Des méthodes électrochimiques classiques, et des caractérisations photoélectrochimiques (PEC), utiliséesex-situ et in-situ, ont permis d’étudier le phénomène de Shadow Corrosion, considéré ici comme une corrosion galvanique entre des alliages de zirconium et de nickel, corrosion influencée par l’environnement chimique et l’irradiation de ces alliages. Une cellule électrochimique simulant les conditions d’un réacteur à eau bouillante (REB), permettant l’illumination UV--Visible des échantillons et le contrôle de la chimie de l’eau, a été conçue, développée et validée. Cette cellule a permis de mesurer pour la première fois des spectres en énergie de photocourant d’un alliage de zirconium, in-situ en milieu REB simulé. Par ailleurs, les résultats expérimentaux obtenus tendent à montrer que les impuretés de type cations métalliques jouent un rôle important dans le mécanisme d’activation du couplage galvanique, donc potentiellement dans le mécanisme d’activation du phénomène de Shadow Corrosion, alors que la présence d’oxygène et/ou de peroxyde d’hydrogène n’induit pas de différences significatives du comportement électrochimique des échantillons. Il est montré également que l’illumination UV--Visible des échantillons, qui amplifie notablement les courants de couplage, est un paramètre important du phénomène de Shadow Corrosion. / Conventional electrochemical methods as well as photoelectrochemical characterisations (PEC), performedex-situ et in-situ, were used to study the Shadow corrosion phenomenon, considered as a galvanic corrosion between Zr-based and Ni-based alloys. The Shadow corrosion is influenced by the chemical environment and the irradiation of these alloys. An electrochemical cell , simulating the conditions of a boiling water reactor (BWR), allowing the illumination of the samples with UV--Visible as well as monitoring the water chemistry was designed, developed and validated. The cell allowed, for the first time, recording of emph{in-situ} photocurrent energy spectra on a Zr-based alloy in simulated BWR environment. Furthermore, the obtained experimental results pointed out that the metallic cation impurities played an important role in the activation mechanism of the galvanic coupling, thus potentially in the activation mechanism of the Shadow corrosion phenomenon, whereas the presence oxygen and/or hydrogen peroxide did not induce significant differences in terms of electrochemical behavior of the samples. It was also shown that the illumination of the sample with UV--visible light, which significantly amplified the galvanic current, is an important parameter of the Shadow corrosion phenomenon.

Reatores PWR empregam, como combustível nuclear, pastilhas de UO2 acondicionadas em tubos de ligas de zircônio, chamados de encamisamento. Na sua fabricação são gerados cavacos de usinagem que não podem ser descartados, pois a reciclagem deste material é estratégica quanto aos aspectos de tecnologia nuclear, econômicos e ambientais. As ligas nucleares têm altíssimo custo e não são produzidas no Brasil, sendo importadas para a fabricação do combustível nuclear. Neste trabalho são abordados dois métodos para reciclar os cavacos de Zircaloy. No primeiro, os cavacos foram fundidos utilizando um forno elétrico a arco para obter lingotes. O segundo usa a técnica da metalurgia do pó, onde os cavacos foram submetidos à hidretação e o pó resultante foi moído e isostaticamente prensado e, a seguir, sinterizado a vácuo. A composição química, as fases presentes e a dureza no material foram determinadas. Os lingotes foram tratados termicamente e laminados, sendo que as microestruturas foram caracterizadas por microscopia óptica e eletrônica de varredura. Os resultados para ambos os métodos mostraram que a composição do Zircaloy reciclado cumpre as especificações químicas e apresentaram microestrutura adequada para uso nuclear. Os bons resultados do método de metalurgia do pó sugerem a possibilidade de produzir pequenas peças, como as tampas do encamisamento - end-caps, usando a sinterização no formato quase final (near net shape). / PWR reactors employ, as nuclear fuel, UO2 pellets with Zircaloy clad. In the fabrication of fuel element parts, machining chips from the alloys are generated. As the Zircaloy chips cannot be discarded as ordinary metallic waste, the recycling of this material is important for the Brazilian Nuclear Policy, which targets the reprocess of Zircaloy residues for economic and environmental aspects. This work presents two methods developed in order to recycle Zircaloy chips. In one of the methods, Zircaloy machining chips were refused using an electric-arc furnace to obtain small laboratory ingots. The second one uses powder metallurgy techniques, where the chips were submitted to hydriding process and the resulting material was milled, isostatically pressed and vacuum sintered. The ingots were heat-treated by vacuum annealing. The microstructures resulting from both processing methods were characterized using optical and scanning electron microscopies. Chemical composition, crystal phases and hardness were also determined. The results showed that the composition of recycled Zircaloy comply with the chemical specifications and presented adequate microstructure for nuclear use. The good results of the powder metallurgy method suggest the possibility of producing small parts, like cladding end-caps, using near net shape sintering.

Ce travail vise à étudier la rupture du gainage avec des essais mécaniques plus représentatifs des conditions RIA, en prenant en compte les blisters d'hydrures ainsi que le niveau élevé de biaxialité du chargement mécanique et des vitesses de déformation. Nous avons formé par thermodiffusion en laboratoire des blisters similaires à ceux observés sur des gaines de Zircaloy-4 irradiées en réacteur. Les caractérisations par métallographie, nanodureté, DRX et ERDA ont montré qu'un blister est constitué d'hydrures delta dont la concentration dans la matrice varie entre 80% et 100%, et que la matrice sous-jacente contient des hydrures radiaux. Nous avons modélisé la cinétique de croissance des blisters avec l'hystérésis de la limite de solubilité de l'hydrogène,puis défini le gradient thermique seuil permettant leur formation. Notre étude sur le comportement dilatométrique du zirconium hydruré montre le rôle important de la texture cristallographique du matériau, ce qui peut expliquer des différences de morphologie des blisters. En parallèle, des essais suivis par caméra infrarouge ont montré que des vitesses de déformation supérieures à 0,1/s induisent des échauffements locaux importants qui favorisent la localisation précoce de la déformation. Enfin, nous avons optimisé l'essai d'Expansion Due to Compression pour atteindre un niveau de biaxialité de déformation plane (essais HB-EDC et VHB-EDC), ce qui réduit fortement la déformation à rupture à 25°C et 350°C, mais seulement en l'absence de blisters. Un critère de rupture est proposé pour rendre compte de la baisse de ductilité des gaines en Zircaloy-4 non irradiées en présence de blisters. / Our aim is to study the cladding fracture with mechanical tests more representative of RIA conditions, taking into account the hydrides blisters, representative strain rates and stress states. To obtain hydride blisters, we developed a thermodiffusion setup that reproduces blister growth in reactor conditions. By metallography, nanohardness, XRD and ERDA, we showed that they are constituted by 80% to 100% of delta hydrides in a Zircaloy-4 matrix, and that the zirconium beneath has some radially oriented hydrides. We modeled the blister growth kinetic taking into account the hysteresis of the hydrogen solubility limit and defined the thermal gradient threshold for blister growth. The modeling of the dilatometric behavior of hydrided zirconium indicates the important role of the material crystallographic texture, which could explain differences in the blister shape. Mechanical tests monitored with an infrared camera showed that significant local heating occurred at strain rates higher than 0.1/s. In parallel, the Expansion Due to Compression test was optimized to increase the biaxiality level from uniaxial stress to plane strain (HB-EDC and VHB-EDC tests). This increase in loading biaxiality lowers greatly the fracture strain at 25°C and 350°C only in homogeneous material without blister. Eventually, a fracture criterion of unirradiated Zircaloy-4 cladding tube taking into account the blister depth is proposed.

La texture marquée des tôles et des gaines en alliages de zirconium se traduit par une forte anisotropie du comportement mécanique. L'objectif de l'étude est de caractériser et de modéliser le comportement anisotrope de tôles en alliage de Zircaloy-4 recristallisé. La caractérisation de l'anisotropie du comportement est réalisée au travers d'essais mécaniques conventionnels (chargements en traction et en cisaillement) sur tôles en utilisant la méthode de corrélation d'images numériques. Dans un premier temps, un modèle a été identifié à partir de cette base expérimentale sur tôle. La loi est validée par des calculs éléments finis d'essais de traction sur éprouvettes plates entaillées. Dans un second temps, la transférabilité du modèle de la tôle vers le tube a été étudiée. Pour les chargements uniaxiés, la transférabilité est avérée. En revanche, pour les chargements biaxiés la transférabilité est moins bonne. Une réidentification des paramètres gérant l'anisotropie du comportement en intégrant à la base d'identification un essai équibiaxié sur tube a été nécessaire. / Zirconium alloy sheet and clad are strongly textured materials, resulting in sharp anisotropic mecanical behavior. The purpose of this work is to characterize and model the anisotropic behavior of recrystallized Zircaloy-alloy sheets. Anisotropy is investigated by usual mechanical tests (tensile and shear loadings) performed on sheet material using digital image correlation measurments. A model is identified from this experimental database obtained on sheet material. Finite element simulations of tensile notched tests are used to validate the law. Then, the model transferability from the sheet to the rod is studied. The transferability is suitable for uniaxial loading. On the contrary, the transferability is not fully adequate for biaxial loadings. Therefore, a new identification of parameters dealing with anisotropy from enriched database with an equibiaxial rod test is necessary.

In dry interim storage of spent nuclear fuel, thermal creep is one of the major threats to the fuel cladding integrity due to the constant decay heat generation from the fission products and minor actinides in the fuel, and the increase in fuel rod internal pressure which is present after burnup. Plenty of research has been done on either empty cladding tubes irradiated in a research reactor, or on spent fuel that is defueled prior to the examination. This type of research excludes the effects of the pellet-cladding bonding that may be present after burnup, where the bonding might have significant effects on the thermal creep behavior. Therefore, this work aims to construct and validate an experimental model that is designed to perform thermal creep tests on as-received spent nuclear fuel, where the pellet-cladding bonding is still intact, in order to gain knowledge in the causal relation that the pellet-cladding bonding has on the thermal creep phenomenon during dry storage. The experimental model is validated by a number of qualification tests, as well as a series of creep tests on unirradiated Zircaloy-4 tubes. The results are compared to the literature which shows the reproducibility of the model, which further supports its validity. / I torrt mellanförvar av använt kärnbränsle så är termisk krypning en av de främsta farhågorna för kapslingens fysiska integritet genom sönderfallsvärmen från klyvningsprodukter och andra aktinider, samt det förhöjda interna trycket i kapslingen som är närvarande efter utbränning. Åtskillig forskning har gjorts på antingen tomma kapslingsrör som är bestrålade i en forskningsreaktor, eller på använt bränsle där bränslekutsen är urborrad innan undersökning. Denna typ av forskning utelämnar effekten av kuts-kapsling bindningen som uppstår under uppbränning, där bindningen kan ha en betydande effekt på kapslingens beteende under termisk krypning. Därför så dedikeras detta arbete till att konstruera och validera en experimentell modell som är designad till att utföra tester av termisk krypning på opåverkat använt bränsle, där kuts-kapsling bindningslagret fortfarande är intakt, för att lära om bindningslagrets effekt på fenomenet termisk krypning under torrt mellanförvar. Den experimentella modellen är validerad genom ett antal kvalificeringstester, samt en serie av kruypningstester på obestrålade Zircaloy-4 rör. Resultaten är jämförda mot litteraturen, vilket visar reproducerbarheten hos modellen, som i sin tur understöder modellens validitet.

This work was produced as part of a multidisciplinary study of the corrosion of zirconium alloys undertaken by a consortium of universities working in the MUZIC program; Oxford, Manchester and The Open University. The objective of the project as a whole was to further the understanding of the mechanisms of the breakaway oxidation process and to characterise these corrosion processes within a number of fuel rod cladding materials. This thesis describes laser 3D atom probe characterisation of the nano-scale chemical redistribution of oxygen and other solutes that occurs at the metal-oxide interface during corrosion, and a large body of technique development that was required to achieve this goal. The development of the metal-oxide interface of ZIRLO, a Zr-Nb-Sn-Fe-O alloy, is followed by generating 3D atomic scale reconstructions at four different stages of corrosion. The formation of a sub-oxide ZrO layer is seen during pre-transition oxide development. The ZrO interfacial layer is consumed by the rapid formation of oxide after the breakaway transition. After transition the chemistry of the interface is similar to the early pre-transition case, although an oxygen-saturated layer of metal adjacent to the interface formed during corrosion remains. The ZrO interfacial layer (Zr-ZrO-ZrO₂) and the region of oxygen-saturated material ahead of the metal-oxide interface alter the distribution of minor alloying additions such as niobium and iron. The ZrO layer increases the acceptance of niobium into the oxide, which is otherwise seen to be rejected at the Zr-ZrO2 interface along with iron. Niobium is seen to precipitate out of solution as nano-scale particles near the interface after around 100 days of corrosion. This is not seen in the bulk metal matrix of the corroded material due to the absence of other factors driving the process: the stress at the interface and a very high oxygen concentration in the metal ahead of the interface. The nano-scale niobium particles are found to be of a meta-stable composition. Iron is seen to redistribute in the corroded material and can be correlated with the local oxygen concentration. Similarities are seen in the behaviour of solutes within pre-transition ZIRLO and Zircaloy-4 (Zr-Sn-Fe-O). In both cases no redistribution of tin is seen at the metal-oxide interface. A Zr-Nb-Ti alloy with very poor corrosion resistance was also analysed in this way, and the similarities and differences with chemically-similar ZIRLO are discussed. The segregation of solutes to grain boundaries and solute clustering within the matrix are also examined before and after corrosion.

This work was performed as part of an EPSRC Leadership Fellowship [EP/I005420/1] for the study of irradiation damage in Zr alloys, and is supported heavily by industrial contributors and especially by Westinghouse, Studsvik and Rolls-Royce plc. for the investigation of mechanisms relating to irradiation-induced growth (IIG). This thesis is an analysis of the microchemical and microstructural evolution of Zircaloy-2 under both proton and neutron irradiation. Comparisons are made between the effects of the different irradiative species through the use of scanning transmission electron microscopy (STEM) and energy dispersive X-ray spectroscopy (EDS). The work takes advantage of advances in EDS capability with large solid angles of collection 0.7 srad coupled with an aberration-corrected FEI Titan ChemiSTEMTM with a high brightness X-FEG electron source.2 MeV proton irradiation experiments were performed to doses of 2.3, 4.7 and 7.0 displacements per atom (dpa) at a dose rate of ~6.7 x10-6 dpa s-1 and at 350 °C. Electropolished TEM foils from Zircaloy-2 cladding and channel components of a BWR were supplied by Westinghouse in the fluence range 8.7 to 14.7 x1025 n m-2 ~14.5 to 24.5 dpa. Comparisons have been made in relation to SPP chemical composition, grain boundary chemistry, dislocation density, correlations between dislocation evolution and microchemical segregations and the nature of irradiation-induced precipitates. Proton irradiation-induced dissolution was observed for both Zr(Fe,Cr)2 and Zr2(Fe,Ni) SPPs, the depletion of Fe was preferentially from the edge region in the former SPP and from throughout the whole SPP in the latter. While no proton-induced amorphisation was observed for the Zr(Fe,Cr)2, the compositional changes in all SPPs agreed well with the reports of other authors. All grain boundaries display Fe and Ni segregation prior to irradiation, which disperses into the matrix after both proton and neutron irradiation, while Sn segregates to the boundary. Sn and the light transition elements Fe, Cr and Ni have shown contrasting behaviour in the matrix also. After irradiation by both protons and neutrons, a-component dislocation loops (a-loops) align parallel to the basal plane and Fe, Cr and Ni segregate to the a-loop positions. Sn, conversely, segregates to between a-loop positions parallel to the basal trace. The threshold dose in c-component dislocation loop (c-loop) nucleation under proton irradiation (~4.5 dpa) is shown as similar to that due to neutron irradiation (~5 dpa). We observe that a-loop density decreases at the onset of c-loop nucleation and that the position of c-loops are in alignment with the a-loops but that they are anticorrelated in position along the basal trace. We therefore propose that chemical ordering promotes the alignment of a-loops, which then provides the conditions necessary for c-loop nucleation. Nanoprecipitation is evident in the matrix after both proton and neutron irradiation. After proton irradiation to ~2 dpa, parallel atom probe tomography and STEM-EDS investigations have shown the nano-rods to be of composition Zr4(Fe0.67Cr0.33), tending towards Zr3(Fe0.69Cr0.31) as the rod volume increases. The rods are higher in density than the a-loops by a factor of ~3 and so are likely to be a significant influence on mechanical properties and IIG phenomena.

Thesis (M.S.)--Massachusetts Institute of Technology, Dept. of Materials Science and Engineering, 1980. / MICROFICHE COPY AVAILABLE IN ARCHIVES AND SCIENCE. / Includes bibliographical references. / by Pierre Leon Jacques Chemelle. / M.S.

Dissertação (mestrado) - Universidade Federal do ABC. Programa de Pós-Graduação em Energia

L'objectif de cette étude est de caractériser et de modéliser le comportement mécanique et la rupture en situation accidentelle d'injection de réactivité de gaines de crayons combustibles en Zircaloy-4 détendu vierges, hydrurées ou irradiées en réacteurs nucléaires à eau pressurisée. Un modèle est proposé pour décrire le comportement viscoplastique anisotrope du matériau en fonction de la température (de 20°C à 1100°C), la vitesse de déformation (de 3.10-4 s-1 à 5 s-1), la fluence (de 0 à 1026 n.m-2) et des conditions d'irradiation. Le comportement plastique anisotrope et la rupture du matériau non irradié hydruré jusqu'à 1200 ppm est étudié à l'aide d'essais de traction axiale, traction circonférentielle, expansion due à la compression et traction plane circonférentielle réalisés à 25°C, 350°C et 480°C. La résistance mécanique et l'écrouissage du<br />matériau dépendent de la température et des teneurs en hydrogène en solution solide et en hydrures précipités. L'anisotropie plastique du matériau est peu modifiée par l'hydrogène. A température ambiante le matériau est fragilisé par les hydrures, qui rompent pour des déformations plastiques d'autant plus faibles que la teneur en hydrogène est élevée. La ductilité du matériau, croissante en fonction de la température, n'est pas affectée par l'hydrogène à 350°C et 480°C. Les modes de rupture macroscopiques et les mécanismes d'endommagement diffèrent selon la géométrie des éprouvettes, la température et la teneur en hydrogène. Un modèle de type Gurson est finalement proposé pour représenter le comportement viscoplastique anisotrope et la rupture ductile du matériau en fonction de la température et de la teneur en hydrogène.