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Search indexed NASA NTRS and DOE OSTI research on propulsion, heat transfer, battery materials and energy systems. Follow report and document links to the original sources.

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At least 253 records · Page 14
All sourcesNASA NTRSDOE OSTI
DOE OSTI2022

Emulation of neutron-irradiated microstructure of austenitic 21Cr32Ni model alloy using dual-ion irradiation

Here, in this study, the capability of heavy-ion irradiation to emulate neutron irradiation was demonstrated on an austenitic 21Cr32Ni type ternary model alloy. The model alloy used in this study is chemically analogous but compositionally simpler than of alloy 800H, which is a candidate austenitic Ni alloy which has been proposed for use in Generation IV reactors. The microstructure of the 21Cr32Ni model alloy irradiated in the BOR-60 fast reactor to 17.1 dpa and 35 dpa at ~380°C was characterized using transmission electron microscopy (TEM). The 17.1 dpa BOR-60 irradiated microstructure was then compared with the microstructure of the same material developed under dual-ion (DI) irradiation using various He/dpa ratios between 0.1 and 16.6 appm He/dpa in the temperature range of 430°C-500°C. The results showed that both neutron and DI irradiation of 21Cr32Ni model alloy produced dislocations in the form of a dislocation network as well as {111}-type faulted dislocation loops, cavities, and radiation-induced Ni enrichment at radiation-induced sinks. When the dose and the He/dpa ratio were kept similar to those in neutron irradiation, DI irradiation of the 21Cr32Ni model alloy at 460°C resulted in over-nucleation of small cavities and in a high density of faulted dislocation loops compared to those observed in the fast-neutron irradiated alloy of the same heat irradiated at ~380°C. The optimal condition for reproducing the neutron-irradiated microstructure was DI irradiation at 460°C and 0.1 appm He/dpa. In that case, the faulted loop and cavity size distributions in the BOR-60 irradiated 21Cr32Ni model alloy samples closely matched with those measured in the DI irradiated 21Cr32Ni model alloy sample. The fact that the He/dpa is an order of magnitude smaller than the helium generation rate for fast neutron irradiation, stops over nucleation and allows for the development of a similar microstructure as for neutron irradiation.

21Cr32Ni model alloy↗
DOE OSTI2022

Transmission electron microscopy study of a high burnup U-10Zr metallic fuel

To support the development of U-10 wt.% Zr (U-10Zr) metallic fuel for Gen IV sodium-cooled fast reactors, we analysed a solid, Na-bonded, U-10Zr (by weight percent) fuel cross section that was irradiated to a burnup of ~ 12.4 % at.% at the Fast Flux Test Facility (FFTF). Advanced characterization techniques, including site specific sample preparation by focused ion beam (FIB) and phase/chemical determination by transmission electron microscopy (TEM), were used to reveal the constituent redistribution of Zr, characterize the fuel phases and the secondary phases (such as solid fission products) present at the end of life. It is shown that the fuel pin cross section is divided into three major concentric zones: a Zr-rich central region, a Zr-lean intermediate region, and a Zr intermediate peripherical region. The phase characterization revealed that the irradiation environment enhanced the development and stabilization of phases not predicted by the standard equilibrium U-Zr phase diagrams. Comparing the current results with the ones from previous studies, it is reaffirmed that the irradiation temperature and the time spent in the reactor, rather than the fuel burnup, are the two factors that most influence the formation of redistribution zones and their extension along the fuel cross section. Various solid fission products precipitated inside the fission gas pores, such as lanthanides, ZrRu, BaTe, CsI, and Ba and Sr oxides. Here, this study provides unprecedented nanoscale understandings in the U-10Zr fuel system that may benefit fuel performance modelling and advanced fuel development.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Irradiation Effects on Stability of δ-UZr2 phase in U-50 wt% Zr Alloy

U-50wt%Zr is a candidate metallic nuclear fuel with potential application in light water reactors due to its excellent thermal properties and high radiation tolerance. The Zr-rich UZr fuels possess greater swelling resistance and fission gas release characteristics compared with U-rich UZr fuels. In this current study, the δ-phase U-50wt%Zr is proton irradiated at room temperature to 1 displacement per atom (dpa) to provide insights on phase stability under irradiation conditions. High resolution characterization of Transmission Electron Microscopy (TEM) and Atom Probe Tomography (APT) characterization techniques are used to elucidate microstructural changes due to irradiation. TEM and APT results show highly oriented bcc β-Zr-rich platelet precipitates nucleating adjacent to α-U phases inside the UZr 2 matrix. Formation of this platelet morphology is characteristic of Widmänstatten structure which can be attributed to a variety of factors such as differences in thermal expansion coefficient between the phases, grain size, alloy composition, and cooling rate. The phases present are distinctly different than those observed through in situ annealing, but irradiation accelerates diffusion and phase separation kinetics. These microstructural changes in U-50wt%Zr are different from those achieved by pure thermodynamic or high temperature heavy ion irradiation experiments. Finally, our work, together with previous ones, highlight the necessity to study the U-Zr phase diagram under non-equilibrium thermodynamics conditions to support this material's deployment as a viable nuclear fuel form.

36 MATERIALS SCIENCE↗
DOE OSTI2023

Early self-organization of fission gas bubble superlattice formation in neutron-irradiated monolithic U-10Mo fuels

Self-organization of defect superlattices in far-from-equilibrium systems presents a promising way to mitigate swelling concerns in nuclear materials. The gas bubble superlattice (GBS) is a highly ordered, three-dimensional complex defect structure that can retain fission gasses in Uranium-Molybdenum (U-Mo) fuels. Transmission electron microscopy (TEM) investigation of monolithic U-10Mo fuel irradiated to 1.15 × 10 21 fissions/cm 3 and 1.30 × 10 21 fissions/cm 3 revealed that early-stage ordering preferentially occurs at the grain boundaries (GB) and that the critical bubble size for complete ordering is ~3 nm. Once formed at the GB, the GBS extends towards the grain interior; however, the spread in distance from the GB varies likely depending on the type and strength of the GB sink. TEM results also showed a possible correlation between the growth and evolution of the intragranular disordered bubbles and large dislocation networks. The fission product distribution in and outside of the GBS was also investigated confirming the presence of xenon in the GBS, as well as other fission products including cesium, barium, lanthanum, and cerium.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

The role of UC inclusions in the development of fission gas bubble superlattice neutron-irradiated monolithic U-10Mo fuels

Uranium Carbide (UC) inclusions are the most prevalent impurities in Uranium-Molybdenum (U-Mo) fuel and are considered undesirable because they could potentially affect fuel performance. This work revealed that, like grain boundaries (gBs), UC inclusions could help facilitate the formation of the fission gas bubble superlattice (GBS). The GBS is a highly organized complex defect structure that can effectively store fission gases, thereby inhibiting fuel swelling. Transmission electron microscopy (TEM) showed that GBS self-organization can initiate at the UC/U-Mo interfaces in U-10Mo fuel irradiated to low fission density. Furthermore, this study also revealed that the UC boundary in U-10Mo irradiated to low fission density is wavy and periodic in morphology and that GBS formation is semi-coherent with the UC boundary. Because the initiation of GBS occurs at gBs and UC/U-Mo interfaces, the fission gas inventory is always highest at those regions compared to the U-Mo grain interior. Consequently, at higher fission densities, it is likely that high burnup structure (HBS) development via grain refinement will begin at gBs and UC/U-Mo interfaces.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Flash electropolishing of BCC Fe and Fe-based alloys

The preparation of transmission electron microscopy (TEM) samples is a critical step in the characterization of materials, and the focused ion beam (FIB) technique is a commonly used method. However, a significant limitation of this technique is the FIB-induced damages on the foil surfaces, which can obscure the real features of interest, particularly in radiation effects studies. To overcome this limitation, this study presents a detailed description of the flash electropolishing technique, which can be used to remove the FIB damage from samples. The flash electropolishing technique has been successfully applied to a range of materials, including Fe, Fe-based model alloys, commercial Fe-Cr alloys, and advanced Fe-Cr alloys, both in their as-received and ion-irradiated states. Furthermore, the parameters used for Fe-Cr model alloys can be adjusted for commercial and advanced alloys with minimal modifications. Further, the study also examined the effects of electropolishing variables, such as perchloric acid concentration, electropolishing temperature, Cr concentration, and voltage. Qualitatively, a general trend and scoping test strategy is explored in our experiments. Overall, the flash electropolishing technique offers a promising solution to the challenges posed by FIB-induced damages in the preparation of TEM samples.

36 MATERIALS SCIENCE↗
DOE OSTI2023

Microstructural characterization of U-20Pu-10Zr-4Sb and U-20Pu-10Zr-4Sb-4Ln

Antimony is being investigated as a potential additive to metallic fuel to control fuel-cladding chemical interactions (FCCI). The most detrimental elements involved in FCCI are fission product lanthanides, leading to brittle intermetallics and low melting eutectic phases. Previous investigations of Sb as an additive focused on U-10Zr, in wt. %, as the fuel. The current investigation expands that to include Pu in the fuel. Here, two alloys, U-20Pu-10Zr-4Sb and U-20Pu-10Zr-4Sb-4Ln (wt. %, Ln=53Nd-25Ce-16Pr-6La) have been investigated using scanning electron microscopy (SEM) and transmission electron microscopy (TEM) to characterize the fuel as-cast microstructure and the microstructure after introduction of lanthanides. Sb reacts with Zr initially, forming Zr 2 Sb and Zr 5 Sb 3 , with as much as 20 at. % interstitial Pu present. In the presence of lanthanides, Sb forms Ln 4 Sb 3 with the lanthanides, containing ~14 at. % Pu. The Pu is substitutional for the lanthanides in the crystal lattice.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Defect-specific strength factors and superposition model for predicting strengthening of ion irradiated Fe18Cr alloy

Here, a high-purity binary alloy Fe18Cr was subjected to heavy-ion irradiation in order to provide improved understanding of the irradiation effect on radiation hardening associated with dislocation loop and network formation, α’ precipitation. The specimens were irradiated with 8 MeV Fe ions (~2 μm ion range) to midrange doses of 0.37 and 3.7 dpa at 300, 350 and 450 °C using dose rates of ~10 –5 –10 –3 dpa/s. Nanoindentation testing was performed to extract the bulk equivalent hardness at low depth from these ion irradiated specimens. High-quality transmission electron microscopy (TEM) images were acquired by utilizing a flash electropolishing method on the Focused Ion Beam (FIB) prepared liftouts. An accurate calculation of the strength factor was provided based on detailed TEM characterization of the irradiated microstructures. A newly refined hardening superposition method was applied to combine the strengthening components from each microstructure. The good agreement between microstructure predicted strength and measured strength further demonstrated the fidelity to use hardening model to quantify the mechanical properties of ion irradiated materials.

36 MATERIALS SCIENCE↗
DOE OSTI2024

Investigation of W-SiC compositionally graded films as a divertor material

W-SiC composite material is a promising plasma-facing material candidate alternative to pure W due to the low neutron activation, low impurity radiation, and low tritium diffusivity of SiC while leveraging the high erosion resistance of the W armor. Additionally, W and SiC have high thermomechanical compatibility given their similar thermal expansion rates. The present study addresses the synthesis and performance of compositionally graded W-SiC films fabricated by pulsed-DC magnetron sputtering. Compositional gradients were characterized using transmission electron microscopy (TEM) and energy-dispersive X-ray spectroscopy (EDS), and crystallographic information was obtained using electron diffraction and X-ray diffraction (XRD). Samples were exposed to L-mode deuterium plasma discharges in the DIII-D tokamak using the Divertor Material Evaluation System (DiMES). Post-mortem characterizations were performed using scanning electron microscopy (SEM) and XRD. Electron diffraction and XRD showed that the compositionally graded W-SiC films were composed of polycrystalline W and amorphous SiC with amorphous W+SiC interlayers. No macroscopic delamination or microstructural changes were observed under mild exposure conditions. Here, this study serves as a preliminary examination of W-SiC compositionally graded composites as a potential candidate divertor material in future tokamak devices.

36 MATERIALS SCIENCE↗
DOE OSTI2024

Insight into the interfacial microstructure and chemistry of hot isostatically pressed AA6061-AA6061 bonds for U-10Mo fuel cladding application

Here, in the present study, hot isostatic pressing (HIP) bonding was carried out using AA6061-AA6061 plates for U-10Mo monolithic fuel cladding application. HIP was performed at 560°C with an applied isostatic pressure of 103 MPa for 90 min. Interfacial microstructure characterizations were performed using scanning electron microscopy (SEM) combined with electron backscatter diffraction (EBSD) to quantify precipitate fractions and study recrystallization, grain growth, and orientation relationships across the interface. EBSD results confirmed that no interfacial grain growth occurred during the HIP process. Transmission electron microscopy (TEM) was also employed to quantify composition and identify phases of various precipitates and an oxide layer that formed at the interface during HIP bonding. Microstructural analyses revealed the formation of a large fraction of Mg 2 Si precipitate and a ~50 nm thick Mg-oxide layer at the interface. Further TEM analysis confirmed that the oxide layer was not continuous, and Mg 2 Al 2 O 5 oxide particles with an average size of 18 nm were present in the oxide layer. Finite element modeling analysis was performed to study the stress distribution at the interface during HIP processing.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Comparison of hardening and microstructures of ferritic/martensitic steels irradiated with fast neutrons and dual ions

Ferritic/martensitic steels T91 and HT9 were irradiated with neutrons (BOR-60 reactor) and dual ions (9 MeV Fe 3+ and 3.42 MeV energy degraded He 2+ ) from 369 to 520 °C and damage levels of 16.6 to 72 dpa to quantify the possibility of using ion irradiation to simulate neutron irradiation in terms of microstructures and mechanical properties. Nanoindentation testing was performed to obtain the bulk equivalent hardness of the dual-ion irradiated samples. For the neutron irradiated samples, both nanoindentation and Vickers hardness testing were conducted. Transmission Electron Microscopy (TEM) characterizations of the cavities, dislocation loops and precipitates were conducted to account for the strengthening contribution of each microstructure element. The good agreement between the microstructure-predicted (dispersed barrier hardening) and measured strength of the irradiated specimens demonstrated the accuracy of the strengthening model and the nanoindentation tests. Furthermore, the comparison of mechanical property and microstructure changes in ion and neutron irradiated structural materials indicated that ion irradiation replicated many neutron irradiation features. However, a single 70 °C temperature shift is insufficient to match all complex microstructures of neutron vs. ion irradiation over the irradiation temperature range of 369–520 °C.

36 MATERIALS SCIENCE↗
DOE OSTI2024

Correlating microstructure and mechanical properties of harvested high dose Zorita light water reactor internals

Here, in this study, microstructural studies and micro-mechanical testing of an ex-plant material harvested from the decommissioned pressurized water reactor (PWR) is carried out. Irradiated 304 stainless steel (SS) components were harvested by the Electric Power Research Institute, U.S. Nuclear Regulatory Commission, and members of the José Cabrera Nuclear Power Station (Zorita) in Spain. The bi-crystalline micro-tensile specimens, atom probe tomography (APT) tips, and transmission electron microscopy (TEM) lamellae were fabricated from the baffle plate materials irradiated to 0.05, 15 and 50 dpa to elucidate microstructural, microchemical, and local mechanical properties changes as a function of dose. TEM and APT studies reveal radiation induced Ni, Cr, and Si rich precipitates in the 15 dpa sample that become smaller and partially redissolve into the matrix of the 50 dpa sample (decreasing size and number density). Dislocation loop and cavity number density and size were quantified as a function of dose as well as swelling. Intergranular Ni and Si enrichments with concomitant Fe and Cr depletion were observed in the 15 dpa sample and were more pronounced in the 50 dpa sample. Micro-tensile testing performed in the scanning electron microscope (SEM) at room temperature and 300 °C shows that the material's local yield strength and ultimate tensile strength decreases with increased dose and elevated test temperature and provides further insights via localized strain mapping. Current and previous mechanical data was compared with calculated values using the dispersed barrier hardening model with inputs of dislocation loops, precipitates, and cavities from microstructural characterization.

36 MATERIALS SCIENCE↗
DOE OSTI2025

In-situ irradiation of uranium carbide

Uranium carbide (UC) is a leading candidate fuel for Generation IV reactors due to its high uranium density and thermal conductivity. However, its irradiation performance—particularly gas bubble swelling and defect dynamics—remains poorly characterized. Using in-situ transmission electron microscopy (TEM), we irradiated UC with 300 keV Xe + and 1 MeV Kr 2+ ions at temperatures up to 900 °C to quantify swelling behavior and dislocation loop evolution. The swelling remained below 0.6 % across all temperatures, suggesting the dimensional stability of UC under irradiation at these temperatures. Dislocation loops grew faster in UC than in UO 2 or UN, correlating with its lower homologous temperature. Notably, nanograin structures emerged in thin regions of the lamellar, mirroring phenomena previously observed in UO 2 and ZrC. These results address critical knowledge gaps in the radiation tolerance of UC and provide insight into its suitability for advanced reactor systems.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Microstructurally informed synchrotron x-ray analysis revealing helium defect transitions in ultrafine grained tungsten

The formation of insoluble gaseous defects in materials due to nuclear transmutation or ion implantation involves the diffusion of impurity atoms to form atomic defect clusters that coalesce into bubbles or cavities and ultimately degrade the material properties. Transmission electron microscopy (TEM) is limited in its ability to resolve sub-nanometer gas clusters whereas X-ray diffraction (XRD) provides information pertaining to local atomic changes. Here, in this study, helium (He) implanted ultrafine grained tungsten is explored through a multimodal defect characterization campaign combining TEM-informed Small Angle X-ray Scattering (SAXS) analysis, XRD lattice parameter measurements, and nanoscale He cluster quantification from a region of reciprocal space accessible via Wide Angle X-ray Scattering (WAXS). Moderate elevated temperature implantations are shown to produce high concentrations of sub-nanoscale He clusters and small, homogeneously distributed cavities, which collectively are linked to lattice expansion and further substantiated through complementary atomistic simulations. Increased implantation temperatures encourage the diffusion of these defects to the grain boundaries (GBs), leading to lattice relaxation and the growth of large GB cavities manifesting as bimodal size distributions in the SAXS analysis. Overall, our results demonstrate the utility of multimodal synchrotron X-ray analysis in bridging the gap between microscale He cavity quantification and atomic-scale defect analysis.

36 MATERIALS SCIENCE↗
DOE OSTI2025

In-situ ion irradiation induced nanograin growth in a spent UO 2 fuel

This study investigates the irradiation-driven evolution of nanograins in the early-stage restructured rim region of medium burnup spent uranium dioxide (UO 2 ) fuel. Transmission electron microscopy (TEM) lamellas prepared from Belgium Reactor 3 (BR-3) fuel were subjected to in-situ 300 keV Xe ion irradiations under varying doses, fluxes, and temperatures to evaluate their effect on the evolution of the nanograins. Our results reveal that the nanograins grow during ion irradiation. Additionally, the growth is most pronounced at elevated temperatures (300 °C), moderate at room temperature, and negligible at cryogenic temperature (−223 °C). This behavior indicates that thermal activation, alongside irradiation effects, is essential to overcome grain boundary pinning by fission gas bubbles, metallic precipitates, and porosity. Furthermore, while nanograins (<200 nm) consistently coarsened under irradiation, larger grains did not undergo further restructuring, which can be attributed to the strong defect annihilation at TEM lamella surfaces combined with the limited electronic stopping power of low energy Xe ions used in this work. These findings highlight the roles of thermal spike effects, defect mobility, and impurity pinning in governing grain evolution in the rim region of spent UO 2 fuel during ion irradiation, providing key insights for predictive models of restructuring and performance of the high burnup nuclear fuel.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Correlative electron and ion beam analysis of the electrochemical performances of LiV 3 O 8 cathode films as a function of microstructures

Degradation mechanisms in LiV 3 O 8 cathode films (~500 nm) with different initial microstructures i.e., globular nanocrystals in amorphous matrix and needle-like nanocrystals, synthesized by annealing the cathodes at 150 °C and 300 °C for 6 h are studied by Transmission Electron Microscopy (TEM) and Auger Electron Spectroscopy (AES). In addition, a unique in-situ combination of TEM and Secondary Ion Mass Spectrometry (SIMS) imaging techniques, namely Parallel Ion and Electron beam Spectrometry (PIES) is carried out to measure the 3D distribution of V and Li with nanoscale resolution. These characterization studies focus on correlating the pre- and post-cycling microstructure to observed electrochemical performance. The LiV 3 O 8 cathode film with needle-like nanocrystals (300 °C) shows a higher initial capacity but it degrades rapidly compared to the film with globular nanocrystals embedded in amorphous matrix (150 °C). We observed that LiV 3 O 8 films with needle-like nanocrystals are more susceptible to cathode decohesion and vanadium dissolution than the films with globular nanocrystals embedded in amorphous matrix, explaining the differences in their electrochemical performance. The findings from this study have relevance to the development of thin film electrode microstructures for high capacity and cyclic stability.

25 ENERGY STORAGE↗
DOE OSTI2023

Aging gracefully? Investigating iridium oxide ink's impact on microstructure, catalyst/ionomer interface, and PEMWE performance

Here, in this study, we conducted a thorough investigation of the impact of aging iridium oxide (IrO 2 ) perfluorosulfonic acid ionomer ink for up to 14 days on the properties of the ink and the resulting catalyst layers. We examined ink properties, such as zeta potential, dynamic light scattering (DLS), density, surface tension, and rheology, as functions of ink aging time. To evaluate the microstructure and catalyst/ionomer interface, we employed transmission electron microscopy (TEM), X-ray scattering, and X-ray photoelectron spectroscopy (XPS) techniques. Furthermore, we assessed the effect of ink aging on the performance of proton exchange membrane water electrolyzers (PEMWEs). Our findings reveal that most ink properties remain stable for 14 days. The variations in PEMWE cell performance are minimal, and no clear trend is observed in relation to ink aging time. This study demonstrates that the effects of aging the inks for 14 days on ink properties, catalyst layer structure, catalyst/ionomer interface, and PEMWE performance are negligible, indicating a substantial time window after ink preparation without any significant changes in its properties. These insights provide crucial guidance for the commercial production and coating processes of ink, which is necessary for scaling up PEM technologies to meet future demand.

30 DIRECT ENERGY CONVERSION↗
DOE OSTI2023

The effect of ink ball milling time on interparticle interactions and ink microstructure and their influence on crack formation in rod-coated catalyst layers

This work investigates the influence of ballmilling (sometimes also referred to as jar roller milling) time on cathode catalyst layer (CL) inks and electrode properties using formulations and coating methods relevant for industrial manufacturing. Four CL inks with the same composition were milled for 24, 48, 72, or 96 h. Rheological investigation of these inks showed a reduction of elastic moduli and steady-shear viscosity with continuous ink milling, which is correlated to a decrease in particle-particle interactions as well as formation of smaller agglomerates. Optical microscopy (OM) analysis of the fabricated electrodes revealed a trend in surface crack formation; formulations milled for 24 h contained the lowest average surface crack area percentages of 0.370% at heavy-duty loadings of ~0.300 mg Pt cm -2 , compared to 2.418% for the ink milled for 96 h. Further characterization of the CL through transmission electron microscopy (TEM) imaging showed a decrease in the mean agglomerate and pore size with milling time. Furthermore, these smaller electrode features were consistent with reduced fracture resistance and, hence, development of larger stresses during drying. Our results highlight the need to consider ink processing as an important component in defect-free CL manufacturing.

30 DIRECT ENERGY CONVERSION↗