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At least 163 records · Page 9
All sourcesNASA NTRSDOE OSTI
DOE OSTI2012

Process Development Studies for Zirconium Recovery/Recycle from Used Nuclear Fuel Cladding

Process development studies are being done to recover and reuse zirconium from used nuclear fuel (UNF) zirconium alloy cladding. This has been shown to be economically viable and is a major initiative in waste reduction. The studies have included recovery processes based on dry pyrochemical iodination, chlorination, and hydrochlorination. Chlorination, using either or both chlorine and hydrogen chloride, was shown to provide better performance and is more compatible for interfacing with existing industrial chlorination processes used for fuel rod cladding manufacture from natural zirconium. The chlorination recovery process could be used for chemical decladding of intact UNF assemblies or for treatment of empty cladding segment hulls. Future work is planned to optimize recovery process conditions, equipment design, and scale-up operations.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Assessment of wear coefficients of nuclear zirconium claddings without and with pre-oxidation

In the cores of pressurized water nuclear reactors, water-flow induced vibration is known to cause claddings on the fuel rods to rub against their supporting grids. Such grid-to-rod-fretting (GTRF) may lead to fretting wear-through and the leakage of radioactive species. The surfaces of actual zirconium alloy claddings in a reactor are inevitably oxidized in the high-temperature pressurized water, and some claddings are even pre-oxidized. As a result, the wear process of the surface oxide film is expected to be quite different from the zirconium alloy substrate. In this paper, we attempt to measure the wear coefficients of zirconium claddings without and with pre-oxidation rubbing against grid samples using a bench-scale fretting tribometer. Results suggest that the volumetric wear coefficient of the pre-oxidized cladding is 50 to 200 times lower than that of the untreated cladding. In terms of the linear rate of wear depth, the pre-oxidized alloy wears about 15 times more slowly than the untreated cladding. Finally, fitted with the experimentally-determined wear rates, a stage-wise GTRF engineering wear model demonstrates good agreement with in-reactor experience in predicting the trend of cladding lives.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2026

Nuclear activation analysis of zirconium-90 isomeric and ground-state reactions at the OMEGA Laser Facility

Nuclear activation is a well-established technique for inferring neutron yields in laser direct-drive deuterium–tritium (DT) and deuterium–deuterium (D 2 ) implosions at the OMEGA Laser Facility. Zirconium has long been considered an excellent candidate for measuring DT neutron fusion yields by observing decays of the 90 Zr(n,2n) 89 Zr ground-state reaction. As it has a higher energy threshold than present activation detectors utilizing copper, zirconium provides a means to infer primary neutron yields that are less susceptible to being skewed due to neutron scattering within the experimental environment. However, with a 78.41-h half-life, it is not operationally practical to utilize this reaction for OMEGA experiments, which have a 1-h shot cycle. Zirconium’s 90 Zr(n,2n) 89 mZr reaction presents itself as a viable candidate to infer neutron yields within a shot cycle, given its half-life of 4.16 min. Here, we present an overview of the approach and methodology, utilizing first principles techniques, to infer the primary neutron yields from OMEGA DT fusion experiments by using both the isomeric and the ground-state reaction. Yields inferred from both reactions are compared, which are in good agreement between the two.

Activation analysis↗
DOE OSTI2019

Elastic strain tensor of zirconium hydrides in Zr2.5%Nb pressure tubes by synchrotron X-ray diffraction

Zirconium alloys are used in fuel cladding and structural components of nuclear power plants. Hydrogen enters the Zr matrix during plant operation and precipitates as hydride particles that degrade the mechanical properties of the alloy, limiting service life. Knowledge of the stress state within hydride precipitates is important to understand stress-induced degradation mechanisms such as delayed hydride cracking, but no direct quantification has yet been reported in the literature. Here, measurements are reported of the average elastic strain tensor within δ zirconium hydride precipitates in Zr2.5%Nb pressure tube material from CANDU power plants. Complete intensity and strain pole figures for the hydride were obtained by synchrotron X-ray diffraction experiments on specimens with hydrogen contents ranging from ~100 wt p.p.m. hydrogen to nearly 100% δ-hydride. Zirconium hydride precipitates by a process involving a martensitic transformation, with two hydride variants possible from a single α-Zr grain. A synthetic model of the hydride crystallographic texture allowed the interpretation of the measured strain pole figures and quantification of the elastic strain tensor for both texture components. It was found that the two variants appear in nearly equal proportion but with different stress states, differing in the sign of the shear strain components (~±3000 µ$\epsilon$). This difference is possibly associated with the shear movement of Zr atoms during the phase transformation. This suggests that hydride clusters are composed of stacks of smaller hydrides in alternating hydride variants. Stresses were estimated from a set of rather uncertain hydride elastic constants. Overall, both variants showed compressive strains along the tube axial direction (~5000 µ$\epsilon$). For low hydrogen concentrations, the hydrides' stress tensor is dominated by compressive stresses of ~300 MPa along the axial direction, probably caused by the elongated morphology of hydride clusters along this direction, and variant-dependent shear stresses of ~±100 MPa, probably from the shear movement of the Zr atoms involved in the phase transformation.

36 MATERIALS SCIENCE↗
DOE OSTI2013

Understanding the Irradiation Behavior of Zirconium Carbide

Zirconium carbide (ZrC) is being considered for utilization in high-temperature gas-cooled reactor fuels in deep-burn TRISO fuel. Zirconium carbide possesses a cubic B1-type crystal structure with a high melting point, exceptional hardness, and good thermal and electrical conductivities. The use of ZrC as part of the TRISO fuel requires a thorough understanding of its irradiation response. However, the radiation effects on ZrC are still poorly understood. The majority of the existing research is focused on the radiation damage phenomena at higher temperatures (>450{degree}C) where many fundamental aspects of defect production and kinetics cannot be easily distinguished. Little is known about basic defect formation, clustering, and evolution of ZrC under irradiation, although some atomistic simulation and phenomenological studies have been performed. Such detailed information is needed to construct a model describing the microstructural evolution in fast-neutron irradiated materials that will be of great technological importance for the development of ZrC-based fuel. The goal of the proposed project is to gain fundamental understanding of the radiation-induced defect formation in zirconium carbide and irradiation response by using a combination of state-of-the-art experimental methods and atomistic modeling. This project will combine (1) in situ ion irradiation at a specialized facility at a national laboratory, (2) controlled temperature proton irradiation on bulk samples, and (3) atomistic modeling to gain a fundamental understanding of defect formation in ZrC. The proposed project will cover the irradiation temperatures from cryogenic temperature to as high as 800{degree}C, and dose ranges from 0.1 to 100 dpa. The examination of this wide range of temperatures and doses allows us to obtain an experimental data set that can be effectively used to exercise and benchmark the computer calculations of defect properties. Combining the examination of radiation-induced microstructures mapped spatially and temporally, microstructural evolution during post-irradiation annealing, and atomistic modeling of defect formation and transport energetics will provide new, critical understanding about property changes in ZrC. The behavior of materials under irradiation is determined by the balance between damage production, defect clustering, and lattice response. In order to predict those effects at high temperatures so targeted testing can be expanded and extrapolated beyond the known database, it is necessary to determine the defect energetics and mobilities as these control damage accumulation and annealing. In particular, low-temperature irradiations are invaluable for determining the regions of defect mobility. Computer simulation techniques are particularly useful for identifying basic defect properties, especially if closely coupled with a well-constructed and complete experimental database. The close coupling of calculation and experiment in this project will provide mutual benchmarking and allow us to glean a deeper understanding of the irradiation response of ZrC, which can then be applied to the prediction of its behavior in reactor conditions.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

IER 516: Zirconium Test Assembly CED Phase-1: Preliminary Experiment Design

IER-516, Zirconium Test Assembly (ZTA), is a campaign to design, execute, and document a series of high-fidelity critical benchmark experiments to validate current and future zirconium (Zr) and zirconium hydride (ZrH x ) nuclear data evaluations. ZTA is a collaborative project between Los Alamos National Laboratory and the French Autoritè de Sûretè Nuclèaire et de Radioprotection. The experiments will be fueled with highly enriched uranium (HEU), and utilize the Comet critical assembly machine at the National Criticality Experiments Research Center. A total of ten preliminary critical experiments were designed and optimized for Zr and ZrHx nuclear data sensitivities using the MCNP-Particle Swarm Optimization methodology in the thermal, epithermal, intermediate, and fast neutron energy regions.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Comparison of Zirconium Redistribution in BISON EBR-II Models Using FIPD and IMIS Databases with Experimental Post Irradiation Examination

Metallic fuels have seen increased interest for future sodium fast reactors due to their material properties: high thermal conductivities and advantageous neutronic properties allow for greater fission densities. One drawback to typical metallic fuels is zirconium redistribution, which impacts this advantageous material and its neutronic properties. Unfortunately, the processes behind zirconium migration behavior are understood using first principles, so before these fuels are implemented in future fast reactors, characterization and fuel qualification regimes must be completed. These activities can be supported through the use of robust modeling using the most accurate empirical models currently available to fuel researchers around the world. The tool that allows researchers to model this complex coupled thermo-mechanical behavior and nuclear properties is BISON. Additionally, BISON model parameters need to be compared against PIE measurements. The current work utilizes two fuel pins from EBR-II experiment X441 to optimize various model parameters, including porosity correction factor, thermal conductivity, phase transition temperature, and diffusion coefficient multipliers, before implementing the final model for seven fuel pins with differing characteristics. To properly evaluate the BISON simulations, the results are compared to PIE metallography data for each fuel pin, to ensure the zirconium redistribution is properly reflected in the simulation results. Six out of seven analyzed fuel pins demonstrate good agreement between the metallography images and BISON results, showing alignment of the Zr-rich, Zr-depleted, and moderately Zr-enriched zones at various axial heights along the fuel pins. Further work is needed to refine the model parameters for general pin use.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
NASA NTRS1995

Containerless Liquid to Solid Nucleation Pathways in Two Representative Grades of Commercially Available Zirconium

Experimental measurements were conducted to determine the solid metal nucleation pathways of radiatively cooling, molten zirconium spheres of two different commercially available purity grades in a high-vacuum, high-temperature electrostatic levitator. The ensemble distribution of maximum undercooling temperatures was interpreted using Poisson statistics to determine the temperature dependence of the solid metal nucleation rate. For a sample of nominally 99.95% pure zirconium, the results are consistent with heterogeneous solid metal nucleation either on static catalyst particles at least ~30 nm diameter or on a surface coating. For a sample of nominally 99% pure zirconium, however, it appears that heterogeneous solid metal nucleation occurred either on a polydispersion of ~10 nm (mean diameter) static catalyst particles or on dynamic catalyst particles which precipitated from a solution which became supersaturated as the melt cooled. A quantitative analysis of the compositions of both samples suggests that the catalysts probably consisted of one or more of the following elements: niobium, molybdenum, hafnium, and oxygen.

containerless, nucleation, zirconium↗
DOE OSTI2016

Corrosion resistant coatings for zirconium-alloy cladding with improved accident tolerance

The Westinghouse-led accident tolerant fuel (ATF) program is evaluating Accident Tolerant Fuel (ATF) coatings of MAX phase (Ti{sub 2}AlC) and multi-layer Ti{sub x}Al{sub 1-x}N/TiN to enhance the oxidation resistance of zirconium alloy fuel cladding in both normal operation and beyond design-based conditions of light water reactors. A variety of coating technologies have been applied to deposit these coating materials onto Zircaloy-4 or Zirlo sheets and tubes, including cold spray and cathodic arc-physical vapor deposition (CA-PVD). Neutronic analysis using the elemental compositions of the coating materials showed that the coating thicknesses should be less than 30 μm for most coating materials to avoid non-negligible economic penalty. The coated samples were autoclave tested in water at 360 Celsius degrees and 150 bar or in steam at 427 Celsius and 103 bar for corrosion resistance evaluation. The initial Ti{sub 2}AlC coating made by using large particle size MAX phase powder and cold spray deposition did not provide the desired corrosion resistance due to un-optimized microstructure and porosity. Subsequent optimization including reduction in particle size and tuning of cold spray deposition parameters resulted in denser coatings less than 30 microns in thickness. Optimization of process conditions resulted in reduced oxidation kinetics and showed that the coating acts as an effective barrier to prevent the zirconium substrate from oxidizing in accident condition temperatures. High purity Ti{sub 2}AlC powder is needed to further improve oxidation resistance of the Ti{sub 2}AlC coating. The oxidation resistance of pure monolithic MAX phase compound was also found to be remarkable, and the weight gain was only 1.85 mg/cm{sup 2} for 96 hours exposure in 1200 C. degrees steam, which is consistent with the literature. The corrosion resistance of the TiAlN/TiN multilayer coating was excellent. The weight gain was negligible and no oxidation was observed on the surface of zirconium-alloy at the coating-substrate interface. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
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 OSTI2013

Synthesis and structural characterization of a new chiral porous hybrid organic–inorganic material based on γ-zirconium phosphates and L-(+)-phosphoserine

In the present work, a chiral layered derivative of γ-zirconium phosphate (γ-ZrP) containing L-(+)-phosphoserine (γ-ZrP-PS*) covalently attached to inorganic layers has been prepared by means of topotactic exchange reaction. This organic–inorganic derivative is characterized by X-ray diffractometry, Solid {sup 13}C–NMR and FT-IR spectrophotometries and thermal analyses. A maximum level of topotactic replacement of 20% is achieved. Under both the acidic environment of the interlayer region of γ-ZrP and the acidic synthesis conditions, the hydrolysis of the ester bond of PS* is expected to take place to some extent. For this reason, it was impossible to exceed the recent percentage, which in turn reflects the relative moderate stability of the above mentioned bond under these conditions. In order to be more certain with regard to an expected further hydrolysis for this bond after separation, a sample of γ-ZrP-PS* was stored in a desiccator over a saturated solution of BaCl{sub 2} (90% relative humidity) for three months, and then the sample re-analyzed once again. Surprisingly, the results show that the sample still keeps almost the same level of exchange (i.e., 20%). Second, it is revealed that the sample almost gives the same spectroscopic and thermal behavior. This could be attributed to the less acidic character of the partially exchanged inorganic layers of the sample in comparison with that of the precursor γ-ZrP. Therefore, the PS* molecules persist and stay there into the interlayer gallery without further hydrolysis. - Graphical abstract: • Red: oxygen • White: zirconium • Cyan: carbon • Yellow: phosphorus • Blue: nitrogen. Highlights: • L-(+)-Phosphoserine (PS*) is exchanged with γ-ZrP by means of topotactic exchange. • The maximum exchange level is 20%. • γ-ZrP is functionalized with chiral amino acid group. • γ-ZrP-PS* has large chiral space for huge guest molecules to be intercalated.

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

Novel nanocomposites based on polythiophene and zirconium dioxide

Highlights: • Crystallinity of polythiophene decreased with doping process. • Zirconium dioxide changed the surface morphology of PT significantly. • Doping process increased the glass transition temperature of polythiophene. • Charge transport mechanism of all samples is consistent with CBH model. - Abstract: This paper reports charge transport mechanism of polythiophene (PT) and its nanocomposites having various weight of zirconium dioxide (ZrO{sub 2}). Fourier transform infrared (FT-IR) spectral studies showed that the Zr ions mostly affected the C−S bond of the polythiophene. X-ray diffraction (XRD) patterns indicated that ZrO{sub 2} powder was a mixture of monoclinic (64.62%) and tetragonal (35.38%) phase. XRD measurements revealed that crystallinity of PT decreased with increasing doping level. Significant morphological differences were observed in scanning electron microscopy (SEM) images of PT depending on the doping process. Thermal analysis carried out by differential scanning calorimetry (DSC) indicated that the doping processes increased glass transition temperature of PT. The alternating current (ac) conductivity mechanism analysis by universal power law indicated that charge transport mechanism of both PT and PT/ZrO{sub 2} nanocomposites was consistent with correlated barrier hopping (CBH) model. The doping process and frequency affected the hopping distance between the sites.

36 MATERIALS SCIENCE↗
DOE OSTI2019

Studying Changes in the Microhardness of a Zirconium Surface Modified by Laser Processing

The results of an investigation of the process of laser-induced hardening of zirconium (E110 alloy) are reported. It has been established that the laser treatment in air results in the formation of a uniformly distributed microrelief characterized by the presence of microparticles with various configurations. Overlap of the laser beam trajectories leads to the formation of a layered structure. The composition of laser-modified surface depends on the regime of processing and mostly comprises a mixture of zirconium oxides and nitrides. In addition, the formation of diffuse transition zones between the bulk metal and modified surface layer is established. The dependence of the thickness of the hardened layer on the regime of pulsed laser action is determined.

75 CONDENSED MATTER PHYSICS, SUPERCONDUCTIVITY AND↗
DOE OSTI2013

Establishment of the roadmap for chlorination process development for zirconium recovery and recycle

Process development studies are being conducted to recover, purify, and reuse the zirconium (about 98.5% by mass) in used nuclear fuel (UNF) zirconium alloy cladding. Feasibility studies began in FY 2010 using empty cladding hulls that were left after fuel dissolution or after oxidation to a finely divided oxide powder (voloxidation). In FY 2012, two industrial teams (AREVA and Shaw-Westinghouse) were contracted by the Department of Energy Office of Nuclear Energy (NE) to provide technical assistance to the project. In FY 2013, the NE Fuel Cycle Research and Development Program requested development of a technology development roadmap to guide future work. The first step in the roadmap development was to assess the starting point, that is, the current state of the technology and the end goal. Based on previous test results, future work is to be focused on first using chlorine as the chlorinating agent and secondly on the use of a process design that utilizes a chlorination reactor and dual ZrCl{sub 4} product salt condensers. The likely need for a secondary purification step was recognized. Completion of feasibility testing required an experiment on the chemical decladding flowsheet option. This was done during April 2013. The roadmap for process development will continue through process chemistry optimization studies, the chlorinated reactor design configuration, product salt condensers, and the off-gas trapping of tritium or other volatile fission products from the off-gas stream. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2019

Reticular chemistry in the rational synthesis of functional zirconium cluster-based MOFs

In the past two decades, reticular chemistry has developed into a powerful tool for the design and synthesis of porous, crystalline framework materials. The discovery of the first hexanuclear, zirconium cluster-based MOF (i.e. UiO-66; fcu net) led to a vast library of functional Zr-MOFs with various properties. The versatile connectivity of zirconium hexanuclear clusters and the adaptable tunability of organic ligands have resulted in the rational synthesis of a large set of Zr-MOFs based on edge-transitive nets; these nets commonly exists in crystalline network structures as suggested by reticular chemistry. In this review, we summarize recent advances in the synthesis of zirconium cluster-based MOFs in the light of reticular chemistry design principles. Isoreticular tuning of MOF parent structures and post-synthetic modification of Zr-MOFs for targeted applications are also deliberated.

Reticular chemistry↗
DOE OSTI2015

Phase transformation of oxide film in zirconium alloy in high temperature hydrogenated water

The effect of the variation of the dissolved hydrogen concentration on the oxide phase transformation under high-temperature hydrogenated water conditions was investigated using in situ Raman spectroscopy. The Raman spectrum in 50 cm(3)/kg of dissolved hydrogen concentration indicated the formation of monoclinic and tetragonal zirconium oxide at the water-substrate interface. As the dissolved hydrogen concentration decreased to 30 cm(3)/kg, the Raman peaks corresponding to the zirconium oxide phase changed, indicating an oxide phase transformation. And, the results of SEM and TEM analyses were compared with those of in situ analyses obtained for the oxide structure formed on the zirconium alloy.

Kim, Taeho↗
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↗