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Results for “zirconium analysis”

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 37 records · Page 2
All sourcesNASA NTRSDOE OSTI
DOE OSTI2021

The local structure of 0.5Ba(Zr 0.2 Ti 0.8 )O 3 -0.5(Ba 0.7 Ca 0.3 )TiO 3 from neutron total scattering measurements and multi-edge X-ray absorption analysis

Neutron total scattering measurements and multi-edge X-ray absorption analysis were performed on a 0.5Ba(Zr 0.2 Ti 0.8 )O 3 -0.5(Ba 0.7 Ca 0.3 )TiO 3 powder sample. The TiIII-II edges showed that titanium is distorted in the rhombohedral [111] direction (in a BaTiO 3 -like chemical and bonding environment). Extended X-ray fine structure analysis (EXAFS) was performed on the zirconium K-edge and it was found that zirconium resides in the center of the ZrO 6 octahedra in cubic symmetry. Furthermore, the local structure was initially modelled with the orthorhombic Amm2 space group with unsatisfactory results. To better model the local structure, the results from the EXAFS analysis were incorporated into the small-box pair distribution function (PDF) refinements and the best fit of the observed data was a combination of barium calcium titanate (BCT) and barium calcium zirconate (BCZ) phases.

36 MATERIALS SCIENCE↗
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 OSTI2021

Generation of the TSL for Zirconium Hydrides from Ab Initio Methods

Zirconium hydride (ZrHx) is a moderator material used in TRIGA and other reactors that may exist in multiple phases with varying stoichiometry, which include the δ phase and the ϵ phase. Current ENDF/B-VIII.0 ZrHx thermal scattering law (TSL) evaluations do not distinguish between phases. These sub-libraries were generated with the LEAPR module of NJOY using historic phonon spectra derived from a central force model and assume incoherent elastic scattering for both bound hydrogen and zirconium, which neglects the effects of crystal structures important for scattering from zirconium bound in ZrHx. In this work, the TSLs for hydrogen and zirconium bound in δ-ZrHx and ϵ-ZrH2 were generated from phonon spectra derived from modern ab initio lattice dynamics methods and ab initio molecular dynamics. Subsequently, TSLs for hydrogen and zirconium in ZrHx and ZrH2 were generated using the Full Law Analysis Scattering System Hub (FLASSH) code. The built-in generalized coherent elastic routine was used to generate the previously neglected elastic contribution from zirconium for this material. The present TSLs provide both a re-evaluation of the current ZrH sub-libraries and expansion of the set of TSLs available for the examination of neutrons in systems with zirconium hydride, permitting explicit treatment of δ and ϵ phases.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2023

Chemical Vapor Deposition of Zirconium Compounds: A Review

Coatings of zirconium compounds are used in a wide variety of fields, yet an understanding and descriptions of deposition mechanisms are scant in the public literature. The mechanisms of deposition for metallic zirconium, ZrC, ZrN, ZrO 2 , ZrB 2 , and zirconium silicides are discussed based on the direct vapor deposition research of those compounds where possible or compared to complementary titanium systems when direct research is lacking. Both inorganic and organometallic deposition systems are discussed. As a class of compounds, an understanding of the vapor deposition mechanisms can be significantly improved by investigations on metallic zirconium deposition by zirconium halides and hydrogen and by in situ analysis techniques such as Fourier-transform infrared (FTIR) spectroscopy or x-ray photoelectron spectroscopy (XPS).

organometallic↗
DOE OSTI2025

Novel Synthesis and Characterization of Uranium-Zirconium Carbonitride by Direct Casting

UZrCN has exhibited thermophysical properties beneficial to high temperature reactor applications such as nuclear thermal rockets. The present work investigates a novel liquid phase synthesis method involving the admission of nitrogen gas during arc melting of uranium, zirconium, and carbon. Initial microstructural examinations using SEM indicated that the samples remain heterogenous with zirconium-rich cores in a uranium-rich matrix. Heterogeneity resulted from large differences in melting temperature of the major constituents and rapid solidification. Additional EDS, combustion analysis, and inert gas fusion analysis proved that the core regions have both uranium and zirconium and establish that carbon and nitrogen are retained during the fabrication process. XRD clarifies that the light elements did not form compounds with the uranium matrix but rather incorporated into the core region forming a sub-stoichiometric UZrCN. Heat treatment performed on equimolar U-Zr-C resulted in an increase in the homogenous phase present with an XRD pattern reflecting a ternary UZrCN despite small amounts of segregation in the final button.

36 - MATERIALS SCIENCE↗
DOE OSTI2025

On Uranium-Zirconium Carbonitride Formation by Direct Casting - A Novel Synthesis Study

UZrCN has exhibited thermophysical properties beneficial to high temperature reactor applications such as nuclear thermal rockets. The present work investigates a novel liquid phase synthesis method involving the admission of nitrogen gas during arc melting of uranium, zirconium, and carbon. Initial microstructural examinations using SEM indicated that the samples remain heterogenous with zirconium-rich cores in a uranium-rich matrix. Heterogeneity resulted from large differences in melting temperature of the major constituents and rapid solidification. Additional EDS, combustion analysis, and inert gas fusion analysis proved that the core regions have both uranium and zirconium and establish that carbon and nitrogen are retained during the fabrication process. XRD clarifies that the light elements did not form compounds with the uranium matrix but rather incorporated into the core region forming a sub-stoichiometric UZrCN. Heat treatment performed on equimolar U-Zr-C resulted in an increase in the homogenous phase present with an XRD pattern reflecting a ternary UZrCN despite small amounts of segregation in the final button.

36 - MATERIALS SCIENCE↗
DOE OSTI2020

BISON BWR Fuel Modeling Capability: Material Models (FY2020)

Until this year the BISON fuel performance code has been developed primarily for pressurized water reactor (PWR), accident tolerant fuel (ATF), and advanced reactor (e.g., tri-structural isotropic (TRISO), metallic) analyses. The main focus of this work is to develop and document required material models that would aid predicting cladding failure with a specific hydride distribution to be investigated for boiling water reactor (BWR) applications in the next fiscal year. In terms of materials, the primary differences between a BWR and PWR is the use of gadolinia doping of the UO 2 fuel, the use of Zircaloy-2 for the cladding, and the addition of a liner on the inner surface of the cladding for improved pellet-clad mechanical interaction (PCMI)performance. The liner is typically made of pure zirconium or a lower tin content zirconium alloy (e.g. Zr-0.3%Sn). This work details the development and documentation of material and behavioral models for BWR analysis such as the addition of material and behavior models for gadolinia doped UO 2 , pure zirconium, Zr-0.3%Sn, and Zircaloy-2 as well as the ease-of-use additions to allow internal meshing capabilities to include liners. Another important aspect affecting cladding performance is hydrogen diffusion and hydride precipitation in the cladding. A model was added previously for non-lined, homogeneous claddings. In this work, instructions are provided on how to apply the model to BWR liner cladding. Representative experimental values are provided for the hydrogen diffusion parameters and precipitation-dissolution kinetics for typical liner materials of interest.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Fuel Performance Analysis of Chromium-Coated Cladding under Burst Conditions

To reduce the oxidation of zirconium-based alloy cladding at high temperatures, accident tolerant fuel systems have been proposed. Of the concepts identified, chromium-coated cladding has been shown to slow oxidation without greatly impacting the fuel system geometry or neutronic performance. To determine how coated-cladding tubes will perform under high-temperature accident conditions, pressurized-tube burst tests have been performed using the Severe Accident Test Station at Oak Ridge National Laboratory. To begin modeling these tubes to better understand how the coating will impact cladding behavior, these burst tests were simulated with the BISON fuel performance code. Cladding tube surface temperatures for the burst test were developed by fitting thermocouple data into axial and azimuthal profiles, while pressure data were compared until cladding failure. The temperatures at failure and the pressure evolution show relatively good agreement between the simulation and experiment results. This is the first step of a larger effort to simulate the cladding deformation process under high-temperature transient conditions and assess the cladding margin to failure more accurately.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Final-stage densification kinetics of direct current–sinteredZrB2

Final-stage sintering was analyzed for nominally phase pure zirconium diboride synthesized by borothermal reduction of high-purity ZrO2. Analysis was conducted on ZrB2 ceramics with relative densities greater than 90% using the Nabarro–Herring stress–directed vacancy diffusion model. Temperatures of 1900°C or above and an applied uniaxial pressure of 50 MPa were required to fully densify ZrB2 ceramics by direct current sintering. Ram travel data were collected and used to determine the relative density of the specimens during sintering. Specimens sintered between 1900 and 2100°C achieved relative densities greater than 97%, whereas specimens sintered below 1900°C failed to reach the final stage of sintering. The average grain size ranged from 1.0 to 14.7 μm. The activation energy was calculated from the slope of an Arrhenius plot that used the Kalish equation. The activation energy was 162 ± 34 kJ/mol, which is consistent with the activation energy for dislocation movement in ZrB2. The diffusion coefficients for dislocation motion that controls densification were 5.1 × 10−6 cm2/s at 1900°C and 5.1 × 10−5 cm2/s at 2100°C, as calculated from activation energy and average grain sizes. This study provides evidence that the dominant mechanism for final-stage sintering of ZrB2 ceramics is dislocation motion.

36 MATERIALS SCIENCE↗
DOE OSTI2020

Stability of zirconium hydride

This reports complete Milestone M2.2.1 Stability of zirconium hydride: Experimental investigation of the hydrogen concentration of encapsulated zirconium hydride before and after joining. In this milestone, we first report the successful fabrication of crack-free bulk zirconium hydride by using the ORNL static hydriding system. The fabricated hydride was characterized by using X-ray diffraction (XRD) and X-ray computed tomography (XCT). XRD results showed that the δ-phase zirconium hydride, α-zirconium, and ε-phase hydride were present. XCT confirmed the absence of the internal cracking. The hydrogen concentration was determined by the weight change before and after hydrogen loading, given the fact that the oxide phase was not found during XRD analysis. Two metallic cladding materials, i.e., molybdenum and FeCrAl, were selected to encapsulate the fabricated zirconium hydride pellets. Demonstration capsules were designed and machined. Mo crucible was successfully sealed by using a combination of electron beam welding and gas tungsten arc welding. Due to the undesired neutronics property of Mo, further investigation of this cladding material was not pursued. Electron beam welding and laser beam welding were employed to seal the FeCrAl capsules. XRD were performed on the zirconium hydride before and after the sealing. The results showed that the welding process has negligible impact on the hydrogen content in the FeCrAl-canned zirconium hydride. The presented work fulfilled the proposed work under this milestone and paves a pathway to fabricate single-phase zirconium hydride and to develop robust cladding materials to can bulk zirconium hydride.

36 MATERIALS SCIENCE↗
DOE OSTI2022

A study on texture stability and the biaxial creep behavior of as-hydrided CWSR Zircaloy-4 cladding at the effective stresses from 55 MPa to 65 MPa and temperatures from 300°C to 400°C

The creep rupture of high burnup used nuclear fuel (> 45 GWD/MTU) cladding is regarded as one of the failure mechanisms during long-term dry storage. A high amount of zirconium hydride in the cladding matrix would degrade the mechanical properties of the cladding especially leading to delayed hydride cracking. To better understand the influence of zirconium hydride on the biaxial thermal creep performance and the crystalline texture stability of Zircaloy-4 cladding, the pressurized tube technique is applied to test the durability of the as-hydrided material. Tests were performed on as-received Zircaloy-4 tubular specimens as well as-hydrided ones with targeted 300 wt parts per million (wppm) or 750 wppm hydrogen. The biaxial creep experiments were conducted at temperatures from 300 degrees C to 400 degrees C and at equivalent stresses at mid-wall from 55 MPa to 65 MPa. The hydridation process prior to creep tests induces the formation of FCC delta-hydride platelets along the circumferential direction of the tube. This alignment and phase structure of hydride show no significant change after biaxial creep tests. The creep strain-rate negatively depends on the hydrogen content. The synchrotron wide-angle X-ray diffraction (WAXD) technique and electron backscatter diffraction (EBSD) analysis were applied for the study of the crystallographic orientation relationship. zirconium-to-hydride grain orientation follows Shoji-Nishiyama crystallographic relationship. This relationship is stable before and after creep deformation. These results of creep performance and texture stability of Zircaloy-4 claddings can help support the design basis of interim and long-term dry storage facilities.

creep↗
DOE OSTI2021

Observation of Fundamental Mechanisms in Compression-Induced Phase Transformations Using Ultrafast X-ray Diffraction

As theoretically hypothesized for several decades in group IV transition metals, we have discovered a dynamically stabilized body-centered cubic (bcc) intermediate state in Zr under uniaxial loading at sub-nanosecond timescales. Under ultrafast shock wave compression, rather than the transformation from α-Zr to the more disordered hex-3 equilibrium ω-Zr phase, in its place we find the formation of a previously unobserved nonequilibrium bcc metastable intermediate. We probe the compression-induced phase transition pathway in zirconium using time-resolved sub-picosecond x-ray diffraction analysis at the Linac Coherent Light Source. We also present molecular dynamics simulations using a potential derived from first-principles methods which independently predict this intermediate phase under ultrafast shock conditions. In contrast with experiments on longer timescale (> 10 ns) where the phase diagram alone is an adequate predictor of the crystalline structure of a material, our recent study highlights the importance of metastability and time dependence in the kinetics of phase transformations.

36 MATERIALS SCIENCE↗
DOE OSTI2024

Analysis of the Surface Morphology and Chemical Composition of Zr-Nb3Sn Alloys with Zr Different Concentrations

The inclusion of zirconium (Zr) in niobium-tin (Nb3Sn) significantly enhances the performance of Nb3Sn radiofrequency cavities in high magnetic fields. This research project is dedicated to characterizing the surface properties and chemical composition of Zr-doped Nb3Sn. Two different concentrations of Zr were used for doping: approximately 0.5% and 24%. Various spectroscopy techniques were employed to analyze how the surface morphology and chemical composition of Nb3Sn change with increasing Zr content. The findings of this study indicate that the grain size decreases as the Zr concentration increases. In samples with 24% Zr, nanometer-sized particles, likely oxides, were observed. X-ray photoelectron spectroscopy (XPS) data revealed that the thickness of niobium oxides (NbOx) decreases with increasing Zr, while the thickness of tin oxides (SnOx) increases. Additionally, grain size distribution analysis showed that the average grain size is around 5300 nm , with a grain area density of about 165 grains/μm .

Sue, Micah↗
DOE OSTI2023

Photocatalytic degradation of orange II dye via high pore volume polymorph zirconium-doped titania under visible light irradiation

A series of mixed-phase Zr-doped TiO 2 was utilized for the photocatalytic degradation of orange II dye. These materials were synthesized through a sol–gel technique by varying the amounts of zirconium in the photocatalysts. These materials were characterized using X-ray diffraction (XRD), transmission electron microscopy (TEM), X-ray fluorescence (XRF), Raman spectroscopy, thermogravimetry analysis (TGA), and X-ray photoelectron spectroscopy (XPS) among others. XRF and XPS data showed that the elemental composition of the photocatalysts consists of zirconium, titanium, and oxygen. In addition, XRD and Raman spectroscopy confirmed the existence of anatase and rutile phases of TiO 2 . The TGA analyses showed that the undoped TiO 2 was the most stable, with no significant weight loss during the heating range. Furthermore, the materials can effectively degrade orange II dye solution up to 7 cycles through asymmetric cleavage of the azo bond.

36 MATERIALS SCIENCE↗
DOE OSTI2025

High resolution microstructural, chemical studies and localized burnup analysis in an irradiated U–10Zr metallic fuel

This study involves high resolution characterization of a sodium bonded solid uranium (U)-10 wt% zirconium (Zr) metallic fuel irradiated in Fast Flux Test Facility (FFTF). The fuel centerline temperature during irradiation was estimated to be around 675 °C with the peak burnup 13.1 atomic percent. Samples for transmission electron microscopy (TEM) and atom probe tomography (APT) were prepared from different regions/zones in the fuel cross section radially to elucidate the microstructural changes and chemical redistribution of solute elements as well as fission products during irradiation experiment. TEM results indicate the irradiation in fast flux testing leads to the formation of extensive Zr-rich precipitates with varying sizes in the U–Zr fuel matrix. APT analysis performed to investigate redistribution of Zr, U and fission products along the radial direction of fuel pin showed Zr-rich precipitates entrapping the fission products in higher concentration as compared to the α-U phases. Here, the local burnup ( 235 U depletion) is found to be consistent, calculated by quantification of 235 U, 236 U and 238 U isotopes from mass spectrum obtained from APT along the radial direction. Zr-rich precipitation and its implication on fuel constitutional redistribution are discussed based on SEM, TEM and APT results.

APT↗
DOE OSTI2022

Capability to Process and Characterize Uranium-Zirconium (U-Zr) and Uranium-Zirconium-Plutonium (U-Zr-Pu) Alloys

This work investigated co-reduction of anhydrous compounds of uranium, zirconium, and plutonium to produce uranium rich ternary nuclear fuel alloys. Metallothermic co-reduction is a novel method to produce all-metal nuclear fuels. Metal fuels offer thermal, fissility, compatibility, and security benefits over oxide fuels. Alloys of uranium–10% zirconium with plutonium contents of 0%, 2.5%, 5%, and 10% were produced, with yields of 60–85% of theoretical values in a traditional calciothermic bomb reduction apparatus. Microstructural analysis indicated transformation of uranium phase from gamma to beta and then alpha, in alternating lamellar plates typical of alpha phase uranium and delta phase uranium-zirconium, with zirconium and carbides at prior grain boundaries. Some of the analyses were inconclusive in their results and therefore require additional testing. Successful ternary co-reduction would simplify production of homogeneous feedstock and thereby streamline manufacture of homogeneous ternary metallic fuel.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

In situ synchrotron tensile investigations on ultrasonic additive manufactured (UAM) zirconium

The microstructure evolution during room temperature uniaxial tensile straining of ultrasonic additive manufactured (UAM) zirconium was evaluated using the Advanced Photon Source (APS) facility. Miniature dog-bone tensile specimens of two orientations were cut from a UAM-fabricated zirconium bar for in situ synchrotron tensile tests. Wide-angle X-ray scattering (WAXS) scanning was used to unveil the changes in microstructure of the entire gauge regions throughout the straining. A series of WAXS data analysis methods were utilized to quantify both elastic and plastic deformation mechanisms within the strained specimens. Stress concentrations were identified during early stages of plastic deformation, which become candidate necking positions and eventually lead to failure. Additionally, fracture surface analysis implied that these stress concentration locations may be correlated to the fabrication defects, providing insightful guidance for future improvements of the UAM zirconium fabrication process.

36 MATERIALS SCIENCE↗
DOE OSTI2023

ORNL Analysis of Leach-Burn-Leach Round-Robin Test Samples

An international round-robin test to examine the consistency in leach-burn-leach (LBL) analysis of tristructural-isotropic- (TRISO-) coated particle fuel was conducted by three research organizations from the Generation IV International Forum member countries of the People’s Republic of China, the Republic of Korea, and the United States of America. Two sets of round-robin test samples were exchanged for analysis. One set of samples consisted of a series of nonuranium-bearing, TRISO-coated zirconium dioxide particles seeded with up to four depleted uranium-bearing, TRISO-coated uranium dioxide (UO 2 ) particles, which had intentionally damaged coating layers to simulate either particles with either exposed-kernel defects (i.e., particles with a cracked TRISO coating that should be detected during preburn leaching) or particles with silicon carbide (SiC) defects (i.e., particles with an intact pyrocarbon coating and a hole in the SiC layer that should be detected during postburn leaching). These simulated samples also contained added powder with known quantities of impurities from a coal standard. The other sample set consisted of representative sublots of UO 2 -TRISO particles fabricated in a production-scale coater, except they all contained depleted uranium instead of enriched uranium. In this report, the methodology used at Oak Ridge National Laboratory to conduct LBL analysis of the round-robin samples is presented, and the general results are summarized.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗