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At least 235 records · Page 13

Determination of Thermal Expansion, Defect Formation Energy, and Defect-Induced Strain of α-U Via ab Initio Molecular Dynamics

Uranium (U) is often alloyed with molybdenum (Mo) or zirconium (Zr) in order to stabilize its high-temperature body-centered cubic phase for use in nuclear reactors. However, in all metallic fuel forms, the α phase of U remains in some fraction. This phase decomposition due to temperature or compositional variance can play an outsized role on fuel performance and microstructural evolution. Relatively little is known about fundamental point defect properties in α -U at non-zero temperatures, from either computational or experimental studies. This work performs the first thorough evaluation of the α phase of U via ab initio molecular dynamics (AIMD). A number of thermophysical properties are calculated as a function of temperature, including equilibrium lattice parameters, thermal expansion, and heat capacity. These results indicate a two-region behavior, with the transition at 400 K. The thermal expansion/contraction in the a / b direction occurs rapidly from 100 up to 400 K, after which a more linear and gradual change in the lattice constant takes place. The volumetric expansion matches experiments quantitatively, but the individual lattice constant expansion only matches experiments qualitatively. Point defect formation energies and induced lattice strains are also determined as a function of temperature, providing insight on defect populations and the fundamentals of irradiation growth in α -U. Interstitials induce significantly more strain than vacancies, and the nature of that strain is highly dependent on the individual lattice directions. The direction of point defect-induced lattice strain is contrary to the irradiation growth behavior of α -U. This work shows that isolated point defects cannot be the primary driving force responsible for the significant irradiation-induced growth of α -U observed experimentally.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Ab-initio informed cluster dynamics simulation of self- and Xe diffusivity in uranium mononitride under irradiation

Uranium mononitride (UN) is one of the ceramic nuclear fuel alternatives to oxide fuels considered for light water reactor and advanced reactor designs, as it presents significant advantages such as high uranium density (better economics) and high thermal conductivity and melting point (increased safety). Self- and fission gas diffusivities need to be better understood, given that they influence key fuel performance phenomena like swelling and fission gas release. Recently, radiation enhanced diffusivity was investigated in UN by means of cluster dynamics simulations relying on empirical potential-based parameterizations, the reliability of which highly depends on the interatomic potential accuracy. Here, in this work, we refine this approach by determining, using ab-initio calculations, the properties of defect clusters containing vacancies, self-interstitials and Xe impurities. We also consider larger clusters than previous studies. The obtained dataset (formation enthalpies, entropies, and migration barriers) is used to parameterize a cluster dynamics model of mobile clusters, and to calculate the defect cluster concentrations under irradiation. This gives us access to the radiation enhanced self- and fission gas diffusivities. Although the resulting diffusivities are close to the values reported in the literature, we find important qualitative differences in the diffusion mechanisms. Capturing the correct mechanisms is crucial to properly describe the chemistry and fission rate dependence of the model.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Micro X-ray computed tomography examination of mini plate fuel with hot isostatic pressed aluminum cladding

In order to minimize proliferation risks and improve security of nuclear material, the United States high performance research and test reactors (USHPRR) program is tasked with converting nuclear reactors that are fueled with highly enriched uranium (HEU) fuels to operate with low-enriched uranium (LEU) nuclear fuels. One favorable LEU fuel configuration is plate fuel with a metallic uranium-molybdenum foil clad within an aluminum alloy (AA 6061). In this fuel, the aluminum cladding is bonded with a hot isostatic pressing (HIP) method to seal the cladding around the fuel meat. However, the HIP process parameters influence the cladding performance, as a defective or incomplete bond can cause a pathway for corrosion. Micro X-ray computed tomography (XCT), a nondestructive technique that provides volumetric imaging, can be applied to inspect fuel plate cladding at the engineering scale. In this work, XCT methodology was developed and successfully utilized to not only observe the bond line of unirradiated mini fuel plates, but to also identify subsurface abnormalities in the plates’ cladding. Importantly, in future work, this technique could be applied to fuel plates, pre- and post-irradiation, to quantify irradiation effects on cladding defects and bond line integrity.

36 MATERIALS SCIENCE↗

Vacancy formation energies and migration barriers in multi-principal element alloys

Multi-principal element alloys (MPEAs) continue to garner interest as structural and plasma-facing materials due to their structural (phase) stability and increased resistance to radiation damage. Despite sensitivity of mechanical behavior to irradiation and point-defect formation, there has been scant attention on understanding vacancy stability and diffusion in refractory-based MPEAs. Using density-functional theory, we examine vacancy stability and diffusion barriers in body-centered cubic (Mo 0.95 W 0.05 ) 0.85 Ta 0.10 (TiZr) 0.05 . The results in this MPEA show strong dependence on environment, originating from local lattice distortion associated with charge-transfer between neighboring atoms that vary with different chemical environments. We find a correlation between degree of lattice distortion and migration barrier: (Ti, Zr) with less distortion have lower barriers, while (Mo, W) with larger distortion have higher barriers, depending up local environments. Under irradiation, our findings suggest that (Ti, Zr) are significantly more likely to diffuse than (Mo, W) while Ta shows intermediate effect. Finally, material degradation caused by vacancy diffusion can be controlled by tuning composition of alloying elements to enhance creep strength at extreme operating temperatures and harsh conditions.

36 MATERIALS SCIENCE↗

Kinetic interplay between chemical short-range order and grain boundaries in NiCoCr alloys under irradiation

Chemical short-range order (CSRO) and grain boundary (GB) engineering are routes to enhance radiation damage tolerance in alloys. Here, we reveal that CSRO and GB interact in a sink-strength-dependent manner under irradiation in NiCoCr. Near a weak sink (Σ3 GB), CSRO reduces defect cluster growth by slowing interstitial diffusion and enhancing vacancy-interstitial recombinations. In contrast, near strong sinks such as Σ5 GBs, CSRO and GB act competitively for interstitial accumulation but synergistically to suppress large stacking-fault tetrahedra growth via enhanced recombination. Such mechanistic duality underscores the need for coordinated control of CSRO stability and GB sink strength to enhance radiation damage tolerance.

36 MATERIALS SCIENCE↗

Preferential hole defect formation in monolayer W Se 2 by electron-beam irradiation

Monolayer transition-metal dichalcogenides (TMDCs) have been extensively studied due to their wide range of physical properties and applications. It has been demonstrated that the electron beam in a transmission electron microscope (TEM) or scanning TEM (STEM) generates chalcogen vacancies that agglomerate into dispersed linelike or holelike defects. Here we employ a STEM electron beam and demonstrate that, in WSe 2 , beam-induced chalcogen vacancies initially form certain dispersed multivacancy structures, as seen in TMDCs in prior work. However, with suitable control of the STEM focused electron beam, these multivacancies gradually evolve into a dense network of ten-, 12-, 14-, and 16-member ring round holes, whereas the same process leads predominantly to chalcogen-vacancy line defects in other trigonal-prismatic TMDCs. Density functional theory calculations find that the underlying atomic-scale processes lead preferentially to defect structures that lower the total energy so that we are able to track the formation of the observed multivacancy complexes, which then lead to the formation of dense large round holes in WSe 2 . The same processes in WS 2 , however, are energetically unfavorable, while linear multivacancy defects are preferred, as observed. The demonstrated control of the formation of unique high-density round holes in WSe 2 has potential for applications such as atomic and molecular sieving.

36 MATERIALS SCIENCE↗

Effects of dose rate on the microstructure and deuterium retention in γ-LiAlO 2 pellets

This report presents experimental findings obtained from November 2024 to September 2025. Defect accumulation and microstructural evolution during ion irradiation at elevated temperatures are governed by two competing processes: defect production, driven by dose rate, and defect recovery, controlled by defect diffusion, interaction, and annihilation. At a given dose, the resulting microstructural evolution depends on both the dose rate and irradiation temperature. As a continuation of our FY24 tritium science project, which investigated temperature effects at a fixed dose rate, this study focuses on dose-rate effects at a fixed temperature to provide deeper insights into the irradiated microstructure and compositional changes in γ-LiAlO 2 pellets. The study aims to elucidate the impact of dose rate on microstructure, precipitate morphology, deuterium retention, and lithium volatility in γ-LiAlO 2 pellets under sequential 120 keV He + and 80 keV D 2 + ion irradiation. Three dose rates of 7.3×10 4 , 2.9×10 4 and 6.8×10 5 dpa/s were applied to achieve the same total ion fluence of 2×10 17 (He + +D + )/cm 2 at 500 °C, corresponding to a maximum combined dose of 7.55 dpa at ~255 nm. The irradiated pellets were subsequently characterized using scanning transmission electron microscopy (STEM) and time-of-flight secondary ion mass spectrometry (ToF-SIMS). The microstructural response of γ-LiAlO 2 to ion irradiation was found to be strongly dose-rate dependent. At medium and high dose rates, irradiation produced a surface amorphized layer and an underlying crystalline layer containing LiAl 5 O 8 precipitates, with implanted gases accumulating and forming blisters at the crystalline-amorphous interface. It remains to be investigated whether the amorphized layer was produced by He + ion irradiation prior to D 2 + ion irradiation. In contrast, at low dose rates, the material remained crystalline, with cavities, likely gas-filled, distributed around precipitates, within the γ-LiAlO 2 matrix, and along grain boundaries. While precipitate morphology exhibited anisotropy, their size showed little sensitivity to dose rate in the applied range of this study. This result, however, does not rule out the possibility that further reductions in dose rate could influence precipitate size. High dose-rate irradiation enhanced protonium–deuterium isotopic exchange and lithium depletion in the amorphized region. Collectively, the results show that dose rate governs amorphization, gas redistribution, isotopic exchange, and lithium depletion, providing important insights into the mechanisms underlying structural evolution in γ-LiAlO 2 pellets under reactor-relevant irradiation conditions.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

Simultaneous exposure to neutron radiation and hydrogen environment: Effect on hydrogen retention

Here, this study demonstrates a novel methodology that simultaneously exposes materials to both neutron irradiation and a hydrogen environment, enabling study on in pile interaction between radiation-induced defects and hydrogen in tungsten. The hydrogen environment was established by releasing hydrogen from vanadium hydride within an irradiation capsule. The hydrogen pressure during irradiation was estimated as 14.2 Torr. Post-irradiation thermal desorption spectroscopy analysis showed increased hydrogen retention in irradiated tungsten compared to unirradiated samples. Two dominant desorption peaks at ∼470 °C and ∼700 °C were observed in irradiated and unirradiated tungsten samples, suggesting similar trapping mechanisms between the two samples. However, an additional desorption peak at ∼800 °C in irradiated tungsten from the hydride capsule suggests an interplay between irradiation induced defects and hydrogen exposure, leading to the formation of additional hydrogen trapping sites. These findings provide critical insights into hydrogen retention in tungsten under fusion-relevant conditions and stimulate future studies to investigate the synergistic effects of neutron irradiation and hydrogen exposure.

08 HYDROGEN↗

Effects of recoil spectra and electronic energy dissipation on defect survival in 3C-SiC

The coincidence of electronic and damage energy dissipation from energetic ions to an atomic lattice can significantly affect damage production along the ion trajectory due to spatial overlap of inelastic and elastic processes. Damage production and disordering in single crystal 3C-SiC from 5 MeV Si and 10 MeV Au ions is investigated using ion-channeling experiments. While defects are created by damage energy dissipation via elastic scattering, electronic energy dissipation via electron-phonon coupling decreases defect survival along the ion trajectory for Si ions. Finally, the more energetic recoil spectrum for 10 MeV Au ions leads to weaker spatial coupling of electronic and damage energy dissipation processes, and damage production is only weakly affected.

36 MATERIALS SCIENCE↗

Primary intrinsic defects and their charge transition levels in β–Ga 2 O 3

A steady-state photocapacitance (SSPC) setup directly connected to the beamline of a MeV ion implanter is utilized to study primary intrinsic defects in β – Ga 2 O 3 generated by He implantation at cryogenic temperatures (120 K). At low temperatures, the migration of defects is suppressed, and hence the generation of primary intrinsic defects is expected to prevail. SSPC measurements reveal defect-related optical transitions in halide vapor-phase epitaxy (HVPE) -grown β – Ga 2 O 3 thin films with onset energies at 1.3 ( T 1 ) , 1.7 ( T 2 ) , 1.9 ( T 3 ) , 2.6 ( T 4 ) , 3.7 ( T 5 ) , and 4.2 eV ( T 6 ) . T 2 , T 4 , T 5 , and T 6 were observed in as-received HVPE-grown β – Ga 2 O 3 thin films, whereby T 2 is only sporadically observed. The introduction rates for T 3 , T 4 , as well as T 6 indicate an origin related to primary intrinsic defects. Notably, T 1 and T 3 are only observed after He implantation at cryogenic temperatures. Hybrid-functional calculations were performed to estimate the optical absorption cross-section spectra for the gallium ( Ga i ) and oxygen ( O i ) interstitials as well as the corresponding vacancies ( V Ga and V O , respectively), and compared with the measured onsets for optical absorption found by SSPC measurements. Indeed, we propose T 3 to be associated with Ga i ( + / + 2 ) and/or V GaI ( - 3 / - 2 ) , while T 4 is suggested to be related to V O K ( 0 / + ) ( K = I , II , III ) and/or V GaII ( - 3 / - 2 ) . Additionally, several further charge-state transition levels associated with V GaI and V GaII may contribute to T 4 and T 6 . We further studied the kinetics of the defects created with He implantation by exposing the sample to room temperature. The kinetics observed for T 3 and T 4 further support the proposed assignments of the corresponding defect signatures.

36 MATERIALS SCIENCE↗

Microstructure and in-service degradation of baffle former bolts – in-core components of light water reactors

This report presents a microstructure evaluation of the baffle former bolt, a commercial pressurized water reactor component. Using SEM, EDS, and EBSD analysis the material was identified as AISI 316 steel with an annealed austenite microstructure. Findings include abnormal grain growth in a fraction of the grains and the absence of retained ferrite. Notable observations include an overlapping of pre- and post-irradiation deformation, including defect-free channel formation as an active deformation mechanism in the near-surface layer of BFB. Specimen surfaces, exposed to high-temperature, high-pressure water, exhibited signs of in-service corrosion degradation. EBSD and EDS analyses highlighted intergranular corrosion, possible grain boundary oxidation with depths less than 3 µm, and, unexpectedly, short cracks filled with Cr-rich oxides measuring around 5-6 µm. The presence of defect-free channels from EBSD data suggests episodes of high mechanical stresses during service. The specimens provide a potentially unique insight into the in-service degradation of NPP components, strain localization, and crack initiation processes. An intact corrosion layer reveals a complex component's in-service history. Further research is underway for a detailed analysis.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

Production of Precursor Materials for the Development of Iron-Based Superconductor (IBS) Wire

The goal of this agreement will be devoted to the understanding of the radiation-induced defects landscape produced on irradiated bulk samples and of the intrinsic properties of the stoichometric phases with special regards to the difference interaction with multiple chemical additions and doping elements. Samples will be produced by ENEA Frascati per Fermilab specifications and then shipped to Fermilab for the characterization testing and the execution of irradiation experiments.

75 CONDENSED MATTER PHYSICS, SUPERCONDUCTIVITY AND↗

Web-based Preprocessing and Visualization of 3D FIB Tomography Data for Nuclear Fuel Characterization

Three-dimensional (3D) focused ion beam (FIB) tomography enables reconstruction of internal nuclear fuel features that can't be fully evaluated through surface imaging alone. This capability supports characterization of fuel constituents and defects under thermal and irradiation conditions relevant to microreactor development. However, large tomography datasets can create data-handling, loading, and visualization challenges, especially when image-stack preparation and file conversion must be completed with separate tools. The Computational Ultraspatial Tomography Toolkit for High-Resolution Object Analysis Tools (CUTTRHOAT) is an open-source web application being developed to display FIB tomography datasets available through the Nuclear Research Data System (NRDS). The current alpha version requires prepared HDF5 datasets and has limited integrated data-preparation capabilities. This project improves CUTTHROAT by adding dataset-folder selection, automatic input detection, dataset scanning, missing-slice identification, blank-slice insertion, and image-stack-to-HDF5 conversion. Two applications will be compared: the baseline CUTTHROAT alpha workflow and the updated application containing the integrated data-handling and preprocessing functions. Evaluation will consider dataset detection accuracy, conversion success, loading time, rendering responsiveness, application stability, and user interaction. Preliminary results demonstrate successful loading of existing HDF5 files and converted image stacks, while testing also identified performance reductions caused by excessive blank-slice generation. The updated workflow reduces reliance on external preparation tools and supports more direct movement from image stacks to color-code 3D visualization. Future work includes refining missing-slice handling, integrating additional preprocessing functions, like a denoising feature, parsing TIFF metadata for automatic voxel scaling, and adding manual X, Y, and Z voxel-spacing inputs for PNG and JPEG.

36 - MATERIALS SCIENCE↗

Microstructural Engineering of Cu-Rich Nanoprecipitate formation in NiCoFeCrCu0.12 High-Entropy Alloy via Severe Plastic Deformation for Enhanced Irradiation Tolerance

This study demonstrates a defect-engineering approach for controlling Cu-rich precipitates in FeNiCrCoCu0.2 high-entropy alloys (Cu-HEAs), delivering a novel pathway for next-generation nuclear reactor materials with superior irradiation resistance. This work establishes that severe plastic deformation (SPD) processing via Shear Assisted Processing and Extrusion (ShAPE) and Friction Stir Layer Deposition (FSLD) creates dense dislocation networks and subgrain boundaries that fundamentally alter precipitation behavior under identical thermal treatments. Atom probe tomography (APT) indicates that SPD produces a metastable, atomically homogeneous solid solution that, upon moderate heat treatment (500°C/10 hour), develops remarkedly stronger Cu clustering than the as-cast counterpart. High-temperature exposure (800°C/100 h) produces near-pure Cu precipitates (~90 at% Cu) with significantly enhanced defect-sink efficacy in SPD-processed alloys: precipitate sizes of 50-60 nm and number densities of 2.7-3.8 × 10¹7 m?³, compared to 89 nm and 0.44 × 10¹7 m?³ in as-cast materials. Collectively, the findings establish defect-mediated precipitation control as a scalable, high-impact route to tailor sink density and distribution in HEAs, enabling microstructures optimized for irradiation tolerance and mechanical robustness in nuclear reactor environments.

Meher, Subhashish↗

Modeling of fission gas diffusion and release for Gd 2 O 3 doped UO 2

Uranium dioxide (UO 2 ) is the primary nuclear fuel in light water reactors, and its excess neutronic reactivity can be controlled by adding burnable absorbers, such as Gd 2 O 3 . This burnable absorber has a large neutron absorption cross-section, lowering the high reactivity of the reactor's initial fuel load. However, there needs to be more understanding of how added Gd 2 O 3 influences the properties of UO 2 under irradiation. To understand the behavior of defects and fission gas in the UO 2 /Gd 2 O 3 system under irradiation, we use cluster dynamics modeling supported by density functional theory calculations. First, we calculate the formation energies of Gd point and cluster defects, and evaluate the temperature-dependent defect concentrations using the defect formation energies and entropies. We show that Gd is soluble in UO 2 , introducing a negative charge in the system. Using this information, we adapted the cluster dynamics code Centipede to model the influence of Gd on U self-diffusion and Xe diffusion in UO 2 with 10 wt% Gd 2 O 3 . Also, we analyzed the Xe diffusion as a function of Gd 2 O 3 concentration, showing that the Xe diffusivity is decreased, which means that the athermal diffusivity due to electronic stopping persists at higher temperatures. In conclusion, the decrease in Xe diffusion means that more Xe stays in the matrix, decreasing the Xe release, and lowering its influence of fission gas release on the thermomechanical properties of UO 2 .

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Athermal annealing of pre-existing defects in crystalline silicon

Systematic investigations of electronic energy loss (S e ) effects on pre-existing defects in crystalline silicon (Si) are crucial to provide reliance on the use of ionizing irradiation to anneal pre-existing defects, leading to successful implementation of this technology in the fabrication of Si-based devices. In this regard, the S e effects on nonequilibrium defect evolution in pre-damaged Si single crystals at 300 K has been investigated using intermediate-energy ions (12 MeV O and Si ions) that interact with the pre-damaged surface layers of Si mainly by ionization, except at the end of their range where the nuclear energy loss (S n ) is no longer negligible. Furthermore, under these irradiation conditions, experimental results and molecular dynamics simulations have revealed that pre-existing disorder in Si can be almost fully annealed by subsequent irradiation with intermediate-energy incident ions with S e values as low as 1.5–3.0 keV/nm. Selective annealing of pre-existing defect levels in Si at room temperature can be considered as an effective strategy to mediate the transient enhanced diffusion of dopants in Si. This approach is more desirable than the regular thermal annealing, which is not compatible with the processing requirements that fall below the typical thermal budget.

36 MATERIALS SCIENCE↗

Defect evolution in Ni and solid-solution alloys of NiFe and NiFeCoCr under ion irradiation at 16 and 300 K

Single-phase concentrated solid-solution alloys (SP-CSAs) have shown unique chemical complexity at the levels of electrons and atoms, and their defect evolution is expected to be different from conventional dilute alloys. Single crystals of Ni, NiFe and NiFeCoCr are chosen as model systems to understand the chemical complexity on defect formation and damage accumulation in SP-CSAs under ion irradiation. The high-quality crystals were irradiated at 16 and 300 K to different ion fluences, to form irradiated region with little to heavy damages. Additionally, the ion-induced damage was determined using Rutherford backscattering spectrometry technique along a channeling direction (RBS/C) and the level of lattice damage in irradiated Ni and SP-CSAs was quantified from Monte Carlo (MC) simulations. The results are interpreted using the Multi Step Damage Accumulation model to reveal material damage accumulation kinetics. Key findings of the study are that in case of room temperature irradiations the damage level measured for complex alloys at the highest irradiation fluence of 2 × 10 15 cm -2 (~3 dpa) is significantly higher than that obtained for pure nickel samples and suggest two-step damage accumulation process with a defect transformation taking place at a fluence of about 1.5 × 10 15 cm -2 . Furthermore, structural and damage kinetic differences clearly imply that, with increasing degree of chemical complexity and high solid-solution strengthening effects from Ni to NiFe and to NiFeCoCr, the enhanced lattice stiffness resists to randomization of atomic configurations and inhibits the growth of extended defects.

36 MATERIALS SCIENCE↗