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At least 91 records · Page 5
All sourcesNASA NTRSDOE OSTI
DOE OSTI2023

Tailoring the Angular Mismatch in MoS 2 Homobilayers through Deformation Fields

Ultrathin MoS 2 has shown remarkable characteristics at the atomic scale with an immutable disorder to weak external stimuli. Ion beam modification unlocks the potential to selectively tune the size, concentration, and morphology of defects produced at the site of impact in 2D materials. Combining experiments, first-principles calculations, atomistic simulations, and transfer learning, it is shown that irradiation-induced defects can induce a rotation-dependent moiré pattern in vertically stacked homobilayers of MoS 2 by deforming the atomically thin material and exciting surface acoustic waves (SAWs). Furthermore, the direct correlation between stress and lattice disorder by probing the intrinsic defects and atomic environments are demonstrated. The method introduced in this paper sheds light on how engineering defects in the lattice can be used to tailor the angular mismatch in van der Waals (vdW) solids.

2D materials↗
DOE OSTI2024

Understanding the effect of crystal anisotropy on grain growth, texturing and transport via the orthorhombic ?-U system [Slides]

The alpha phase of uranium exhibits the orthorhombic crystal structure, resulting in significant physical property anisotropy. These complex properties make the material challenging to work with in engineering settings but provide a rich arena to investigate the fundamental, multi-scale effects of anisotropy on physical behaviors such as grain growth and microstructure evolution under irradiation. We apply molecular dynamics and phase field modeling to study mass transport and grain growth in alpha-uranium. We characterize the mobilities of crystalline defects and the interfacial energy, and We find that the grain boundary energy is highly variable depending on the interface structure and that the relative rates of defect transport change with temperature. We also find that anisotropic thermal expansion has a significant impact on grain growth kinetics and texture development. Our results are used to explain experimental observations and to provide a basis for further hypotheses into the complex physical behavior of alpha-uranium.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Molecular dynamics simulation of vacancy and void effects on strain-induced martensitic transformations in Fe-50 at.% Ni model concentrated solid solution alloy

Strain-induced martensitic transformations can improve the strength and ductility of some face centered cubic (FCC) metals and alloys. Irradiation-induced defects such as vacancies, dislocation loops, and voids activate martensitic transformations over a wider range of mechanical loading conditions than in pristine material. Furthermore, the mechanisms underlying irradiation-enabled martensite transformations remain unclear. In this work, we use molecular dynamics simulations to study the effect of vacancies and voids on strain-induced martensitic transformations in a model concentrated solid solution alloy Fe-50 at.% Ni. It has been found that single vacancies have no resolvable effect on the transformation because they reduce the stacking fault energy by a relatively insignificant margin and do so only if the vacancy is located on the stacking fault plane. Voids, however, activate the martensite transformation through shear strain accumulation around the void due to dislocation pileup. The larger the void, the more pronounced this effect.

36 MATERIALS SCIENCE↗
DOE OSTI2022

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↗
DOE OSTI2023

Enhanced Exciton-to-Trion Conversion by Proton Irradiation of Atomically Thin WS 2

Defect engineering of van der Waals semiconductors has been demonstrated as an effective approach to manipulate the structural and functional characteristics toward dynamic device controls, yet correlations between physical properties with defect evolution remain underexplored. Here, using proton irradiation, we observe an enhanced exciton-to-trion conversion of the atomically thin WS 2 . The altered excitonic states are closely correlated with nanopore induced atomic displacement, W nanoclusters, and zigzag edge terminations, verified by scanning transmission electron microscopy, photoluminescence, and Raman spectroscopy. Density functional theory calculation suggests that nanopores facilitate formation of in-gap states that act as sinks for free electrons to couple with excitons. The ion energy loss simulation predicts a dominating electron ionization effect upon proton irradiation, providing further evidence on band perturbations and nanopore formation without destroying the overall crystallinity. This study provides a route in tuning the excitonic properties of van der Waals semiconductors using an irradiation-based defect engineering approach.

77 NANOSCIENCE AND NANOTECHNOLOGY↗
DOE OSTI2024

Transmission electron microscopy of ion irradiated ODS MA956 samples

Oxide dispersion strengthened (ODS) alloys are promising candidate materials for the next generation advanced nuclear reactors due to their superior irradiation resistance and mechanical properties. To better understand the effect of irradiation on MA956, it is essential to study higher dose (50-100 dpa) samples, so that the general trend of microstructural evolution and the resulting radiation-hardening can be deduced. Currently, ion irradiations are considered the only way to achieve doses beyond ~50 dpa in a practical time frame relevant to alloy and welding development programs. This dataset contains TEM characterization results of ODS MA956 samples that were ion irradiated under different conditions: Sample #5 (1.25 dpa at 190℃); Sample #9 (50 dpa at 190℃); Sample #15 (1.25dpa under 320℃); and Sample 17 (25 dpa at 320℃). TEM characterization focused on the irradiation induced defects (dislocation lines and loops) using the on-zone axis bright field STEM technique. These data were collected using a FEI Tecnai G2 F30 S/TEM at Microscopy and Characterization Suite (MaCS) at Center for Advanced Energy Studies (CAES), Idaho Falls, ID. This project (ion irradiation and TEM studies) was supported by the U.S. Department of Energy, Office of Nuclear Energy under DOE Idaho Operations Office Contract DE-AC07-05ID14517 as part of Nuclear Science User Facilities award #18-14784 (PI: Ramprashad Prabhakaran, PNNL).

Prabhakaran, Ramprashad↗
DOE OSTI2024

Response of 11 B enriched ZrB 2 ultra-high temperature ceramic to neutron irradiation at elevated temperatures

ZrB 2 , an ultra-high temperature ceramic (UHTC) is being considered for use in fusion reactor first-wall structures, yet its response to irradiation remains poorly understood. This study employed scanning/transmission electron microscopy (S/TEM), synchrotron X-ray diffraction (XRD), finite element calculations, and thermal property measurements to thoroughly investigate the neutron-irradiation effects on 11 B-enriched ZrB 2 . Neutron irradiations were conducted at 220 °C and 620 °C, with a neutron fluence of 2.2 × 10 25 neutron/m 2 (energy > 0.1 MeV), resulting in 3.9 dpa and 4200 appm He. The study revealed the unusual prevalence of prism loops and a > c anisotropic lattice swelling, likely linked to the low c/a ratio of ZrB 2 , leading to grain boundary microcracking. Reducing the grain sizes was effective in reducing intergranular cracking and macroscopic swelling. The observation of cavities in ZrB 2 irradiated at 620 °C, as opposed to 220 °C, prompts questions about the temperature at which vacancies in ZrB 2 become mobile, and the role of neutron absorption by 10 B in elevating irradiation temperatures. Isotopic enrichment in 11 B proves to be a viable strategy for mitigating helium production in transition-metal diborides, which is a critical consideration for nuclear applications. Irradiation-induced defects reduce the thermal diffusivity and conductivity of ZrB 2 by a factor of 4–9, which has important implications for its role as a plasma-facing material in fusion reactors that drive high heat fluxes through first-wall materials. Here, this comprehensive study lays the foundation for understanding ZrB 2 behavior under neutron irradiation and highlights important phenomena to consider for various material applications.

36 MATERIALS SCIENCE↗
DOE OSTI2022

The effect of helium on cavity swelling in dual-ion irradiated Fe and Fe-10Cr ferritic alloys

We report Ferritic-martensitic (FM) steels for in-vessel components in proposed fusion reactors are expected to suffer from high levels of displacement damage and helium (He) generation by neutron transmutation. However, a thorough understanding of the He synergistic effects on the cavity swelling in FM steels is still not well established. To gain fundamental insights into the He effect on cavity swelling, high purity Fe and Fe-10 wt.% Cr ferritic model alloys were irradiated with 8 MeV Ni ions and co-implanted He ions at 400 to 550 °C up to 30 displacements per atom (dpa) with He implantation rates of 0.1, 10 and 50 appm He/dpa. The current study focuses on the 50 appm He/dpa, 500 °C behavior vs. 0.1 and 10 appm He/dpa. Irradiation-induced defects, including cavities, dislocation loops, and dislocation networks were characterized using transmission electron microscopy (TEM). In the grain interior, a bimodal cavity size distribution was observed in the 10 and 50 appm He/dpa samples, but not for 0.1 appm He/dpa. Cavity swelling was maximized at intermediate He implantation rates of ~10 appm He/dpa for both ion-irradiated Fe and Fe-10Cr alloys. The cavity swelling behavior as a function of He implantation rate appears to be controlled by the He/dpa-dependent variation of cavity sink strengths. Treating small bubbles as biased sinks for interstitial absorption can significantly increase the ratio of biased to unbiased sink strengths (Q) and results in maximized cavity swelling for a Q ratio close to one.

36 MATERIALS SCIENCE↗
DOE OSTI2023

FeCrAl fuel/clad chemical interaction in light water reactor environments

This article investigates the fuel-cladding chemical interaction (FCCI) behavior of two commercial FeCrAl alloys, APMT composition (Fe-21Cr-5Al-3Mo wt.%) and C35M (Fe-13Cr-5Al-2Mo-0.2Si-0.03Y wt.%), after neutron irradiation. “H-cup” diffusion multiples of FeCrAl alloys and ceramic UO 2 fuel were irradiated at a temperature of ~300 °C to a total estimated burnup of 26 GWd/tHM. Post-irradiation Examination results demonstrate the excellent degradation resistance of FeCrAl alloys as accident tolerant fuel (ATF) cladding materials in light water reactor conditions. The study concludes that there was no irradiation-induced defects observed in either of the two commercial FeCrAl claddings. The formation of amorphous Al/U mixed oxide was observed at the fuel-clad interface, which can serve as a tritium permeation barrier and protect against potential chemical attack from the fuel. Finally, the study attributed the formation of amorphous Al/U mixed oxide to the low temperature and limited time of neutron irradiation. APMT forms more distinct Cr and Cr-Fe intermetallic at the FeCrAl-UO 2 interface than C35M due to the higher bulk Cr:Al ratio.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2023

Novel effects of grain size and ion implantation on grain boundary segregation in ion irradiated austenitic steel

Nuclear reactor environments are extreme in nature due to the combination of exposure to corrosive coolant, mechanical stresses, and neutron irradiation damage. This combination of environments can lead to irradiation induced/assisted precipitation, segregation, and embrittlement coupled with enhanced creep and stress corrosion cracking. While nanostructured materials are of interest in a variety of applications due to their unique properties including high mechanical strength, for nuclear applications nanostructured materials have the added benefit of irradiation resistance due to the high volume density of sinks for irradiation-induced defects. In this study, we show unique radiation-induced segregation (RIS) behavior in nanocrystalline 304 stainless steel (SS304). Typical RIS in austenitic steels includes Ni enrichment and Cr depletion within the grain boundaries. In nanostructured SS304 this RIS is reduced. Furthermore, interestingly, some grain boundaries in the nanocrystalline SS304 in this study were found to be enriched in Cr after Fe 2+ ion irradiation at 500 °C up to 50 displacements per atom (dpa). To understand this unique observation, lattice-based atomic kinetic Monte Carlo simulations were performed to investigate the effects of grain size, thermal segregation, as well as injected interstitials (introduced as Fe self-ions). The results indicate that the Cr enrichment at grain boundaries may be caused by combined effects of grain size, self-ion injection and preferential Cr diffusion via the interstitial mechanism. In conclusion, the findings also suggest a possible difference in RIS between ion and neutron irradiations.

36 MATERIALS SCIENCE↗
DOE OSTI2025

Implications of point defect accumulation on UO 2 thermal conductivity and fission gas release under accelerated fuel irradiation

Evaluation of thermal properties is a crucial factor for nuclear fuel performance. During reactor operation, the accumulation of fission products and irradiation-induced lattice defects are responsible for degradation in thermal conductivity. Consequently, it affects fuel temperature and fission gas release (FGR) among other Multiphysics processes important for economics and safety analysis. We analyze the implications of point defects (PD) accumulation described using a rate theory (RT) Model on lattice thermal conductivity of UO 2 . Here, we demonstrate that fission rate-dependent point defect concentrations have the largest impact on in-pile thermal conductivity in the periphery of light water reactor fuels below a temperature threshold governed by the migration barrier of defects. Our analysis provides a mechanistic description of this phenomena which current fuel performance codes treat empirically. The reduction of thermal conductivity in the low -temperature rim region acts as additional thermal resistance and leads to a temperature notably larger than suggested by Lucuta thermal conductivity correlation. These effects are anticipated to have notable impacts when fuels are exposed to accelerated radiation. The impact of such point defect-informed treatment of thermal conductivity on fuel performance is evaluated by a detailed analysis of fission gas behavior and its release. We consider several models capturing different stages of fission gas bubble evolution and fission gas release (FGR). Finally, a new fission rate-dependent correction to the Lucuta correlation is proposed. The results show a significant reduction in thermal conductivity at the fuels’ periphery and an increase in fuel centerline temperature specifically at low burnups. Ultimately a modified LC shows a higher FGR compared to the original LC, while the acceleration process results in a reduction in overall FGR.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Orientation-selected micro-pillar compression of additively manufactured 316L stainless steels: Comparison of as-manufactured, annealed, and proton-irradiated variants

In this work, irradiation response and deformation mechanisms of additively manufactured (AM) 316L stainless steel were studied by atomic scale characterization and micro-pillar compression. The AM 316L stainless steels were fabricated by direct energy deposition, a laser-based additive manufacturing process. Irradiation with 2 MeV protons at 360°C was performed to create ~1.8 displacements-per-atom (dpa) damage in AM 316L. Deformation behaviors of the as-manufactured, annealed, and proton-irradiated variants were studied, focusing on the effects of manufacturing-induced pores, residual stress, and irradiation-introduced defects (dislocation loops and voids). Micro-pillars were prepared from grains of pre-selected orientation, avoiding contributions of grain boundaries and allowing determination of resolved shear stress on {111} glide planes. Transmission electron microscopy was used to characterize the pre- and post-deformation microstructure. It was found that in the as-manufactured alloy variant, moving dislocations were the major deformation carrier, with noticeable blocking by fabrication-induced pores, In the annealed variant, hardness was reduced, and deformation was also accomplished by dislocation gliding. In the proton-irradiated variant, significant twinning was observed. Comparing measured resolved shear stress and predicted critical stress for dislocation dissociation, we conclude that irradiation hardening became high enough to activate twinning. Therefore, the deformation mechanism changes from dislocation gliding to twinning. The study is important for both processing optimization and performance evaluation of AM alloys for reactor applications.

36 MATERIALS SCIENCE↗
DOE OSTI2023

Role of structural defects in mediating disordering processes at irradiated epitaxial Fe 3 O 4 / Cr 2 O 3 interfaces

Disordering processes in oxide materials are complicated by the presence of interfaces, which can serve as either point defect sinks or accumulation sites; the response depends on factors such as interfacial structure, chemistry, and termination. Here, we have characterized the disordering of epitaxial Fe 3 O 4 (111) / Cr 2 O 3 (0001) thin film heterostructures after 400 keV Ar 2+ radiation at room temperature. The density of misfit dislocations in both the Fe 3 O 4 overlayer and Cr 2 O 3 buffer layer is varied by changing the thickness of Cr 2 O 3 to be pseudomorphically strained to the Al 2 O 3 (0001) substrate (5 nm thick) or partially relaxed (20 nm thick), as confirmed by Bragg filtering analysis of scanning transmission electron microscopy images. In both cases, irradiation leads to damage accumulation on the Fe 3 O 4 side of the heterointerface, as shown by Rutherford backscattering spectrometry measurements in the channeling geometry. However, the interface with more misfit dislocations exhibits disordering at a faster rate than the less-defective interface. Likewise, layer-resolved electron energy loss spectroscopy reveals interfacial reduction of Fe after irradiation at the more defective interface. Intermixing of Cr across the interface is observed by atom probe tomography, which is likely facilitated by the generation of Cr interstitials in Cr 2 O 3 under irradiation.

36 MATERIALS SCIENCE↗
DOE OSTI2025

Superior plastic flow stability of self-patterned carbide – amorphous ceramic nanostructures

Amorphous ceramics and carbides exhibit superb strength but poor plasticity. Here, we synthesized TiC-SiOC nanostructures with TiC-nanocarbides embedded in amorphous ceramic SiOC by co-sputtering followed by high-temperature annealing and/or irradiation. TiC-SiOC nanostructures exhibit high strength and good plastic flow stability even after heavy irradiation, 7 GPa at room temperature and 3.6 GPa at 700 ℃ with a uniform strain of about 10%∼18%. The uniform deformation is accommodated by the shearing of amorphous ceramic and the rotation of nanocarbides. Nanocarbides inhibit the propagation of shear banding in amorphous SiOC, and amorphous-crystal interfaces act as sinks to manage irradiation-induced defects.

36 MATERIALS SCIENCE↗
DOE OSTI2025

Anisotropic growth of Ni 2 (Cr,Mo) ordered phase in proton irradiated Ni-Cr-Mo alloys

Ni-Cr-Mo alloys are widely used in the nuclear industry as structural materials due to their high temperature strength and corrosion resistance. Ni-based alloys containing around 33 at.% (Cr+Mo) developed a long-range ordered Ni 2 (Cr,Mo) phase after thermal aging and/or irradiation. The ordering mechanism for thermally-aged Ni 2 (Cr,Mo) phase is well-understood, characterized to be sluggish, homogeneous, and isotropic. The ordering mechanism for irradiation-induced Ni 2 (Cr,Mo) phase is not fully understood, characterized as having rapid formation and demonstrating anisotropic precipitation. Here, this work elucidates the anisotropic precipitation and anisotropic precipitation mechanism of Ni 2 (Cr,Mo) after proton irradiation in Ni-Cr-Mo alloys. Selected area electron diffraction and bright-field scanning transmission electron microscopy imaging are used to image superlattice reflections from the ordered phase and irradiation-induced defects, respectively. A higher degree of anisotropic precipitation is observed with increasing dislocation loop and void size; a phenomenon not observed in thermally aged samples.

36 - MATERIALS SCIENCE↗
DOE OSTI2023

Probing Defects and Spin‐Phonon Coupling in CrSBr via Resonant Raman Scattering

Abstract Understanding the stability limitations and defect formation mechanisms in 2D magnets is essential for their utilization in spintronic and memory technologies. Here, defects in mono‐ to multilayer CrSBr are correlated with structural, vibrational, and magnetic properties. Resonant Raman scattering is used to reveal distinct vibrational defect signatures. In pristine CrSBr, it is shown that bromine atoms mediate vibrational interlayer coupling, allowing for distinguishing between surface and bulk defect modes. Environmental exposure is shown to cause drastic degradation in monolayers, with the formation of intralayer defects. This is in contrast to multilayers that predominantly show bromine surface defects. Through deliberate ion irradiation, the formation of defect modes is tuned: these are strongly polarized and resonantly enhanced, reflecting the quasi‐‐1D electronic character of CrSBr. Strikingly, pronounced signatures of spin‐phonon coupling of the intrinsic phonon modes and the ion beam‐induced defect modes are observed throughout the magnetic transition temperature. Overall, defect engineering of magnetic properties is possible, with resonant Raman spectroscopy serving as a direct fingerprint of magnetic phases and defects in CrSBr.

2D magnetic semiconductors↗
DOE OSTI2022

In-situ irradiation-induced studies of grain growth kinetics of nanocrystalline UO 2

The thermal conductivity of UO 2 fuel needs to be high enough to dissipate the heat generated from the fission reaction. Since grain size affects thermal conductivity and grain size can evolve with irradiation, it is critical to understand in-reactor UO 2 grain growth. Most studies of grain growth in UO 2 are based on thermally driven processes at elevated temperatures. However, studies have shown that grain growth can occur even at cryogenic temperatures by ballistic processes. Such irradiation-induced grain growth in UO 2 is yet to be studied. Advanced in-situ Kr ion irradiation and transmission electron microscopy were systematically performed on nanocrystalline UO 2 thin films at temperatures ranging from 50 K to 1073 K; grain growth was observed at all temperatures. A combination of manual and machine learning techniques was used to measure and plot grain size evolution against irradiation fluence at various irradiation temperatures. The machine learning method has significantly improved the analysis efficiency and reduced human labors. The grain diameter data were fitted using classical grain growth and thermal spike models to describe grain growth kinetics with and without irradiation effect. Grain growth during low temperature irradiation (≤ 475 K) can be well described by the thermal spike model. Above 475 K, there were additional thermally assisted processes that further accelerate the grain growth. At the highest irradiation temperature about 1075 K, both irradiation-induced dislocation loops and cavities/bubbles were observed to form in the UO 2 . In this report, the effects of irradiation-induced defects on grain growth kinetics are discussed.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Dislocation channel broadening–A new mechanism to improve irradiation-assisted stress corrosion cracking resistance of additively manufactured 316 L stainless steel

Additively manufactured (AM) 316 L stainless steel (SS) after hot isostatic pressing (HIP) was found to exhibit superior resistance to irradiation-assisted stress corrosion cracking (IASCC) in high-temperature water, as compared to wrought 316 L SS. The well-accepted IASCC factors of radiation-induced segregation (RIS) and radiation hardening are not accurate descriptions of IASCC susceptibility in this case. A decreased strain localization along grain boundaries, caused by dislocation channel broadening, was confirmed to suppress crack initiation. A unique distribution of irradiation-induced defects in HIP AM 316 L SS eased dislocation cross-slip compared to those in the wrought counterpart, thus increasing the channel width near the grain boundaries. Here for the first time, this study highlights the importance of dislocation channel broadening as a potential mechanism to further improve the IASCC resistance of 316 L SS and provides direct experimental evidence based on commercial-grade materials.

316L stainless steel↗