Relations of Microstructural Attributes and Strength-Ductility of Zirconium Alloys with Hydrides

As the first safety barrier of nuclear reactors,zirconium alloy cladding tubes have attracted extensive attention because of its good mechanical properties.The strength and ductility of zirconium alloy are of great significance to the service process of cladding tubes,while brittle hydrides precipitate and thus deteriorate the overall performance.Based on the cohesive finite element method,the effects of cohesive strength,interfacial characteristics,and hydrides geometric characteristics on the strength and ductility of two-phase material(zirconium alloy with hydrides)are numerically simulated.The results show that the fracture behavior is significantly affected by the cohesive strength and that the overall strength and ductility are sensitive to the cohesive strength of the zirconium alloy.Furthermore,the interface is revealed to have prominent effects on the overall fracture behavior.When the cohesive strength and fracture energy of the interface are higher than those of the hydride phase,fracture initiates in the hydrides,which is consistent with the experimental phenomena.In addition,it is found that the number density and arrangement of hydrides play important roles in the overall strength and ductility.Our simulation provides theoretical support for the performance analysis of hydrogenated zirconium alloys during nuclear reactor operation.

OUT-OF-PILE CORROSION RESISTANCE OF NEW ZIRCONIUM ALLOYS IN 500 ℃/10 3MPa STEAM

he out of pile corrosion resistance of new Zirconium alloys in 500℃/10 3MPa steam has been investigated and the effect of alloying elements, Sn,Nb,Fe,Cr on the property has been analyzed. The results show that the new alloys have better corrosion resistance than Zircaloy 4. That no nodular corrosion is found during test cycles shows the nodular corrosion resistance can be dramatically improved by addition of alloying elements Nb. The Fe/Cr ratio should be properly controlled if there is Cr addition in the alloys when developing new alloys.

The Influence of Different Contents of Bi Addition on the Corrosion Behavior of Various Zirconium-Based Alloys

Zr-4(Zr-1.5Sn-0.2Fe-0.1Cr,wt%), S5(Zr-0.8Sn-0.34Nb-0.39Fe-0.1Cr), T5(Zr-0.7Sn-1.07Nb-0.32Fe-0.08Cr) and Zr-1Nb were adopted to prepare Bi-containing zirconium alloys for systematically investigating the effect of Bi addition on the corrosion resistance of zirconium alloys. The specimens were corroded in superheated steam at 400℃/10.3 MPa, and in lithiated water with 0.01 M LiOH or in deionized water at 360℃/18.6 MPa by autoclave testing. Results show that the corrosion resistance increases with the increasing of Bi content dissolved in α-Zr. But the presence of Bi-con- taining second phase particles (SPPs) is unfavorable for the enhancement of corrosion resistance. This indicates that the Bi dissolved in α-Zr matrix plays an important role in improving the corrosion resistance, while the precipitation of the Bi-containing SPPs does harm to the corrosion resistance.

Microstructure of Bi-containing zirconium alloys

The effect of Bi addition(0.05 wt%-0.50 wt%)on the microstructure of zirconium alloys,including T5(Zr-0.7 S n-1.0 Nb-0.3 Fe-0.1 Cr),S5(Zr-0.8 Sn-0.35 Nb-0.4 Fe-0.1 Cr),Zr-4(Zr-1.5 Sn-0.2 Fe-0.1 Cr) and Zr-1 Nb,was investigated by transmission electron microscopy(TEM),energy-dispersive spectroscopy(EDS) and selected area electron diffraction(SAED).Results show that with the increase in Bi content,h-Zr(Fe,Cr,Nb)2,o-Zr(Fe,Sn,Bi)_(2) and Zr-Fe-Cr-Nb-Sn-Bi second-phase particles(SPPs) precipitate successively in the T5+xBi and S5+xBi alloys;in the Zr-4+xBi alloys,h-Zr(Fe,Cr)2,o-Zr(Fe,Sn,Bi)2,Zr-Fe-Cr-Sn-Bi and Zr-Fe-Cr-Bi SPPs are detected successively.While as for Zr-1 Nb+xBi alloys,Bi-free SPPs appear.The addition of Bi promotes the precipitation of SPPs with Sn in the alloys.The concentration of Bi dissolved in α-Zr matrix increases with the decrease in Sn content in the alloys.Adding reasonable Bi has little influence on the solid solution content of Nb in α-Zr matrix.

Corrosion behavior of Zr–Nb–Cr cladding alloys

Zr-Nb-Cr alloys were used to evaluate the effects of alloying elements Nb and Cr on corrosion behavior of zirconium alloys. The microstructures of both Zr substrates and oxide films formed on zirconium alloys were characterized. Corrosion tests reveal that the corro- sion resistance of ZrxNb0.1Cr （x = 0.2, 0.5, 0.8, 1.1; wt%） alloys is first improved and then decreased with the increase of the Nb content. The best corrosion resistance can be obtained when the Nb concentration in the Zr matrix is nearly at the equilibrium solution, which is closely responsible for the formation of columnar oxide grains with protective characteristics. The Cr addition degrades the corrosion resistance of the Zrl.lNb alloy, which is ascribed to Zr（Cr,Fe,Nb）2 precipitates with a much larger size than β-Nb.

Eutectic reaction and cored dendritic morphology in yttrium doped Zr-based amorphous alloys

The microalloying effect of yttrium on the crystallization behaviors of （Zr0.525Al0.10Ti0.05Cu0.179Ni0.146）100-xYx, and （Zr0.55Al0.15- Ni0.10Cu0.20）100-xYx （x=0, 0.4, and 1, thus the two alloy systems were denoted as Zr52.5, Zr52.5Y0.4, Zr52.5Y1, and Zr55, Zr55Y0.4, Zr55Y1, respectively） was studied. Transmission electron microscopy （TEM） results suggested that the crystalline phases were different in the two Zr-based alloys and with different yttrium contents. ZrNi-phase and Al3Zr5 phase precipitations can be well explained by the mechanisms of nucleation and growth. Al3Zr5 phase is mainly formed by a peritectic-like reaction, while ZrNi-phase by a eutectic reaction. The contents of elements Y, A1, and Ti may dominate the reaction types. The orientation relationship between Y203 particles and A13Zr5 phase is also discussed.

Formation and properties of high Fe-content Fe-(B-Si)-Zr bulk amorphous alloys

The present work is devoted to the development of Fe-（B-Si）-Zr amorphous alloys with high glass-forming ability and good magnetic properties. Using the cluster-plus-glue-atom model proposed for ideal amorphous structures, [FeFe11B3Si]（Fe1-xZrx） was determined as the cluster formula of Fe-（B-Si）-Zr alloys. The glass formation and thermal stability of the serial alloys, namely, [FeFel^B3Si]（Fel_xZrx） （x = 0, 0.1, 0.2, 0.3, 0.4, 0.5, 0.6, 0.75, and 1.0）, were studied by the combination of copper mold casting, X-ray diffraction, and differential thermal analysis techniques. The maxima of glass-forming ability and thermal stability were found to occur at the compositions of [FeFe11B3Si] （Fe0.6Zr0.4） and [FeFe11B3Si]（Fe0.5Zr0.5）. The alloys can be cast into amorphous rods with 1.5 ram diameter, and upon reheating, the amorphous alloys exhibit a large undercooled liquid span of 37 K. The saturation magnetization of the [FeFe11B3Si]（Fe0.5Zr0.5） amorphous alloy was measured to be 1.4 T.

EFFECT OF MELT OVER-HEATING AND ZIRCONIUM ALLOYING ON THE MORPHOLOGY OF Al_9FeNi PHASE AND MECHANICAL PROPERTIES OF 2618 ALLOY

The effect of melt over-heating on the morphology of Al_9FeNi phase in 2618aluminum alloy with high contents of Fe and Ni and 0.22 wt. percent zirconium has been investigatedby optical microscopy. SEM and TEM. The mechanical properties of 2618 aluminum alloy after hotextrusion and quenching/aging have been tested. The results show: melt over-heating treatment of2618 alloy with high contents of Fe and Ni at 960 deg C led to finer and better-distributedneedle-like Al_9FeNi phase in cast microstructure and fine Al_9FeNi particles after hot extrusion;the grain size of the alloy after hot extrusion could also be refined evidently by alloying ofzirconium; the ambient and high temperature tensile strength and elongation of 2618 alloy have beenapparently enhanced due to fine Al_9FeNi particles and dispersed Al_3Zr as well as fine grain size.

Interfacial reaction between zirconium alloy and graphite mold/yttrium oxide ceramic mold

Zirconium alloys are active in the molten state and tend to react with the mold during casting. The casting technology of zirconium is not yet well established; especially in selecting the mold materials, which are diffi cult to determine. In the present work, the interfacial reactions between zirconium casting and casting mold were studied. The zirconium alloy was melted in a vacuum arc skull furnace and then cast into the graphite mold and ceramic mold, respectively. The zirconium casting samples were characterized using SEM, EDS and XRD with an emphasis on the chemical diffusion of elements. A reaction layer was observed at the casting surface. Chemical analysis shows that chemical elements C, O and Y from the mold are diffused into the molten zirconium, and new phases, such as ZrC, Zr3O, YO1.335 and Y6ZrO11, are formed at the surface. In addition, an end product of zirconium valve cast in a yttria mold has a compact structure and good surface quality.

Fretting Wear Characteristics of Nuclear Fuel Cladding in High-Temperature Pressurized Water

In pressurized water reactor(PWR),fretting wear is one of the main causes of fuel assembly failure.Moreover,the operation condition of cladding is complex and harsh.A unique fretting damage test equipment was developed and tested to simulate the fretting damage evolution process of cladding in the PWR environment.It can simulate the fretting wear experiment of PWR under different temperatures(maximum temperature is 350℃),displacement amplitude,vibration frequency,and normal force.The fretting wear behavior of Zr-4 alloy under different temperature environments was tested.In addition,the evolution of wear scar morphology,profile,and wear volume was studied using an optical microscope(OM),scanning electron microscopy(SEM),and a 3D white light interferometer.Results show that higher water temperature evidently decreased the cladding wear volume,the wear mechanism of Zr-4 cladding changed from abrasive wear to adhesive wear and the formation of an oxide layer on the wear scar reduced the wear volume and maximum wear depth.

CO_2 selective hydrogenation to synthetic natural gas(SNG) over four nano-sized Ni/ZrO_2 samples:ZrO_2 crystalline phase & treatment impact

Two type zirconia (monoclinic and tetragonal phase ZrO2) carriers were synthesized via hydrothermal route, and nano-sized zirconia supported nickel catalysts were prepared by incipient impregnation then followed thermal treatment at 300?°C to 500?°C, for the CO2selective hydrogenation to synthetic natural gas (SNG). The catalysts were characterized by XRD, CO2-TPD-MS, XPS, TPSR (CH4, CO2) techniques. For comparison, the catalyst NZ-W-400 (monoclinic) synthesized in water solvent exhibited a better catalytic activity than the catalyst NZ-M-400 (tetragonal) prepared in methanol solvent. The catalyst NZ-W-400 displayed more H2absorbed sites, more basic sites and a lower temperature of initial CO2activation. Then, the thermal treatment of monoclinic ZrO2supported nickel precursor was manufactured at three temperature of 350, 400, 500?°C. The TPSR experiments displayed that there were the lower temperature for CO2activation and initial conversion (185?°C) as well as the lower peak temperature of CH4generation (318?°C), for the catalyst calcined at 500?°C. This sample contained the more basic sites and the higher catalytic activity, evidenced byCO2-TPD-MS and performance measurement. As for the NZ-W-350 sample, which exhibited the less basic sites and the lower catalytic activity, its initial temperature for CO2activation and conversion was higher (214?°C) as well as the higher peak temperature of CH4formation (382?°C). ? 2016 Science Press

CREEP BEHAVIOR OF Zr-1.5Nb-0.4Sn-0.1Fe-0.1Cu ALLOY

Creep behavior of the Zr-1.5Nb-0.4Sn-0.1Fe-0.1Cu alloy sheet is investigated from 300℃ to 400℃ in the stress range from 50 MPa to 180 MPa along the rolling direction. The measured strain rates range from 8.8 × 10^-10 s^-1 to 4.7 × 10^-7 s^-1. The activation energies are estimated to assess the creep deformation mechanisms in this alloy. The strain rate is slightly different at low stress, while it shows a distinct difference at high stresses. Stress exponents of this alloy increase with increasing applied stress at all testing temperatures. It is concluded that the creep deformation of the Zr-1.5Nb-0.4Sn-0.1Fe-0. 1 Cu alloy is controlled by the diffusion creep at low stress region and by the climbing of dislocations at high stress region.

Electroplating of Zr-Ti Alloy

The electrodeposition of Zr-Ti alloy has been studied in molten fluoride system consisting of fluorides of both zirconium and titanium dissolved in FLINAK (KF:NaF:LiF = 42.0:11.5:46.5 mol). A coherent dense deposit of the alloy with a smooth surface is obtained on different material substrates. X-ray diffraction has demonstrated that the alloy deposit is a Ti-Zr solid solution with Ti mainly and its preferred orientation is strongly along Ti[110] direction. The properties of the deposit are better than those of a pure zirconium coating. The optimum operating parameters are a temperature of about 800°C, a current density less than 40mA/cm2, a solution about 15wt% ZrF4 and 15wt% K2TiF6 in FLINAK as well as pure zirconium as anode material.

Effects of Zr and Y on the Mechanical Properties and the Microstructure of Ti-Al-Sn-Nb-Mo-Si Alloy

In the present paper the effects of additions of Zr and Y on the microstructure and mechanical properties for Ti-(6.0 approximately 6.5)Al-(2.0 approximately 3.0)Sn-(1.5 approximately 6.0)Zr-(0.8 approximately 1.0)Mo-1.0Nb-0.25Si alloys are reported. The experimental results shows that: with increasing of Zr content, tensile strength and creep resistance of the alloys increase, and reduction in area and thermal stability of the alloys decrease. Decrease in thermal stability of the alloys mainly caused by surface thermal unstability. After heat treatment Y addition can make grain size of the alloys refine. The reduction in area and thermal stability of the alloys with Y addition are improved, and tensile strength slightly decreases and creep resistance is essentially the same as the alloy without Y addition. These phenomena are explained in brief.

Oxide formation mechanism of a corrosion-resistant CZ1 zirconium alloy

Tailoring microstructure and microchemistry by altering elemental compositions and thermomechanical treatment parameters enables superior corrosion performance in zirconium alloys for nuclear applications.However,our understanding of the relationship between various defects and the corrosion process remains limited in the newly developed zirconium alloys.Here we report the oxide formation mechanism of a CZ1 zirconium alloy with corrosion resistance surpassing many other zirconium alloy systems,such as Zircaloy-4 and Zr-1Nb-1Sn alloys.Autoclave experiments of CZ1 alloy and Zr1Nb-1Sn model alloy were performed in 360°C water for up to 820 d.We quantitively determined oxide phases by transmission Kikuchi diffraction(TKD)and examined lateral cracks,nano-porosity,and second-phase particles in oxide scales by transmission electron microscopy(TEM).Compared to the Zr-1Nb-1Sn model alloy,CZ1 alloy with lower Nb and Sn concentrations has shown smaller and lower-density lateral cracks but slightly larger oxide grains,reducing the diffusion route for oxidating species.Using analytical scanning and transmission electron microscopy,we demonstrate that due to the lower content of Sn(∼0.9 wt.%),there is less tetragonal ZrO_(2) phase formed in the oxide,and the level of tetragonal to the monoclinic phase transition is reduced.Although the Nb content(0.1 wt.%–0.3 wt.%)is lower than the solid solution limit of Nb in Zr,by introducing minor elements such as Fe,Cr,and Cu,there are still a reasonable number of second-phase particles to relieve the high stress associated with the metal-to-oxide transformation.These mechanisms have substantially changed the density and distribution of lateral cracks in the oxide,thus reducing the corrosion rate of zirconium alloys.

Microstructure and properties of as-cast Zr-2.5Nb-1X(X=Ru,Mo,Ta and Si) alloys for biomedical application

The micro structure and properties of as-cast Zr-2.5Nb-1X(X=Ru,Mo,Ta and Si) alloy are screened to explore novel biomedical zirconium alloys for magnetic resonance applications.Corresponding micro structure and phase transformation were characterized using X-ray diffraction(XRD),scanning electron microscope(SEM)and transmission electron microscope(TEM).Hardness test,magnetic detection and electrochemical corrosion measurements are taken to present properties.The results show that all alloys consist of α-Zr,β-Zr and ω-Zr.α-Zr and β-Zr mainly exist in the form of parallel and intersecting plates,and nanoscale ω-Zr is dispersed in β-Zr plate.Especially,blocky ω-Zr with needle-like α-Zr is only found in plate-free blocks of Zr-2.5Nb-1Mo/Ru alloy.The orientation relationship(OR) between α-Zr and ω-Zr follows [1120]_α//[1101]_ω and(0001)_α//([1011]_ω 011)_ω.Combining this OR with the OR between β-Zr and ω-Zr,the transformation relationship between β-Zr/ω-Zr and α-Zr is also discussed.Zr-2.5Nb-1Ru alloy with high corrosion potential(-0.500 V),low corrosion rate(0.949 μm·year^(-1)) and low magnetic susceptibility(92×10^(-6)) shows great potential to be a novel biomedical implant with magnetic resonance imaging compatibility.Based on the experimental results,the possible relationship among alloying elements,micro structure and properties has been established in these Zr-2.5Nb-1X alloys.

Effect of Normal Force on Fretting Wear Behavior of Zirconium Alloy Tube in Simulated Primary Water of PWR

The effect of normal force on fretting wear behavior of zirconium alloy tube mated with grid dimple in simulated primary water of pressurized water reactor nuclear power plant was investigated.Results showed that the maximum wear depth,wear volume and wear coefficient of Zr alloy tube in simulated primary water at 315℃ gradually increased with increasing normal force,while the friction coefficient gradually decreased.Fretting process could be divided into four stages according to the variation of friction coefficient during test.When normal force exceeds 30 N,the fretting regime would transition from gross slip regime to partial slip regime after 3×10^(7 )cycles.Delamination was aggravated with increasing normal force,while abrasive wear became slighter.A thicker third-body layer with monoclinic ZrO_(2) was formed by the tribo-sintering mechanism under higher normal force.In addition,the schematic evolution processes of delamination and third-body layer formation were displayed according to morphology observation.

Optimization of N18 Zirconium Alloy for Fuel Cladding of Water Reactors

In order to optimize the microstructure and composition of N18 zirconium alloy （Zr-1Sn-0.35Nb-0.35Fe-0.1Cr, in mass fraction, %）, which was developed in China in 1990s, the effect of microstructure and composition variation on the corrosion resistance of the N18 alloy has been investigated. The autoclave corrosion tests were carried out in super heated steam at 400 ~C/10.3 MPa, in deionized water or lithiated water with 0.01 mol/L LiOH at 360 ~C/18.6 MPa. The exposure time lasted for 300-550 days according to the test temperature. The results show that the microstructure with a fine and uniform distribution of second phase particles （SPPs）, and the decrease of Sn content from 1% （in mass fraction, the same as follows） to 0.8% are of benefit to improving the corrosion resistance; It is detrimental to the corrosion resistance if no Cr addition. The addition of Nb content with upper limit （0.35%） is beneficial to improving the corrosion resistance. The addition of Cu less than 0.1% shows no remarkable influence upon the corrosion resistance for N18 alloy. Comparing the corrosion resistance of the optimized N18 with other commercial zirconium alloys, such as Zircaloy-4, ZIRLO, E635 and Ell0, the former shows superior corrosion resistance in all autoclave testing conditions mentioned above. Although the data of the corrosion resistance as fuel cladding for high burn-up has not been obtained yet, it is believed that the optimized N18 alloy is promising for the candidate of fuel cladding materials as high burn-up fuel assemblies. Based on the theory that the microstructural evolution of oxide layer during corrosion process will affect the corrosion resistance of zirconium alloys, the improvement of corrosion resistance of the N18 alloy by obtaining the microstructure with nano-size and uniform distribution of SPPs, and by decreasing the content of Sn and maintaining the content of Cr is discussed.

Double glow plasma carburization on zirconium to improve surface hardness and wear resistance

Though some important progress in the excel- lent mechanical properties of zirconium alloys have been reported, their high surface hardness and good wear prop- erties need to be explored further. In this work, a carbur- ized layer was formed on the surface of commercially pure zirconium by a double glow plasma hydrogen-free car- burizing technique. Commercial high-purity graphite was used as the carbon source material. X-ray diffraction （XRD）, scanning electron microscopy （SEM）, energy dis- persive spectroscopy （EDS）, Vickers hardness test, friction and wear test were used to characterize the samples car- burized. The carburized layer could be clearly observed under a microscope. XRD patterns indicate that the zirco- nium carbide phase is formed in the carburized layer. The surface hardness of the sample increases significantly after carburization. Friction and wear tests results show that wear resistance and friction coefficient of zirconium are improved considerably after carburization. Surface plastic deformation is arrested to a low extent in contrast with pure zirconium because of the presence of ZrC phases during the wear test. The results may provide new insight into methods for surface strengthening of zirconium alloys.

Transmission electron microscopy characterization of dislocation loops in irradiated zirconium

Characterization of irradiation defects is of great importanceto mitigate irradiation damage,reduce irradiation growth and tune mechanical properties in Zr alloys.Here,we describe a practical method to characterize the dislocation loops in irradiated Zr using conventional transmission electron microscopy(TEM).Vacancy or interstitial nature of dislocation loops is determined using the inside and outside contrast method.The habit plane of dislocation loops is determined by tilting the sample to multiple zone axes and judged based on the projected loop shape.The size ofloops is measured by tilting the sample to an edge-on position and the loop number is counted under a weak-beam dark-field TEM condition.<c>loops have a line contrast under viewing direction of a-axis and a circular shape under viewing direction of c-axis.In addition,a large number of triangle-shaped vacancy platelets(TVPs)were formed on the basal plane.With increasing the irradiation damage from 0.5 to 1.5 dpa,the number density ofloops keeps constant,while the number density of TVPs increased significantly,owing to the anisotropic diffusion and accumulation of point defects within basal plane.The methods introduced here are easy to follow and extend into other related investigations.