Dissertationen zum Thema „Réacteurs à neutrons rapides – Accidents“
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Lainault, Franck. „Modélisation de la libération d'énergie liée aux accidents graves dans les réacteurs nucléaires à neutrons rapides“. Poitiers, 1997. http://www.theses.fr/1997POIT2308.
Andriolo, Lena. „Impact des combustibles sphere-pac innovants sur les performances de sûreté des réacteurs à neutrons rapides refroidis au sodium“. Thesis, Université Grenoble Alpes (ComUE), 2015. http://www.theses.fr/2015GREAI067/document.
Future sodium cooled fast reactors (SFRs) have to fulfill the GEN-IV requirements of enhanced safety, minimal waste production, increased proliferation resistance and high economical potential. This PhD project is dedicated to the evaluation of the impact of innovative fuels (especially minor actinides bearing oxide sphere-pac fuels) on the safety performance of advanced SFRs with transmutation option. The SIMMER-III code, originally tailored to mechanistically analyze later phases of core disruptive accidents, is employed for accident simulations. During the PhD project, the code has been extended for a better simulation of the early accident phase introducing the treatment of thermal expansion reactivity effects and for taking into account the specifics of sphere-pac fuels (thermal conductivity and gap conditions). The entire transients (from the initiating event to later accident phases) have been modeled with this extended SIMMER version. Within this PhD work, first the thermo-physical properties of sphere-pac fuel have been modeled and casted into SIMMER-III. Then, a new computational method to account for thermal expansion feedbacks has been developed to improve the initiation phase modeling of the code. The technique has the potential to evaluate these reactivity feedbacks for a fixed Eulerian mesh and in a spatial kinetics framework. At each time step, cell-wise expanded dimensions and densities are calculated based on temperature variations. Density factors are applied to the expanded densities to get an equivalent configuration (in reactivity) with original dimensions and modified densities. New cross sections are calculated with these densities and the reactivity of the equivalent configuration is computed. The developed methods show promising results for uniform and non-uniform expansions. For non-uniform expansions, model improvement needs have been identified and neutronics simulations have been carried out to support future SIMMER extensions. Preliminary results are encouraging. In the third part of the PhD, two core designs with conventional and sphere pac fuels are compared with respect to their transient behavior. These designs were established in the former CP-ESFR project: the working horse core and the optimized CONF2 core (with a large sodium plenum above the core for coolant void worth reduction). The two fuel design options are compared for steady state and transient conditions (Unprotected Loss of Flow accident, ULOF) either at beginning of life (BOL) or under irradiated conditions. Analyses for sphere-pac fuel reveal two main phases to consider at BOL. At start-up, the non-restructured sphere-pac fuel shows a low thermal conductivity compared to pellet fuel of same density. However, the fuel restructures quickly (in a few hours) due to the high thermal gradients and its thermal conductivity recovers. The fuel then shows a behavior close to the pellet one. The study also shows that the CONF2 core leads to a very mild transient for a ULOF accident at BOL. The large upper sodium plenum seems to effectively prevent large positive reactivity insertions. However, stronger reactivity and power peaks are observed under irradiated conditions or when americium is loaded in the core and lower axial blanket. This PhD work demonstrates, under current simulation conditions, that sphere-pac fuels do not seem to cause specific safety issues compared to standard pellet fuels, when loaded in SFRs. The accurate simulation of core thermal expansion reactivity feedbacks by means of the extended SIMMER version plays an important role in the accident timing (simulations confirm the expected delay in the first power peak) and on the energetic potential compared to the case where these feedbacks are omitted. The analyses also confirm the mitigating impact of a large sodium plenum on transients with voiding potential. The behavior of sphere-pac fuel in these conditions opens a perspective to its practical application in SFRs
Mathe, Emmanuel. „Comportement des radiocontaminants dans les confinements d’un réacteur à neutrons rapides refroidi au sodium en situation accidentelle“. Electronic Thesis or Diss., Lille 1, 2014. http://www.theses.fr/2014LIL10102.
In the context of the Generation IV initiative, the consequences of a severe-accident (SA) in a sodium-cooled fast reactor must be studied. A SFR (Sodium cooled Fast Reactor) severe accident involves the disruption of the core by super-criticality involving the destruction of a certain number of fuel assemblies. Subsequently the interaction between hot fuel and liquid sodium can lead to a vapor explosion which could create a breach in the primary system. Some contaminated liquid sodium would thus be ejected into the containment building. In this situation, the evaluation of potential releases to the environment (the source term) must forecast the quantity and the chemical speciation of the radiocontaminants likely to be released from the containment building
Mathe, Emmanuel. „Comportement des radiocontaminants dans les confinements d’un réacteur à neutrons rapides refroidi au sodium en situation accidentelle“. Thesis, Lille 1, 2014. http://www.theses.fr/2014LIL10102/document.
In the context of the Generation IV initiative, the consequences of a severe-accident (SA) in a sodium-cooled fast reactor must be studied. A SFR (Sodium cooled Fast Reactor) severe accident involves the disruption of the core by super-criticality involving the destruction of a certain number of fuel assemblies. Subsequently the interaction between hot fuel and liquid sodium can lead to a vapor explosion which could create a breach in the primary system. Some contaminated liquid sodium would thus be ejected into the containment building. In this situation, the evaluation of potential releases to the environment (the source term) must forecast the quantity and the chemical speciation of the radiocontaminants likely to be released from the containment building
Guyot, Maxime. „Neutronics and thermal-hydraulics coupling : some contributions toward an improved methodology to simulate the initiating phase of a severe accident in a sodium fast reactor“. Thesis, Aix-Marseille, 2014. http://www.theses.fr/2014AIXM4345.
This project is dedicated to the analysis and the quantification of bias corresponding to the computational methodology for simulating the initiating phase of severe accidents on Sodium Fast Reactors. A deterministic approach is carried out to assess the consequences of a severe accident by adopting best estimate design evaluations. An objective of this deterministic approach is to provide guidance to mitigate severe accident developments and recriticalities through the implementation of adequate design measures. These studies are generally based on modern simulation techniques to test and verify a given design. The new approach developed in this project aims to improve the safety assessment of Sodium Fast Reactors by decreasing the bias related to the deterministic analysis of severe accident scenarios.During the initiating phase, the subassembly wrapper tubes keep their mechanical integrity. Material disruption and dispersal is primarily one-dimensional. For this reason, evaluation methodology for the initiating phase relies on a multiple-channel approach. Typically a channel represents an average pin in a subassembly or a group of similar subassemblies. Inthe multiple-channel approach, the core thermal-hydraulics model is composed of 1 or 2 D channels. The thermal-hydraulics model is coupled to a neutronics module to provide an estimate of the reactor power level.In this project, a new computational model has been developed to extend the initiating phase modeling. This new model is based on a multi-physics coupling. This model has been applied to obtain information unavailable up to now in regards to neutronics and thermal-hydraulics models and their coupling
Jadon, Ankita. „Interactions between sodium carbonate aerosols and iodine fission-products“. Electronic Thesis or Diss., Université de Lille (2018-2021), 2018. http://www.theses.fr/2018LILUR021.
The safety analysis of Generation IV sodium-cooled fast neutron reactors requires the study of the consequences of a severe accident in case of release into the environment of sodium and the radionuclides it carries (term chemical and radiological source). The global source term therefore depends on both the chemical speciation of sodium aerosols, resulting from the combustion of sodium in the containment, and their interactions with radionuclides. During this thesis, the interactions between sodium carbonate and iodinated gaseous fission products (I2 and HI) were studied at the atomic and macroscopic scales, via a combined theoretical and experimental approach. An analytical expression of the adsorption isotherm has been developed. The relative stability of the sodium carbonate surfaces was determined by ab initio calculations using density functional theory. The reactivity of iodine has been studied for the most stable surfaces and the adsorption isotherms evaluated. In parallel, the kinetics of capture of molecular iodine by sodium carbonate has been determined experimentally for different boundary conditions.The results show an effective capture of the molecular iodine by sodium carbonate at 373 K, varying according to the partial pressure of iodine and the surface of the carbonate sorbent. For the representative conditions of a severe accident, the adsorption sites of the most favorable sodium carbonate surfaces will be mostly bare or doubly occupied depending on the partial pressure of molecular iodine; leading to an equilibrium pressure of less than 2x10-4 bar at 373 K
Singh, Shifali. „Radioscopie X pour les interactions corium-sodium lors d'un scénario d'accident grave“. Thesis, Université Paris-Saclay (ComUE), 2019. http://www.theses.fr/2019SACLS114.
In Sodium-cooled Fast Reactors (SFR), hypothetical failure of the core cooling system or the plant protection system may lead to a severe accident scenario. In such a scenario, core materials (fuel and cladding) melt down generating a hot molten mixture called corium. This corium may interact with the coolant (liquid sodium) leading to Fuel Coolant Interaction (FCI) which can generate energetic events and hence jeopardize the reactor structures. The yield of these energetic events strongly depends on the state of the corium-sodium mixture prior to the energetic event. Therefore, the knowledge of the features of the mixture composed of three-phases (i.e., corium, liquid sodium, and sodium vapor) is crucial. The lack of knowledge on the phenomenology of the interaction emphasizes the need to study it with the help of experiments. PLINIUS-2, the future large-mass experimental platform of CEA Cadarache, will be dedicated to experiments aiming at understanding the interaction phenomenology of prototypic corium with coolant (sodium and water). The present research aims to develop a high-energy X-Ray imaging system for this facility, to visualize and better understand the corium-sodium interaction. An image-processing algorithm to analyze the three-phase repartition is also developed to contribute to the improvement of numerical modeling. This Ph.D. research has been executed in three steps. In the first step, a bibliographic study of the past experiments was carried out to better understand the physics of the interaction and the mechanism of fragmentation during corium-sodium interaction. This bibliographic study, along with a statistical analysis of the particle size distribution data of various experiments conducted in the past, revealed that the particles formed in these tests are extremely fine fragments with characteristic diameters smaller than 1 mm. Due to the small particle size and the detection limitations of corium fragments in sodium with our X-Ray system, clouds of particles were detected instead of individual particles. In the second phase, the simulation of clouds of corium particles followed by the designing of phantoms (3D mock-ups) representing the 3-phase medium was carried out. Simulations of clouds of corium fragments in liquid sodium and vapor were performed using the CEA Cadarache in-house tool MODHERATO. Based on the results obtained from the simulations, certain phantoms were designed to conduct some physical experiments. These phantoms representative of the FCI interaction zone were manufactured to experimentally evaluate the performance of the radioscopy system and to facilitate the development and calibration of the image processing software. The third step of this work was dedicated to performing experiments with the phantoms and analyzing the radiographic images by developing an image processing algorithm. Experiments were carried out with phantoms in several configurations with the X-Ray radiography system at the CEA Cadarache KROTOS facility. The radioscopic images obtained were treated by developing a new comprehensive image processing and analysis code called PICSEL to identify the three phases composing the medium. Further verification and validation of the PICSEL software were carried out on a test conducted between corium and water at the KROTOS facility under the Euro-Chinese project “ALISA”. Thus, in this Ph.D. research, an X-Ray imaging system was qualified to visualize the corium-sodium interaction in the future PLINIUS-2-FR facility. A qualitative analysis of the images produced by this system was also performed using the PICSEL software to better characterize the evolution of the three-phase mixture and understand the FCI phenomenon, knowledge of which is deemed essential to improve the safety and designs of future sodium-cooled fast reactors
Lacourcelle, Claire. „Optimisation du procédé de décontamination des composants de réacteurs à neutrons rapides“. Aix-Marseille 3, 1994. http://www.theses.fr/1994AIX30065.
Czernecki, Sébastien. „Avancées dans le calcul neutronique des réacteurs à neutrons rapides : démonstration sur le réacteur Super-Phénix“. Aix-Marseille 1, 1998. http://www.theses.fr/1998AIX11066.
Rodet, Jean-Claude. „Contribution à l'étude de la turbulence en écoulement moyen tri-dimensionnel : cas des réacteurs nucléaires“. Ecully, Ecole centrale de Lyon, 1985. http://www.theses.fr/1985ECDL0012.
Dancre, Mathieu. „Analyse d'images tridimensionnelles ultrasonores pour l'inspection en service des réacteurs à neutrons rapides“. Aix-Marseille 2, 1999. http://www.theses.fr/1999AIX22068.
Angeli, Pierre-Emmanuel. „Simulation multi-résolution/multi-échellesde la thermohydraulique des assemblages de réacteurs à neutrons rapides“. Phd thesis, Ecole Centrale Paris, 2011. http://tel.archives-ouvertes.fr/tel-00678241.
Messaoudi, Nadia. „Etude d'un réacteur à neutrons rapides (RNR) dédié à l'incinération des actinides mineurs“. Aix-Marseille 1, 1996. http://www.theses.fr/1996AIX11011.
Brizi, Julie. „Cycles uranium et thorium en réacteurs à neutrons rapides refroidis au sodium : Aspects neutroniques et déchets associés“. Phd thesis, Université Paris Sud - Paris XI, 2010. http://tel.archives-ouvertes.fr/tel-00545616.
Privas, Edwin. „Contribution à l’évaluation des incertitudes sur les sections efficaces neutroniques, pour les réacteurs à neutrons rapides“. Thesis, Université Grenoble Alpes (ComUE), 2015. http://www.theses.fr/2015GREAI109/document.
The thesis has been motivated by a wish to increase the uncertainty knowledge on nuclear data, for safety criteria. It aims the cross sections required by core calculation for sodium fast reactors (SFR), and new tools to evaluate its.The main objective of this work is to provide new tools in order to create coherent evaluated files, with reliable and mastered uncertainties. To answer those problematic, several methods have been implemented within the CONRAD code, which is developed at CEA of Cadarache.After a summary of all the elements required to understand the evaluation world, stochastic methods are presented in order to solve the Bayesian inference. They give the evaluator more information about probability density and they also can be used as validation tools. The algorithms have been successfully tested, despite long calculation time.Then, microscopic constraints have been implemented in CONRAD. They are defined as new information that should be taken into account during the evaluation process. An algorithm has been developed in order to solve, for example, continuity issues between two energy domains, with the Lagrange multiplier formalism. Another method is given by using a marginalization procedure, in order to either complete an existing evaluation with new covariance or add systematic uncertainty on an experiment described by two theories. The algorithms are well performed along examples, such the 238U total cross section.The last parts focus on the integral data feedback, using methods of integral data assimilation to reduce the uncertainties on cross sections. This work ends with uncertainty reduction on key nuclear reactions, such the capture and fission cross sections of 238U and 239Pu, thanks to PROFIL and PROFIL-2 experiments in Phénix and the Jezebel benchmark
Khamakhem, Wassim. „Etude de l'évolution du combustible dans des réacteurs rapides de quatrième génération : impact des données nucléaires sur leur performance“. Paris 11, 2010. http://www.theses.fr/2010PA112173.
The objective of this PhD topic is to contribute to the understanding of the variations of the core neutronic characteristics of the 4th generation reactors (Sodium Cooled Fast Reactors (SFR) and Gas Cooled Fast Reactors (GFR)) during fuel depletion. The neutron characteristics of interest are of course the burn up reactivity swing and the breeding gain but also the Doppler effect and the coolant void effect. Fuel depletion leads to a degradation of the core safety parameters. The study of these variations and their associated uncertainties contributes to justify 4th generation reactor core designs as envisaged in their last developments. These last developments concerned Sodium Cooled Fast Reactors (SFR) and Gas Cooled Fast Reactors (GFR) which were reshaped in order to meet Generation IV goals on economics, safety and reliability, sustainability and proliferation resistance. They exhibit very innovative characteristics compared to the European Fast Reactor (EFR) whose design was very much in line with those of Phenix and Super Phenix. Recent CEA studies had led to large 3600 MWth SFR cores using oxide fuel and to large 2400 MWth GFR cores using carbide fuel. Since the designs have to balance between positive breeding gain and safety characteristics such as rather low void reactivity effects (SFR) or rather sm ail core pressure drop (GFR), scoping studies for breakthrough SFR cores were performed using dense fuels either carbide (already taken as a reference for the GFR core) or metal. These preliminary breakthrough SFR images are characterized by high power density and highly positive breeding gain (Breeding Gain = 0. 17). As a first step towards the development of GFR plants, a low power experimental GFR called ALLEGRO is being envisaged and has been studied for its peculiar characteristics. To study the main neutronic characteristics of these cores, one can use analyses based on the sensitivity methods of the deterministic computer code ERANOS (neutronic code system). These methods are available in statics without the possibility of taking into account fuel depletion. Ln order to mitigate this insufficiency, a subsequent part of the thesis consisted in developing the depletion perturbation theory which requires to couple Boltzmann and Bateman equations and allows a more precise understanding of the behaviour of the previous cores. The method is now able to calculate the sensitivity of the actinides and fission products concentrations and of neutron characteristics of interest such as breeding gain, Doppler reactivity effect and the coolant void reactivity coefficient effect. Ln order to illustrate these sensitivity developments, uncertainties of the neutron characteristics have been calculated using a preliminary variance covariance matrix called BOLNA. The uncertainty analyses highlight the contribution of each isotope to the neutron characteristics of the various core designs. This determination has given relatively small uncertainty variations with burn up when possible modifications of nuclear data are applied. The in-depth study performed on sodium nuclear data evaluations (ENDFB-VII, JEFF-3. 1, JENDL-3. 3) highlight the difficulty of creating accurate enough nuclear data and their associated covariance matrix. It appears hence that although the feasibility of these core designs are not questioned (relatively optimistic values being calculated are within the target value of 700 pcm for the reactivity swing and 7% for the reactivity coefficient), their performance will require integral experiments both to confirm what has been evaluated with nuclear data covariance matrices and to reduce nuclear data uncertainties. Lastly, the sensitivity methods are used to explain the peculiar behaviour of integral characteristics Iike the void effect or the Doppler effect with depletion in the GFR and ALLEGRO cores. One reason was track back to the difference in size of the two cores but also to the different structural materials being used. Furthermore, the building up of Pu239 fission products and the change in Pu239 and Pu241 isotopes being different induce divergent behaviour of both Doppler with time. For the SFR, the distribution of the void effect in the various core zones which present different fuel depletion histories is finally analyzed to be compared to that of the power distribution and finally to that of the breeding gain. It appears that the SFR core design with a rather flat internai breeding gain has, as a consequence, a rather flat void effect which is another nice feature. One concludes on the advantages resulting from the last core designs as weil as their degree of performance from the view point of computational tools very dependent at first on the nuclear data knowledge
Pilarski, Stevan. „Etude du potentiel de concepts innovants de réacteurs à neutrons rapides (RNR) vis-à-vis des exigences du développement durable“. Paris 11, 2008. http://www.theses.fr/2008PA112287.
This thesis investigates innovative fast reactors (FRs) and the possibility of using different liquid-metal coolants (Pb, Pb-Li7, Pb-Mg). The comparison is made for the requirements of “fourth generation” reactors, as defined in the framework of the Generation IV international forum, and in particular reactor safety. More specifically, two important safety criteria are studied in detail: the issue of the void reactivity effect, which is an inherent drawback of FRs; and the behaviour of the reactor during unprotected transients. The potential of these innovative coolants is investigated through simulations of their use in a concept similar to the Russian project BREST-300. Two options were considered for the fuel: an oxide fuel, which is a mature technology, and an innovative, denser, nitride fuel. A study at low power was complemented by investigations of industrial size cores. A concept similar to the SFR reactor (Sodium Fast Reactor) is taken as the state-of-the-art reference point. These concepts were initially dimensioned using parametric studies covering the main core geometry options, taking into account both neutronics and thermal-hydraulics aspects in steady-state conditions. Furthermore, we propose a systematic approach for evaluating the tolerance to the main accident types (loss of primary flow, transient over-power, loss of heat sink, and over-cooling) based on a quasi-static reactivity approach complemented by the use of a dynamics code. In addition to these studies based on the U-Pu fuel cycle, we show the advantage of using the Th-U fuel cycle as an effective way of reducing the void effect for these innovative FR concepts
Maury, Cécile. „Spectroscopies analytiques innovantes pour l'amélioration de la sûreté des réacteurs nucléaire à neutrons rapides refroidis au sodium“. Paris 6, 2012. http://www.theses.fr/2012PA066428.
Lefevre, Emmanuel. „Mise au point et validation d'un nouveau formulaire adapté au calcul des protections neutroniques des réacteurs à neutrons rapides“. Aix-Marseille 1, 1996. http://www.theses.fr/1996AIX11055.
Tran, Van De. „Contribution au remplacement des revêtements durs par traitement de surface non conventionnel dans les réacteurs à neutrons rapides“. Thesis, Châtenay-Malabry, Ecole centrale de Paris, 2014. http://www.theses.fr/2014ECAP0063/document.
This thesis contributes to the replacement of the coating of Stellite 6 which is used in friction areas for the primary circuit of the reactor fast neutron.It contains three parts:1) A literature review for selecting the materials and the deposition process2) A parametric study to get healthy deposits (good adhesion with the substrate, little porosity, no cracks, low dilution)3) A study wear behavior of deposits obtained, at high temperature (200°C) under an atmosphere inert gas, to determine the wear resistance of materials selected without the influence of an eventual oxidation layer.From the literature review, it appears the following choices implemented in our study : * the method laser cladding with advantages such as: - Good adhesion (metallurgical) - High cooling speed - Low dilution rate - Wide parametric range * two nickel-based alloys: Colmonoy-52 and Tribaloy-700. These alloys have good dry wear behavior and could be deposited by the laser. In the manufacturing part of the healthy deposit, firstly, we characterized the metal powder. Then, a parametric study was performed to look for a good parametric range that makes us getting a healthy deposit of Stellite 6 (reference) of Colmonoy-52 and Tribaloy-700. In this case, relationships among three main process parameters laser cladding (laser beam power, surface scanning speed, rate of powder) with the microstructure and chemical composition of the deposit are studied. In study the wear behavior, a pin-on-disc type of tribological was used and tests were carried out in argon at room temperature and 200°C. We investigated the wear mechanism of the best deposition of Stellite 6, Colmonoy-52 and Tribaloy-700. The wear resistance of these materials were thourghly compared
Maury, Cécile. „Spectroscopies analytiques innovantes pour l'amélioration de la sûreté des réacteurs nucléaires à neutrons rapides refroidis au sodium (RNRNa)“. Phd thesis, Université Pierre et Marie Curie - Paris VI, 2012. http://tel.archives-ouvertes.fr/tel-00807954.
Na, Byung Chan. „Etude de conception neutronique des coeurs de RNR [Réacteurs à Neutrons Rapides] visant à améliorer leur potentiel de sureté“. Aix-Marseille 1, 1996. http://www.theses.fr/1996AIX11085.
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
Benoit, Jean-Christophe. „Développement d'un code de propagation des incertitudes des données nucléaires sur la puissance résiduelle dans les réacteurs à neutrons rapides“. Phd thesis, Université Paris Sud - Paris XI, 2012. http://tel.archives-ouvertes.fr/tel-01064275.
Benoit, Jean-christophe. „Développement d’un code de propagation des incertitudes des données nucléaires sur la puissance résiduelle dans les réacteurs à neutrons rapides“. Thesis, Paris 11, 2012. http://www.theses.fr/2012PA112254/document.
This PhD study is in the field of nuclear energy, the back end of nuclear fuel cycle and uncertainty calculations. The CEA must design the prototype ASTRID, a sodium cooled fast reactor (SFR) and one of the selected concepts of the Generation IV forum, for which the calculation of the value and the uncertainty of the decay heat have a significant impact. In this study is developed a code of propagation of uncertainties of nuclear data on the decay heat in SFR.The process took place in three stages.The first step has limited the number of parameters involved in the calculation of the decay heat. For this, an experiment on decay heat on the reactor PHENIX (PUIREX 2008) was studied to validate experimentally the DARWIN package for SFR and quantify the source terms of the decay heat.The second step was aimed to develop a code of propagation of uncertainties : CyRUS (Cycle Reactor Uncertainty and Sensitivity). A deterministic propagation method was chosen because calculations are fast and reliable. Assumptions of linearity and normality have been validated theoretically. The code has also been successfully compared with a stochastic code on the example of the thermal burst fission curve of 235U.The last part was an application of the code on several experiments : decay heat of a reactor, isotopic composition of a fuel pin and the burst fission curve of 235U. The code has demonstrated the possibility of feedback on nuclear data impacting the uncertainty of this problem.Two main results were highlighted. Firstly, the simplifying assumptions of deterministic codes are compatible with a precise calculation of the uncertainty of the decay heat. Secondly, the developed method is intrusive and allows feedback on nuclear data from experiments on the back end of nuclear fuel cycle. In particular, this study showed how important it is to measure precisely independent fission yields along with their covariance matrices in order to improve the accuracy of the calculation of the decay heat
Grosjean, Cédric. „Mesure de la section efficace de fission induite par neutrons rapides des noyaux ²³²Th / ²³³U dans le cadre des cycles de combustible innovants“. Bordeaux 1, 2005. http://www.theses.fr/2005BOR12967.
The thorium ²³²Th- ²³³U fuel cycle might provided safer and cleaner nuclear energy than the present Uranium/ Pu fuelled reactors. Over the last 10 years, a vast campaign of measurements has been initiated to bring the precision of neutron data for the key nuclei (²³²Th, ²³³Pa and ²³³U) at the level of those of the U- Pu cycle. This is the framework of these measurements, the energy dependent neutron induced fission cross section of ²³²Th and ²³³U has been measured from 1 to 7 MeV with a target accuracy lesser than 5 %. These measurements imply the accurate determination of the fission rate, the number of the target nuclei as well as the incident neutron flux impinging on the target, the latter has been obtained using the elastic scattering (n, p). The cross section of which is the very well known in a large neutron energy domain (~ 0,5 % from 1 eV to 50 MeV) compared to the 235U(n, f) reaction. This technique has been applied for the first time to the 232Th(n, f) and ²³³U(n, f) cases. A Hauser- Feshbach statistical model has been developed. It consists of describing the different decay channels of the compound nucleus 234U from 0,01 to 10 MeV neutron energy. The parameters of this model were adjusted in order to reproduce the measured fission cross section of ²³³U. From these parameters, the cross sections from the following reactions could be extracted: inelastic scattering ²³³U(n, n'), radiative capture ²³³U(n, ) and ²³³U(n, 2n). These cross sections are still difficult to measure by direct neutron reactions. The calculated values allow to fill the lack of experimental data for the major fissile nucleus of the thorium cycle
Casalta, Sylvie. „Etude des propriétés du système Am-O en vue de la transmutation de l'Americium 241 en réacteur à neutrons rapides“. Aix-Marseille 1, 1996. http://www.theses.fr/1996AIX11036.
Redon, Bruno. „Étude et modélisation du piégeage sur matériaux carbones du césium 137 lors de la purification du sodium primaire des réacteurs à neutrons rapides“. Grenoble INPG, 1998. http://www.theses.fr/1998INPG0098.
Khatcheressian, Nayiri. „Développement d’un modèle de transferts couplés pour l’aide à la conception et à la conduite des systèmes de purification du sodium des réacteurs à neutrons rapides“. Phd thesis, Toulouse, INPT, 2013. http://oatao.univ-toulouse.fr/10883/1/khatcheressian.pdf.
Cabrero, Julien. „Amélioration de la conductivité thermique des composites à matrice céramique pour les réacteurs de 4ème génération“. Thesis, Bordeaux 1, 2009. http://www.theses.fr/2009BOR13877/document.
Cai, Li. „Condensation et homogénéisation des sections efficaces pour les codes de transport déterministes par la méthode de Monte Carlo : Application aux réacteurs à neutrons rapides de GEN IV“. Thesis, Paris 11, 2014. http://www.theses.fr/2014PA112280/document.
In the framework of the Generation IV reactors neutronic research, new core calculation tools are implemented in the code system APOLLO3® for the deterministic part. These calculation methods are based on the discretization concept of nuclear energy data (named multi-group and are generally produced by deterministic codes) and should be validated and qualified with respect to some Monte-Carlo reference calculations. This thesis aims to develop an alternative technique of producing multi-group nuclear properties by a Monte-Carlo code (TRIPOLI-4®).At first, after having tested the existing homogenization and condensation functionalities with better precision obtained nowadays, some inconsistencies are revealed. Several new multi-group parameters estimators are developed and validated for TRIPOLI-4® code with the aid of itself, since it has the possibility to use the multi-group constants in a core calculation.Secondly, the scattering anisotropy effect which is necessary for handling neutron leakage case is studied. A correction technique concerning the diagonal line of the first order moment of the scattering matrix is proposed. This is named the IGSC technique and is based on the usage of an approximate current which is introduced by Todorova. An improvement of this IGSC technique is then presented for the geometries which hold an important heterogeneity property. This improvement uses a more accurate current quantity which is the projection on the abscissa X. The later current can represent the real situation better but is limited to 1D geometries.Finally, a B1 leakage model is implemented in the TRIPOLI-4® code for generating multi-group cross sections with a fundamental mode based critical spectrum. This leakage model is analyzed and validated rigorously by the comparison with other codes: Serpent and ECCO, as well as an analytical case.The whole development work introduced in TRIPLI-4® code allows producing multi-group constants which can then be used in the core calculation solver SNATCH in the PARIS code platform. The latter uses the transport theory which is indispensable for the new generation fast reactors analysis. The principal conclusions are as follows:-The Monte-Carlo assembly calculation code is an interesting way (in the sense of avoiding the difficulties in the self-shielding calculation, the limited order development of anisotropy parameters, the exact 3D geometries) to validate the deterministic codes like ECCO or APOLLO3® and to produce the multi-group constants for deterministic or Monte-Carlo multi-group calculation codes. -The results obtained for the moment with the multi-group constants calculated by TRIPOLI-4 code are comparable with those produced from ECCO, but did not show remarkable advantages
Daudin, Kevin. „Contribution à la prédiction des effets réactions sodium-eau : application aux pertes de confinement dans un bâtiment générateur de vapeur d'un réacteur à neutrons rapides refroidi au sodium“. Thesis, Compiègne, 2015. http://www.theses.fr/2015COMP2212/document.
Study of sodium-water reaction (SWR) consequences in open air represents a challenge in the frame of safety assessments of sodium fast reactors (SFR). In case of major accident and to predict consequences of SWR, it is necessary to better appreciate phenomena and especially quantity and rate of the energy releasement. The objective is thus to strengthen the understanding of such reactions in order to predict with lore accuracy its consequences on mechanical equipment in the surroundings. This work focuses on three areas : research of accidental sequences, experimental investigation, and phenomenological analysis before the explosive contact. At first, a tree structure risk analysis with calculations of dangerous phenomena permitted to suggest how the contact between reactants may happen. Then, demonstrative experimental studies were performed to deepen some practical aspects of the phenomenology, like the influence of the way the reactants get in contact. Data analysis conducted to the development of a phenomenological model, implemented into a software platform for numerical simulations. Although numerous hypothesis, transient heat transfer consideration enables to reproduce experimental observations, especially the influence of mixing conditions (sodium mass and initial temperatures) on the phenomenology. This study of the premixing step of sodium-water explosion is relevant in the frame of current prediction methods of mechanical loadings on structures
Dumas, Jean-Christophe. „Étude des conditions de formation du joint oxyde-gaine dans les combustibles des réacteurs à neutrons rapides : observations et proposition d'un modèle de comportement des produits de fission volatils“. Grenoble INPG, 1995. http://www.theses.fr/1995INPG0098.
Grosjean, Cédric. „Mesure de la section efficace de fission induite par neutrons rapides des noyaux 232Th / 233U dans le cadre des cycles de combustiblesinnovants“. Phd thesis, Université Sciences et Technologies - Bordeaux I, 2005. http://tel.archives-ouvertes.fr/tel-00404551.
Bouret, Cyrille. „Etudes des contre-réactions dans un réacteur à neutrons rapides à caloporteur sodium : impact de la conception et de la neutronique sur les incertitudes“. Thesis, Clermont-Ferrand 2, 2014. http://www.theses.fr/2014CLF22508/document.
Fast reactors (FR) can give value to the plutonium produced by the existing light water reactors and allow the transmutation of a significant part of the final nuclear waste. These features offer industrial prospects for this technology and new projects are currently studied in the world such as ASTRID prototype in France. Future FRs will have also to satisfy new requirements in terms of competitiveness, safety and reliability. In this context, the new core concept envisaged for ASTRID incorporate innovative features that improve the safety of the reactor in case of accident. The proposed design achieves a sodium voiding effect close to zero: it includes a fertile plate in the middle of the core and a sodium plenum in the upper part in order to increase the neutron leakage in case of sodium voiding. This heterogeneous design represents a challenge for the calculation tools and methods used so far to evaluate the neutronic parameters in traditional homogeneous cores. These methods have been improved over the thesis to rigorously treat the neutron streaming, especially at the mediums interfaces. These enhancements have consisted in the development of a specific analysis methodology based on perturbation theory and using a modern three dimensional Sn transport solver. This work has allowed on the one hand, to reduce the bias on static neutronic parameters in comparison with Monte Carlo methods, and, on the other hand, to obtain more accurate spatial distributions of neutronic effects including the reactivity feedback coefficients used for transient analysis. The analysis of the core behavior during transients has also allowed estimating the impact of reactivity feedback coefficients assessment improvements. In conjunction with this work, innovative methods based on the evaluation of local sensitivities coefficients have been proposed to assess the uncertainties associated to local reactivity effects. These uncertainties include the correlations between the different local parameters. The propagation during transients with these methods has allowed an estimation of temperature distributions achieved in the core and also to determine the available safety margins before sodium boiling
Guo, Hui. „Design of innovative systems for the optimized control of reactivity in Gen-IV fast neutron reactors“. Thesis, Aix-Marseille, 2019. http://www.theses.fr/2019AIXM0245.
The Generation-IV reactors could benefit from the fast neutron spectrum to maximize the utilization of uranium resources, improve the management of fissile materials, and help the transmutation of nuclear waste. As the absorption cross-sections decrease with incident neutron energy, the fast spectrum challenges its reactivity control.The conventional control rod is a cluster of open pins with boron carbide (B4C) as the absorber. ^10B enrichment can be adjusted to satisfy the requirements of different cores. However, the operating lifetime of B4C is limited due to its characteristics under irradiation. Alternative absorbers such as gadolinium oxide (Gd2O3), europium oxide (Eu2O3) and hafnium diboride (HfB2) may present some advantages and be used with local addition of moderators to optimize the design of control rods in sodium fast reactors (SFRs).In the conventional fast reactors, the control rod is usually the only reactivity control system, which would lead to fuel melting in control rod withdrawal (CRW) accidents. Therefore, two burnable poison (BP) designs are investigated to reduce core excess reactivity and thus improve the inherent safety performance of reactors. The first BP design load minor actinides in homogenous or hybrid mode. The second BP design combines depleted B4C and moderators in dedicated assemblies.These designs are investigated using the advanced calculation scheme in APOLLO3® that is developed and validated in this thesis. These designs are applied in a large industrial SFR and a small modular SFR, which proves their excellent flexibility to optimize reactivity control in a wide range of fast reactors
Garti, Sara. „Spectrométrie gamma haute résolution et bas bruit Compton pour la détection des ruptures de gaine dans les réacteurs rapides refroidis au sodium“. Thesis, Université Grenoble Alpes, 2020. https://tel.archives-ouvertes.fr/tel-02862768.
Fourth generation reactors are being developed for renewing the nuclear energy industry with safer reactors that optimize uranium resources and produce less long-lived radioactive waste. France has focused on the development of SFR “Sodium Fast Reactor”. In the past, prototypes have been built and operated such as Rapsodie, Phénix, Superphénix. Since 2006, the CEA (French Atomic Energy Commission) has overseen the design of the future prototype of this technology: ASTRID. It is within this framework that teams at the French Atomic Energy Commission are working on the development of instrumental means to guarantee the absolute integrity of the first containment barrier: the fuel cladding. Indeed, the safety level to be achieved in the fourth-generation program requires the operator to continuously monitor the fission products released into the primary coolant. To ensure this safety function, various systems were implemented, such as delayed neutron detection (DND) and gamma spectrometry systems. These systems have been upgraded since their first operations through R&D programs in line with technological advances and new digital design methods.The historical gamma spectrometry instrumentation for clad failure monitoring is compromised by the presence of significant background that leads to delay lines of about 15 minutes necessary for the 23Ne (T1/2 =38 s) decay. In addition, the other background source, i.e. 41Ar, with a half-life of 110 minutes, was not filtered by this instrumentation. Indeed, it loses only 7 % of its activity in these delay lines.Regarding the safety requirement set for SFR, we will study in the frame of this work the potential contribution of a low background instrumentation by means of a Compton suppression system integrated into the SFR measuring device. The main goal of such an instrumentation would be, first, to strengthen the diagnostic of the detection of short-period fission products (about 3 min) which could indicate an early contact between the fuel and the coolant that can be dangerous for the reactor safety, second, to ensure a fast detection by improving the measuring response time.First, the source term of fission products was characterized by putting into equation various physical phenomena that govern the behavior of fission gas in an SFR environment following a clad failure. Expected fission products activities have been estimated. Second, we performed a numerical study, by means of the Monte Carlo method, of a low background instrumentation, from its experimental validation to its optimization. Finally, we implemented the model for the problem of cladding failure in the case of SFR environment. Minimum detectable activities have been estimated then compared to expected activities. The added value of this instrumentation for applications to the early detection of clad failures in an SFR environment will then be exposed
Roumiguier, Lena. „Frittage par Spark Plasma Sintering de céramiques de carbure de bore : modélisation numérique du procédé et optimisation des nano-,microstructures pour l’amélioration des performances des absorbants en réacteurs à neutrons rapides“. Thesis, Limoges, 2019. http://www.theses.fr/2019LIMO0109.
This study aims at developing new boron carbide materials, used as neutron absorbers for fast neutron reactors. The defined strategy is to refine the microstructure to limit the anisotropic swelling of grains under irradiation, responsible for the premature deterioration of pellets. To this end, submicronic and nanometric powders were densified by Spark Plasma Sintering. Two materials were elaborated by Spark Plasma Sintering with submicronic or nanometric microstructures, allowing a reduction in grain sizes compared to the reference material historically used in fast reactors by the CEA. SPS and reference materials were characterized and compared in terms of chemical composition, mechanical and thermal properties. This study led to the selection of the submicronic material and to further investigation regarding flexural strength and thermal shock resistance. The performance was improved compared to the reference. Moreover, the creep behavior at high temperature was characterized and creep parameters were identified. Furthermore, the production of absorbent pellets require to increase the height/diameter ratio compared to classical SPS pellets. Process data necessary for this modeling were obtained using specific thermal and electrical measurements. In addition, the densification parameters of the SPS material were determined from a nonlinear viscous flow model. The thermal, electrical and mechanical phenomena numerically described were then validated by confrontation with experimental monitoring of boron carbide sintering
Pantera, Laurent. „Application d'une méthodologie statistique à la compréhension du phénomène de corrosion du surgénérateur Phénix“. Compiègne, 1992. http://www.theses.fr/1992COMPD509.
Moriot, Jérémy. „Détection vibro-acoustique passive d’une réaction sodium-eau par formation de voies dans un générateur de vapeur d’un réacteur nucléaire à neutrons rapides refroidi au sodium“. Thesis, Lyon, INSA, 2013. http://www.theses.fr/2013ISAL0151/document.
This thesis deals with a new method to detect a sodium-water reaction in a steam generator of a fast sodium-cooled nuclear reactor. More precisely, the objective is to detect a micro-leak of water (flow < 1 g/s) in less than 10 seconds by measuring the external shell vibrations of the component. The strong background noise in operation makes impossible the use of a detection system based on a threshold overrun. A beamforming method applied to vibrations measured by a linear array of accelerometers is developed in this thesis to increase the signal-to-noise ratio and to detect and locate the leak in the steam generator. A numerical study is first realized. Two models are developed in order to simulate the signals measured by the accelerometers of the array. The performances of the beamforming are then studied in function of several parameters, such as the source location and frequency, the damping factor, the background noise considered. The first model consists in an infinite plate in contact with a heavy fluid, excited by an acoustic monopole located in this fluid. Analyzing the transverse displacements in the wavenumber domain is useful to establish a criterion to sample correctly the vibration field of the plate. A second model, more representative of the system is also proposed. In this model, an elastic infinite cylindrical shell, filled with a heavy fluid is considered. The finite dimensions in the radial and circumferential directions lead to a modal behavior of the system which impacts the beamforming. Finally, the method is tested on an experimental mock-up which consists in a cylindrical pipe made in stainless steel and filled with water connected to hydraulic circuit. The water flow speed can be controlled by varying the speed of the pump. The acoustic source is generated by a hydrophone. The performances of the beamforming are studied for different water flow speeds and different amplitude and frequencies of the source
Jadon, Ankita. „Interactions between sodium carbonate aerosols and iodine fission-products“. Thesis, Lille 1, 2018. http://www.theses.fr/2018LIL1R021/document.
The safety analysis of Generation IV sodium-cooled fast neutron reactors requires the study of the consequences of a severe accident in case of release into the environment of sodium and the radionuclides it carries (term chemical and radiological source). The global source term therefore depends on both the chemical speciation of sodium aerosols, resulting from the combustion of sodium in the containment, and their interactions with radionuclides. During this thesis, the interactions between sodium carbonate and iodinated gaseous fission products (I2 and HI) were studied at the atomic and macroscopic scales, via a combined theoretical and experimental approach. An analytical expression of the adsorption isotherm has been developed. The relative stability of the sodium carbonate surfaces was determined by ab initio calculations using density functional theory. The reactivity of iodine has been studied for the most stable surfaces and the adsorption isotherms evaluated. In parallel, the kinetics of capture of molecular iodine by sodium carbonate has been determined experimentally for different boundary conditions.The results show an effective capture of the molecular iodine by sodium carbonate at 373 K, varying according to the partial pressure of iodine and the surface of the carbonate sorbent. For the representative conditions of a severe accident, the adsorption sites of the most favorable sodium carbonate surfaces will be mostly bare or doubly occupied depending on the partial pressure of molecular iodine; leading to an equilibrium pressure of less than 2x10-4 bar at 373 K
Tillard, Léa. „Impact du déploiement de réacteurs de type ASTRID sur la gestion dynamique du plutonium dans des scénarios de transitions électronucléaires“. Thesis, Université Paris-Saclay (ComUE), 2019. http://www.theses.fr/2019SACLS494.
All the laws, decrees and public debates relating to the energy transition, emphasize the importance of the study of electronuclear fleet evolution scenarios. One of the reference strategies for the French electronuclear fleet evolution considers the step by step deployment of Generation IV Sodium-cooled Fast Reactors (SFR). A proper assessment of the possible transitions scenarios requires a thorough study of the different possible trajectories and its associated consequences on the entire fuel cycle.In this framework, this Ph.D. work aims at analyzing the impact on plutonium and minor actinide dynamic management, of ASTRID-like reactor deployment scenarios, a Generation IV SFR developed by the CEA and its industrial partners. The modeling of two ASTRID-like reactor configurations, one plutonium break-even, and one burner, allow the validation of the calculation hypothesis, the quantification of associated bias and the verification of reactor safety coefficients. It was observed that the variation of initial fuel composition had a drastic impact on the system configuration. Within the framework of this research, the dynamic fuel cycle simulator CLASS, developed by the CNRS/IN2P3 and the IRSN was further modified, to meet the requirement of new dedicated complex physics models. These new developments using multidimensional and nonlinear interpolators allow modeling of the fresh fuel fabrication and irradiation while maintaining the reactor heterogeneity throughout the simulations. With these multizone models, effects of SFR deployment is studied, and potential constraints on in-cycle materials are identified by the simulation of transition scenarios, from a Pressurized Water Reactor fleet to a mixed fleet integrating SFRs. An academic analysis of the scenarios presented within the energy transition law is proposed to extend this work
Jourdy, Benjamin. „Analyse des effets d'échelle sur le comportement thermo-hydraulique de jets impactants“. Electronic Thesis or Diss., Université de Lorraine, 2023. http://docnum.univ-lorraine.fr/ulprive/DDOC_T_2023_0063_JOURDY.pdf.
Safety studies and numerical modelling implemented within the framework of studies on fast neutron reactors imply a need for validation of calculation codes using representative mock-ups. A safety issue identified in this type of reactor is the rise of the radial jet resulting from the impact of hot jets from the core to the Upper Core Structure (UCS). This rise induces thermal fluctuations and thermal stratification in the upper plenum causing thermal fatigue on the components. Understanding the various mechanisms at work in the phenomenon of radial jet raising is necessary in order to be predictive about the operating conditions for which this phenomenon occurs. For this purpose, the MICAS mock-up is used, which geometrically reproduces the upper plenum of the ASTRID reactor on a 1/6 scale, using water as a simulating fluid instead of sodium. In order to validate the ASTRID simulation codes, the representativeness of the MICAS mock-up must be validated with respect to its reduced scale but also to its simulating fluid. In order to study the transposition of the jet behaviour between different scales, a method of scale effects analysis is introduced, based on three mock-ups at different scales but representative of the reactor. A phenomenological study is carried out on a simplified model, PIGNIA, allowing to verify experimentally the theoretical dependence of the trajectory of the jet on the densimetric Froude number of the radial jet, justifying its conservation as a similarity parameter for the transposition of the jet behaviour. A transfer function to relate the radial jet conditions to the nozzle outlet conditions is established theoretically and validated experimentally. To obtain a critical value of the densimetric Froude number in representative geometry, tests are performed on the MICAS model. Based on the experimental conditions of MICAS leading to the rise of the jet, two representative mock-ups are sized: MOJIT-Eau and MOJI/4, respectively at scales 1/2,5 and 1/4 relative to MICAS. A study of the flow under isothermal conditions on the MOJIT-Eau and MOJI/4 mock-ups shows an attachment of the radial jet to the surface of the core, due to Coanda effect. This effect depends only on the geometrical quantities of the system. A thermal-hydraulic study is finally carried out to verify the conditions of extrapolation of the recovery phenomenon. A binary behaviour is observed on the rise of the jet, which is either attached and runs along the core or detached and rises in the plenum. The Coanda effect plays a major role on the stability of the jet. The densimetric Froude number alone is not sufficient to characterize this phenomenon, which occurs on MICAS. A new dimensionless number L, representing the competition between inertia, buoyancy and Coanda effects, is then defined. This number allows for the correct transposition of the detachment and thus the rise of the jet between MICAS, MOJIT-Eau and MOJI/4. The conditions expected on ASTRID by the conservation of this number allow to study the distortion of the water-sodium similarity on the heat exchanges. Water is then verified as a relevant choice for the representativeness of small-scale mock-ups with respect to ASTRID by estimating the Péclet number (representing the heat exchange modes) of the reactor: the change of fluid corrects the change of scale on these aspects
Gutierrez, Gaëlle. „Étude du comportement thermique et sous irradiation du xénon dans l'oxycarbure de zirconium“. Phd thesis, Université Claude Bernard - Lyon I, 2011. http://tel.archives-ouvertes.fr/tel-00670024.
Cheik, Njifon Ibrahim. „Modélisation des modifications structurales, électroniques et thermodynamiques induites par les défauts ponctuels dans les oxydes mixtes à base d'actinides (U,Pu)O2“. Electronic Thesis or Diss., Aix-Marseille, 2018. http://www.theses.fr/2018AIXM0356.
(U,Pu)O2 (commonly called MOX) is currently used as nuclear fuel in pressurized water reactors with a Pu content of around 10 wt.%, and is envisaged as the reference fuel in Generation IV sodium fast reactors (SFR) with a Pu content of around 25 wt.%. Under operation, (U,Pu)O2 is submitted to fission reactions which generate a large quantity and variety of point defects, as well as fission products. By migrating, point defects and gaseous fission products can aggregate into nano-voids, dislocations and fission gas bubbles, which lead to the modification of the fuel microstructure. Therefore, a better description of the fuel behaviour at the atomic scale, and especially of the elementary mechanisms involved in the diffusion of point defects and fission products, is necessary to refine the models used in the fuel performance codes used to simulate the behaviour of fuels at the macroscopic scale. We use electronic structure calculations based on the DFT+U method combined with the occupation matrix control scheme (OMC) to investigate (U,Pu)O2 properties for various Pu contents. Static energy minimizations and ab initio molecular dynamics were used. We have first determined bulk structural, electronic and thermodynamics properties of (U,Pu)O2. We then studied the stability of point defects in (U,Pu)O2 and (U,Ce)O2, as well as the structural and electronic modifications induced by these point defects, in (U,Pu)O2 and the common experimental surrogate (U,Ce)O2. Finally, the fission gas (Kr and Xe) and helium (He) trapping and solubility in (U,Pu)O2 matrix are investigated
Chaia, Nabil. „Mise au point de revêtements protecteurs pour le gainage du combustible en alliage de vanadium V-4Cr-4Ti destiné aux RNR-Na“. Thesis, Université de Lorraine, 2013. http://www.theses.fr/2013LORR0149/document.
The use of vanadium alloy V-4Cr-4Ti as fuel cladding in the generation IV sodium cooled fast reactor (SFR) is considered with a great interest thanks to its attractive physico-chimicals properties namely: a good compatibility with liquid sodium, a high neutronic transparency,a good mechanical properties even under irradiation. However, the dissolution of oxygen in vanadium leads to its hardening. This behavior imposes, consequently, the use of on external protection as coatings that can be considered as a barrier against oxygen diffusion contained in liquid sodium at very low concentrations (a few ppm). In this work, binary and ternary diffusional silicides coatings are produced mainly by halide activated pack cementation. Their ability to protect the substrate in media simulating a SFR’s conditions, with a low oxidation potential of O2, is proved according to the results of oxidation tests in impure helium at 650 ° C and corrosion in sodium liquid at 550 ° C (CorroNa test at CEA de Saclay). Other air oxidation tests (cyclic, isothermal and creep-bending 3 points) showed good resistance of coatings at temperatures above 900°C due to the formation of a protective layer of SiO2, adherent and compact. In another part of this work, the microstructural stability of the cladding/coating system in accidental conditions is studied. This required the calculation of interdiffusion coefficients using models of multilayer growth as proposed by Wagner and mutual consumption as proposed by Buscaglia. Finally, the isothermal section at 1200 ° C and the liquidus projection of V-Cr-Si system are studied. This step, preliminary to the study of quaternary V-Cr-Ti-Si system, should allow as a perspective the optimization of the architecture of the coating and help to understand the oxidation mechanisms
Brazzale, Pietro. „Numerical and experimental study at the pilot scale of the hydrogen injection into liquid sodium by permeation through nickel membrane“. Thesis, Toulouse, INPT, 2020. http://www.theses.fr/2020INPT0101.
In the framework of the SFR (Sodium-cooled Fast Reactors), the management of tritium contamination in sodium circuits and the control of its release in atmosphere is fundamental. In order to capture and recover the tritium, it is necessary to maintain a certain amount of hydrogen dissolved in the liquid sodium stream. The hydrogen injection by permeation through nickel dense membranes has been proposed to provide a continuous hydrogen intake to liquid sodium stream, thus allowing the desired hydrogen concentration to be reached. Similar nickel-based membranes have been developed in the past for SFR (i.e. hydrogen-meters), but a lack of knowledge and applications is found for what concerns the hydrogen injection by permeation. In this study, an original permeator prototype has been designed and an experimental activity at pilot-scale has been carried out on an experimental sodium loop, under different operating conditions (temperature: 375°C-450°C; hydrogen supply partial pressures: 5 kPa-28 kPa). A dedicated measurement system, based on the gas chromatography on the retentate side, coupled to the hydrogen detection inside sodium (through a dedicated hydrogen-meter using mass spectrometry), has provided an accurate estimation of the hydrogen permeation flowrate. Tests are carried out for both a gas-vacuum and a gas-sodium configuration: in both cases, the global hydrogen permeation flowrate depends linearly on the square root of the hydrogen partial pressure in the feed side up to 20 kPa, thus demonstrating that the process in this range is limited by the hydrogen diffusion inside the nickel membrane. In particular, the presence of sodium in the permeate side does not affect significantly the whole mass transfer process. The results, compared to the permeation theoretical laws, provide an experimental permeability coefficient, specific to the prototype geometry and configuration. Comparison to values from the literature results for small nickel samples, showed that some metal-lattice phenomenon, probably linked to the membranes deformation by cold-working, could affect the hydrogen permeation in this study. In fact, slightly higher permeation coefficient with a lower activation energy is found here if compared to the literature. Finally, the experimental process has been successfully validated, thus demonstrating the feasibility of this application at the pilot-scale. An analytical 1D model has been set up with a multi-physics approach, in order to assess the radial hydrogen mass transfer in steady conditions over three physical domains, including gas, nickel and liquid sodium. It includes benchmark literature correlations for the convective mass transfer inside gas and sodium phase in tubular geometry, the Sieverts law for the H-Ni and H-Na equilibrium, coupled to the Richardson’s law for the hydrogen permeation through the nickel membrane, assumed to be diffusion-limited. CFD simulations, performed in a 2D axial-symmetric geometry with the software Comsol Multiphysics, have provided a better comprehension of the transport phenomena taking place and have confirmed the results of the straightforward 1D model under certain conditions, specific to the experimental prototype. Finally, the experimental results have shown a good agreement with the 1D model and CFD simulations in the whole temperature interval and up to a hydrogen partial pressure of 20 kPa. By resuming all the elements provided by this study, both at the experimental and numerical stage, a single equation law has been defined to describe the prototype performance and to enhance the industrial scale-up design activity
Gutierrez, Gaëlle. „Étude du comportement thermique et sous irradiation du xénon dans l’oxycarbure de zirconium“. Thesis, Lyon 1, 2011. http://www.theses.fr/2011LYO10276/document.
Refractory ceramics are considered for the GEN IV reactors (GFR). Transition metal carbides, like ZrC, are candidates as components for fuel elements owing to their good thermal stability and their neutronic performance. An extensive study was carried out to elucidate the role of temperature on the diffusion of xenon, an abundant and volatile radionuclide, in zirconium oxycarbide. For that purpose, dense zirconium carbide samples ZrC0.8O0.2 and ZrC0.95O0.05 were synthesized using Spark Plasma Sintering and Hot Pressing. 136Xe2+ ions were implanted at three fluencies: 1015, 1016 and 1017 at/cm2, at an energy of 800 keV. Thermal annealing were carried out under vacuum in a temperature range of 1500°C to 1800°C. The Xe distribution profiles were measured either by Rutherford Backscattering Spectrometry or by Secondary Ion Mass Spectrometry before and after the different treatments. Our results show that the ZrC0.8O0.2 stoichiometry is not stable at high temperature and for the ZrC0.95O0.05 stoichiometry, the Xe migration behaviour depends on the implantation fluence. The role of the implantation defects, their evolution during annealing and the trapping of Xe into bubbles was evidenced using Positron Annihilation Lifetime Spectroscopy and Transmission Electron Microscopy. In order to simulate the effects due to neutron irradiation, irradiation experiments were carried out at the JANNUS irradiation platform at CEA Saclay and the Alto Tandem accelerator at Orsay taking into account the respective roles of the ballistic and electronic processes. We observed that no xenon migration occurred after irradiation
Plevacova, Kamila. „Etude des matériaux sacrificiels absorbants et diluants pour le contrôle de la réactivité dans le cas d'un accident hypothétique de fusion du coeur de réacteurs de quatrième génération“. Phd thesis, Université d'Orléans, 2010. http://tel.archives-ouvertes.fr/tel-00592463.
Ingremeau, Jean-Jacques. „Méthodologie d’optimisation d’un coeur de réacteur à neutrons rapides, application à l’identification de solutions (combustible, coeur, système) permettant des performances accrues : étude de trois concepts de coeurs refroidis à gaz, à l’aide de l’approche FARM“. Thesis, Paris 11, 2011. http://www.theses.fr/2011PA112253/document.
In the study of any new nuclear reactor, the design of the core is an important step. However designing and optimising a reactor core is quite complex as it involves neutronics, thermal-hydraulics and fuel thermomechanics and usually design of such a system is achieved through an iterative process, involving several different disciplines. In order to solve quickly such a multi-disciplinary system, while observing the appropriate constraints, a new approach has been developed to optimise both the core performance (in-cycle Pu inventory, fuel burn-up, etc…) and the core safety characteristics (safety estimators) of a Fast Neutron Reactor. This new approach, called FARM (FAst Reactor Methodology) uses analytical models and interpolations (Metamodels) from CEA reference codes for neutronics, thermal-hydraulics and fuel behaviour, which are coupled to automatically design a core based on several optimization variables. This global core model is then linked to a genetic algorithm and used to explore and optimise new core designs with improved performance. Consideration has also been given to which parameters can be best used to define the core performance and how safety can be taken into account.This new approach has been used to optimize the design of three concepts of Gas cooled Fast Reactor (GFR). For the first one, using a SiC/SiCf-cladded carbide-fuelled helium-bonded pin, the results demonstrate that the CEA reference core obtained with the traditional iterative method was an optimal core, but among many other possibilities (that is to say on the Pareto front). The optimization also found several other cores which exhibit some improved features at the expense of other safety or performance estimators. An evolution of this concept using a “buffer”, a new technology being developed at CEA, has hence been introduced in FARM. The FARM optimisation produced several core designs using this technology, and estimated their performance. The results obtained show that this innovative feature leads to much higher performing and/or safer cores. The FARM approach has also been applied to a GFR concept using a vanadium cladding. However the large uncertainties involved do not really enable one to evaluate the performance of this promising concept.In summary, the feasibility of a global multi-disciplinary optimization has been demonstrated. Although the resulting method (FARM) is less accurate than the conventional method, it allows fast optimization and permits a large number of cores to be explored quickly, and is ideally suited for the preliminary designs studies before further refinement of the core design