Zircaloy-4 Samples Research Articles

Characterisation of Deuterium Distributions in Corroded Zirconium Alloys Using High-Resolution SIMS

Hydrogen diffusion through the oxide grown on Zr alloys by aqueous corrosion processes plays a critical role in determining the rate of hydrogen pickup which can result in embrittlement of fuel cladding and limit the burnup of the nuclear fuel it encapsulates. Mapping the hydrogen/deuterium distributions in these oxide layers, especially in the barrier layer close to the metal/oxide interface, is a powerful way to understand the mechanism of both oxidation and hydrogen pickup. Here we have characterised by high-resolution SIMS analysis the deuterium distribution in oxide layers on a series of Zr alloys, including autoclave-oxidised Zircaloy-4, Zr-1Nb and Zr-2.5Nb alloys, and in-flux and out-of-flux corroded Zr-2.5Nb samples. Pre-transition Zircaloy-4 samples show a high deuterium trapping ratio in the oxide and a higher diffusion coefficient than in oxides on the Nb-containing samples. Neutron irradiation increases the deuterium diffusion coefficient, the deuterium concentration in the oxide and the pickup fraction in Zr-2.5 Nb samples. Comparative NanoSIMS and EDX/SEM analysis demonstrates that the deuterium is not preferentially trapped at SPPs in the oxides on any of the alloys studied, but there is direct evidence for trapping at the surfaces of small oxide cracks especially in Zircaloy-4 samples. The high resolution mapping of these hot-spots in 3D can provide unique information on the mechanisms of hydrogen uptake, and suggests that the development of interconnected porosity in the oxide may be the critical rate-determining mechanism that controls HPU in the aqueous corrosion of zirconium alloys in water-cooled reactors.

Texture Analysis and Fracture Behavior of Zircaloy-4 Processed Through Swaging

Zirconium alloy (Zircaloy-4) was cold-worked using the swaging process and subsequently subjected to mechanical and microstructural characterization in the present work. The influence of feed rate (0.7, 1.25, 2 m/min) during swaging on phase analysis, tensile properties and fracture behavior of the alloy was investigated. The swaged Zircaloy-4 samples were annealed at 732 °C for 3 h holding time to study the microstructural evolution and changes in its fracture behavior. Tensile test results show the improved tensile strength and ductility in the case of 1.25 m/min feed rate. The three-point bend test was performed on swaged and annealed samples to study the fracture behavior of Zircaloy-4 alloys. The apparent fracture toughness value of swaged samples (76.20 MPa.m0.5 for 0.7 m/min, 69.22 MPa.m0.5 for 1.25 m/min, and 57.52 MPa.m0.5 for 2 m/min) and annealed samples (45.12 MPa.m0.5 for 0.7 m/min, 44.37 MPa.m0.5 for 1.25 m/min and 41.81 MPa.m0.5 for 2 m/min) are reported in this work. The result shows the higher fracture toughness for swaged samples due to higher dislocation density. The deformed and annealed Zircaloy-4 samples were characterized by scanning electron microscopy–electron backscattered diffraction (SEM–EBSD) to elucidate the fracture behavior as well as phase analysis of the alloy. Texture analysis was performed using EBSD on the samples subjected to swaging at different feed rates. It reveals different types of texture on swaged and annealed rod samples. The swaged samples show basal texture 〈0002〉, and annealed samples show combination of basal 〈0002〉 and prismatic $$ \langle 10\bar{1}0\rangle $$. Different types of textures are observed with varying feed rates and are found to be directly correlated with tensile and fracture properties.

Selection of deformation modes and related texture evolution in Zircaloy-4 during one pass cold pilgering

Zircaloy-4 is usually used as cladding tubes and structure components in nuclear reactor. Cladding tubes are manufactured mainly by hot extrusion and subsequent cold pilgering steps. The texture of cladding tubes is one of key factors influencing the final performance and becomes a hot issue in recent years. This work focuses on enclosing selection of deformation modes and related texture evolution during one-pass pilgering process. Electron backscatter diffraction (EBSD) samples of Zircaloy-4 were cut from a tapered tube with 30%, 40% and 50% area reduction respectively to find the effect of area reduction on texture evolution. The selection of deformation modes with various reduction amount was also measured with in-grain misorientation axes method. The relationship between texture evolution and deformation modes was then discussed with Schmid factor distribution calculation and activity of slip modes based on visco-plastic self-consistent (VPSC) simulation during pilgering deformation. The results show that basal slip activity indeed exists when deformation strain increases to a sufficiently high level. For low strains, the basal pole is changing from transverse direction (TD) to radial direction (RD) caused by the combination of prismatic and pyramidal slip. For higher strains, the cause of the same texture change tendency is primarily related to basal slip.

Microstructure, corrosion resistance and oxidation behavior of Cr-coatings on Zircaloy-4 prepared by vacuum arc plasma deposition

Cr-coated Zircaloy-4 samples are prepared by vacuum arc plasma deposition and their microstructure, corrosion resistance and oxidation behavior are investigated. The as-received coating is dense with no pores or defects observed. The corrosion rate of Cr coating is significantly lower than Zircaloy-4 under the condition simulating the coolant of both PWRs and BWRs. Cr coating can prevent zirconium from harsh and fatal oxidation in air. Bubbles are found on the surface of the coated samples after oxidizing at 1200°C. These bubbles are within the Zr-substrate, locating at the interface of α Zr (O) layer and prior β Zr.

Enhanced wear resistance of CrAl-coated cladding for accident-tolerant fuel

For enhancing accident tolerance in current operating LWRs, Zr-based claddings with a protective layer, which have outstanding resistance to high-temperature oxidation, are selected as strong candidates for accident-tolerant fuel claddings. One of the key factors determining the successful application of these coated claddings is their ability to maintain their coating layer during normal operation without significant damage. This study aimed to verify the scratch and fretting wear behaviors of coated Zircaloy-4 claddings deposited with a Cr–10Al binary alloy target (CrAl) by the arc ion plating method. In the scratch tests, the commercial Zircaloy-4 samples showed accumulated scratch grooves with ductile chipping caused by adhesive failure. However, there was no gross spallation, except for well-developed scratch traces caused by plastic deformation, on the entire stroke length of the CrAl-coated Zircaloy-4. The wear resistance of the CrAl-coated Zircaloy-4 against the as-received Zr grid was improved significantly because of its high load support capability owing to the coating layer and the formation of a load-bearing layer by the material transfer when compared with the results of the uncoated Zircaloy-4. Furthermore, when the Zr grid was pre-oxidized, the wear amount of the CrAl-coated Zircaloy-4 became negligible owing to the formation of a glaze layer by the interaction with the surface oxide of the Zr grid. According to the results of the high-temperature steam oxidation tests, the defected region of the CrAl coating layer caused by the scratch and fretting wear showed no further significant damage and still acted as a protective layer for enhancing the accident tolerance of the fuel cladding.

Structural characterization of Zircaloy-4 subjected to helium ions irradiation of variable fluence

Zircaloy-4 is widely used as structural material for the fuel cladding of nuclear reactors due to its low neutron absorption cross-section, excellent corrosion resistance and good mechanical stability under neutron irradiation. In nuclear reactors, helium ions (He+) are produced during (n, α) reactions that can affect the structural performance of the Zircaloy-4. Therefore, it is important to investigate the structural properties of Zircaloy-4 after irradiation with helium ions. In this work, structural properties of He+irradiated Zircaloy-4 are investigated at different ion fluences. The samples of Zircaloy-4 were irradiated with 300 keV He+ at different fluences (1 × 1013, 1 × 1015 and 1 × 1016 ions/cm2) by using Pelletron Accelerator at room temperature. The ion range inside the samples was 0.84 µm as estimated through the SRIM software. Structural properties of the pristine and He+irradiated samples were studied through x-ray diffraction that revealed polycrystalline Zr peaks whose intensity and width were changed after irradiation. The diffraction peaks of Zr corresponding to (002) plane were found to be higher in intensity as compared to its other diffraction peaks. The changes in peaks positions at lower and higher diffraction angles were also noticed that was attributed to the generation of stresses inside the irradiated samples. The Williamson-Hall (W-H) analysis of the samples depicted a decrease in material's crystallinity after exposure to He+ up to 1 × 1015 ions/cm2 whereas the crystallinity was started to improve with increase of the ion fluence to 1 × 1016 ions/cm2. These results were explained on the basis of collisional cascades process and defects annihilation under the thermal spike model.

Comparison of electron backscatter and X-ray diffraction techniques for measuring dislocation density in Zircaloy-2

Two methods for measuring dislocation density were applied to a series of plastically deformed tensile samples of Zircaloy-2. Samples subjected to plastic strains ranging from 4 to 17% along a variety of loading paths were characterized using both electron backscatter diffraction (EBSD) and synchrotron X-ray line profile analysis (LPA). It was found that the EBSD-based method gave results which were similar in magnitude to those obtained by LPA and followed a similar trend with increasing plastic strain. The effects of microscope parameters and post-processing of the EBSD data on dislocation density measurements are also discussed. The typical method for estimating uncertainty in dislocation density measured via EBSD was shown to be overly conservative, and a more realistic method of determining uncertainty is presented as an alternative.

The effect of cold work on the transformation kinetics and texture of a zirconium alloy during fast thermal cycling

The effect of cold-work on the transformation kinetics and texture evolution after fast temperature cycling were studied in a dilute zirconium alloy. It was found that cold-work delays the onset of phase transformation and helps randomise the texture after the transformation. Samples of Zircaloy-4 in two conditions, cold-rolled to 70% reduction and fully recrystallised, were heated above the β-transus at a fast rate of 100 °C s−1 using resistive heating and without constraint. Electrical resistivity was used to measure the phase fraction during heating and electron back-scatter diffraction was used to measure the texture before and after the thermal cycle. Whereas previous work on titanium suggested that cold-work leads to texture strengthening after transformation, these new experiments show that, when the heating rate is fast enough, recrystallisation is incomplete on heating, which slows down the start of transformation, leading to a random α-texture after cooling. The texture in the recrystallised material after β heat-treatment is non-random but different from the start, a consequence of both a stronger β-texture after grain growth and stronger variant selection on cooling. This variant selection can be mostly explained by α-nucleation at special β-grain boundaries. These results demonstrate that prior deformation has a strong effect on the textures produced after β-heat treatment when heating rates are fast. This has implications for the anisotropy of nuclear cladding components after loss-of-coolant (LOCA) and reactivity-initiated accidents (RIA) but also more widely for the manufacturing of zirconium and titanium using welding or additive layer manufacturing where the heating rates are high.

H-D Analysis Employing Energy Transfer from Metastable Excited-State He in Double-Pulse LIBS with Low-Pressure He Gas.

A laser-induced-breakdown-spectroscopy (LIBS) experiment with a unique double-pulse setup and operated in low-pressure (3 kPa) He ambient gas is performed to study the detection of light elements, such as hydrogen (H) and deuterium (D), as well as elements of high excitation energies, such as fluorine (F) and chlorine (Cl), which are usually difficult to detect using ordinary LIBS techniques. A nanosecond Nd:YAG laser operated in its fundamental wavelength with energy of 54 mJ is focused onto the Al target to generate the He plasma. Another picosecond Nd:YAG laser operated in its fundamental wavelength with energy of 2 mJ is focused onto the sample surface and activated 2 μs before the operation of the nanosecond laser. The application to polyvinyl chloride (PVC) and polytetrafluoroethylene (PTFE) samples produces sharp and high-intensity Cl- and F-emission lines. Meanwhile, the sharp and well-resolved H-D-emission lines with merely 0.18 nm wavelength separation are also clearly detected from a zircaloy sample. Further measurement of a set of zircaloy samples containing different concentrations of D yields a linear calibration curve with a zero intercept. The detection limit of D is found to be about 10 ppm.

Progressive modelling and experimentation of hydrogen diffusion and precipitation in anisotropic polycrystals

Hydrogen embrittlement of zirconium alloys is a major concern in nuclear industry. It is known that diffusion of hydrogen atoms in zirconium lattice depends on the localized state of stresses, the patterning of which is significantly affected by the elastic and plastic anisotropy of zirconium crystals. This study focuses on developing a model to understand the effects of such localized stress fields on hydrogen diffusion and precipitation. For this purpose, a crystal plasticity finite element code is updated and coupled with a series of newly developed subroutines for simulating stress-assisted-diffusion of hydrogen atoms in zirconium polycrystals. An electron backscatter diffraction experiment is also conducted on a hydrided Zircaloy-2 sample to compare the measured distribution of the hydride phase with the calculated one. It is shown that even in the absence of macroscopic loads, thermal residual stresses within zirconium grains can disturb the uniform distribution of hydrogen atoms by inducing large hydrostatic stresses at the vicinity of grain boundaries. Such stress concentrations can further affect the sequence of hydrides nucleation. Also, it is shown that by alternating local stress fields, hydride induced transformation strain can change the distribution of the hydrogen atoms in zirconium matrix.

Release and speciation of carbon from Zircaloy-4 in anaerobic and highly alkaline conditions: Comparison of simple immersion and potentiostatic corrosion tests

ABSTRACTThe gas release and speciation of carbon species from irradiated and unirradiated Zircaloy-4 samples, representative for the fuel cladding as used in Belgian nuclear power plants, were studied in a saturated Ca(OH)2 solution in anaerobic conditions. This environment is relevant for the Belgian Supercontainer design, as perceived for the geological disposal of high-level nuclear waste. To achieve this, we performed simple immersion and potentiostatic corrosion tests. Potentiodynamic polarization curves, recorded prior to the potentiostatic tests, revealed that irradiation seems to induce changes on the Zircaloy-4 corrosion behavior, such as a shift of the corrosion potential. Potentiostatic corrosion tests on unirradiated Zircaloy-4 provided a corrosion rate of ~54 nm/yr over a 7 day-experiment, whilst a corrosion rate of only ~4 nm/yr was calculated for the irradiated sample. Gas chromatography revealed that during simple immersion tests, which lasted 195 days, hydrogen, methane, ethane, and CO2 were produced, with methane being the major compound. Assuming that all carbon released from the metal was transformed into gaseous compounds, this yields to a corrosion rate ranging from 57 to 84 nm/yr for the irradiated sample. However, caution has to be taken on these corrosion rate and more tests should be performed to confirm these results.

Evaluation of mechanical properties of irradiated zirconium alloys in the vicinity of the metal-oxide interface

The mechanical properties of oxides play an important role on the protection of underlying metals from further oxidation. Such properties should be examined to provide further mechanistic understanding of oxidation phenomena. This paper presents a novel method for determining the mechanical properties of material at the two sides of the metal-oxide (M/O) interface. It is applied to Zircaloy-2 cladding from a boiling water reactor. Two samples irradiated for 3 and 9 years have been examined. Micropillars were fabricated in the vicinity of M/O interface in the oxide, metal, and transition region using focused ion beam (FIB). The pillars were then uniaxially compressed using nano-indentation. Results prove the ability to pinpoint and consistently measure Young's modulus and yield strength of specific regions around the M/O interface. The Young's modulus and yield strength of metal and oxide regions for the two Zircaloy-2 samples are reported.

Effect of temperature on the selection of deformation modes in Zircaloy-4

The effect of deformation temperature (77, 298 and 573 K) on the evolution of slip and twinning modes in Zircaloy-4 was studied. Samples of Zircaloy-4 were uniaxially compressed to different strains at a nominal strain rate of 10−3 s−1. The compression direction was chosen so as to activate twinning. Microstructural examination was carried out using electron backscatter diffraction to relate the evolution of microstructure and texture to the micro-mechanics and work-hardening behaviour. In conjunction with experiments a visco-plastic self-consistent crystal plasticity model was used to determine the evolution of deformation modes. Extensive twinning was observed at all the test temperatures. Twinning activity at 298 and 573 K was comparable, but was lower than that at 77 K. A strong basal texture parallel to the compressive direction was found at 298 and 573 K. The VPSC model predicted well the texture evolution at 298 and 573 K. The basal texture developed at 77 K was found to have partially reoriented back to the initial texture due to secondary twinning. Prism slip was the dominant deformation mode with secondary slip becoming more active at higher temperatures.

Effects of an alloying element on a c-component loop formation and precipitate resolution in Zr alloys during ion irradiation

ABSTRACT It is important to clarify the mechanisms of the dislocation loop formation, dissolution of precipitates to understand the degradation behavior of the fuel cladding tubes in light water reactors (LWR) under neutron irradiation. In this study, 3.2 MeV Ni ion irradiation was carried out at 400°C on Zircaloy-2 and two types of model alloys with and without Fe (Zr-1.5Sn-0.3Fe and Zr-1.5Sn). To understand the effects of hydrogen, 60 and 300 ppm pre-injected Zircaloy-2 samples were also irradiated. The microstructure was observed with a conventional transmission electron microscopy. Additionally, the dissolution of precipitates and the enrichment of the alloying element due to irradiation were analyzed using a spherical aberration (Cs)-corrected high-resolution analytical electron microscope. After ion irradiation at 400°C, the dissolution of Fe-enriched precipitates and the c-component dislocation loops were observed in the region of peak ion damage. Observations by STEM-EDS showed that Fe atoms were enriched in the c-component dislocation loops. On the contrary, the c-component dislocation loops were detected in Fe-containing alloys (Zircaloy-2 and Zr-1.5Sn-0.3Fe alloy) but were not in the Zr-1.5Sn alloy. These results indicate that the dissolution of Fe-enriched precipitates and the enhanced formation of c-component dislocation loops are essential for the degradation of LWR fuel cladding under irradiation.

Deformation-induced phase transformation in Zircaloy-4 below the beta transus

Dynamic transformation is shown to take place in steels and titanium alloys at temperatures above and below the beta transus, respectively. Here the hard phase transforms into a softer phase during deformation. This unusual metallurgical phenomenon has previously been explained in terms of transformation softening (Ghosh et al., 2016). In the present work, compression tests were performed on samples of Zircaloy-4 at temperatures below the beta transus. The microstructures and microhardness measurements indicate that alpha transforms displacively into beta during deformation at temperatures down to about 75 °C below the transus.

Nondestructive hydrogen analysis of steam-oxidized Zircaloy-4 by wide-angle neutron scattering

A nondestructive neutron scattering method to precisely measure the hydrogen content in high-temperature steam-oxidized Zircaloy-4 cladding was developed. Zircaloy-4 cladding was used to produce hydrided specimens with hydrogen content up to ≈500 wppm. Following hydrogen charging, the hydrogen content of the hydrided specimens was measured using the vacuum hot extraction method, by which the samples with desired hydrogen concentrations were selected for the neutron study. The hydrided samples were then oxidized in steam up to ≈6.0 wt. % at 1100 °C. Optical microscopy shows that our hydriding procedure results in uniform distribution of circumferential hydrides across the wall thickness, and uniform oxide layers were formed on the sample surfaces by the steam oxidation. Small- and wide-angle neutron scattering were simultaneously performed to provide a quick (less than an hour per sample) measurement of the hydrogen content in various types of hydrided and oxidized Zircaloy-4. Our study demonstrates that the hydrogen in pre-oxidized Zircaloy-4 cladding can be measured very accurately by both small- and wide-angle neutron scattering. For steam-oxidized samples, the small-angle neutron scattering is contaminated with coherent scattering from additional structural features induced by the steam oxidation. However, the scattering intensity of the wide-angle neutron scattering increases proportionally with the hydrogen charged in the samples. The hydrogen content and wide-angle neutron scattering intensity are highly linearly correlated for the oxidized cladding samples examined in this work, and can be used to precisely determine the hydrogen content in steam-oxidized Zircaloy-4 samples. Hydrogen contents determined by neutron scattering of oxidation samples were also found to be consistent with the results of chemical analysis within acceptable margins for error.

High- and Low-Temperature Deformation Behavior of Different Orientation Hot-Rolled Annealed Zircaloy-4

In this paper, the hot-rolled annealed Zircaloy-4 samples with different orientation were subjected to uniaxial compression with a strain rate of 0.001 s−1 to obtain the stress-strain curves of different initial orientation samples at different temperatures. Electron backscatter diffraction (EBSD) technique and transmission electron microscope (TEM) technique were used to analyze the microstructures and textures of compressed samples. The mechanical properties and microstructural evolution of rolling directions (RD), transverse directions (TD) and normal directions (ND) were investigated under the conditions of – 150 °C low temperature, room temperature and 200 °C high temperature (simulated lunar temperature environment). The results show that the strength of Zircaloy-4 decreases with the increase in deformation temperature, and the strength in three orientations is ND > TD > RD. The deformation mechanism of hot-rolled annealed Zircaloy-4 with different orientation is different. In RD, $$\{ 10\bar{1}0\}$$ $$\left\langle {\text{a}} \right\rangle$$ prismatic slip has the highest Schmid factor (SF), so it is most easy to activate the slip, followed by TD orientation, and ND orientation is the most difficult to activate. The deformed grains abide slip→twinning→slip rule, and the different orientation Zircaloy-4 deformation mechanisms mainly are the twinning coordinated with the slip.

Positron annihilation analysis of the atomic scale changes in oxidized Zircaloy-4 samples

The process of oxidation in zirconium alloys is not straightforward. It includes cyclic periods, where the oxidation rate is decelerating as a function of increasing oxide thickness. Transition to the higher oxidation rate happens approximately after every 2 μm oxide growth and each cycle is associated with a layer of oxide with little or no cracking and a similar thickness layer with a large number of micro-scale lateral cracks. This work uses positron annihilation Doppler broadening spectroscopy to investigate the development of atomic-scale defect features in the oxides on Zircaloy-4 samples exposed to alkaline water at 350 °C. Results from the Doppler broadening measurements supported by Density Functional Theory calculations indicate that complexes involving both zirconium-site and oxygen-site vacancies exist in the oxide. The implications of the vacancies in the oxide layer are considered for oxygen diffusion and the oxidation rate.

Effect of He implantation on the microstructure of zircaloy-4 studied using in situ TEM

Zirconium alloys are of great importance to the nuclear industry as they have been widely used as cladding materials in light-water reactors since the 1960s. This work examines the behaviour of these alloys under He ion implantation for the purposes of developing understanding of the fundamental processes behind their response to irradiation. Characterization of zircaloy-4 samples using TEM with in situ 6 keV He irradiation up to a fluence of 2.7×1017ions·cm−2 in the temperature range of 298 to 1148 K has been performed. Ordered arrays of He bubbles were observed at 473 and 1148 K at a fluence of 1.7×1017ions·cm−2 in αZr, the hexagonal compact (HCP) and in βZr, the body centred cubic (BCC) phases, respectively. In addition, the dissolution behaviour of cubic Zr hydrides under He irradiation has been investigated.

Oxygen segregation in pre-hydrided Zircaloy-4 cladding during a simulated LOCA transient

Oxygen and hydrogen distributions are key elements influencing the residual ductility of zirconium-based nuclear fuel cladding during the quench phase following a Loss Of Coolant Accident (LOCA). During the high temperature oxidation, a complex partitioning of the alloying elements is observed. A finite-difference code for solving the oxygen diffusion equations has been developed by Institut de Radioprotection et de Sûreté Nucléaire to predict the oxygen profile within the samples. The comparison between the calculations and the experimental results in the mixed α+β region shows that the oxygen diffusion is not accurately predicted by the existing modeling. This work aims at determining the key parameters controlling the average oxygen profile within the sample in the two-phase regions at 1200 °C. High temperature steam oxidation tests interrupted by water quench were performed using pre-hydrided Zircaloy-4 samples. Experimental oxygen distribution was measured by Electron Probe Micro-Analysis (EPMA). The phase distributions within the cladding thickness, was measured using image analysis to determine the radial profile of α(O) phase fraction. It is further demonstrated and experimentally checked that the α-phase fraction in these regions follows a diffusion-like radial profile. A new phase fraction modeling is then proposed in the cladding metallic part during steam oxidation. The modeling results are compared to a large set of experiments including the influence of exposure duration and hydrogen content. Another key outcome from this modeling is that oxygen average profile is straightforward derived from the proposed modeling.