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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 109 records · Page 6
All sourcesNASA NTRSDOE OSTI
DOE OSTI2021

Characterization of high thermal conductivity fuel surrogates before and after ion irradiation

High thermal conductivity nuclear fuels offer important potential advantages over traditional oxide-based fuels such as higher burnup, reduction in fission gas release, and better overall safety of the system. One proposed approach to high thermal conductivity fuels utilizes high thermal conductivity nonfissile additives with UO 2 fuel to lower the fuel operating temperature and thereby take advantage of the highly favorable radiation resistance of UO 2 at lower operating temperatures. However, differential swelling in the matrix and high conductivity additive phases during high dose irradiation could lead to internal cracking and poor performance. In the current study, ceria (CeO 2 ) and zirconia (ZrO 2 ) surrogate matrices were used to model UO2 behavior. Additives of 10 vol. % Al 2 O 3 or SiC in the form of short fibers or platelets were used for the high conductivity second phase. The nuclear fuel surrogates were sintered to achieve densities greater than 93% of the ideal values. Scanning electron microscopy (SEM) imaging and X-ray diffraction confirmed the uniform distribution of the second phase and that no intermetallic second phase was formed during sintering. The thermal conductivity of the sintered samples was measured from 50 °C to 900 °C and confirmed the desirable increase compared to pure CeO 2 /ZrO 2 pellets. Samples were irradiated with 20 MeV Ni 6+ ions at midrange doses ranging from 1 to 15 displacements per atom (dpa) and temperatures from 300 °C to 700 °C. Post irradiation characterization revealed a good stability of the samples at low to medium doses with matrix lattice parameter swelling of < 0.14 % but showed a significant microstructural deterioration and decrease of the mechanical properties at 15 dpa.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
NASA NTRSDate not supplied

Multidisciplinary Optimization of A Transonic Truss-Braced Wing Aircraft Using Aviary

The continuous push to decrease fuel burn of single-aisle commercial aircraft has led to interest in a Transonic Truss-Braced Wing (TTBW) concept vehicle. Sporting high-slung wings that are long and slender to increase aerodynamic efficiency, a TTBW can also accommodate higher-bypass turbine engines. The combination of these two changes potentially leads to an overall decrease in fuel consumption. In this paper, a TTBW concept vehicle is assembled in the Aviary open-source tool for conceptual aircraft design. The conceptual-level aerodynamics and propulsion systems that come prepackaged with Aviary are replaced with higher-fidelity vortex lattice method for aerodynamics (VSPAERO) and a one-dimensional cycle analysis tool for propulsion (pyCycle). The vehicle is then optimized to minimize fuel burn for a representative commercial mission. Design parameters for the vehicle include the electrified turbine size, the size of the electric motors which are used for takeoff and climb assist, and battery capacity. The TTBW concept vehicle presented in this paper represents the first application of Aviary to an aircraft design problem.

Optimization↗
DOE OSTI2021

An evaluation of tri-valent oxide (Cr 2 O 3 ) as a grain enlarging dopant for UO 2 nuclear fuels fabricated under reducing environment

A study was performed to evaluate the microstructure and crystallography of nominally 500–2000 Cr 2 O 3 -doped UO 2 fabricated in a temperature range of 1150–1750°C under reducing experimental conditions. We observed an increase in grain size of the samples with the increase in heat treating temperature as expected. For a given sintering temperature (1700–1750°C), an increase in the grain size was also observed with the increase in Cr 2 O 3 concentration up to a value of ~1000–1200 wppm. A decrease in fission gas release as a function of grain size was estimated for the Cr 2 O 3 -doped UO 2 samples assuming specified post-irradiation annealing conditions. A nearly linear decrease was obtained in the lattice parameter of the Cr 2 O 3 -doped UO 2 fcc phase with the increase in Cr 2 O 3 concentration, especially up to a nominal value of 1000 wppm. The lattice parameter decrease was also persistent with the increase in the average grain size as a result of addition of Cr 2 O 3 into the UO 2 lattice. An increase in the crystallite size and a decrease in the microstrain of the $fcc$ phase were observed with the increase in the average grain size of the samples, indicating a higher crystallinity of the Cr 2 O 3 -doped samples than that of the undoped UO 2 sample.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Tuning of lattice oxygen reactivity and scaling relation to construct better oxygen evolution electrocatalyst

Developing efficient and low-cost electrocatalysts for oxygen evolution reaction is crucial in realizing practical energy systems for sustainable fuel production and energy storage from renewable energy sources. However, the inherent linear scaling relation for most catalytic materials imposes a theoretical overpotential ceiling, limiting the development of efficient electrocatalysts. Herein, using modeled Na x Mn 3 O 7 materials, we report an effective strategy to construct better oxygen evolution electrocatalyst through tuning both lattice oxygen reactivity and scaling relation via alkali metal ion mediation. Specifically, the number of Na + is linked with lattice oxygen reactivity, which is determined by the number of oxygen hole in oxygen lone-pair states formed by native Mn vacancies, governing the barrier symmetry between O–H bond cleavage and O–O bond formation. On the other hand, the presence of Na + could have specific noncovalent interaction with pendant oxygen in *OOH to overcome the limitation from linear scaling relation, reducing the overpotential ceiling. Combining in situ spectroscopy-based characterization with first-principles calculations, we demonstrate that an intermediate level of Na + mediation (NaMn 3 O 7 ) exhibits the optimum oxygen evolution activity. This work provides a new rational recipe to develop highly efficient catalyst towards water oxidation or other oxidative reactions through tuning lattice oxygen reactivity and scaling relation.

36 MATERIALS SCIENCE↗
DOE OSTI2025

Additive Manufacturing of Lattice Structures for Catalyst Applications

Abstract The design and fabrication of Inconel 718 open-pore lattice structures via Laser Powder Bed Fusion (LPBF) has been investigated in this research, focusing on applications such as catalyst supports in jet fuel production. The study explores the impact of laser power and scanning speed on the geometrical resolution of these structures aiming to achieve high porosity (porosity > 60%) and specific pore sizes ranging from 500–1000 μm, intending to serve as catalyst supports, replacing conventionally manufactured foams to reduce costs. Results demonstrate the significant influence of processing parameters on the geometrical aspects of printed lattice structures, with laser power having a more pronounced effect on geometrical accuracy than scanning speed. Additionally, the mechanical properties of the printed lattice structures showed a correlation with the lattice strut sizes, as lattices with less porosity and thicker struts resulted in higher maximum shear stress.

36 MATERIALS SCIENCE↗
DOE OSTI2025

Additive Manufacturing of Lattice Structures for Catalyst Applications

Abstract The design and fabrication of Inconel 718 open-pore lattice structures via Laser Powder Bed Fusion (LPBF) has been investigated in this research, focusing on applications such as catalyst supports in jet fuel production. The study explores the impact of laser power and scanning speed on the geometrical resolution of these structures aiming to achieve high porosity (porosity > 60%) and specific pore sizes ranging from 500–1000 μm, intending to serve as catalyst supports, replacing conventionally manufactured foams to reduce costs. Results demonstrate the significant influence of processing parameters on the geometrical aspects of printed lattice structures, with laser power having a more pronounced effect on geometrical accuracy than scanning speed. Additionally, the mechanical properties of the printed lattice structures showed a correlation with the lattice strut sizes, as lattices with less porosity and thicker struts resulted in higher maximum shear stress.

Ghanadi, Nahal [Oregon State University] (ORCID:00↗
DOE OSTI2023

Ab Initio Evaluation of Plutonium Dioxide $S(α,β)$ and Thermal Neutron Cross Sections

Plutonium Dioxide (PuO 2 ) is an advanced fuel-cycle nuclear fuel material for thermal and fast reactors, with a melting point at 3261.8 K and thermal conductivity above 3 Wm -1 K -1 for temperatures below 1500 K. Plutonium from spent UO 2 is reprocessed to form chemically stable mixed oxide (MOX) fuel. In contrast to metallic fuels, the low PuO 2 conductivity results in high thermal gradients between fuel center and surface, resulting in structural annealing and efficient fission product migration over burnup. Crystal binding affects the doppler broadening of epithermal resonances for nuclear fuels and scattering behavior of thermal neutrons. Currently, there is no ENDF/B Thermal Scattering Law, i.e., $S(α,β)$, evaluation for PuO 2 , only free atom cross sections for Plutonium and Oxygen, respectively, which neglect lattice contributions to total cross section. Herein, ab initio lattice dynamics (AILD) techniques are employed to calculate the phonon density of states (DOS) using spin-orbit-coupling density functional theory (DFT) to predict the PuO2 paramagnetic ground-state structure. These results highlight the evaluation of $S(α,β)$ for Plutonium and Oxygen, respectively in PuO 2 and consequential generation of thermal neutron scattering cross sections for high fidelity criticality safety analysis and reactor calculations.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

In-pile thermal conductivity of uranium dioxide at low burnup

In-pile thermal conductivity of uranium dioxide (UO 2 ) was investigated at low burnup levels (<0.2 kWd/kg-UO 2 ) to elucidate the concurrent effects of the fission damage and the thermal recovery. In-pile experiments were performed in the Halden reactor. The fuel centerline and the cladding temperatures were measured during the experiments, and the power level of specially designed test rodlets were also monitored. The uncertainty of the pellet-cladding gap's thermal resistivity was minimized with a liquid-metal bond, while comparable measurements were made in reference test rodlets without liquid metal bonding. The experimental data were analyzed using the inverse heat transfer approach. The heat conduction equation was solved by applying the measured temperature of the cladding as a boundary condition, and the calculated fuel centerline temperature was compared with the measured fuel centerline temperature to determine the relative matrix resistivity. Results showed that the in-pile matrix resistivity, the phonon-lattice interaction terms of the thermal conductivity, was ~1.5–2.5 times higher than its unirradiated value due to fission-induced damage. Finally, the in-pile annealing caused a significant recovery of the matrix resistivity, and the amount of recovery increased from ~20 to ~33%, while the annealing temperature increased from 700 to 1,000 °C.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Out-of-pile and postirradiated examination of lanthanide and lanthanide-palladium interactions for metallic fuel

Palladium is being investigated as a fuel additive to bind with and potentially immobilize lanthanide fission products. A primary cause of fuel-cladding chemical interaction (FCCI) is the lanthanide fission products migrating to the fuel periphery and interacting with the cladding. This interaction will lead to wastage of the cladding and eventually to a cladding breach. Palladium has previously been identified as a promising additive used to prevent or decrease FCCI by reacting with the lanthanide fission products. In the current study, an alloy cast from the four highest abundant lanthanides found in irradiated metallic fuel, Nd, Ce, Pr, and La, with and without Pd, has been characterized using neutron diffraction, scanning electron microscopy, and electron probe microanalysis. In the lanthanide-Pd intermetallic compounds, all of the constituent compounds, i.e. Nd-Pd, Ce-Pd, La-Pd and Pr-Pd are known. There is very good agreement, both structurally and compositionally, between the out-of-pile lanthanide alloy and lanthanide fission products characterized in irradiated fuels. In both cases, the lanthanide elements form a solid solution in a hexagonal crystal structure. The out-of-pile lanthanide alloy follows Vegard's Law, with the measured and calculated (weighted average of constituents) lattice parameters being within 1% for both the a and c parameters. Pd bonds with the lanthanides (Ln) forming the phases LnPd and Ln7Pd3. The results indicate the properties of lanthanide compounds in irradiated metallic fuel can be reliably simulated in out-of-pile experiments.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Strain-Modulated Platinum–Palladium Nanowires for Oxygen Reduction Reaction

Electrocatalytic activity of alloy nanocatalytsts can be manipulated effectively by tuning their physical properties (ensemble, geometric, and ligand effects) to afford optimal surface structure and compositions for proton exchange membrane fuel cell (PEMFC) application. Herein, highly catalytic platinum-palladium nanowires (Pt n Pd 100-n NWs) with a subtle lattice strain and Boerdijk-Coxeter helix type morphology are synthesized through a surfactant-free, thermal single phase solvent method. X-ray diffraction results show that Pt n Pd 100-n NWs are exposed through the (111) facets and their shrinking or expanding lattice parameters can be modulated by the alloy compositions. Electrochemical results reveal that their high catalytic activity correlates with the lattice shrinking, facets, and bimetallic compositions, showing higher activity when the ratio of Pt and Pd is similar to ~78:22, which is further supported by DFT results. Compared to the nanoparticle type platinum-palladium alloyed catalysts with similar metal compositions (Pt n Pd 100-n NPs), the Pt n Pd 100-n NWs exhibit significantly improved electrocatalytic activity and stability for the oxygen reduction reaction. Our findings open new strategies to design the highly active and stable alloy nanocatalysts with controllable compositions.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2023

Highly Oxidized Oxide Surface toward Optimum Oxygen Evolution Reaction by Termination Engineering

The oxygen evolution reaction (OER) is a critical step for sustainable fuel production through electrochemistry process. Maximizing active sites of nanocatalyst with enhanced intrinsic activity, especially the activation of lattice oxygen, is gradually recognized as the primary incentive. Since the surface reconfiguration to oxyhydroxide is unavoidable for oxygen-activated transition metal oxides, developing a surface termination like oxyhydroxide in oxides is highly desirable. In this work, we demonstrate an unusual surface termination of (111)-facet Co 3 O 4 nanosheet that is exclusively containing edge-sharing octahedral Co 3+ similar to CoOOH that can perform at approximately 40 times higher current density at 1.63 V (vs RHE) than commercial RuO 2 . It is found that this surface termination has an oxidized oxygen state in contrast to standard Co-O systems, which can serve as active site independently, breaking the scaling relationship limit. Finally, this work forwards the applications of oxide electrocatalysts in the energy conversion field by surface termination engineering.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2021

Structure-property relationships and mobility optimization in sputtered La-doped BaSn O 3 films: Toward 100cm 2 V -1 s -1

he wide band gap semiconducting perovskite BaSnO 3 is of high current interest due to outstanding room temperature mobility at high electron density, fueled by potential applications in oxide, transparent, and power electronics. Due in part to a lack of lattice-matched substrates, BaSnO 3 thin films suffer from high defect densities, however, limiting electron mobility. Additionally, the vast majority of BaSnO 3 thin film research has focused on pulsed laser deposition or molecular beam epitaxy. In this work, we present an exhaustive optimization of the mobility of Ba 0.98 La 0.02 SnO 3 films grown by a scalable, high-throughput method: high-pressure-oxygen sputter deposition. Considering target synthesis conditions, substrate selection, buffer layer structure, deposition temperature, deposition rate, thickness, and postdeposition annealing conditions, and by combining high-resolution x-ray diffraction, reciprocal space mapping, rocking curve analysis, scanning transmission electron microscopy, atomic force microscopy, and temperature-dependent electronic transport measurements, detailed understanding of synthesis-structure-property relationships is attained. Optimized room temperature mobility of 96 cm 2 V –1 s –1 is achieved in vacuum-annealed GdScO 3 (110)/BaSnO 3 (120 nm)/Ba 0.98 La 0.02 SnO 3 (200 nm) heterostructures, as well as 92 cm 2 V –1 s –1 on unbuffered substrates and 87 2 V –1 s –1 without postdeposition annealing. These results, including important trends in defect densities and a surprising dependence of mobility on lattice mismatch, substantially expand the understanding of the interplay between deposition conditions, microstructure, and transport in doped BaSnO 3 films, establishing competitive mobilities in films fabricated via a scalable, high-throughput, industry-standard technique.

36 MATERIALS SCIENCE↗
DOE OSTI2022

Reactor physics benchmark of Westinghouse PWR core design suite for high burnup/high enrichment fuel - Part 2: 2D and 3D core simulations

Westinghouse has performed a comprehensive set of code-to-code benchmark comparisons to corroborate application of its state-of-the-art NEXUS-based code suite to high burnup and high enrichment (HB/HE) fueled PWR core design and reload analysis. The benchmark consisted of a progression in complexity from modeling single 2D pin cells, through 2D lattices and finally to whole core 2D and then 3D scenarios. The accompanying paper (Part 1) presented the results for a variety of pin cell and lattices representative of high energy cores and established the ability of the MPACT code to accurately match the predictions of the Monte-Carlo code SERPENT2 such that the former may be used as a surrogate reference code in spatially complex whole core modeling applications where using Monte-Carlo methods is somewhat impractical. This paper focuses on comparisons at the whole core level using a representative high power density 3-loop Westinghouse NSSS core operating for 24 month cycles employing high enriched fuel with use of combined burnable absorbers to hold down excess reactivity. The evolution of the core reactivity and power distribution characteristics throughout the cycle are calculated using the Westinghouse NEXUS-based code suite and compared to the results using MPACT. The results of this whole core benchmark confirm the conclusions that the Westinghouse code suite accurately models HB/HE fuel designs, with very good prediction of core reactivity and power distribution throughout the cycle. The industry can be confident that the NEXUS code system can be deployed to support PWR reload safety and operational analyses for cores featuring HB/HE fuel and 24-month cycle operation. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
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 OSTI2025

Defect-induced phonon-resonant scattering and its influence on thermal transport of irradiated thorium-dioxide

Thermal transport in proton irradiated thorium-dioxide (ThO 2 ) is investigated. Using a combination of experiments and first-principles computational framework, the role of lattice defects on thermal conductivity is analyzed. A resonant-phonon scattering mechanism beyond the traditionally considered Rayleigh scattering is found to significantly influence low-temperature thermal transport in the presence of irradiation-induced point defects. The existence of localized phonon modes associated with irradiation-induced defects is suggested by the inability of the first-principles based thermal conductivity model—which considers only three-phonon interactions and phonon-defects scattering using the Tamura formalism—to predict the experimental results, unless a resonant scattering mechanism is included. The emergence of additional peaks in the Raman spectra in the proximity of phonon-resonant frequency provides further evidence for the existence of localized modes. Coupled with a microstructure evolution model, this analysis enables more accurate analysis for contrasting the contributions of different phonon scattering mechanisms across all irradiation doses and temperatures.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Status Report on Design of In-situ Thermomechanical Testing at LANSCE

Nuclear fuel encounters severe thermomechanical environments in which its mechanical response is determined by its microstructure, temperature and stress level histories. Simulating the response of such microstructures is crucial for predicting both performance and transient fuel mechanical responses and experimental verification of such predictions is therefore of great interest. While most of the deformation in a nuclear fuel rod occurs in the cladding, deformation of the fuel itself is still of interest with deformation mechanisms at operating temperature and above including creep, swelling, cracking as well as pellet-clad interaction. Characterization of these properties and understanding of the underlying deformation phenomena at operating or excursion temperatures is therefore of great importance for development and ultimately licensing of improved and novel nuclear fuel forms. Diffraction techniques offer unique insight on the atomistic (e.g. crystal structure) and microstructure (e.g. phase transformations, texture, defects) length scales and have a long history of providing unique data to inform relevant deformation models that enable the required predictive capabilities. For example, dislocations lead to diffraction peak broadening that can be characterized to estimate the dislocation density and study the role of dislocations on the deformation while measuring lattice strains allows to studie load sharing in two phase materials. In this report the requirements for a sample environment for high temperature deformation of nuclear fuels are defined. The HIPPO neutron time-of-flight diffractometer at LANSCE will host this sample environment and is also described. This instrument covers diffraction angles from 140° to 40° and is also equipped with an event-mode neutron imaging detector system, enabling energy-resolved neutron imaging in parallel with the diffraction that could measure sample temperature from Doppler broadening of neutron absorption resonances or measure pore densities from changes in the attenuation. Designs of devices to characterize thermomechanical properties of nuclear fuel without diffraction are also considered to guide the design. While this report is focused on applications for nuclear fuels, the device can also characterize cladding, moderator or structural materials and therefore contribute to other fields of research and development for advanced reactors. The temperatures planned to be reached are above 2000℃, thus enabling characterization of LWR reactor fuels under accident scenarios but also reaching temperatures of fuels developed for nuclear thermal propulsion and providing opportunities to characterize those. In conjunction with the energy-resolved neutron imaging detector, this setup would allow to measure neutron cross-sections at high temperatures, filling a gap towards development of reactors operating at high temperatures.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Structure characterization and steam oxidation performance of U 3 Si 2 with Zr alloying additions

Research and development of high uranium density fuels supports the on-going effort to enhance reactor technology, safety, performance, and economics. Among high uranium density fuels, U 3 Si 2 is a favored fuel candidate due to its higher uranium density and improved thermal conductivity when compared to traditional UO 2 . While U 3 Si 2 demonstrates an improved fuel economy, it suffers from rapid degradation when exposed to water at T > 300°C. In order for U 3 Si 2 to be implemented in water cooled reactors, the water reaction must be understood and mitigated. Reported here is the performance of Zr alloying additions to U 3 Si 2 exposed to high temperature steam oxidation environments. Here, the investigation shows that homogenization heat treatment contributes to a delayed onset of oxidation and an improved homogeneity throughout the arc-melt fabricated alloyed compositions. Of the pre- and post-annealed samples, the post-annealed 7 vol% Zr addition exhibited the highest onset temperature of oxidation. A preliminary computational investigation to complement the experimental analysis assesses the solubility of Zr in U 3 Si 2 and its incorporation in the crystal lattice.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗