Dans le cadre de ce travail, nous nous sommes intéressés à un concept innovant d’empoisonnement homogène des neutrons par insertion de terres rare (erbium et gadolinium) dans les gaines en alliage de zirconium pour les réacteurs à eau pressurisée. L’étude des équilibres entre phases des alliages erbium–zirconium et gadolinium–zirconium est indispensable comme préalable à la mise en oeuvre industrielle de ce procédé prometteur d’empoisonnement. Ce travail a consisté à déterminer expérimentalement le diagramme de phases du système erbium–zirconium. Nous avons, par le biais de différentes caractérisations, obtenu des données diagrammatiques. Avec celles-ci, nous proposons un nouveau tracé du diagramme de phases. Celui-ci est radicalement différent de celui disponible dans la littérature. Nous avons modélisé le système par l’approche CALPHAD. Nous avons également déterminé les limites de solubilité des solutions solides terminales du système gadolinium–zirconium. Les données obtenues expérimentalement sont en accord avec le tracé expérimental de la littérature et avec le modèle thermodynamique disponible. Afin de prendre en compte l’oxydation des gaines en service, nous nous sommes également intéressés aux systèmes erbium–oxygène–zirconium et gadolinium–oxygène–zirconium. Le premier système a fait l’objet d’une étude expérimentale. Nous avons mis en place un procédé de synthèse par métallurgie des poudres, incluant la synthèse de celles-ci à partir de métaux massifs. La caractérisation des échantillons ternaires nous a permis de proposer deux coupes isothermes (800°C et 1100°C). Pour le système gadolinium–oxygène–zirconium, nous avons prédit les équilibres entre phases à différentes températures à l’aide de calculs effectués à partir d’une base de données que nous avons construite avec les modèles thermodynamiques de la littérature des systèmes oxygène–zirconium, gadolinium–zirconium et sesquioxyde de gadolinium–zircone. Enfin, nous avons travaillé avec des alliages erbium–zirconium fabriqués en milieu industriel. Nous nous sommes intéressés à leurs propriétés mécaniques en traction, en lien avec leur microstructure. Nous avons mis en évidence l’effet durcissant de l’erbium, notamment à 325‰°C / This work is a contribution to the development of innovating concepts for fuel cladding in pressurized water nuclear reactors. This concept implies the insertion of rare earth (erbium and gadolinium) in the zirconium fuel cladding. The determination of the phase equilibrium in the systems is essential prior to the realisation of such a promising solution. This study consisted in the experimental determination of the erbium–zirconium phase diagram. For this, we used many different techniques to get diagram data like solubility limits and solidus, liquidus or invariant temperatures. With these data, we were able to give a new diagram, very different from the literature one. With the experimental data we collected, we also assessed the diagram, using the CALPHAD approach. In this work, we also determined the solubility limits of the gadolinium–zirconium system. Those limits had never been determined before, and the values we obtained are in excellent agreement with the experimental and with the assessed diagrams. Because these alloys are subjected to oxygen diffusion throughout their life, we focused our attention on the erbium–oxygen–zirconium and gadolinium–oxygen–zirconium system. The first system has been investigated experimentally.We used many different synthesis techniques, and we finally have opted for a powder metallurgy one. As raw material, we fabricated powder from erbium and zirconium bulk metals using hydrogen absorption/desorption. With the formed ternary pellets, we investigated the phase equilibria at 800°C and 1100°C. With the obtained data, we propose two isotherms at those two temperatures. For the gadolinium–oxygen–zirconium system, we calculated the phase equilibria at temperatures ranging from 800°C to 1100‰°C, using a homemade database compiled from literature assessments of the oxygen–zirconium, gadolinium–zirconium et gadolia-zirconia systems. We also determined the mechanical properties, in connexion with the microstructure, of industrial quality alloys in order to identify the influence of erbium content. We highlighted the hardening influence of erbium at 325‰°C

Um dos principais mecanismos de falha que causam riscos de fratura a reatores de água pressurizada é a corrosão sob tensão de ligas metálicas em água do circuito primário (CSTAP). É causada por uma combinação das tensões de tração, meio ambiente em temperatura e microestruturas metalúrgicas susceptíveis. Ela pode ocorrer, dentre outros locais, nos bocais do mecanismo de acionamento das barras de controle. Essa fratura pode causar acidentes que comprometem a segurança nuclear através do bloqueio das barras de controle e vazamentos de água do circuito primário reduzindo a confiabilidade e a vida útil do reator. O objetivo desta Tese de Doutorado é o estudo de modelos e uma proposta de modelagem para fraturas por corrosão sob tensão em liga 75Ni15Cr9Fe (liga 600), em água de circuito primário de reator de água pressurizada nesses bocais. São superpostos modelos eletroquímicos e de mecânica da fratura e validados com dados obtidos em experimentos e na literatura. Na parte experimental foram utilizados resultados obtidos pelo CDTN no equipamento recém-instalado de ensaio por taxa de deformação lenta. Na literatura está proposto um diagrama que exprime a condição termodinâmica de ocorrerem diversos modos de CSTAP na liga 600: partiu-se de diagramas de potencial x pH (diagramas de Pourbaix), para a liga 600 imersa em água primária à alta temperatura (3000C a 3500C). Sobre ele, determinaram-se os submodos de corrosão, a partir de dados experimentais. Em seguida acrescentou-se uma dimensão adicional ao diagrama, correlacionando uma variável a que se denominou fração de resistência à corrosão sob tensão. No entanto, é possível acrescentar-se outras variáveis que exprimem a cinética de iniciação e/ou crescimento de trinca, provenientes de outras modelagens de CSTAP. A contribuição original deste trabalho se insere nessa fase: partindo-se de uma condição de ensaio de potencial versus pH, foram iniciadas as modelagens de um modelo empírico-comparativo, um semi-empírico-probabilístico, um de tempo de iniciação e um de taxa de deformação, a partir dos ensaios experimentais e superpostas a essa condição. Esses exprimem respectivamente a susceptibilidade à CSTAP, o tempo de falha, e nos dois últimos o tempo de iniciação de falha por corrosão sob tensão. Os resultados foram comparados com os da literatura e se mostraram coerentes. Através desse trabalho, obteve-se uma metodologia de modelagem a partir de dados experimentais. / One of the main failure mechanisms that cause risks to pressurized water reactors is the primary water stress corrosion cracking (PWSCC) occurring in alloys. It can occurs, besides another places, at the control reactor displacement mechanism nozzles. It is caused by the joint effect of tensile stress, temperature, susceptible metallurgical microstructure and environmental conditions of the primary water. These cracks can cause accidents that reduce nuclear safety by blocking the rods displacement and may cause leakage of primary water, reducing the reactors life. In this work it is proposed a study of the existing models and a modeling proposal to primary water stress corrosion cracking in these nozzles in a nickelbased Alloy 600. It is been superposed electrochemical and fracture mechanics models, and validated using experimental and literature data. The experimental data were obtained at CDTN-Brazilian Nuclear Technology Development Center, in a recent installed slow strain rate testing equipment. In the literature it is found a diagram that indicates a thermodynamic condition for the occurrence of some PWSCC submodes in Alloy 600: it was used potential x pH diagrams (Pourbaix diagrams), for Alloy 600 in high temperature primary water (3000C till 3500C). Over it, were located the PWSCC submodes, using experimental data. It was added a third parameter called stress corrosion strength fraction. However, it is possible to superpose to this diagram, other parameters expressing PWSCC initiation or growth kinetics from other models. Here is the proposition of the original contribution of this work: from an original experimental condition of potencial versus pH, it was superposed, an empiric-comparative, a semi-empiric-probabilistic, an initiation time, and a strain rate damage models, to quantify respectively the PWSCC susceptibility, the failure time, and in the two lasts, the initiation time of stress corrosion cracking. It was modeling from our experimental data. The results were compared with the literature and it showed to be coherent. From this work was obtained a modeling methodology from experimental data.

The environmentally-assisted fatigue crack growth behaviour of austenitic stainless steels in deoxygenated high temperature water at low strain rates has been reported to be greatly affected by the sulphur (S) content of the specimen, with high S specimens exhibiting significant reduced crack growth rates (retardation) when compared to low S specimens. To further the understanding of the mechanistic behaviour, fatigue crack growth experiments have been performed on high and low sulphur Type 316 austenitic stainless steel specimens tested in high temperature water and evaluated via microstructural characterisation techniques. At high strain rates the enhanced crack growth for both specimens appeared to be crystallographic and associated with slip localization. Furthermore, matching fracture surface analysis indicated discrepancy of the slip steps and micro-cleavage cracks between the matching surfaces, suggesting that slip steps and micro-cleavage cracking occurred after the crack-tip had advanced. It was also postulated that their formation may involve cathodically-produced hydrogen and shear deformation on the fracture surface. However, when the loading frequency was decreased, the high S specimens retarded the crack growth and the path was no longer crystallographic. Significant differences in the crack-tip opening displacements were observed in both materials, with blunt crack-tips in the high sulphur specimen and sharp tips in the low sulphur specimen when the strain rate was low. EBSD analysis at the crack-tips of both specimens showed that the strain was more localised at the crack-tip of the low sulphur specimen whist the strain ahead of the high sulphur specimen was more homogenous. It is thus postulated that retardation occurs when slip localisation is no longer the dominant factor. The localised deformation during enhancement is believed to have been caused by hydrogen enhanced localised plasticity (HELP) which causes the crack-tip to sharpen. The diffused strain distribution during crack growth retardation is believed to have been caused by hydrogen enhanced creep (HEC) which causes the crack-tip to blunt. It is also believed that both enhancement and retardation mechanisms are associated with contrasting effects deriving from hydrogen enhanced plasticity. Oxide induced crack closure was excluded as a mechanism responsible for retardation of fatigue crack growth when the stress ratio is high. Effects of hydrogen induced alpha' and ε martensite phases on oxidation behaviour of austenitic stainless steels in deoxygenated high temperature water have also been studied. Microstructural characterisation shows that hydrogen induced alpha' martensite enhances oxidation of austenitic stainless steels in deoxygenated high temperature water. The implications of this finding on environmentally assisted cracking of austenitic stainless steels in deoxygenated high temperature water is discussed.

Orientador: João Manoel Losada Moreira / Dissertação (mestrado) - Universidade Federal do ABC, Programa de Pós-Graduação em Energia, 2011

The purpose of this research is to develop a methodology for a thorium fuel recycling analysis that provides results for isotopics and radio-toxicity evaluation and analysis. This research is motivated by the need to reduce the long term radiological hazard in spent nuclear fuel, which mitigates the mixing hazard (radiotoxicity and chemical toxicity) and decay heat load on the repository. The first part of the thesis presents comparison of several once-through cases with uranium and thorium fuels to show how transuranics build up as fuel is depleted. The once-through analysis is performed for the following pairs of comparison cases: low enriched uranium dioxide (UOX) vs. thorium dioxide with 233UOX (233U-ThOX), natural uranium dioxide mixed with transuranic oxides (U-TRUOX) vs. thorium dioxide mixed with transuranic oxides (Th-TRUOX), natural uranium dioxide mixed with weapons grade plutonium dioxide (U-WGPuOX) vs. thorium dioxide mixed with weapons grade plutonium dioxide (Th-WGPuOX), natural uranium dioxide mixed with reactor grade plutonium dioxide (U-RGPuOX) vs. thorium mixed with reactor grade plutonium dioxide (Th-RGPuOX). The second part of the research evaluates the thorium fuel equilibrium cycle in a pressurized water reactor (PWR) and compares several recycling cases with different partitioning schemes. Radio-toxicity results of the once-through cycle and multi-recycle calculations demonstrate advantages for thorium fuel and reprocessing with respect to long term nuclear waste management.

Axial offset anomaly (AOA) in pressurized water reactors (PWR) refers to deviation of the measured neutron flux in the top half of the core from the predicted values. Among other difficulties, AOA reduces the shutdown margin, and may force the plant to reduce power output. AOA is believed to be caused by three related phenomena occurring in the core while operating at full power: sub-cooled nucleate boiling concentrated mainly in the upper half of the core, corrosion product deposition on the cladding surface (crud), and the deposition of boron within the porous crud layer in regions of vigorous sub-cooled boiling. This study replicates the conditions within the PWR primary coolant; specifically, the temperature, pressure, peak surface heat flux, coolant velocity and water chemistry are simulated in order to produce prototypical crud on an electrically heated Zircaloy-4 test element. At the conclusion of each test run, the heated Zircaloy-4 test element is rapidly isolated from the coolant in order to trap any soluble boron species that may be present in the crud layer. The results of this investigation indicate that prototypical crud with significant boron deposition can be produced. The deposited boron compound has been determined to be lithium tetraborate (Li2B4O7). Comparative experiments have been run to determine the effect of coolant pH, concentration and type of additives, and duration of exposure on the thickness of the crud deposit. The data obtained in this investigation can be used to validate mechanistic models for crud deposition and AOA in pressurized water reactors.

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