Teses / dissertações sobre o tema "Alliages à base de zirconium"
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Borroto, Ramírez Alejandro. "Synthesis, structure and properties of zirconium-based binary alloy thin films". Electronic Thesis or Diss., Université de Lorraine, 2019. http://www.theses.fr/2019LORR0057.
Texto completo da fonteIn this thesis, we demonstrate that original nanostructures can be obtained by working around the crystalline-to-amorphous transition in sputter-deposited thin films. In particular, we study two systems, Zr-Mo and Zr-W, in which such transition occurs. By decreasing the Mo content in the Zr-Mo system, a structural transition from a nanocrystalline solid solution of Zr in the bcc lattice of Mo to an amorphous structure can be achieved around 60 at% Mo. The films obtained present high hardness H, low Young's modulus E and, consequently, high H/E ratio compared with bulk Zr and Mo. Furthermore, we demonstrate that a self-separation of the nanocrystalline and the amorphous phases occurs at the composition intermediate to those necessary to form single-phased amorphous and nanocrystalline films. The particular geometry in which the nanocrystalline phase grows in competition with the amorphous phase is exploited to achieve a thickness-controlled surface morphology which allows to tune the film reflectance. A model was developed to describe the kinetics of the competitive growth between the nanocrystalline and the amorphous phases. Furthermore, it allows to construct a thickness-composition phase diagram evidencing that the nanocrystalline/amorphous competitive growth is easily hidden experimentally. Finally, we demonstrate that massive monocrystalline grains with lateral size larger than 1 µm can be obtained by working at low Ar pressure if the composition of the films approaches to the edge of the amorphous transition. Our results suggest that the phenomena reported here for Zr-Mo and Zr-W can be extended to other systems
Lafaye, Paul. "Développement d’outils de modélisation thermodynamique pour la prédiction de l’état métallurgique d’alliages à base zirconium". Thesis, Paris Est, 2017. http://www.theses.fr/2017PESC1125/document.
Texto completo da fonteCurrently, zirconium alloys are used as fuel cladding materials in PWR (Pressurized Water Reactors). The claddings stand in a very corrosive and radiative environnement, and can be submitted to temperature variations. In addition, the claddings will be subjected to mechanical stresses in reactor or accidental conditions. Thus, it appears useful to have a better understanding of phase transformations occurring in these alloys, as a function of temperature and chemical composition variations, but also to forecast the precipitation of fragile phases induced by the addition of alloying elements. At last, the ability to test new alloy compositions may allow to optimize it.The most suitable method for the thermodynamic modeling of multicomponent systems is the Calphad method (CALculation of PHAse Diagrams). The Calphad method is a widely used technique of semi-empirical modelling of phase diagrams. It consists in the description of the Gibbs energies of the different phases by fitting parameters allowing to describe the experimental data.This report details the design of a thermodynamic database considering the five following elements Zr, Cr, Fe, Nb, and Sn. The originality of this database lies in a systematic use of DFT calculations. Indeed, DFT calculations are performed to predict the formation enthalpy of the intermetallic phases appearing in these systems. Moreover, the SQS method (Special Quasirandom Structure) is used to predict the mixing enthalpy of the fcc, bcc and hcp binary solid solutions. Besides, experimental investigations are an important step of this thesis. Since no experimental data were available for the Cr-Fe-Sn, Cr-Nb-Sn, Cr-Sn-Zr and Fe-Nb-Sn ternary systems, new experimental data are provided, within this study, on the isothermal sections of these systems at different temperatures. All these calculated data in addition to the experimental data and the data from literature are used as input data for the Calphad modelling of the twenty binary and ternary systems which are then combined in the new database. A last part is dedicated to comparisons between predictions obtained with our new database and experimental results on industrial quinary alloys and a new concept of claddings
Toffolon-Masclet, Caroline. "Etude metallurgique et calculs de diagrammes de phases des alliages base zirconium du systeme : zr-nb-fe-(o,sn)". Paris 6, 2000. http://www.theses.fr/2000PA066457.
Texto completo da fonteOlier, Patrick. "Alliages à mémoire de forme de base TiNi : influence du mode de fabrication, de la teneur en oxygène et de l'ajout de zirconium ou d'hafnium sur les caractéristiques métallurgiques et les propriétés thermomécaniques /". Gif-sur-Yvette : Direction de l'information scientifique et technique, CEA Saclay, 1996. http://catalogue.bnf.fr/ark:/12148/cb35851337w.
Texto completo da fonteWu, Alexia. "Etude du comportement hors et sous irradiation aux ions d'un gainage combustible REP innovant base zirconium revêtu de chrome". Thesis, Paris 6, 2017. http://www.theses.fr/2017PA066611/document.
Texto completo da fonteIn Light Water Reactors (LWR) under hypothetical accidental conditions such as LOss of Coolant Accident (LOCA), zirconium alloy fuel claddings undergo significant oxidation at high temperatures. To limit this phenomenon, innovative chromium coated nuclear fuel claddings are studied at CEA. However, the integrity under neutron irradiation of such coating for in-service conditions must be preserved..The main objective of this PhD work is to study the behavior under ion irradiation of this new cladding concept. We especially focus on Zr/Cr interface microstructure evolution under irradiation, since the latter controls the adhesion of the coating to the substrate. Ion irradiations were performed to simulate the damage caused by neutrons in LWR. A multi-scale approach is used to characterize the samples before and after irradiation. In particular, Transmission Electron Microscopy (TEM) was used to characterize, at an atomic scale, the microstructure of the Zr/Cr interface. A first type of Zr/Cr interface is observed and shows the presence of nanometric phases of Zr(Fe,Cr)2 C14 and ZrCr2 C15 types. After irradiation, the C14 phase seems to be stabilized over the C15 phase, by segregation of iron at the interface. For a second interface, obtained using different deposition conditions, only C15 phase is observed. Under in-situ TEM irradiation, dissolution of the C15 phase is observed. Whatever the Zr/Cr interface type, preservation of the continuity of crystallographic planes before and after irradiation throughout the interface is demonstrated and thus induces a good adhesion of the coating to the substrate
Skocic, Milan. "Etude (photo)-électrochimique en réacteur simulé du phénomène de shadow corrosion des alliages de zirconium". Thesis, Université Grenoble Alpes (ComUE), 2016. http://www.theses.fr/2016GREAI015/document.
Texto completo da fonteConventional electrochemical methods as well as photoelectrochemical characterisations (PEC), performedex-situ et in-situ, were used to study the Shadow corrosion phenomenon, considered as a galvanic corrosion between Zr-based and Ni-based alloys. The Shadow corrosion is influenced by the chemical environment and the irradiation of these alloys. An electrochemical cell , simulating the conditions of a boiling water reactor (BWR), allowing the illumination of the samples with UV--Visible as well as monitoring the water chemistry was designed, developed and validated. The cell allowed, for the first time, recording of emph{in-situ} photocurrent energy spectra on a Zr-based alloy in simulated BWR environment. Furthermore, the obtained experimental results pointed out that the metallic cation impurities played an important role in the activation mechanism of the galvanic coupling, thus potentially in the activation mechanism of the Shadow corrosion phenomenon, whereas the presence oxygen and/or hydrogen peroxide did not induce significant differences in terms of electrochemical behavior of the samples. It was also shown that the illumination of the sample with UV--visible light, which significantly amplified the galvanic current, is an important parameter of the Shadow corrosion phenomenon
Pecheur, Dominique. "Evolution des précipités à base de zirconium lors de l'oxydation et de l'irradiation d'alliages de zirconium : impact sur la cinétique d'oxydation d'alliages de zirconium". Grenoble INPG, 1993. http://www.theses.fr/1993INPG0013.
Texto completo da fonteWu, Alexia. "Etude du comportement hors et sous irradiation aux ions d'un gainage combustible REP innovant base zirconium revêtu de chrome". Electronic Thesis or Diss., Paris 6, 2017. https://accesdistant.sorbonne-universite.fr/login?url=https://theses-intra.sorbonne-universite.fr/2017PA066611.pdf.
Texto completo da fonteIn Light Water Reactors (LWR) under hypothetical accidental conditions such as LOss of Coolant Accident (LOCA), zirconium alloy fuel claddings undergo significant oxidation at high temperatures. To limit this phenomenon, innovative chromium coated nuclear fuel claddings are studied at CEA. However, the integrity under neutron irradiation of such coating for in-service conditions must be preserved..The main objective of this PhD work is to study the behavior under ion irradiation of this new cladding concept. We especially focus on Zr/Cr interface microstructure evolution under irradiation, since the latter controls the adhesion of the coating to the substrate. Ion irradiations were performed to simulate the damage caused by neutrons in LWR. A multi-scale approach is used to characterize the samples before and after irradiation. In particular, Transmission Electron Microscopy (TEM) was used to characterize, at an atomic scale, the microstructure of the Zr/Cr interface. A first type of Zr/Cr interface is observed and shows the presence of nanometric phases of Zr(Fe,Cr)2 C14 and ZrCr2 C15 types. After irradiation, the C14 phase seems to be stabilized over the C15 phase, by segregation of iron at the interface. For a second interface, obtained using different deposition conditions, only C15 phase is observed. Under in-situ TEM irradiation, dissolution of the C15 phase is observed. Whatever the Zr/Cr interface type, preservation of the continuity of crystallographic planes before and after irradiation throughout the interface is demonstrated and thus induces a good adhesion of the coating to the substrate
Tcheliebou, Frederic. "Etude des alliages d'oxydes à base de ZrO2 obtenus par évaporation thermique réactive au double canon à électrons". Montpellier 2, 1995. http://www.theses.fr/1995MON20037.
Texto completo da fonteVERMOYAL, JEAN-JEROME. "Contribution a l'identification des processus cinetiquement limitants de l'oxydation des alliages de zirconium caracterisation en electrochimie des solides des films d'oxyde formes a haute temperature". Université Joseph Fourier (Grenoble), 2000. http://www.theses.fr/2000GRE10077.
Texto completo da fonteOlier, Patrick. "Alliages a memoire de forme de base tini : influence du mode de fabrication, de la teneur en oxygene et de l'ajout de zirconium ou d'hafnium sur les caracteristiques metallurgiques et les proprietes thermomecaniques". Paris 11, 1995. http://www.theses.fr/1995PA112485.
Texto completo da fonteLelievre, Gwenn. "Etude du rôle des précipités intermétalliques dans l'absorption d'hydrogène lors de la corrosion aqueuse d'alliages de zirconium". Université Joseph Fourier (Grenoble), 1998. http://www.theses.fr/1998GRE10174.
Texto completo da fonteJacques, Patrick. "Contribution à l'étude de l'amorçage des fissures de corrosion sous contrainte dans le zirconium et le zircaloy-4". Grenoble INPG, 1994. http://www.theses.fr/1994INPG0046.
Texto completo da fonteBéchade, Jean-Luc. "Texture et écrouissage de tôles en zircaloy-4 : évolutions en fonction des paramètres de laminage à froid : influence sur le comportement élastique, la dilatation thermique et l'anisotropie plastique". Ecole centrale de Nantes, 1993. http://www.theses.fr/1993NANT2083.
Texto completo da fonteWuchner, Sip Sibylle. "Etude des processus d'aimantation de tri-couches magnétiques à base d'alliages amorphes de terres rares et de cobalt". Université Joseph Fourier (Grenoble), 1995. http://www.theses.fr/1995GRE10070.
Texto completo da fonteSaintoyant, Lucie. "Couplage fluage-recristallisation dans les alliages de zirconium". Grenoble INPG, 2009. http://www.theses.fr/2009INPG0161.
Texto completo da fonteThe thesis aim is to understand and to model the behavior of Zircaloy-4 during the transport of spent fuel. We developed a model coupling the recrystallisation, the recovery and the mechanical behavior. Electron microscopy confirmed the pre-existent observations on static recrystallisation and showed that the application of a stress influences both the nucleation and the growing of new grain during recrystallisation. The former effect is physically described by adding a term of friction to the recrystallisation model and the latter implicitly included in the coupled modeI. The creep is modelled by a polycristallin mode] to take into account the etTect of the anisotropic behavior of zirconium and microstructure heterogeneity. The local behavior is described by a physically based model which adds a predictive potential to the previous model
Halley-Demoulin, Isabelle. "Oxydation d'alliages titane zirconium : Cinétique - Morphologique - Contraintes mécaniques". Dijon, 1992. http://www.theses.fr/1992DIJOS022.
Texto completo da fonteAlexandre, Pascal. "Purification sous vide du pseudo-alliage Zr-Mg-MgCl2 : approche théorique et étude expérimentale". Vandoeuvre-les-Nancy, INPL, 1988. http://www.theses.fr/1988NAN10326.
Texto completo da fonteChauvy, Cédric. "Traitements thermomécaniques dans le haut domaine alpha du zircaloy-4 trempé-bêta". Saint-Etienne, EMSE, 2004. http://www.theses.fr/2004EMSE0015.
Texto completo da fonteZircaloy-4 is a Zr base alloy mainly used for nuclear applications. This study deals with its behaviour during forming, with a special interest for physical mechanisms acting in the upper a-range. The b-treated Zircaloy-4 is first described in terms of microstructure and texture. The a plates are organised in colonies with alternating crystallographic orientations. The Burgers relationships have also been checked. The mechanical behaviour shows two distinct domains (with or without hardening). This could be linked to interactions between the solutes (Sn, O) and mobile dislocations. The evolution of crystallographic texture is characterised by X-ray diffraction. At 550ʿC, twinning is shown to be the main deformation mechanism under specific experimental conditions. Globularization of the initial lamellar structure is identified as a continuous dynamic recrystallization process, more efficient at higher temperature
Bossis, Philippe. "Mécanismes de corrosion du zircaloy-4 et de l'alliageZr-1Nb en eau pressurisée hors et sous irradiation : rôle des interfaces". Grenoble INPG, 1999. http://www.theses.fr/1999INPG0114.
Texto completo da fonteBerchi, Tarek. "Comportement micromécanique des alliages de zirconium en grandes déformations". Nantes, 2006. http://www.theses.fr/2006NANT2067.
Texto completo da fonteZirconium and its alloys are hexagonal close packed materials which present high plastic anisotropy and a variety of active deformation modes. As a consequence of these properties, second-order strains (and stresses) are generated during a plastic forming process between the grains having different crystallographic orientation. The mechanical behaviour of these alloys has been studied in the large strain framework for two cases: cold rolled Zy-4 plates and cold pilgering M5 zirconium alloy cladding tubes. Using X-ray diffraction technique, we have determined the residual stresses and crystallographic texture evolutions for different total strains. The elastoplastic self-consistent model was used to predict the mechanical state at the different scales. This model was developed to large deformations. A new formulation of the crystal plasticity has been proposed. The influence and the role of elastoplastic anisotropy have also been studied and explained in this work. A good agreement has been found between experimental and predicted crystallographic textures. The contribution and the magnitude of the first as well as the second-order residual stresses have been correctly evaluated from the measured strain. The prismatic slip is the most active deformation mode in these alloys at large strains
Dewobroto, Natanael. "Etude de l'évolution de texture lors de la recristallisation et de la croissance de grains d'alliages de titane et de zirconium". Metz, 2004. http://docnum.univ-lorraine.fr/public/UPV-M/Theses/2004/Dewobroto.Natanael.SMZ0412.pdf.
Texto completo da fonteThe objective is to explain mechanisms leading to texture change during annealing for commercially pure Ti (T40) and a zirconium alloy (Zr702) by making relationships between phenomena occurred during deformation, recrystallization and grain growth. In this experimental study, X-Ray goniometry SEM-EBSD and TEM were used to give complete descriptions on microstructure and texture evolution from deformed state until grain growth stage. Texture evolution for both materials are similar during cold rolling and annealing. Texture changes mainly during grain growth. Recrystallization changes slightly the rolling texture (maxima at {Phi1=0° PHI Phi2=0°}. Different deformation behaviors of both materials during cold rolling give differences in recrystallization mechanisms. T40 and Zr702 showed a not oriented nucleation. The slight growth selection since recrystallization stage develops the texture component {Phi1=0° PHI Phi2=30°} which becomes main component at advanced stage of grain growth. T40 evolves by normal grain growth for annealing temperature at 600, 700 and 800°C. Grain growth kinetics parameters for T40 were determined. Zr702 showed lower grain growth kinetics due to precipitates which are also the responsible for abnormal grain growth when heat treatments were done for long time at 800°C
Zumpicchiat, Guillaume. "Modélisation numérique de la diffusion-corrosion des alliages de zirconium". Thesis, Université Paris-Saclay (ComUE), 2015. http://www.theses.fr/2015SACLS238/document.
Texto completo da fonteIn Pressurized Water Reactor (PWR), zirconium-based alloy cladding tubes are immersed in high pressure water containing boron (1000 wt. boron) and lithium (2 wt. ppm) at high temperature (320 °C). The corrosion induced by this environment is mainly due to the oxidation of the zirconium which transforms in zirconia. This phenomenon is one of the limiting factors of the in-pile fuel rod lifetime (~ 5 years). Therefore, it is important to predict the corrosion process of zirconium based alloys in PWR conditions. Zirconium-based alloys oxidation is sub-parabolic inlike the Wagner theory which predicts a parabolic kinetics. Two finite element models were 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 were studied 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 were simulated. The sub-parabolic behavior of the oxidation kinetics can be explained by a non-uniform compressive stress level into the oxide layer. Simulations of oxygen diffusion throught polycristalline layer of zirconia were performed. Zirconia grains are modelled by Voronoï tesselation and a space between grains is meshed to model grain boundaries. These numerical samples are used to study the effect of zirconia microstructure and microtexture on oxygen diffusion. Experimental data from thin foils of zirconia formed on Zircaloy-4 and zirconium hydrure are used in the simulations
Bouvier, Pierre. "Étude Raman des distributions de phase et de contrainte dans des couches d'oxydation d'alliages de zirconium : étude spectroscopique des effets de pression et de température sur différentes zircones nanométriques". Grenoble INPG, 2000. http://www.theses.fr/2000INPG0136.
Texto completo da fonteDunlop, John. "Approche variable interne de fluage et recristallisation des alliages en zirconium". Grenoble INPG, 2005. http://www.theses.fr/2005INPG0055.
Texto completo da fonteZircaloy-4 fuel cladding is the primary barrier between nuclear fuel pellets and associated fission products, and the external environment. Though the behaviour of Zircaloy-4 during service has been thoroughly studied, the response of cladding after service, particularly dur-ing transport and temporary storage, is less well understood. For the different scenarios that have been proposed for storage of spent fuel, some may induce transients in temperatures and stresses owing to residual thermal power within the fuel pellets. To ensure cladding in-tegrity during these transients it is desirable to be able to predict the microstructural and me-chanical response under these conditions. This thesis has two aims. These are: firstly, to develop a model describing the plastic de-formation of Zircaloy-4 under the conditions expected after service, and secondly, to develop a model describing the microstructural evolution expected at higher temperatures. A coupled model for recovery, nucleation and growth of recrystallisation, is developed and is applied to isothermal data of recovery and recrystallisation of Zircaloy-4. The model successfully pre-dicts critical strains and temperatures for recrystallisation and can describe recrystallisation kinetics under non-isothermal conditions. The internal variable plasticity model of Kocks, Mecking and Estrin is modified to include kinematic hardening associated with the develop-ment of incompatibility stresses between the grains, and applied to the plastic deformation of Zircaloy-4. A model for unpinning of mobile dislocations is developed to describe the yield point seen in fully recrystallised material. This is successful in describing th
Szymanski, Raymond. "Préparation, caractérisation et étude des propriétés catalytiques d'alliages platine-zirconium sur supports de carbone ou de zircone". Lyon 1, 1985. http://www.theses.fr/1985LYO19012.
Texto completo da fonteDrouet, Julie. "Étude expérimentale et modélisation numérique du comportement plastique des alliages de zirconium sous et après irradiation". Toulouse 3, 2014. http://thesesups.ups-tlse.fr/3401/.
Texto completo da fonteRecrystallized zirconium alloys are widely used as constitutive material of claddings and cladding tubes in Pressurized Water Reactors (PWR). During their lifetime in reactor, these elements are submitted to irradiation, creating a large amount of defects and changing their mechanical behavior. Despite the broad knowledge of macroscopic modifications due to irradiation, microscopic mechanisms involved remain partially known and understood. This study aims to clear up that point using two different means to investigate interactions between moving dislocations and dislocation loops created by irradiation. The experimental one is based on straining pre irradiated Zircaloy-4 samples containing dislocation loops in a transmission electron microscope (TEM). Mobile dislocations are observed to interact with these loops, following different mechanisms. Loops can act as strong obstacles to moving dislocations, pinning their further glide and hardening the material. Therefore, this type of mechanism participates in irradiation hardening. They have also been observed to be absorbed by dislocations, showing the ability of dislocations to clear up defects. This mechanism explains the formation of clear bands observed in the material after irradiation and mechanical testings. The numerical one is based on Discrete Dislocation Dynamics simulations of interactions between mobile dislocations in prismatic or basal planes of the HCP lattice and loops, using NUMODIS. The results of this study are consistent with a recent study of interactions of dislocations in a prismatic plane and loops studied by molecular dynamics (MD). The counterpart of this study with gliding dislocations in the basal plane, performed only using DD simulations, show interesting explanations of the observed clear band formation in basal and prismatic planes, with broader channels in basal planes. Some interesting clues have been found to explain differences in quantification of critical stresses needed to overcome defaults between DD and MD simulations. Looking towards multiscale simulation of plasticity, these clues have to be investigated to fill the gap between DD and MD simulations. A situation observed during in situ TEM experiments has been simulated using DD, and the result of the simulation is spatially and temporally consistent with the experimental observations. This reveals the ability of the DD codes to mimic in situ TEM experiments with a good agreement at time and space scale, when parameters are fitted on data extracted directly from TEM experiments. This offers opportunities to fit DD parameters not only on MD simulation results but also on experimental results, closer to the real behavior of materials. A preliminary study of microscopic mechanisms responsible for irradiation creep of zirconium alloys observed in reactor has also been carried out. Combined in situ and post mortem TEM observations of pre strained samples under irradiation at room temperature have not yet allowed to observe evidence of climb of dislocations. This study has to be pursued at higher temperatures in order to allow activation of diffusion mechanisms
Onimus, Fabien. "Approche Expérimentale et Modélisation Micromécanique du Comportement des Alliages de Zirconium Irradiés". Phd thesis, Ecole Centrale Paris, 2003. http://tel.archives-ouvertes.fr/tel-00006513.
Texto completo da fonteOnimus, Fabien. "Approche expérimentale et modélisation micromécanique du comportement des alliages de zirconium irradiés /". [Gif-sur-Yvette] : [CEA Saclay, Direction des systèmes d'information], 2004. http://catalogue.bnf.fr/ark:/12148/cb39254231s.
Texto completo da fonteLa couv. porte en plus : "Direction de l'énergie nucléaire" Bibliogr. p. 313-321. Résumé en français et en anglais.
Queylat, Benoît. "Compréhension de l'évolution de la fraction d'hydrogène absorbée par les gaines en alliages de zirconium". Thesis, université Paris-Saclay, 2020. http://www.theses.fr/2020UPASN026.
Texto completo da fonteIn nuclear Pressurized Water Reactor (PWR), nuclear fuel is contained in zirconium alloy cladding tubes. As the cladding tube acts as the first containment barrier of fissile materials, protecting its physical integrity is fundamental. Under harsh environment (high temperature and pressure, neutron irradiation), claddings undergo oxidation and hydriding processes. Hydrogen uptake is detrimental to the integrity of claddings: when the hydrogen content reaches the solubility limit, precipitation of hydrides occurs, which may embrittle them. For the last decades, development of new zirconium-niobium alloys, as M5Framatome, allowed a large decrease of the hydrogen uptake by claddings. In addition, it has been observed that, opposite to Zircaloys, M5Framatome absorbs less hydrogen after ion-irradiation. Nevertheless, the reasons of this decrease in niobium containing alloys are still not well understood. Consequently, this PhD thesis follows multiple aims:· Understanding the hydrogen pickup evolution of the M5Framatome under simulated PWR conditions;· Understanding niobium’s role in hydrogen pickup;· Understanding irradiation effects on hydrogen pickup.To reach these goals, samples of M5Framatome (with 1 wt% niobium) and model alloys (containing 0.2 and 0.4 wt% of niobium) have been corroded in static autoclave in order to monitor the evolution of oxidation and hydrogen pick up kinetics. Neutron irradiation has been simulated by series of ion irradiations of pre-oxidized M5Framatome samples. Chemical analyses have been performed in order to understand mechanisms involved in the absorption and diffusion processes of hydrogen in the oxide layer at different spatial scales. All these tests have led to the following conclusions :· As long as the oxide layer thickness stays below 2.5 µm, the hydrogen pickup by the metal underneath is close to zero. Above that limit, the hydrogen pickup is proportional to the oxide layer thickness.· The hydrogen pickup in the metal is not limited by its absorption nor its diffusion in the oxide but by the transition from the oxide layer into the metal.· When the niobium content is below or equal to the solubility limit in α-Zr, the hydrogen pickup is proportional to the oxide thickness.· Ion irradiation in post-transition corrosion state decreases the oxidation and hydrogen pickup rates but has no effect on the hydrogen pickup fraction.Based on these various results, a hydrogen pickup mechanism in corrosion conditions has been suggested for the M5Framatome. It attempts to take into account the role of both the niobium inclusion and the irradiation effects observed in this work. In addition, empirical laws describing the evolution of hydrogen pickup fraction have been proposed and compared to data taken from nuclear reactors
Cataldo, Laurent. "Contribution à l'élaboration et à l'optimisation d'alliages magnétiques permanents Sm-Co-Cu-Fe-Zr". Lyon 1, 1996. http://www.theses.fr/1996LYO10206.
Texto completo da fonteHervier, Paul. "Fonctionnalisation de surface de verres métalliques base Zirconium". Thesis, Université Grenoble Alpes (ComUE), 2017. http://www.theses.fr/2017GREAI088/document.
Texto completo da fonteMetallic glasses are recent materials. First developed in the 60s, they are well-known for their high mechanical resistance and their ability to become viscous at relatively low temperatures. Functionalization of their surfaces is a promising way to further increase their properties. However, their amorphous structure is in a metastable state and maintaining them at too high temperatures leads systematically to their crystallization, and thus the loss of their unique properties. Most of surface treatment techniques are performed at high temperatures and thus are not adapted to these materials. In this work, two innovative techniques which are thermoforming and ultra-short pulse duration laser treatment have been used and allow to texture the surfaces of the alloys by avoiding their crystallization. This thesis is focused on the effect of these two processing techniques on physical and chemical properties of Zr-based bulk metallic glasses and thus on the modification of their surface properties such as wettability or corrosion resistance. We will see that both techniques present their advantages and can be particularly adapted for biomedical applications
Ribis, Joël. "Approche expérimentale et modélisation micromécanique du comportement en fluage des alliages de zirconium irradiés". Grenoble INPG, 2007. http://www.theses.fr/2007INPG0177.
Texto completo da fonteUsed as cladding tubes in the Pressurized Water Reactor, the zirconium alloys are hardened by dislocation loops induced by irradiation. The study of the creep behavior of the irradiated zirconium alloys was conducted with an experimental approach (TEM, mechanical testing, microhardness) and a numerical approach where the microstructure evolution during a heat treatment was modeled (cluster dynamic). This study allows to understand the creep behavior of the irradiated alloy which is strongly dependant of the thermal recovery and the sweeping of loops. In the end, a micromechanical modeling was used for a predictive approach of the creep behavior of the irradiated zirconium alloys in dry storage conditions
Sun, Fan. "Alliages nanostructurés : à base d'aluminium et des beta-métastables à base de titane". Rennes, INSA, 2009. http://www.theses.fr/2009ISAR0013.
Texto completo da fonteIn this work thermal analyses were carried out on the Al88Ni6Sm6 amorphous alloy of and metastable beta Ti-Mo alloys. By means of DSC, electrical resistivity and dilatometry, the nanophase transformation mechanisms and the kinetics growth in both alloys were investigated under isothermal and non-isothermal condition. On the Ti-based alloys, thermo-mechanical treatments were carried out in order to enhance the tensile strength by controlling the dispersion of the alpha nanoprecipitates through the beta matrix. Tensile test results indicated a very high strengthening effect, which is particularly huge with the Ti-12Mo alloy where a tensile strength as high as around 1600MPa was obtained after a two-step annealing treatment. This highly enhanced tensile strength was attributed to the complex intragranular nanostructure observed by transmission electron microscopy, consisting of two-scale alpha nanoprecipitates inside sub-micrometer beta grains
STEPHAN, MAURICE. "Solidification rapide d'alliages a base fer et a base zinc. Application a l'amelioration du rendement d'un inducteur". Le Mans, 1994. http://www.theses.fr/1994LEMA1021.
Texto completo da fonteProff, Christian. "Aspects microstructuraux de l'oxydation d'alliages de Zirconium". Phd thesis, Université de Grenoble, 2011. http://tel.archives-ouvertes.fr/tel-00609232.
Texto completo da fonteDiz, Jésus. "Evaluation à l'aide de modèles des paramètres structuraux importants pour la prévision du comportement élastique, de dilatation thermique et de la croissance sous irradiation d'alliages de zirconium polycristallin". Metz, 1992. http://docnum.univ-lorraine.fr/public/UPV-M/Theses/1992/Diz.Jesus.SMZ9227.pdf.
Texto completo da fonteOne of the aims of this work was to identify, through the study of elastic and thermal expansion properties, the main parameters necessary for a satisfactory modelling of deformation properties, which show some analogies in their mathematical formulation. Thus we developed and tested various modelling of these behaviors for hexagonal materials. Our study of the influence of some metallurgical parameters on the physical properties of zirconium alloys led us to the following conclusions. Some parameters play an equally important influence on properties as different as elasticity, thermal expansion and irradiation growth. So, we have verified that the crystallographic texture plays a key role in performing such predictions. Secondly, we have observed that other metallurgical parameters such as the grain size, the chemical composition and internal stresses are more or less important according to the investigated property. In the case of irradiation growth a complete model, taking all these parameters into account, is not available today. It is nevertheless possible, according to the metallurgical state of the studied material and for a given chemical composition, to perform qualitative and even quantitative predictions
Lê, Đưc Huy. "Contribution à l'étude structurale et vibrationnelle des couches minces de zircone ZrO2 déposées sur alliage Zy-4". Le Mans, 2004. http://cyberdoc.univ-lemans.fr/theses/2004/2004LEMA1021.pdf.
Texto completo da fonteAdami, Lahebib. "Etude des alliages zirconium-fer par mesure de pouvoir thermoélectrique et microscopie électronique". Lyon, INSA, 1988. http://www.theses.fr/1988ISAL0021.
Texto completo da fontePoollay, Mootien Sattyvel. "Fluage des alliages à base d'aluminium-lithium". Aix-Marseille 2, 1992. http://www.theses.fr/1992AIX22011.
Texto completo da fonteAjao, John. "Borures dans quelques alliages à base nickel". Grenoble INPG, 1988. http://www.theses.fr/1988INPG0025.
Texto completo da fonteAjao, John. "Borures dans quelques alliages à base nickel". Grenoble 2 : ANRT, 1988. http://catalogue.bnf.fr/ark:/12148/cb37611105t.
Texto completo da fonteHaurais, Florian. "Evaluate the contribution of the fuel cladding oxidation process on the hydrogen production from the reflooding during a potential severe accident in a nuclear reactor". Thesis, Université Paris-Saclay (ComUE), 2016. http://www.theses.fr/2016SACLS375/document.
Texto completo da fonteIn nuclear power plants, a severe accident is a very unlikely sequence of events during which components of the reactor core get significantly damaged, through chemical interactions and/or melting, because of very high temperatures. This may potentially lead to radiotoxic releases in the containment building and to air ingress in the reactor core. In that context, this thesis work led at EDF R&D aimed at modeling the deterioration of the nuclear fuel cladding, made of zirconium alloys, in accidental conditions: high temperature and either pure steam or air-steam mixture. The final objective was to improve the simulation by the MAAP code of the cladding oxidation and of the hydrogen production, in particular during a core reflooding with water. Due to the progressive thickening of a dense and protective ZrO2 layer, the oxidation kinetics of Zr in steam at high temperatures is generally (sub-)parabolic. However, at certain temperatures, this oxide layer may crack, becoming porous and not protective anymore. By this “breakaway” process, the oxidation kinetics becomes rather linear. Additionally, the temperature increase can lead core materials to melt and to relocate down to the vessel lower head whose failure may induce air ingress into the reactor core. In this event, oxygen and nitrogen both react with the pre-oxidized claddings, successively through oxidation of Zr (thickening the ZrO2 layer), nitriding of Zr (forming ZrN particles) and oxidation of ZrN (creating oxide and releasing nitrogen). These self-sustained reactions enhance the cracking of the cladding and of its ZrO2 layer, inducing a rise of its open porosity.In order to quantify this cladding porosity, an innovative two-step experimental protocol was defined and applied: it consisted in submitting ZIRLO® cladding samples first to various accidental conditions during several time periods and then to measurements of the open porosity through porosimetry by mercury intrusion. The tested corrosion conditions included numerous temperatures ranging from 1100 up to 1500 K as well as both pure steam and a 50-50 mol% air-steam mixture. For the ZIRLO® samples oxidized in pure steam, except at 1200 and 1250 K, the “breakaway” kinetic transitions do not occur and the open porosity remains negligible along the oxidation process. However, for all other samples, corroded in air-steam or oxidized in pure steam at 1200 or 1250 K, “breakaway” transitions are observed and the porosimetry results show that the open porosity increases along the corrosion process, proportionally to the mass gain. Moreover, it was evidenced that the pore size distribution of ZIRLO® samples significantly extends during corrosion, especially after “breakaway” transitions. Indeed, the detected pore sizes ranged from 60 μm down to around: 2 μm before the transition, 50 nm just after and 2 nm longer after. Finally, a two-step numerical model was developed in the MAAP code to improve its simulation of the cladding oxidation. First, thanks to the proportionality between open porosity and mass gain of cladding samples, porosity correlations were implemented for each tested corrosion condition. Second, the calculated porosity values are used to proportionally enhance the cladding oxidation rate. This improved model thus simulates not only chemical reactions of Zr-based claddings (oxidation and nitriding) but also their mechanical degradation and its impact on their oxidation rate. It was validated by simulating QUENCH tests (-06, -08, -10 and -16), conducted at KIT to study the behavior of claddings in accidental conditions with a final reflooding. These simulations show a better cladding thermal behavior and a hydrogen production significantly higher and so closer to experimental values, in particular during the reflooding
Adami, Lahebib. "Etude des alliages zirconium-fer par mesure de pouvoir thermoélectrique et microscopie électronique". Grenoble 2 : ANRT, 1988. http://catalogue.bnf.fr/ark:/12148/cb376111619.
Texto completo da fonteGaume, Marine. "Etude des mécanismes de déformation des alliages de zirconium après et sous irradiation". Thesis, Toulouse 3, 2017. http://www.theses.fr/2017TOU30220/document.
Texto completo da fonteIn Pressurized Water Reactors, the neutron flux leads to a change in the mechanical properties of the fuel cladding tubes made of zirconium alloys. Although their macroscopic behavior is well known, the microscopic deformation mechanisms of zirconium alloys still need to be characterized. In order to simulate the neutron irradiation, charged particles irradiations (ion and electron) were carried out at 400°C and 450°C on a zirconium alloy: RXA Zircaloy-4. The experimental analysis of the irradiated microstructure, performed by using a Transmission Electron Microscope (TEM), have shown some crystalline defects: dislocation loops with a Burgers vector. Their evolution (size and density) and their characteristics (nature and habit plane) have been determined and discussed based on the point defects diffusion. The results suggest a weak anisotropy in the self-interstitial diffusion. In-situ tensile tests were performed using a TEM, after ion irradiation, in order to activate the dislocation glide and to observe their interaction with the loops. Some of the experimental cases of interaction have been simulate using Dislocation Dynamics for a better understanding of the mechanisms. The simultaneous effect of the stress and of the irradiation on the deformation mechanisms have been then studied. In-situ electron and ion irradiations were conducted, with and without an applied stress. Deformation mechanisms involving dislocation climb have thus been demonstrated. Through this study, models based on the identified mechanisms may be suggested, in order to predict the behavior of zirconium alloys in the reactor
JOMARD, GERALD. "Approche ab initio de processus d'oxydation du zircaloy-4". Université Joseph Fourier (Grenoble), 2000. http://www.theses.fr/2000GRE10087.
Texto completo da fonteLebon, Cyril. "Etude expérimentale et simulation numérique des mécanismes de plasticité dans les alliages de zirconium". Phd thesis, Université de La Rochelle, 2011. http://tel.archives-ouvertes.fr/tel-00808627.
Texto completo da fonteSimonot, Claude. "Evolutions microstructurales des alliages de zirconium sous irradiation liens avec le phenomene de croissance". Paris 11, 1995. http://www.theses.fr/1995PA112258.
Texto completo da fonteSimonot, Claude. "Évolutions microstructurales des alliages de zirconium sous irradiation : liens avec le phénomène de croissance /". Gif-sur-Yvette : Direction de l'information scientifique et technique, CEA Saclay, 1996. http://catalogue.bnf.fr/ark:/12148/cb35840082f.
Texto completo da fonteRibis, Joël. "Approche expérimentale et modélisation micromécanique du comportement en fluage des alliages de zirconium irradiés /". Gif-sur-Yvette : CEA Saclay, Direction des systèmes d'information, 2008. http://catalogue.bnf.fr/ark:/12148/cb41399294w.
Texto completo da fonteNotice réd. d'après la couv. La couv. porte en plus : "Direction de l'énergie nucléaire, Direction des activités nucléaires de Saclay" Bibliogr. p. 225-230. Résumé en français et en anglais.