Zr-Nb Alloys Research Articles

Study of high burnup effect on steam oxidation of Zircaloy and its regulatory implications via the developement of a pre-transient oxide model of TRANOX

A mechanisitc model that gives oxygen distribution of Zircaloy cladding with pre-transient oxide and absorbed hydrogen under steam oxidation environments has been developed. The model has been incorporated into TRANOX. The model has been validated against steam oxidation experiments with Zircaloy-4 specimens that were pre-oxidized in simulated PWR environments and past experimental data available in open literature. The high burnup effects on cladding oxidation owing to pre-transient oxide and hydrogen can be categorized into: thinning of initial Zr-matrix by the formation of pre-transient ZrO2, increasing diffusion resistance of oxygen with pre-transient ZrO2, and increasing oxygen diffusion rate with a thicker β-phase by β-stabilizing hydrogen. Relative magnitudes of these effects with temperature, pre-transient oxide thickness, and hydrogen content determine the dynamics of high burnup fuel oxidation. Mechanistically modeling the high burnup effect, the model comprehensively simulates the integrated effects of pre-transient oxide and absorbed hydrogen. This study confirms that the corrected Cathcart Pawel correlation-based Equivalent Cladding Reacted evaluation method employed by the U.S Nuclear Regulatory Commission is an effective, yet accurate, simplification of the high burnup cladding oxidation modeling for safety analysis of representative LBLOCA of current PWR systems. The application of the developed model to safety analysis revealed allowable discharge burnup limits of 59 and 76 MWd/kgU for Zircaloy-4 and Zr-Nb alloy, respectively from the view point of fuel accident safety.

Structural Transformations of Ceramics Forming during High-Temperature Nitridation of Zr–Nb Alloys

Structural Transformations of Ceramics Forming during High-Temperature Nitridation of Zr–Nb Alloys

Structural Stability, Thermodynamic and Elastic Properties of Cubic Zr0.5Nb0.5 Alloy under High Pressure and High Temperature

Structural stability, sound velocities, elasticity, and thermodynamic properties of cubic Zr0.5Nb0.5 alloy have been investigated at high pressure and high temperature by first-principles density functional calculations combined with the quasi-harmonic Debye model. A pronounced pressure-induced shear wave velocity stiffening in Zr0.5Nb0.5 alloy is observed at pressures above ~11 GPa, owing to its structural instability under high pressure, whose anomalous behavior is also observed in the end members of Zr-Nb alloys for Zr at ~13 GPa and for Nb at ~6 GPa upon compression, respectively. In addition, high-pressure elasticity and elastic-correlated properties of cubic Zr0.5Nb0.5 are reported, as compared with previous studies on Zr-Nb alloys with different compositions. A comprehensive study of the thermodynamic properties of cubic Zr0.5Nb0.5, such as heat capacity (Cv), thermal expansion coefficients (α), and Debye temperature (ΘD), are also predicted at pressures and temperatures up to 30 GPa and 1500 K using the quasi-harmonic Debye model.

High-temperature oxidation of Zr[sbnd]1Nb zirconium alloy with protective Cr/Mo coating

The oxidation and interdiffusion behavior of Cr- and Cr/Mo-coated Zr1Nb zirconium alloy were investigated. A single-layer Cr (8 um) and multilayer Cr (8 μm)/Mo (3 μm) coatings were deposited by magnetron sputtering. The coated Zr1Nb alloy samples were oxidized in air at 1100 °C for 15–60 min. Both coating types had a protective scale during high-temperature (HT) oxidation. The use of Mo sublayer resulted in preventing CrZr interdiffusion under high temperature. Scanning electron microscopy, optical microscopy, X-ray diffraction before and after HT oxidation and in situ X-ray diffraction up to 1250 °C were used to identify interdiffusion behavior of the system of Cr/Mo/Zr. Some aspects of applying of Cr/Mo coatings for Zr nuclear fuel claddings are discussed.

Phase Transformations Accompanying High-Temperature Nitridation of Zr–Nb Alloys

Phase Transformations Accompanying High-Temperature Nitridation of Zr–Nb Alloys

Characterization of Zr-Nb-Fe(-Cr) precipitates in Zr-based alloys using density functional theory

The presence of iron (and chromium) impurities in Zr-Nb based alloys leads to the formation of intermetallic Zr-Nb-Fe(-Cr) precipitates. These precipitates are known to affect the thermomechanical properties of Zr-Nb alloys. This study seeks to model these intermetallic precipitates from first principles using density functional theory (DFT) and ab initio Metropolis Monte Carlo (MC). A wide variety of Laves-type atomic structures were considered, composed of Zr, Nb, Fe and Cr in proportions close to experimental values. An exhaustive ensemble of elemental permutations was taken into account. The calculations show that hexagonal ternary alloys of equiatomic composition can be designated as a Zr(Nb, Fe)2 Laves phase. They also show that hexagonal precipitates found in Canadian Deuterium Uranium (CANDU) pressure tube material can be designated as Zr(Zr, Nb, Fe)2 and Zr(Zr, Nb, Fe, Cr)2 Laves phases. Ab initio MC simulations indicate that the lowest-energy Zr(Nb, Fe)2 phases considered in this work are stable under annealing at 1073 K. The DFT-optimized Zr(Zr, Nb, Fe, Cr)2 atomic positions are consistent with the plane d-spacings observed in TEM images.

Athermal ω Phase and Lattice Modulation in Binary Zr-Nb Alloys.

To further explore the potential of Zr-based alloys as a biomaterial that will not interfere with magnetic resonance imaging (MRI), the microstructural characteristics of Zr-xat.% Nb alloys (10 ≤ x ≤ 18), particularly the athermal ω phase and lattice modulation, were investigated by conducting electrical resistivity and magnetic susceptibility measurements and transmission electron microscopy observations. The 10 Nb alloy and 12 Nb alloys had a positive temperature coefficient of electrical resistivity. The athermal ω phase existed in 10 Nb and 12 Nb alloys at room temperature. Alternatively, the 14 Nb and 18 Nb alloys had an anomalous negative temperature coefficient of the resistivity. The selected area diffraction pattern of the 14 Nb alloy revealed the co-occurrence of ω phase diffraction and diffuse satellites. These diffuse satellites were represented by gβ + q when the zone axis was [001] or [113], but not [110]. These results imply that these diffuse satellites appeared because the transverse waves consistent with the propagation and displacement vectors were q = <ζ ζ¯ 0>* for the ζ~1/2 and <110> directions. It is possible that the resistivity anomaly was caused by the formation of the athermal ω phase and transverse wave. Moreover, control of the athermal ω-phase transformation and occurrence of lattice modulation led to reduced magnetic susceptibility, superior deformation properties, and a low Young’s modulus in the Zr-Nb alloys. Thus, Zr-Nb alloys are promising MRI-compatible metallic biomaterials.

Hydrogen solubility in Zr–Nb alloys

We present a fast and accurate method for predicting the thermodynamics of hydrogen solubility and trapping in alloys, which is two orders of magnitude faster than conventional ab-initio approaches. The model hinges on the finding that the solubility of H is dominated by its nearest neighbour environment. We apply the method to the problem of hydrogen redistribution in Nb-containing Zr nuclear fuel cladding, and validated it against brute-force ab-initio approaches. We find that hydrogen preferentially dissolves into the β-Zr phase found in as-manufactured 2.5%Nb alloys, and H is likely to redistribute into the α phase following the irradiation-induced decomposition of β-Zr. β-Nb particles found in as-manufactured 1%Nb alloys may act as weak sinks for H, however irradiation-induced change of composition of β-Nb particles increases their hydrogen-trapping strength. Nano-platelets formed under irradiation in these alloys are potentially even stronger hydrogen sinks, especially if induced by proton irradiation rather than neutrons.

Microstructure and corrosion performance of Zr–1Nb alloy laser welded joints treated by micro-arc oxidation

The protective coatings on laser-welded joints of Zr–1Nb alloy were prepared by the micro-arc oxidation (MAO) process. The morphology, microstructure and phase composition of MAO coatings at different regions of Zr–1Nb alloy laser-welded joints were investigated by scanning electron microscopy (SEM) and X-ray diffraction (XRD). Potentiodynamic polarization and electrochemical impedance spectroscopy (EIS) were used to evaluate the corrosion behaviors of the coatings. The MAO coatings at the fusion zone (FZ) and heat-affected zone (HAZ) were smooth with some small holes, while MAO coatings at base metal (BM) were relatively rough with large holes. The cross-sectional topography at FZ coatings was quite dense with less distribution of pores and cracks, which can effectively improve the protective properties of the coatings. HAZ coatings were divided into outer loose layer and inner dense layer. The BM coating layer possessing many holes was a typical sparse layer structure. The phase compositions of welded joints treated by MAO were diverse at different zones of the joints, but all of them consisted of the monoclinic ZrO2 and tetragonal ZrO2 phases. The coatings of FZ had the best corrosion resistance, compared with HAZ and BM. The results of EIS indicated that the thickness and defects of the inner compact layer played a key role in improving the corrosion resistance of the MAO coatings on the surface of the Zr–1Nb alloy laser-welded joints.

Hydride embrittlement resistance of Zircaloy-4 and Zr-Nb alloy cladding tubes and its implications on spent fuel management

• Mechanical tests of hydrided Zircaloy-4 and Zr-Nb alloy with Ring Compression Test. • EBSD analyses on texture of reactor-grade hydrided Zircaloy cladding tubes. • Abrupt ductile to brittle transition at 560 and 490 wppm for Zr-4 and Zr-Nb alloy. • Larger grain size of Zr-4 reduces interlink of inter-granular hydrides. • Reduced hydride interlink increases hydride embrittlement resistance. In the spent fuel storage phase, nuclear fuel cladding is subjected to increased embrittlement owing to a large amount of hydride precipitation. This study compares the differences in the hydrogen-induced cladding embrittlement of cold work stress-relief annealed (CWSR or SRA) Zircaloy–4 and Zr-Nb alloy cladding with ring compression test at the temperature of the spent fuel pool, which is approximately 40 °C. Experiments demonstrate that an abrupt ductile to brittle (DTB) transition occurs at the critical hydrogen content of 560 and 490 wppm for Zircaloy-4 and the tested Zr-Nb cladding tubes, respectively. Even beyond the critical hydrogen content, sufficiently high cladding ductility with the offset strain >10% is maintained up to ∼ 90 MWd/kgU for both cladding materials on the rod-average basis. Extensive EBSD analyses coupled to thermodynamic modeling demonstrate that this is primarily due to the slightly larger grain diameter of Zircaloy-4 tube, which reduces the number of available sites for inter-granular hydride precipitation. Reduced inter-granular hydride precipitation prevents the extent of hydride interlink, thereby improving the hydride embrittlement resistance. Nevertheless, the tested Zr-Nb alloy cladding presents an extened discharge burnup limit for abrupt DTB transition owing the reduced in-core cladding oxidation rate. The presented understanding of microstructural effect on hydride interlink and resulting embrittlement may provide a basis for understanding the general hydride embrittlement phenomena of Zircaloy cladding which include, but not limited to, wet storage, dry storage, and post-accident ductility of high burnup Zircaloy cladding.

Tem Characterizations of a Zro2/Cr Composite Coating on Zr-1nb Alloy after 1200°C Steam Oxidation

Tem Characterizations of a Zro2/Cr Composite Coating on Zr-1nb Alloy after 1200°C Steam Oxidation

Oxidation of Typical Zr-1nb Alloy in High-Temperature Air Under 700–900 °C

Oxidation of Typical Zr-1nb Alloy in High-Temperature Air Under 700–900 °C

Evolution of β-Nb precipitates during hydrogenation of a zirconium-niobium alloy

We firstly reported that during hydrogenation of Zr-Nb alloy, β-Nb particles transformed into rods with their major axes parallel to each other within neighboring ε-ZrH2 grains, while obeying orientation relationship with the matrix: [101] ZrH2 // [1¯13] β-Nb. Parallel Moiré fringes were observed viewing along [101] ZrH2 due to the low mismatch angle between (101) ZrH2 and (101¯) β-Nb.

Study of Coatings Formed on Zirconium Alloy by Plasma Electrolytic Oxidation in Electrolyte with Submicron Yttria Powder Additives

Coatings with thickness 40 to 150 μm were formed by plasma electrolytic oxidation (PEO) on the zirconium alloy Zr-1Nb (Zr-1% Nb) in the slurry electrolyte containing 9 g/L Na2SiO3 5H2O, 5 g/L Na(PH2O2) and 6 g/L submicron Y2O3 yttria powder during 60 min under the AC electrical mode at current densities 20; 30 and 40 A/dm2. The surface morphology, structure, composition, and corrosion-protective ability of the formed coatings have been analyzed. At PEO current density 30 A/dm2, a predominantly tetragonal phase of zirconia was formed in coatings. Increasing the PEO current density up to 40 A/dm2 promoted the formation of the coating surface layer containing submicron yttria particles. Electrochemical polarization studies in 0.5% LiOH solution showed that PEO coatings demonstrated high corrosion-protective ability. The dependence of the polarization currents on the PEO current density was found to be inconsequential.

Silicon carbide coatings on zirconium alloy for light water reactor fuel cladding studies

Coating of zirconium alloys is a promising approach for improving various properties of the cladding of fuel elements, including physical, mechanical properties, corrosion resistance and reduction of radiation damage. In this paper, SiC coatings with thicknesses of 100, 200, and 300 microns were deposited by selective laser sintering (SLS) on Zr-1Nb alloy substrates at a laser power of 150 W. The 200 μm layer deposition registered the least contact depth during macro-scratch test to a maximum depth of about 25 μm into the coating. The uniformity and less porosity of the coatings is observed on the transverse sections in the entire variant. Analysis of the phase composition of the coating revealed the formation of silicon carbide phases (6H & 4H), with protective oxides of Al2O3, Y2O3, SiO2 and YAlO3. High coating adhesion was noticed with results of the scratch tests showing no delamination regardless of the higher loads used. However, partial chippings and transverse cracks were observed. The results obtained indicate the need for further research to improve the porosity of the surface structure of the coating and to conduct other necessary tests.

Plastic relaxation and solute segregation to β-Nb second phase particles in Zr-Nb alloys: A discrete dislocation plasticity study

There is clear evidence in the literature that iron segregates to the interface of second phase particles (SPPs) in unirradiated Zr-Nb alloys, and that it does not do so in the presence of radiation damage. In this work, a discrete dislocation plasticity model is developed that takes into account the long-range stress field of the SPP interface. A simple analytical model is also outlined, providing an upper bound for estimating the amount of interstitial segregation. The model provides a possible mechanism to explain both the iron segregation to coherent SPPs and its subsequent loss after irradiation. Qualitatively, the model proved to be insensitive to variations of all geometrical and computational parameters, allowing for general conclusions to be drawn. The model suggests that the segregation originates from a tensile field of order 1 GPa induced by the dislocations generated during the plastic relaxation around the SPP. This leads to the six-fold increase in the iron concentration observed in experiments. In the model, the loss of SPP/matrix coherency after irradiation causes the dislocations to drift away from the interface, and the iron concentration is homogenised accordingly. The hydrogen concentration was also predicted and found to be about 50% higher than in the bulk zirconium matrix at room temperature. The computational framework is built to be fast, making possible a statistical analysis on over five hundred simulations for improved reliability of the predictions.

Studying the Effect of Fuel Elements Structural Materials Corrosion on their Operating Life

The paper deals with experimental data regarding the effect of internal and external factors on the corrosion decay of Zr1Nb alloy fuel elements. Based on the analysis results, losses of zirconium that transfers to oxide or coolant as per the fuel element wall weight and thickness as well as economic losses from their corrosion decay have been theoretically calculated. To avoid a state-level emergency occurrence, an increase in the fuel element wall thickness up to 660 μm is proposed, which can increase the operating life under the conditions of trouble-free coolant mass transfer hydrodynamic mode.

Atomic oxygen effect of Zr-Al-C coatings on ZrNb alloys used in space environment

The preparation of suitable coating materials is crucial for enhancing the atomic oxygen erosion of zirconium alloys. Zr-Al-C ceramics composed of MAX phase have excellent inherited properties of metal and ceramic, making them useful as spacecraft materials for atomic oxygen protection. The MAX phases are a group of layered ternary compounds with the general formula Mn+1AXn (M: early transition metal; A: group A element; X: C and/or N; n = 1–3), which combine some properties of metals, such as machinability, low hardness and damage tolerance, with those of ceramics, such as high elastic moduli, high temperature strength, oxidation and corrosion resistance. In this paper, Zr-Al-C coatings were successfully prepared by the magnetron sputtering method. The macroscopy morphology, microstructure, adhesion and the effect of atomic oxygen of Zr-Al-C coatings were all investigated. The results showed the successful deposition of Zr-Al-C coatings consisting of Zr2Al3C4, Zr3Al3C5, and ZrAl3 phases on zirconium alloy surface by adding an intermediate layer using magnetron sputtering route. The adhesion between the coating and the substrate was estimated to 18 N by scratch testing. After 30 cycles of cold and hot impact tests, the Zr-Al-C coating remained undamaged without crack or peel off. The oxidation kinetics of ZIRLO with Zr-Al-C coating followed approximatively a linear variation law. The oxidation rate of ZIRLO containing Zr-Al-C coating was lower than that of uncoated Zr alloy within 30 h of atomic oxygen exposure. In other words, Zr-Al-C coating could protect Zr alloy under certain exposure dosage of atomic oxygen. Meanwhile, further increase in exposure time of atomic oxygen led to the formation of ZrO, ZrO2, Al2O3 on Zr-Al-C coating. Also, more pores appeared on Zr-Al-C coating surface due to the oxidation of carbon elements in the coatings to form volatile oxides, such as CO and CO2. As atomic oxygen erosion continued, the existing holes became connected to form complex channel-like labyrinthine two-dimensional network structures. The latter penetrated further the atomic oxygen into the interior of Zr-Al-C coating to continue the corrosion effect of atomic oxygen. By considering the protection effect of Zr-Al-C coating on Zr alloy under a certain exposure dosage of atomic oxygen, the protection time can be calculated and experiments can be designed according to the linear law obtained as a function of the thickness of Zr-Al-C coating. In sum, these results look promising for future application of MAX or MAX-like phase in spacecraft environment.

Optimized interatomic potential for atomistic simulation of Zr-Nb alloy

We present a new classical interatomic potential for a study of the binary Zr-Nb system, taking into account a wide range of the components concentrations. The potential was developed by virtue of the force-matching method that is capable of ensuring a high accuracy at the description of the complex systems containing diverse crystal phases. At simulation of pure Zr, the potential correctly describes a relative stability of Zr phases (α-Zr, β-Zr and ω-Zr) and qualitatively reproduces the right arrangement of these phases in the phase diagram. It is remarkable that β-Zr phase is found to have a dynamically unstable structure at the low temperature, in agreement with the ab initio calculations. The potential can also play a role in considering the tasks related to the crystal defects in the Zr-Nb system. In support of this statement, we show the simulation results proving adequate representation of a number of key properties of the crystal defects in Zr-Nb system. In particular, the offered potential reproduces formation/solution energies of point defects with well accuracy. To illustrate wide application possibilities for the model, we made a prediction of atomic self-diffusion and impurity diffusion in Zr and Nb. Also, the potential ensures correct description of a screw dislocation in niobium, which is a crucial point for the investigation of plasticity.

Study on the dissolution of β-precipitates in the Zr–1Nb alloy under the influence of Ne ion irradiation

Abstract The stability of β-precipitates in the Zr–1Nb alloy has been studied under Ne ion irradiation of energy 250 keV by insitu transmission electron microscope as a function of irradiation dose. The irradiation was carried out up to ∼136 dpa at 573 K. Microstructural investigations have shown that up to ∼38 dpa, precipitates showed an increase in size, and for irradiation doses &amp;gt;38 dpa, the size of the precipitates was noticed to reduce. Post-irradiation energy-dispersive spectrometry of the specimens revealed the Nb concentration throughout the matrix to be ∼0.8–1.5%. Three-dimensional atom probe tomography was also carried out for irradiated specimens to look for the presence of any nanoclusters. However, Nb clustering was not observed in the specimens. It is proposed that the dissolution of the precipitates may be facilitated by an increase in the solubility limit of Nb in Zr caused by irradiation. The solubility limit may increase by the introduction of defects generated by irradiation and by the destabilization of the β-phase. This may result in back-diffusion of Nb atoms to the matrix by radiation-enhanced diffusion to lower the strain produced by the defects, resulting in the dissolution of the precipitates.