During the corrosion in primary water of zirconium, hydrogen from the water diffuses through the ... more During the corrosion in primary water of zirconium, hydrogen from the water diffuses through the oxide. To better understand this process, we use Density Functional Theory with hybrid functionals to calculate the energetics of interstitial hydrogen ions in defect-free monoclinic zirconia. While there is only one stable site for hydride ions in zirconia, protons have four different sites. We calculate the migration paths and energies between insertion sites to obtain the diffusion coefficients of hydrogen. We find that protons diffuse orders of magnitude faster than hydride ions, proving that protons are responsible for diffusion of hydrogen in monoclinic zirconia.
This is a PDF file of an article that has undergone enhancements after acceptance, such as the ad... more This is a PDF file of an article that has undergone enhancements after acceptance, such as the addition of a cover page and metadata, and formatting for readability, but it is not yet the definitive version of record. This version will undergo additional copyediting, typesetting and review before it is published in its final form, but we are providing this version to give early visibility of the article. Please note that, during the production process, errors may be discovered which could affect the content, and all legal disclaimers that apply to the journal pertain.
Understanding the corrosion processes of fuel cladding in pressurized water reactors
Nuclear Corrosion, 2020
Abstract The cladding made of zirconium alloy provides the first containment barrier for fission ... more Abstract The cladding made of zirconium alloy provides the first containment barrier for fission products, which is why its mechanical integrity is a prerequisite for nuclear safety. The external corrosion of the zirconium alloy cladding is one of the factors limiting the fuel rod’s lifetime. Understanding the corrosion processes of the zirconium alloys is consequently very important industrial issue for safety and efficiency of light water reactors. This review of international works is divided into two parts, dealing with the oxidation behavior and the hydrogen pickup (HPU) of zirconium alloys in pressurized water reactor environments. The first one describes the growth mechanism of the oxide, the oxidation kinetics and its modeling, the crystallographic phases and microstructure of the oxide layer, the alloying element effect and the irradiation impact on the corrosion rate. The second part is focused on the HPU mechanism and the link between oxidation and hydriding processes.
Significant improvement in the safety of light water reactors should be achieved by the use of al... more Significant improvement in the safety of light water reactors should be achieved by the use of alternative cladding materials which are more robust than those currently serving as the first fuel containment barrier, in Loss of Coolant Accident (LOCA) and beyond-LOCA conditions. CEA is assessing the solution of a coated cladding aiming to protect the current zirconium alloy cladding from high temperature steam oxidation. The present paper particularly focuses on some recent results obtained at CEA on Cr coated Zircaloy-4 base materials that have experienced several hours exposure under steam at 1000°C ("post-breakaway" conditions for the bare Zircaloy-4 material).
Influence of water vapor on high-temperature oxidation of titanium and zirconium and their alloys
Titanium and zirconium are Group IV transition metals that possess the hexagonal crystal structur... more Titanium and zirconium are Group IV transition metals that possess the hexagonal crystal structure at room temperature. At temperatures above 865°C, Zr undergoes an allotropic transformation from the alpha to the beta phase, which has a face-centered cubic crystal structure. This transformation occurs at 882°C in Ti. The α-Zr and the α-Ti matrix can contain a high amount of dissolved oxygen at interstitial sites. Hydrogen can also be introduced into the metallic matrix, but this process is highly temperature-dependent. These properties are of great importance to the oxidation behavior of titanium and zirconium metals and alloys. Indeed, water vapor in the oxidizing environment does not only affect the oxide scale growth mechanism, but also influences the evolution of the substrate. Another distinctive characteristic has to do with the anion transport properties of titanium and zirconium oxides: stresses that develop at the metal/oxide interface and within the scale during its growth...
Understanding of Corrosion Mechanisms after Irradiation: Effect of Ion Irradiation of the Oxide Layers on the Corrosion Rate of M5
Irradiation damage in fuel cladding material is mainly caused by the neutron flux that results fr... more 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 dif...
Zirconium in the Nuclear Industry: 18th International Symposium, 2018
The corrosion process (oxidation and hydriding) of the zirconium alloy cladding is one of the lim... more The corrosion process (oxidation and hydriding) of the zirconium alloy cladding is one of the limiting factors on the fuel rod lifetime, in particular for the Zircaloy-4 alloy. The corrosion rate of this alloy shows indeed a great acceleration at high burn-up in Light Water Reactors. Understanding the corrosion behavior under irradiation for this alloy is an important technological issue for the safety and efficiency of LWRs. In particular, understanding the effect of irradiation on the metal and the oxide layers is a key parameter in the study of corrosion behavior of zirconium alloys. Zircaloy-4 samples have undergone helium and proton ion-irradiation up to 0.3 dpa forming a uniform defect distribution up to 1 µm deep. Both as-received and pre-corroded samples have been irradiated in order to compare the effect of metal irradiation to that of oxide layer irradiation. After irradiation, samples have been corroded in order to study the impact of irradiation defects in the metal and in preexisting oxide layers on the formation of new oxide layers. Synchrotron X-ray micro-diffraction and micro-fluorescence are used to follow the evolution of oxide crystallographic phases, texture and stoichiometry both in the metal and in the oxide in cross-section. In particular, the tetragonal oxide phase fraction, which has been known to have an important role on corrosion behavior, is mapped in both unirradiated and irradiated metal at the sub-micron scale and appears to be significantly affected by irradiation. These observations, complemented with electron microscopy analyses on samples in carefully chosen areas of interest are combined in order to fully characterize changes due to irradiation in metal and oxide phases of both alloys.
Early studies on Cr-Coated Zircaloy-4 as enhanced accident tolerant nuclear fuel claddings for light water reactors
Journal of Nuclear Materials
Abstract Coatings with thicknesses between a few microns and ∼10 μm deposited on a Zircaloy-4 su... more Abstract Coatings with thicknesses between a few microns and ∼10 μm deposited on a Zircaloy-4 substrate have been studied with the objective to provide a significant reduction in the oxidation-induced embrittlement of the nuclear fuel cladding, especially in accidental conditions, such as LOss-of-Coolant-Accident (LOCA) conditions. This paper deals with the early studies carried out at CEA, several years before the Fukushima-Daiishi events, on different types of coatings obtained by a physical vapor deposition process. The studied coatings included ceramic, nitride and metallic multi-layered ones. The results of this screening analysis showed that the first generation of chromium-based coatings exhibited the most promising behavior: good compromise between oxidation resistance and adhesion to the metallic substrate, good fretting resistance and improved resistance to oxidation in steam at high temperature (Design Basis Accident LOCA conditions and slightly beyond).
Effect of ion irradiation of the metal matrix on the oxidation rate of Zircaloy-4
Corrosion Science
Abstract The oxidation kinetics acceleration observed in Zircaloy-4 cladding in Pressurized Water... more Abstract The oxidation kinetics acceleration observed in Zircaloy-4 cladding in Pressurized Water Reactor could be partially due to metal irradiation damage occuring in-pile. To check this assumption, several ion irradiation tests of the Zircaloy-4 metal matrix were performed to reproduce the evolution of the metallurgical state in reactor. This study showed that the amorphization process of the intermetallic precipitates (SPPs) does not change significantly the oxidation rate of Zircaloy-4. Implantation of iron in the matrix to simulate iron dissolution from the SPPs has no real impact on the corrosion kinetics. However proton irradiations at 350 °C producing dislocation loops resulted in a significant increase in the oxidation rate.
O isotopic experiments). Moreover, during oxidation, the hydrogen initially present in the hydrid... more O isotopic experiments). Moreover, during oxidation, the hydrogen initially present in the hydride phase remains in the metallic matrix and leads to the allotropic transformation δ−ZrH 1 ,66 ε−ZrH 2 .
Monitoring of the microstructure of ion-irradiated nuclear ceramics by in situ Raman spectroscopy
Raman spectroscopy is an efficient technique for studying the evolution of microstructure of mate... more Raman spectroscopy is an efficient technique for studying the evolution of microstructure of materials under irradiation. For that purpose, a Raman spectrometer has been recently installed at the JANNUS-Saclay platform. In this paper, we describe the new setup for in situ experiments. These in situ experiments allowed following the microstructural evolution of different materials (SiC, ZrO 2 and B 4 C) as a function of ion fluence on a single sample (either single crystal or polycrystalline ceramics) under the same irradiation conditions. Our results show that Raman spectroscopy is a versatile non-contact technique for studying on-line crystalline phase changes or amorphization of irradiated iono-covalent solids. A detailed analysis of Raman spectra is provided for the three materials (SiC, ZrO 2 and B 4 C) investigated in this study, revealing quite different behaviors upon irradiation. Basically, Raman spectroscopy gives insight on these evolutions at the level of bonds given by s...
Finite element modelling of the oxidation kinetics of Zircaloy-4 with a controlled metal-oxide interface and the influence of growth stress
Corrosion Science, 2015
Abstract Experimentally, zirconium-based alloys oxidation kinetics is sub-parabolic, by contrast ... more Abstract Experimentally, zirconium-based alloys oxidation kinetics is sub-parabolic, by contrast with the Wagner theory which predicts a parabolic kinetics. Two finite element models have been developed to simulate this phenomenon: the diffuse interface model and the sharp interface model. Both simulate parabolic oxidation kinetics. The growth stress effects on oxygen diffusion are studied to try to explain the gap between theory and experience. Taking into account the influence of the hydrostatic stress and its gradient into the oxygen flux expression, sub-parabolic oxidation kinetics have been simulated. The sub-parabolic behaviour of the oxidation kinetics can be explained by a non-uniform compressive stress level into the oxide layer.
Influence of light ion irradiation of the oxide layer on the oxidation rate of Zircaloy-4
Corrosion Science, 2015
ABSTRACT A strong increase of the oxidation rate of Zircaloy-4 fuel cladding is observed for burn... more ABSTRACT A strong increase of the oxidation rate of Zircaloy-4 fuel cladding is observed for burnups above 35 GWd/MtU in PWR. This kinetic acceleration could be in large part due to the irradiation damage. To quantify the effect of the oxide layer irradiation on the Zircaloy-4 corrosion rate, a new approach was used. Defects created by light ion irradiation increases the oxidation rate of the alloy up to 23 days in autoclave in agreement with the irradiation defect annealing characterized by RAMAN spectroscopy. A simple model taking into account the annealing of the irradiation defect is proposed to explain the results.
The oxidation by oxygen of a zirconium based alloy, M5 TM (which is a ZrNbO alloy, containing 1% ... more The oxidation by oxygen of a zirconium based alloy, M5 TM (which is a ZrNbO alloy, containing 1% of Nb) has been studied. The M5 TM alloy, like many zirconium alloys, undergoes a kinetic transition. The aim of the present work is to achieve a better understanding of the oxidation in the pre-transition stage, and to clearly identify the differences between the pre-and post transition stages from the kinetic point of view. The oxidation of M5 TM was followed by isothermal gravimetry at 490°C, under a controlled partial pressure of oxygen (in the range 7 to 200hPa). First, we have verified the steady state assumption, by coupling thermogravimetry and differential scanning calorimetry (DSC) : it is shown that the system is in a steady state from the beginning of the oxidation, in the pre-and post-transition stages. Then, the existence of a rate-limiting step was verified in the pretransition stage using an experimental method based on temperature or pressure jumps; this assumption is no more verified in the post-transition stage, which means that the oxidation does not proceed in the same way as in the pre-transition stage. Finally, we have obtained the variations of zirconia growth reactivity Φ with the oxygen pressure, in the pre-transition stage (using the pressure jump method). The oxygen pressure has a slightly accelerating effect, which cannot be interpreted by the diffusion of oxygen vacancies through a dense oxide layer (in that case no effect of the oxygen pressure would be observed). In the post-transition stage, the oxygen effect is more important.
The kinetic curves of oxidation of Zircaloy-4 exhibit a transition, which is a sharp increase in ... more The kinetic curves of oxidation of Zircaloy-4 exhibit a transition, which is a sharp increase in the oxidation rate when the oxide thickness reaches a critical value. The pre-transition stage is controlled by the diffusion of oxygen vacancies in the oxide layer. In the posttransition stage, oxygen or water vapour have an accelerating effect on the oxidation (whereas they have no influence during the pre-transition) and the oxide layer is damaged, with large cracks parallel to the metal/oxide interface and connected to the gaseous atmosphere by pores. Consequently, it is clear that the post-transition stage cannot be accounted for by the same mechanism as in pre-transition. In this paper, we propose a geometrical modelling allowing to describe the progressive transformation of the oxide layer during the transition. This model is based on a random appearance of pores (connected to the external surface) which leads to the transformation, from a pre-transition stage to the post-transition stage, of small sections s0 of the oxide layer (analogy with the models of thermal transformations of powders, involving the processes of nucleation and growth of a new phase). The model allows to describe the kinetic curves obtained for the oxidation by water vapour of Zircaloy-4.
Significant improvement in the safety of light water reactors should be achieved by the use of al... more Significant improvement in the safety of light water reactors should be achieved by the use of alternative cladding materials which are more robust than those currently serving as the first fuel containment barrier, in Loss of Coolant Accident (LOCA) and beyond-LOCA conditions. CEA is assessing the solution of a coated cladding aiming to protect the current zirconium alloy cladding from high temperature steam oxidation. The present paper particularly focuses on some recent results obtained at CEA on Cr coated Zircaloy-4 base materials that have experienced several hours exposure under steam at 1000°C ("post-breakaway" conditions for the bare Zircaloy-4 material).
Imaging by Photoelectrochemical Techniques of Laves-Phases γ Zr (Fe, Cr) 2 Thermally Oxidized on Zircaloy-4
The oxidation of γ-Zr(Fe,Cr)2 intermetallic particles during the thermal exposition of Zircaloy-4... more The oxidation of γ-Zr(Fe,Cr)2 intermetallic particles during the thermal exposition of Zircaloy-4 at 470°C in oxygen was investigated with PhotoElectroChemical techniques (PEC). Via the measurement of bandgap, haematite Fe2O3 (2.2 eV), rhomboedric solid solution (FexCr1-x)2O3 (2.6 eV) and chromia Cr2O3 (3.0 eV) phases were identified as components of oxidised particles. Evolution of size, lateral distribution and density of these particles was studied in function of zirconia scale thickness. During the first stage of oxidation, the density of oxidised particles increased with thickness but decreased during a second stage, highlighting in an innovative way the phenomenon of haematite and chromia dissolution in the zirconia matrix. It is concluded that PEC techniques represent a sensitive and powerful way to locally analyse the various semiconductor phases in an oxide scale at the micron scale.
During the corrosion in primary water of zirconium, hydrogen from the water diffuses through the ... more During the corrosion in primary water of zirconium, hydrogen from the water diffuses through the oxide. To better understand this process, we use Density Functional Theory with hybrid functionals to calculate the energetics of interstitial hydrogen ions in defect-free monoclinic zirconia. While there is only one stable site for hydride ions in zirconia, protons have four different sites. We calculate the migration paths and energies between insertion sites to obtain the diffusion coefficients of hydrogen. We find that protons diffuse orders of magnitude faster than hydride ions, proving that protons are responsible for diffusion of hydrogen in monoclinic zirconia.
This is a PDF file of an article that has undergone enhancements after acceptance, such as the ad... more This is a PDF file of an article that has undergone enhancements after acceptance, such as the addition of a cover page and metadata, and formatting for readability, but it is not yet the definitive version of record. This version will undergo additional copyediting, typesetting and review before it is published in its final form, but we are providing this version to give early visibility of the article. Please note that, during the production process, errors may be discovered which could affect the content, and all legal disclaimers that apply to the journal pertain.
Understanding the corrosion processes of fuel cladding in pressurized water reactors
Nuclear Corrosion, 2020
Abstract The cladding made of zirconium alloy provides the first containment barrier for fission ... more Abstract The cladding made of zirconium alloy provides the first containment barrier for fission products, which is why its mechanical integrity is a prerequisite for nuclear safety. The external corrosion of the zirconium alloy cladding is one of the factors limiting the fuel rod’s lifetime. Understanding the corrosion processes of the zirconium alloys is consequently very important industrial issue for safety and efficiency of light water reactors. This review of international works is divided into two parts, dealing with the oxidation behavior and the hydrogen pickup (HPU) of zirconium alloys in pressurized water reactor environments. The first one describes the growth mechanism of the oxide, the oxidation kinetics and its modeling, the crystallographic phases and microstructure of the oxide layer, the alloying element effect and the irradiation impact on the corrosion rate. The second part is focused on the HPU mechanism and the link between oxidation and hydriding processes.
Significant improvement in the safety of light water reactors should be achieved by the use of al... more Significant improvement in the safety of light water reactors should be achieved by the use of alternative cladding materials which are more robust than those currently serving as the first fuel containment barrier, in Loss of Coolant Accident (LOCA) and beyond-LOCA conditions. CEA is assessing the solution of a coated cladding aiming to protect the current zirconium alloy cladding from high temperature steam oxidation. The present paper particularly focuses on some recent results obtained at CEA on Cr coated Zircaloy-4 base materials that have experienced several hours exposure under steam at 1000°C ("post-breakaway" conditions for the bare Zircaloy-4 material).
Influence of water vapor on high-temperature oxidation of titanium and zirconium and their alloys
Titanium and zirconium are Group IV transition metals that possess the hexagonal crystal structur... more Titanium and zirconium are Group IV transition metals that possess the hexagonal crystal structure at room temperature. At temperatures above 865°C, Zr undergoes an allotropic transformation from the alpha to the beta phase, which has a face-centered cubic crystal structure. This transformation occurs at 882°C in Ti. The α-Zr and the α-Ti matrix can contain a high amount of dissolved oxygen at interstitial sites. Hydrogen can also be introduced into the metallic matrix, but this process is highly temperature-dependent. These properties are of great importance to the oxidation behavior of titanium and zirconium metals and alloys. Indeed, water vapor in the oxidizing environment does not only affect the oxide scale growth mechanism, but also influences the evolution of the substrate. Another distinctive characteristic has to do with the anion transport properties of titanium and zirconium oxides: stresses that develop at the metal/oxide interface and within the scale during its growth...
Understanding of Corrosion Mechanisms after Irradiation: Effect of Ion Irradiation of the Oxide Layers on the Corrosion Rate of M5
Irradiation damage in fuel cladding material is mainly caused by the neutron flux that results fr... more 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 dif...
Zirconium in the Nuclear Industry: 18th International Symposium, 2018
The corrosion process (oxidation and hydriding) of the zirconium alloy cladding is one of the lim... more The corrosion process (oxidation and hydriding) of the zirconium alloy cladding is one of the limiting factors on the fuel rod lifetime, in particular for the Zircaloy-4 alloy. The corrosion rate of this alloy shows indeed a great acceleration at high burn-up in Light Water Reactors. Understanding the corrosion behavior under irradiation for this alloy is an important technological issue for the safety and efficiency of LWRs. In particular, understanding the effect of irradiation on the metal and the oxide layers is a key parameter in the study of corrosion behavior of zirconium alloys. Zircaloy-4 samples have undergone helium and proton ion-irradiation up to 0.3 dpa forming a uniform defect distribution up to 1 µm deep. Both as-received and pre-corroded samples have been irradiated in order to compare the effect of metal irradiation to that of oxide layer irradiation. After irradiation, samples have been corroded in order to study the impact of irradiation defects in the metal and in preexisting oxide layers on the formation of new oxide layers. Synchrotron X-ray micro-diffraction and micro-fluorescence are used to follow the evolution of oxide crystallographic phases, texture and stoichiometry both in the metal and in the oxide in cross-section. In particular, the tetragonal oxide phase fraction, which has been known to have an important role on corrosion behavior, is mapped in both unirradiated and irradiated metal at the sub-micron scale and appears to be significantly affected by irradiation. These observations, complemented with electron microscopy analyses on samples in carefully chosen areas of interest are combined in order to fully characterize changes due to irradiation in metal and oxide phases of both alloys.
Early studies on Cr-Coated Zircaloy-4 as enhanced accident tolerant nuclear fuel claddings for light water reactors
Journal of Nuclear Materials
Abstract Coatings with thicknesses between a few microns and ∼10 μm deposited on a Zircaloy-4 su... more Abstract Coatings with thicknesses between a few microns and ∼10 μm deposited on a Zircaloy-4 substrate have been studied with the objective to provide a significant reduction in the oxidation-induced embrittlement of the nuclear fuel cladding, especially in accidental conditions, such as LOss-of-Coolant-Accident (LOCA) conditions. This paper deals with the early studies carried out at CEA, several years before the Fukushima-Daiishi events, on different types of coatings obtained by a physical vapor deposition process. The studied coatings included ceramic, nitride and metallic multi-layered ones. The results of this screening analysis showed that the first generation of chromium-based coatings exhibited the most promising behavior: good compromise between oxidation resistance and adhesion to the metallic substrate, good fretting resistance and improved resistance to oxidation in steam at high temperature (Design Basis Accident LOCA conditions and slightly beyond).
Effect of ion irradiation of the metal matrix on the oxidation rate of Zircaloy-4
Corrosion Science
Abstract The oxidation kinetics acceleration observed in Zircaloy-4 cladding in Pressurized Water... more Abstract The oxidation kinetics acceleration observed in Zircaloy-4 cladding in Pressurized Water Reactor could be partially due to metal irradiation damage occuring in-pile. To check this assumption, several ion irradiation tests of the Zircaloy-4 metal matrix were performed to reproduce the evolution of the metallurgical state in reactor. This study showed that the amorphization process of the intermetallic precipitates (SPPs) does not change significantly the oxidation rate of Zircaloy-4. Implantation of iron in the matrix to simulate iron dissolution from the SPPs has no real impact on the corrosion kinetics. However proton irradiations at 350 °C producing dislocation loops resulted in a significant increase in the oxidation rate.
O isotopic experiments). Moreover, during oxidation, the hydrogen initially present in the hydrid... more O isotopic experiments). Moreover, during oxidation, the hydrogen initially present in the hydride phase remains in the metallic matrix and leads to the allotropic transformation δ−ZrH 1 ,66 ε−ZrH 2 .
Monitoring of the microstructure of ion-irradiated nuclear ceramics by in situ Raman spectroscopy
Raman spectroscopy is an efficient technique for studying the evolution of microstructure of mate... more Raman spectroscopy is an efficient technique for studying the evolution of microstructure of materials under irradiation. For that purpose, a Raman spectrometer has been recently installed at the JANNUS-Saclay platform. In this paper, we describe the new setup for in situ experiments. These in situ experiments allowed following the microstructural evolution of different materials (SiC, ZrO 2 and B 4 C) as a function of ion fluence on a single sample (either single crystal or polycrystalline ceramics) under the same irradiation conditions. Our results show that Raman spectroscopy is a versatile non-contact technique for studying on-line crystalline phase changes or amorphization of irradiated iono-covalent solids. A detailed analysis of Raman spectra is provided for the three materials (SiC, ZrO 2 and B 4 C) investigated in this study, revealing quite different behaviors upon irradiation. Basically, Raman spectroscopy gives insight on these evolutions at the level of bonds given by s...
Finite element modelling of the oxidation kinetics of Zircaloy-4 with a controlled metal-oxide interface and the influence of growth stress
Corrosion Science, 2015
Abstract Experimentally, zirconium-based alloys oxidation kinetics is sub-parabolic, by contrast ... more Abstract Experimentally, zirconium-based alloys oxidation kinetics is sub-parabolic, by contrast with the Wagner theory which predicts a parabolic kinetics. Two finite element models have been developed to simulate this phenomenon: the diffuse interface model and the sharp interface model. Both simulate parabolic oxidation kinetics. The growth stress effects on oxygen diffusion are studied to try to explain the gap between theory and experience. Taking into account the influence of the hydrostatic stress and its gradient into the oxygen flux expression, sub-parabolic oxidation kinetics have been simulated. The sub-parabolic behaviour of the oxidation kinetics can be explained by a non-uniform compressive stress level into the oxide layer.
Influence of light ion irradiation of the oxide layer on the oxidation rate of Zircaloy-4
Corrosion Science, 2015
ABSTRACT A strong increase of the oxidation rate of Zircaloy-4 fuel cladding is observed for burn... more ABSTRACT A strong increase of the oxidation rate of Zircaloy-4 fuel cladding is observed for burnups above 35 GWd/MtU in PWR. This kinetic acceleration could be in large part due to the irradiation damage. To quantify the effect of the oxide layer irradiation on the Zircaloy-4 corrosion rate, a new approach was used. Defects created by light ion irradiation increases the oxidation rate of the alloy up to 23 days in autoclave in agreement with the irradiation defect annealing characterized by RAMAN spectroscopy. A simple model taking into account the annealing of the irradiation defect is proposed to explain the results.
The oxidation by oxygen of a zirconium based alloy, M5 TM (which is a ZrNbO alloy, containing 1% ... more The oxidation by oxygen of a zirconium based alloy, M5 TM (which is a ZrNbO alloy, containing 1% of Nb) has been studied. The M5 TM alloy, like many zirconium alloys, undergoes a kinetic transition. The aim of the present work is to achieve a better understanding of the oxidation in the pre-transition stage, and to clearly identify the differences between the pre-and post transition stages from the kinetic point of view. The oxidation of M5 TM was followed by isothermal gravimetry at 490°C, under a controlled partial pressure of oxygen (in the range 7 to 200hPa). First, we have verified the steady state assumption, by coupling thermogravimetry and differential scanning calorimetry (DSC) : it is shown that the system is in a steady state from the beginning of the oxidation, in the pre-and post-transition stages. Then, the existence of a rate-limiting step was verified in the pretransition stage using an experimental method based on temperature or pressure jumps; this assumption is no more verified in the post-transition stage, which means that the oxidation does not proceed in the same way as in the pre-transition stage. Finally, we have obtained the variations of zirconia growth reactivity Φ with the oxygen pressure, in the pre-transition stage (using the pressure jump method). The oxygen pressure has a slightly accelerating effect, which cannot be interpreted by the diffusion of oxygen vacancies through a dense oxide layer (in that case no effect of the oxygen pressure would be observed). In the post-transition stage, the oxygen effect is more important.
The kinetic curves of oxidation of Zircaloy-4 exhibit a transition, which is a sharp increase in ... more The kinetic curves of oxidation of Zircaloy-4 exhibit a transition, which is a sharp increase in the oxidation rate when the oxide thickness reaches a critical value. The pre-transition stage is controlled by the diffusion of oxygen vacancies in the oxide layer. In the posttransition stage, oxygen or water vapour have an accelerating effect on the oxidation (whereas they have no influence during the pre-transition) and the oxide layer is damaged, with large cracks parallel to the metal/oxide interface and connected to the gaseous atmosphere by pores. Consequently, it is clear that the post-transition stage cannot be accounted for by the same mechanism as in pre-transition. In this paper, we propose a geometrical modelling allowing to describe the progressive transformation of the oxide layer during the transition. This model is based on a random appearance of pores (connected to the external surface) which leads to the transformation, from a pre-transition stage to the post-transition stage, of small sections s0 of the oxide layer (analogy with the models of thermal transformations of powders, involving the processes of nucleation and growth of a new phase). The model allows to describe the kinetic curves obtained for the oxidation by water vapour of Zircaloy-4.
Significant improvement in the safety of light water reactors should be achieved by the use of al... more Significant improvement in the safety of light water reactors should be achieved by the use of alternative cladding materials which are more robust than those currently serving as the first fuel containment barrier, in Loss of Coolant Accident (LOCA) and beyond-LOCA conditions. CEA is assessing the solution of a coated cladding aiming to protect the current zirconium alloy cladding from high temperature steam oxidation. The present paper particularly focuses on some recent results obtained at CEA on Cr coated Zircaloy-4 base materials that have experienced several hours exposure under steam at 1000°C ("post-breakaway" conditions for the bare Zircaloy-4 material).
Imaging by Photoelectrochemical Techniques of Laves-Phases γ Zr (Fe, Cr) 2 Thermally Oxidized on Zircaloy-4
The oxidation of γ-Zr(Fe,Cr)2 intermetallic particles during the thermal exposition of Zircaloy-4... more The oxidation of γ-Zr(Fe,Cr)2 intermetallic particles during the thermal exposition of Zircaloy-4 at 470°C in oxygen was investigated with PhotoElectroChemical techniques (PEC). Via the measurement of bandgap, haematite Fe2O3 (2.2 eV), rhomboedric solid solution (FexCr1-x)2O3 (2.6 eV) and chromia Cr2O3 (3.0 eV) phases were identified as components of oxidised particles. Evolution of size, lateral distribution and density of these particles was studied in function of zirconia scale thickness. During the first stage of oxidation, the density of oxidised particles increased with thickness but decreased during a second stage, highlighting in an innovative way the phenomenon of haematite and chromia dissolution in the zirconia matrix. It is concluded that PEC techniques represent a sensitive and powerful way to locally analyse the various semiconductor phases in an oxide scale at the micron scale.
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Papers by Marc Tupin