The focus of this study is about a new experimental approach for a separate effects study of the secondary hydriding phenomenon under LOCA conditions. Many nuclear institutes perform semi-integrals tests to study the cladding behaviour during a LOCA transient. Those tests combined several phenomena and performing a detailed analysis of the secondary hydriding phenomenon using these tests can be challenging. A dedicated experimental protocol aiming at simulating secondary hydriding has been set up. Separate effects tests (SETs) were then carried out using this protocol to study the effects of both oxidation duration and temperature, on the hydrogen absorption during the oxidation stage of the LOCA transient on M5Framatome1 cladding. The effects of gap size were also investigated. Metallographic analysis has been used to characterise the M5Framatome clad metallurgical transformation after the high-temperature (HT) oxidation.

Hydride precipitation in zirconium alloys leads to embrittlement, making it essential to understand their prevalence and stability in the microstructure. Dictionary indexing of Kikuchi patterns, along with orientation relationship analysis and x-ray diffraction, confirmed the presence of both delta and gamma hydride phases in Zircaloy-4. Both phases were found to be stable in recrystallised zirconium, with the gamma phase exhibiting a distinct orientation relationship with the matrix. Delta hydride morphology and orientation were influenced by local stresses, resulting in a change in orientation during precipitation. By analysing the orientation relationships, the evolution of hydride phases could be visualised, providing insights into the room temperature stability of both delta and gamma hydrides.

After in reactor use, the nuclear fuel assembly of pressurized water reactors is stored in water and is eventually transported, in a dry cask, to a long-term storage facility or to a reprocessing plant. During transportation in dry environment the fuel rod temperature and internal pressure both increase. This leads to simultaneous creep deformation and annealing of the cladding tubes made of M5 Framatome1 zirconium alloy. This study shows that post -irradiation creep, or heat treatment, leads to the recovery of the radiation induced hardening and also to the recovery of the uniform elongation because of -loop annealing. On the other hand, it is shown that cladding tubes made of M5 Framatome alloy exhibit low post -irradiation creep rates, especially at low applied stress. This strain rate is significantly lower than for the unirradiated material and similar to the as -irradiated material, despite the significant loop annealing. The low creep rates are mainly attributed to the Nb-rich nano precipitates that act as dispersed strengthening particles. These nano -precipitates appear under irradiation and are stable at temperatures up to 450 degrees C explaining their strengthening effect even after significant annealing.

Based on the M5Framatome alloy metallurgy, several Zr1NbxSnyFe alloys were developed to make structural components, with ultra-low tin addition and slightly increased iron content (Sn = 0, 0.3, and 0.5 wt.%; Fe = 1,000 and 2,000 wt. ppm). This paper details the microstructure of five different alloys, including M5Framatome and Q12, and their microstructural evolution after neutron irradiation, in the same campaign, as fuel rods in a PWR up to high fluence. Previous studies have detailed microstructural changes of zirconium alloys under irradiation and have underlined the influence of these changes on oxidation behavior, mechanical properties, creep, and growth. The presence of tin and iron (by iron dissolution out of the precipitates) in the matrix is suspected to influence the irradiation-induced microstructural features such as -loop alignments (corduroys) and size, -component loop nucleation, and “needle like” β-Nb precipitate spatial distribution and size. Relevant microstructural observations are needed to decorrelate the tin's influence from that of the iron on the microstructural changes under irradiation in alloys containing niobium, tin, and iron. In this study, the comparison between M5Framatome and the Zr1Nb0.1Fe alloy has determined the influence of iron on alloys without tin. The effect of iron at a fixed tin content was obtained by comparing Zr1Nb0.3Sn0.1Fe and Zr1Nb0.3Sn0.2Fe alloys. Finally, the comparison of Zr1Nb0.1Fe, Zr1Nb0.3Sn0.1Fe, and Q12 (Zr-1Nb0.5Sn0.1Fe) alloys addressed the effect of different tin contents with the same iron content. The microstructural features were studied on all five alloys for fast neutron fluences up to 13 × 1025 n/m2(E > 1 MeV) with analytical transmission electron microscopy. Significant differences were brought out, particularly concerning the -loop distribution, the Laves phase dissolution, and the -component loop linear density and spatial distribution. All of these results prompt a reconsideration of the influence of iron and tin contents on microstructural evolution under irradiation of Zr1NbxSnyFe alloys.

Hydride precipitation and reorientation have the potential to embrittle zirconium alloys. This study aims to better understand the influence of the zirconium microstructure on hydride precipitation and reorientation. Specifically, the crystallography, phase stability, and morphology of hydride precipitation were correlated to microstructural variations due to changes in the metallurgical state of the zirconium alloy. This work highlights that microstructural features induced during recrystallization have a significant influence on the distribution and orientation of hydrides when no external stress is applied. The stability of γ hydride was shown to be dictated by metallurgical state, whereby its formation was promoted in the recrystallized sample owing to its reduced strength. The influence of grain orientation on γ stability was also explored. It was highlighted that upon cooling, grains oriented in the <101̅0> direction are under compression such that γ-hydride formation is suppressed. This study suggests that the extent of reorientation is driven by differences in hydrogen content of the alloy as well as the applied stress during reorientation, while the influence of the metallurgical state still remains unclear. Quantification of the dislocation density in both the matrix and hydride during precipitation highlighted that extensive matrix recovery takes place during hydriding. It was also shown that the dislocation density in the hydride is lower after thermomechanical loading, whereby the presence of dislocation nests left behind after initial hydride precipitation and dissolution could provide more space for the hydride to precipitate into.

Secondary hydriding may occur during Loss of Coolant Accidents (LOCA) when nuclear fuel cladding is exposed to steam at high temperature. Indeed, during a LOCA, the cladding may burst, and the steam ingress results in cladding inner surface oxidation. The hydrogen released during this oxidation can be partly absorbed by the cladding, leading to a localized high hydrogen content. This study aims at characterizing the influence of different parameters on the maximum hydrogen content, localization and distribution regarding secondary hydriding. The oxidation duration has been studied using an axisymmetric testing procedure reproducing LOCA conditions. Metallographic analysis and hydrogen measurements have been performed. EPMA and µ-LIBS micro-analysis have been used to map oxygen and the hydrogen local content.

Whilst substantial progress has been made in understanding the influence that hydrides have on the mechanical properties of zirconium alloys, there is currently an urgent need for a transparent, reproducible image analysis workflow for their characterisation. In this study, an open-source software package for the analysis of hydride networks, HAPPy (Hydride Analysis Package in Python), is introduced to calculate the radial hydride fraction (RHF) and mean hydride length, as well as characterising the connectivity of the microstructure both quantitatively and qualitatively. In this study, we used the Hough line transform to calculate the orientation distribution of the hydride segments within a micrograph, and its projection on to the radial direction is used to determine the RHF. The proposed methodology is validated, and its robustness is demonstrated over a wide range of microstructures. The image processing prior to analysis as well as the projection method used has been shown to have a significant influence on the calculated RHF, highlighting the need for standardized image analysis workflows to facilitate accurate comparisons and correlations across different studies in the literature. Finally, this paper introduces a new damage susceptibility parameter termed the branch length fraction, which can be used in conjunction with a path of lowest cost algorithm to visualise the most plausible crack path as well as the connectivity evolution over an entire micrograph.

Under irradiation, zirconium (Zr) alloys experience free growth, following a three-step kinetic: a transitioning phase with fast growth, followed by a stabilization phase, and finally an acceleration phase (fluence 6 to 10 × 1021 n/cm2). While the initial growth kinetics are related to the dislocation loops, the acceleration phase is driven by the nucleation and growth of dislocation loops. Corrosion of zirconium alloys in a pressurized water reactor environment leads to an increase in their hydrogen content, with hydrogen being a subproduct of the oxidation reaction. Experiments have evidenced that increasing hydrogen content in zirconium alloys leads to higher loop linear density. Atomistic modeling also shows that increasing the hydrogen concentration lowers the stacking fault energy of the loops and thus stabilizes them. Using cluster dynamics calculations, this study attempts to simulate the microstructural evolution observed in zirconium alloys under irradiation, based on the diffusional anisotropy difference model, which is well adapted for zirconium. The hydrogen effect is included in this model by means of its influence on the stacking fault energy of loops. After neutron irradiation in the research reactor, the microstructure of prehydrided Zr alloy samples has been characterized. A correlation between the hydrogen content in the samples and the nucleation and growth of dislocation loops is highlighted. The modeling results are then compared to the experimental observations and a good agreement is obtained.

Les alliages de zirconium sont utilises comme materiaux de structure des crayons de combustible nucleaire pour les centrales nucleaires a eau legere en raison de leurs proprietes neutroniques, mecaniques satisfaisantes et de leur resistance a la corrosion. En conditions nominales de fonctionnement (environ 150 bar et 350°C), les tubes en alliage de zirconium sont sujets a une oxydation maitrisee en presence d’eau pressurisee, servant de caloporteur. La demonstration de la surete des reacteurs nucleaires integre l’etude des scenarios accidentels hypothetiques, tels que l’Accident de Perte de Refrigerant Primaire (APRP). Dans ce scenario, une breche dans le circuit primaire du reacteur peut conduire les crayons de combustible a etre oxydes en milieu de vapeur d’eau a haute temperature (jusqu’a 1200°C). Il a ete demontre [1]–[3] que dans l’intervalle de temperature [900-1050°C] et apres un certain temps d’oxydation sous vapeur d’eau, le phenomene de breakaway peut apparaitre. Il se manifeste par une acceleration de la cinetique d’oxydation, par la perte du caractere protecteur de la couche d’oxyde et par l’absorption d’hydrogene du materiau. Un tel phenomene implique donc une deterioration des proprietes mecaniques de la gaine et menace son integrite pendant et apres le transitoire accidentel. Les mecanismes fondamentaux a l’origine du breakaway sont encore mal connus. Dans ce cadre, ce travail de these a pour objectif de caracteriser les conditions d’apparitions du breakaway et d’ameliorer la comprehension des mecanismes physiques sous-jacents. Pour ce faire, des essais d’oxydation sur des echantillons de Zy-4 en thermobalance symetrique ont permis d’etudier l’influence de la temperature, de la pression partielle de vapeur d’eau et du temps sur l’occurrence du breakaway. La couche d’oxyde formee a ete caracterisee par des observations en microscopie optique et des analyses en DRX in-situ pour suivre l’evolution des phases de la zircone durant l’oxydation.

Zirconium alloys used as cladding tubes for the fuel of pressurized water reactors can undergo high applied stress during power transients. In these conditions, biaxial loading may lead to plastic deformation of the cladding. A comprehensive understanding of the material mechanical response during loading-path change tests is an important step toward predicting the behavior in these specific conditions. Using a nonstandard mechanical testing machine, we conducted loading-path change tests at 623 K on as-received recrystallized Zircaloy-4 tubes. These tests consisted of an axial tensile loading and unloading followed by an internal pressure (or pure hoop tension) loading and unloading. These tests examined the kinematic and isotropic hardening components of the strain hardening behavior of the thin cladding tubes. The isotropic hardening was attributed to dislocation multiplication and dislocation-dislocation interactions. The kinematic hardening was attributed to the interaction of the grains with each other. A polycrystalline model was also used and improved in order to simulate the tests. A good prediction of the isotropic and kinematic hardening was provided by the modeling.

In order to improve the understanding of the microscopic phenomena occurring during irradiation in zirconium alloys, ion beam irradiations are performed at 350 degrees C (dose-rate of 2 x 10(-5) dpa/s) on recrystallized Zy-4 and M5((R)) 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 degrees C, 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((R)) undergo a uniform amorphisation, while the native beta Nb 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((R)) 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. (C) 2017 Elsevier B.V. All rights reserved.

Welding can highly modify the mechanical properties of materials due to the extreme thermal solicitations applied. For precipitation hardened materials, such as aluminium alloy 6xxx, a welding operation implies a modification of the microstructural state and, consequently, of the mechanical properties, both phenomena being highly nonlinear. The purpose of this paper is to propose a methodology to predict the post-welding mechanical properties of a welded joint. For this, three models are coupled: (i) a thermal finite element model of the welded structure that allows the prediction of the material's thermal history at every point; (ii) a precipitation model to predict the microstructural state in the joint using the thermal history; and (iii) a mechanical model to link the microstructural state to the mechanical properties, i.e. hardness, yield limit and hardening. A coupling between these models and a finite element commercial code is then performed to predict the precipitation state and mechanical properties of a 6xxx-T6 aluminium alloy after welding. To validate this methodology a tensile test is performed on a specimen extracted from a 6061-T6 welded plate. Using Digital Image Correlation, the in-plane strain fields across the weld are measured and compared with the finite element simulation of the tensile test, thereby providing good prediction.

The aim of this research is to link the microstructural state and the mechanical properties of an age hardening alloy. during a fast heat treatment such as encountered during welding. A coupled model between precipitation state and mechanical properties is used to predict the yield strength and hardening behavior that can be observed experimentally. The method permits the identification of the kinematic and isotropic contributions in the hardening model. The methodology is applied to a 6061-T6 aluminium alloy which is used in the Jules Horowitz reactor vessel. The general idea of this methodology is to couple an efficient microstructural model to a mechanical one based on the dislocation theory and ad'hoc experiments. The theoretical background is based on the work of Kampmann and Wagner, known as the KWN model, to account for nucleation, growth/dissolution and coarsening of precipitates. This analysis requires transient thermo-mechanical experimental data. The efficiency of these models and their coupling are shown for a serie 6XXX aluminium alloy which contains beta '' and beta' precipitates. Ultimately these models are coupled to a FEA model and allows to predict the distribution of precipitates within each element of the mesh, and subsequently its mechanical behavior

In age-hardening alloys, high-temperature processes, such as welding, can strongly modify the precipitation state, and thus degrade the associated mechanical properties. The aim of this paper is to present a coupled approach able to describe precipitation and associated yield stresses for non-isothermal treatments of a 6061 aluminium alloy. The precipitation state (in terms of volume fraction and precipitate size distribution) is modelled thanks to a recent implementation of the classical nucleation and growth theories for needle-shaped precipitates. The precipitation model is validated through small-angle neutron scattering and transmission electron microscopy experiments. The precipitation size distribution is then used as an entry parameter of a micromechanical model for the yield strength of the alloy. Predicted yield stresses are compared to tensile tests performed with various heating conditions, representative of the heat-affected zone of a welded joint.