Showing papers on "Zirconium alloy published in 2017"
TL;DR: In this article, isothermal isothermal oxidation tests with Zircaloy-4 in steam-nitrogen mixtures have been performed at 600, 800, 1000, and 1200°C using thermogravimetry.
49 citations
TL;DR: In this paper, the texture of Zirconium grains is controlled by twinning or prior texture, depending on the strain path and deformation level, and the texture can be controlled through nucleation, growth and texture control.
47 citations
TL;DR: In this article, Brouwer diagrams of the defect populations with respect to oxygen partial pressure were calculated and presented in the form of brouwerwer diagram, showing that low oxidation state Nb ions (2+ or 3+) charge balance the build-up of positive space charge in the oxide layer, increasing oxygen vacancy and electron mobility, leading to near-parabolic corrosion kinetics and a reduced hydrogen pick-up.
37 citations
TL;DR: The air oxidation behavior of zirconium-silicide coatings for three stoichiometries, namely Zr2Si, ZrSi, and Zr Si2, at 700°C has been investigated in this paper.
Abstract: The air oxidation behavior of zirconium-silicide coatings for three stoichiometries, namely, Zr2Si, ZrSi, and ZrSi2, at 700 °C has been investigated. These three coatings were deposited on a zirconium-alloy substrate using a magnetron sputter process at a low temperature. Argon gas pressure was observed to have a profound effect on the coating microstructure, with lower pressures favoring a denser and more protective microstructure. Coatings of ZrSi2 stoichiometry clearly showed superior oxidation resistance presumably due to the formation of a thin protective oxide layer, consisting of nanocrystalline SiO2 and ZrSiO4 in amorphous Zr-Si-O matrix. The thermal stability of the coatings was evaluated by annealing in an argon environment, and this also assisted in eliciting the effects of oxidation-induced inward Si migration. Thicker coatings of ZrSi2 were prepared and evaluated for oxidation resistance at 700 °C for longer exposure times, as well as at 1000 °C and 1200 °C. Once again the thin oxide layer provided for significant oxidation resistance. Pre-oxidizing the samples at 700 °C prior to 1000 °C and 1200 °C oxidation tests substantially reduced the extent of oxidation. Insights into the fundamental mechanisms of the oxidation behavior of zirconium-silicide coatings were obtained using a combination of scanning electron microscopy, X-ray diffraction, and X-ray photoelectron spectroscopy techniques. One potential application of these coatings is to enhance the oxidation resistance of zirconium-alloy fuel cladding in light water reactors under normal and accident conditions.
36 citations
TL;DR: In this article, an iron-chromium-aluminum (FeCrAl) cladding is proposed to replace zirconium alloy cladding in current commercial light water power reactors.
Abstract: After a tsunami caused plant black out at Fukushima, followed by hydrogen explosions, the US Department of Energy partnered with fuel vendors to study safer alternatives to the current UO2 -zirconium alloy system. This accident tolerant fuel alternative should better tolerate loss of cooling in the core for a considerably longer time while maintaining or improving the fuel performance during normal operation conditions. General electric, Oak ridge national laboratory, and their partners are proposing to replace zirconium alloy cladding in current commercial light water power reactors with an iron-chromium-aluminum (FeCrAl) cladding such as APMT or C26M. Extensive testing and evaluation is being conducted to determine the suitability of FeCrAl under normal operation conditions and under severe accident conditions. Results show that FeCrAl has excellent corrosion resistance under normal operation conditions and FeCrAl is several orders of magnitude more resistant than zirconium alloys to degradation by superheated steam under accident conditions, generating less heat of oxidation and lower amount of combustible hydrogen gas. Higher neutron absorption and tritium release effects can be minimized by design changes. The implementation of FeCrAl cladding is a near term solution to enhance the safety of the current fleet of commercial light water power reactors.
36 citations
TL;DR: In this article, the authors describe three-dimensional nodal kinetics simulations of a reactivity-initiated accident (RIA) in a representative state-of-the-art pressurized water reactor with both nuclear-grade iron-chromium-aluminum (FeCrAl) and silicon-carbide (SiC-SiC)-based cladding materials.
36 citations
TL;DR: In this paper, isotopic tracers and SIMS experiments were performed on oxides grown on Zircaloy-4 alloy under simulated primary water conditions and the oxide film in pre-transition is divided into two sub-layers, an external one, highly permeable to hydrogen and an inner protective one Thermal Desorption Spectroscopy analyses revealed two interaction sites of hydrogen, located in each oxide sublayer.
34 citations
TL;DR: In this paper, it is hypothesized that the corrosion acceleration is caused by the formation of hydrides, which leads to earlier loss of oxide protectiveness in the form of more frequent oxide kinetic transitions.
33 citations
TL;DR: In this article, the surface oxide film of a Zr−2.5Nb alloy subjected to long term corrosion at 633 K in simulated primary coolant of pressurized water reactors has been analyzed.
31 citations
TL;DR: In this paper, the US Department of Energy (DOE) is partnering with fuel vendors to develop enhanced accident tolerant nuclear fuels for Generation III water cooled reactors, which should better tolerate loss of cooling in the core for a considerably longer time while maintaining or improving the fuel performance during normal operation conditions.
Abstract: The US Department of Energy (DOE) is partnering with fuel vendors to develop enhanced accident tolerant nuclear fuels for Generation III water cooled reactors. In comparison with the standard current uranium dioxide and zirconium alloy system UO2-Zr), the proposed alternative accident tolerant fuel (ATF) should better tolerate loss of cooling in the core for a considerably longer time while maintaining or improving the fuel performance during normal operation conditions. General Electric, Oak Ridge National Laboratory and their partners have proposed to replace zirconium based alloy cladding in current commercial power reactors with an iron-chromium-aluminum (FeCrAl) alloy cladding such as APMT. The use of FeCrAl alloys will greatly reduce the risk of operating the power reactors to produce electricity.
30 citations
TL;DR: In-situ electrochemical impedance spectroscopy experiments were performed on Zircaloy-4 and Zr-2.5Nb alloys in 360°C water as mentioned in this paper.
TL;DR: In this paper, Chrobak et al. proposed a method to solve the Bykova problem in the context of nuclear power plants, which is the basis for this paper.
Abstract: I. Bykova,∗, C.P. Chrobak, T. Abrams, D.L. Rudakov, E.A. Unterberg, W.R. Wampler, E.M. Hollmann, R.A. Moyer, J.A. Boedo, B. Stahl, E.T. Hinson, J.H. Yu, C.J. Lasnier , M. Makowski , A.G. McLean University of California (San Diego), La Jolla, USA General Atomics, P.O. Box 85608, San Diego, CA, 92186-5608, USA Oak Ridge National Laboratory, Oak Ridge, TN, 37830, USA Sandia National Laboratories, Albuquerque, NM, and Livermore, CA, USA University of Wisconsin-Madison, Wisconsin, WI, 53706, USA Lawrence Livermore National Laboratory, Livermore, CA 94550, USA
TL;DR: In this article, the atomic simulations show that a strong repulsion exists only when the O atoms lie in the dislocation core and belong to the prismatic dislocation habit plane.
TL;DR: The thorough analytical and numerical treatment presented here quantifies the rich coupling between defect chemistry, thermodynamics and electrostatics which can be used to design and control oxide hetero-interfaces.
Abstract: We present a multi-scale approach to predict equilibrium defect concentrations across oxide/oxide hetero-interfaces. There are three factors that need to be taken into account simultaneously for computing defect redistribution around the hetero-interfaces: the variation of local bonding environment at the interface as epitomized in defect segregation energies, the band offset at the interface, and the equilibration of the chemical potentials of species and electrons via ionic and electronic drift-diffusion fluxes. By including these three factors from the level of first principles calculation, we build a continuum model for defect redistribution by concurrent solution of Poisson's equation for the electrostatic potential and the steady-state equilibrium drift-diffusion equation for each defect. This model solves for and preserves the continuity of the electric displacement field throughout the interfacial core zone and the extended space charge zones. We implement this computational framework to a model hetero-interface between the monoclinic zirconium oxide, m-ZrO2, and the chromium oxide Cr2O3. This interface forms upon the oxidation of zirconium alloys containing chromium secondary phase particles. The model explains the beneficial effect of the oxidized Cr particles on the corrosion and hydrogen resistance of Zr alloys. Under oxygen rich conditions, the ZrO2/Cr2O3 heterojunction depletes the oxygen vacancies and the sum of electrons and holes in the extended space charge zone in ZrO2. This reduces the transport of oxygen and electrons thorough ZrO2 and slows down the metal oxidation rate. The enrichment of free electrons in the space charge zone is expected to decrease the hydrogen uptake through ZrO2. Moreover, our analysis provides a clear anatomy of the components of interfacial electric properties; a zero-Kelvin defect-free contribution and a finite temperature defect contribution. The thorough analytical and numerical treatment presented here quantifies the rich coupling between defect chemistry, thermodynamics and electrostatics which can be used to design and control oxide hetero-interfaces.
TL;DR: In this article, it was shown that the main driving force for oxide texture development in single-phase zirconium alloys is the compressive stress caused by the Zr ZrO 2 transformation.
TL;DR: In this article, a novel mechanism of nodular corrosion related to Sn segregation at the interface between oxide and matrix (O/M) is proposed, which results from the anisotropy of Sn segregation formed on S N, S R and S T surfaces with different grain orientations.
TL;DR: In this paper, the authors measured Young's modulus and hardness for metal matrix of Zircaloy-4 cladding and δ-hydride embedded within its matrix using nanoindentation for the temperature range of 300 k −773 k.
TL;DR: In this article, a comprehensive first principles understanding of the oxidation of zirconium alloys by water was reiterated, and two channels were taken to jointly constitute to the oxidation process: one according to classical oxidation theory involving hydrogen evolution and the second reflected by inwards transport of protons causing hydrogen pick-up.
Abstract: A comprehensive first principles understanding of the oxidation of zirconium alloys by water was reiterated. Two channels were taken to jointly constitute to the oxidation process: one according to classical oxidation theory involving hydrogen evolution and the second reflected by inwards transport of protons causing hydrogen pick-up. The two were associated with charged and uncharged oxygen vacancies, respectively. The purpose of the present study was to clarify the nature of the effective anode during oxidation of zirconium as to the detailed role of the metal. Oxygen dissolution in the alloy resulted in a “pre-anodic” property associated with the formation of oxygen vacancy VO in the oxide, i.e., preceding VO2+/2e− separation. Atomistic perspective on the metal/oxide interface before nucleation of VO was provided. The rapid convergence of the model interface to bulk properties in spite of the local structural variability provided new insight as to the nature of an amorphous metal/oxide interface.
TL;DR: In this article, hot compression tests were performed to understand the deformation behavior of Zr-1Nb alloy in the temperature range of 700-1050°C, which envelopes α-phase, (α+β) phase, and β-phase.
Abstract: In nuclear water reactors, zirconium alloys are extensively used as fuel cladding material and in other structural applications. Uniaxial hot compression tests were performed to understand the deformation behavior of Zr-1Nb alloy. Therefore, hot compression tests were performed in the temperature range of 700-1050°C, which envelopes α-phase, (α+β) phase, and β-phase. True stress-strain curves, processing maps, microstructural observation and kinetic analysis were used to discuss the deformation behavior of Zr-1Nb alloy. Deformation at a strain rate of 10-2 s-1 reveals softening at lower temperatures and steady state behavior at higher temperatures. Processing map also reveals domain of high efficiency at 10-2 s-1 strain rate for a wide range of deformation temperatures. The flow softening and high power dissipation efficiency predicts dynamic recrystallization or dynamic recovery during the hot deformation of studied alloy.
TL;DR: In this paper, high-temperature steam oxidation experiments were performed at 1012-1207°C on Zr-1Nb-1Sn-0.1Fe fuel cladding tubes to study their weight gains and microstructural characteristics.
TL;DR: In this article, the deformation behaviors of cubic zirconia and a cubic ZIRconia thin film on top of an hcp Zirconium substrate were investigated using molecular dynamics nanoindentation simulation.
TL;DR: In this article, the microstructure of pristine U3Si2 and Zricaloy-4 interdiffusion products were examined using scanning electron microscopy (SEM) and TEM equipped with an energy dispersive X-ray spectroscopy (EDS) system.
TL;DR: In this article, the second phase particle changes and the influence of these changes on the microstructural evolution of the material during irradiation were investigated, and it was shown that the effect of both dose-rate and temperature on second phase particles behavior under irradiation and point out the complexity of iron rejection influence on the basal -component loops.
TL;DR: In this article, the radiation damage evolution in recrystallized Zircaloy-4 has been studied using charged particles irradiation using High Voltage Electron Microscope (HVEM) under 1 MeV electron irradiation at 673 and 723 K and 600 K ion irradiations at the same temperature.
TL;DR: In this paper, a series of uniaxial and multi-xial ratcheting tests were conducted on zirconium alloy tubes at 350°C, and the experimental results showed that the temperature significantly affects the properties of zirconsium alloy, and its ductility at 350 °C is poorer than that at room temperature.
TL;DR: In this article, the generalized stacking fault energies (GSFEs) of Zr-X binary alloys were computed using first-principles calculations, and the effects of atomic volumes, electronegativities, and first ionization energies of the alloying elements were also analyzed.
TL;DR: A detailed study on a Chernobyl “hot” particle collected from contaminated soil was performed as discussed by the authors, where optical and electron microscopy, as well as quantitative x-ray microbeam analysis methods were used to determine the properties of the sample.
TL;DR: This article used positron annihilation Doppler broadening spectroscopy to investigate the development of atomic-scale defect features in the oxides on Zircaloy-4 samples exposed to alkaline water at 350°C.
TL;DR: In this article, a phase field model is proposed to study the oxidation behavior of zirconium alloys at high temperatures, where the diffusion process is described by the time-dependent Cahn-Hilliard equation, and the elasto-plastic deformation is predicted using an elastic-perfectly plastic model and the Norton power law creep equation.