Thèses sur le sujet « Combustibles nucléaires – Gaines – Matériaux »
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Yang, Hongyue. « Approche thermomécanique du phénomène d'interaction pastille-gaine dans un crayon combustible ». Lyon, INSA, 1995. http://www.theses.fr/1995ISAL0015.
The objective of this study is to investing ate the thermo-mechanical behaviour of a fuel rod which is composed of a tube containing cylindrical uranium dioxide fuel pellets. A lot of work has been particularly devoted to the design and the set up of an original experimental device which allows the simulation of the pellet-cladding mechanical interaction during the uprating of water pressurised reactors. The mode) is a two-dimensional plane stress one which has not taken the effects of irradiation and chemical transformation into account. Both experimental and numerical analyses on the cladding stresses concentration have been carried out. Different effects have been studied including the presence of cracks in the pellets, the coefficient of friction between pellet-cladding, the initial gap at the pellet-cladding interface, the cladding external pressure and the variation of power, The comparison of experimental and numerical results, for a particular case, enables a better understanding of pellet-cladding mechanical interaction et demonstrates that the modelisation of the joint element used in the pellet-cladding interface is appropriate. It should be pointed out that a simple elastic approach had been carried out before the thermoplastic experiment was done, aiming at analysing the stress concentrations due to the presence of cracks in the pellet
Ougier, Michaël. « Etude de l’élaboration de revêtements autocicatrisants pour le développement de matériaux robustes en condition nucléaire ». Thesis, Université Paris-Saclay (ComUE), 2019. http://www.theses.fr/2019SACLE028/document.
This study aims to improve oxidation resistance of nuclear fuel claddings in accident conditions. In this context, Cr-Al-C and Cr2AlC coatings deposition and their behavior were studied. Firstly, we investigated the influence of HiPIMS process parameters on the properties of the plasma and the deposited films. Despite more intense ionic fluxes due to the HiPIMS process, coatings do not crystallize without an additional energy supply. Partially crystallized Cr2AlC thin films were obtained by a 500°C annealing of as-deposited Cr-Al-C coatings. This two-step process is a viable solution to protect nuclear claddings with Cr2AlC coating while maintaining the metallurgical properties of the zirconium-based substrates. Secondly, the assessment of the oxidation resistance of as-deposited and annealed coatings revealed significant protective effect against rapid oxidation under dry and wet air at high temperatures (up to 1200°C) owing to the formation of a continuous oxide layer. During the first stages of oxidation, this layer is made of α Al2O3 and Cr2O3 for as-deposited coating while only α-Al2O3 is present for the annealed one. Because of Al depletion, coatings later deteriorate and form a residual and porous intermediate chromium carbide (Cr7C3) layer which further fully oxidizes. It was shown that the inward diffusion of Al with Zr also accelerates the coating deterioration. To improve the oxidation resistance of these coatings, multilayered architectures were developed. Adding a molybdenum interlayer as diffusion barrier globally decreased the oxidation resistance of the coating. In contrast, topping Cr-Al-C and Cr2AlC with a Cr layer improved oxidation behavior over single-layer coatings
Pantera, Laurent. « Application d'une méthodologie statistique à la compréhension du phénomène de corrosion du surgénérateur Phénix ». Compiègne, 1992. http://www.theses.fr/1992COMPD509.
Zouari, Ahmed. « Comportement des gaines en alliages de zirconium en conditions thermo-mécaniques représentatives d’un accident RIA ». Thesis, Université Paris sciences et lettres, 2020. http://www.theses.fr/2020UPSLM058.
The aim of this work is to enhance the understanding of the thermomechanical behavior at rupture of the fuel rod cladding during an accidental transient of the RIA type. A new mechanical test has been developed in order to apply a strain biaxiality ratio ɛzz/ɛ00 between -0,2 et 1. It allows reproducing loading conditions close to the ones occurring during a RIA accident. An experimental campaign at room temperature carried out with this device made it possible to study the effects of strain biaxiality ratios and strain rate on the fracture of the cladding. The tests carried out show a significant effect of the biaxiality on the hoop strain at failure which has a minimum of a plane strain state where the strain biaxiality is close to 0. A slight decrease in ductility was also recorded during the increase in the strain rate for all biaxiality levels. The damage mechanisms and failure modes of specimens are identified from the surfaces and failure profiles depending on the stress conditions. Numerical finite element simulations were performed with the CAST3M code to model the test and simulate the failure of cladding with biaxial loading. A second experimental device has been developed to couple the effects of biaxial mechanical loading and rapid thermal loading. The objective is to heat the cladding with temperature rise rates greater than 100 °C.s-1 in order to avoid the restoration of the defects linked to the hydriding and to the irradiation during the test. The method was used to perform thermomechanical tests at high heating rates, high strain rates, and high biaxiality levels to reproduce full loadings in a reactivity accident. The first results show, for a virgin cladding, that the strain at the break was not affected by temperature or by the rate of heating. Finite element simulations were undertaken to model the different heating methods tested and to optimise the chosen method. These simulations made it possible in particular to model the passage of electric current and heat flow through solid-solid interfaces
Khelifi, Nour-Eddine. « Modélisation de la croissance sous irradiation de feuillards de zircaloy-4 détendus en fonction de leur texture ». Metz, 1991. http://docnum.univ-lorraine.fr/public/UPV-M/Theses/1991/Khelifi.Nour_eddine.SMZ9124.pdf.
Autones, Lucas. « Élaboration d’aciers ODS (Oxide Dispersion Strengthened) par fabrication additive laser et cold spray : compréhension des relations procédés - microstructures ». Electronic Thesis or Diss., Université de Lille (2022-....), 2022. http://www.theses.fr/2022ULILR004.
ODS (Oxide Dispersion Strengthened) steels are materials that exhibit very good resistance to creep and swelling under irradiation. These properties make them good candidates for cladding materials in Generation IV reactors, or for structural materials in thermonuclear fusion reactors. The dispersion of the nano-oxides, which reinforce the material, is obtained by powder metallurgy. Mechanical-alloying of an atomized steel powder with an oxide powder (Y2O3) results in the oxide dissolution in the matrix. During hot consolidation (hot isostatic pression or hot extrusion), the precipitation of the nano-oxides takes place. Designs of component with these materials and their final geometry could be improved using additive manufacturing.Since the 2010s, recent developments in additive manufacturing technologies could enable to reduce lead times and costs, while increasing the geometric, hierarchical and functional complexity of parts. They pave the way to new freedom of design compared to conventional subtractive manufacturing processes.The objective of this thesis work was to assess the potentials of different additive manufacturing techniques (SLM, DMD, and Cold Spray) for ODS steels.Thus, three types of ODS powder (mechanically-alloyed, composite and STARS) were obtained to determine the most interesting powder-process combinations. The materials produced from these different combinations have been characterized at several scales. The amount of macroscopic defects (porosities, cracks) was analyzed in order to optimize the manufacturing parameters. Their granular microstructure was observed before and after annealing at 1100 °C by optical and electron microscopy (SEM, EBSD). The nano-precipitation was analyzed by SEM, TEM and by small angle X-rays scattering. An image analysis method combining high definition electron microscopy images and a machine learning software was implemented. Finally, the high temperature tensile properties of these different materials were evaluated and are in good agreement with their microstructural characteristics. The comparison of the whole characterization results enabled to select the relevant manufacturing paths.The results obtained indicate that laser additive manufacturing processes (SLM, DMD) lead to ODS steels with low performance, regardless the type of powder used. The yttrium content can greatly decrease after consolidation. It also forms fragile Y-rich coarse phases, and the density of the nano-precipitates population appears very low. These microstructural characteristics induce tensile properties equivalent to those of an unreinforced steel. Nevertheless, the composite powder elaboration method implemented in this work makes it very easy to adapt the nature and content of the reinforcements added to the base powder. Using TiC nano-particles, very fine microstructures composed of equiaxed grains were obtained. These unusual microstructures in laser additive manufacturing offer interesting prospects.ODS steels obtained by cold spray from a mechanically-alloyed powder have characteristics similar to conventional ODS steels. After annealing, these materials have a microstructure similar to the ODS steels obtained by HIP. However, the coarse grains take up a much larger fraction of the microstructure and attest to a more advanced recrystallization. The lower hardness and elastic limit of this material compared to its HIP equivalent confirm this result, which is very encouraging if further shaping should be aimed. The very high density of Y-Ti-O nano-oxides in the Cold Sprayed ODS steel enables to achieve a mechanical resistance at 700 °C which is 50 MPa higher than the HIPed ODS. However, this material exhibits a loss of ductility which will have to be resolved. The analyzes carried out enabled to suggest two mechanisms to explain this damage, which would be caused by the presence of microcracks and porosities in the part
Bakkali, Amin El. « Conception et mise au point d'un essai de courbe R sur des gaines d'éléments combustibles nucléaires ». Châtenay-Malabry, Ecole centrale de Paris, 1987. http://www.theses.fr/1987ECAP0056.
Roussette, Sophie. « Analyse par champs de transformation de matériaux élastoviscoplastiques multiphases : application aux combustibles MOX ». Aix-Marseille 2, 2005. http://www.theses.fr/2005AIX22054.
The description of the overall behavior of nonlinear materials with nonlinear dissipative phases requires an infinity of internal variables. An approximate model involving only a finite number of internal variables, Nonuniform Transformation Field Analysis, is obtained by considering a decomposition of these variables on a finite set of nonuniform transformation fields, called plastic modes. The method is initially developed for incompressible elastoviscoplastic materials. Karhunen-Loève expansion is proposed to optimize the plastic modes. Then the method is extended to porous elastoviscoplastic materials. Finally the transformation field analysis, developed by Dvorak, is applied to nuclear fuels MOX. This method enables to make sensitivity studies to determine the role of some microstructural parameters on the fuel behaviour. Moreover the adequacy of the nonuniform method for fuels MOX is shown, the final objective being to be able to apply the model to the MOX in 3D
Robert-Berat, Laurence. « Influence d'une couche de zircone sur le comportement mécanique des tubes en zircaloy-4 ». Clermond-Ferrand 2, 2001. http://www.theses.fr/2001CLF2A001.
Quaranta, Delphine. « Étude de voies potentielles pour le recyclage du zirconium des gaines en Zircaloy des combustibles nucléaires usés ». Thesis, Toulouse 3, 2019. http://www.theses.fr/2019TOU30038.
Zircaloy-4 is an alloy mainly composed of zirconium (~ 98%wt.) constituting the cladding of nuclear assemblies. Currently, used Zircaloy claddings are intended for deep geological storage due to their contamination by radioelements from the nuclear reaction and the reprocessing process. They are classified as long-lived intermediate-level waste according to ANDRA recommendations (radioactivity: 10 6 - 10 9 Bq/g, periods > 31 years), as they represent 25%wt. of the assembly inventory. Zirconium recycling thus could present an economic interest, either to upgrade the zirconium by remanufacturing sheaths (with the constraint imposed by the residual presence of 93Zr), or to downgrade the cladding wastes into low activity waste. This thesis aims to study the potential routes for the recycling of zirconium contained in spent Zircaloy sheaths, and more precisely electrorefining in molten fluorides. The study of Zircaloy sheath composition of spent nuclear fuel was first carried out to identify the radioelements present in used claddings. These elements are either activation products (Cr, Fe, Ni, Co, Sn, etc.), or fission products (H, Sr (+ Y), Cs (+ Ba), Eu, etc.), or actinides (U, Pu, Am and Cm). An electrochemical study of the zirconium (IV) ions was carried out in LiF-NaF at 750 °C to determine its reduction mechanisms into metallic zirconium. Then, a nucleation / growth study was performed to optimize the operating conditions (ie nature of the cathode, concentration of ZrF4, current density applied, etc.), to obtain an adherent metal zirconium deposit on inert solid cathode. The last part of this work was focused on the electrorefining of "fresh" Zircaloy sections, i.e. before its stay in the reactor. Particular attention was paid to the behavior of the alloy constituents (Fe, Cr and Sn), during the electrolysis process. This work proposes a first scenario for the reprocessing of spent fuel claddings
Pillon, Sylvie. « Étude des diagrammes de phases U-O-Na, Pu-O-Na et U,Pu-O-Na ». Montpellier 2, 1989. http://www.theses.fr/1989MON20045.
Bouloré, Antoine. « Etude et modélisation de la densification en pile des oxydes nucléaires UO2 et MOX ». Grenoble INPG, 2001. http://www.theses.fr/2001INPG4203.
Amongst the many phenomena which take place in the course of the irradiation of UO2 or (U, Pu)O2 nuclear fuels, one of them involves the elimination of a fraction of the as-fabricated porosity. In-pile densification or sintering can reach 2. 5%, i. E. Approximately half the initial volume of pores is likely to disappear. Our literature survey indicates that the amplitude and kinetics of the phenomenon are both heavily dependent on the initial fuel microstructure. Micro-structural characterisation techniques of oxide fuels have therefore been developed in conjunction with quantitative image analysis methods. The ensuing methodology enables a quantitative comparison of micro-structural features in different fuels and has been applied to ascertaining the influence of the local fission rate and temperature on in-pile densification. It is thus revealed that in-pile operation eliminates a significant fraction of pores smaller than 3 microns in diameter. The experimental data generated has been used to set up a semi-empirical and a mechanistic model. The former is based on experimental results and is not essentially predictive. The inability of this model to predict the in-pile densification of oxide fuels is illustrated by the fact that the maximum fraction of pores that disappears is proportional to an empirical function of fission rate, and temperature. The proportionality factor appears to be difficult to correlate quantitatively to any given micro-structural feature. The model has however been applied to the interpretation of an in-pile densification experiment carried out in the Halden reactor (Norway). The latter model is mechanistic, i. E. It is based on the solution to a set of equations that describe the coupled temperature and radiation induced phenomena which occur in-pile. These can broadly be broken down into three categories : the fission fragment-pore interaction, the creation of point defects as the fission fragments slow down, and the diffusion of these point defects to sinks. The model calculates the evolution of the pore size distribution and has successfully been applied to modelling the in-pile densification behaviour of a fuel pellet characterised before and after irradiation
Caranoni, Laurent. « Incidence d'additifs à base de soufre sur la microstructure des combustibles nucléaires : Elaboration et caractérisations ». Limoges, 2002. http://www.theses.fr/2002LIMO0010.
Even though the global reactor working of MOX fuel is good, the fission gas emission now represents the limitating factor of its use at high burn-up
Gossard, Alban. « Synthèse d'oxydes par voie sol-gel colloïdale : application aux précurseurs de combustibles nucléaires ». Thesis, Montpellier, Ecole nationale supérieure de chimie, 2014. http://www.theses.fr/2014ENCM0010/document.
One of the main objectives for the future nuclear fuel cycle is the recycling of the minor actinides. Different options are considered: their integration into a new fuel for a prospect of a closed fuel cycle or their transmutation in order to significantly decrease the long-term radiotoxicity of ultimate wastes. In both cases, the synthesis of new advanced materials integrating the actinides jointly is required.Sol-gel processes allow the organization of the material at the colloidal scale or the insertion of controlled porosity using « templates ». Furthermore, the possibility to work in a « wet environment » prevents the formation of pulverulent powders which are contaminant in the case of materials incorporating radioactive elements. The main purpose of this work is to demonstrate the adaptability of this route to the nuclear field.Firstly, a methodology of synthesis from a colloidal sol-gel route was set up on a non-radioactive zirconium-based system in order to characterize and understand of the different mechanisms of this synthesis. Then, studies on shaping, including insertion of porosity, were performed. Zirconia monoliths have been obtained thanks to a coupling between a colloidal sol-gel process and the formation of an emulsion stabilized by clusters of solid particles. Finally, a transposition of this work to an uranium-based system was introduced, pointing out different promising perspectives specially concerning the possibilities of shaping of the final material
March, Philippe (1970. « Caractérisation et modélisation de l'environnement thermohydraulique et chimique des gaines de combustible des réacteurs à eau sous pression en présence d'ébullition ». Aix-Marseille 1, 1999. http://www.theses.fr/1999AIX11068.
Fauque, de Maistre Jules. « Modèle d’ordre réduit en mécanique du contact. Application à la simulation du comportement des combustibles nucléaires ». Thesis, Paris Sciences et Lettres (ComUE), 2018. http://www.theses.fr/2018PSLEM073/document.
The model order reduction of mechanical problems involving contact remains an important issue in computational solid mechanics.An extension of the hyper-reduction method based on a reduced integration domain to frictionless contact problems written by a mixed formulation is proposed.As the potential contact zone is naturally reduced through the reduced domain, the dual reduced basis is chosen as the restriction of the dual full-order model basis.A hybrid hyper-reduced model combining empirical modes for primal variables with finite element approximation for dual variables is then obtained.If necessary, the inf-sup condition of this hybrid saddle point problem can be enforced by extending the hybrid approximation to the primal variables. This leads to a hybrid hyper-reduced/full-order model strategy. By this way, a better approximation on the potential contact zone is furthermore obtained.A post-treatment dedicated to the reconstruction of the contact forces on the whole domain is introduced.In order to optimize the snapshots selection, an efficient error indicator is coupled to a greedy sampling algorithm leading to a robust reduced-order model
Chaieb, Ahmed. « Comportement anisotherme et rupture des gaines combustibles en alliages de zirconium : Application à la situation d'accident d'insertion de réactivité (RIA) ». Thesis, Paris Sciences et Lettres (ComUE), 2019. http://www.theses.fr/2019PSLEM005.
Fuel clads made of zirconium alloys are the first safety barrier in the nuclear power plants. This work aims to enhance the understanding of the thermomechanical behavior of Zirlcaoy-4 during RIA accidental scenario. Indeed, the current experimental databases are mainly constituted of uniaxial tensile tests carried out under isothermal conditions. The anisothermal character of the loading, coupled or not with the biaxiality of the mechanical loading, has been poorly studied. The aim of the thesis is to develop new experimental setups to highlight the effect of anisothermal loading. A first experimental test device was developed to study the effects of temperature transient on the cladding material during uniaxial tensile test. The experimental setup allows to reproduce loading conditions close to the ones occuring during a RIA accident. It allows clad testing up to 600 °C.s-1 heating rates coupled to rapid mechanical loading reaching 5 s-1 in terms of strain rate. First experiments showed first effects of anisothermal loading and allowed us to establish as a second step a comparison between isothermal and anisothermal states. A marked effect of anisothermal loading was observed at low strain rates and high heating rates : the flow stress is much higher than that expected from the isothermal tests. A study of the recrystallization of the material under dynamic conditions has shown that a delay in triggering the recrystallization process would be the cause of the anisothermal e ects observed during the tensile tests. A second experimental device was developed to couple effects of biaxial and anisothermal loading. A sheet of Zircaloy-4 was tested along its two main directions (rolling and transverse) with an induction heating system. Several heating rates and biaxiality ratios were explored and failure strains were determined for each experimental condition. The analysis of the tests showed that the multiaxiality of the loading is the dominant parameter with regard to the ductility of the material, no significant influence of the anisothermal loading was observed during these tests. In support of the analysis of uniaxial and biaxial anisothermal tensile tests, numerical FEM calculations were undertaken using a macroscopic mechanical behavior model developed in this study. These simulations made it possible to determine the stress fields of the biaxial tests and showed that the tests carried out were in the field of interest of the RIA studies
Roche, Stéphane. « Modélisation simplifiée de l'écoulement radial d'un mélange de matériaux fondus à travers des crayons combustibles dans un coeur REP ». Aix-Marseille 1, 1996. http://www.theses.fr/1996AIX11058.
Lozano, Nathalie. « La subdivision d'un solide induite par l'évolution de sa composition chimique : intérêt pour la céramique nucléaire a fort taux d'irradiation ». Dijon, 1998. http://www.theses.fr/1998DIJOS067.
Poitou, Benoît. « Analyse de la fissuration au voisinage d'une interface dans les matériaux fragiles : applications aux composites à matrice céramique et aux combustibles nucléaires ». Bordeaux 1, 2007. http://www.theses.fr/2007BOR13448.
Pflieger, Rachel. « Mass spectrometric study of the laser vaporisations of graphite and uranium dioxide up to 4000k ». Université Louis Pasteur (Strasbourg) (1971-2008), 2006. https://publication-theses.unistra.fr/restreint/theses_doctorat/2006/PFLIEGER_Rachel_2006.pdf.
A new method of high-temperature mass spectrometry (TOF MS) was developed, where the specimen surface is heated by a laser pulse of approx. 20 ms. During it, time-resolved measurements of mass spectra and of the temperature are performed. Each experiment covers an entire temperature interval. The method was applied to pyrolytic graphite and uranium dioxide. In graphite study, it was clearly shown that the sublimation occurs in a Langmuir-like mode (free surface vaporisation), despite the very high temperatures and thus pressures. Relative partial pressures of C1, C2, C3, C4 and C5 were measured up to 4100 K. Obtained sublimation enthalpies of the main three vapour species are in a good agreement with literature values. Relative vaporisation coefficients of C1-C5 were estimated by comparison of the present partial pressures at 4000 K with equilibrium ones given in the literature. The vapour pressure curve of UO2 over stoichiometric uranium dioxide was measured between 2800 and 3400 K. Obtained sublimation and vaporisation enthalpies are in agreement with the literature. The vaporisation enthalpy of UO3 was measured for the first time. Relative partial pressure ratios p(UO2)/p(UO), p(UO2)/p(UO3) and p(UO2+)/p(UO+) were measured at around 3300 K and indicate that the vaporisation occurs in a regime close to thermodynamic equilibrium. This method is suitable for the fast and time-resolved mass spectrometric measurements of refractory materials up to very high temperatures, and could now be applied to the study of chemically unstable materials such as hyperstoichiometric urania and some carbides and nitrides. Key words: pyrolytic graphite, HOPG, uranium dioxide, laser vaporisation, TOF MS, vaporisation coefficients, Langmuir evaporation
Schäffler, Isabelle. « Modélisation du comportement elasto-viscoplastique anisotrope des tubes de gaine du crayon combustible entre zéro et quatre cycles de fonctionnement en réacteur à eau pressurisée ». Besançon, 1997. http://www.theses.fr/1997BESA2076.
Minne, Jean-Baptiste. « Contribution à la modélisation du couplage mécanique-chimique de l'évolution de l'interface pastille-gaine sous irradiation ». Thesis, Dijon, 2013. http://www.theses.fr/2013DIJOS085/document.
Pas de résumé en anglais
Issaoui, Amal. « Comportement sous irradiation des aciers ODS (Oxide Dispersion Strengthened) pour le gainage combustible des réacteurs de 4ème génération ». Thesis, Lille 1, 2020. http://www.theses.fr/2020LIL1R008.
The extreme operating conditions envisaged for the fuel cladding of generation IV reactors (high temperature: 400°C-700°C, and high dose of irradiation: up to 150 dpa) require the development of new materials. Ferritic/martensitic steels reinforced by a dispersion of nanometric oxides (ODS: Oxide Dispersion Strengthened) are now one of the options for fissile cladding materials dedicated to the high combustion rates of a SFR. In fact, these steels exhibit a good resistance to swelling for high doses, up to 150 dpa, and a good resistance to creep deformation at high temperature thanks to the presence of nanometric oxides. However, neutron irradiation induces microchemical changes in the structure of these materials such as the separation of α-α ’phases and Cr depletion at the grain boundaries. These microstructural modifications can considerably affect the mechanical properties of these steels and could notably degrade the resistance to creep deformation and the resistance to swelling. These phenomena have been relatively little studied in ODS steels, in particular the precipitation of the ’ phase and its impact on the hardening of materials. Thus, the objective of the thesis work is to study the phenomenon of separation of the α-α ’phases as well as the behavior of grain boundaries under thermal aging, under ion irradiation and also under neutron irradiation. Excluding irradiation, the results obtained show that the precipitate ’ is formed by a non classical mechanism in ODS steels after thermal aging. It has been found that the oxide nanoreinforcements serve as a heterogeneous germination site for ’phases, thus accelerating the latter’s growth kinetics. If these phases initially harden the material significantly, their hardening effect is dependent on their kinetics of precipitation. In addition to the formation of these Cr-rich phases, Cr segregation at the grain boundaries has been demonstrated. It has been shown that enrichment in Cr is strongly dependent on the disorientation of the grain boundary and could, in the case of highly disoriented joints, cause a spinodal decomposition localized at the grain boundary. Under ion irradiation, it has been shown that the defects generate an induced Cr segregation depleting the grain boundaries, in particular in the case of an ODS Fe-14Cr alloy. ’-isolated droplets are 6 observed in the case of Fe-18Cr ODS while a mechanism of spinodal decomposition induced under irradiation has been observed in the case of Fe-14Cr ODS. The mechanisms highlighted in thermal ageing and under ion irradiation made it possible to understand the microstructures observed after neutron irradiation
Temmar, Mourad. « Simulation multiphysique du phénomène de rattrapage du jeu pastille-gaine dans les aiguilles combustibles des réacteurs à neutrons rapides ». Thesis, Aix-Marseille, 2019. http://www.theses.fr/2019AIXM0611.
The aim of this thesis is to improve the comprehension and modeling the phenomena responsible of the closure of the gap, separating initially the fuel from its surrounding cladding. A realistic simulation of the gap closure phenomenon leads to a better evaluation of the fuel temperature, which is of the first importance to meet the fuel non-fusion criterion requirement. Firstly, phenomena responsible of the fuel-to-cladding gap closure are identified. The size reduction of the fuel-to-cladding gap seems to be mainly related to two phenomena. The first one, is the effect of fuel fragmentation. The second one is related to the migration phenomenon of porosities. Thanks to 3D simulations, the impact of these two phenomena is represented. In a second step, a 1D formulation derived from 3D simulations is proposed. This formulation includes the two identified phenomena. The fuel-to-cladding gap closure is simulated by an inelastic strain called relocation strain while the porosities migration is modeled through an advection equation. This formulation is then implemented in the multiphysics computation scheme of the GERMINAL SFR 1D software. Thanks to these new developments, the fuel temperature obtained is in better agreement with the experimental results. In our 1D modeling, we have assumed that the migration velocities of the closed and open porosities are the same. However in the literature, only the closed porosity migration velocity has been evaluated. Our hypothesis therefore remains to be validated. A contribution to this validation is proposed with a 2D analysis of the evaporation condensation transfer mechanism near the free surfaces created by cracks
Lecraz, Catherine. « Etude des réactions entre l'oxyde mixte d'uranium-plutonium et le nitrure d'uranium et entre l'oxyde d'uranium et le nitrure d'uranium ». Toulouse, INPT, 1993. http://www.theses.fr/1993INPT039G.
Dowek, Rébecca. « Les gaz de fission dans les combustibles REP irradiés : un état détaillé à fort taux de combustion ». Electronic Thesis or Diss., Aix-Marseille, 2021. http://www.theses.fr/2021AIXM0248.
During the irradiation of nuclear fuel pellets, fission reactions lead to a progressive accumulation of new atoms, some of which are gaseous. These fission gases, and the bubbles they form, contribute significantly to the fuel’s behavior, whether during operation in nominal conditions or in incidental or accidental cases. This Ph.D. work provides a better description of the fission gases’ state at a micrometer scale of high burn up PWR UO2 fuels, thanks to experimental characterization campaigns and improved or new methodologies for data acquisition, processing and analysis. These campaigns were carried out in a high activity laboratory (LECA-STAR), with different microanalysis devices. Two types of fuels were examined, with different initial grain sizes and different burn-ups. The study was conducted along three main axes: morphology of the bubbles, thanks to 2D and mainly 3D FIB-SEM examinations, the microstructural evolutions, thanks to EBSD characterizations, and the quantification of the gases in order to estimate the pressure of the fission gases in the bubbles, by combining microprobe analysis, SIMS and SEM-FIB measurements. This work has allowed to establish new methodologies for fission gas and microstructure analysis. The combination of the results obtained in these different areas has led to a synthetic representation of the gas state, as a function of the radial position, the burn-up, and the initial microstructure of the fuel. This work will allow to enrich, feed and validate the calculation codes for the modeling of the UO2 fuel behavior at high burn-up
Ziouane, Yannis. « Dissolution de poudres d'oxydes mixtes (U,Pu)O2 monophasées ». Thesis, Montpellier, 2017. http://www.theses.fr/2017MONTS005.
The main objective of this study is to acquire data on the dissolution of (U, Pu)O2 compounds to support the understanding of the phenomena occurring during the dissolution steps of MOX fuels irradiated in light water or sodium fast reactors. Previous studies, in particular on unirradiated MOX fuel, have highlighted the complexity of understanding the dissolution mechanisms through a direct approach. Indeed, the dissolution depends on a large number of parameters, which are mainly chemical dissolution parameters (acidity, temperature…). But it also depends on the physico-chemical characteristics of the fuel pellets (plutonium content, homogeneity of the plutonium content, microstructure, geometry...), a majority of which being highly dependent on the manufacturing process used. To avoid getting averaged responses due to the presence of heterogeneity in the Pu distribution in pellets, it is proposed to carry out a study on single-phase compounds in the shape of powders characterized by a well-defined stoichiometry (U and Pu) and a perfectly determined morphology. A step approach allowed the determination of the key parameters controlling the dissolution kinetics of these actinide oxides (specific surface area, crystal size, Pu content, activity of nitrate ions, dissolution temperature).A global kinetics law describing the dissolution kinetics of U1-xPuxO2 oxides was established from 45 dissolution tests (with 0≤x≤1, [HNO3] and temperature ranging from 1.5 to 8.5M and from 50 to 95°C respectively). Despite the 5 orders of magnitude between dissolution kinetics of UO2 and PuO2, the model shows a good precision. Additional dissolution tests were conducted on different single-phase oxide powders to validate the predictive quality of this model
Ciszak, Clément. « Etude de l'accrochage pastille/gaine des crayons combustibles des réacteurs à eau pressurisée ». Thesis, Bourgogne Franche-Comté, 2017. http://www.theses.fr/2017UBFCK045/document.
Durability and integrity of materials used in nuclear power plants is a continuous concern of the nuclear power plant owners and developers. During the fuel irradiation in pressurised water reactors (PWR), the whole fuel-clad assembly is subjected to several irradiation-induced modifications. In particular, the fuel element expansion concomitant to the cladding creeping, leads to the contacting of both materials, allowing the oxidation of the inner side of the clad, locally at first, then tending to affect the overall cladding inner surface. At high burnup, a bonding of the fuel periphery with the metallic cladding can be observed, forming the fuel-clad bonding phenomenon, which conditions the thermal transfers and the mechanical behaviour of the fuel rods. The main objective of this PhD, is to further the knowledges of the physic-chemical interaction between fuel and clad, by identifying especially the origin of their bonding and its evolution with burnup. For that purpose, studies on inter-diffusion couples were performed on model materials both under ionic irradiation and not, completing detailed characterisations of the fuel|clad interface of samples irradiated in PWR. Comparison of the results obtained on model materials with those obtained on samples irradiated in PWR, allows making reliable assumptions on the nature, the origin and the evolution of the fuel-clad bonding in PWR
Julien, Jérôme. « Modélisation multi-échelles du couplage physico-chimie mécanique du comportement du combustible à haute température des réacteurs à eau sous pression ». Aix-Marseille 1, 2008. http://theses.univ-amu.fr.lama.univ-amu.fr/2008AIX11077.pdf.
In the Pellet-Cladding Interaction (PCI) problems of a fuel rod, it is necessary to adopt a good description of the thermomecanical behaviour of the fuel. When the fuel is subject to fluctuations in power, one of the main strains is due to the phenomenon of gaseous swelling induced by irradiation. Indeed, fuel is a porous ceramic of U02 containing several types of cavities and the accumulation of fission products in gaseous form in these cavities causes swelling of the pellet. However, this gaseous swelling has an influence on the mechanical behaviour of the pellet and particularly the viscoplastic behaviour. To improve the description of this behavior, it was necessary to develop a micromechanical model capable of coupling two phenomena modelled independently : the transfer of gas between the various cavities and the estimation of mechanical viscoplastic strains of the fuel. This thesis is to link these two disciplines from the cavities present in the fuel: mechanics calculates changes in the volume fraction of cavities according to their pressure and physical reflects the evolution of the volume fraction of cavities to calculate an internally consistent pressure. In order to describe a microstructure much richer, a new micromechanics model was developed using a multi-scale to describe the viscoplastic behavior of nuclear fuel
Kerleguer, Valentin. « Apport de l'étude de matériaux modèles U1-xPuxO2 à la compréhension des mécanismes d'altération des combustibles UOx et MOx en stockage géologique ». Electronic Thesis or Diss., Université Paris sciences et lettres, 2020. http://www.theses.fr/2020UPSLM061.
The effects of the geological disposal environment on the leaching of the oxide matrices of UOx and MOx fuels were investigated following a step-by-step procedure: carbonated water, synthetic porewater of the Collovo-Oxfordian (COx) argillite, synthetic porewater in presence of iron or argillite samples. Two types of alpha-emitting materials were considered, UO2 pellets doped with a low Pu content, and pellets of homogenous U0.73Pu0.27O2 MOx fuel. The experimental protocols did not show any significant effect of the argillite on the UO2alteration. Plutonium decreased the oxidative dissolution of U0.73Pu0.27O2 and enhanced the disproportionation of the H2O2 produced by water radiolysis. The dissolution of the MOx matrix decreased in COx water. It was strongly inhibited in presence of iron which anoxic corrosion liberated Fe2+ in solution that fully reacted with the radiolytic H2O2, leading to magnetite precipitation on the pellet surface. Geochemical (CHESS code) and reactive transport (HYTEC code) models, which were developed for the homogeneous MOx alteration,correctly simulated the main experimental data and the underlying mechanisms. The alteration processes of UOx and MOx matrices were found to be very similar under the present environmental conditions
Gasca, Petrica. « Zirconium – modélisation ab initio de la diffusion des défauts ponctuels ». Thesis, Lille 1, 2010. http://www.theses.fr/2010LIL10111/document.
Zirconium is the main element of the cladding found in pressurized water reactors, under an alloy form. Under irradiation, the cladding elongate significantly, phenomena attributed to the vacancy dislocation loops growth in the basal planes of the hexagonal compact structure. The understanding of the atomic scale mechanisms originating this process motivated this work. Using the ab initio atomic modeling technique we studied the structure and mobility of point defects in Zirconium. This led us to find four interstitial point defects with formation energies in an interval of 0.11 eV. The migration paths study allowed the discovery of activation energies, used as entry parameters for a kinetic Monte Carlo code. This code was developed for calculating the diffusion coefficient of the interstitial point defect. Our results suggest a migration parallel to the basal plane twice as fast as one parallel to the c direction, with an activation energy of 0.08 eV, independent of the direction. The vacancy diffusion coefficient, estimated with a two-jump model, is also anisotropic, with a faster process in the basal planes than perpendicular to them. Hydrogen influence on the vacancy dislocation loops nucleation was also studied, due to recent experimental observations of cladding growth acceleration in the presence of this element
Vauchy, Romain. « Etude du rapport O/M dans des nouveaux combustibles oxydes à base d'U et Pu : élaboration et caractérisation de matériaux modèles U1-y PuyO2-x ». Thesis, Grenoble, 2014. http://www.theses.fr/2014GRENI053/document.
Uranium-plutonium mixed oxides are considered within the scope of the development of nuclear fuel for the next generation of nuclear reactors (Sodium-cooled fast reactors). Because of some technological choices and safety constraints, the mixed oxide fuel will exhibit an oxygen hypostoichiometry, i.e. its Oxygen/Metal ratio (noted O/M) will be lower than 2.00. The control of this deviation from stoichiometry is essential as the O/M ratio influences numerous of the fuel properties irradiation (thermal conductivity, melting temperature, dilatation, etc.) which in turn strongly affect the behavior under. First, a special attention was paid to the fabrication of mixed oxide pellets U1-yPuyO2-x with different plutonium contents (y = 0.15 ; 0.28 and 0.45) by powder metallurgy. The two main goals were to obtain: o A homogeneous U-Pu distribution in order to have suitable materials for a thermodynamic study. o A high density of the resulting pellets in order to determine oxygen chemical diffusion coefficients within the three compounds by thermogravimetric analysis and oxygen self-diffusion coefficients by secondary ion mass spectrometry (SIMS). The second part of this study was focused on associating the O/M ratio values to the micro- and crystallographic structures of the fabricated samples. Beforehand, the qualification of the used gravimetric and thermogravimetric experimental devices dedicated to the O/M ratio measurements was performed. A reliable experimental method was then proposed for the determination of the oxygen stoichiometry of uranium-plutonium mixed oxides taking into account the presence of americium within the samples generated by natural decay of plutonium. With the aim of controlling the O/M ratio of U1-yPuyO2-x during fabrication, the influence of the cooling rate on the oxygen stoichiometry during sintering was investigated. Particularly, the crystallographic and microstructural effects of a variation in the O/M ratio during cooling were studied both at high and room-temperatures. Moreover, these effects made it possible to obtain new data on the kinetics and mechanisms of the phase separation occurring in the hypostoichiometric mixed oxides at high Pu content. Finally, the stability of U1-yPuyO2-x at room-temperature during standard storage conditions was investigated by thermogravimetry, X-ray absorption spectroscopy and X-ray diffraction. Finally, an experimental thermodynamic study of U1-yPuyO2-x was performed by thermogravimetric analysis and high-temperature X-ray diffraction as a function of temperature and oxygen partial pressure. The main factor allowing the establishment of the thermodynamic equilibrium being the oxygen diffusion, the associated chemical and self-diffusion coefficients were determined by thermogravimetry and SIMS after 16O – 18O isotopic exchange. These innovative results will allow a better understanding of the U-Pu mixed oxide properties on the basis of the point defect chemistry
Gojon, Christine. « Préparation par procédé sol-gel et évaluation des performances analytiques d'un capteur chimique spécifique de l'hydrazine ». Montpellier 2, 1996. http://www.theses.fr/1996MON20203.
Rouxel, Baptiste. « Développement d’aciers austénitiques avancés résistant au gonflement sous irradiation ». Thesis, Lille 1, 2016. http://www.theses.fr/2016LIL10187/document.
In the framework of studies about Sodium Fast Reactors (SFR) of generation IV, the CEA is developing new austenitic steel grades for the fuel cladding. These steels demonstrate very good mechanical properties but their use is limited because of the void swelling under irradiation. Beyond a high irradiation dose, cavities appear in the alloys and weaken the material. The reference material in France is a 15Cr/15Ni steel, named AIM1, stabilized with titanium. This study try to understand the role played by various chemical elements and microstructural parameters on the formation of the cavities under irradiation, and contribute to the development of a new grade AIM2 more resistant to swelling. In an analytical approach, model materials were elaborated with various chemical compositions and microstructures. Ten grades were casted with chemical variations in Ti, Nb, Ni and P. Four specific microstructures for each alloy highlighted the effect of dislocations, solutes or nano-precipitates on the void swelling. These materials were characterized using TEM and SANS, before irradiation with Fe2+ (2 MeV) ions in the order to simulate the damages caused by neutrons. Comparing the irradiated microstructures, it is demonstrated that the solutes have a dominating effect on the formation of cavities. Specifically titanium in solid solution reduces the swelling whereas niobium does not show this effect. Finally, a matrix enriched by 15% to 25% of nickel is still favorable to limit swelling in these advanced austenitic stainless steels
Lainé, Maxime. « Etude du comportement de matériaux argileux sous rayonnement ionisant ». Thesis, Université Paris-Saclay (ComUE), 2017. http://www.theses.fr/2017SACLS192/document.
The aim of this PhD thesis is to study and understand, by proposing reaction mechanisms, the behavior under irradiation of various clay materials. The systems of interest were first synthetic talc, which is the prototype of a non-swelling material. Under irradiation by accelerated electrons, the production of dihydrogen in this system, due solely to surface hydroxyl groups, is of the same order of magnitude as the one obtained in liquid water. This yield is divided by 30 in the case of natural talc from Luzenac, thus highlighting the importance of the impurities as scavengers of the precursors of dihydrogen. Synthetic smectites, which are swelling materials, were then studied.The results evidence the radiolysis of water confined in the interlayer space, leading to H2 yields which may be two to three times higher than those measured in water. Moreover, they are similar for montmorillonite and saponite, evidencing that the charge location plays only a minor role. Finally, the study of double layered hydroxides or anionic clays shows that, in this case, the nature of the anion in the interlamellar space controls the reactivity. Parallel to these measurements, electron paramagnetic spectroscopy experiments have enabled proposing reaction mechanisms. Finally, all these results are of interest in the context of the disposal of radioactive waste
Vaugoude, Adrien. « Contribution au développement d’aciers austénitiques avancés résistants au gonflement sous irradiation ». Thesis, Lille 1, 2019. http://www.theses.fr/2019LIL1R054.
In the framework on 4th generation reactors, the CEA is developing new grades of austenitic steels that will be usable, for example, for the cladding of fuels for sodium-cooling fast neutron reactors (RNR-Na). Thanks to their excellent mechanical properties and good corrosion resistance, they can be used up to 100 dpa, although their service life may be limited by the phenomenon of swelling under irradiation. Swelling is due to the formation of cavities in the material following irradiation and can cause geometric deformations and weaken the fuel claddings. The reference alloy, developed thanks to previous R&D on French RNRs, is an austenitic 15Cr/15Ni titanium stabilized steel called AIM1. This work focuses on studying and understanding the mechanisms leading to the formation of cavities under irradiation to contribute to the development of a more swell-resistant AIM2 grade. Different chemical and microstructural optimizations were investigated using an analytical approach. Three model alloys were used to study the double stabilization of titanium and niobium and several model microstructures were defined to highlight the role of microstructural parameters influencing swelling (dislocations, solutes, nanoprecipitates). Characterizations by SEM, DRX and DNPA have allowed a better understanding of the microstructural evolutions of the three grades, model microstructures and also to study their ability to form a fine network of nanoprecipitates. Very high-dose irradiations with Fe3+ ions (2MeV and 10MeV) to induce the formation of cavities have highlighted the major role of microstructure on swelling resistance. A new methodology for the study of swelling induced by ion irradiation has been proposed. It allows a statistical study of cavity formation and is based on the use of scanning microscopy. Indeed, the new detectors can acquire high definition images that can contain several thousand cavities on the same micrograph. These images are then analyzed using a supervised learning artificial intelligence algorithm to automatically recognize the cavities but as well as different objects present in the microstructure (precipitates, grain joints, etc.). An example of a study of the effect on the swelling of the irradiation damage gradient, characteristic of heavy ion irradiation, is presented as an illustration of this methodology called MEBIA. Cluster dynamic calculations simulated the impact of nanoprecipitates and the initial density of dislocations on swelling. These results inspired the creation of new microstructures that were irradiated and began to be characterized. This work will have to be continued to validate the relevance of optimized microstructures. Results presented in this manuscript illustrate the difficulties encountered in studying the microstructures of austenitic steels irradiated at very high doses, but it shows that new approaches can also be put in place to facilitate this work
Haller, Xavier. « Modélisation du comportement élastique des matériaux nanoporeux : application au combustible UO2 ». Thesis, Montpellier, 2015. http://www.theses.fr/2015MONTS232.
The irradiated uranium dioxide (UO2), which is the nuclear fuel of pressurized water reactors, contains two populations of cavities saturated by fission gaz: i. intergranular cavities almost lenticular in shape whose size ranges between few tens to several hundred nanometers, ii. intragranular cavities, almost spherical in shape whose size is of the order of the nanometer. Recent studies have shown the existence of a surface effect at the scale of nanometric cavities, which influences the effective elastic behavior of the nuclear fuel. In this work, an analytical micromechanical model, which is able to take into account this heterogeneous microstructure and the surface effect at the nanometric scale, is proposed to describe the macroscopic behavior of the irradiated UO2. The approach is based on a multiscale modeling and homogenization techniques in mechanics of materials. The irradiated UO2 is described as a porous media, which contains pressurized spherical nanocavities (intragranular cavities) and randomly oriented pressurized spheroidal cavities (intergranular cavities). The surface effect is taken into account with imperfect coherent interfaces between the matrix and the cavities. A novel model based on the morphologically representative pattern approach has been developed to describe the effective elastic behavior of this heterogeneous medium. The proposed model relies on assumptions whose relevance is evaluated with finite element simulations which require a specific formulation to take into account the imperfect coherent interfaces
Henry, Ronan. « Caractérisation locale des propriétés à la rupture du combustible nucléaire irradié ». Thesis, Lyon, 2019. http://www.theses.fr/2019LYSEI031.
The nuclear fuel UO2 of Pressurized Water Reactor (PWR) is a refractory ceramic sintered into pellets. During service, the heat produced by the nuclear reaction is transferred to the coolant by thermal conduction, leading to a significant difference of temperature between the pellet center, around 1000°C, and the pellet rim, around 500°C. At the first power rise, this gradient generates systematically large cracks which divide pellets into a few pieces. Moreover, during power transients, additional cracking is generated at the pellet rim and for simulated accidental situations, important rises of temperature lead to a complete fracturing of the fuel. Numerical simulations of the nuclear fuel behavior under irradiation needs specific properties of the material. To model the brittle cracking of the fuel in PWRs, it is necessary to experimentally measure its fracture properties and their evolution with irradiation. Nevertheless, because of pellet cracking, it is impossible to manufacture macroscopic specimens on irradiated fuel. The goal of this PhD work was to develop methods of fracture properties measurement adapted to the irradiated nuclear fuel at a room temperature. To this end, micromechanical tests has been set up to make measurements into the pieces of the cracked fuel. Two kind of tests has been studied. The first method is the nano-indentation, which has already been studied before, and were completed in this work. This method consist to make a print with a pyramidal tip on the polish surface of a sample. Depending on the load applied, cracks appear around the indentation print and the fracture toughness can be evaluated. The second method is a conventional bending test adapted to the microscopic scale. It allows the measurement of fracture toughness when the specimen is notched, and fracture stress measurement when there is no notch. To prepare such micro-specimens, a focalized ion beam (FIB) is used and a nano-indenter is employed to bend them up to fracture. To set up and validate measurements of the two methods, a model material was first used: the cubic zirconia ZrO2. The ceramic material has crystallographic and mechanical properties close to the UO2 fuel, and is not, during setting up steps, submitted to constraints linked to the nuclear environment. Then, the measurements methods has been applied to both fresh and irradiated in PWR nuclear fuel. This work showed the complementarity between the two studied methods. Indentation is a very convenient and versatile technique, which allows a large number of tests at different radial positions of irradiated fuel pellets. Micro-cantilever bending is longer to set up and use and needs several laboratory equipment, but is closer to conventional mechanical tests. It also gives needed results about fracture stress on irradiated fuels, and allows an evaluation of the resistance of specific crystallographic planes or grain boundaries, which were not accessible before on the nuclear fuel
Fallet, Alexis. « Influence des ions oxydants issus de la dissolution du combustible nucléaire usé sur le comportement des matériaux de structures ». Thesis, Montpellier, 2015. http://www.theses.fr/2015MONTS009/document.
The reprocessing of spent nuclear fuels by the PUREX process (Plutonium and Uranium Refining by Extraction) is based on a preliminary stage of dissolution which takes place in hot concentrated nitric acid. The high oxidizing power of dissolution media can induce corrosion phenomena and weaken the structural equipment exposed to it, especially stainless steels such as 304L steel. Although nitric acid is responsible of corrosion, the presence of oxidizing ions (Pu, Np ...) can change the cathodic reaction and bring the steel in its transpassive area where it may undergo intergranular corrosion. Therefore knowledge of oxidizing ions, their oxidation state, their behavior in solution and corrosion is necessary to lead to a better understanding and predict the corrosion associated risks.First, a thermodynamic model based on the heat capacity and the free enthalpies was developed to estimate the stoichiometric activity coefficients, water activities and dissociation coefficients of the binary mixture HNO3-H2O at temperatures above 25°C. The acquisition of these data is a first step in understanding the corrosion behavior of stainless steels in nitric acid in the presence of oxidizing ions.Then, an electrochemical experimental study coupled to analytical techniques enabled to understand the electrochemical behavior of the plutonium in HNO3 medium, in particular the oxidation of Pu(IV) to Pu(V) which was not identified in this medium. The acquisition of data needed for an electrochemical modeling was limited by physicochemical factors so a parametric study with a non-radioactive chemical analogue (Ce(IV)/Ce(III)) was undertaken. The determination of some analogies between plutonium and cerium has enabled to estimate the evolution of thermodynamic and kinetic constants of plutonium in condition of temperature and concentration in HNO3 higher than physicochemical limitations.Finally, the study of the corrosion behavior of 304L steel in HNO3 medium in the presence of oxidizing ions consists of two complementary studies. First a study of the electrochemical corrosion was carried out in the presence of Pu(VI) or Ce(IV). On one hand, it reveals that the Pu(VI) does not control the reduction mechanism (contrary to Ce(IV)). On the other hand, it shows that the corrosion products do not have any influence on the corrosion mechanism and highlights a Ce(IV)-Cr(III) complex which inhibits the reduction of Ce(IV). Secondly a study of chemical corrosion was undertaken through immersion tests. It has improved the knowledge of the dissolution mechanism including highlighting the presence of an extreme surface layer of Cr(VI) that could be related to a grain marking and a preliminary step of intergranular corrosion
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.
Geiger, Ernesto. « Study of Fission Products (Cs, Ba, Mo, Ru) behaviour in irradiated and simulated Nuclear Fuels during Severe Accidents using X-ray Absorption Spectroscopy, SIMS and EPMA ». Thesis, Université Paris-Saclay (ComUE), 2016. http://www.theses.fr/2016SACLS064/document.
The identification of Fission Products (FP) release mechanism from irradiated nuclear fuels during a severe accident is of main importance for the development of codes for the estimation of the source-term (nature and quantity of radionuclides released into the environment). Among the many FP Ba, Cs, Mo and Ru present a particular interest, since they may interact with each other or other elements and thus affect their release. In the framework of this thesis, two work axes have been set up in order to identify, firstly, the chemical phases initially present before the accident and, secondly, their evolution during the accident itself. The experimental approach consisted in reproducing nuclear severe accidents conditions at laboratory scale using both irradiated fuels and model materials (natural UO₂ doped with 12 FP). The advantage of these latter is the possibility of using characterization methods such as X-ray Absorption Spectroscopy which are not available for irradiated fuels. Three irradiated fuel samples have been studied, representative to an initial state (before the accident), to an intermediate stage (1773K) and to an advanced stage (2873K) of a nuclear severe accident. Regarding to model materials, many accident sequences have been carried out, from 573 to 1973K. Experimental results have allowed to establish a new release mechanism, considering both reducing and oxidizing conditions during an accident. These results have also demonstrated the importance of model materials as a complement to irradiated nuclear fuels in the study of nuclear severe accidents
Bruycker, Franck De. « High temperature phase transitions in nuclear fuels of the fourth generation ». Thesis, Orléans, 2010. http://www.theses.fr/2010ORLE2060/document.
Understanding the behaviour of nuclear materials in extreme conditions is of prime importance for the analysis of the operation limits of nuclear fuels, and prediction of possible nuclear reactor accidents, relevant to the general objectives of nuclear safety research. The main purpose of this thesis is the study of high temperature phase transitions in nuclear materials, with special attention to the candidate fuel materials for the reactors of the 4th Generation. In this framework, material properties need to be investigated at temperatures higher than 2500K, where equilibrium conditions are difficult to obtain. Laser heating combined with fast pyrometer is the method used at the European Institute for Transuranium Elements (JRC – ITU). It is associated to a novel process used to determine phase transitions, based on the detection, via a suited low-power (mW) probe laser, of changes in surface reflectivity that may accompany solid/liquid phase transitions. Fast thermal cycles, from a few ms up to the second, under almost container-free conditions and control atmosphere narrow the problem of vaporisation and sample interactions usually meet with traditional method. This new experimental approach has led to very interesting results. It confirmed earlier research for material systems known to be stable at high temperature (such as U-C) and allowed a refinement of the corresponding phase diagrams. But it was also feasible to apply this method to materials highly reactive, thus original results are presented on PuO2, NpO2, UO2-PuO2 and Pu-C systems
Ferrer, Alexandre. « Modélisation des mécanismes de formation sous ébullition locale des dépôts sur les gaines de combustible des Réacteurs à Eau sous Pression conduisant à des activités volumiques importantes ». Phd thesis, Université de Strasbourg, 2013. http://tel.archives-ouvertes.fr/tel-01072543.
Macdonald, Vincent. « Détermination d’un critère de rupture des gaines de Zircaloy-4 détendu hydruré contenant un blister d’hydrures, en conditions d’accident d’injection de réactivité ». Thesis, Paris Sciences et Lettres (ComUE), 2016. http://www.theses.fr/2016PSLEM038/document.
This study deals with the determination of a fracture criterion for hydrided, cold worked and stress relieved Zircaloy-4 fuel cladding tubes with hydride blister, during a reactivity initiated accident. Two types of fracture profiles were identified, depending on the temperature, thanks to a bibliographical study, mechanical tests and fracture profiles analysis : brittle fracture at 25°C, and ductile fracture at 350°C.At 25°C, brittle fracture was studied by a global analysis in elasto-plastic fracture mechanic. Numerical simulations were performed by a finite element method with the CAST3M code, based on mechanical tests on fuel cladding tubes with blisters. Crack tip J-integral calculations were carried out to identify a mean fracture toughness of 13,8 +/- 3,1 MPa.m1/2.At 350°C, internal pressure combined to axial tensile tests were performed on Zircaloy-4 fuel cladding tubes with hydride blisters, at stress biaxialities corresponding to those of a RIA. It was observed a ductile fracture for tubes with and without blister. It was shown that hoop strain at failure decreases when blister thickness increases, and that stress biaxiality has no effect on cladding tubes bearing a thick blister. A ductile fracture model based on the GTN model was employed and a nucleation of voids due to shear stress was introduced, based on the Lode parameter. Stress triaxiality and Lode parameter were assessed in numerical simulations to understand some experimental observations
Gouze, Benoît. « Auto organisation de semifluoroalcanes amphiphiles en milieux non-aqueux : vers un carbure de silicium à mésoporosité contrôlée ». Thesis, Montpellier, Ecole nationale supérieure de chimie, 2016. http://www.theses.fr/2016ENCM0002/document.
Silicon carbide (SiC) is a light material with numerous interesting properties: strong mechanical resistance, weak thermal expansion, good heat conductivity and chemically inert on a large range of temperatures. These characteristics make SiC an appropriate material for various applications in extreme conditions, from catalyst to generation IV nuclear fuel cladding material. Nevertheless, to fulfill these application specificities, SiC has to show high specific surface area, and a controlled porosity.We have studied the possibility to synthetize mesoporous SiC by a soft templating approach using semifluorinated alkanes (SFA) to structure a SiC molecular precursor, the 1,3,5-trisilacyclohexane (TSCH). The TSCH polymerization into polycarbosilane around SFA aggregates can structure the matrix, that will create porosity after the template removal. Then polycarbosilane is converted into a SiC by a calcination process conserving the porosity.In a first time, we studied the self-aggregation capacities of SFA in cyclohexane as model solvent, and then in TSCH, by X-ray scattering techniques and simulations of scattering patterns. We discussed the behavior of SFA and determined the parameters controlling the size of the aggregates. Then, we proceeded to SiC synthesis from TSCH in presence of SFA.As resulting materials didn’t show the expected specific surface area and porosity characteristics, we enlarged our studies to other templates such as a triblock copolymer styrene-butadiene-styrene, which finally allowed us to obtain mesoporous SiC, amorphous or crystalline, by an approach involving the grafting of the SiC precursor onto the copolymer
Meynard, Joane. « Influence de la taille, de la morphologie et de la distribution spatiale des pores sur la conductivité thermique de céramiques UO2 ». Thesis, Aix-Marseille, 2019. http://www.theses.fr/2019AIXM0607.
Inside a nuclear reactor core, the behavior of fuels is largely controlled by thermal phenomena. That is why it is very important to model the thermal behavior of fuels very precisely.The objective of this study is to develop a model that indicates the influence of porosity on thermal conductivity at 50° that is representative of the thermal behavior of the UO2 fuels. UO2 fuels were manufactured and their microstructures were studied using optical microscopy, SEM-FIB and X-ray tomography. Two types of porosity were identified: 1) sealed and near-spherical pores which are located in UO2 aggregates, and 2) an interconnected "assembly" porosity located at the interfaces of aggregates. Several descriptive parameters were estimated by immersion measurements and image analysis. Studies based on analytical and numerical homogenization were conducted. Numerical calculations using the Fast Fourier Transform method were performed on images of slice planes obtained with imaging technologies or 3D simulated microstructures generated with an original morphological model reproducing some characteristics of the observed porosity networks. The significant impact of the spatial distribution and the interconnection of the assembly porosity on the thermal conductivity of manufactured UO2 fuels were highlighted. Finally, the proposed model was compared with experimental thermal diffusivity measurements obtained by the Flash method.Discrepancies between the model and the experimental measurements have been largely reduced with the proposed model compared with the standard models, which means that the developed model is more representative of the UO2 thermal behavior
Migeon, Valérie. « Application des isotopes du molybdène en traçage des matériaux du cycle nucléaire ». Thesis, Lyon, 2016. http://www.theses.fr/2016LYSEN008/document.
Nuclear forensics aims at determining the age, provenance as well as industrial or storage history of uranium ores and uranium ore concentrates that are part of the nuclear fuel cycle. Several potential tracers have already been identified for this purpose. However, these tracers are not providing always unambiguous information. This study is focused on establishing Mo isotopes as a new tracer of uranium ore provenance and of ore processing for its application in nuclear forensics. Molybdenum and uranium share a number of common geochemical properties. In the nuclear fuel cycle, molybdenum is an impurity that is difficult to separate during uranium extraction and purification processes, while its concentration is required to be lower than some specification limits. We focused this study on the first part of the nuclear fuel cycle, from the uranium ores extraction to the production of uranium ore concentrates.We developed an enhanced separation method for Mo from a uranium-rich matrix (uranium ores, uranium minerals, uranium ore concentrates) in order to analyze the mass fractionation induced by processes typical of the nuclear fuel cycle. Molybdenum isotopic compositions in uranium ores depend of adsorption and precipitation processes. The δ98Mo values of sedimentary uranium ores is shifted to negative values relative to magmatic ores. This provides a means of distinguishing these types of uranium ores. Uranium ores concentrates produced from both uranium ore natures (magmatic and sedimentary) have Mo isotope compositions similar to the uranium ores. These results suggest that molybdenum isotopes have a strong potential of as a tracer for identifying the origin of the uranium ore concentrates. However, Mo isotopes fractionations were established during the production of uranium ore concentrates in the both Niger mills. We reproduced in laboratory the lixiviation, solvent extraction and precipitation processes to explain these observations. The Mo isotopes fractionation is positive for the lixiviation process, negative for the solvent extraction and precipitation with hydrogen peroxide, and null for ammonia precipitation. In the case of the Niger samples, the sum of these processes is negative and agrees with our experimental data. Mo isotopes have a strong potential as a tracer for identifying the origin and transformation of uranium in the nuclear fuel cycle, in the framework of nuclear forensics
Konarski, Piotr. « Thermo-chemical-mechanical modeling of nuclear fuel behavior : Impact of oxygen transport in the fuel on Pellet Cladding Interaction ». Thesis, Lyon, 2019. http://www.theses.fr/2019LYSEI080.
The goal of this thesis is to study the impact of oxygen transport on thermochemistry of nuclear fuel and pellet cladding interaction. During power ramps, nuclear fuel is exposed to high temperature gradients. It undergoes chemical and structural changes. The fuel swelling leads to a mechanical contact with the cladding causing high mechanical stresses in the cladding. Simultaneously, chemically reactive gas species are released from the hot pellet center and can interact with the cladding. The combination of these chemical and mechanical factors may lead to the cladding failure by iodine stress corrosion cracking. It has been proven that oxygen transport under high temperature gradients affects irradiated fuel thermochemistry, a phenomenon which may be of importance for stress corrosion cracking. This thesis presents 3D simulations of power ramps in pressurized water reactors with the fuel performance code ALCYONE, which is part of the computing environment PLEIADES. The code has been upgraded to couple the description of irradiated fuel thermochemistry already available with oxygen transport taking into account oxygen thermal diffusion. The impact of oxygen redistribution during a power transient on irradiated fuel thermochemistry in the fuel and on chemically reactive gas release from the fuel (consisting of I(g), I2(g), CsI(g), TeI2(g), Cs(g) and Cs2(g), mainly) is studied. The simulations show that oxygen redistribution, even if moderate in magnitude, leads to the reduction of metallic oxides (molybdenum dioxide, cesium molybdate, chromium oxide) at the fuel pellet center and consequently to the release of a much greater quantity of gaseous cesium, in agreement with post-irradiation examinations. The three-dimensional calculations of the quantities of importance for iodine stress corrosion cracking (hoop stress, hoop strain, iodine partial pressure at the clad inner wall) are then used in simulations of clad crack propagation
Pennisi, Vanessa. « Contribution à l'identification et à l'évaluation d'un combustible UO2 dopé à potentiel oxygène maîtrisé ». Thesis, Bordeaux, 2015. http://www.theses.fr/2015BORD0191/document.
Temperature and oxygen partial pressure (PO2) of nuclear oxide fuels are the main parametersgoverning both their thermochemical evolution in reactor and the speciation of volatile fissionproducts such as Cs, I or Te. An innovative way to limit the risk of cladding rupture by corrosionunder irradiation consists in buffering the oxygen partial pressure of the fuel under operation in a PO2domain where the fission gas are harmless towards Zr clad, by using solid redox buffers as additives.Niobium, with its NbO2/NbO and Nb2O5/NbO2 redox couples has been found to be a promisingcandidate to this end. A manufacturing process of a buffered UO2 fuel, doped with niobium has beenoptimized, in order to fulfill usual specifications (density, microstructure). The experimental study ofthe UO2-NbOx system has shown the existence of a liquid phase between UO2 and NbOx at 810°C,which was not reported in the literature. The characterization of Nb containing phases present in UO2both in solid solution and as precipitates has lead us to propose a solubility thermodynamic model ofniobium in UO2 at 1700°C. An extensive study of the niobium precipitates shows the co-existence inthe fuel of NbO2 and NbO as major phases, together with small amounts of metallic Nb. The coexistenceof niobium under two oxidation states inside the fuel is a key element of demonstration of apossible in-situ buffering effect, which is likely to impact some properties of the material that aredependent upon PO2, such as densification. These results confirm the promising potential of oxygenbuffered fuels as regard to their performance in reactor