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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 271 records · Page 15
All sourcesNASA NTRSDOE OSTI
DOE OSTI2021

Key properties of mixed cerium and zirconium microspheres prepared by the internal gelation process with previously boiled HMTA and urea

The internal gelation process using previously boiled hexamethylenetetramine-urea was used in exploratory study to produce CeO 2 –ZrO 2 microspheres, which can serve as a simulant for ceramic transuranic fuel particles and as a viable three-way catalyst. The calcined CeO 2 and ZrO 2 microspheres with Barrett–Joyner–Halenda (BJH) pore size and volumes of 8–10 nm and 0.19–0.20 mL/g, respectively, had many more surface cracks than their relatively smooth CeO 2 –ZrO 2 counterparts even though the BJH pore sizes and volumes of mixed oxide spheres were lower at 3 nm and 0.11–0.12 mL/g, respectively. The Brunauer–Emmett–Teller (BET) surface areas for the CeO 2 and ZrO 2 microspheres were 74 and 93 m2/g, respectively, and the BET surface areas for the mixed oxides were much greater at 155–158 m 2 /g, which should improve catalytic performance. Overall, the catalytic activity of each microsphere composition was confirmed through the oxidation of carbon monoxide.

36 MATERIALS SCIENCE↗
DOE OSTI2025

Effect of Sn microalloying on the nucleation of L1 2 Al 3 Zr precipitates in a dilute aluminum-zirconium alloy

While L1 2 -Al 3 Zr nanoprecipitates provide a balance between strengthening and good electrical conductivity, the precipitation of L1 2 -Al 3 Zr in aluminum requires aggressive heat treatments. An improved age-hardening response was observed during isochronal aging of an Al-0.24Zr (wt%) alloy when microalloyed with Sn. A new mechanism termed Low melting point Element-Assisted Nucleation (LEAN) is proposed to explain the lower temperature nucleation of L1 2 -Al 3 Zr precipitates observed in this alloy based on the addition of a low melting point element, such as Sn. Characterization verified the first-principles density functional theory prediction that Zr and Sn atoms cluster during homogenization owing to the favorable binding energy of Zr-Sn-vacancy triplets. Direct microstructural observations revealed these clusters form Sn nanoprecipitates that assist the nucleation of L1 2 -Al 3 Zr at 200°C, where L1 2 -Al 3 Zr precipitation is not expected due to the low diffusivity of Zr atoms in Al. At higher temperatures (≳350°C), the acceleration of L1 2 -Al 3 Zr precipitation is driven by faster Zr diffusion in Al with Sn microalloying and the nuclei formed via the LEAN mechanism. In conclusion, this combination of mechanisms explains the improvement in age hardening through L1 2 -Al 3 Zr precipitation with Sn microalloying.

36 MATERIALS SCIENCE↗
DOE OSTI2020

Efficient SF 6 /N 2 separation at high pressures using a zirconium-based mesoporous metal–organic framework

Adsorptive separation of SF 6 /N 2 mixtures is an important issue since SF 6 is a significant greenhouse gas. Although separation performances should be evaluated under high pressure conditions to find an efficient adsorbent candidate for pressure swing adsorption (PSA) processes, no studies on metal–organic frameworks (MOFs) have focused on SF 6 /N 2 separation at high pressures above 1 bar. In this study, we evaluated the potential of three hydrothermally stable adsorbents, including two MOFs (UiO-66 and UiO-67) and zeolite-13X, for SF 6 /N 2 separation under normal PSA operation pressures (~10 bar). Interestingly, UiO-67 with a high surface area and large pore size exhibited very high SF6/N2 selectivity (30~37) at 10 bar as well as a significantly large SF6 working capacity (5.94 mmol g-1) between 10 bar and 1 bar. Moreover, UiO-67 showed facile regeneration at the ambient temperature as well as good cyclic adsorption and desorption behavior during 20 cycles. These results show that UiO-67 is a potential adsorbent for adsorptive SF 6 /N 2 separation.

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

Irradiation creep measurement and microstructural analysis of chromium nitride–coated zirconium alloy using pressurized tubes

Environmental barrier coatings for Zr-based materials are currently under development to reduce oxidation and embrittlement in light-water reactors. Chromium nitride is one such candidate for this application, particularly as accident-tolerant fuel cladding. However, quantifying the impact of coatings on the irradiation-induced creep of zircaloy (Zry) is critical as this mechanism often exceeds thermal creep rates under light-water reactor operating conditions and can be a limiting design characteristic. Additionally, examining irradiation effects in the microstructure at the coating interface is key to understanding the compatibility of the material system. Here, to accelerate the experimental measurement of irradiation creep and microstructure evolution in CrN-Zry, compact, pressurized creep tubes were fabricated and irradiated in the High Flux Isotope Reactor (HFIR) at Oak Ridge National Laboratory. Miniature, thin-walled rodlets fabricated from annealed Zr-Sn barstock were coated with CrN using physical vapor deposition (PVD) to nominal thicknesses of 4 and 8 μm. Coated and uncoated rodlet specimens were internally pressurized and welded, generating nominal circumferential hoop stresses of 0, 90, or 180 MPa under 300°C irradiation conditions. Twelve specimens were measured diametrically prior to irradiation using a low-cost, automated, contactless laser profilometer developed for this work. Specimens were irradiated in sealed capsules for one 25-day HFIR cycle, accumulating approximately 1.8 $\times$ 10 21 n/cm 2 fast fluence (E n > 10 MeV). The irradiated samples were retrieved and remeasured using the same profilometry system in a shielded hot cell facility. Irradiation creep between specimens was compared using standard statistical tests and showed that both thicknesses of CrN coating had a negligible effect on the irradiation creep strain of the Zry material. Microstructure characterization of pre- and post-irradiated CrN-Zry specimens showed minimal changes due to irradiation but did show a substantial O-rich region at the Zry-CrN interface.

36 MATERIALS SCIENCE↗
DOE OSTI2024

First-principles investigation of lanthanides diffusion in HCP zirconium via vacancy-mediated transport

The diffusion of lanthanide fission products plays an important role in the growth of the fuel-cladding chemical interaction (FCCI) region in metallic fuels. The use of a Zr interdiffusion barrier may mitigate the transport of lanthanides from the fuel to the cladding, but the efficacy of such a liner is not yet known. In this paper, the stability and vacancy-mediated diffusion of La, Ce, Pr, and Nd in hexagonal close-packed (HCP) Zr is investigated via density functional theory (DFT) calculations and self-consistent mean field (SCMF) analysis. DFT is used to calculate the formation, binding, and migration energies of vacancies and vacancy-solute pairs. The DFT energetics are used in the KineCluE code to calculate the Onsager transport coefficients. La is found to be the fastest diffusing species in HCP Zr and experiences an almost isotropic diffusion behavior. The other three species (Ce, Pr, and Nd) demonstrate anisotropic diffusion where the diffusion in the basal planes is significantly faster than that along the c-axis. Further, the calculated lanthanide diffusivities in HCP Zr are fitted to an Arrhenius relation and the activation energies and prefactors are reported for the first time. Furthermore, the vacancy drag and the segregation tendencies were analyzed using the calculated off-diagonal transport coefficients. According to our vacancy-mediated diffusion model, lanthanides are expected to be enriched at vacancy sinks at low temperatures, while at high temperatures, lanthanides are depleted at sinks and will preferably diffuse into the bulk. The enrichment/depletion transition temperature depends on the diffusion direction (basal or axial) and hence will be controlled by the grain texture and orientation.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗