33 research outputs found

    Mécanismes de déformation et effets d’irradiation dans les alliages de zirconium. Une étude multi-échelle

    No full text
    Zirconium alloys have been used for more than 30 years in the nuclear industry as structural materials for the fuel assemblies of pressurized water reactors. In particular, the cladding tube, made of zirconium alloys, constitutes the first barrier against the dissemination of radioactive elements. It is therefore essential to have a good understanding and prediction of the mechanical behavior of these materials in various conditions. The work presented in this dissertation deals with an experimental study and numerical simulations, at several length scales, of the deformation mechanisms and the mechanical behavior of zirconium alloys before irradiation, but also after irradiation and under irradiation. The mechanical behavior of zirconium single crystal has been determined, during an original study, using tensile test specimens containing large grains. Based on this study, crystal plasticity constitutive laws have been proposed. A polycrystalline model has also been developed to simulate the behavior of unirradiated zirconium alloys. A thorough Transmission Electron Microscopy (TEM) study has been able to clarify the deformation mechanisms of zirconium alloys occurring after irradiation. The clearing of loops by gliding dislocations leading to the dislocation channeling mechanism has been studied in details. This phenomenon has also been simulated using a dislocation dynamics code. The macroscopic consequences of this process have also been analyzed. A polycrystalline model taking into account the specificity of this mechanism has eventually been proposed. This approach has then been extended to the post-irradiation creep behavior. The recovery of radiation defects during creep tests has been characterized by TEM and modeled using cluster dynamics method. Deformation modes during creep have also been studied and a simple model for the creep behavior has eventually been proposed. Finally, the mechanism responsible for the acceleration of irradiation growth that occurs at high doses, the nucleation and growth of loops, has been particularly studied. The effects of the hydrogen pick up and of an external applied stress on loops have been characterized by TEM. This work, which already contributes to a better understanding of deformation mechanisms and mechanical behavior of zirconium alloys, should improve, in a future prospect, the reliability and performance of pressurized water reactors fuel assemblies.Les alliages de zirconium sont utilis√©s depuis plus de 30 ans dans l‚Äôindustrie nucl√©aire comme mat√©riaux de structure de l‚Äôassemblage combustible des R√©acteurs √† Eau Pressuris√©e. En particulier, les gaines des crayons combustibles, en alliages de zirconium, constituent la premi√®re barri√®re de s√©curit√© vis-√†-vis de la diss√©mination d'√©l√©ments radioactifs. La bonne compr√©hension et pr√©diction du comportement m√©canique de ces mat√©riaux dans diff√©rentes situations s‚Äôav√®re donc essentielle. Les travaux, pr√©sent√©s dans ce m√©moire d‚ÄôHabilitation √† Diriger des Recherches, portent sur l‚Äô√©tude exp√©rimentale et la simulation num√©rique, √† diff√©rentes √©chelles, des m√©canismes de d√©formation et du comportement m√©canique des alliages de zirconium non irradi√©s, apr√®s irradiation et √©galement sous irradiation. Le comportement monocristallin du zirconium a √©t√© d√©termin√© gr√Ęce √† une √©tude exp√©rimentale originale utilisant des √©prouvettes contenant des grands grains. Un mod√®le de plasticit√© cristalline d√©crivant ce comportement a √©t√© propos√©. Un mod√®le polycristallin permettant de simuler le comportement des alliages de zirconium non irradi√©s a √©galement √©t√© d√©velopp√©. Une √©tude exp√©rimentale approfondie par Microscopie Electronique en Transmission (MET) a pr√©cis√© les m√©canismes de d√©formation des alliages de zirconium apr√®s irradiation. Le m√©canisme de balayage des boucles et de canalisation des dislocations a √©t√© particuli√®rement √©tudi√©. Ce ph√©nom√®ne a √©t√© simul√© par dynamique des dislocations. Les cons√©quences macroscopiques de ce processus ont √©t√© √©galement analys√©es. Un mod√®le polycristallin int√©grant les sp√©cificit√©s de ce m√©canisme a √©t√© finalement propos√©. Cette d√©marche a ensuite √©t√© √©tendue au comportement m√©canique en fluage thermique post-irradiation. Le recuit des d√©fauts d‚Äôirradiation qui se produit lors du fluage a √©t√© caract√©ris√© par MET et mod√©lis√© par une m√©thode de dynamique d‚Äôamas. Les modes de d√©formation lors du fluage ont √©t√© √©tudi√©s et un mod√®le simple de comportement en fluage a finalement √©t√© propos√©. Enfin, le m√©canisme responsable de l‚Äôacc√©l√©ration de croissance qui se produit sous irradiation √† forte dose, la nucl√©ation et croissance des boucles , a √©t√© particuli√®rement √©tudi√©. L‚Äôinfluence de la prise d‚Äôhydrog√®ne ainsi que l‚Äôinfluence d‚Äôune contrainte appliqu√©e sur les boucles ont √©t√© caract√©ris√©es par MET. Ces travaux, qui contribuent d√©j√† √† une meilleure compr√©hension des m√©canismes de d√©formation et du comportement m√©canique des alliages de zirconium, permettront, √† terme, d‚Äôam√©liorer la fiabilit√© et la performance des assemblages combustibles des r√©acteurs nucl√©aires √† eau pressuris√©e

    Mécanismes de déformation et effets d’irradiation dans les alliages de zirconium. Une étude multi-échelle

    No full text
    Zirconium alloys have been used for more than 30 years in the nuclear industry as structural materials for the fuel assemblies of pressurized water reactors. In particular, the cladding tube, made of zirconium alloys, constitutes the first barrier against the dissemination of radioactive elements. It is therefore essential to have a good understanding and prediction of the mechanical behavior of these materials in various conditions. The work presented in this dissertation deals with an experimental study and numerical simulations, at several length scales, of the deformation mechanisms and the mechanical behavior of zirconium alloys before irradiation, but also after irradiation and under irradiation. The mechanical behavior of zirconium single crystal has been determined, during an original study, using tensile test specimens containing large grains. Based on this study, crystal plasticity constitutive laws have been proposed. A polycrystalline model has also been developed to simulate the behavior of unirradiated zirconium alloys. A thorough Transmission Electron Microscopy (TEM) study has been able to clarify the deformation mechanisms of zirconium alloys occurring after irradiation. The clearing of loops by gliding dislocations leading to the dislocation channeling mechanism has been studied in details. This phenomenon has also been simulated using a dislocation dynamics code. The macroscopic consequences of this process have also been analyzed. A polycrystalline model taking into account the specificity of this mechanism has eventually been proposed. This approach has then been extended to the post-irradiation creep behavior. The recovery of radiation defects during creep tests has been characterized by TEM and modeled using cluster dynamics method. Deformation modes during creep have also been studied and a simple model for the creep behavior has eventually been proposed. Finally, the mechanism responsible for the acceleration of irradiation growth that occurs at high doses, the nucleation and growth of loops, has been particularly studied. The effects of the hydrogen pick up and of an external applied stress on loops have been characterized by TEM. This work, which already contributes to a better understanding of deformation mechanisms and mechanical behavior of zirconium alloys, should improve, in a future prospect, the reliability and performance of pressurized water reactors fuel assemblies.Les alliages de zirconium sont utilis√©s depuis plus de 30 ans dans l‚Äôindustrie nucl√©aire comme mat√©riaux de structure de l‚Äôassemblage combustible des R√©acteurs √† Eau Pressuris√©e. En particulier, les gaines des crayons combustibles, en alliages de zirconium, constituent la premi√®re barri√®re de s√©curit√© vis-√†-vis de la diss√©mination d'√©l√©ments radioactifs. La bonne compr√©hension et pr√©diction du comportement m√©canique de ces mat√©riaux dans diff√©rentes situations s‚Äôav√®re donc essentielle. Les travaux, pr√©sent√©s dans ce m√©moire d‚ÄôHabilitation √† Diriger des Recherches, portent sur l‚Äô√©tude exp√©rimentale et la simulation num√©rique, √† diff√©rentes √©chelles, des m√©canismes de d√©formation et du comportement m√©canique des alliages de zirconium non irradi√©s, apr√®s irradiation et √©galement sous irradiation. Le comportement monocristallin du zirconium a √©t√© d√©termin√© gr√Ęce √† une √©tude exp√©rimentale originale utilisant des √©prouvettes contenant des grands grains. Un mod√®le de plasticit√© cristalline d√©crivant ce comportement a √©t√© propos√©. Un mod√®le polycristallin permettant de simuler le comportement des alliages de zirconium non irradi√©s a √©galement √©t√© d√©velopp√©. Une √©tude exp√©rimentale approfondie par Microscopie Electronique en Transmission (MET) a pr√©cis√© les m√©canismes de d√©formation des alliages de zirconium apr√®s irradiation. Le m√©canisme de balayage des boucles et de canalisation des dislocations a √©t√© particuli√®rement √©tudi√©. Ce ph√©nom√®ne a √©t√© simul√© par dynamique des dislocations. Les cons√©quences macroscopiques de ce processus ont √©t√© √©galement analys√©es. Un mod√®le polycristallin int√©grant les sp√©cificit√©s de ce m√©canisme a √©t√© finalement propos√©. Cette d√©marche a ensuite √©t√© √©tendue au comportement m√©canique en fluage thermique post-irradiation. Le recuit des d√©fauts d‚Äôirradiation qui se produit lors du fluage a √©t√© caract√©ris√© par MET et mod√©lis√© par une m√©thode de dynamique d‚Äôamas. Les modes de d√©formation lors du fluage ont √©t√© √©tudi√©s et un mod√®le simple de comportement en fluage a finalement √©t√© propos√©. Enfin, le m√©canisme responsable de l‚Äôacc√©l√©ration de croissance qui se produit sous irradiation √† forte dose, la nucl√©ation et croissance des boucles , a √©t√© particuli√®rement √©tudi√©. L‚Äôinfluence de la prise d‚Äôhydrog√®ne ainsi que l‚Äôinfluence d‚Äôune contrainte appliqu√©e sur les boucles ont √©t√© caract√©ris√©es par MET. Ces travaux, qui contribuent d√©j√† √† une meilleure compr√©hension des m√©canismes de d√©formation et du comportement m√©canique des alliages de zirconium, permettront, √† terme, d‚Äôam√©liorer la fiabilit√© et la performance des assemblages combustibles des r√©acteurs nucl√©aires √† eau pressuris√©e

    Approche Expérimentale et Modélisation Micromécanique du Comportement des Alliages de Zirconium Irradiés

    No full text
    Zirconium alloys cladding tubes containing nuclear fuel of the Pressurized Water Reactors constitute the first safety barrier against the dissemination of radioactive elements. Thus, it is essential to predict the mechanical behavior of the material in-reactor conditions. This study aims, on the one hand, to identify and characterize the mechanisms of the plastic deformation of irradiated zirconium alloys and, on the other hand, to propose a micromechanical modeling based on these mechanisms. The experimental analysis shows that, for the irradiated material, the plastic deformation occurs by dislocation channeling. For transverse tensile test and internal pressure test this channeling occurs in the basal planes. However, for axial tensile test, the study revealed that the plastic deformation also occurs by channeling but in the prismatic and pyramidal planes. In addition, the study of the macroscopic mechanical behavior, compared to the deformation mechanisms observed by TEM, suggested that the internal stress is higher in the case of irradiated material than in the case of non-irradiated material, because of the very heterogeneous character of the plastic deformation. This analysis led to a coherent interpretation of the mechanical behavior of irradiated materials, in terms of deformation mechanisms. The mechanical behavior of irradiated materials was finally modeled by applying homogenization methods for heterogeneous materials. This model is able to reproduce adequately the mechanical behavior of the irradiated material, in agreement with the TEM observations.Les tubes en alliages de zirconium renfermant le combustible nucléaire des Réacteurs à Eau Pressurisée constituent la première barrière de sécurité vis-à-vis de la dissémination d'éléments radioactifs. Il est donc essentiel de garantir leur tenue mécanique en réacteur. Cette étude a pour objectifs d'une part d'identifier et caractériser les mécanismes de déformation plastique des alliages de zirconium irradiés, d'autre part de modéliser le comportement macroscopique sur la base des mécanismes identifiés. L'analyse expérimentale a mis en évidence que, sur matériau irradié, la déformation plastique se produit par canalisation des dislocations. Cette canalisation a lieu suivant les plans de base, pour des sollicitations de traction sens travers et de pression interne. En revanche, pour la sollicitation de traction axiale, l'étude a révélé que la canalisation se produit dans les plans prismatiques et pyramidaux. L'étude du comportement macroscopique, en lien avec les mécanismes de déformation observés en Microscopie Electronique en Transmission, a suggéré que la contrainte interne est plus élevée dans le cas du matériau irradié que dans le cas du matériau non irradié, du fait du caractère très hétérogène de la déformation. Cette analyse a permis d'interpréter de façon cohérente l'ensemble des caractéristiques du comportement du matériau irradié, en termes de mécanismes de déformation. Le comportement mécanique du matériau irradié a enfin été modélisé en appliquant les méthodes d'homogénéisation des matériaux hétérogènes. Ce modèle permet de reproduire l'ensemble des caractéristiques du comportement mécanique du matériau irradié, en accord avec les observations MET

    Approche expérimentale et modélisation micromécanique du comportement des alliages de zirconium irradiés

    No full text
    Les tubes en alliage de zirconium renfermant le combustible nucléaire des Réacteurs à Eau Pressurisée constituent la première barrière de sécurité vis-à-vis de la dissémination d'éléments radioactifs. Il est donc essentiel de garantir leur tenue mécanique en réacteur. Cette étude a pour objectifs d'une part d'identifier les mécanismes de déformation plastique des alliages de zirconium irradiés, d'autre part de modéliser le comportement macroscopique sur la base des mécanismes identifiés. L'analyse expérimentale a mis en évidence que, sur matériau irradié, la déformation plastique se produit par canalisation des dislocations. Cette canalisation a lieu suivant les plans de base, pour des sollicitations de traction sens travers et de pression interne. En revanche, pour la sollicitation de traction axiale, l'étude a révélé que la canalisation se produit dans les plans prismatiques et pyramidaux. L'étude du comportement macroscopique, en lien avec les mécanismes de déformation observés en Microscopie Electronique en Transmission, a suggéré que la contrainte interne est plus élevée dans le cas du matériau irradié que dans le cas du matériau non irradié, du fait du caractère très hétérogène de la déformation. Cette analyse a permis d'interpréter de façon cohérente l'ensemble des caractéristiques du comportement du matériau irradié, en termes de mécanismes de déformation. Le comportement mécanique du matériau irradié a enfin été modélisé en appliquant les méthodes d'homogénéisation des matériaux hétérogènes. Ce modèle permet de reproduire l'ensemble des caractéristiques du comportement mécanique du matériau irradié, en accord avec les observations MET.CHATENAY MALABRY-Ecole centrale (920192301) / SudocSudocFranceF

    Dislocation electron tomography: A technique to characterize the dislocation microstructure evolution in zirconium alloys under irradiation

    No full text
    International audienceDiffraction-contrast electron tomography was used to analyse the 3D geometry of the dislocation microstructure in a zirconium alloy before and after ion irradiation. The material had been strained at room temperature prior to irradiation. After straining, the material exhibited mainly screw dislocations with Burgers vectors. From the analysis of the habit plane of dislocations with non-screw segments, it was deduced that they have glided mainly in the prismatic planes and to a lesser extent in the first order pyramidal planes. After irradiation, dislocation loops with Burgers vectors were observed. It was shown that the loops are not pure edge and have their habit plane located around the planes {10-10}, tilted up to 20¬į towards the planes (0001) and {11-20}. Furthermore, it was proven that the initial screw dislocations have climbed under irradiation. Several dislocations were also found to have interacted with loops during climb. The climb of dislocations under irradiation is an important mechanism that can explain part of the in-reactor deformation of zirconium alloys when subjected to simultaneous mechanical loading and irradiation. Interactions between dislocations and loops occurring during dislocation climb may also play a significant role on the in-reactor deformation of zirconium alloys

    Polycrystalline simulations of in-reactor deformation of recrystallized Zircaloy-4 tubes: Fast Fourier Transform computations and mean-field self-consistent model

    No full text
    International audienceFuel cladding and structural components made of zirconium alloys, used in light and heavy water nuclear reactors, exhibit, during normal operation, significant in-reactor deformation. Fast Fourier Transform (FFT) simulations have been conducted on large grain aggregates to simulate the in-reactor behavior of recrystallized Zircaloy-4. Original constitutive equations have been proposed to account, at the microscopic scale, for thermal creep, irradiation creep and irradiation induced growth. The evolution of irradiation defects with irradiation is taken into account, especially to deduce the local growth strain. A good description of the in-reactor behavior is obtained with irradiation defects evolution consistent with Transmission Electron Microscopy observations. The FFT simulations are compared to a self-consistent model. A good agreement is obtained when the behavior is linear (irradiation creep and growth) while the nonlinear response (thermal creep) is underestimated by the self-consistent model. The FFT simulations are also compared to the lower-bound model which neglects the interactions between grains. The lower-bound model underestimates the growth strain proving the importance of using an accurate polycrystalline model to predict the growth strain from the knowledge of the irradiation defect evolution

    Polycrystalline simulations of in-reactor deformation of recrystallized Zircaloy-4 tubes: Fast Fourier Transform computations and mean-field self-consistent model

    No full text
    International audienceFuel cladding and structural components made of zirconium alloys, used in light and heavy water nuclear reactors, exhibit, during normal operation, significant in-reactor deformation. Fast Fourier Transform (FFT) simulations have been conducted on large grain aggregates to simulate the in-reactor behavior of recrystallized Zircaloy-4. Original constitutive equations have been proposed to account, at the microscopic scale, for thermal creep, irradiation creep and irradiation induced growth. The evolution of irradiation defects with irradiation is taken into account, especially to deduce the local growth strain. A good description of the in-reactor behavior is obtained with irradiation defects evolution consistent with Transmission Electron Microscopy observations. The FFT simulations are compared to a self-consistent model. A good agreement is obtained when the behavior is linear (irradiation creep and growth) while the nonlinear response (thermal creep) is underestimated by the self-consistent model. The FFT simulations are also compared to the lower-bound model which neglects the interactions between grains. The lower-bound model underestimates the growth strain proving the importance of using an accurate polycrystalline model to predict the growth strain from the knowledge of the irradiation defect evolution

    Mobility of „Äąc +a„ÄČ dislocations in zirconium

    No full text
    International audiencePlasticity in hexagonal close-packed zirconium is mainly controlled by the glide of dislocations with 1/3 Burgers vectors. As these dislocations cannot accommodate deformation in the [0001] direction , twinning or glide of dislocations, i.e. dislocations with 1/3 Burgers vector, have to be activated. We have performed in situ straining experiments in a transmission electron microscope to study the glide of dislocations in two different zirconium samples, pure zirconium and Zircaloy-4, at room temperature. These experiments show that dislocations exclusively glide in first-order pyramidal planes with cross-slip being activated. A much stronger lattice friction is opposing the glide of dislocations when their orientation corresponds to the direction defined by the intersection of their glide plane with the basal plane. This results in long dislocations straightened along which glide either viscously or jerkily. This direction governs the motion of segments with other orientations, whose shape is merely driven by the minimization of the line tension. The friction due to solute atoms is also discussed
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