified to handle the complexities presented by the different Pu phases. As a case in point, we show the possible anomalous temperature dependence of vacancy properties in δ-Pu, caused by entropic contributions from defect-induced spin-lattice fluctuations. Such complications may affect defect-defect interactions and thus alter the void swelling bias.

75 CONDENSED MATTER PHYSICS, SUPERCONDUCTIVITY AND↗
DOE OSTI2026

Electron beam irradiation effects on bulk metals: a comparative study of polycrystalline versus single-crystalline structures

This study investigates the effects of electron beam (e-beam) irradiation on the mechanical and structural properties of eight bulk metallic samples, comprising both polycrystalline (PC) and single-crystalline (SC) forms of Ni, Cr, V, and Ti. These metals were evaluated as potential candidates for beam exit windows in high-power (MW-class) particle accelerators. The primary objective is to identify metals capable of withstanding the conditions of high-power/MW-class e-beam accelerators and serve effectively as exit windows. Selection criteria were based on each metal’s intrinsic properties, power dissipation capability, and irradiation-induced changes in mechanical behavior, including hardness, elastic modulus, and defect density. Comprehensive characterization was conducted using field-emission scanning electron microscopy (FESEM), atomic force microscopy (AFM), scanning electron microscopy with energy-dispersive X-ray spectroscopy (SEM–EDS), X-ray diffraction (XRD), high-resolution transmission electron microscopy (HRTEM), selected area electron diffraction (SAED), and nanoindentation, performed both before and after exposure to a ~¿66 kGy dose from a 10 MeV e-beam accelerator. Results revealed that e-beam irradiation induced hardening in PC Ni, whereas PC Ti, commonly used in beam exit windows, exhibited softening. The observed softening in PC Ti is attributed to grain coarsening, elongation, and the formation of twins and twin boundaries, in contrast to the smaller, compressed grains in the pristine (Pr) PC Ti samples, consistent with the Hall–Petch relationship. The stresses due to twinning are small and insignificant in influencing the overall hardening of the PC Ti irradiated sample when compared to the stresses due to the dislocation density. Conversely, SC Ti samples exhibited irradiation-induced hardening. The SC Ti irradiated samples developed additional irradiation-induced modifications in crystallographic texture of (100), (101), (110), (200), (112), (004), and (211) as evidenced from the XRD results, which could probably explain the hardening effect that is caused by irradiation.

36 MATERIALS SCIENCE↗
DOE OSTI2021

Optical transmission and dimensional stability of single-crystal sapphire after high-dose neutron irradiation at various temperatures up to 688 °C

The use of single-crystal sapphire optical fibers has been considered to extend fiber-optic sensing to the extreme temperature (>1000 °C) environments encountered in nuclear applications. However, before these sapphire fiber–based sensors can be deployed, their optical transmission and dimensional stability (which impacts drift of some sensors) must be characterized under representative testing conditions. Data regarding the optical transmission of sapphire following high-dose neutron irradiation at temperatures >100 °C is extremely limited. This work provides measurements of optical density (i.e., attenuation) and directional dimensional changes in bulk single-crystal sapphire materials irradiated to a fast neutron fluence of 2.4 × 10 21 n/cm 2 (3.5 displacements per atom) at temperatures ranging from 95 to 688 °C. Optical density measured after irradiation at 95 and 298 °C showed ultraviolet and visible absorption bands corresponding to known defect centers and temperature trends that were generally consistent with previous ex situ and in situ measurements made at much lower neutron fluence. However, optical density measured after irradiation at 688 °C was as much as two orders of magnitude higher, indicating that the fundamental mechanism for radiation-induced attenuation changes at this irradiation temperature. Additional analysis and comparison with previous works suggest that the attenuation may result from void formation, leading to increased Rayleigh scattering losses in the material and increased swelling that would also result in drift of Bragg grating-based sensors in sapphire fibers. These results pose serious questions regarding the feasibility of sapphire fiber–based sensors for high-temperature nuclear applications.

36 MATERIALS SCIENCE↗
DOE OSTI2022

Tuning of the electronic and vibrational properties of epitaxial MoS 2 through He-ion beam modification

Abstract Atomically thin transition metal dichalcogenides (TMDs), like MoS 2 with high carrier mobilities and tunable electron dispersions, are unique active material candidates for next generation opto-electronic devices. Previous studies on ion irradiation show great potential applications when applied to two-dimensional (2D) materials, yet have been limited to micron size exfoliated flakes or smaller. To demonstrate the scalability of this method for industrial applications, we report the application of relatively low power (50 keV) 4 He + ion irradiation towards tuning the optoelectronic properties of an epitaxially grown continuous film of MoS 2 at the wafer scale, and demonstrate that precise manipulation of atomistic defects can be achieved in TMD films using ion implanters. The effect of 4 He + ion fluence on the PL and Raman signatures of the irradiated film provides new insights into the type and concentration of defects formed in the MoS 2 lattice, which are quantified through ion beam analysis. PL and Raman spectroscopy indicate that point defects are generated without causing disruption to the underlying lattice structure of the 2D films and hence, this technique can prove to be an effective way to achieve defect-mediated control over the opto-electronic properties of MoS 2 and other 2D materials.

77 NANOSCIENCE AND NANOTECHNOLOGY↗
DOE OSTI2021

Healing of nanocracks by collision cascades in nickel

We use molecular dynamics simulations to investigate the interaction between collision cascades and pre-existing nano-cracks in nickel (Ni). We find that collision cascades cause the cracks to heal whenever a cascade initiates a thermal spike whose core overlaps with the crack. Moreover, crack healing affects the distribution of radiation-induced defects generated during a collision cascade: defect clusters, such as dislocations and stacking fault tetrahedra, dominate in the presence of crack healing, while isolated point defects dominate in its absence. Furthermore, this work indicates that irradiation may reduce the number of pre-existing crack-like flaws in metallic components, potentially improving their mechanical resilience.

36 MATERIALS SCIENCE↗
DOE OSTI2024

Technical report Letter: RAFM, ODS steels and MMLC for Nuclear energy application

The lifetime, thermodynamic efficiency, safety and economic viability of new generation fission and fusion reactor concepts can largely be tied to the mechanical performance and stability of structural alloys under extreme environments. In this context, engineered nano materials could have broad-reaching impact on the future of advanced nuclear fuel-cycle and reactors. These systems are characterized by a large number density of interfaces which are efficient sinks for point defects and moderately biased; therefore limiting the deleterious effects of irradiation. Broadly, nuclear nano-technology deals with the use of the latest engineered-nanomaterials for improving the nuclear power performances and safety in all areas of nuclear energy production to bring new generations of nuclear power units. New advanced fuel assembly designs also have implications for securities and safeguards. To support the readiness for potential future license applications, an understanding of the technologies that would enable new reactor designs in the areas of component performance and domestic safeguards is necessary. This technical report letter work explores the technical issues and potential regulatory considerations associated with developing and adopting fuel claddings made of advanced nano- materials. Specifically three classes of nanomaterials are considered: (i) reduced activation ferritic/martensitic (RAFM) steels, (ii)oxide dispersed steels (ODS) and (iii) multi-metallic layered composites (MMLC).

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2025

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↗
DOE OSTI2023

Understanding grain boundary segregation under irradiation: A mesoscale modeling study

Grain boundaries (GBs) are an important component in the design of irradiation resistant materials as they play a crucial role as point defect sinks for non-equilibrium vacancies and self-interstitials. However, in the presence of sustained point defect fluxes, the coupled transport of atomic components with inherently different diffusivities often leads to radiation-induced segregation (RIS), with undesired effects on properties such as intergranular corrosion and stress-assisted cracking. In this talk, we investigate the fundamental mechanisms leading to GB segregation in BCC FeCr-based alloys using multiscale modeling and experiments. A phase-field model is employed to account for both thermodynamic and kinetic segregation mechanisms. The full Onsager transport matrix is derived using kinetic Monte Carlo, molecular dynamics, and kinetic cluster expansion (KineCluE) calculations. The thermodynamics of GB segregation is described via an atomic density-based modification to the CALPHAD free energy. The modeling predictions are then compared against segregation data from atom probe characterization for various temperatures, compositions, and neutron irradiation conditions. We will conclude by discussing the utility of the multiscale modeling approach to shed light on the fundamental mechanisms leading to solute-GB interaction.

36 MATERIALS SCIENCE↗
DOE OSTI2023

Bridging microscale to macroscale mechanical property measurements of FeCrAl alloys by crystal plasticity modeling

FeCrAl alloys are candidates for accident tolerant fuel cladding of light water reactors. In this work, a microstructure- and temperature-dependent crystal plasticity model is employed to bridge microscale to macroscale mechanical property measurements of FeCrAl alloys. With the visco-plastic self-consistent (VPSC) polycrystal plasticity framework, a mechanism-based single crystal plasticity (MSCP) model adopts the Arrhenius type rate equation to describe the dependence of the critical resolved shear stress for dislocation slips on their temperature-dependent intrinsic frictional resistance and the microstructure-dependent irradiation hardening. The intrinsic frictional resistance associated with {110}<111> and {112}<111> slip systems were measured by in-situ micromechanical testing on unirradiated/irradiated samples at 25-500 °C. The irradiation hardening is estimated by the Bacon-Kocks-Scattergood (BKS) model with density and size of radiation-induced defects measured from microstructural characterization. Several features associated with thermo-mechanical behavior of unirradiated/irradiated polycrystalline FeCrAl alloys are captured. High density of deformation-induced dislocations and radiation-induced defects results in obvious hardening at room temperature, which is weakened at high temperature, and facilitates damage evolution during deformation. Moreover, both high temperature and radiation-induced defects, which facilitate dislocation multiplication, trigger large hardening rate. Finally, the proposed method together with application of accelerator-based ion irradiation technique is a surrogate approach to simulate neutron damage, improving the efficiency associated with evaluation of mechanical properties of FeCrAl alloys exposed to temperature, stress and radiation conditions.

36 MATERIALS SCIENCE↗
DOE OSTI2024

Structure and migration of heavily irradiated grain boundaries and dislocations in Ni in the athermal limit

The microstructural evolution at and near preexisting grain boundaries (GBs) and dislocations in materials under high radiation doses is still poorly understood. In this work, we use the creation relaxation algorithm (CRA) developed for atomistic modeling of high-dose irradiation in bulk materials to probe the athermal limit of saturation of GB and dislocation core regions under irradiation in fcc Ni. We find that, upon continuously subjecting a single dislocation or GB to Frenkel pair creation in the athermal limit, a local steady-state disordered defect structure is reached with excess properties that fluctuate around constant values. Case studies are given for a straight screw dislocation which elongates into a helix under irradiation and several types of low- and high-angle GBs, which exhibit coupled responses such as absorption of extrinsic dislocations, roughening, and migration. A positive correlation is found between the initial GB energy and the local steady-state GB energy under irradiation across a wide variety of GB types. Metastable GB structures with similar density in the defect core region but different initial configurations are found to converge to the same limiting structure under CRA. The mechanical responses of pristine and irradiated dislocations and GB structures are compared under an applied shear stress. Irradiated screw and edge dislocations are found to exhibit a hardening response, migrating at larger flow stresses than their pristine counterparts. Mobile GBs are found to exhibit softening or hardening responses depending on GB character. Although some GBs recover their initial pristine structures upon migration outside of the radiation zone, many GBs sustain different flow stresses corresponding to altered mobile core structures.

36 MATERIALS SCIENCE↗
DOE OSTI2025

Be₂C as a neutron moderator for molten salt reactors: A first-principles study of structural, electronic, and defect properties

Beryllium carbide (Be 2 C), valued for its high neutron moderation efficiency and low absorption cross section, is a promising high-temperature neutron moderator for molten salt reactors. Its practical adoption, however, demands significant technological maturation, requiring comprehensive theoretical and experimental studies of its response to different conditions, including high temperature and irradiation. Here, we report initial results on the fundamental properties and radiation-induced defects of Be 2 C, focusing on antisites, vacancies, interstitial atoms, and Frenkel pairs in the Be and C sublattices. Using density functional theory (DFT) and ab initio molecular dynamics (AIMD), we calculate the defects formation and binding energies, evaluating their dependence on the supercell size, charge states, and chemical environment. In general, carbon defects exhibit higher formation energies, greater sensitivity to cell size, and stronger impacts on the density of states compared to beryllium defects, with charged state the effects being more pronounced. Static DFT reveals multiple metastable interstitial configurations, while AIMD identifies ground states as C-C <100> dumbbells and octahedral Be interstitials. In conclusion, the diversity of metastable configurations and defect states complicates the diffusion mechanisms, requiring further molecular dynamics analysis to elucidate the mechanisms and rates of radiation-induced atomic transport, as well as the structural stability of Be2C.

Ab initio modeling↗
DOE OSTI2024

The microstructure effects on irradiation response of ferritic – martensitic steels

Microstructural optimization to achieve greater mechanical strength has been one of the focuses in ferritic–martensitic steels development. However, these optimized microstructures’ effects on the radiation response are not well known. In this work, two ferritic–martensitic steels (9Cr-NbMo and 9Cr-Ta) underwent neutron irradiation in the High Flux Isotope Reactor, and their room-temperature post-irradiation tensile properties and microstructure evolutions were investigated and compared. These two steels exhibit similar pre-irradiation tensile behavior, and their yield strengths are higher than that of other ferritic–martensitic steels by about 200–250 MPa. Microstructural characterization on pre-irradiated materials reveals a smaller grain size in 9Cr-Ta (2.8 ± 0.3 μm in 9Cr-Ta versus 4.3 ± 0.5 μm in 9Cr-NbMo) but higher dislocation density and precipitate density in 9Cr-NbMo. As is common for ferritic–martensitic steels at low irradiation temperatures (less than about 0.45T m ), irradiation-induced hardening at 400 °C was observed for both alloys. Irradiation at 490 °C causes the two alloys to exhibit different tensile behavior: 9Cr-Ta softens by 208 MPa in yield stress, whereas 9Cr-NbMo maintains strength. Additionally, microstructural characterizations were performed, including precipitate growth, dislocation, and defect formation. Using the barrier hardening model for microstructure–property correlation, the softening in irradiated 9Cr-Ta is primarily attributed to the significant dislocation recovery, while the strength lost from the slight dislocation recovery in 9Cr-NbMo was compensated by the additional strength from the irradiation-induced cavities. The microstructure effect (primarily precipitate, dislocation and boundary) on the radiation response is discussed herein.

36 MATERIALS SCIENCE↗
DOE OSTI2023

A deep learning model for automatic analysis of cavities in irradiated materials

Transmission electron microscopy (TEM) is a commonly used technique in materials science for defect investigation. Quantitative analysis of defects is important for understanding the properties of a material, but manual analysis of TEM micrographs can be time-consuming and prone to error, especially when the defects have irregular shapes rather than spherical shapes. Many existing methods or deep learning models do not handle a wide range of sizes for the same object type within a single image. In this work, we present a framework that enables users to train an instance segmentation model called Mask R- CNN on any microstructure dataset, perform multi-detection on the same image at different scales, and obtain properties (e.g., size, area) of the objects based on the desired shape (e.g., circle, ellipse, rectangle). Additionally, we have developed a parallel detection module that uses multiple GPUs to increase the efficiency of the object detection process. We demonstrate the capabilities of our framework using a set of TEM images of cavities with different shapes, size distributions, and background contrasts. Finally, we show that the performance of our model in terms of density, size, and swelling of the cavities is comparable to the human average and that our model achieves the highest recall value compared to existing methods due to the use of image multi-rescaling.

36 MATERIALS SCIENCE↗
DOE OSTI2023

Molecular dynamics simulations of radiation response of $\mathrm{LiAlO_2}$ and $\mathrm{LiAl_5O_8}$

Here, we report the findings of a computational investigation of defect production and migration in two lithium aluminate ceramics: LiAlO 2 and LiAl 5 O 8 . Although the first ceramic, LiAlO 2 has been widely investigated computationally, especially via molecular dynamics (MD), there have been no MD studies done on LiAl 5 O 8 up to date, due to the lack of interatomic potentials for this system. This work implemented an existing set of Li-Al-O coulombic and Buckingham potential parameters for the LiAl 5 O 8 system from the literature, followed by validating it based on experimental values of density, crystallinity, bond lengths and melting point. The Li + diffusion in the two ceramics was examined and was found to be one order of magnitude slower in LiAl 5 O 8 (D 0 = 3.17 x 10 -11 m 2 /s) as compared to that in LiAlO 2 (D 0 = 4.02 x 10 -10 m 2 /s) at 600 K. The lithium vacancy migration barrier of 4.15 eV in LiAl 5 O 8 was more than three times that in LiAlO 2 (1.31 eV) possibly due to stronger ordering in the Li-O 6 octahedra of LiAl 5 O 8 than in Li-O 4 tetrahedra of LiAlO 2 . The Li displacement threshold energy (E d ) in LiAl 5 O 8 was found to have a median value of 68 eV which is much higher than that in LiAlO 2 (40 eV). The greater E d for Li signifies a lower likelihood for defect formation and clustering and thus lower tendency for amorphization in LiAl 5 O 8 . The presented results show the difference in defect dynamics in the two ceramics that may help us understand the susceptibility of the two ceramics to amorphization caused by irradiation.

36 MATERIALS SCIENCE↗
DOE OSTI2022

Electronic and optical characterization of bulk single crystals of cubic boron nitride (cBN)

Cubic boron nitride (cBN) is a relatively less studied wide bandgap semiconductor despite its many promising mechanical, thermal, and electronic properties. We report on the electronic, structural, and optical characterization of commercial cBN crystal platelets. Temperature dependent transport measurements revealed the charge limited diode behavior of the cBN crystals. The equilibrium Fermi level was determined to be 0.47 eV below the conduction band, and the electron conduction was identified as n-type. Unirradiated dark and amber colored cBN crystals displayed broad photoluminescence emission peaks centered around different wavelengths. RC series zero phonon line defect emission peaks were observed at room temperature from the electron beam irradiated and oxygen ion implanted cBN crystals, making this material a promising candidate for high power microwave devices, next generation power electronics, and future quantum sensing applications.

36 MATERIALS SCIENCE↗