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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 OSTI2022

Thermal Neutron Scattering Law Evaluations for Zirconium Carbide and Critical Mass Calculations [Slides]

Zirconium Carbide (ZrC) is a refractory material currently being investigated for high temperature reactor applications. ZrC was studied as a base of (U,Zr)C solid solution and UC 2 composite fuels for NERVA-type cores during the Rover program. Testing of the material was performed at Nuclear Furnace 1 (NF-1) located at Los Alamos. Thermal Scattering Law (TSL) evaluations generated with FLASSH are a new contribution to ENDF/B-VIII.1.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2023

New Cross Section Measurements and Evaluations of Zirconium Isotopes [Slides]

This presentation gives a look to new cross section measurements and evaluations of Zirconium (Zr) isotopes. This lecture includes highlights of accomplishments made in 2022. Additionally, it explains the motivation for investigation, new measurements, and re-evaluation of the Zr isotopes. This presentation concludes with a preliminary evaluation result from RPI 90 Zr evaluation and preliminary results from Eigenvalue sensitivity study on Zr sensitive critical benchmark experiments.

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2024

Investigation of Degradation Mechanisms of Cr-coated Zirconium Alloy Cladding in Reactivity-Initiated Accident

Reactivity-initiated accident testing has been performed on two baseline fuel rodlets which were fabricated from Zr-4 cladding and utilized fresh UO 2 fuel pellets. These experiments are part of a larger series of transient irradiations to explore Cr-coated zirconium alloy cladding under Reactivity-initiated accident conditions to evaluate ballooning and burst and sever oxidation characteristic of late-stage failure. The first two irradiations have been completed and post-transient examinations have been performed. In one rod, ballooning and burst occurred, along with oxidation. The second rod leaked and as a result did not balloon as much during the transient. This second rod also experienced increased oxidation compared to the first due to higher temperatures.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Comparison of zirconium redistribution in BISON EBR-II models using FIPD and IMIS databases with experimental Post Irradiation Examination

The BISON code framework is utilized to model and analyze zirconium redistribution within UPuZr metallic fuels. The simulation results are compared against past experimental metallography data to validate the BISON code models. Various BISON models are implemented and compared to properly reflect the Zr present in past data.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2026

Electrode and Microstructure Dependence of Oxygen Diffusion in Ferroelectric Hafnium Zirconium Oxide Thin Films

Hafnia-based ferroelectrics hold promise to reduce energy demand for computing by enabling compute-in-memory and as non-volatile memories. The ferroelectric phase in this material system is, in part, stabilized by oxygen vacancies. While oxygen vacancies may be a necessity for phase stability, they limit device endurance through diffusion and accumulation into conducting channels. Herein, it is shown that oxygen diffusion is spatially variable within individual grains of ferroelectric hafnium zirconium oxide (HZO). Using 18 O tracers and finite difference modeling, it is shown that grain boundaries and regions near electrode interfaces allow for relatively rapid oxygen diffusion, with values as much as 10 4 larger than the grain cores. Further, the selection of electrode material affects the diffusion coefficients across all microstructural regions. HZO films in contact with TiN electrodes result in more oxygen-deficient HZO films and higher oxygen diffusion coefficients. Tungsten electrodes result in fewer vacancies and lower diffusion coefficients. Diffusion activation energy differences between the HZO with the two electrodes is reconciled by differing populations of charged and uncharged oxygen vacancies. This insight into the local vacancy populations and diffusion pathways provides a platform for designing hafnia-based films, deposition processes, and integration strategies to reduce vacancy gradients and improve performance.

36 MATERIALS SCIENCE↗
DOE OSTI2024

Zirconium‐Based Metal–Organic Frameworks with Free Hydroxy Groups for Enhanced Perfluorooctanoic Acid Uptake in Water

Perfluorooctanoic acid (PFOA) is a highly recalcitrant organic pollutant, and its bioaccumulation severely endangers human health. While various methods are developed for PFOA removal, the targeted design of adsorbents with high efficiency and reusability remains largely unexplored. Here, in this work, the rational design and synthesis of two novel zirconium‐based metal‒organic frameworks (MOFs) bearing free ortho ‐hydroxy sites, namely noninterpenetrated PCN‐1001 and twofold interpenetrated PCN‐1002, are presented. Single crystal analysis of the pure ligand reveals that intramolecular hydrogen bonding plays a pivotal role in directing the formation of MOFs with free hydroxy groups. Furthermore, the transformation from PCN‐1001 to PCN‐1002 is realized. Compared to PCN‐1001, PCN‐1002 displays higher chemical stability due to interpenetration, thereby demonstrating an exceptional PFOA adsorption capacity of up to 632 mg g −1 (1.53 mmol g −1 ), which is comparable to the reported record values. Moreover, PCN‐1002 shows rapid kinetics, high selectivity, and long‐life cycles in PFOA removal tests. Solid‐state nuclear magnetic resonance results and density functional theory calculations reveal that multiple hydrogen bonds between the free ortho ‐hydroxy sites and PFOA, along with Lewis acid‐base interaction, work collaboratively to enhance PFOA adsorption.

36 MATERIALS SCIENCE↗
DOE OSTI2022

Catalytic Degradation of Polyethylene Terephthalate Using a Phase‐Transitional Zirconium‐Based Metal–Organic Framework

Abstract Polyethylene terephthalate (PET) is utilized as one of the most popular consumer plastics worldwide, but difficulties associated with recycling PET have generated a severe environmental crisis with most PET ending its lifecycle in landfills. We report that zirconium‐based metal–organic framework (Zr‐MOF) UiO‐66 deconstructs waste PET into the building blocks terephthalic acid (TA) and mono‐methyl terephthalate (MMT) within 24 hours at 260 °C (total yield of 98 % under 1 atm H 2 and 81 % under 1 atm Ar). Extensive structural characterization studies reveal that during the degradation process, UiO‐66 undergoes an intriguing transformation into MIL‐140A, which is another Zr‐MOF that shows good catalytic activity toward PET degradation under similar reaction conditions. These results illustrate the diversity of applications for Zr‐MOFs and establish MOFs as a new class of polymer degradation catalysts with the potential to address long‐standing challenges associated with plastic waste.

Wu, Yufang↗
DOE OSTI2022

Catalytic Degradation of Polyethylene Terephthalate Using a Phase-Transitional Zirconium-Based Metal–Organic Framework

Polyethylene terephthalate (PET) is utilized as one of the most popular consumer plastics worldwide, but difficulties associated with recycling PET have generated a severe environmental crisis with most PET ending its lifecycle in landfills. We report that zirconium-based metal–organic framework (Zr-MOF) UiO-66 deconstructs waste PET into the building blocks terephthalic acid (TA) and mono-methyl terephthalate (MMT) within 24 hours at 260 °C (total yield of 98 % under 1 atm H 2 and 81 % under 1 atm Ar). Extensive structural characterization studies reveal that during the degradation process, UiO-66 undergoes an intriguing transformation into MIL-140A, which is another Zr-MOF that shows good catalytic activity toward PET degradation under similar reaction conditions. Here these results illustrate the diversity of applications for Zr-MOFs and establish MOFs as a new class of polymer degradation catalysts with the potential to address long-standing challenges associated with plastic waste.

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

Phosphate‐functionalized Zirconium Metal–Organic Frameworks for Enhancing Lithium–Sulfur Battery Cycling

Abstract Lithium–sulfur batteries are promising candidates for next‐generation energy storage devices due to their outstanding theoretical energy density. However, they suffer from low sulfur utilization and poor cyclability, greatly limiting their practical implementation. Herein, we adopted a phosphate‐functionalized zirconium metal–organic framework (Zr‐MOF) as a sulfur host. With their porous structure, remarkable electrochemical stability, and synthetic versatility, Zr‐MOFs present great potential in preventing soluble polysulfides from leaching. Phosphate groups were introduced to the framework post‐synthetically since they have shown a strong affinity towards lithium polysulfides and an ability to facilitate Li ion transport. The successful incorporation of phosphate in MOF‐808 was demonstrated by a series of techniques including infrared spectroscopy, solid‐state nuclear magnetic resonance spectroscopy, and X‐ray pair distribution function analysis. When employed in batteries, phosphate‐functionalized Zr‐MOF (MOF‐808‐PO4) exhibits significantly enhanced sulfur utilization and ion diffusion compared to the parent framework, leading to higher capacity and rate capability. The improved capacity retention and inhibited self‐discharge rate also demonstrate effective polysulfide encapsulation utilizing MOF‐808‐PO4. Furthermore, we explored their potential towards high‐density batteries by examining the cycling performance at various sulfur loadings. Our approach to correlate structure with function using hybrid inorganic–organic materials offers new chemical design strategies for advancing battery materials.

Liu, Bingqian↗
DOE OSTI2023

Grain size dependence of thermally induced oxidation in zirconium carbide

Here complementary analytical approaches were employed to probe the effect of grain size on thermally induced oxidation of zirconium carbide (ZrC) utilizing thermogravimetric analysis, differential scanning calorimetry, and Raman spectroscopy, as well as synchrotron-based and laboratory-based X-ray diffraction (XRD) experiments. The oxidation mechanism and phase behavior of nanocrystalline ZrC (grain size ~ 20 nm) were compared with that of the more documented microcrystalline ZrC (grain size ~ 1 µm). Synchrotron XRD at the Advanced Photon Source with a hydrothermal diamond anvil cell (HDAC) used as a sample chamber revealed that the onset of oxidation is at ~ 380 °C for microcrystalline ZrC which is in agreement with previous work. In contrast, the critical oxidation temperature was ~ 330 °C for nanocrystalline ZrC. Additional high-temperature synchrotron XRD experiments at the National Synchrotron Light Source II using a lamp furnace in combination with Raman analysis showed that tetragonal ZrO 2 forms as an initial oxidation product and transforms at higher temperatures to the monoclinic phase. Thermogravimetric analysis (TGA) coupled with differential scanning calorimetry (DSC) confirmed the X-ray results of a lower critical oxidation temperature for the nanocrystalline sample. The phase transformations in the oxide phase with associated critical temperatures were also evident in the thermodynamic data as exothermic heat events.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Brazing characteristics, microstructure, and wettability of laser powder bed fusion additive manufactured GRCop-84 compared to CuCrZr and OFC, and brazing to titanium-zirconium-molybdenum alloy limiters

Laser Powder Bed Fusion (L-PBF) of Glenn Research Copper 84 (GRCop-84), a Cr 2 Nb (8 at. % Cr, 4 at. % Nb) precipitation hardened alloy, produces a fully dense, high conductivity alloy with a yield strength of 500 MPa and ultimate tensile strength (UTS) of 740 MPa with 20% elongation; superior to other competing copper alloys. Braze wetting characteristics of GRCop-84 with Ag-Cu-X, and Au-Cu brazes were similar to CuCrZr, but less than oxygen free copper. No difference in wetting was observed between infill and surface contour areas in L-PBF GRCop-84. Wet sanding to 240 grit (R a =0.24 µm) was considered the optimal surface condition. Silver diffusing through GRCop-84 depleted Cr 2 Nb precipitates from the copper grain and deposited agglomerations of coarsened precipitates within silver-rich regions of intergranular diffusion once a density threshold was reached. Microstructure modification was minimized with 50Au-50Cu braze implying that silver caused precipitate coarsening and agglomeration, and not high temperature exposure. Coarsened precipitates were observed on the surface within braze pools implying a contribution to braze wetting. Palcusil-25, Ticusil, CuSil-ABA, and 50Au-50Cu brazes were suitable for brazing to unplated Titanium-Zirconium-Molybdenum (TZM), while sulfamate nickel plating to allows wetting with CuSil or other non-active brazes. Additionally, vacuum brazing techniques were developed to join a 1 mm thick layer of TZM to the front of additive manufactured GRCop-84 waveguides considering the brazing characteristics of both GRCop-84, TZM, and internal stress from the difference in coefficient in thermal expansion.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2025

Thermal and structural analysis of a new WEST reciprocating probe head with Titanium-Zirconium-Molybdenum armour material

The Tore Supra tokamak was transformed into an X-point divertor fusion device in the frame of the WEST (W-for tungsten-Environment in Steady-state Tokamak) project, launched in support of the ITER tungsten divertor strategy. WEST began operation in 2017. A new reciprocating probe head, equipped with a Titanium-Zirconium-Molybdenum (TZM) heat shield, was recently built and operated on WEST during the C9 Campaign in early 2024. TZM was chosen as armour material for its very good mechanical properties between 700 °C and 1400 °C. This material was already extensively used for probe heads and other plasma facing components on the Alcator C-Mod tokamak. Here, this paper presents the design of the new WEST probe head. It is equipped with tungsten filaments that can be heated to emit electrons and provide a direct measurement of the plasma potential in the scrape-off layer, and it is also equipped with a pair of Langmuir probes. The probe plunges vertically into the plasma, reaches its maximum dive depth, then returns to its protected resting position in about 0.25 s. Thermal loads are calculated using as input the thermal properties of TZM and radial profiles of the heat flux measured in the past by other reciprocating probes, and scaled upwards to the expected heat loads in the WEST scrape-off layer. An analytical estimation of Electromagnetic (EM) loads will also be presented, as well as a thermal and structural analysis of the probe for 15 cycles at maximum plunge depth. The probe is observed by an infrared camera. Comparisons between the heat flux deduced by the Langmuir probe and the infrared radiance of the probe housing during plasma exposure are finally presented.

Diagnostic↗
DOE OSTI2021

Microstructural characterization of as-fabricated monolithic plates with boron carbide, aluminum boride, and zirconium boride burnable absorbers

The use of burnable absorbers can be beneficial for nuclear reactors by extending the fuel’s operational cycle, providing additional criticality control, and flattening the power profile. In this work, three burnable absorber materials (boron carbide, aluminum boride, and zirconium boride) embedded in aluminum have been fabricated into foils and clad in AA-6061 for potential use in high performance research reactors. The as-fabricated boron-containing phases were determined using transmission electron microscopy to be AlB2, B4C, and ZrB2. TEM also revealed weak bonding at the B4C-matrix interface. SEM showed a relatively uniform spatial distribution of boron-containing phases for all the candidate materials. Higher porosity was observed in the foil containing ZrB2 in its as-rolled condition. The porosity in the ZrB2 foil was reduced by hot isostatic pressing. The size and shape distributions of the boron-containing phases were analyzed on the criteria of cross-sectional area, perimeter, roundness, circularity, and aspect ratio. A method of converting the 2D burnable absorber dispersoids seen in cross-sectional microscopy images into 3D volumes was derived using both spherical and ellipsoidal geometry models. The difference in calculated burnable absorber dispersoid average volume between the two models ranges from 20% to 100%, which could impact burnable absorber burnout rates due to differences in neutron self-shielding.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

High-temperature oxidation and quenching of chromium-coated zirconium alloy ATF cladding tubes with and w/o pre-damage

Chromium-coated zirconium alloys are one of the promising candidates for accident-tolerant fuel cladding (ATF) tubes for light water reactors (LWRs). In this study, the high temperature oxidation and degradation of two types of Cr coatings (cold spray and physical vapor deposition) with and without pre-damage by scratches were investigated on prototype rod segment samples filled with ZrO 2 pellets and tightly sealed with welded end caps. Isothermal tests at 1100 and 1200 °C were terminated by quenching with water; transient tests were performed up to 1500–1600 °C until complete coating failure. The positive effect of both types of Cr coatings was observed in all tests. Pre-damaged specimens showed only locally increased oxidation of the scratched zone, but no negative effects on the adjacent Cr coating. Finally, the behavior of the two types of coatings is compared and the degradation mechanisms are discussed based on hydrogen release data, extensive metallographic post-test examinations, and the current state of global research.

36 MATERIALS SCIENCE↗
DOE OSTI2022

Properties and microstructure evolution of silicon nitride and zirconium nitride following Ni ion irradiation

We report that silicon nitride and zirconium nitride have been proposed as potential materials for multiple nuclear applications (inert matrix fuels, accident tolerant fuels, space nuclear power, fusion reactor design), yet knowledge on their behavior under irradiation remains limited. Ion irradiations were performed using 15 MeV Ni 5+ ions on Si 3 N 4 and ZrN samples, with midrange doses (around 3 µm) from 1 to 50 dpa and temperatures from 300 to 700°C. Volumetric lattice swelling was determined by grazing incidence X-ray diffraction, defect production and evolution were tracked using Transmission Electron Microscopy, and nanoindentation was performed to quantify the ceramics’ mechanical properties evolution. The results from these irradiation studies on nitride ceramics help fill the current gap present in the literature. Behavior consistent with past work on irradiated Si 3 N 4 was observed with respect to mechanical properties and defect formation up to 15 dpa and 500°C. Failure of the grain boundary sintering aid in Si 3 N 4 was observed above these conditions. Different behavior was observed in both nitrides at 50 dpa and 700°C, where lattice swelling increased past potential saturation values. Unreported cavity formation was witnessed in both materials under all irradiation conditions, with stable number density and slight size increase above 15 dpa. The mechanism for the cavity formation remains to be determined.

36 MATERIALS SCIENCE↗
DOE OSTI2021

Ultrasonic additive manufacturing of zirconium: Pilot results

Ultrasonic additive manufacturing (UAM) was successfully applied to the Zirconium material system to create a three-dimensional component prototype. The UAM process resulted in grain size refinement and significant localized deformation. Electron backscatter diffraction analysis revealed activated slip systems only applicable at high deformation temperatures, while texture analysis showed a decrease in measurable texture of the UAM build in comparison to the initial Zr-foil. Average tensile strengths in the X (sonotrode travel), Y (vibration), and Z (build) directions averaged 435 MPa, 458 MPa, and 359 MPa, respectively, demonstrating the viability of UAM for Zr-based materials. Furthermore, the delamination of some Z specimens along foil boundaries during loading suggests an interplay between interfacial Ti impurities introduced during welding and the spatial dependence of weld quality for this material system.

36 MATERIALS SCIENCE↗
DOE OSTI2021

Facile microwave synthesis of zirconium metal-organic framework thin films on gold and silicon and application to sensor functionalization

Zirconium-based metal-organic frameworks, including UiO-66 and related frameworks, have become the focus of considerable research in the area of chemical warfare agent (CWA) decontamination. However, little work has been reported exploring these metal-organic frameworks (MOFs) for CWA sensing applications. For many sensing approaches, the growth of high-quality thin films of the active material is required, and thin film growth methods must be compatible with complex device architectures. Several approaches to synthesize thin films of UiO-66 have been described but many of these existing methods are complex or time consuming. In this work, we describe the development of a simple and rapid microwave assisted synthesis of oriented UiO-66 thin films on unmodified silicon (Si) and gold (Au) substrates. Thin films of UiO-66 and UiO-66-NH 2 can be grown in as little as 2 min on gold substrates and 30 min on Si substrates. The film morphology and orientation are characterized and the effects of reaction time and temperature on thin film growth on Au are investigated. Both reaction time and temperature impact the overgrowth of protruding discrete crystallites in the thin film layer but, surprisingly, no strong correlation is observed between film thickness and reaction time or temperature. We also briefly describe the synthesis of Zr/Ce solid solution thin films of UiO-66 on Au and report the first synthesis of a solid solution thin film MOF. Finally, we demonstrate the utility of the microwave method for the facile functionalization of two sensor architectures, plasmonic nanohole arrays and microresonators, with UiO-66 thin films.

36 MATERIALS SCIENCE↗
DOE OSTI2024

Swift heavy ion irradiation effects in zirconium and hafnium carbides

The behavior of microcrystalline zirconium carbide (ZrC) and hafnium carbide (HfC) was studied under highly ionizing irradiation conditions at room temperature. The induced structural modifications were characterized via synchrotron-based X-ray diffraction experiments. Unit-cell expansion and buildup of microstrain were determined across a wide fluence range and linked to chemical compositions of the target material. The observed swelling resulting from irradiation with 198 MeV Xe ions in both carbide materials is characterized by two distinct mechanisms that operate within different fluence regimes. Unit-cell expansion initially proceeds by a direct-impact behavior that reaches saturation, followed at higher fluences by a second, linear swelling regime. The overall behavior, particularly the direct-impact regime, is similar for ZrC and HfC, with a more pronounced second defect accumulation process in HfC. Swelling in ZrC shows the same two distinct mechanisms upon irradiation with 198 MeV Xe ions and 946 MeV Au ions, but swelling induced by the lower-energy ions is greater across the entire fluence series. Accounting for the difference in energy deposition density between the two irradiation conditions reveals that the first swelling mechanism (direct-impact behavior) is likely related to the formation of more simple defects. In contrast, the second damaging mechanism at higher fluences (linear increase) cannot be fully explained by the induced energy density, and swelling remains somewhat higher for the low-velocity Xe irradiation. This may suggest that more complex defects and defect clusters are responsible for this swelling regime, with either their size and/or morphology modified at different energy densities.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗