The thesis first introduces the topic of nuclear energy and provides a brief section on plant familiarisation, after which zirconium nuclear fuel cladding is explained, and an in-depth literature review is presented on the in-service degradation of this component from hydriding. The concept of synchrotron X-ray diffraction is elucidated, and examples of its use are given, relevant to the topic of this work. The experimental section discusses an initial quantification of the Zircaloy-4 material used throughout the present work, and documents in minutia the process of collecting and analysing in-situ synchrotron X-ray diffraction data. The experimental campaign discussed within involved a series of consecutive thermal cycles designed to investigate the redistribution of hydrogen as a function of thermal and concentration gradients; the kinetics of precipitation during isothermal dwells at reactor relevant temperatures; and the evolution of strain in the matrix and hydride during these dwells. As an alternative style thesis, these three topics are separated into three independent proposed manuscripts, produced in a format ready for publication. The diffusion and redistribution paper observes localised enrichment and depletion that occurs as a function of time and temperature, investigating the flux of hydrogen that results from concentration and thermal gradients, and introduces the concept of hydrogen trapping. The second manuscript documents evidence of the rate limiting kinetics for hydride precipitation seen at elevated temperatures, and describes a model for nucleation, developed to support the experimentally produced results. The final manuscript investigates the nature of the strains that evolve in the matrix and hydride phases during precipitation and growth, highlighting slow-strain rate relaxation in both phases and examining the constraining effect that the matrix has on the hydride precipitates. Lastly, the themes from each of the three manuscripts are drawn together in a final conclusion, after which further experimental analysis that is to be performed as part of this experimental campaign is outlined.

Thesis (Nucl.E.)--Massachusetts Institute of Technology, Dept. of Nuclear Engineering, 1981. / MICROFICHE COPY AVAILABLE IN ARCHIVES AND SCIENCE. / Includes bibliographical references. / by John Edward Rivera. / Nucl.E.

Thesis (Nucl.E.)--Massachusetts Institute of Technology, Dept. of Nuclear Engineering, 1981. / MICROFICHE COPY AVAILABLE IN ARCHIVES AND SCIENCE. / Includes bibliographical references. / by El-Khider Si Ahmed. / Nucl.E.

The core of a BWR constitutes a highly complex radiational and chemical environment. Nuclear fission is utilized to generate power and as a consequence, large quantities of ionizing radiation are produced. Gamma and beta-particles along with neutrons have the sufficient ranges to escape the fuel rods and deposit energy in the reactor coolant. By doing so, radiolysis is initiated. The radiolysis species present in the coolant can interact chemically with the fuel rod cladding and cause corrosion. Different forms of corrosion are found on BWR fuel, with some effects being very local.  This thesis work outlines a method developed to investigate local corrosion phenomena, with a high spatial resolution. The purpose was to study Zircaloy corrosion and more particularly, to investigate an observed jump in corrosion thickness around the lower enrichment step on BWR fuel rod cladding. The corrosion thickness jump is a very local effect, hence the need for high spatial resolution. Monte-Carlo simulations were performed in a GEANT4-model of an inoperation reactor, to study the energy deposited from ionizing radiation in the coolant, around the low enrichment-step of the fuel rods. The energy dose data was then used as input to lwrChem to compute electrochemical potential and equilibrium concentrations of radiolysis species. These are the quantities needed to compute an equivalent corrosion rate on the cladding surface, although this was not performed within this project. The main focus was to successfully develop the two-program method using GEANT4 and lwrChem and this was achieved.  The project was performed Uppsala University with financial contribution from Vattenfall Nuclear Fuel AB and scientific contribution from Studsvik Nuclear AB.

No description available.

Les gaines de zircaloy-4, assurent la première barrière de confinement des combustibles des Réacteurs à Eau Pressurisée. Plusieurs situations accidentelles au cours desquelles les gaines de crayons combustibles sont exposées l’air à haute température ont été envisagées. L’azote généralement utilisé en tant que gaz inerte, joue un rôle primordial lorsqu’il est combiné à l’oxygène à haute température. Les courbes cinétiques obtenues par la technique de thermogravimétrie révèlent la présence de deux domaines cinétiques : le domaine pré-transitoire et le domaine post-transitoire. Durant le domaine post-transitoire, la vitesse de corrosion augmente. Les images obtenues en microscopie optique révèlent l’existence de régions corrodées caractérisées par une couche de zircone poreuse et par la présence de précipités de nitrure de zirconium (ZrN) à l’interface métal - oxyde. La corrosion des plaquettes de Zy4 à 850°C sous mélanges oxygène - azote a été étudiée durant le domaine post-transitoire. Trois réactions successives permettent d’expliquer la présence des différentes phases. Ainsi, la dégradation catastrophique du métal est due à la progression auto-catalysée par ZrN du front de croissance des zones attaquées.Les hypothèses de modélisation ont été validées durant le domaine post-transitoire. L’étape déterminante a été identifiée. La réaction d’interface externe du mécanisme d’oxydation des précipités de ZrN impose sa vitesse aux autres étapes du mécanisme de croissance des régions corrodées. Par analogie avec les modèles de germination - croissance utilisés dans le cadre de la transformation thermique des poudres, nous avons pu décrire l’évolution des zones attaquées. / Zircaloy-4 claddings provide the first containment of UO2 fuel in Pressurised Water Reactors. It has been demonstrated that the fuel assemblies cladding could be exposed to air at high temperature in several accidental situations. When mixed to oxygen at high temperature, the nitrogen, usually used as an inert gas, causes the accelerated corrosion of the cladding.The kinetic curves obtained by thermogravimetry reveal two stages: a pre-transition and a post-transition one. In the post-transition stage, the kinetic rate increases with time. Images obtained by optical microscopy of a sample in the post-transition stage reveal the presence of corroded zones characterized by a porous scale with zirconium nitride precipitates at metal - oxide interface. Corrosion of Zy4 plates at 850°C under mixed oxygen - nitrogen atmospheres has been studied during the post-transition stage. A sequence of three reactions is proposed to explain the mechanism of nitrogen-enhanced corrosion. The accelerating effect of nitrogen in the corrosion scale can therefore be described on the basis of an autocatalytic effect of the zirconium nitride precipitates. Then, it is demonstrated that the steady-state approximation as well as the existence of an elementary step controlling the growth process are valid during the post-transition stage. The rate-determining step is identified as the external interface reaction step of the oxidation of the zirconium nitride precipitates. Finally, a nucleation and growth model used for thermal reactions in powders is used to describe both the nucleation and the growth of the attacked regions.

La corrosion sous contrainte par l’iode (CSC-I) est l’un des mécanismes de rupture potentiels des crayons combustibles en alliage de zirconium, pouvant intervenir au cours des transitoires de puissance des réacteurs nucléaires. La fissuration par CSC-I comporte trois étapes : amorçage de la fissure, développement intergranulaire puis propagation transgranulaire. Le but du travail est d’identifier des paramètres critiques gouvernant les transitions entre ces différentes étapes. En premier lieu, des essais sur des éprouvettes en Zircaloy présentant des finitions de surface et des états métallurgiques variés permettent de discriminer l’influence de différents paramètres sur l’amorçage des fissures. Nous mettons en évidence le rôle critique du niveau des contraintes résiduelles, de leur répartition en surface ainsi que de leur profil au sein du matériau. La sensibilité des alliages à l’amorçage des fissures n’est pas directement corrélée à la rugosité de la surface. Cependant, la dispersion des paramètres de rugosité traduit l’irrégularité du profil, l’hétérogénéité du niveau des contraintes résiduelles, et donc l’existence de zones où les contraintes résiduelles sont localement moins protectrices. Dans un second temps, des éprouvettes de Zircaloy-4 possédant différents états d’écrouissage sont sollicitées sous charge constante, en présence de méthanol iodé. Les modifications microstructurales induites par l’écrouissage favorisent l’apparition de la propagation transgranulaire des fissures de CSC-I. Des observations des faciès de rupture en MET révèlent que la transition inter/transgranulaire intervient dans des zones où les grains sont fortement désorientés les uns par rapport aux autres, suite à l’augmentation des contraintes locales résultant des incompatibilités de déformation grain à grain. / Iodine-induced stress corrosion cracking (I-SCC) is one of the potential failure modes of zirconium alloy fuel claddings during power transients in nuclear reactors. I-SCC failures are usually described in three steps: initiation of cracks, intergranular development and transgranular propagation. The objective of this work is to identify critical parameters controlling transitions between crack propagation modes. First of all, experiments conducted on Zircaloy samples with various surface conditions and metallurgical states lead to discriminate the influence of several parameters responsible for cracks initiation. The critical role of residual stresses level, their distribution at the subsurface and their evolution in the bulk of the material is evidenced. Sensitivity to I-SSC is not directly correlated to surface roughness. However, dispersion in roughness parameters indicates the presence of surface irregularities, heterogeneities of residual stresses and the existence of surface areas where residual stresses are less protective. In a second step, Zircaloy-4 samples with various strain-hardening pre-treatments are submitted to constant load tests in an iodine methanol solution. Microstructural modifications induced by a strain-hardening pre-treatment enhance transgranular propagation of I-SCC cracks. TEM observations of fracture surfaces show that the intergranular to transgranular crack transition takes place preferentially where the relative crystallographic orientation is large between two adjacent grains, because of local stress concentrations resulting from strain incompatibilities between neighbouring grains.

Dans le cadre des études visant à garantir l'intégrité de la gaine du crayon combustible, EDF est amené à caractériser la ductilité de la gaine après un Accident de Perte de Réfrigérant Primaire (APRP). La thèse porte sur la caractérisation du comportement à rupture des gaines en Zircaloy-4 détendu pour lesquels les conditions d'APRP ont été simulées en laboratoire par une oxydation à haute température suivie d'un refroidissement. L'oxydation est effectuée à 1100°C et à 1200°C pour différentes durées ce qui conduit à des niveaux d'oxydation de 3% à 30% d'ECR (Equivalent Cladding Reacted). Deux types de refroidissement sont mis en oeuvre : la trempe à l'eau et le refroidissement à l'air. Les gaines oxydées comportent deux couches fragiles, la couche de zircone externe ZrO2 et la couche α(O), et une couche présentant une ductilité résiduelle, la couche ex-β.Les gaines oxydées ont fait l'objet de caractérisations en microscope optique, par analyse à la microsonde et par nano-indentation. Une corrélation entre la teneur en oxygène et la nano-dureté et le module d'Young a été proposée.L'essai Expansion due à la Compression (EDC) a été développé avec une instrumentation par stéréo-corrélation d'images puis a été utilisé pour caractériser le comportement mécanique des gaines oxydées. Le comportement des gaines oxydées a été étudié à partir de l'analyse des courbes macroscopiques de l'essai EDC et à partir des observations des échantillons rompus ou pré-déformés.Un scénario de rupture des gaines oxydées a été proposé. Ce scénario a été validé d'une part par la réalisation d'essais sur gaines sablées ne comportant que la couche ex-β et d'autre part par la modélisation de l'essai par la méthode des éléments finis. Un critère de rupture des gaines oxydées a par ailleurs été établi. La modélisation du comportement et le critère de rupture proposés ont été validés par la modélisation des essais de compression d'anneau. / In order to guarantee the integrity of nuclear fuel rod cladding, it is necessary for EDF to characterize the ductility of cladding after a Loss of Coolant Accident (LOCA). The thesis is about the characterization of the fracture behavior of cold-worked stress-relieved Zircaloy-4 claddings which have undergone LOCA conditions simulated in laboratory by a high temperature oxidation followed by a cooling. The high temperature oxidation is carried out at 1100°C and 1200°C with different times, which leads to different oxidation levels varying from 3% to 30% ECR (Equivalent Cladding Reacted). The high temperature oxidation is followed by two types of cooling: water quench and air cooling. The oxidized claddings contain two fragile layers - the outer zirconium oxide ZrO2 layer and the middle α(O) layer, and a layer which can have residual ductility - the inner ex-β layer.Characterizations by means of optical microscopy, electron probe micro analysis and nano-indentation have been carried out on the oxidized claddings. A correlation between the oxygen concentration and the nano-hardness and the Young's modulus has been proposed.The Expansion Due to Compression (EDC) test has been developed with an instrumentation of stereo digital image correlation, and then used to characterize the mechanical behavior of the oxidized claddings. The behavior of the oxidized claddings has been studied via macroscopic EDC test curves and observations of fractured or pre-deformed test samples.A fracture scenario of the oxidized claddings has been proposed. The fracture scenario has then been validated via EDC tests on oxidized claddings whose ZrO2 and α(O) layers have been removed, and via finite element modeling of EDC tests. Moreover, a fracture criterion has been established. The mechanical behavior modeling and the proposed fracture criterion have been validated by modeling of ring compression test.

No description available.

Zirconium alloys (zircaloy) have been widely used in light water reactors due to their good thermomechanical properties, corrosion resistance, and low thermal neutron absorption rate. As one of the most important safety barriers, cladding is not only used to encapsulate nuclear fuel, but also to prevent the nuclear fission products from leaking into the coolant. During the operation of nuclear reactors, hydride will form in zircaloy and significantly degrade the tensile strength, ductility, fracture toughness, and creep behavior of the cladding, and eventually leading to the failure of cladding. Therefore, understanding the material properties of zircaloy and its hydrides is crucial to the safety of power plants. In this study, the mechanical Raman spectroscopy and nano-mechancial testing techniques were used to perform thermomechanical measurements and damage analysis of zircaloy-4. The Raman thermometry method was used to measure localized spatially resolved thermal conductivity and establish the potential linkage of microstructure to thermal and mechanical properties of zircaloy-4. The local thermal conductivity values showed to increase with increase in grain size. Nanoindentation and nano-scale impact techniques were used to obtain the viscoplastic constitutive relation of hydrides at elevated temperatures. Based on the obtained viscoplastic model, fracture strength of hydrides was predicted by using finite element method (FEM) simulations. An extended Gurson-Tvergaard-Needleman (GTN) model was used to study the macro-scale fracture behavior of hydrided zircaloy-4 structures. Good agreement between calculated and experimental results was obtained for various boundary conditions.