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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 19 records
All sourcesNASA NTRSDOE OSTI
DOE OSTI2015

Topology-Guided Design and Syntheses of Highly Stable Mesoporous Porphyrinic Zirconium Metal-Organic Frameworks with High Surface Area

Through a topology-guided strategy, a series of Zr-6-containing isoreticular porphyrinic metalorganic frameworks (MOFs), PCN-228, PCN-229, and PCN-230, with ftw-a topology were synthesized using the extended porphyrinic linkers. The bulky porphyrin ring ligand effectively prevents the network interpenetration which often appears in MOFs with increased linker length. The pore apertures of the structures range from 2.5 to 3.8 nm, and PCN-229 demonstrates the highest porosity and BET surface area among the previously reported Zr-MOFs. Additionally, by changing the relative direction of the terminal phenyl rings, this series replaces a Zr-8 cluster with a smaller Zr-6 cluster in a topologically identical framework. The high connectivity of the Zr-6 cluster yields frameworks with enhanced stability despite high porosity and ultralarge linker. As a representative example, PCN-230, constructed with the most extended porphyrinic linker, shows excellent stability in aqueous solutions with pH values ranging from 0 to 12 and demonstrates one of the highest pH tolerances among all porphyrinic MOFs. This work not only presents a successful example of rational design of MOFs with desired topology, but also provides a strategy for construction of stable mesoporous MOFs.

Liu, TF↗
DOE OSTI2018

Optimized anion exchange column isolation of zirconium-89 (89Zr) from yttrium cyclotron target: Method development and implementation on an automated fluidic platform

Zirconium-89 (89Zr), produced by the (p,n) reaction from naturally monoisotopic yttrium (natY), is a promising positron emitting isotope for immunoPET imaging. Its long half-life of 78.4 h is sufficient for evaluating slow physiological processes. A prototype automated fluidic system, coupled to on-line and in-line detectors, has been constructed to facilitate development of new 89Zr purification methodologies. The highly reproducible reagent delivery platform and near-real time monitoring of column effluents allows for efficient method optimization. The separation of Zr from dissolved Y metal targets was evaluated using several anion exchange resins. Each resin was evaluated against its ability to quantitatively capture Zr from a load solution that is high in dissolved Y. The most appropriate anion exchange resin for this application was identified, and the separation method was optimized. The method is capable of a high Y decontamination factor (>105) and has been shown to separate Fe, an abundant contaminant in Y foils, from the 89Zr elution fraction. Finally, the performance of the method was evaluated using cyclotron bombarded Y foil targets. The separation method was shown to achieve >95% recovery of the 89Zr present in the foils. The 89Zr eluent, however, was in a chemical matrix not immediately conducive to labeling onto proteins. The main intent of this study was to develop a tandem column 89Zr purification process, wherein the anion exchange column method described here is the first separation in a dual-column purification process.

Zirconium-89, ImmunoPET, medical isotope, automati↗
DOE OSTI2015

Non-equilibrium oxidation states of zirconium during early stages of metal oxidation

The chemical state of Zr during the initial, self-limiting stage of oxidation on single crystal zirconium (0001), with oxide thickness on the order of 1 nm, was probed by synchrotron x-ray photoelectron spectroscopy. Quantitative analysis of the Zr 3d spectrum by the spectrum reconstruction method demonstrated the formation of Zr 1+ , Zr 2+ , and Zr 3+ as non-equilibrium oxidation states, in addition to Zr 4+ in the stoichiometric ZrO 2 . This finding resolves the long-debated question of whether it is possible to form any valence states between Zr 0 and Zr 4+ at the metal-oxide interface. As a result, the presence of local strong electric fields and the minimization of interfacial energy are assessed and demonstrated as mechanisms that can drive the formation of these non-equilibrium valence states of Zr.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2017

Use of multiscale zirconium alloy deformation models in nuclear fuel behavior analysis

Accurate prediction of cladding mechanical behavior is a key aspect of modeling nuclear fuel behavior, especially for conditions of pellet-cladding interaction (PCI), reactivity-initiated accidents (RIA), and loss of coolant accidents (LOCA). Current approaches to fuel performance modeling rely on empirical constitutive models for cladding creep, growth and plastic deformation, which are limited to the materials and conditions for which the models were developed. To improve upon this approach, a microstructurally-based zirconium alloy mechanical deformation analysis capability is being developed within the United States Department of Energy Consortium for Advanced Simulation of Light Water Reactors (CASL). Specifically, the viscoplastic self-consistent (VPSC) polycrystal plasticity modeling approach, developed by Lebensohn and Tomé [1], has been coupled with the BISON engineering scale fuel performance code to represent the mechanistic material processes controlling the deformation behavior of light water reactor (LWR) cladding. A critical component of VPSC is the representation of the crystallographic nature (defect and dislocation movement) and orientation of the grains within the matrix material and the ability to account for the role of texture on deformation. A future goal is for VPSC to obtain information on reaction rate kinetics from atomistic calculations to inform the defect and dislocation behavior models described in VPSC. The multiscale modeling of cladding deformation mechanisms allowed by VPSC far exceed the functionality of typical semi-empirical constitutive models employed in nuclear fuel behavior codes to model irradiation growth and creep, thermal creep, or plasticity. This paper describes the implementation of an interface between VPSC and BISON and provides initial results utilizing the coupled functionality.

75 CONDENSED MATTER PHYSICS, SUPERCONDUCTIVITY AND↗
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 OSTI2018

Development of Self-Healing Zirconium-Silicide Coatings for Improved Performance Zirconium-Alloy Fuel Cladding

Zirconium-alloy fuel claddings have been used successfully in Light Water Reactors (LWR) for over four decades. However, under high temperature accident conditions, zirconium-alloys fuel claddings exhibit profuse exothermic oxidation accompanied by release of hydrogen gas due to the reaction with water/steam. Additionally, the ZrO 2 layer can undergo monoclinic to tetragonal to cubic phase transformations at high temperatures which can induce stresses and cracking. These events were unfortunately borne out in the Fukushima-Daiichi accident in in Japan in 2011. In reaction to such accident, protective oxidation-resistant coatings for zirconium-alloy fuel claddings has been extensively investigated to enhance safety margins in accidents as well as fuel performance under normal operation conditions. Such surface modification could also beneficially affect fuel rod heat transfer characteristics. Zirconium-silicide, a candidate coating material, is particularly attractive because zirconium-silicide coating is expected to bond strongly to zirconium-alloy substrate. Intermetallic compound phases of zirconium-silicide have high melting points and oxidation of zirconium silicide produces highly corrosion resistant glassy zircon (ZrSiO 4 ) and silica (SiO 2 ) which possessing self-healing qualities. Given the long-term goal of developing such coatings for use with nuclear reactor fuel cladding, this work describes results of oxidation and corrosion behavior of bulk zirconium-silicide and fabrication of zirconium-silicide coatings on zirconium-alloy test flats, tube configurations, and SiC test flats. In addition, boiling heat transfer of these modified surfaces (including ZrSi 2 coating) during clad quenching experiments is discussed in detail.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2023

A novel protocol to recycle zirconium from zirconium alloy cladding from used nuclear fuel rods

A potential new zirconium recycling protocol has been demonstrated using unirradiated Zircaloy-based claddings and depleted uranium oxide for the chemical removal of zirconium alloy cladding in used nuclear fuel (UNF) rods from light water reactors. This protocol is based on the application of three new scientific findings. First, a new lower temperature chlorination reaction of zirconium with sulfur monochloride is described. Second, the high solubility of zirconium chloride in thionyl chloride is used to separate it from uranium oxide fuel and fission and activation contaminants. Finally, in the third step, the zirconium chloride is purified by simple recrystallization from thionyl chloride. Utilizing sulfur chloride solvents for a lower temperature liquid-based chemical digestion and purification of zirconium alloy claddings reduces technical complications experienced by current high temperature gas phase chlorination strategies, such as contamination of product streams. This novel protocol has been demonstrated on a 3–50 g scale of unirradiated zirconium alloy, with no significant change in the time required for complete chlorination (3–4 h). This zirconium chlorination protocol has also been performed in the presence of depleted uranium oxide pellets. The depleted fuel pellets do not affect zirconium chlorination, and the chlorination protocol does not chemically or physically alter the fuel pellets. Finally, a preliminary description of an industrial protocol to recycle nuclear grade zirconium from UNF rods is presented.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2017

Pressure-dependence of the phase transitions and thermal expansion in zirconium and hafnium pyrovanadate

Low or negative thermal expansion (NTE) has been previously observed in members of the ZrP{sub 2}O{sub 7} family at temperatures higher than their order-disorder phase transitions. The thermoelastic properties and phase behavior of the low temperature superstructure and high temperature negative thermal expansion phases of ZrV{sub 2}O{sub 7} and HfV{sub 2}O{sub 7} were explored via in situ variable temperature/pressure powder x-ray diffraction measurements. The phase transition temperatures of ZrV{sub 2}O{sub 7} and HfV{sub 2}O{sub 7} exhibited a very strong dependence on pressure (∼700 K GPa), with moderate compression suppressing the formation of their NTE phases below 513 K. Compression also reduced the magnitude of the coefficients of thermal expansion in both the positive and negative thermal expansion phases. Additionally, the high temperature NTE phase of ZrV{sub 2}O{sub 7} was found to be twice as stiff as the low temperature positive thermal expansion superstructure (24 and 12 GPa respectively). - Graphical abstract: The temperature at which ZrV{sub 2}O{sub 7} transforms to a phase displaying negative thermal expansion is strongly pressure dependent. The high temperature form of ZrV{sub 2}O{sub 7} is elastically stiffer than the low temperature form. - Highlights: • The order-disorder phase transition temperatures in ZrV{sub 2}O{sub 7} and HfV{sub 2}O{sub 7} are strongly pressure dependent (∼700 K.GPa). • The high temperature (disordered) phase of ZrV{sub 2}O{sub 7} is much stiffer than the ambient temperature (ordered) phase. • Compression reduces the magnitude of the negative thermal expansion in the high temperature phase of ZrV{sub 2}O{sub 7}.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2016

Development of Zirconium-Silicide Coatings for Improved Performance Zirconium-Alloy Fuel Cladding

There is considerable interest in the development of coatings for zirconium-alloy fuel cladding used in light water reactors (LWR) to improve their oxidation resistance during high temperature accident conditions and to enhance performance under normal operating conditions. These coatings should also enhance the heat transfer characteristics of the cladding surface so that heat from the fuel rod is efficiently transferred into the water coolant. One potential candidate coating material is zirconium-silicide. Transition metal silicides have been generally shown to have outstanding oxidation resistance. Zirconium-silicide is particularly attractive because Zr is strongly oxophilic, so that the Zr-silicide layer is expected to bond more strongly to the inner, native ZrO{sub 2} layer. The relative proximity of the free energy of formation of oxides of Zr and Si can lead to alternating multilayered oxide structures with radiation damage resistance. Additionally, oxidation of zirconium silicide can produce excellent corrosion resistant glassy zircon (ZrSiO{sub 4}) and silica (SiO{sub 2}), possessing self-healing qualities. This paper reports results of our results of high temperature oxidation tests of bulk zirconium-silicide and coatings of zirconium-silicide deposited on zirconium-alloy test flats and cladding tube configurations using plasma vapor deposition (PVD) process. We also report our initial results on the experimental facility for studying boiling characteristics of coated cladding tubes in prototypic pressurized water conditions. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Adding to the Arsenal of Zirconium‐Based Metal–Organic Frameworks: the Topology as a Platform for Solvent‐Assisted Metal Incorporation

Gaining control over the assembly of mesoporous Zr‐based metal–organic frameworks (MOFs) with modifiable metal nodes remains a challenge. The the topology is intriguing, particularly if considering postsynthetic modification, as it presents large sodalite‐like cages surrounded by 12 interconnected 1D channels that allow for efficient diffusion of guest molecules throughout the structure. Herein, we demonstrate the first Zr MOF with the topology, NU‐1200, which was synthesized by controlling the geometry of the organic building unit. In addition, by using NU‐1200 as a platform, we perform solvent‐assisted metal incorporation to form single‐site metalated nodes, a process that is characterized by single‐crystal X‐ray diffraction.

Liu, Tian‐Fu↗
DOE OSTI2014

A Highly Stable Porphyrinic Zirconium Metal–Organic Framework with shp-a Topology

Through a kinetically controlled synthetic process, we synthesized PCN-223, a new porphyrinic Zr-MOF constructed from the newly reported hexagonal prismatic 12-connected Zr6 cluster through an unusual disordered arrangement, giving rise to the first example of the shp-a network in MOFs. With its extremely high connectivity, PCN-223 shows high stability in aqueous solutions with a wide range of pH. Cationic PCN-223(Fe) formed by postsynthetic treatment is an excellent recyclable heterogeneous catalyst for the hetero-Diels–Alder reaction.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2014

The {sup 13}C-pocket structure in AGB models: constraints from zirconium isotope abundances in single mainstream SiC grains

We present postprocess asymptotic giant branch (AGB) nucleosynthesis models with different {sup 13}C-pocket internal structures to better explain zirconium isotope measurements in mainstream presolar SiC grains by Nicolussi et al. and Barzyk et al. We show that higher-than-solar {sup 92}Zr/{sup 94}Zr ratios can be predicted by adopting a {sup 13}C-pocket with a flat {sup 13}C profile, instead of the previous decreasing-with-depth {sup 13}C profile. The improved agreement between grain data for zirconium isotopes and AGB models provides additional support for a recent proposal of a flat {sup 13}C profile based on barium isotopes in mainstream SiC grains by Liu et al.

79 ASTRONOMY AND ASTROPHYSICS↗
DOE OSTI2017

Multi-Functional Zirconium-Silicide Coatings on Zirconium-alloy for Improved Accident Tolerance

Recently, development of protective coatings on zirconium-alloy (Zr-alloy) fuel cladding in light water reactors has been extensively investigated to enhance safety margins in accident scenarios. One accident scenario is degradation of the reactor core under a coolant-limited condition, where a large amount of Zr-alloy claddings would react with high temperature steam, leading to oxidation, hydrogen production, and loss of intrinsic mechanical properties of the fuel cladding. Although, total replacement of Zr-alloy with SiC-SiCf composites or FeCrAl is being considered for the long-term, application of oxidation resistant coatings on zirconium-alloy is being investigated as a near-term solution to mitigate oxidation of Zr-alloy and increase the accident coping time. Coatings, such as chromium, Ti{sub 2}AlC, SiC, and Cr{sub 3}C{sub 2}- NiCr, deposited by various deposition techniques have been investigated. In addition, recent studies on modified heating surface for heat transfer enhancement provides basis on the optimal design of coated Zr-alloy claddings so that the new cladding can be achieve accident tolerance and heat transfer efficiency for both accident and normal operation conditions. Therefore, boiling heat transfer from the coated surface to water environment must be evaluated in parallel. In heat transfer performance, boiling heat transfer is characterized by nucleate boiling, critical heat flux (CHF), transition boiling, minimum film boiling temperature (T{sub min}), and film boiling. Zirconium-silicide (ZrSi{sub 2}) is a potentially attractive protective coating material for cladding applications. Although this material has not been studied extensively, it appears to be a promising candidate because of the high melting points, and the low neutron penalty, and the generally good oxidation resistance to transition metal silicides. It has been reported that oxidation of ZrSi{sub 2} powder in a flowing air environment at elevated temperatures produced ZrO{sub 2}, SiO{sub 2}, and ZrSiO{sub 4}, pointing to the possibility of achieving highly protective surface layers of SiO{sub 2} and ZrSiO{sub 4} that would protect the material from further oxidation. In this study, magnetron sputter deposition of ZrSi{sub 2} on zirconium-alloy flats and cylindrical rods was performed. The coated samples were oxidized in high temperature air (700 deg. C, 1000 deg. C and 1200 deg. C) to evaluate their oxidation resistance. The coated cylindrical rods were quenched in saturated water to observe transition boiling, T{sub min}, film boiling regime, while the coated flats were tested in pool boiling facility to investigate nucleation boiling and CHF.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2014

K{sub α} x-ray imaging of laser-irradiated, limited-mass zirconium foils

X-ray fluorescence measurements to determine the effect of target heating on imaging efficiency, at a photon energy of 15.7 keV corresponding to the K{sub α} line of zirconium, have been carried out using limited-mass foils irradiated by the Texas Petawatt Laser. Zirconium foils that ranged in volume from 3000 × 3000 × 21 μm{sup 3} to 150 × 150 × 6 μm{sup 3} were irradiated with 100 J, 8 ps-long pulses and a mean intensity of 4 × 10{sup 19} W/cm{sup 2}. The K{sub α} emission was measured simultaneously using a highly ordered pyrolytic graphite crystal spectrometer and a curved quartz imaging crystal. The measured ratio of the integrated image signal to the integrated spectral signal was, within the experimental error, constant, indicating that the imaging efficiency's dependence on temperature is weak throughout the probed range. Based on our experience of target heating under similar conditions, we estimate a temperature of ∼200 eV for the smallest targets. The successful imaging of K{sub α} emission for temperatures this high represents an important proof of concept for Zr K{sub α} imaging. At these temperatures, the imaging of K{sub α} emission from lower-Z materials (such as Cu) is limited by temperature-dependent shifts in the K{sub α} emission energy.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2014

Solid-state coexistence of {Zr 12 } and {Zr 6 } zirconium oxocarboxylate clusters

Ligand metathesis, Co(II) coordination, and partial condensation reactions of an archetypal {Zr6} zirconium oxocarboxylate cluster result in the first example of the coexistence of the distinct zirconium oxide frameworks {Zr6O8} and {Zr12O22}. Even minor modifications to the reaction conditions push this apparent equilibrium towards the {Zr6O8}-based product.

36 MATERIALS SCIENCE↗
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↗