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High temperature steam oxidation of chromium-coated zirconium-based alloys: Kinetics and process

Corrosion science, 2020-05, Vol.167, p.108537, Article 108537

2020

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Early studies on Cr-Coated Zircaloy-4 as enhanced accident tolerant nuclear fuel claddings for light water reactors

Journal of nuclear materials, 2019-04, Vol.517, p.268-285

2019

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Breakaway oxidation of zirconium alloys exposed to steam around 1000 °C

Corrosion science, 2020-11, Vol.176 (108936), p.108936-21, Article 108936

2020

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Fatigue of short carbon fiber reinforced PEEK under compression: Influence of the load ratio and predictions from heat buildup measurements

Fatigue & fracture of engineering materials & structures, 2023-07, Vol.46 (7), p.2396-2410

2023

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

Journal of nuclear materials, 2022-01, Vol.558, p.153332, Article 153332

2022

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A model to describe the cyclic anisotropic mechanical behavior of short fiber-reinforced thermoplastics

Mechanics of time-dependent materials, 2020-12, Vol.24 (4), p.481-503

2020

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Combined effects of temperature and of high hydrogen and oxygen contents on the mechanical behavior of a zirconium alloy upon cooling from the βZr phase temperature range

Journal of nuclear materials, 2021-10, Vol.554 (554), p.153069, Article 153069

2021

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Fatigue criteria for short fiber-reinforced thermoplastic validated over various fiber orientations, load ratios and environmental conditions

International journal of fatigue, 2020-06, Vol.135, p.105574-16, Article 105574

2020

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Evaluation of Equivalent Cladding Reacted parameters of Cr-coated claddings oxidized in steam at 1200 °C in relation with oxygen diffusion/partitioning and post-quench ductility

Journal of nuclear materials, 2020-05, Vol.533, p.152106-1 - 152106-16, Article 152106

2020

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COMBINED TECHNIQUES AND RELEVANT IMAGE PROCESSING FOR QUANTITATIVE STATISTICAL CHARACTERIZATION OF INCLUSIONS IN ELASTOMERS

Rubber chemistry and technology, 2023-01, Vol.96 (1), p.59-89

2023

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

Surface & coatings technology, 2018-09, Vol.349, p.1048-1057

2018

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In-situ X-ray diffraction analysis of zirconia layer formed on zirconium alloys oxidized at high temperature

Journal of nuclear materials, 2015-03, Vol.458, p.245-252

2015

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The SFIR Test: An Innovative Hydrostatic Compression Test to Characterize the Volumetric Behavior of Polymeric Foams

Experimental mechanics, 2024-08

2024

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Experimental investigation and theoretical modelling of induced anisotropy during stress-softening of rubber

International journal of solids and structures, 2016-10, Vol.97-98, p.554-565

2016

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A model to describe the mechanical behavior and the ductile failure of hydrided Zircaloy-4 fuel claddings between 25 °C and 480 °C

Journal of nuclear materials, 2015-11, Vol.466, p.43-55

2015

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Statistical characterization of microcellular polyurethane foams microstructure based on 2D and 3D image analysis

Journal of cellular plastics, 2023-11, Vol.59 (5-6), p.395-417

2023

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Phase transformations during cooling from the βZr phase temperature domain in several hydrogen-enriched zirconium alloys studied by in situ and ex situ neutron diffraction

Acta materialia, 2020-10, Vol.199, p.453-468

2020

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

Corrosion science, 2023-08, Vol.221, p.111328, Article 111328

2023

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In-situ time-resolved study of structural evolutions in a zirconium alloy during high temperature oxidation and cooling

Materials characterization, 2019-12, Vol.158, p.109971, Article 109971

2019

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Effect of a pre-oxide on the high temperature steam oxidation of Zircaloy-4 andM5 alloys

Journal of nuclear materials, 2019, Vol.518, p.386-399

2019

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DLI-MOCVD Crx Cy coating to prevent Zr-based cladding from inner oxidation and secondary hydriding upon LOCA conditions

Journal of nuclear materials, 2021, Vol.550

2021

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New insights in structural characterization of zirconium alloys oxidation at high temperature

Journal of nuclear materials, 2012-10, Vol.429 (1-3), p.19-24

2012

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Study of secondary hydriding at high temperature in zirconium based nuclear fuel cladding tubes by coupling information from neutron radiography/tomography, electron probe micro analysis, micro elastic recoil detection analysis and laser induced breakdown spectroscopy microprobe

Journal of nuclear materials, 2017-05, Vol.488, p.267-286

2017

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Early studies on Cr-Coated Zircaloy-4 as enhanced accident tolerant nuclear fuel claddings for light water reactors

Journal of nuclear materials, 2019-04, Vol.517, p.268-285

2019

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Behavior and failure of uniformly hydrided Zircaloy-4 fuel claddings between 25 °C and 480 °C under various stress states, including RIA loading conditions

Engineering failure analysis, 2010-04, Vol.17 (3), p.683-700

2010

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High temperature steam oxidation of chromium-coated zirconium-based alloys: Kinetics and process

Early studies on Cr-Coated Zircaloy-4 as enhanced accident tolerant nuclear fuel claddings for light water reactors

Breakaway oxidation of zirconium alloys exposed to steam around 1000 °C

Fatigue of short carbon fiber reinforced PEEK under compression: Influence of the load ratio and predictions from heat buildup measurements

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

A model to describe the cyclic anisotropic mechanical behavior of short fiber-reinforced thermoplastics

Combined effects of temperature and of high hydrogen and oxygen contents on the mechanical behavior of a zirconium alloy upon cooling from the βZr phase temperature range

Fatigue criteria for short fiber-reinforced thermoplastic validated over various fiber orientations, load ratios and environmental conditions

Evaluation of Equivalent Cladding Reacted parameters of Cr-coated claddings oxidized in steam at 1200 °C in relation with oxygen diffusion/partitioning and post-quench ductility

COMBINED TECHNIQUES AND RELEVANT IMAGE PROCESSING FOR QUANTITATIVE STATISTICAL CHARACTERIZATION OF INCLUSIONS IN ELASTOMERS

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

In-situ X-ray diffraction analysis of zirconia layer formed on zirconium alloys oxidized at high temperature

The SFIR Test: An Innovative Hydrostatic Compression Test to Characterize the Volumetric Behavior of Polymeric Foams

Experimental investigation and theoretical modelling of induced anisotropy during stress-softening of rubber

A model to describe the mechanical behavior and the ductile failure of hydrided Zircaloy-4 fuel claddings between 25 °C and 480 °C

Statistical characterization of microcellular polyurethane foams microstructure based on 2D and 3D image analysis

Phase transformations during cooling from the βZr phase temperature domain in several hydrogen-enriched zirconium alloys studied by in situ and ex situ neutron diffraction

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

In-situ time-resolved study of structural evolutions in a zirconium alloy during high temperature oxidation and cooling

Effect of a pre-oxide on the high temperature steam oxidation of Zircaloy-4 andM5 alloys

DLI-MOCVD Crx Cy coating to prevent Zr-based cladding from inner oxidation and secondary hydriding upon LOCA conditions

New insights in structural characterization of zirconium alloys oxidation at high temperature

Study of secondary hydriding at high temperature in zirconium based nuclear fuel cladding tubes by coupling information from neutron radiography/tomography, electron probe micro analysis, micro elastic recoil detection analysis and laser induced breakdown spectroscopy microprobe

Early studies on Cr-Coated Zircaloy-4 as enhanced accident tolerant nuclear fuel claddings for light water reactors

Behavior and failure of uniformly hydrided Zircaloy-4 fuel claddings between 25 °C and 480 °C under various stress states, including RIA loading conditions

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