Zircaloy Research Articles

Vacuum sealing-assisted processing of titanium–zirconium alloys: synthesis, microstructure, hardness, friction, wear, and corrosion studies for biomedical application

Vacuum sealing-assisted processing of titanium–zirconium alloys: synthesis, microstructure, hardness, friction, wear, and corrosion studies for biomedical application

On the Construction of a Calibration Curve Using a Glow Discharge Emission Spectrometer for Measuring the Hydrogen Content in Zirconium Alloys

One of the main directions of modern activity of metrology researchers is to provide the real sector of the economy of the Russian Federation with reference materials that have no analogues in the country. The article discusses the research of highly efficient materials for fuel cladding (fuel elements) in the active zones of thermal reactors, the scope of which is wide. The purpose of the study is the development of samples of the composition of the zirconium alloy to establish and control the stability of the calibration curves of spectrometers when determining the mass fraction of hydrogen, provided that the metrological and technical characteristics of the reference material meet the requirements of the measurement procedure. The main methods for determining the hydrogen content in materials are analyzed. It has been established that the method of glow discharge optical emission spectroscopy has found the widest application. It has also been established that zirconium alloys are chosen as the main material in most cases.The need to create samples for constructing a calibration curve for spectrometers for measuring the hydrogen content in a zirconium alloy is noted. Samples of zirconium alloy Zr-1Nb (grade Э110) have been developed to construct a calibration curve using a glow discharge emission spectrometer for measurements with a mass fraction of hydrogen from 0.034 to 0.498 %. Calibration is carried out using the developed samples to obtain a calibration curve using the example of an emission glow discharge spectrometer of the GD Profiler2 type. The relative error of the hydrogen mass fraction obtained during calibration does not exceed ± 10 %.The practical significance of the study lies in the development of samples that can be used to calibrate spectrometers based on the method of glow discharge optical emission spectroscopy.

Tribological and corrosion behavior of hydrided zirconium alloy

Zirconium alloys (Zr-alloy) are used for the core components of nuclear reactors due to their exceptional properties, which make them suitable for the given applications such as low neutron absorption cross-section, good corrosion resistance, good wear resistance, good mechanical stability at high temperatures etc. The coolant water that flows inside the pressure tube contains LiOH to control the pH value of the coolant water, which can cause corrosion. The flow of coolant water produces low amplitude vibration, which can cause fretting wear of the components. Further, the coolant water at high temperature causes the hydride formation in the material, degrading the properties of zirconium alloys. The safe working of nuclear reactors is the prime objective, and the corrosion and wear behavior of the core components must be explored to prevent possible failure. Fretting wear or tribo-corrosion experiments were conducted on Zr-2.5Nb alloy against SS-410 for varied fretting duration to understand the tribological response. Electrochemical corrosion experiments were performed under the aqueous solution of LiOH (10.4 pH) using Electrochemical Impedance Spectroscopy (EIS) and potentiodynamic polarization technique. The comparison was made between the tribological and corrosion response of unhydrided-Zr-2.5Nb alloy (Zr-2.5Nb) and hydrided- Zr-2.5Nb alloy (Zr-2.5Nb-H). The hydride formation causes the reduction in corrosion resistance of Zr-2.5Nb alloy. The oxide layer is formed during the wear and corrosion process which protects the material from further wear and corrosion respectively.

Effect of processing temperature on interface microstructure of diffusion welded joint of super-duplex stainless steel and zirconium alloy with nickel alloy interlayer

In the present study, super-duplex stainless steel (SDSS) and zirconium alloy (ZrA) was vacuum welded using nickel alloy (NiA) interlayer by the diffusion welding process. The welded joints were processed from 800 to 950 °C in steps of 50 °C using 4-MPa compressive load for 75 min. The joints were characterised using optical microscopy, scanning electron microscopy, electron prob micro-analyser, and x-ray diffractometer. The SDSS│NiA interface is planar in nature for all welding temperatures; however, layer-wise Ni5Zr, Ni10Zr7, NiZr, and NiZr2 reaction products were observed at the NiA│ZrA interfaces. The reaction products and width of these reaction products increases with welding temperature. At both the interfaces, hardness values gradually increases with increase in welding temperature. NiA│ZrA interface exhibited higher hardness as compared to the SDSS│NiA interface due to the presence of layer wise intermetallics. Joints reached a maximum tensile strength of ∼370 MPa (with an elongation of ∼7.9%) when the joints were processed at 900 °C due to optimal reaction zone width and the least embrittling effect of the respective intermetallics. At lower temperatures, fracture surfaces of welded assemblies showed featureless morphology and voids present due to the poor coalescence of mating surfaces. At higher welding temperatures, the fracture surface of the transition joint preferred brittle fracture mode due to the cleavage planes with different alignments with insignificant volume fraction of voids on the fractography.

A novel nano-structured zirconium alloy prospective for high temperature applications

Zirconium alloys have many advantages for structural applications in nuclear reactors, but their use has been generally restricted at temperatures below 400°C. In this study, a novel nano-structured Zr alloy was successfully fabricated by ball milling and hot-isostatic-pressing, using Zr, Sn, ZrH2 and nano-Y2O3 as raw powders. The alloy microstructure was characterized to be uniquely featured with a very fine grain structure (mean size: ∼3 μm) and a high number density of highly-coherent, ultra-fine nano-precipitates (mean size: 4.2nm, density: ∼2.52×1023m−3). These nano-precipitates were identified as Y2Zr2O7, with the interfacial orientation relation of 001(010)Y2Zr2O7//000121̅1̅0Zr. This new alloy achieved remarkably-improved strength and heat resistance from room temperature up to 600°C for at least 1000h, which is superior to all the current commercial Zr alloys, especially to Zr-2 and Zr-4 alloys with a more comparable matrix composition.

EXPERIMENTAL RESEARCH CONCERNING WATERSIDE CORROSION OF FUEL CLADDING IN CANDU REACTOR

EXPERIMENTAL RESEARCH CONCERNING WATERSIDE CORROSION OF FUEL CLADDING IN CANDU REACTOR

Impact of ECAP at 300 °C on the microstructure and mechanical properties of the quenched Zr–2.5%Nb alloy

We investigated the microstructure of the Zr–2.5%Nb zirconium alloy after subjecting it to equal-channel angular pressing (ECAP) and found that ECAP at 300 °C increases the strength by 140 to 180 %. Notably, unlike other studies, our alloy did not show complete dissolution of niobium particles, which may be due to the reduced diffusion rates at the lower deformation temperature of 300 °C. Pre-treatment involving quenching before severe plastic deformation was also studied, which developed a lamellar structure introducing additional boundaries that facilitated grain refinement during subsequent ECAP. The strength of the alloy was further enhanced by solid-solution hardening, achieved through the complete dissolution of the Nb particles into the matrix post-quenching. This process resulted in a 2.3-fold increase in yield strength after quenching plus ECAP compared to the initial coarse-grained state.

Simulation of nuclear fuel clad high-temperature ballooning under loss-of-coolant accident conditions considering anisotropic creep

Zirconium (Zr) alloys are standard clad material in commercial light water reactors. However, it is vulnerable to ballooning and burst under high-temperature conditions such as LOCA (Loss-Of-Coolant Accident), due to large internal/external pressure difference. Considering that clad ballooning is detrimental in terms of nuclear reactor safety especially for high-burnup fuel rod, high-fidelity numerical simulation of Zircaloy clad ballooning is desirable. This paper presented a finite-element based implementation for the simulation of Zircaloy clad ballooning. The developed simulation capability predicts progressive clad ballooning up to failure given empirical burst limits. The material models coupled together are phase transformation and high-temperature creep. Especially, anisotropic creep was considered for α phase Zircaloy. The implemented algorithm for stress update is implicit in time integration, which allows relatively large time step size to efficiently simulate the process that can last for hundreds of seconds. An analytic benchmark, the PUZRY ballooning and burst test, and the Hardy stress-rupture test were utilized to verify and validate the simulation capability. The anisotropic creep calculation conforms to the analytical solution. The comparison with PUZRY and Hardy test data shows that considering anisotropic creep in the α phase region is essential. 3D ballooning was also simulated by considering circumferential temperature difference for both conventional and novel annular clad designs. Interestingly, it is the first time to note that the risk of buckling failure is predicted for the double-cooled annular nuclear fuel clad during the accident condition, only after considering the anisotropic creep algorithm.

Heat of transport for hydrogen and hydrides in Zircaloy

ABSTRACT Hydrogen diffuses in zirconium alloys in response to gradients in hydrogen concentration, temperature, and stress. This essay discusses the results of several evaluations of the coefficient describing the effects of temperature gradients called the heat of transport. The values distinguish between when hydrogen is in solution, (25 ± 3) kJ/mol, and when stable hydrides are present, (116 ± 17) kJ/mol.

Comprehensive integrity assessment of spent nuclear fuel cladding during normal and postulated conditions of transportation

As a part of Spent Nuclear Fuel (SNF) management, it is important to maintain the cladding integrity to prevent the potential release of hazardous radioactive materials. Accordingly, assessing the integrity of SNF under the generated shock and vibration loads during normal conditions is essential for safe transportation. In this study, failure criteria for two zirconium alloys were derived by utilizing data from tensile property tests, mechanical model calculations, and machine learning estimations. Integrity assessment under normal and postulated transportation conditions was also performed via finite element analyses. Loading conditions were generated by real time-acceleration data collected during transportation tests in the Republic of Korea. Shock and vibration loads were applied to a simplified light-water reactor assembly. Furthermore, postulated transportation conditions and pre-cracked cladding were assumed for more challenging conditions. Consequently, a comprehensive integrity assessment of SNF cladding under shock and vibration loads for both free-defect and defective (pre-cracked) cladding was conducted. In each case, the resultant maximum principal strains, stress intensity amplitudes, and stress intensity factors were significantly smaller than the strain-based failure criteria, S-N curve, and fracture toughness, respectively. Therefore, the integrity of SNF cladding under transportation conditions remained intact even when assuming further challenging postulated transportation load conditions.

Porous TiZr alloys by dealloying of Ti20Zr20Cu60 precursor in Mg melt for biomedical applications

This research aims to develop porous titanium–zirconium alloy for implant applications by the liquid metal dealloying method. The results indicated the formation of a homogeneous three-dimensional porous morphology with a density of 3 ± 0.3 g/cm3. The XRD analysis shows that the porous titanium–zirconium alloy is a single-phase hcp solid solution. The unique porous scaffold can be introduced as a promising candidate for dental implants owing to bone-like elastic modulus of 3.5 GPa and compressive strength of 550 MPa. The Ashby-type diagram confirms the achievement of new biomaterial with a good combination of high strength and low stiffness.

Insight into the pitting corrosion behavior of laser‐welded R60702 zirconium alloy in chloride electrolyte

AbstractThis work aims to analyze the passivation and pitting corrosion behaviors of laser beam welded R60702 zirconium alloy in neutral and acidic chloride‐containing electrolytes. Potentiodynamic polarization and electrochemical impedance spectroscopy measurements are carried out to investigate the electrochemical performance of the welding joint. Scanning electron microscope, X‐ray diffraction, and three‐dimensional profile digital microscope are utilized to reveal the microstructures and corrosion morphologies. The electrochemical results show that the heat‐affected zone (HAZ) and weld zone have nearly equal corrosion performance, and both of them are more corrosion‐resistant than the base metal. The corrosion morphologies suggest that the HAZ has the lowest sensitivity to pitting corrosion. Moreover, it is unraveled that, in chloride‐containing electrolytes, the quantity and distribution of Zr(Fe,Cr)2 particle phases are the main factors that determine the different corrosion performances of the welding joint.

An Overview of Mechanisms of the Degradation of Promising ATF Cladding Materials During Oxidation at High Temperatures

AbstractAccident tolerant fuel (ATF) cladding is a new type of nuclear fuel cladding designed to improve the safety and performance of nuclear reactors. In this paper, the kinetics and degradation mechanisms during high-temperature oxidation in steam of the three most promising ATF cladding materials, i.e., chromium-coated zirconium alloys, FeCrAl alloys, and silicon carbide-based composites, are described. Each system has its own degradation mechanisms leading to different maximum survival temperatures. After providing general information and data to understand the oxidation and degradation processes, illustrative examples obtained at the Karlsruhe Institute of Technology are given for each type of cladding. The maximum temperatures at which the barrier effect of the cladding can be maintained for a reasonable period of time during nuclear accident scenarios are 1200–1300 °C for Cr-coated Zr alloys, 1400 °C for FeCrAl alloys, and 1700 °C for SiC-based composite claddings.

Fretting corrosion running regime and damage mechanism evolution of zirconium alloy mated with Inconel 718 alloy in high temperature high pressure water environment

Fretting corrosion behavior and damage mechanism of Zr-4 alloy mated with Inconel 718 alloy under different fretting regimes in high temperature high pressure water were studied. The corresponding running condition fretting corrosion map (RCFM) and material response fretting corrosion map (MRFM) were established. The results showed that the change of fretting regime led to the differences of fretting corrosion behavior and damage mechanism. Based on the damage mechanism, the MRFM could be divided into partial slip regime with only slight surface damage; mixed fretting regime-I dominated by cracks; mixed fretting regime-II dominated by cracks and abrasive wear; and gross slip regime dominated by severe adhesive wear and oxidation. The transformation of fretting regime also resulted in different oxides.

High-temperature steam oxidation experiment of molten zirconium alloy

In this study, we investigate the steam oxidation behaviors of molten zirconium alloy at exceptionally high temperatures of 1870 ℃ and 1970 ℃. A specialized high-temperature steam oxidation apparatus was established to conduct the steam oxidation experiment. The experiments yielded crucial data including oxidation weight gain results, macroscopic appearance of the oxide surface, and section morphology. Several new phenomena: α-Zr(O) precipitates and two-layer oxide structure, were observed in this study. Based on these observations, a complete mechanism of molten zirconium oxidation in steam was developed, providing an interpretive framework for the oxidation observed in this experiment. This newly proposed mechanism reevaluates the formation process of nonceramic precipitates in the molten bulk and its consequential impact on oxidation kinetics at temperatures below 2000 ℃.

The Corrosion and Mechanical Behavior of Zirconium Alloy for Alkali Fusion Process at High Temperature

The Corrosion and Mechanical Behavior of Zirconium Alloy for Alkali Fusion Process at High Temperature

Experimental Study on Coefficient of Restitution of Small-Sized Spherical Particles during Low-Speed Impact

This study presents experimental investigations on the normal restitution coefficients of a titanium bead (Ti), zirconia bead (ZrO2), and amorphous zirconium alloy sphere (Amor). The research explores the influence of particle diameter and collision velocity on the normal restitution coefficient between two independent, identical spherical particles of different materials. The experimental findings demonstrate that increasing the particle diameter results in more effective plastic deformation, leading to higher energy losses and, subsequently, smaller coefficients of restitution. Similarly, higher particle velocities cause more energy dissipation during collisions, resulting in smaller restitution coefficients. Comparing particles of different materials, those with larger yield strengths exhibit more elastic behavior, experience less initial energy loss due to deformation, and reach the maximum restitution coefficient (elastic state) with fewer collisions. This finding suggests that material properties significantly influence the overall energy dissipation and elastic response in the particles. To validate the experimental results, existing models are compared and discussed. Furthermore, potential physical mechanisms responsible for the observed behavior are explored, providing valuable insights into the collision dynamics in spherical particle interactions. Overall, this study contributes to a better understanding of the factors affecting the normal restitution coefficient in particle collisions, enabling the design and optimization of particle systems for diverse applications in condensed matter and related fields.

A comparative study on the corrosion of pure titanium and titanium–12%zirconium alloy after different exposure periods of time in sodium chloride solution

In this work, the powder metallurgy technique was employed to manufacture pure titanium (Ti) and 88% titanium–12% zirconium (TiZr) alloy. The electrochemical corrosion investigations for pure Ti and the TiZr alloy were carried out after exposure for 30 min and 3 days in 3.5% NaCl solutions. The Nyquist and Bode plots obtained from the electrochemical impedance spectroscopy experiments revealed that the presence of Zr remarkably magnifies the corrosion resistance of Ti via increasing the impedance and degree of the phase angle, as well as the polarization and solution resistances. The potentiodynamic cyclic polarization measurements revealed that the presence of 12% Zr highly enhances the corrosion resistance of Ti. These polarization results showed that Zr addition reduces the corrosion of Ti via decreasing its corrosion rate. The intensity of the current when measured with increasing time of the experiment at −0.10 mV (Ag/AgCl) indicated that the addition of 12% Zr greatly decreases the absolute current, which indicates that alloying Zr within Ti reduces the severity of its corrosion in the chloride electrolyte. The morphology of the surfaces and the possible surface layer(s) for the corroded Ti and TiZr samples were analyzed using a scanning electron microscope and energy dispersive x rays. Results collectively depicted that the presence of Zr increases the corrosion resistance when alloyed with Ti.

In-situ TEM investigation of zirconium alloy under Kr+ single-beam and Kr+-He+ dual-beam synergetic irradiation

The in-situ TEM irradiation experiments of zirconium alloy were conducted at 573 K, 673 K, and 773 K utilizing a 400 keV Kr + single beam and a 400 keV Kr+ and 30 keV He + dual beam. The results show that a large number of dislocation loops have been characterized in the matrix of the zirconium alloy under irradiation. With increasing the irradiation damage dose, some dislocation loops have reacted with one another to form a larger dislocation loop, which has finally formed dislocation lines or other defect structures. In zirconium alloys irradiated with Kr + single beam and Kr+ and He + dual-beam radiation, the proportion of <a> type dislocation loops with different Burgers vectors is essentially the same at low damage doses, but the proportion of interstitial type dislocation loops with the same Burgers vectors is obviously different. The amorphization of the second phase and the dissolution of the small-sized second phase were also pointed out. With the increase in temperature, the density of the dislocation loop in zirconium alloy gradually decreases, and the size of dislocation loop first increases and then decreases. Kr+ and He + dual beam irradiation increases the size of dislocation loops but decreases their density as compared with Kr + single beam irradiation.

Crystal plasticity approach for predicting mechanical responses in wire-arc directed energy deposition of NbZr1 refractory alloy

The refractory alloy, such as Niobium-1 wt% Zirconium (NbZr1) alloy, is a promising candidate for diverse applications in extreme environments. This study investigates an integrated experimental and computational methodology that can predict the mechanical properties of wire-arc directed energy deposited (DED) refractory NbZr1 alloy. The effect of large columnar grains with strong texture induced by the wire-DED process has been successfully incorporated into the representative volume element (RVE). Performing crystal plasticity (CP) simulation on RVE of NbZr1, deformation behavior and stress-strain relationship are predicted to explore the process-structure-property (PSP) relationship. The RVE has been created in synthetic microstructure generation software utilizing the grain statistics derived from electron backscatter diffraction (EBSD) analysis, capturing essential microstructural features. A phenomenological constitutive model and appropriate boundary conditions have been applied and solved, using the CP fast Fourier transform method. The CP model has been calibrated and validated utilizing the tensile test data along the build direction and deposition direction, respectively. The error in predicting yield strength and ultimate tensile strength is 0.91% and 0.67%, respectively, for the calibration simulation, whereas these values are 2.51% and 1.81% for the validation simulation. The results suggest strong agreement between the simulation and experimental observation. Even though global deformation behavior remains consistent across multiple RVEs with the same microstructural features, local stress-strain values vary, indicating structural anisotropy, which is also consistent with the digital image correlation (DIC) result. It is proven that this proposed CP method can effectively predict the mechanical responses of components with large grains and strong textures resulting from the wire-arc DED.