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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 325 records · Page 18
All sourcesNASA NTRSDOE OSTI
DOE OSTI2018

Zirconium tetrachloride revisited

We present that zirconium tetrachloride, ZrCl 4 , is a strategic material with wide-ranging applications. Until now, only one crystallographic study on ZrCl 4 has been reported [Krebs (1970). Z. Anorg. Allg. Chem. 378 , 263–272] and that was more than 40 years ago. The compound used for the previous determination was prepared from ZrO 2 and Cl 2 –CCl 4 , and single-crystal X-ray diffraction (SCXRD) studies on ZrCl 4 obtained from Zr metal have not yet been reported. In this context, we prepared ZrCl 4 from the reaction of Zr metal and Cl 2 gas in a sealed tube and investigated its structure at 100, 150, 200, 250, and 300 K. At 300 K, the SCXRD analysis indicates that ZrCl 4 crystallizes in the orthorhombic space group Pca 2 1 [ a = 6.262 (9), b = 7.402 (11), c = 12.039 (17) Å, and V = 558.0 (14) Å 3 ] and consists of infinite zigzag chains of edge-sharing ZrCl 6 octahedra. This chain motif is similar to that observed previously in ZrCl 4 , but the structural parameters and space group differ. Finally, in the temperature range 100–300 K, no phase transformation was identified, while elongation of intra-chain Zr...Zr [3.950 (1) Å at 100 K and 3.968 (5) Å at 300 K] and inter-chain Cl...Cl [3.630 (3) Å at 100 K and 3.687 (9) Å at 300 K] distances occurred.

36 MATERIALS SCIENCE↗
DOE OSTI2019

Comparison of Different Methods of Atomistic Simulation To Calculate the Temperature of Phase Transition Using the Example of Zirconium

Comparison of various methods for calculating the phase transition temperature based on atomistic simulation such as two-phase simulation, single-phase simulation, calculations of thermodynamic potentials using phonon spectra is performed. Based on the simulation of zirconium, melting and transition between different crystalline phases are studied. It is shown that the Lindemann criterion is quite rigorous when describing the melting curve in the phase diagram. However, the threshold values of the Lindemann parameter corresponding to the phase transition may differ in different models. In addition, the results of the study allow us to estimate the accuracy of calculating the phase transition temperature based on the thermodynamic approach using the concept of entropy.

75 CONDENSED MATTER PHYSICS, SUPERCONDUCTIVITY AND↗
DOE OSTI2026

Progress and Perspectives: Zirconium Electrodeposition from Different Electrolytes

Zirconium (Zr) possesses outstanding properties, including exceptional chemical resistance and a high melting point, and is therefore desirable for use in a wide variety of challenging environments, such as in nuclear reactors and the chemical processing industry. To minimize the amount of Zr required for any given application, developing methods for generating metallic Zr coatings is highly advantageous. Owing to Zr’s highly negative reduction potential, electrodeposition in traditional solvents at near ambient temperatures remains challenging and thus not well understood. Due to the extreme conditions required to deposit this metal, extensive work has been conducted in molten salt electrolytes, however the broad applicability of this methodology is limited due to its corrosivity. The primary focus of this article is to present an overview of Zr’s electrochemical behavior and to consolidate the efforts of researchers in exploring electrodeposition techniques for Zr involving aqueous, organic, ionic liquid, deep eutectic, and molten salt solvents. With this information, we highlight trends across solvent systems and opportunities for future research.

42 ENGINEERING↗
DOE OSTI2024

In situ potassium and hydrogen ion exchange into a cubic zirconium silicate microporous material

The selective separation of ions from aqueous systems, and even in the human body, is a crucial to overall environmental management and health. Nanoporous materials are widely known for their selective removal of cations from aqueous media, and therefore have been targeted for use as a pharmaceutical to treat hyperkalemia. This study investigated the detailed crystallographic molecular mechanisms that control the potassium ion selectivity in the nanoporous cubic zirconium silicate (CZS) related materials. Using time-resolved in situ Raman spectroscopy and time-resolved in situ X-ray diffraction, the selectivity mechanisms were determined to involve a synchronous cation-cation repulsion process that serves to open a favorable coordination bonding environment for potassium, not unlike the ion selectivity filter process found in potassium ion channels in proteins. Enhancement of ion exchange was observed when the CZS material was in a partial protonated state (≈3:1 Na:H), causing an expansion of the unit-cell volume, enlargement of the 7 member-ring window, and distortion of framework polyhedra, which allowed increased accessibility to the cage structures and resulted in rapid irreversible potassium ion exchange.

59 BASIC BIOLOGICAL SCIENCES↗
DOE OSTI2016

Long-term irradiation growth In zirconium alloys - 2016-0034

This paper reviews the experimental research program on irradiation growth in zirconium alloys performed by the United Kingdom Atomic Energy Authority between 1975 and 1988. The original technological objectives of the program, key aspects of rig and test specimen design, experiments aimed at improving fundamental understanding of the growth process, and key final results on long-term growth behavior are described. Our recollections of the program are given in relation to the understanding of irradiation growth mechanisms that prevailed at the beginning and end of the research program. (authors)

36 MATERIALS SCIENCE↗
DOE OSTI2016

Effect of alloying elements, cold work, and hydrogen on the irradiation-induced growth behavior of zirconium alloy variants - 2016-0040

In-reactor dimensional changes in zirconium-based alloys result from a complex interplay of many factors, such as (1) alloy type and composition, including the addition of elements such as niobium, iron, and tin; (2) fabrication process, including cold work, texture, and residual stresses; (3) irradiation temperature; and (4) hydrogen levels. In many cases, the observed dimensional changes in light water reactor fuel-assembly components-especially at high exposures-cannot be fully explained based on current growth and creep models. Therefore, a systematic approach was taken in this multi-year (2005-2011) Nuclear Fuel Industry Research Program investigation. The objective was to measure stress-free irradiation-induced growth (IIG) of specially fabricated alloys through irradiation under controlled conditions in the BOR-60 fast-flux test reactor up to a high fluence of approximately 2X10{sup 26} m{sup -2} (E>1 MeV)-equivalent to maximum of approximately 37 dpa exposure-followed by postirradiation examinations (PIEs). Irradiation temperature was within a narrow temperature range (320±10 deg. C). The PIEs included dimensional-change and microhardness measurements, metallography and hydride etching, and scanning transmission electron microscopy (STEM) or transmission electron microscopy (TEM). All irradiation samples (typically flat rectangular coupons or curvilinear cutouts of cladding tubes sized 35 by 6.5 by 0.8 mm) were pre-filmed to avoid the uptake of impurity hydrogen from sodium-cooled BOR-60. A wide variety of samples representing standard LWR cladding alloys with and without pre-hydriding (approximately 116 to approximately 718 ppm) as well as special compositions with iron contents (100 to 4,000 ppm) were irradiated. The irradiation in BOR-60 was done in five different stages (eight micro-cycles) and lasted approximately 18 months with interim and final growth measurements made using a high precision-length measurement device. Results of the extensive investigation include: significant effects of Fe, Nb, and hydrogen additions; quantification of growth rates from low to very high fluences (dpas); measurement of volume changes; and correlation of growth with -component dislocation densities. (authors)

36 MATERIALS SCIENCE↗
DOE OSTI2016

The effect of iron on dislocation evolution in model and commercial zirconium alloys - 2016-0068

Although the evolution of irradiation-induced dislocation loops has been well correlated with irradiation-induced growth phenomena, the effect of alloying elements on this evolution remains elusive, especially at low fluences. To develop a more mechanistic understanding of the role iron has on loop formation, we used state-of-the-art techniques to study a proton-irradiated Zr-0.1Fe alloy and proton- and neutron-irradiated Zircaloy-2. The two alloys were irradiated with 2-MeV protons up to 7 dpa at 350 deg. C and Zircaloy-2 up to 14.7 X 10{sup 25} n. m{sup -2}, approximately 24 dpa, in a boiling water reactor at approximately 300 deg. C. Baseline transmission electron microscopy showed that the Zr{sub 3}Fe secondary-phase particles in the binary system were larger and fewer in number than the Zr(Fe,Cr){sub 2} and Zr{sub 2}(Fe,Ni) particles in Zircaloy-2. An analysis of the irradiated binary alloy revealed only limited dissolution of Ze{sub 3}Fe, suggesting little dispersion of iron into the matrix, while at the same time a higher -loop density was observed compared with Zircaloy-2 at equivalent proton dose levels. We also found that the redistribution of iron during irradiation led to the formation of iron nano-clusters. A delay in the onset of -loop nucleation in proton-irradiated Zircaloy-2 compared with the binary alloy was observed. The effect of iron redistributed from secondary-phase particles because of dissolution on the density and morphology of and loops is described. The implication this may have on irradiation-induced growth of zirconium fuel cladding is also discussed. (authors)

36 MATERIALS SCIENCE↗
DOE OSTI2016

Quantifying irradiation defects in zirconium alloys: a comparison between transmission electron microscopy and whole-pattern diffraction line-profile analysis - 2016-0089

Irradiation-induced dislocations significantly affect the mechanical properties of zirconium alloys, altering slip and influencing creep and growth. Thus, the quantitative characterization of irradiation defects as a function of fluence, cold work, and/or thermal treatments is important for models that attempt to predict their impact on properties. Whole-pattern diffraction line-profile analysis (DLPA) is a well-established modern tool for microstructure characterization based on first-principle physical models for dislocation density measurements in plastically deformed materials. However, applying these DLPA methods directly to irradiated materials yields higher than expected dislocation density values compared with historical transmission electron microscopy (TEM) measurements and past line-broadening analysis studies calibrated to TEM observations. In an effort to understand these differences, a new microstructural model was developed for DLPA to specifically address dislocation structures consisting of elliptical - and -component loops. To compare the refined DLPA method with TEM measurements, high-resolution neutron diffraction patterns on nonirradiated and irradiated Zr-2.5Nb samples were collected with the Neutron Powder Diffractometer instrument at the Los Alamos Neutron Science Center and were evaluated. High-resolution TEM measurements were performed at the Reactor Materials Testing Laboratory, Queen's University, for comparison with the DLPA results. The capabilities and inherent uncertainties of both the refined DLPA and TEM methods are compared and discussed in detail. We show that the differences between the density values provided by DLPA and TEM are inherent to the methods and can be reconciled with the interpretation of the data. (authors)

36 MATERIALS SCIENCE↗
DOE OSTI2016

Strain-path change tests and physically based polycrystalline modeling of the behavior of recrystallized zirconium alloys - 2016-0092

Zirconium alloys used as cladding tubes for the fuel of pressurized water reactors can undergo high applied stress during power transients. In these conditions, biaxial loading may lead to plastic deformation of the cladding. A comprehensive understanding of the material mechanical response during loading-path change tests is an important step toward predicting the behavior in these specific conditions. Using a nonstandard mechanical testing machine, we conducted loading-path change tests at 623 K on as-received recrystallized Zircaloy-4 tubes. These tests consisted of an axial tensile loading and unloading followed by an internal pressure (or pure hoop tension) loading and unloading. These tests examined the kinematic and isotropic hardening components of the strain hardening behavior of the thin cladding tubes. The isotropic hardening was attributed to dislocation multiplication and dislocation-dislocation interactions. The kinematic hardening was attributed to the interaction of the grains with each other. A polycrystalline model was also used and improved in order to simulate the tests. A good prediction of the isotropic and kinematic hardening was provided by the modeling. (authors)

36 MATERIALS SCIENCE↗
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 OSTI2015

Spark Plasma Sintering of Commercial Zirconium Carbide Powders: Densification Behavior and Mechanical Properties

Commercial zirconium carbide (ZrC) powder is consolidated by Spark Plasma Sintering (SPS). Processing temperatures range from 1650 to 2100 °C. Specimens with various density levels are obtained when performing single-die SPS at different temperatures. Besides the single-die tooling setup, a double-die tooling setup is employed to largely increase the actual applied pressure to achieve higher densification in a shorter processing time. In order to describe the densification mechanism of ZrC powder under SPS conditions, a power-law creep constitutive equation is utilized, whose coefficients are determined by the inverse regression of the obtained experimental data. The densification of the selected ZrC powder is shown to be likely associated with grain boundary sliding and dislocation glide controlled creep. Transverse rupture strength and microhardness of sintered specimens are measured to be up to 380 MPa and 24 GPa, respectively. Mechanical properties are correlated with specimens’ average grain size and relative density to elucidate the co-factor dependencies.

36 MATERIALS SCIENCE↗
DOE OSTI2016

Zirconium Carbide Produced by Spark Plasma Sintering and Hot Pressing: Densification Kinetics, Grain Growth, and Thermal Properties

Spark plasma sintering (SPS) has been employed to consolidate a micron-sized zirconium carbide (ZrC) powder. ZrC pellets with a variety of relative densities are obtained under different processing parameters. The densification kinetics of ZrC powders subjected to conventional hot pressing and SPS are comparatively studied by applying similar heating and loading profiles. Due to the lack of electric current assistance, the conventional hot pressing appears to impose lower strain rate sensitivity and higher activation energy values than those which correspond to the SPS processing. A finite element simulation is used to analyze the temperature evolution within the volume of ZrC specimens subjected to SPS. The control mechanism for grain growth during the final SPS stage is studied via a recently modified model, in which the grain growth rate dependence on porosity is incorporated. Finally, the constant pressure specific heat and thermal conductivity of the SPS-processed ZrC are determined to be higher than those reported for the hot-pressed ZrC and the benefits of applying SPS are indicated accordingly.

36 MATERIALS SCIENCE↗