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Study of secondary intermetallic phase precipitation/dissolution in Zr alloys by high temperature–high sensitivity calorimetry

Journal of nuclear materials, 2008-01, Vol.372 (2), p.367-378

2008

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

Journal of nuclear materials, 2021-07, Vol.550, p.152953, Article 152953

2021

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An approach to study oxidation-induced stresses in Zr alloys oxidized at high temperature

Corrosion science, 2018-08, Vol.140, p.79-91

2018

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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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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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Influence of grain size on the high-temperature creep behaviour of M5Framatome1 zirconium alloy under vacuum

Journal of nuclear materials, 2022-03, Vol.560, p.153503, Article 153503

2022

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Creep flow and fracture behavior of the oxygen-enriched alpha phase in zirconium alloys

Scripta materialia, 2016-05, Vol.117, p.20-23

2016

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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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Effect of a pre-oxide on the high temperature steam oxidation of Zircaloy-4 and M5Framatome alloys

Journal of nuclear materials, 2019-05, Vol.518, p.386-399

2019

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Mechanical behavior of reduced-activation and conventional martensitic steels after neutron irradiation in the range 250–450°C

Journal of nuclear materials, 2000-12, Vol.283-287, p.353-357

2000

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Influence of chemical composition on the microstructure and phase transformations of Fe-14Cr ferritic steels

Materialia, 2019-09, Vol.7, p.100388, Article 100388

2019

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Numerical modeling of oxygen diffusion in the wall thickness of Low-Tin Zircaloy-4 fuel cladding tube during high temperature (1100–1250 °C) steam oxidation

Journal of nuclear materials, 2010-05, Vol.400 (3), p.196-204

2010

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Heterogeneous austenite grain growth in 9Cr martensitic steels: influence of the heating rate and the austenitization temperature

Materials science & engineering. A, Structural materials : properties, microstructure and processing, 2003-05, Vol.348 (1), p.122-132

2003

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A thermodynamic database for zirconium alloys

Journal of nuclear materials, 1999-11, Vol.275 (3), p.287-295

1999

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Experimental determination of creep properties of Zirconium alloys together with phase transformation

Scripta materialia, 2004-09, Vol.51 (6), p.515-519

2004

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Microstructural influence on high temperature creep flow of Zr–1%NbO alloy in near- α, ( α + β), and β temperature ranges in a high vacuum environment

Journal of nuclear materials, 2011, Vol.408 (1), p.116-124

2011

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Quantification and local distribution of hydrogen within Zircaloy-4 PWR nuclear fuel cladding tubes at the nuclear microprobe of the Pierre Süe Laboratory from μ-ERDA

Nuclear instruments & methods in physics research. Section B, Beam interactions with materials and atoms, 2008-05, Vol.266 (10), p.2424-2428

2008

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Oxidation kinetics and oxygen diffusion in low-tin Zircaloy-4 up to 1523K

Journal of nuclear materials, 2008-07, Vol.377 (2), p.359-369

2008

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Physical metallurgy and mechanical behaviour of FeCrWTaV low activation martensitic steels: Effects of chemical composition

Journal of nuclear materials, 1998-10, Vol.258, p.1228-1235

1998

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Relationship between Crystallographic Texture and Dilatometric Behaviour of a Hexagonal Polycrystalline Material

Materials science forum, 1998-01, Vol.273-275, p.529-534

1998

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Pre- and post-irradiation mechanical properties of ferritic-martensitic steels for fusion applications: EM10 base metal and EM10/EM10 welds

Journal of nuclear materials, 1994, Vol.212, p.588-593

1994

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Fast Reactor Cores, Ferritic–Martensitic Steels for

Encyclopedia of Materials: Science and Technology, 2011, p.2863-2866

2011

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Intercomparison experiments of systems for the measurement of xenon radionuclides in the atmosphere

Applied radiation and isotopes, 2004-06, Vol.60 (6), p.863-877

2004

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Study of secondary intermetallic phase precipitation/dissolution in Zr alloys by high temperature–high sensitivity calorimetry

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

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

An approach to study oxidation-induced stresses in Zr alloys oxidized at high temperature

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

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

Influence of grain size on the high-temperature creep behaviour of M5Framatome1 zirconium alloy under vacuum

Creep flow and fracture behavior of the oxygen-enriched alpha phase in zirconium alloys

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

Effect of a pre-oxide on the high temperature steam oxidation of Zircaloy-4 and M5Framatome alloys

Mechanical behavior of reduced-activation and conventional martensitic steels after neutron irradiation in the range 250–450°C

Influence of chemical composition on the microstructure and phase transformations of Fe-14Cr ferritic steels

Numerical modeling of oxygen diffusion in the wall thickness of Low-Tin Zircaloy-4 fuel cladding tube during high temperature (1100–1250 °C) steam oxidation

Heterogeneous austenite grain growth in 9Cr martensitic steels: influence of the heating rate and the austenitization temperature

A thermodynamic database for zirconium alloys

Experimental determination of creep properties of Zirconium alloys together with phase transformation

Microstructural influence on high temperature creep flow of Zr–1%NbO alloy in near- α, ( α + β), and β temperature ranges in a high vacuum environment

Quantification and local distribution of hydrogen within Zircaloy-4 PWR nuclear fuel cladding tubes at the nuclear microprobe of the Pierre Süe Laboratory from μ-ERDA

Oxidation kinetics and oxygen diffusion in low-tin Zircaloy-4 up to 1523K

Physical metallurgy and mechanical behaviour of FeCrWTaV low activation martensitic steels: Effects of chemical composition

Relationship between Crystallographic Texture and Dilatometric Behaviour of a Hexagonal Polycrystalline Material

Pre- and post-irradiation mechanical properties of ferritic-martensitic steels for fusion applications: EM10 base metal and EM10/EM10 welds

Fast Reactor Cores, Ferritic–Martensitic Steels for

Intercomparison experiments of systems for the measurement of xenon radionuclides in the atmosphere

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