En rampe de puissance, le combustible nucléaire est soumis à d'importantes contraintes thermiques et mécaniques, et subit une modification profonde de son environnement chimique. Le combustible contraint fortement la gaine, notamment au niveau des zones inter-pastilles, ce qui, associé au relâchement de produits de fission corrosifs, peut conduire à sa rupture par corrosion sous contraintes. Les évolutions simultanées de la mécanique, de la thermique et de la chimie du combustible sont liées, et participent au bon ou mauvais comportement de l'UO2 en rampe de puissance. L'objectif de ce travail est de modéliser à l'échelle d'une pastille de combustible, l'évolution couplée de la chimie, de la thermique et de la mécanique, et de préciser l'impact de ces couplages sur le comportement de l'UO2 en rampe de puissance. La finalité est d'évaluer un terme source en relâchement d'iode pour alimenter les modèles de corrosion sous contraintes dédiés aux études d'Interaction Pastille-Gaine. / Nuclear fuels under power transient undergo high thermal and mechanical stresses, as well as deep chemical modifications. Stresses on the cladding at the inter-pellet plane due to the pellet thermal expansion, associated to the corrosive fission product release, can lead to clad failures, resulting from a stress corrosion cracking mechanism. The thermal, mechanical and chemical properties of the UO2 irradiated fuel are closely dependent and play a major role on the behavior of the material during a power transient. The aim of this work is to model at the pellet scale the chemical, thermal and mechanical coupled changes of the UO2 fuel during a power transient scenario and to evaluate the consequences on the fuel behavior. The final objective is to obtain an evaluation of the iodine release source term to be used in I-SCC modelling codes dedicated to Pellet-Clad-Interaction studies.

L’identification des mécanismes de relâchement des Produits de Fission (PF) hors de combustible nucléaire irradié lors d’un accident grave est primordiale pour le développement de codes capables d’estimer précisément le terme source (nature et quantité des radionucléides émis dans l’environnement). Parmi les différents PF, les Ba, Cs, Mo et le Ru sont particulièrement intéressants, car ils peuvent interagir entre eux ou avec d’autres éléments et donc affecter leur relâchement. Dans le cadre de cette thèse, deux axes de travail ont été mis en place avec l’objectif d’identifier les phases chimiques présentes avant l’accident et leur évolution au cours de l’accident lui-même. L’approche expérimentale a consisté à reproduire les conditions d’un accident nucléaire à l’échelle du laboratoire, en utilisant des échantillons de combustibles irradiés et des matériaux modèles (UO₂ vierge dopés en 12 PF). Le principal avantage de ces derniers est l’utilisation de méthodes de spéciation chimique comme la Spectroscopie d’Absorption des rayons X, qui n’est pas aujourd’hui encore disponible pour les combustibles irradiés. Trois échantillons de combustible irradié ont été étudies, représentatifs de l’état initial (i.e. avant l’accident), d’une étape intermédiaire en température (1773K) et d’un état avancé d’accident nucléaire (2873K). Pour les matériaux modèles, plusieurs séquences accidentelles (de 573K à 1973K) ont été réalisés. Les résultats expérimentaux ont permis d’établir un nouveau mécanisme de relâchement des PF en en fonction des conditions oxydantes et réductrices du scénario accidentel. Ces résultats ont démontré aussi l’importance des matériaux modèles pour l’étude des accidents nucléaires graves, en complémentarité aux combustibles irradiés. / 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.

Lors de l’irradiation en réacteur, des gaz de fission tels que le xénon et le krypton sont produits. Ces gaz diffusent dans le combustible, mais peuvent également précipiter sous forme de bulles. En outre,les réactions de fission conduisent à la formation de défauts ponctuels (lacunes ou interstitiels) et sous forme d’amas (dislocations ou cavités). L’obtention de données expérimentales sur la migration des gaz de fission en présence de défauts est nécessaire afin d’améliorer la compréhension et la modélisation du comportement du combustible sous irradiation. La démarche mise en place dans ce travail a pour objectif d’étudier la diffusion thermique des gaz et de comprendre leur interaction avec les défauts d’irradiation. Elle repose sur la réalisation d’études à effets séparés couplant des irradiations/implantations aux ions à des techniques de caractérisation fines. La Spectroscopie d’Annihilation des Positons (SAP) complétée par la Microscopie Electronique en Transmission (MET)permet de caractériser les défauts (ponctuels et/ou sous forme d’amas) générés par l’irradiation et de suivre leur évolution en température. En parallèle, la modélisation des cinétiques de relâchement des gaz rares mesurées par désorption thermique couplée à la spectrométrie de masse, permet d’obtenir les coefficients de diffusion des gaz et de mettre en lumière les phénomènes de piégeage opérants. La synthèse de ces résultats expérimentaux nous amène à identifier les mécanismes de migration des gaz et à décrire leurs interactions avec les défauts d’irradiation. / During in-reactor irradiation, fission gases such as xenon or krypton are produced. In the fuel, those gases diffuse and precipitate to form bubbles. In addition, fission reactions induce small defects(vacancies and interstitials) and larger defects (cavities and dislocations) formation. Data acquire menton fission gases migration considering radiation-induced defects is thus necessary to better understand and improve models of in-pile fuel behavior. The experimental approach developed in this work aims to study thermal diffusion of rare gases and to understand their interaction with radiation-induced defects.To do this, separated effect studies were performed coupling ion implantations/irradiations to fine characterization techniques. Positron Annihilation Spectroscopy (PAS) coupled to Transmission Electron Microscopy (TEM) observations allows for defects characterizations (vacancies and/or cavities induced by ion implantation) and for their thermal behavior study. On the other hand, gas release measurements are performed by thermal desorption spectrometry. Simulation of gas kinetic release allows to determine diffusion coefficients and to lighten trapping mechanisms. The synthesis of those various experimental results brings us to identify gas migration mechanism and to describe their interaction with radiation-induced defects.

Le Commissariat à l'Energie Atomique et aux Energies Alternatives (CEA) a en charge la conception d'un réacteur à neutrons rapides de quatrième génération.L'instrumentation neutronique de ce futur réacteur s'appuiera sur des chambres à fission placées en cuve. Ces chambres à fission à haute température (CFHT) devront fonctionner à pleine puissance à une température comprise entre 400°C et 650°C.Un bilan récent de la technologie CFHT a révélé que certains points sont à améliorer afin d'en garantir une plus grande fiabilité.En particulier, on recherche une meilleure compréhension du phénomène de décharges partielles. Celles-ci engendrent des impulsions non discernables de celles produites par les fragments de fission du dépôt fissile.Par ailleurs, elles pourraient accélérer le vieillissement des isolants minéraux.En s'appuyant sur une démarche expérimentale et théorique, ce travail de thèse a apporté plusieurs résultats.Les tests sur les différentes chambres à fission ont permis de caractériser les signaux de DP, vis à vis des signaux neutroniques et de trouver une méthode efficace de discrimination DP-neutron. De la même manière, les signaux DP ont été localisés et une solution technologique a été proposée et mise en oeuvre avec succès pour les éliminer.Un outil de calcul pour la simulation des impulsions neutroniques a été conçu et testé avec succès.Une expérience sur l'effet de la température sur la courbe de Paschen, dans un volume de gaz fermé, a été conçue et réalisée en donnant les premiers résultats intéressants. / The Commission for Atomic and Alternative Energy (CEA) is in charge of the fourth generation fast neutron reactor design. The instrumentation for neutron flux measurement of this future reactor will be based on fission chambers placed in-core. These high temperature fission chambers (HTFC) will have to operate at full reactor power, and thus at a temperature between 400°C and 650°C.A recent review of HTFC technology has revealed that some points need improvement to ensure greater reliability.In particular, a better understanding of the phenomenon of partial discharges (PD), which are observed in the fission chambers at high temperature, is needed. These PD pulses are indistinguishable from those produced by the products of fission caused by collision with neutrons with the fissile deposit within thechambers.In addition, they could accelerate aging of the ceramic insulators used in the chambers.Based on both experimental and theoretical approaches, this PhD work found several results.Tests on different fission chambers made it possible to characterize the DP signals vis-a-vis the neutron signals and to find an operational DP-neutron discrimination method. The DP signals were localized and a technological solution was proposed and successfully implemented to eliminate them.A calculation tool for neutron pulse simulation was also designed and tested successfully.An experiment on the effect of temperature on the Paschen curve, in a closed gas volume, was designed and carried out giving initial interesting results.

(U,Pu)O2 (aussi appelé MOX) est actuellement utilisé comme combustible dans les réacteurs nucléaires à eau pressurisée (REP) avec une teneur massique en Pu d’environ 10 %. Il est également envisagé comme combustible de référence pour les réacteurs à neutrons rapides à caloporteur sodium, avec une teneur massique en Pu d’environ 25 %. En conditions opérationnelles, (U,Pu)O2 est soumis à des réactions de fission qui génèrent une grande quantité de défauts et de produits de fission. Par migration, ces défauts et produits de fission gazeux peuvent s'agréger en nano-cavités, dislocations et bulles de gaz, conduisant à une modification de la microstructure. Une meilleure description du comportement du combustible à l’échelle atomique, notamment des mécanismes élémentaires impliqués dans la diffusion des défauts et des produits de fission, est donc nécessaire pour affiner les modèles utilisés dans les codes de performance des combustibles. Pour l’étude des propriétés de (U,Pu)O2, nous avons effectué des calculs de structure électronique basés sur la méthode DFT+U combinée au contrôle des matrices d’occupation des orbitales corrélées. Des minimisations d’énergie ainsi que la dynamique moléculaire ab initio ont été utilisées. Nous avons étudié dans un premier temps les propriétés du cristal de (U,Pu)O2 pour différentes teneurs en Pu. Nous avons ensuite étudié la stabilité des défauts ponctuels ainsi que les modifications structurales et électroniques induites par ces défauts ponctuels dans (U,Pu)O2 et (U,Ce)O2, matériau utilisé comme simulant de (U,Pu)O2. Enfin, nous avons étudié le piégeage et la solubilité des gaz de fission (Kr, Xe) et de l’hélium dans la matrice de (U,Pu)O2 / (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

<p> Part A Abstract </p> <p> Temperature coefficients of reactivity for an 37-element reference design of a thorium oxide fuelled, heavy water moderated and cooled reactor, are calculated. The physical processes which determine magnitude and sign of the coefficients are identified and discussed. Results are given for fresh fuel containing equilibrium concentrations of the fission product Xe-135 and with boron control in the moderator. Results are. also -given for fresh fuel with the equilibrium concentration of Xe-135 but without boron contorl for fuel with an exposue of 1.513 n/k barn and for fuel with an exposure of 3.13 n/k barn; each containing appropriate concentrations of 50 separate nuclides and one-pseudo fission product. The fuel temperature coefficient of reactivity is negative for all the cases studied, while the coolant temperature coefficient of reactivity is positive for all the cases studied. The void effect is an increase in reactivity for all cases studied. </p> ////////////////////// <p> Part B Abstract </p> <p> This work has been done with the purpose of studying the validity of Bragg Kleeman rule which states that for combinations of elements, the atomic stopping cross-sections are additive. The validity of Bragg Kleeman rule for low energy He ions has not been conclusively tested for solids. In this work, the comparison with the experimental stopping power of SiC with the additive stopping powers of Si and C has been made experimentally. </p> <p> A thick target technique in the experimental evaluation of the stopping powers is used. This method has some simplicity over the thin target techniques. </p> <p> A calibration of the McMaster University Van-de Graff accelerator was done. Experiments were conducted later using the calibration curves produced. </p> <p> The report contains a brief account on different sources of errors due to the Van-de-Graff accelerator calibration and due to stopping power experiments. </p> / Thesis / Doctor of Philosophy (PhD)

In collaboration with Westinghouse Electric AB this project aims to study the possibilities with using machine learning methods to predict fuel failure in a Boiling Water Reactors (BWRs). The main objective has been to create a dataset consisting of both empirical measurements and simulated samples from a physics model and evaluate different machine learning algorithms, that use these datasets to predict fuel defects. The simulated data is created using a physics model derived from the ANS-5.4 standard which allows for good control over specific parameter values. Three machine learning algorithms were deemed fit for this type of problem and used throughout the project: Random Forest (RF), K-Nearest Neighbor (KNN) and Neural Network (NN). Both classification and regression type problems have been assessed. All three methods showed good results for the classification problems, where the goal was to predict if there was a fuel failure or not. All models reached an accuracy above 97% and performed well, the RF model had the highest overall, with an accuracy of 98.2 %. However, the NN method made the fewest false negative predictions and can therefore be seen as the best model for this purpose. For the regression, problems with the aim of predicting escape rates, both the RF and KNN had similar promising results with very small errors overall. Yet, there is a slight increase in errors when predicting higher escape rates for both models. This is most likely due to the available data being of mostly low escape rates. The NN did not perform well with this problem, the predictions having large error for both low and high escape rates, a possible explanation is the lack of data. To improve the results, and create even better models, the empirical measurements need to contain more information such as defect location and fuel failure size, also an increase in the number of samples taken at fuel failure operation would be valuable.

Safety and reliability are one of the key objectives for future Generation IV nuclear energy systems. The neutron flux monitoring system forms an integral part of the safety design of a nuclear reactor and must be able to detect any irregularities during all states of reactor operation. The work in this thesis mainly concerns the detection of in-core perturbations arising from unwanted movements of control rods with in-vessel neutron detectors in a sodium-cooled fast reactor. Feasibility study of self-powered neutron detectors (SPNDs) with platinum emitters as in-core power profile monitors for SFRs at full power is performed. The study shows that an SPND with a platinum emitter generates a prompt current signal induced by neutrons and gammas of the order of 600 nA/m, which is large enough to be measurable. Therefore, it is possible for the SPND to follow local power fluctuations at full power operation. Ex-core and in-core detector locations are investigated with two types of detectors, fission chambers and self-powered neutron detectors (SPNDs) respectively, to study the possibility of detection of the spatial changes in the power profile during two different transient conditions, i.e. inadvertent withdrawal of control rods (IRW) and one stuck rod during reactor shutdown (OSR). It is shown that it is possible to detect the two simulated transients with this set of ex-core and in-core detectors before any melting of the fuel takes place. The detector signal can tolerate a noise level up to 5% during an IRW and up to 1% during an OSR.

Approved for public release, distribution is unlimited / Thermophotovoltaic energy conversion offers a means of efficiently converting heat into electrical power. This has potential benefits for space nuclear reactor power systems currently in development. The primary obstacle to space operation of thermophotovoltaic devices appears to be the low heat rejection temperatures which necessitate large radiator areas. A study of the tradespace between efficiency and radiator size indicates that feasible multi-junction TPV efficiencies result in substantial overall system mass reduction with manageable radiator area. The appendices introduce the endothermodynamic model of a TPV cell and briefly assess the utility of advanced carbon-carbon heat pipe radiator concepts. / Lieutenant, United States Navy

O 131I é um radioisótopo de iodo amplamente utilizado em medicina nuclear, pode ser utilizado tanto para diagnóstico quanto para tratamento devido às suas características físicas de decaimento - e sua elevada emissão de raios-y. Sua produção no IPEN é realizada utilizando um reator nuclear a partir da reação indireta: 130Te (n,y) 131mTe 131Te 131I, onde são irradiados alvos contendo Te. Pode também ser produzido via produto de fissão de 235U, onde, o 235U irradiado produz cerca de 300 elementos diferentes, entre eles o 131I. O 131I produzido nesse método apresenta altas atividade específica e concentração radioativa, o que facilita a produção de compostos marcados com o radionuclídeo. O objetivo deste trabalho é desenvolver um método de recuperação de 131I no processo de produção de 99Mo pela rota de dissolução ácida de alvos de 235U, com a qualidade necessária para ser utilizado em Medicina Nuclear. O 131I encontra-se em 2 fases no processo, tanto na fase gasosa produzida na dissolução ácida dos alvos de U metálico e a menor parte em solução. Foram utilizados diversos materiais para captura e recuperação de 131I nas 2 fases do processo, a fase gasosa e a solução de dissolução dos alvos de U. Foram testadas colunas de alumina com Cu, alumina ácida com Cu, nanoesferas de Ag, cartuchos aniônicos, resina aniônica, colunas de carvão ativado, microesferas de Ag e microesferas de Cu. Soluções contendo 131I em NaOH 0,1 mol.L-1 foram percoladas pelos materiais e os eluídos foram analisados em calibrador de dose. Foi também estudada a precipitação de AgI e dissolução desse precipitado em NH4OH 0,1 mol L-1 e Na2S2O3 5%. Dentre os testes realizados, a princípio, os resultados de recuperação variaram de acordo com o material, o carvão ativado apresentou rendimento de recuperação entre 42% a 83%. Já o rendimento de recuperação da coluna de alumina com Cu variou de 20% a 85%. Os testes com nanoesferas de Ag apresentaram rendimento de recuperação de 26% utilizando NaOH 0,1 mol L-1 e 72% utilizando Na2S2O3 como eluentes. Testes com cartuchos aniônicos apresentaram os melhores resultados com uma porcentagem de recuperação de 81 a 90%. Testes utilizando 131I na sua forma gasosa apresentaram uma retenção de 66,45% e não foram realizados testes para recuperação do 131I retido. Nos testes utilizando precipitação de AgI a porcentagem de retenção de 131I foi de 100%. É possível concluir que os cartuchos aniônicos e a precipitação de AgI foram as melhores opções para a retenção de 131I, e as colunas de alumina com Cu tem um grande potencial para eluição do radionuclídeo 131I na forma química adequada. / 131I is an iodine radioisotope widely used in nuclear medicine that can be used either for diagnostic or for treatment due to its physical decay by - and its high emission of -rays. It is produced at IPEN using the indirect reaction: 130Te(n,)131mTe 131Te 131I where TeO2 targets are irradiated in a Nuclear Reactor. There is also the possibility of producing 131I by the fission of 235U, where about 300 different elements are produced together with 131I. The 131I produced through this method presents high specific activity and radioactive concentration suitable for the labeling of molecules. The aim of this work was to develop a recovery method of 131I with the required quality to be used in Nuclear Medicine in the 99Mo production process through the route of acid dissolution of metallic 235U targets. 131I can appear in two phases of the process, both in the gaseous phase produced during the dissolution of metallic U targets and in the dissolution solution. This work studied the recovery of 131I in these two phases. Several materials were used for the capture and recovery of 131I at the two phases of the process, the gaseous one and the solution of dissolution of U targets. Columns of alumina with Cu, acid alumina with Cu, Ag microspheres, Cu microspheres, Ag nanospheres, anionic cartridges, Ag cartridges, anion exchange resin and activated charcoal columns were tested. Solutions containing 131I in 0.1 mol.L-1 NaOH were percolated through the materials and the eluted solutions were analyzed in a dose calibrator. The precipitation of AgI was also studied wth further dissolution of this precipitate with 0.1 mol L-1 NH4OH and 5% Na2S2O3. The recovery results varied according to the material, activated charcoal showed recovery yields between 42% and 83% but the recovery yield of the alumina column with Cu ranged from 20% to 85%. Tests with Ag nanospheres showed recovery yield of 26% using 0.1 mol L-1NaOH and 72% for Na2S2O3. Tests with anionic cartridges showed the best results with a recovery percentage ranging between 81 to 90%. Tests using 131I in the gaseous phase presented retention of 66.45% and its elution was not studied. The experiments with the AgI precipitation showed total retention of 131I. It can be concluded that the anionic cartridges and the precipitation of AgI have higher affinity for the retention of 131I, and alumina columns with Cu have great potential for its elution in a suitable chemical form.