Search

International audience; Zirconium alloys used as cladding tubes for the fuel of Pressurized Water Reactors can undergo high applied stress during power transients. In these conditions, biaxial loading may lead to plastic deformation of the cladding. A comprehensive understanding of the material mechanical response during loading-path change tests is an important step toward the prediction of the behavior in these specific conditions. Using a non-standard mechanical testing machine, loading-path change tests have been conducted at 623 K on as-received recrystallized Zircaloy-4 tubes. These tests consist of an axial tensile loading and unloading followed by an internal pressure (or pure hoop tension) loading and unloading. These tests are able to examine the kinematic and isotropic hardening components of the strain hardening behavior of the thin cladding tubes. The isotropic hardening is attributed to dislocation multiplication and dislocation-dislocation interactions. The kinematic hardening is attributed to the interaction of the grains with each other. A polycrystalline model has also been used and improved in order to simulate the tests. A good prediction of the isotropic and kinematic hardening is provided by the modeling.

International audience; Irradiation damage in fuel cladding material is mainly caused by the neutron flux that results from fission reactions occurring in the fuel. To avoid the constraints inherent in handling radioactive material, the irradiation effects on the corrosion resistance of zirconium alloys can be studied by irradiating the materials with ions. We performed an original experiment using ion irradiation to more specifically study the influence of irradiation damage in the oxide on the corrosion rate of M5®. It has been established that irradiation with a 1.3-MeV helium ion at a fluence of 1017 cm−2 results in significant modifications of oxide properties, oxygen diffusion flux, and oxidation kinetics, as evidenced by Raman spectroscopy, secondary ion mass spectrometry (SIMS) analyses, and measurements of mass gains. A newly identified Raman vibration band at 712 cm−1 was linked to the presence of irradiation defects and allowed the evolution of their concentrations to be followed. The oxygen diffusion flux was significantly reduced after irradiation partly due to a surface concentration decrease of oxygen. The defects remained present in the oxide after 100 days of annealing in pressurized water reactor (PWR) conditions and were thus very stable in PWR conditions, which probably means that these defects would be stable in the reactor. According to the kinetics and in agreement with the results obtained by SIMS analyses, the oxidation rate was significantly reduced after ion irradiation, and this effect remained beyond 100 days in agreement with the high stability of irradiation defects in PWR conditions. An original model described quite well the oxidation kinetic results.

The desire to improve the corrosion resistance of Zr cladding material to allow high burnup has resulted in a general trend among fuel manufacturers to develop alloys with reduced levels of Sn. While the detrimental effect of Sn on high temperature aqueous corrosion performance is widely accepted, the reason for it remains unclear. High-Energy synchrotron X-ray diffraction was used to characterise the oxides formed by autoclave exposure on ZrSn-Nb alloys with tin concentrations ranging from 0.01 to 0.92 wt.%. The alloys studied included the commercial alloy ZIRLO®1 and two variants of ZIRLO with significantly lower tin levels, referred to here as A-0.6Sn and A-0.0Sn. The nature of the oxide grown on tube samples from each alloy during autoclave testing at 360°C was investigated by crosssectional Scanning and Transmission Electron Microscopy (SEM & TEM). Non-destructive synchrotron X-ray diffraction analysis on the oxides revealed that the monoclinic and tetragonal oxide phases display highly compressive in-plane residual stresses with the magnitudes dependent on both phase and alloy. Additional in situ Synchrotron X-ray diffraction experiments during oxidation at 550°C provided further confirmation of the 1 ZIRLO® is a registered trademark of Westinghouse Electric Company LLC in the United States and may be registered in other countries throughout the world. All rights reserved. Unauthorized use is strictly prohibited. 17th International Symposium on Zirconium in the Nuclear Industry 2 trends seen for autoclave tested samples and demonstrated the presence of elevated levels of tetragonal phase in the initial stages of oxidation. In situ and ex-situ measurements demonstrate unambiguously that the amount of tetragonal phase present and, more importantly, the degree of transformation from tetragonal to monoclinic oxide both decrease with decreasing tin levels, suggesting that tin stabilises the tetragonal phase. It is proposed that in Zr-Nb-Sn alloys with low Sn, the tetragonal phase is mainly stabilised by very small grain size and therefore remains stable throughout the corrosion process. By contrast, in alloys with higher tin levels larger, stress stabilised, tetragonal grains can form initially, but then transform as the corrosion front progresses inwards and stresses in the existing oxide relax.

The desire to improve the corrosion resistance of Zr cladding material to allow high burnup has resulted in a general trend among fuel manufacturers to develop alloys with reduced levels of Sn. While the detrimental effect of Sn on high temperature aqueous corrosion performance is widely accepted, the reason for it remains unclear. High-Energy synchrotron X-ray diffraction was used to characterise the oxides formed by autoclave exposure on Zr-Sn-Nb alloys with tin concentrations ranging from 0.01 to 0.92 wt.%. The alloys studied included the commercial alloy ZIRLO® and two variants of ZIRLO with significantly lower tin levels, referred to here as A-0.6Sn and A-0.0Sn. The nature of the oxide grown on tube samples from each alloy during autoclave testing at 360°C was investigated by cross-sectional Scanning and Transmission Electron Microscopy (SEM & TEM). Non-destructive synchrotron X-ray diffraction analysis on the oxides revealed that the monoclinic and tetragonal oxide phases display highly compressive in-plane residual stresses with the magnitudes dependent on both phase and alloy. Additional in-situ Synchrotron X-ray diffraction experiments during oxidation at 550°C provided further confirmation of the trends seen for autoclave tested samples and demonstrated the presence of elevated levels of tetragonal phase in the initial stages of oxidation. In-situ and ex-situ measurements demonstrate unambiguously that the amount of tetragonal phase present and, more importantly, the degree of transformation from tetragonal to monoclinic oxide both decrease with decreasing tin levels, suggesting that tin stabilises the tetragonal phase. It is proposed that in Zr-Nb-Sn alloys with low Sn, the tetragonal phase is mainly stabilised by very small grain size and therefore remains stable throughout the corrosion process. By contrast, in alloys with higher tin levels larger, stress stabilised, tetragonal grains can form initially, but then become unstable as the corrosion front progresses inwards and stresses in the existing oxide relax.

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 415C 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 36°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 360C displays a substantially different chemistry and microstructure compared to the base alloy with Fe segregating to the Zr/ZrO interface. Copyright © 2011 by ASTM International.

The previous CEA corrosion code COCHISE provided satisfactory simulations of in-reactor corrosion of the fuel cladding when used in its validity range. In contrast, it could lead to hazardous predictions if applied out of this range due to the strongly linked parameters mainly based on the analysis of French pressurized water reactor (PWR) data. To predict the oxidation kinetics for new operating conditions or new materials, the CEA and EDF decided to develop a new model, named CORCY, which is based on a more phenomenological approach and uses separate parameters deduced from analytical experiments. The aim of this paper is to present the new model for Zircaloy-4 in PWR. The phenomenological approach is described. It is based on out- and in-pile data. Typically, since (1) the oxidation kinetics of zirconium alloys in autoclave are periodic, and (2) the oxide films formed in autoclave, in out-of-pile loop, and in-reactor all exhibit periodic lateral cracks with a period similar to the oxide thickness to transition, the oxidation kinetics in CORCY are characterized by a cyclic repetition of semi-parabolic law. Each model parameter is detailed. They are deduced separately from (a) oxidation tests performed in autoclave on fresh alloys to determine their kinetics; (b) oxidation tests performed in the out-of-pile corrosion loops Corail and Reggae to quantify the effects of thermo-hydraulic conditions; (c) data provided by Testing Material Reactors (OSIRIS and Halden reactor) during isothermal oxidation to determine the effects of irradiation; and (d) oxidation tests performed on pre-hydrided alloys to take into account the accelerated corrosion phase occurring, in reactor, for Zircaloy-4 at high burn-up. After describing both the phenomenological approach and the different model parameters, a comparison of measured and calculated corrosion data from French PWRs at a burn-up up to 65 GWd/tU is provided.

Until now, most of the detailed characterizations of the M5 corrosion behaviour were performed under standard PWR operating conditions, under moderate Li content and moderate temperature of the primary coolant. In this study, in addition to these standard conditions, two demanding operating conditions were explored: increased Li chemistry and elevated temperature. The objective is to establish whether these more demanding conditions have an impact on the structure of the oxide layers formed, on Nb, Li and B contents in these layers and on Hydrogen pickup of the cladding. The structure of oxide layers was studied by microscopy, the Nb content and distribution by Electron Probe Micro Analysis, the Li and B contents and distributions by Nuclear Reaction Analysis and the hydrogen pickup by gas extraction. It was observed that the stability of the corrosion behaviour of M5 is not affected by increased Li or elevated temperature conditions. The hydrogen pickup fraction of M5 is not modified by increased Li conditions or by irradiation temperature with measured contents (

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.

In pressurized water reactors, new operating conditions (higher burnup, new chemistry, etc.) can have an effect on the dimensional stability of the fuel assembly skeleton. Previous studies have shown that the fuel assembly growth is, among others parameters (free growth, creep), strongly driven by corrosion. Oxide layer and hydrides precipitation could both induce an increase of the dimensional parameters. The scope of the present study is, regardless of irradiation effect, to quantify and to understand the separate effects of hydrogen and oxide layers on the Zircaloy-4 dimensional changes. Experimental works have been performed in laboratory on stress relieved annealed (SRA) and recrystallized (RXA) Zircaloy-4 strips. First, the hydrogen impact on dimensional changes has been studied without the effect of the oxide layer. The measurements were performed at room temperature on strips previously pre-hydrided by the gaseous charging method. The hydrogen content of the samples was between 100 ppm and 2000 ppm. Results indicate a linear correlation between hydrogen content and length variation. RXA material is more affected by the hydrogen effect than the SRA material. Nevertheless, in comparison with data issued from out-of-reactor measurement, the impact of hydrides is not sufficient (with irradiation growth) to explain the post-irradiation examinations (PIE) results. To understand these differences, the oxide layer contribution must be quantified. Second, the impact of the oxide layer was therefore studied on RXA Zircaloy-4 strips. Corrosion tests have been performed in autoclave at 360°C in primary water (2 ppm Li-1000 ppm B–H2) on as-received and pre-hydrided materials. To obtain thicker oxide layer within a shorter duration, samples have been also oxidized in furnace at 415°C. Moreover, as no significant hydriding occurs during oxidation in air, we are able to characterize properly the specific effect of the oxide layer. As for hydrides’ effect, an increase of strain is observed as the oxide thickness becomes thicker. The contributions of hydrides and oxide layer are then discussed with regard to the metallurgical properties of the alloy. Finally, all these results are compared with PIE observations. Free growth, hydride precipitation, and oxide thickness seem to be the three main parameters to explain the dimensional changes in Zircaloy-4 observed in reactor.