Showing papers on "Zirconium alloy published in 2013"
TL;DR: In this paper, high-energy synchrotron X-ray diffraction was used to characterize the oxides formed by autoclave exposure on Zr-Sn-Nb alloys with tin concentration ranging from 0.01 to 0.92.
117 citations
TL;DR: In this article, the viability of advanced oxidation-resistant Fe-base alloys to protect zirconium from rapid oxidation in high-temperature steam environments has been examined, and the applicability of these protective layers in light-water-reactor nuclear fuel structures is offered.
111 citations
TL;DR: In this article, a detailed residual stress and phase fraction analysis was carried out for the oxides formed on Zircaloy-4 after autoclave exposure at 360°C for various times by means of synchrotron X-ray diffraction.
101 citations
19 Apr 2013
TL;DR: Zirconium is a highly active metal and is resistant to corrosion by water, steam, mineral acids,alkalies, salts, etc. Zircon has been known as a gem material since biblical times.
Abstract: Zirconium occurs naturally as a silicate in zircon, the oxide baddeleyite, and in other oxide compounds. Zircon has been known as a gem material since biblical times. Zirconium is a highly active metal and is resistant to corrosion by water, steam, mineral acids,alkalies, salts, etc. Zirconium is highly refractory and the ore is cracked with only strong reagents and high temperatures. Various processes accomplish this, such as electric furnace, cautic fusion, chlorination, fluorosilicate fusion. Zirconium and hafnium always occur together in natural minerals and separation of the two are discussed. Zirconium is not considered a hazard to the environment. Zircon ceramics, opacifiers, refractories, foundry application are the leading end uses for zirconium. Other end uses include abrasives, chemicals, metal alloys, and welding rod coatings. There are many zirconium compounds in the form of hydrides, carbides, nitrides, oxides, halides, organometallics, chalcogenides, oxides, etc. Information is given on many of these compounds.
Keywords:
zirconium;
zircon;
zirconium compounds;
hafnium separation;
refractories
73 citations
TL;DR: ZrCo1−xNix (x = 0, 0.2 and 0.3) alloys were prepared and their hydrogen storage behavior were studied in this article, where a van't Hoff plot was constructed using plateau pressure data of each pressure-composition isotherm and the thermodynamic parameters were calculated for desorption of hydrogen in the ZrCo 1−x Nix-H2 systems.
70 citations
TL;DR: In this paper, the precipitate oxidation behavior of binary zirconium alloys containing 1.5% Fe, Ni, Cr or 0.6% Nb was investigated and the thermodynamics with the new approach presented could explain most precisely their behavior, considering precipitate stoichiometry and the free energy of oxidation of the constituting elements.
67 citations
TL;DR: In this article, the phase structure, tribological and corrosion properties were characterized by X-ray diffraction (XRD), pin-on-plate sliding wear test and potentiodynamic polarization corrosion test, respectively.
Abstract: In this study, zirconium oxide coatings were produced on Zr–2.5 wt.% Nb alloy by a plasma electrolytic oxidation (PEO) process using a pulsed DC power source with various frequencies 1000, 2000 and 4000 Hz. When a relatively low termination voltage (i.e. 400 V) was set, the thickness of the coatings was in a range of 7–15 μm and the treatment time to reach the preset voltage was less than 5 min. The coatings' phase structure, tribological and corrosion properties were characterized by X-ray diffraction (XRD), pin-on-plate sliding wear test and potentiodynamic polarization corrosion test, respectively. The results indicate that the thickest coating of 15 μm can be obtained by adopting the highest frequency pulsed DC, i.e., 4000 Hz and the coating has almost the same surface roughness as other two coatings. In spite of the difference in thickness, all the coatings showed a similar phase structure with the main phase of monoclinic ZrO2 and less tetragonal ZrO2 phase. Furthermore, all the PEO coated samples can improve wear and corrosion protection of the Zr alloy, among which the coating produced by the highest frequency (4000 Hz) achieved the highest corrosion polarization resistance, wear resistance and lowest friction coefficient under the testing conditions in this study.
56 citations
TL;DR: In this paper, the effects of solution-treatment temperature on the phase constitution, microstructure and magnetic properties of Sm(Cobal.Fe0.35Cu0.06Zr0.018)7 were investigated systematically.
Abstract: The effects of solution-treatment temperature on the phase constitution, microstructure and magnetic properties of Sm(Cobal.Fe0.35Cu0.06Zr0.018)7.8 were investigated systematically. The formation of Sm2Co7 around the Sm2O3 was confirmed and the volume fraction of Sm2Co7 increases with decreasing solution-treatment temperature. An area without cellular structure was observed around the Sm2Co7 phase in aged specimens solution treated at 1403 K. Thus, it was found that precipitation of Sm2Co7 phase prevents the formation of cellular structure during aging treatment. The coercivity of those magnets was less than 600 kA/m. On the other hand, the obvious formation of cellular structure was confirmed in the case of solution treatment at 1423 K and these magnets showed high coercivity of more than 1000 kA/m. The magnetic properties of Mr = 1.22 T and HcJ = 1011 kA/m have been obtained for Sm(Cobal.Fe0.35Cu0.06Zr0.018)7.8 by solution treatment at 1423 K.
53 citations
TL;DR: In this paper, the influence of zirconium doping on the piezoelectric properties and relaxor characteristics of BNT-6BZT bulk ceramics was investigated.
Abstract: This article details the influence of zirconium doping on the piezoelectric properties and relaxor characteristics of 94(Bi1/2Na1/2)TiO3–6Ba(ZrxTi1−x)O3 (BNT–6BZT) bulk ceramics. Neutron diffraction measurements of BNT–6BZT doped with 0%–15% Zr revealed an electric-field-induced transition of the average crystal structure from pseudo-cubic to rhombohedral/tetragonal symmetries across the entire compositional range. The addition of Zr up to 10% stabilizes this transition, resulting in saturated polarization hysteresis loops with a maximum polarization of 40 μC/cm2 at 5.5 kV/mm, while corresponding strain hysteresis measurements yield a maximum strain of 0.3%. With further Zr addition, the ferroelectric order is progressively destabilized and typical relaxor characteristics such as double peaks in the current density loops are observed. In the strain hysteresis, this destabilization leads to an increase of the maximum strain by 0.05%. These changes to the physical behavior caused by Zr addition are consistent with a reduction of the transition temperature TF-R, above which the field-induced transformation from the relaxor to ferroelectric state becomes reversible.
50 citations
TL;DR: In this article, the radiation-induced dislocation including and dislocation loops is reviewed in various temperature and fluence ranges, and their growth behavior is examined in the same way. But the authors focus on the characteristics of dislocation and growth in zirconium and its alloys.
49 citations
TL;DR: In this paper, high resolution secondary ion mass spectrometry (SIMS) analysis has been used to study the oxidation mechanisms when commercial low tin ZIRLO™1 and Zircaloy 4 materials are exposed to corroding environments containing both 18O and 2H isotopes.
TL;DR: Grain morphology and crystal structure of pre-transition oxides formed on a Zircaloy-4 alloy were investigated in this article, where small-sized tetragonal equiaxed grains were embedded in monoclinic columnar grains.
TL;DR: In this article, a finite element study is used to investigate the development of stresses in the oxide under the combined influence of molar volume expansion during oxide formation, metal/oxide interface geometry and metallic substrate creep.
TL;DR: In this paper, 2.2 MeV proton irradiation was performed on recrystallized zirconium alloys, used as structural materials for Pressurized Water Reactor fuel assemblies, which undergo stress-free growth which accelerates for high irradiation doses.
TL;DR: Investigating possible differences in the surface chemistry and/or surface topography of titanium and titanium-zirconium surfaces after sand blasting and acid etching revealed significant differences on the micro and nanoscale.
TL;DR: In this article, the initiation mechanism of nanoscale pores in the oxide films formed on zirconium alloys was investigated, and it was shown that these porosity is generated by the enrichment of Y at columnar oxide grain boundaries and subsequent unequal diffusion of Zr and Zr cations in oxide crystals.
TL;DR: In this paper, the effects of a nanocrystallisation treatment on high temperature oxidation of a zirconium alloy are specifically studied in order to show benefits of such a nanocystallisation on corrosion resistance and its influence on the mechanical fields.
TL;DR: In this article, a comparative study of cladding tubes with different surface conditions was performed to investigate their effect on the Zircaloy-4 substrate and oxide textures as well as the oxidation kinetic.
TL;DR: In this article, the effects of Al content (2.2 and 6.9 ) on structure and mechanical properties of the hot-rolled ZrTiAlV alloy samples were investigated.
TL;DR: In this article, a combined study based on a micromechanical phase-field model and a previously developed mean-field approach was proposed to analyze the influence of an applied stress on the precipitation of coherent hydrides in zirconium.
TL;DR: In this paper, micro-Raman imaging was used to investigate oxide scales formed on Zircaloy-4 and M5® alloys in air, in the 800-1,000°C temperature range.
Abstract: Micro-Raman imaging was used to investigate oxide scales formed on Zircaloy-4 and M5® alloys in air, in the 800–1,000 °C temperature range. To create the 2D spectral images, the data were processed by different ways. The results clearly show that a microscopic picture of the scales in terms of microstructure and internal stresses can be developed from Raman spectral maps at the micron scale. Data on the microstructure, crystallography, and composition, are presented. They confirm that the crystallographic phases observed for the Zircaloy-4 and M5® alloys are different, since, for Zircaloy-4, we clearly observed additional Raman signatures which most probably track the presence of nitrogen in the layers well before the occurrence of the kinetic transition. In particular, they show the presence of cubic zirconia in the layers, and strongly suggest the presence of zirconium nitride and oxynitride. Results also suggest the presence of strong stress gradients in the oxide scales.
TL;DR: In this article, a two-layer SiO2 scale was formed during exposure to high-pressure steam conditions, composed of a porous cristobalite layer above a thin, dense amorphous surface layer.
Abstract: Silicon carbide is a candidate cladding for fission power reactors that can potentially provide better accident tolerance than zirconium alloys. SiC has also been discussed as a host matrix for nuclear fuel. Chemical vapor–deposited silicon carbide specimens were exposed in 0.34–2.07 MPa steam at low gas velocity (~50 cm/min) and temperatures from 1000°C to 1300°C for 2–48 h. As previously observed at lower steam pressure of 0.15 MPa, a two-layer SiO2 scale was formed during exposure to these conditions, composed of a porous cristobalite layer above a thin, dense amorphous SiO2 surface layer. Growth of both layers depends on temperature, time, and steam pressure. A quantitative kinetics model is presented to describe the SiO2 scale growth, whereby the amorphous layer is formed through a diffusion process and linearly consumed by an amorphous to crystalline phase transition process. Paralinear kinetics of SiC recession were observed after exposure in 0.34 MPa steam at 1200°C within 48 h. High-pressure steam environments are seen to form very thick (10–100 μm) cristobalite SiO2 layers on CVD SiC even after relatively short-term exposures (several hours). The crystalline SiO2 layer and SiC recession rate significantly depend on steam pressure. Another model is presented to describe the SiC recession rate in terms of steam pressure when a linear phase transition kl governing the recession kinetics, whereby the reciprocal of recession rate is found to follow a negative unity steam pressure power law.
TL;DR: In situ neutron diffraction lattice strain measurements have been made during compressive loading tests on Zircaloy-2 at different temperatures as discussed by the authors, and experimental results show changes of the relative contributions of different deformation modes with increasing temperature.
Abstract: In situ neutron diffraction lattice strain measurements have been made during compressive loading tests on Zircaloy-2 at different temperatures The experimental results show changes of the relative contributions of different deformation modes with increasing temperature Among the operating deformation modes, basal slip is found to operate at a higher applied stress than prism slip at ambient conditions, but at a similar applied stress at elevated temperatures Twinning is found to be present at 573 K (300 °C), but is not observed at a temperature of 773 K (500 °C) at the loading rate used
TL;DR: In this article, the hydrogen absorption cracking behavior of Zr-4 has been investigated based on the microstructure, hydrogen absorption amount as well as residual stress, and the results show that the grain size has a pronounced effect on the hydrogen absorbing cracking, with the increase of grain size from 30μm to 500μm, the cracks vary from small cracks of independence to large cracks that connect with each other along the radial direction.
TL;DR: In this article, the hydride formation and its influence on the mechanical performance of hydrided Zircaloy-4 plates containing different hydrogen contents were studied at room temperature.
TL;DR: A review of the current understanding of the delayed hydride cracking behavior of zirconium alloy cladding tubes for fuel rods, focusing on the degradation mechanisms in high burnup fuel rods and transient loading scenarios, is presented in this paper.
Abstract: Zirconium alloy cladding tubes used in nuclear fuel rods are susceptible to delayed hydride cracking, which is a time dependent crack growth process resulting from the stress assisted diffusion of hydrogen to the crack tip, followed by the formation of radial hydrides and the subsequent fracture of the hydrides in the crack tip region. This article reviews the current understanding of the delayed hydride cracking behaviour of zirconium alloy cladding tubes for fuel rods, focusing on the degradation mechanisms in high burnup fuel rods and transient loading scenarios, which could potentially lead to substantial changes in the hydride microstructure and cladding failure by delayed hydride cracking following removal from the reactor and during storage and disposal of spent nuclear fuel rods in a waste repository. A brief summary of the general characteristics of delayed hydride cracking in zirconium alloy cladding is presented first. Relevant information on the cladding stresses under various usage conditions is then compiled and categorised into several characteristic stress transients that can be anticipated during reactor operation. Delayed hydride cracking in cladding tubes under stress transients is then examined under various temperatures, cooling rates, burnup levels and loading conditions.
TL;DR: Using density functional theory calculations, six special quasirandom structure cells are generated based on face centered cubic and face centered tetragonal unit cells to describe ZrH(2-x) (x = 0.25-0.5).
Abstract: The study of the structure and properties of zirconium hydrides is important for understanding the embrittlement of zirconium alloys used as cladding in light water nuclear reactors. Simulation of the defect processes is complicated due to the random distribution of the hydrogen atoms. We propose the use of the special quasirandom structure approach as a computationally efficient way to describe this random distribution. We have generated six special quasirandom structure cells based on face centered cubic and face centered tetragonal unit cells to describe ZrH2−x (x = 0.25–0.5). Using density functional theory calculations we investigate the mechanical properties, stability, and electronic structure of the alloys.
15 Jul 2013-Nuclear Instruments & Methods in Physics Research Section B-beam Interactions With Materials and Atoms
TL;DR: The 18th International Conference on Ion Beam Modifications of Materials (IBMM) as discussed by the authors was held in Qingdao, China from September 2-7, 2012, with a focus on ion beam modifications of materials.
Abstract: Conference Name:18th International Conference on Ion Beam Modifications of Materials (IBMM). Conference Address: Qingdao, PEOPLES R CHINA. Time:SEP 02-07, 2012.
TL;DR: In this paper, a hydroxylated grain boundary model decorated by various transition metal, TM, ions is employed to study the dependence of the hydrogen evolution reaction, HER, on the choice of TM ion and spin state along the hydride-proton recombination pathway.
Abstract: Zirconium alloys are widely used as cladding materials employed to contain the fission fuel in nuclear power plants. A limiting factor for fuel longevity is the corrosion property of the zirconium alloys. In the reactor, the main corrodent is water. The oxide forms thermodynamically during hydrogen evolution. Due to the corrosion mechanism, a fraction of the hydrogen is transferred to the alloy. It has long been known that the alloying elements actually control the hydrogen pick-up fraction, HPUF. A mechanism that explains these observations by means of density functional theory calculations is presented and validated. A hydroxylated grain boundary model decorated by various transition metal, TM, ions is employed to study the dependence of the hydrogen evolution reaction, HER, on the choice of TM ion and spin state along the hydride–proton recombination pathway. The efficiency of the system to utilize the overpotential for hydrogen evolution, originating from the overall corrosivity of the alloy, is found to be decisive for the HPUF. A dual origin of the detrimental effects of Co and Ni additives on the HPUF is identified.
01 Jan 2013
TL;DR: In this article, the authors highlight the various uses and properties of zirconium alloy cladding and structural components used in nuclear power light water reactors and discuss specific attributes including dimensional stability, corrosion resistance, irradiation effects and mechanical properties.
Abstract: This chapter highlights the various uses and properties of zirconium alloy cladding and structural components used in nuclear power light water reactors. Specific attributes including dimensional stability, corrosion resistance, irradiation effects and mechanical properties are discussed in detail.