High-temperature oxidation resistance of chromium-based coatings deposited by DLI-MOCVD for enhanced protection of the inner surface of long tubes

For nuclear safety issues, there is an international effort to develop innovative “Enhanced Accident Tolerant Fuels” (EATF) materials. EATF cladding tubes are of particular interest because they constitute the first barrier against radioactive fission species dispersal in case of accidental scenario such as LOCA (LOss of Coolant Accident). Actual nuclear fuel claddings are made from Zr-based alloys and to increase safety margins, both mechanical strength and resistance to high-temperature oxidation have to be improved. Several alternatives using high-temperature oxidation resistant coatings for outer-wall protection have been proposed worldwide but there is currently no solution for the inner-wall protection. In order to resist to high temperature steam environment upon LOCA transients, internal Cr-based coatings deposited by DLI-MOCVD (Direct Liquid Injection of MetalOrganic precursors) were investigated. These hard metallurgical coatings could also be used in high-temperature corrosive environments as those encountered in aeronautics and other industries to protect 3D complex components. Thanks to a suitable chemistry of the liquid Cr precursor, bis(ethylbenzene)chromium, different coatings were deposited including: metal Cr, chromium carbides CrxCy and mixed carbides CrxSizCy. The high-temperature behavior of these Cr-based coatings under oxidizing atmospheres has been studied using several techniques and various oxidation tests including pure steam environment followed by water quenching down to room temperature to be representative of LOCA situations. Amorphous CrxCy coatings showed the most promising properties. For instance compared to uncoated substrate, they shift the catastrophic oxidation towards higher temperatures and delay the complete oxidation of the substrate at 1473K of >2h. The results are discussed in terms of oxidation mechanisms and protection of the fuel claddings inner surface deduced from fine characterizations of the samples before and after oxidation tests

High Cycle Thermal Fatigue of two austenitic stainless steels

International audienceThermal fatigue is one of the main damage mechanisms that engineers must take into account in the design of current or future nuclear power plants. Fluctuations of temperature in the fluid is indeed inevitable, as a partial transfer of these fluctuations to the solid walls, and the resulting oscillations of thermal gradients must stay low enough to prevent the formation of typical crack networks. A chain of simulations, from thermohydraulics to thermo-mechanics of materials, is usually used to estimate the material loading that needs to be compared to material fatigue curves. There are not so many thermal fatigue experimental results in the literature that can serve to validate the last part of this chain of simulations, especially in the High Cycle Fatigue (HCF) regime. In this study, HCF thermal fatigue tests are performed under helium environment on two types of austenitic stainless steels, with an original apparatus in which cyclic thermal loading is imposed by a pulsed laser beam. Crack initiation and crack network developments are detected by an InfraRed (IR) camera that is also used with IR pyrometers to measure the thermal loading on the area impacted by the laser. The influence of temperature variation, pulse duration and a mean mechanical stress on crack initiation, crack propagation rate and crack network morphology is evaluated. Eventually, a numerical analysis of thermal fatigue tests, fed by proper experimental measurements on the crack initiation surface, allows deducing that the number of cycles to crack initiation is similar to the one obtained in more classical uniaxial isothermal fatigue results

Impact of intragranular misorientation on void swelling and inter-granular cavities after ion irradiation in standard and additive manufacturing 316 L austenitic steels

International audienceAdditive manufacturing (AM) is a promising technology for the design of materials with complex geometries with reduced cost and material waste. In order to be used in the nuclear industry, the capability of AM materials, in term of radiation resistance must be compared with materials elaborated in conventional ways. In this work, the radiation resistance of 316L austenitic stainless steels (ASSs) elaborated by AM is compared to a solution-annealed 316L ASS after irradiation with 5 MeV Fe$^{5+}$ for 3 dpa at 550 °C (873 K). After irradiation, cavities are mainly located near grain boundaries for all studied alloys. Intra-granular cavities are only found in the AM material after heat treatment and are likely to be remaining porosity already present before irradiation. No cavities in intra-granular position are found in the conventional 316L ASS or in the AM material after hot isostatic pressing (HIP) at 1100 °C. It suggests that the void swelling ASSs starts by the formation of cavities at grain boundaries followed by a formation of cavities in intra-granular position, conventionally studied. Loops in the AM material with a hot treatment at 700°C are heterogeneously distributed due to a bimodal distribution of grains in terms of intra-granular misorientation. The intra-granular misorientation drastically reduces the loop density after irradiation. Frank and perfect loops are found to be only interstitial-type loops. The larger cavities' size and the more advanced dislocations network in the AM HIP sample suggests a slightly reduced radiation swelling resistance for the AM material but further investigations at higher irradiation dose have to be done

Contributions of thermokinetic calculations to the understanding of the microstructural evolutions of E-ATF Cr Coated Zr-based nuclear fuel claddings upon high temperature transients

International audienceChromium coated zirconium based nuclear fuel claddings are among the most promising evolutive concepts in the field of Enhanced Accident Tolerant Fuel (EATF) developments [1-2]. The development of these new claddings has been accelerated since 2011 and the Fukushima-DAICHI nuclear accident. Among some other objectives, one of the main ideas is to develop new concepts of nuclear fuel claddings able to delay the High Temperature (HT) steam oxidation kinetics and the associated hydrogen production (explosion risk) and to improve the resultant Post-Quenching (PQ) cladding strength and ductility upon and following hypothetical accident conditions such as Loss-of-Coolant-Accident (LOCA). In this context, Cr-coated Zr based alloys, with 10-20µm thick Cr coating (PVD deposition), show promising behaviour

Principal image decomposition for multi-detector backscatter electron topography reconstruction

International audienceScanning Electron Microscopes (SEMs) often generate images with a shaded appearance which gives a natural 3D impression. Ergo, quite a few methods to reconstruct the 3D surface topography from these using shape-from-shading methods are available in the literature. Here, a novel approach is discussed which uses BackScatter Electron (BSE) images from multiple detectors to reconstruct the topography. Classically, algorithms exist which resort to a quad-BSE detector setup. However, other detector configurations are often found in SEMs. A set of images of these non-conforming detectors still contains enough information to allow for reconstruction, but requires a more general algorithm to do so. This article discusses a method based on a modal decomposition of the principal image components. The resulting method is shown to be efficient and independent of the number of detectors or their orientation. In fact, the orientation is identified as part of the algorithm and thus requires very little calibration

Study of the evolution of stresses and associated mechanisms in zirconia growing at high temperature on Zircaloy-4 by use of synchrotron radiation

International audienceThe present work studied the evolution of stresses and associated mechanisms in the zirconia layer formed during the oxidation of Zircaloy-4 under a He/O2 mixture, at temperatures of 700°C, 800°C and 900°C. Measurements by X-ray diffraction are performed in-situ under synchrotron radiation during oxidation to determine the evolutions of phases and stresses in the oxide layer with time. The results show that the zirconia formed contains a mixture of monoclinic and tetragonal phases. The proportion of the tetragonal phase depends on the oxidation temperature and decreases during oxidation. In order to better understand the influence of this evolution and localize the tetragonal phase, measurements by Raman spectroscopy have also been performed. These two phases are subjected to compressive stresses in directions perpendicular to the oxide layer growth direction. These stresses depend on temperature and relax during oxidation. Two mechanical models either considering the zirconia phases independently or considering an equivalent homogeneous oxide are proposed to describe the evolution of these stresses, considering that it is due to oxide viscoplasticity. The model parameters are analysed to discuss the mechanisms of viscoplastic flow in the oxide. Numerical values for the viscoplastic parameters of the model as well as for the corresponding activation energy are therefore provided

Study of the evolution of stresses and associated mechanisms in zirconia growing at high temperature on Zircaloy-4 by use of synchrotron radiation

International audienceThe present work studied the evolution of stresses and associated mechanisms in the zirconia layer formed during the oxidation of Zircaloy-4 under a He/O2 mixture, at temperatures of 700°C, 800°C and 900°C. Measurements by X-ray diffraction are performed in-situ under synchrotron radiation during oxidation to determine the evolutions of phases and stresses in the oxide layer with time. The results show that the zirconia formed contains a mixture of monoclinic and tetragonal phases. The proportion of the tetragonal phase depends on the oxidation temperature and decreases during oxidation. In order to better understand the influence of this evolution and localize the tetragonal phase, measurements by Raman spectroscopy have also been performed. These two phases are subjected to compressive stresses in directions perpendicular to the oxide layer growth direction. These stresses depend on temperature and relax during oxidation. Two mechanical models either considering the zirconia phases independently or considering an equivalent homogeneous oxide are proposed to describe the evolution of these stresses, considering that it is due to oxide viscoplasticity. The model parameters are analysed to discuss the mechanisms of viscoplastic flow in the oxide. Numerical values for the viscoplastic parameters of the model as well as for the corresponding activation energy are therefore provided

Chemical and structural evolution of nano-oxides from mechanical alloying to consolidated ferritic oxide dispersion strengthened steel

International audienceFerritic Oxides Dispersion Strengthened (ODS) steels are of great interest for nuclear fission and fusion power plants. The nano-oxides embedded into the matrix provide the main contribution to the ODS steel strength. Understanding of the precipitation mechanism of ODS steels is thus critical for optimizing the fabrication process, involving Mechanical Alloying (MA) of Fe-14Cr, Y2O3 and TiH2 powders. In this study, results from small-angle X-ray and neutron scattering, atom probe tomography and electron microscopy have been combined to investigate the nano-oxides evolution throughout the whole consolidation thermal treatment until 1100 °C. After MA clusters are observed, composed of Y, O and Ti. During heating these clusters grow and new ones nucleate, together with a sequential enrichment in Ti (from as-MA to 700 °C) and Y (between 900 and 1100 °C). A small quantity of Al is also found in the nano-oxides between 700 and 1100 °C. At 1100 °C the nano-oxides are found to be mainly Y2Ti2O7 and subsequently progressively transform to Y2TiO5 during isothermal holding. Nano-oxides display however an unchanged extremely low coarsening rate, demonstrating the outstanding stability of both Y2Ti2O7 and Y2TiO5 at 1100 °C

Mechanical behavior of a chromium coating on a zirconium alloy substrate at room temperature

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