Search NASA⌕ Search

SEARCH · Search NASA

Results for “ZIRCONIUM ALLOY”

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.

Quote a phrase for an exact phrase match. Source license links do not imply unrestricted reuse.

187 records · Page 11
All sourcesNASA NTRSDOE OSTI
DOE OSTI2022

Draft Critical Minerals Evaluation Report: Advanced Materials and Manufacturing Technology

Future energy production by sustainable clean energy technologies will be influenced by the availability and projected supply risks of critical minerals. In this report we discuss this context from the perspective of a sustainable nuclear energy technology and the benefits from the deployment of advanced manufacturing. In the past, the demand of critical minerals in the energy sector was minor, but in recent years renewable energy systems do force a substantial increase in the requirements for critical minerals which further induces pressure on the supply of critical minerals such as chromium, cobalt, nickel, niobium, tantalum, titanium, tungsten, vanadium, and zirconium for nuclear energy technology. The demand of critical minerals for nuclear technology is, besides beryllium and hafnium, minor and less than 1 % of the world supply. This alleviates the negative impact as the nuclear industry may not be able to provide the most compelling priority for the suppliers and to other policymakers looking at the overall needs. In this draft report, detailed background information is provided on the mineralogy of the identified critical minerals, supply chain risks followed by the discussion of specific minerals applicable to material relevant to nuclear energy technology. Briefly, the impact of advanced manufacturing on savings of critical minerals and electric energy is discussed and its effect to nuclear energy technology evaluated. This report will be followed with a final report providing a strategic vision for executable actions to implement the savings through advanced manufacturing, which will be further explored in the upcoming months. The Gen-IV relevant critical minerals will also be evaluated in the context of the AMMT programs material priorities as identified in the material score cards, to determine priorities for future actions. Material savings will become even more substantial when commercial Gen-IV reactor technology becomes available because of their needs for structural materials with higher content in alloying elements for enhanced high-temperature properties and improved corrosion resistance. The requirement for critical minerals in nuclear technology (in kg/MWe) will therefore significantly increase by 2050, even though its contribution to the worldwide electric energy production capacity will remain at about 5 %. The share of nuclear on world’s electric energy production is, on grounds of high utilization and availability 10.2 % and is expected to increase to 12 % by 2050. Advanced manufacturing could aid the deployment of Gen-IV nuclear technology since critical materials savings of up to 30 % seems possible, promoting nuclear energy as a true sustainable clean energy technology together with hydroelectric power.

36 MATERIALS SCIENCE↗
DOE OSTI2023

Phase transformations and thermal expansion coefficients of unirradiated U-X wt.% Zr (X = 6, 10, 20, 30) measured via neutron diffraction

This work characterizes the crystallographic evolution of unirradiated U-X wt.% Zr (X = 6, 10, 20, 30) while cooling from equilibration, single phase γ-U-Zr, at 900 °C to ambient temperature using time-of-flight neutron diffraction. The β-U phase was unobserved during cooling at 1 °C/min in all alloys. All alloys followed the phase transformation pathway of γ-U-Zr→γ-U-Zr+α-U→α-U +δ-UZr2 with an observed miscibility gap in γ-U-Zr. The α-U and δ-UZr2 transformation took place simultaneously in the U-30 wt.% Zr sample. These findings strengthen the need to re-approach the U-Zr phase diagram in entirety. Bulk volumetric CTEs agree well with published data, strengthening the quantification of lattice-specific CTEs reported in this study. A compositionally dependent discontinuity in thermal expansion, increasing in magnitude with decreasing U content, occurs during the γ-U-Zr→α-U+δ-UZr2 transformation. The γ-U-Zr lattice parameter was measured to have a compositional dependency.

36 MATERIALS SCIENCE↗
DOE OSTI2025

Variations in GARS powder microstructure as a function of powder chemistry and particle size

The properties of metal produced through powder metallurgy depends on the feedstock used. Powders produced via gas atomization reaction synthesis (GARS) are used to produce oxide dispersion strengthened alloys. The desired powder size range can vary for each consolidation technique. However, powder microstructure also can vary with powder particle size, which in turn can impact the microstructure and properties of the consolidated parts. In this study, GARS powders are characterized via inductively coupled plasma mass spectroscopy, inert gas fusion, and high-resolution x-ray diffraction to determine variations in elemental and phase compositions. Transmission electron microscopy was used to understand microstructure variations as a function of chemistry and size. Across the three batches tested intermetallic content was 0.73–1.35 wt% in the 0-20 μm powder batch and increased to 2.46–3.80 wt% in the coarse 45-106 μm batch. Across all batches, volume percent of surface oxidation decreased with powder diameter, with volume percents within the range of 0.75–1.2 % across 10 μm powder particles, and below 0.4 % across coarse powder particles approximately 100 μm in diameter. These observations were supported by inert gas fusion measurements. However, the oxide layer was thicker in coarse powder particles due to a slower cooling rate. Increasing oxygen content in atomization gas to 2000 ppm and adding yttrium increased both the surface oxidation content and yttrium intermetallic content. Lastly, intermetallic phases within the powder coarsened with powder size. Intermetallic morphology changed from fine spherical intermetallic and columnar dendritic growth to a cellular structure with finer spherical intermetallic, to coarse irregular intermetallic and intermetallic along grain boundaries as a result of slower cooling rate and solidification rate in coarse powder particles. Furthermore, the addition of zirconium does not appear to significantly change intermetallic morphology, but the composition changed from a Y-Fe rich intermetallic to a Y-Zr-Fe intermetallic.

42 ENGINEERING↗
DOE OSTI2025

IMPACT-01 Assembly Overview [Slides]

The Irradiated Material Properties Accelerated Characterization Test (IMPACT) series of experiments will irradiate three metallic fuel alloy specimens with embedded thermal conductivity probes in the Advanced Test Reactor (ATR). Metallic fuel alloys have long been under investigation for use in advanced reactors on account of their high thermal conductivity. Metallic fuel undergoes dramatic microstructural changes early in life due to fission gas swelling until ~2-3 at% burnup when pores interconnect, thus allowing fission gas to escape into the fuel pin plenum and swelling effectively ceases. The evolution of metallic fuel thermal conductivity during this early phase has never been successfully measured in situ. This experiment will be designed to use advances in measurement sciences to characterize how thermal transport properties evolve while in reactor. The IMPACT experiment consists of three metallic fuel rodlets with thermal conductivity probes axially centered within the fuel specimen. This presentation is an overview of the IMPACT-01 assembly process with a focus MFC fuel fabrication and experiment assembly steps.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Investigation of γ - ( U , Z r ) structural properties and its interfacial properties with liquid sodium using ab initio molecular dynamics

In this study, the elastic properties, structural parameters, sound velocity, and Debye temperature of γ–(U,Zr) were computed using ab initio molecular dynamics (AIMD) at temperatures between 1000 K and 1400 K and for Zr content between 0 at. % and 100 at. %. UZr is used as a metallic fuel for Sodium Fast Reactors (SFRs). The study of the mechanical and thermal behavior of these alloys leads to a better data-informed fuel design. The bulk modulus, shear modulus, Young's modulus, and Poisson's ratio were calculated from the elastic constants and their dependence on Zr content and temperature was investigated, comparing the results with previous computational work and the available experimental data in the literature. Interfacial properties between UZr (up to 32 at. % which typically exists in nuclear fuel) and liquid sodium are also of interest due to the presence of a sodium bond between the fuel and the cladding in metallic nuclear fuel. The interfacial energy between γ–(U,Zr) and liquid sodium, the surface tension of liquid sodium, and the work of adhesion were computed at different temperatures and Zr concentrations. It was demonstrated that γ–(U,Zr) is completely wetted by liquid sodium at all the investigated temperatures and Zr concentrations. Finally, this work provides the basis for the determination of interfacial resistances in SFRs and their implementation into heat transfer fuel performance simulations, which will be the subject of future work.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Mechanical properties, strain hardening, and fracture behavior of ultrasonic additively manufactured Zircaloy-4 after low-temperature neutron irradiation

Ultrasonic additive manufacturing (UAM) is a solid-state, layer-by-layer advanced manufacturing process that has the potential to create custom spatially controlled composites with embedded wires and sensors for nuclear component manufacture. For this work, to assess the feasibility of using UAM for nuclear-relevant materials research, the technique was used to produce a 3.5-mm-thick Zircaloy-4 plate for irradiation testing. The UAM Zircaloy-4 specimens were irradiated in the High Flux Isotope Reactor at a target irradiation temperature of 117 °C to 2.9 displacements per atom (dpa) to assess differences in irradiation-hardening behavior as a function of alloy processing path. The UAM and reference baseplate (BP) materials increased in yield strength by 372±27 MPa and 346±21 MPa, respectively, and both suffered significant reductions in uniform and total elongation attributed to irradiation hardening at low-temperature. Although the materials had similar nanoscale defect structures, including nanoscale black dot/loop features and strain-induced dislocation channels, the UAM material’s processing-related defects resulted in accelerated strain localization and failure as demonstrated by lower post-irradiation uniform elongation of UAM specimens (0.5 %) compared to BP (1.5 %) material. The UAM material also showed considerable anisotropy in mechanical response due to crack propagation along weld boundaries, resulting in differences in strength & ductility when tested parallel and perpendicular to the prior UAM build orientation. Therefore, although the fundamental irradiation response of UAM-processed Zircaloy-4 was phenomenologically comparable to that of BP reference material, additional optimization of the UAM processing is needed to produce irradiation-resistant and nuclear-relevant materials.

Digital image correlation↗
DOE OSTI2021

Texture Memory in Hexagonal Metals and Its Mechanism

Texture memory is a phenomenon in which retention of initial textures occurs after a complete cycle of forward and backward transformations, and it occurs in various phase-transforming materials including cubic and hexagonal metals such as steels and Ti and Zr alloys. Texture memory is known to be caused by the phenomena called variant selection, in which some of the allowed child orientations in an orientation relationship between the parent and child phases are preferentially selected. Without such variant selection, the phase transformations would randomize preferred orientations. In this article, the methods of prediction of texture memory and mechanisms of variant selections in hexagonal metals are explored. The prediction method using harmonic expansion of orientation distribution functions with the variant selection in which the Burgers orientation relationship, {110}β//{0001} α-hex <1$\bar{1}$1>β//$\langle2\bar{11}0\rangle$ α-hex , is held with two or more adjacent parent grains at the same time, called “double Burgers orientation relation (DBOR)”, is introduced. This method is shown to be a powerful tool by which to analyze texture memory and ultimately provide predictive capabilities for texture changes during phase transformations. Variation in nucleation and growth rates on special boundaries and an extensive growth of selected variants are also described. Analysis of textures of commercially pure Ti observed in situ by pulsed neutron diffraction reveals that the texture memory in CP-Ti is indeed quite well predicted by consideration of the mechanism of DBOR. The analysis also suggests that the nucleation and growth rates on the special boundary of 90° rotation about $\langle2\bar{11}0\rangle$ α-hex should be about three times larger than those of the other special boundaries, and the selected variants should grow extensively into not only one parent grain but also other grains in α-hex(hexagonal)→β(bcc) transformation. The model calculations of texture development during two consecutive cycles of α-hex→β→α-hex transformation in CP-Ti and Zr are also shown.

36 MATERIALS SCIENCE↗