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At least 343 records · Page 19
All sourcesNASA NTRSDOE OSTI
DOE OSTI2025

Bayesian SegNet for Semantic Segmentation with Improved Interpretation of Microstructural Evolution During Irradiation of Materials

Understanding the relationship between the evolution of microstructures of irradiated LiAlO2pellets and tritium diffusion, retention and release could improve predictions of tritium performance. Given expert-labeled segmented images of irradiated and unirradiated pellets, we trained Deep Convolutional Neural Networks to segment images into defect, grain, and boundary classes. Qualitative microstructural information was calculated from these segmented images to facilitate the comparison of unirradiated and irradiated pellets. We tested modifications to improve the sensitivity of the model, including incorporating meta-data into the model and utilizing uncertainty quantification. The predicted segmentation was similar to the expert-labeled segmentation for most methods of microstructural qualification, including pixel proportion, defect area, and defect density. Overall, the high performance metrics for the best models for both irradiated and unirradiated images shows that utilizing neural network models is a viable alternative to expert-labeled images.

Oostrom, Marjolein T.↗
DOE OSTI2020

Deep level transient spectroscopy investigation of ultra-wide bandgap ($\bar201$) and ($001$) β -Ga 2 O 3

Here, this work reports on a comprehensive examination of the electrical and thermal properties of vertical Schottky diodes fabricated on ($\bar201$) and ($001$)-oriented samples of β-Ga 2 O 3 . The temperature-dependent current–voltage (I–V) and capacitance–voltage (C–V) data were gathered and analyzed down to 60 K. Deep level transient spectroscopy (DLTS) was used to study bulk and interface defects in the two materials from approx. 325 K down to 60 K. In the bulk ($\bar201$) material, an electron trap was observed at E C -0.46 eV, with a capture cross section of 1.6 x 10 -14 cm 2 and a lambda-corrected maximum trap density of 9.08 x 10 15 cm -3 . These results and others indicate that the electron trap is a strong candidate for the well-known E1 defect in β-Ga 2 O 3 based on recent investigations. Additionally, in the ($\bar201$) material, the smooth modulation typical of interface states is evident at temperatures below 275 K. The ($001$) samples manifested what is likely the E2* electron trap at E C -0.68 eV, with a capture cross section of 1.64 x 10 -15 cm 2 and a lambda-corrected maximum trap density of 8.85 x 10 15 cm -3 . The presence of the E2* defect, in particular, is a contrast to the findings of recent DLTS investigations on β-Ga 2 O 3 , which report that E2* emerged only after low-energy proton irradiation. These results help to further map out the defect signatures found in β-Ga 2 O 3 materials, which are of vital importance in the design and fabrication of future β-Ga 2 O 3 devices.

71 CLASSICAL AND QUANTUM MECHANICS, GENERAL PHYSIC↗
DOE OSTI2025

Diffusion behavior of lanthanide fission products in bcc Fe cladding: A first-principles study

Fuel-cladding chemical interaction poses significant challenges in nuclear reactors, where fission products generated from nuclear fuel interact with Fe-based cladding materials, potentially compromising their structural integrity. This study investigates the diffusion behavior of lanthanide fission products, Lanthanum (La), Cerium (Ce), Praseodymium (Pr), and Neodymium (Nd), within body-centered cubic (bcc) Fe cladding using the density functional theory, nudged elastic band method, and self-consistent mean field theory. Our results reveal significant vacancy binding energies, particularly with the 1st and 2nd nearest neighbors, which diminish beyond the 5th nearest neighbor, with La exhibiting the strongest binding affinity, followed by Nd, Ce, and Pr. The nudged elastic band calculations indicate significant high barriers for the dissociation of 1st nearest neighbor vacancy-solute pairs for all fission products. The tracer diffusion coefficients of these fission products were derived in an Arrhenius form, with a magnetic correction that accounts for the high-temperature paramagnetic state. The significant trapping effect of vacancies caused by a very dilute concentration of fission products reduces vacancy mobility, potentially leading to modifications in point defect supersaturation, void nucleation, and swelling under irradiation. These represent critical challenges for irradiated cladding materials. The tracer diffusion coefficients indicate that Nd diffuses the fastest, followed by La, Ce, and Pr. Furthermore, this study provides essential insights for understanding fission product transport in cladding materials and informs future design strategies to mitigate fuel-cladding chemical interaction, ultimately enhancing nuclear reactor safety and performance.

Diffusion↗
DOE OSTI2025

Self‐Strain Suppression of the Metal‐to‐Insulator Transition in Phase‐Change Oxide Devices

Strongly correlated materials exhibiting phase transitions which can be controlled through external stimuli, such as electric fields, are promising for future computing technologies beyond conventional semiconductor transistors. Devices that take advantage of structural phase transitions have inherent built‐in memory, reminiscent of synapses and neurons, and are thus natural candidates for neuromorphic computing. Of particular interest are phase‐change oxides, which allow for control over the metal‐to‐insulator transition. Here, X‐ray nano‐diffraction structural imaging of micro‐devices fabricated with the archetypal phase‐change material vanadium sesquioxide (V 2 O 3 ) is reported. The devices contain a Ga ion‐irradiated region where the metal‐to‐insulator transition critical temperature is lowered, a useful feature for controlling neuron‐like spiking behavior. Results show that strain, induced by crystal lattice mismatch between the pristine and irradiated material, leads to a suppression of the metal‐to‐insulator‐transition. Suppression occurs within the irradiated region or along its edges, depending on the defect‐distribution and the size of the region. The observed self‐straining effect can extend to other phase‐change oxides and dominate as device dimensions are reduced and become too small to dissipate strain within the irradiated region. The findings are important for phase engineering in phase‐change devices and highlight the necessity to study phase transitions at the nanoscale.

77 NANOSCIENCE AND NANOTECHNOLOGY↗
DOE OSTI2025

Phase-field modeling of radiation-induced composition redistribution: An application to additively manufactured austenitic Fe–Cr–Ni

Multicomponent alloys undergoing irradiation damage develop radiation-induced composition redistribution at point defect sinks such as grain boundaries (GBs) and dislocations. Such redistribution results in undesired changes to their mechanical behavior and corrosion resistance. Additively manufactured alloys proposed for future nuclear applications are expected to demonstrate a distinct response to irradiation owing to their unique microstructure with as-solidified dislocation density and chemical microsegregation. To capture the composition redistribution in such systems, we develop a mesoscale model with coupled evolution of atomic and point defect components in the presence of dislocation density, dislocation heterogeneity, and thermodynamic interactions at the GB. The model is parameterized for an FCC Fe–Cr–Ni alloy as a representative system for austenitic stainless steels, and simulations are performed in 1D and 2D as a function of irradiation temperature, dose, dislocation density, and grain size. Radiation-induced segregation (RIS) characterized by Cr depletion and Ni enrichment is predicted at both the GB and the dislocation cell wall, with RIS being lower in magnitude but wider at the cell wall. Strongly biased absorption of self-interstitials by dislocations is found to suppress Ni enrichment but slightly enhance Cr depletion under certain conditions. Thermodynamic segregation at the GB is predicted to be narrower and opposite in sign to RIS for both Cr and Ni. Importantly, non-monotonic segregation is found to occur when both thermodynamic and RIS mechanisms are considered, providing a novel physical interpretation of experimental observations. The model is expected to serve as a key tool in accelerated qualification of irradiated materials.

additively manufactured microstructure↗
DOE OSTI2024

Effects of irradiation temperature on the microstructure and deuterium retention in γ-LiAlO 2 pellets

This report presents the experimental findings obtained from November 2023 to September 2024. The study aims to investigate the effects of temperature on the microstructure, deuterium (D) retention, and lithium (Li) loss in γ-LiAlO 2 pellets subjected to sequential He + and D + ion irradiation to a high dose. Sequential ion irradiation has been a key method in our previous studies to simulate the behavior of γ-LiAlO 2 under neutron irradiation. In addition to dose and dose rate, irradiation temperature is a critical factor influencing microstructural and compositional changes. LiAl 5 O 8 precipitates have been observed in γ-LiAlO 2 pellets irradiated with reactor neutrons at ~300°C. These precipitates also form during ion irradiation at an elevated temperature of 500°C, but not at 300°C. In our ion irradiation experiments, the dose rate is typically three orders of magnitude higher than that of neutron irradiation, leading to a more rapid damage production. To better emulate the microstructural features in neutron-irradiated pellets using ion irradiation, a higher irradiation temperature is needed to accelerate the diffusion of point defects and enhance defect recovery rates, thereby compensating for the effects of the higher dose rate. The microstructural changes observed are the result of competing processes occurring during ion irradiation.

36 MATERIALS SCIENCE↗
DOE OSTI2022

Irradiation-induced grain boundary facet motion: In situ observations and atomic-scale mechanisms

Metals subjected to irradiation environments undergo microstructural evolution and concomitant degradation, yet the nanoscale mechanisms for such evolution remain elusive. Here, we combine in situ heavy ion irradiation, atomic resolution microscopy, and atomistic simulation to elucidate how radiation damage and interfacial defects interplay to control grain boundary (GB) motion. While classical notions of boundary evolution under irradiation rest on simple ideas of curvature-driven motion, the reality is far more complex. Focusing on an ion-irradiated Pt Σ3 GB, we show how this boundary evolves by the motion of 120° facet junctions separating nanoscale {112} facets. Our analysis considers the short- and mid-range ion interactions, which roughen the facets and induce local motion, and longer-range interactions associated with interfacial disconnections, which accommodate the intergranular misorientation. We suggest how climb of these disconnections could drive coordinated facet junction motion. These findings emphasize that both local and longer-range, collective interactions are important to understanding irradiation-induced interfacial evolution.

74 ATOMIC AND MOLECULAR PHYSICS↗
DOE OSTI2020

Carbon nanotube (CNT) metal composites exhibit greatly reduced radiation damage

Radiation damage of structural materials leads to mechanical property degradation, eventually inducing failure. Secondary-phase dispersoids or other radiation defect sinks are often added to materials to boost their radiation resistance. We demonstrate that a metal composite made by adding 1D carbon nanotubes (CNTs) to aluminum (Al) exhibits superior radiation resistance. In situ ion irradiation with transmission electron microscopy (TEM) and atomistic simulations together reveal the mechanisms of rapid defect migration to CNTs, facilitating defect recombination and enhancing radiation tolerance. The origin of this effect is an evolving stress gradient in the Al matrix resulting from CNT transformation under irradiation, and the stability of resulting carbides. Extreme value statistics of large defect behavior in our simulations highlight the role of CNTs in reducing accumulated damage. Furthermore, this approach to controlling defect migration represents a promising opportunity to enhance the radiation resistance of nuclear materials without detrimental effects.

36 MATERIALS SCIENCE↗
DOE OSTI2026

Cluster Dynamics Simulations of Intra-Granular Fission Gas Bubble Size and Pressure Evolution in UO 2

Fission gases such as xenon (Xe) play a critical role in determining the behavior and response of nuclear fuel. Given that Xe has little solubility in UO 2 , it accumulates and forms bubbles, which significantly impact fuel performance. Intra- and inter-granular bubble nucleation and growth can lead to fuel swelling, and once bubbles interconnect at grain boundaries, fission gas can be released into the plenum. At low temperatures, limited uranium vacancy mobility can restrict swelling, therefore causing the bubbles to become highly pressurized. Consequently, this can induce micro-cracking, promote fission gas release (increasing the likelihood of cladding failure), and even lead to fuel pulverization under accident conditions such as a loss of coolant accident. As bubble evolution is strongly influenced by local temperature and fission rate, markedly different behavior occurs across the radial profile of the fuel pellet. Capturing the mechanisms that underpin bubble evolution is therefore important to predict these behaviors in the fuel. Previous models describing important mechanisms informed by lower length scale simulations have been developed under the NEAMS program. These can describe the evolution of a single bubble type (i.e., single value for radius and pressure) at each position in the pellet, for instance using the Centipede cluster dynamic code. However, in reality, a full distribution in bubble sizes and pressures exists within the microstructure at a given position in the pellet. To address this the cluster dynamics code Xolotl, which can predict Xe and vacancy phase space (i.e., bubble distributions) for intra-granular bubbles, has been used before. Prior work benchmarked the Xolotl code against the Centipede cluster dynamics code to ensure compatibility and to verify that mobile defect properties are adequately transferred between the two codes, along with some physics improvements. In this work, we go further by introducing a physics-based set of improvements that will allow us to accurately predict bubble size distributions and internal bubble pressures under representative UO 2 irradiation conditions. The improvements include (i) coupling bubble-defect reaction energies to a virial equation of state (EOS), (ii) including a bubble surface tension contribution, (iii) incorporating radiation-induced re-solution of Xe and vacancies, (iv) enabling pressure-driven dislocation loop punching through an effective emission of interstitial clusters informed by interstitial loop energetics, (v) accounting for radiation induced athermal diffusion of Xe, and (vi) implementing a Booth-type grain boundary sink representation for all mobile defects and defect clusters. After these modifications, we observe good agreement of Xolotl fission gas bubble size and concentration predictions with legacy experimental measurements. Additionally, it allows the distribution of Xe bubble pressures and radius to also be predicted and compared to data produced through the Advanced Fuels Campaign (AFC) program. Here, we have done this by running simulations under conditions similar to the AFC post-irradiation examination (PIE) samples irradiated at North Anna 2 light water reactor (LWR). Our results shows excellent agreement with these experimental measurements.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Proton radiation effects on electronic defect states in MOCVD-grown (010) β -Ga 2 O 3

The impact of 1.8 MeV proton irradiation on metalorganic chemical vapor deposition grown (010) β-Ga 2 O 3 Schottky diodes is presented. It is found that after a 10.8 × 10 13 cm –2 proton fluence the Schottky barrier height of (1.40 ± 0.05 eV) and the ideality factor of (1.05 ± 0.05 ) are unaffected. Capacitance–voltage extracted net ionized doping curves indicate a carrier removal rate of 268 ± 10 cm –1 . The defect states responsible for the observed carrier removal are studied through a combination of deep level transient and optical spectroscopies (DLTS/DLOS) as well as lighted capacitance–voltage (LCV) measurements. The dominating effect on the defect spectrum is due to the E c -2.0 eV defect state observed in DLOS and LCV. This state accounts for ~75 % of the total trap introduction rate and is the primary source of carrier removal from proton irradiation. Of the DLTS detected states, the E c -0.72 eV state dominated but had a comparably smaller contribution to the trap introduction. These two traps have previously been correlated with acceptor-like gallium vacancy-related defects. Several other trap states at E C -0.36, E C -0.63, and E C -1.09 eV were newly detected after proton irradiation, and two pre-existing states at E C -1.2 and E C -4.4 eV showed a slight increase in concentration after irradiation, together accounting for the remainder of trap introduction. However, a pre-existing trap at E c -0.40 eV was found to be insensitive to proton irradiation and, therefore, is likely of extrinsic origin. Furthermore, the comprehensive defect characterization of 1.8 MeV proton irradiation damage can aid the modeling and design for a range of radiation tolerant devices.

42 ENGINEERING↗
DOE OSTI2022

Ion beam-induced bending of TiO 2 nanowires with bead-like and prismatic shapes

Ion beam irradiation is a promising method to manipulate the composition and shape of nanowires. It causes the formation of crystal defects like vacancies and dislocations, and consequently, a volume expansion within the irradiated region, giving rise to the nanowire bending. The bending effect has been extensively discussed within nanowires with different diameters under ion beams with varying energies and ion fluences. However, the behaviors of nanowires with complicated shapes, which may have non-uniform irradiated regions due to the changing angle of incidence and shadowing effect, have remained largely unknown. Herein, the structural changes and bending of TiO 2 nanowires with both bead-like and prismatic shapes are investigated under a Ga + ion beam. The multi-faceted morphology, and consequently, varying angles of incidence, result in inhomogeneous irradiation and volume expansion. As a result, significant bending is only observed in prismatic nanowires. Since irradiation is confined within the half of nanowires facing the ion beam, the bending of nanowires is reversible by changing the direction of the ion beam. In order to provide insights into the tailoring composition and morphology of nanowires, we anticipate that this finding can establish the beam analog at the nanoscale, the bending of which can be tuned by ion irradiation.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2020

Dislocation loop evolution in F/M steel T91 under in-situ ion irradiation: Influence of the presence of initial dislocations

Ferritic/Martensitic (F/M) steel T91 was irradiated in-situ to 4 and 10 dpa at 470 °C using 1 MeV Kr 2+ . Additionally, the microstructure evolution under irradiation was followed and characterized by in-situ Transmission Electron Microscopy (TEM), in terms of dislocation loop Burgers vector, average size and density as a function of dose. Furthermore, the influence of the presence of dislocations in the pre-irradiated matrix was investigated by comparing the radiation induced defect evolution in grains with and without initial dislocations.

36 MATERIALS SCIENCE↗
DOE OSTI2026

A kinetic model for simulating non-equilibrium mass transport in oxides applied to hematite growth under irradiation

We propose an approach to simulating the dynamic evolution and transport of charged point defects within and through the oxide scales that form during corrosion. The method follows the cluster dynamics formalism widely adopted for radiation damage in solids, which can apply in both quasi-static and far from equilibrium conditions, such as in radiation environments. By treating each charged defect as a cluster of an atomic defect and associated unit charges, the proposed model flexibly allows charge state transitions and shifts in the Fermi level by absorption and emission of charge carriers from point defects in a reaction network with rates constrained by local equilibrium. Applying this model to hematite predicts changes in self-diffusion and oxidation kinetics in irradiation environments, surprisingly showing reduced oxidation rates in many conditions, despite enhanced self-diffusion. In conclusion, the origin of this effect lies in a change in Fermi level induced by excess vacancies formed under irradiation, which in turn suppresses the transport of cation interstitials which facilitate hematite growth.

36 MATERIALS SCIENCE↗
DOE OSTI2022

Session 3: Irradiation and Thermal Stesses - NRC Graphite Behavior Model

Graphite anisotropic microstructure Graphite crystal structure and basal planes Response to irradiation and temperature C-axis and a-axis behavior The effect of porosity Nano-and micro-length scale Microscopic to macroscopic response Material property changes Defect length-scale factors Internal stress build-up Irradiation induced dimensional change Temperature effects Stress increases & property changes

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Dynamic Deformation in Nuclear Graphite and Underlying Mechanisms

Time-dependent deformation in nuclear graphite is influenced by the creation and migration of radiation-induced defects in the reactor environment. This study investigates the role of pre-existing defects such as point defect clusters and Mrozowski cracks in nuclear graphite IG-110. Separate specimens were irradiated with a 2.8 MeV Au 2+ beam with a fluence of 4.38 × 10 14 cm −2 and an 8 MeV C 2+ beam with a fluence of 1.24 × 10 16 cm −2 . Microscopic specimens were either mechanically loaded inside a transmission electron microscope (TEM) or subjected to ex situ indentation-based creep loading. In situ TEM tests showed significant plasticity in regions highly localized around the Mrozowski cracks, resembling slip or ripplocation bands. Slip bands were also seen near regions without pre-existing defects but at very high stresses. Ex situ self-ion irradiation embrittled the specimens and decreased the creep displacement and rate, while heavy ion irradiation resulted in the opposite behavior. We hypothesize that the large-sized gold ions (compared to the carbon atoms) induced interplanar swelling as well as cross-plane channels for increased defect mobility. These findings illustrate the role of pre-existing defects in the dynamic relaxation of stresses during irradiation and the need for more studies into the radiation environment’s impact on the mechanical response of nuclear graphite.

creep deformation↗
DOE OSTI2021

Potential defect structure effects in the volatility separation of sub-picogram quantities of fission and activation products from an irradiated UO2 matrix

Here, the use of fluorination as a rapid separation technique for quantification of trace-mass, short-lived isotopes is illustrated. Single crystal 238 UO 2 was exposed to a 14 MeV neutron source to produce mixed fission and activation products. Selective removal of trace volatile fluorides from non-volatile ones produced two results of interest. A decrease in background produced better detection and quantification of select analytes across the gamma spectrum. Second, at the very low burnup condition described, the volatility behaviors of common isotopes were altered relative to their well-established behaviors in higher burnup metal and metal oxide fuels.

Fission and activation product separations↗
DOE OSTI2021

Structural damage response of lanthanum and yttrium aluminate crystals to nuclear collisions and electronic excitation: Threshold assessment of irradiation damage

A comparative analysis is performed on the structural damage response and associated mechanisms in lanthanum aluminate and yttrium aluminate crystals under various irradiation conditions by a combination of experimental and theoretical approaches. Under low-energy Au+ irradiation, the damage accumulation curve shows a higher damage rate for LaAlO 3 crystals than YAlO 3 crystals. The relatively low irradiation tolerance of LaAlO 3 to the action of nuclear collisions is ascribed to the large amorphization cross-section and effective cross-section for defect-stimulated amorphization. Under swift Ar 12+ , Ni 19+ and Kr 17+ irradiation with different ion energies and velocities, the formed highly-disordered/amorphous latent tracks with different morphologies in pristine and predamaged crystals are discussed, and the corresponding electronic energy loss and lattice temperature thresholds are quantitatively determined. Compared to YAlO 3 , LaAlO 3 exhibits lower sensitivity and higher damage tolerance to the electronic energy loss process, attributing to its relatively high recrystallization efficiency during the rapid quenching process. Furthermore, the introduction of lattice defects into LaAlO 3 and YAlO 3 crystals considerably enhances the sensitivity and intensity of thermal spike response to the electronic energy loss, and the induced effective modification of track morphologies demonstrates the synergistic effect between the electronic energy loss and pre-existing defects created by nuclear collisions. In this case, even under the action of electronic energy loss below the threshold, the lattice temperature in the nuclear-collision damaged crystalline system could still meet the criterion for track production. The irradiation energy deposited to atoms and induced lattice temperature evolution discussed in this work provide a deeper insight into the complex processes involved in irradiation-induced latent track behaviors.

36 MATERIALS SCIENCE↗
DOE OSTI2020

Irradiation-induced segregation at dislocation loops in CoCrFeMnNi high entropy alloy

To understand the redistribution of alloying elements in high entropy alloys under irradiation, we irradiated a CoCrFeMnNi alloy with 1 MeV Kr ions at room temperature and at 500°C, and characterized with atom probe tomography and transmission electron microscopy. At 500°C, Co and Ni were enriched around the interstitial, faulted and perfect, dislocation loops resulted from the ion irradiation. In contrast, no segregation was observed at room temperature. The inverse Kirkendall effect through vacancy flux, as opposed to the interstitial binding mechanism, was the primary underlying process attributing to the observed segregation. In addition, a ring-shaped segregation pattern was observed at the faulted dislocation loops, indicating a non-equilibrium nature of the defect clustering and solute segregation process in CoCrFeMnNi under irradiation at high temperature.

36 MATERIALS SCIENCE↗