Showing papers on "Zirconium alloy published in 2011"

Oxidation of Advanced Zirconium Cladding Alloys in Steam at Temperatures in the Range of 600–1200 °C

Abstract: The oxidation kinetics of the classical pressurized water reactors (PWR) cladding alloy Zircaloy-4 have been extensively investigated over a wide temperature range from operational conditions to beyond design basis accident (BDBA) temperatures. In recent years, new cladding alloys optimized for longer operation and higher burn-up are used in Western light water reactors (LWR). This paper presents the results of thermo-gravimetric tests with Zircaloy-4 as the reference material, Duplex DX-D4, M5® (both AREVA), ZIRLO™ (Westinghouse), and the Russian E110 alloy. All materials were investigated in isothermal and transient tests in a thermal balance with steam furnace. Post-test analyses were performed by light-microscopy and neutron radiography for investigation of the hydrogen absorbed by the metal. Strong and varying differences (up to 800%) in oxidation kinetics between the alloys were found at up to 1000 °C, where the breakaway effect plays a role. Less but significant differences (ca. 30%) were observed at 1100 and 1200 °C. Generally, the M5® alloy revealed the lowest oxidation rate over the temperature range investigated whereas the behavior of the other alloys was considerably dependent on temperature. A strong correlation was found between oxide scale structure and amount of absorbed hydrogen.

Classification of the critical resolved shear stress in the hexagonal-close-packed materials by atomic simulation: Application to α-zirconium and α-titanium

Abstract: We have studied the hierarchy of the activation of dislocation glide in zirconium and titanium alloys and presented experimental results in zirconium alloys We have compared the experimental results with simulations obtained by two different approaches The first is by using the stacking fault energy maps (γ surfaces) obtained by molecular dynamics (MD) and by ab initio approaches A good agreement was observed between the two approaches and with recent published work The second is to compare the experimental critical resolved shear stresses (CRSS) with those determined by MD simulations based on embedded atom method (EAM) potentials The CRSS for slip in the -direction for the basal, prismatic (type 1) and pyramidal (type 2) planes for edge dislocations are obtained Finally, we discuss the hierarchy of the glide systems with the energy criterion of the γ surfaces and with the CRSS values and we compare with both experimental and modeling data

EBSD studies on microstructure and crystallographic orientation of δ-hydrides in Zircaloy-4, Zr–1% Nb and Zr–2.5% Nb

Abstract: The Zr–Nb alloys are reported to be showing superior resistance to corrosion and substantially lower hydrogen pickup when compared with Zircaloy. In the present study Zircaloy-4, Zr–2.5Nb and Zr–1Nb (E110) alloys were hydrided using high pressure hydrogen gas furnace at a pressure of 20 MPa. Hydriding was done by thermal cycling on the samples at a temperature of 350 °C for 10 cycles with 5 h soaking time. The microstructural observations on Zircaloy-4 and Zr–2.5Nb samples show long and interlinked hydride like chains, oriented along the circumferential direction. Whereas, in Zr–1Nb (E110) hydrides were short, thick without any preferential orientation with respect to the sample reference direction. Electron backscatter diffraction (EBSD) map shows the presence of β1 phase in two phased Zr–1Nb, Zr–2.5Nb samples. Heavy { 1 0 1 ¯ 2 } 1 0 1 ¯ 1 ¯ tensile twins and { 1 0 1 ¯ 1 } 1 0 1 ¯ 2 compression twins were observed in Zr–1Nb. Hydrides in Zircaloy-4 and Zr–2.5Nb have shown (0 0 0 1)α-Zr//{1 1 1}-ZrH1.5 orientation with α-matrix whereas almost 50% of hydrides in Zr–1Nb (E110) alloy are showing (0 0 0 1) α-Zr//{1 0 0} δ-ZrH1.5 crystallographic relation. The (0 0 0 1)α-Zr//{1 0 0}δ-ZrH1.5 orientation of hydrides in Zr–1Nb (E110) is observed for the hydrides formed at the twin boundaries.

Studies regarding corrosion mechanisms in zirconium alloys

Abstract: Understanding the key corrosion mechanisms in a light water reactor primary water environment is critical to developing and exploiting improved zirconium alloy fuel cladding. In this paper, we report recent research highlights from a new collaborative research programme involving 3 U.K. universities and 5 partners from the nuclear industry. A major part of our strategy is to use the most advanced analytical tools to characterise the oxide and metal/oxide interface microstructure, residual stresses, as well as the transport properties of the oxide. These techniques include three-dimensional atom probe (3DAP), advanced transmission electron microscopy (TEM), synchrotron X-ray diffraction, Raman spectroscopy, and in situ electro-impedance spectroscopy. Synchrotron X-ray studies have enabled the characterisation of stresses, tetragonal phase fraction, and texture in the oxide as well as the stresses in the metal substrate. It was found that in the thick oxide (here, Optimized-ZIRLO, a trademark of the Westinghouse Electric Company, tested at 415°C in steam) a significant stress profile can be observed, which cannot be explained by metal substrate creep alone but that local delamination of the oxide layers due to crack formation must also play an important role. It was also found that the oxide stresses in the monoclinic and tetragonal phases grown on Zircaloy-4 (autoclave testing at 360°C) first relax during the pre-transition stage. Just before transition, the compressive stress in the monoclinic phase suddenly rises, which is interpreted as indirect evidence of significant tetragonal to monoclinic phase transformation taking place at this stage. TEM studies of pre- and post-transition oxides grown on ZIRLO, a trademark of the Westinghouse Electric Company, have used Fresnel contrast imaging to identify nano-sized pores along the columnar grain boundaries that form a network interconnected once the material goes through transition. The development of porosity during transition was further confirmed by in situ electrochemical impedance spectroscopy (EIS) studies. 3DAP analysis was used to identify a ZrO sub-oxide layer at the metal/oxide interface and to establish its three-dimensional morphology. It was possible to demonstrate that this sub-oxide structure develops with time and changes dramatically around transition. This observation was further confirmed by in situ EIS studies, which also suggest thinning of the sub-oxide/barrier layer around transition. Finally, 3DAP analysis was used to characterise segregation of alloying elements near the metal/oxide interface and to establish that the corroding metal near the interface (in this case ZIRLO) after 100 days at 360°C displays a substantially different chemistry and microstructure compared to the base alloy with Fe segregating to the Zr/ZrO interface.

Waterside corrosion in zirconium alloys

Abstract: The influence of the alloy microstructure and microchemistry on uniform waterside corrosion of zirconium alloys is reviewed, with special attention to the various stages of corrosion, such as pre-transition, transition, and breakaway.

In-situ neutron radiography investigations of hydrogen diffusion and absorption in zirconium alloys

Abstract: The fast and non-destructive character of neutron radiography provides the possibility of in-situ investigations of hydrogen uptake and diffusion in zirconium alloys. A special reaction furnace with neutron transparent windows was constructed. The method of quantitative hydrogen determination by neutron transmission measurements was calibrated for each experimental run. Additionally, oxygen is absorbed in the α-Zr phase and precipitated in the oxide layer. The calibration of the correlation between hydrogen and oxygen concentrations and total neutron cross-sections at room temperature and between 1123 and 1623 K are described. Results of in-situ neutron radiography investigations of hydrogen diffusion and absorption are presented in this paper. A linear dependence of the total macroscopic neutron cross section on the H/Zr atomic ratio as well as on the oxygen concentration was found. No significant temperature dependence of the total neutron cross-sections of hydrogen dissolved in β-Zr or oxygen dissolved in the α-Zr or precipitated in the oxide layer was found.

Hydride Platelet Reorientation in Zircaloy Studied with Synchrotron Radiation Diffraction

Abstract: Hydrogen ingress into zirconium alloy fuel cladding in light water reactors can degrade clad- ding performance as a result of the formation of brittle hydrides. In service, hydrides normally precipitate in the circumferential direction and are homogeneously distributed through the cladding thickness in ideal cases. However, temperature and stress gradients in the cladding can promote hydrogen redistribution. This hydrogen redistribution is responsible for the formation of hydride rims, dissolution, and reorientation of hydride precipitates and for the formation of brittle hydrides at stress concentration locations, all of which can reduce cladding resistance to failure. Thus, it is crucial to understand the kinetics of hydride dissolution and precipitation under load and at temperature. Studies of hydrogen behavior in zirconium alloys are normally performed post facto, which causes them to suffer both from a scarcity of data points and from the confounding effects of studying hydrides at room temperature that might be dissolved at higher tempera- ture. In the current study, we have used synchrotron radiation diffraction to study the kinetics of hydride precipitation and dissolution in situ under load and at temperature. Samples of hydrided Zircaloy-4 were examined in transmission by using 80 keV synchrotron radiation while undergoing heating and cooling in a furnace. Temperatures ranged from 20 to 550°C, and loads from 75 to 100 MPa were applied. The hydrides dissolved and reprecipitated in a different orientation when sufficiently high loads were applied. Through careful study of the intensities and full-width half maxima of the diffraction peaks as a function of time, load, and temperature, it was possible to identify the characteristic diffraction patterns for the reoriented hydrides so that the kinetics of dissolution, reprecipitation, and orientation of the hydrides could be followed. The analysis of the diffraction patterns allowed a detailed understanding of the kinetics of hydride evolution under temperature and stress, as presented in this work.

Effects of Secondary Phase Particle Dissolution on the In-Reactor Performance of BWR Cladding

Abstract: Although the complete mechanisms are not yet fully understood, it is well established that the secondary phase particle (SPP) size distribution and chemical composition have a crucial affects in the reactor corrosion rate and hydrogen uptake in boiling water reactor (BWR) cladding. To further study these effects, different selected fuel rod assemblies with standard LK3 cladding materials that have been irradiated in KernKraftwerk Leibstadt (KKL) for three, five, six, seven, and nine annual cycles were investigated using transmission electron microscopy (TEM). TEM analysis of the samples showed that the average size of SPPs increases as the small SPPs are dissolved during irradiation. After three cycles the Fe–Cr bearing SPPs have been identified as completely amorphous over the whole range of examined samples and the Fe–Ni bearing SPPs remained crystalline. The EDX analyses of several Fe–Ni bearing precipitates show that the Fe/Ni ratio stays more or less constant for the irradiated material at 1.5 to 1.6 and drops to about 1 after fast fluence of 17.9×1021 n/cm2 (>1 MeV) for nine cycles sample. Results from this study confirms that the increased oxide thickness, the higher hydrogen content, and the accelerated growth of the rods at rod average burnup of 78 MWd/kgU goes along with a change in the appearance of SPPs by TEM.

Enhancement of glass-forming ability and corrosion resistance of Zr-based Zr-Ni-Al bulk metallic glasses with minor addition of Nb

Abstract: Cu-free Zr-based bulk metallic glasses (BMGs) with high glass-forming ability (GFA) have been discovered in the Zr-Ni-Al-Nb system. The GFA of Zr-Ni-Al alloys can be significantly enhanced by the minor addition of Nb, which increases the glass transition temperature, and lowers the melting and liquidus temperatures. The Zr-Ni-Al-Nb BMGs have critical sample diameters of 15–20 mm as verified by a copper mold casting. They exhibit excellent corrosion resistance in chloride-ion-containing solutions, which is considerably better than that of other known Zr-based BMGs with superhigh GFA. It was revealed that the formation of highly protective Zr-, Al-, and Nb-enriched surface film is responsible for the high corrosion resistance. The BMGs also possess high compressive yield strength of 1786–1847 MPa, large plastic strain of 1.6–3.2%, and a high Poisson’s ratio of 0.365–0.371.

Understanding crack formation at the metal/oxide interface during corrosion of Zircaloy-4 using a simple mechanical model

Abstract: It has been established in previous works that corrosion kinetics in primary water of various zirconium alloys are periodic. Each period is associated with a layer of cracks parallel to the metal-oxide interface. These observations have been made either in autoclave or in pile. This indicates that corrosion processes in autoclave and under irradiation are of similar nature though their absolute kinetics might be different. Taking advantage of this correlation between cracks and corrosion kinetics, the present work aims at identifying the main microstructural parameters controlling cracks appearance in the oxide layer under well-controlled conditions. In order to achieve this, Zircaloy-4 was heat-treated to obtain various metallurgical states (stress-relieved versus recrystallised with different grain sizes) followed by corrosion tests in primary water. The key metallurgical parameters for the various conditions have been analysed (texture, precipitate sizes and grain sizes and distributions) using electron microscopy and synchrotron X-ray diffraction techniques. Corrosion kinetics of the various Zircaloy-4 microstructures are distinct as expected from the literature. Crack morphology in the oxide layer has been analysed and quantified using a dual beam scanning electron microscope/focused ion beam. Crack layers are evident even at small scale of observation. Three dimensional (3D) images of the oxide structure are presented. Cracks observed in this way are typically penny-shaped with a radius of about 100 nm. Near the metal-oxide interface, they are mainly found at the top of metal protrusions in the oxide. The roughness of the metal-oxide interface was measured. It does not exhibit any periodicity. The residual stresses in the oxide layers were measured by high energy (44 keV) synchrotron X-ray diffraction in transmission mode. Large compressive stresses (∼−1 GPa), changing with the metallurgical state and through the oxide scale thickness, were measured. The residual stresses in the oxide layers were measured by high energy (44 keV) synchrotron X-ray diffraction in transmission mode. Large compressive stresses (∼−1 GPa), changing with the metallurgical state and through the oxide scale thickness, were measured. A model of the oxide breaking at the point of transition has been developed. It is based on mechanical considerations and the existence of compressive stress in the oxide layer.

Oxygen, Hydrogen and Main Alloying Chemical Elements Partitioning Upon Alpha←→Beta Phase Transformation in Zirconium Alloys

Abstract: Due to their adequate properties, zirconium alloys are the reference materials for the nuclear fuel cladding tubes of Light Water Reactors (LWR). During some hypothetical accidental High Temperature (HT) transients, the materials should experience heavy steam oxidation and deep metallurgical evolutions. This promotes Alpha-Beta phase transformations and an associated strong partitioning of oxygen/hydrogen and of the main chemical alloying elements (Nb, Sn, Fe and Cr). Moreover, it has been shown quite recently that such chemical elements partitioning during on-cooling Beta-to-Alpha transformation can strongly impact the residual mechanical properties of HT oxidized materials. Thus, it appeared that it was important to better quantify and, if possible, to compute the quite complex phase equilibrium that occurs in multi-alloyed zirconium materials in the presence of both oxygen and hydrogen. For that, systematic studies have been performed on industrial alloys, charged with oxygen and/or hydrogen. After applying different heating/cooling scenarii, both Electron Microprobe using Wave Dispersive Spectrometry (WDS) and Nuclear Microprobe using Elastic Recoil Detection Analysis (ERDA) have been applied. Finally, to support the observed chemical elements partitioning between the Alpha and Beta allotropic phases, some thermodynamic calculations have been performed thanks to the development and the use of a specific thermodynamic database for zirconium alloys called “Zircobase".

Hydrogen Absorption Mechanism of Zirconium Alloys Based on Characterization of Oxide Layer

Abstract: In order to get a better understanding of the mechanism governing hydrogen absorption behavior in Zr-based alloys, various characterization techniques were applied to the oxide layers of three alloys: Zry-2, GNF-Ziron (Zry-2-based alloy with ∼0.26 wt % Fe), and VB (Zr-based alloy containing ∼0.5 wt % Sn, ∼0.5 wt % Fe, and ∼1 wt % Cr). Out-of-pile corrosion tests were carried out in 400 °C steam and 290 °C LiOH water. For both tests, the hydrogen absorption decreased with higher iron content in the alloys, in the order of Zry-2>GNF-Ziron>VB, despite different kinetics of a parabolic law in the former test and a linear law in the latter test. The acceleration of hydrogen absorption in the LiOH water was ascribed to the formation of degraded or open grain boundaries up to locations very near the metal/oxide interface. The pre-transition steam oxides of 1.4–1.7 μm had a double layer structure composed of the outside non-protective oxide of monoclinic ZrO2 with faster diffusivity and the inside barrier layer of predominantly tetragonal ZrO2 with slower diffusivity. The thickness of the barrier layer of about 0.8–0.9 μm was not changed for the different alloys. The diffusion coefficient of deuterium in the VB oxide was approximately half of that in the GNF-Ziron oxide. This factor for the diffusivity was consistent with their hydrogen pickup performance. The higher compressive stress in the barrier layer was directly linked to the higher hydrogen pickup resistance of the alloys. Preferential dissolution of alloy elements from the second-phase particles (SPPs) into the oxide matrix was evinced for iron, and was very limited for chromium and nickel. These two elements had a tendency to exist as precipitates in the oxide layers, chromium mainly as oxide, and nickel mainly as metal. The superior hydrogen absorption performance of VB containing higher iron content and the SPPs with larger size and number density was attributable to the dissolved iron effect and higher compressive stress state generated from the delayed oxidation of the SPPs in the barrier layer.

Simulation of the β→α(O) Phase Transformation due to Oxygen Diffusion during High Temperature Oxidation of Zirconium Alloys

Abstract: The EKINOX numerical code, formerly developed to simulate high temperature oxidation of Ni alloys, has been recently adapted to solve out the issue of high temperature oxidation of Zirconium alloys. This numerical code is a one dimensional model that simulates the growth of an oxide layer using a specific algorithm for moving boundaries problem. In order to simulate the oxygen diffusion inside Zr alloys, an adaptation of the EKINOX code was necessary. It consisted in adding, first, a non-null oxygen equilibrium concentration in the substrate and second, a new interface in order to simulate the β/α(O) phase transformation due to oxygen diffusion. In this study, EKINOX has also been coupled with the thermodynamic database for zirconium alloys ZIRCOBASE (thermocalc formalism) in order to obtain accurate concentrations values in each phases (considering local equilibrium at each interface). The present paper illustrates the simulation ability of the code by comparing experimental and calculated oxygen diffusion profiles corresponding to different cases, from isothermal oxidations at high temperature (900 < T < 1250°C) to the study of dissolution kinetics of a pre-transient oxide layer under a neutral environment. The influence of pre-hydriding from a few hundreds up to thousands weight-ppm is also derived from the calculations.

Modelling and Simulation of Reactor Fuel Cladding under Loss-of-Coolant Accident Conditions

Abstract: We present a unified model for calculation of zirconium alloy fuel cladding rupture during a postulated loss-of-coolant accident in light water reactors. The model treats the Zr alloy solid-to-solid phase transformation kinetics, cladding creep deformation, oxidation, and rupture as functions of temperature and time in an integrated fashion during the transient. The fuel cladding material considered here is Zircaloy-4, for which material property data (model parameters) are taken from the literature. We have modelled and simulated single-rod transient burst tests in which the rod internal pressure and the heating rate were kept constant during each test. The results are compared with experimental data on cladding rupture strain, temperature, and pressure. The agreement between computations and measurements in general is satisfactory. The effects of heating rate and rod internal pressure on the rupture strain are evaluated on the basis of systematic parameter variations of these quantities. In the α-phase ...

Hydrogen Solubility and Microstructural Changes in Zircaloy-4 Due to Neutron Irradiation

Abstract: The present work deals with two essential degradation mechanisms of zirconium alloys: Hydrogen behavior and microstructural transformations due to neutron irradiation. The studies were carried out by using samples taken from Zircaloy-4 components, which remained for 14 years in the Atucha I pressurized heavy water reactor, accumulating neutron fluences of up to 10 22 neutrons / cm 2 . A hydrogen/deuterium solubility increase was observed in both dissolution and precipitation processes. With the aim to follow this evolution with radiation damage recovery, the irradiated samples were annealed for different periods at 350 and 400 ° C in a differential scanning calorimeter (DSC). These temperatures were chosen to reproduce those reached by spent fuels in the early stages of the dry storage (DS). Between intervals, the terminal solid solubility temperatures in dissolution (TTSSd) and in precipitation (TTSSp) were determined. They were also determined after additional annealing at 500 and 600 ° C , trying to simulate long periods at DS. The hydride morphology and the metallurgical state of the Zr ( Fe , Cr ) 2 second-phase particles (SPPs) were followed between annealing by optical microscopy, analytical electron microscopy, synchrotron light X-ray diffraction (SXRD), and DSC. As a brief summary, we found from the first to the last run of the annealing sequence an increase of 55 ° C in TTSSd and 40 ° C in TTSSp on average. The optical and transmission electron microscopies (TEM) indicate that in the material in the irradiated condition, exists a hydride size distribution of four orders of magnitude, from 10 nm to 100 μ m , but biased to sizes smaller than 5 μ m . A complete amorphisation of the SPPs was observed in high fluence samples by TEM and SXRD. Similar TEM observations made in irradiated samples annealed above 400 ° C show a re-precipitation process of nano-SPPs forming clusters. The re-precipitation is affected by kinetics: The SPPs crystallization temperature, T c , rises from 450 to 500 ° C as the heating rate increases from 5 to 20 ° C / min . The measured heat of crystallization was 27.3 ± 2.1 kJ / mol ( Fe + Cr ) .