Nano-scale chemical evolution in a proton-and neutron-irradiated Zr alloy

Proton-and neutron-irradiated Zircaloy-2 are compared in terms of the nano-scale chemical evolution within second phase particles (SPPs) Zr(Fe,Cr)2 and Zr2(Fe,Ni). This is accomplished through ultra-high spatial resolution scanning transmission electron microscopy and the use of energy-dispersive X-ray spectroscopic methods. Fe-depletion is observed from both SPP types after irradiation with both irradiative species, but is heterogeneous in the case of Zr(Fe,Cr)2, predominantly from the edge region, and homogeneously in the case of Zr2(Fe,Ni). Further, there is evidence of a delay in the dissolution of the Zr2(Fe,Ni) SPP with respect to the Zr(Fe,Cr)2. As such, SPP dissolution results in matrix supersaturation with solute under both irradiative species and proton irradiation is considered well suited to emulate the effects of neutron irradiation in this context. The mechanisms of solute redistribution processes from SPPs and the consequences for irradiation-induced growth phenomena are discussed.<br/

INVESTIGATION OF NEUTRON RADIATION EFFECTS ON THE MECHANICAL BEHAVIOR OF RECRYSTALLIZED ZIRCONIUM ALLOYS

International audienceNeutron radiation induces important changes in the mechanical behavior of recrystallized zirconium alloys used as fuel cladding tube. The neutron radiation effects on the mechanical behavior for internal pressure test performed at 350DC have been investigated using a specific analysis in terms of isotropic hardening, kinematic hardening and viscous stress. The impact of irradiation has been interpreted in terms of microscopic deformation mechanisms observed by Transmission Electron Microscopy (TEM). It is proposed that because of the localization of the plastic deformation inside channels and because of the only activation of basal channeling, the kinematic hardening must be strong in irradiated zirconium alloys. A simple unified phenomenological modeling is also used in order to have a coherent description of the radiation effects on the mechanical behavior

INVESTIGATION OF NEUTRON RADIATION EFFECTS ON THE MECHANICAL BEHAVIOR OF RECRYSTALLIZED ZIRCONIUM ALLOYS

International audienceNeutron radiation induces important changes in the mechanical behavior of recrystallized zirconium alloys used as fuel cladding tube. The neutron radiation effects on the mechanical behavior for internal pressure test performed at 350DC have been investigated using a specific analysis in terms of isotropic hardening, kinematic hardening and viscous stress. The impact of irradiation has been interpreted in terms of microscopic deformation mechanisms observed by Transmission Electron Microscopy (TEM). It is proposed that because of the localization of the plastic deformation inside channels and because of the only activation of basal channeling, the kinematic hardening must be strong in irradiated zirconium alloys. A simple unified phenomenological modeling is also used in order to have a coherent description of the radiation effects on the mechanical behavior

Transmission electron microscopy study of second phase particles irradiated by 2 MeV protons at 350 °C in Zr alloys

International audienceIn order to improve the understanding of the microscopic phenomena occurring during irradiation in zirconium alloys, ion beam irradiations are performed at 350 DC (dose-rate of 2 × 10-5 dpa/s) on recrystallized Zy-4 and M5 alloys, with 2 MeV protons. The aim of this study is to determine in which way proton irradiations can be representative of neutron irradiations, considering the second phase particle changes and the influence of these changes on the microstructural evolution of the material during irradiation. The 2 MeV proton irradiation at 350 DC, performed here, seems to reproduces well what happened in Zy-4 in PWR conditions with a progressive amorphisation of the Zr(Fe,Cr)2 Laves phases, but with a lower growth rate and a higher Fe/Cr ratio of the amorphous rim. The Zr(Fe,Nb)2 s phase particles in M5 undergo a uniform amorphisation, while the native βNb precipitates remain fully crystalline as evidenced in neutron irradiation at very low irradiation temperature. No radiation-enhanced precipitation of nanometric BETA-Nb particles was observed. Thus, for M5 alloy, the present irradiation seems to be representative of neutron irradiations at a very low irradiation temperature. Nevertheless it does not reproduce what happens in PWR conditions, where no amorphisation and a drastic loss of iron is reported for the Zr(Fe,Nb)2 Laves phase SPPs. Despite the lower iron rejection from the particles into the matrix during proton irradiation than during neutron irradiation, <c>-component loop distribution is found to be similar after both types of irradiations. These results underline the influence of both dose-rate and temperature on second phase particles behavior under irradiation and point out the complexity of iron rejection influence on the basal <c>-component loops. Indeed, although the <c>-component loop nucleation and growth seem locally correlated to iron dissolution into the matrix, they do not seem to be directly correlated to the global amount of iron rejected

The effect of applied stress on <c>-component dislocation loops in Zr-based alloys

International audienceTMS 2015 Annual Meeting and ExhibitionMaterials and Fuels for the Current and Advanced Nuclear Reactors IVThe effect of applied stress on -component dislocation loops in Zr-based alloysN. Gharbi1 and R.M. Hengstler-Eger2, X. Feaugas3, D. Gilbon4, P.B. Hoffmann2, M.A. Kirk5, J.P. Mardon6, F. Onimus1 1CEA, DEN, Section for Applied Metallurgy Research, 91191 Gif-Sur-Yvette, Cedex, France2AREVA GmbH, Paul-Gossen-Str. 100, 91052 Erlangen, Germany3LaSIE, UMR CNRS 7356, Universite de La Rochelle, 17042 La Rochelle Cedex 01, France4CEA, DEN, Nuclear Materials Department, 91191 Gif-Sur-Yvette, Cedex, France5Argonne National Laboratory, 9700 S. Cass Avenue, Argonne, IL60439, USA6AREVA NP, 10 rue Juliette Recamier, 69456 Lyon Cedex 06, FranceAbstractIrradiation-induced PWR fuel assembly deformation at high burn-ups is correlated with the appearance of specific irradiation defects -component dislocation loops. Industrial feedback suggests that a coupling between axial creep and stress-free growth could exist. Therefore, the effect of an external stress on -loop microstructure was studied through two complementary experiments on recrystallized Zircaloy-4 and M5 samples. Firstly, bending experiments were conducted under Zr irradiation beyond the c-loop incubation dose and thorough TEM analyses were performed after irradiation on many grains.Secondly, in-situ tensile tests were carried out under Kr irradiation to higher dose in a Transmission Electron Microscope allowing the observation of c-loop growth under stress in few grains. TEM observations showed that the applied stress has a minor effect on c-loop incubation dose. Moreover, their linear density decreases when a tensile stress is applied parallel to the c-axis. This effect was observed with different magnitudes, depending on experimental conditions

Le comportement des matériaux de structure et de gainage sous irradiation

La modification des propriétés métallurgiques et mécaniques des matériaux avec l'irradiation limite leur utilisation sous flux. Les mécanismes associés à la dégradation de leurs propriétés (gonflement, croissance et fluage sous flux, durcissement, décalage de la température de transition, ...) ont fait l'objet de nombreuses recherches. Ce chapitre présente l'intérêt des réacteurs expérimentaux pour étudier les performances des matériaux et pour améliorer la prédiction de la durée de vie des composants. La première partie porte sur les matériaux du parc actuel, avec comme enjeux majeurs l'augmentation de la compétitivité économique et de la durée de vie des réacteurs. La seconde partie est consacrée à la génération suivante de matériaux pour les filières du futur. Les irradiations expérimentales doivent elles aussi s'adapter pour se rapprocher des conditions de fonctionnement envisagées et être mieux instrumentées pour suivre in situ l'évolution du comportement sous flux

Effets des transformations de phases sur le comportement mécanique d’alliages base zirconium, pendant et après incursion à haute température en ambiance oxydante (vapeur)

Dans le cœur des réacteurs nucléaires à eau légère, les tubes de gainage en alliage base zirconium constituent la première barrière de confinement du combustible, et leur intégrité ainsi que leur stabilité dimensionnelle sont des paramètres clés aussi bien en conditions nominales de fonctionnement que lors de scénarios hypothétiques accidentels. Dans certains types de transitoires, on suppose que le tube de gainage se trouve soumis à une brusque élévation de température et de pression interne, du fait de la dépressurisation du milieu primaire, de la puissance résiduelle du combustible et de la pression interne due au gaz initial (+ gaz de fission). Le ballonnement du tube qui s’ensuit et son éventuelle rupture doivent être correctement appréhendés afin de vérifier le respect des critères de sûreté pour pouvoir, lors de la phase finale d’un tel transitoire, assurer le refroidissement efficace du cœur (notamment éviter un ballonnement excessif qui conduirait à boucher les canaux entre crayons) et limiter une dispersion des produits et des gaz radioactifs (du fait d’une éventuelle multi-fragmentation des crayons combustible). L’article se propose ainsi de décrire les transformations métallurgiques successives qui se produisent au sein du tube de gaine lorsque celui-ci subit une incursion à haute température (i.e. >800 °C) en ambiance oxydante (vapeur d’eau) jusqu’au retour à basse température, ainsi que leurs conséquences sur le comportement mécanique du matériau

Microstructural Evolution of M5 TM7 Alloy Irradiated in PWRs up to High Fluences—Comparison With Other Zr-Based Alloys

International audienc

Impact of Hydrogen Pick-Up and Applied Stress on C-Component Loops: Toward a Better Understanding of the Radiation Induced Growth of Recrystallized Zirconium Alloys

Impact of hydrogen pickup and applied stress on c-component loops: toward a better understanding of the radiation induced growth of recrystallized zirconium alloys, published in ASTM STP 1543 (17th International Symposium on Zirconium in the Nuclear Industry