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CEA studies on High temperature oxidation and hydriding of Zr based nuclear fuel claddings upon LOCA transients phenomenology, mechanisms and modelling => consequences on mechanical properties

Abstract : During a hypothetical loss-of-coolant accident (LOCA) in a light water reactor, fuel claddings made of zirconium alloys are loaded by internal pressure and exposed to steam at high temperature (HT, potentially up to 1200DC) during a few minutes, before being cooled then water quenched. During such a transient, the cladding material undergoes a number of metallurgical changes. These evolutions have a strong influence on the cladding mechanical properties, which have to remain sufficient for safety reasons. In particular, the cladding is embrittled by its oxidation at HT, due to the growth of an oxide layer, oxygen diffusion through the underlying metal and hydrogen absorption in some conditions. This presentation will provide an overview of the studies carried out at CEA (in collaboration with EDF, AREVA and other laboratories from France and other countries) since the 1990s to (i) understand the phenomenology and the mechanisms of oxidation and hydriding of Zr alloys at HT, (ii) evaluate their consequences on the cladding mechanical properties and (iii) model and simulate the phenomena.The experimental approach relies on HT steam oxidation and water quenching tests performed in dedicated facilities (DEZIROX 1 et 2, CINOG BP et HP with more than 2000 tests performed to date), mechanical testing (ring compression, bending, impact) and multiscale characterization of the material microstructure by using various complementary techniques (microscopy, electron probe micro analysis, micro elastic recoil detection analysis, X-ray and neutron diffraction, neutron radiography/tomography). The modeling work includes the development of a thermodynamic database (Zircobase) and of a multi-component diffusion database (Zircomob), thermodynamic calculations (Thermo-Calc together with the Zircobase database), numerical modeling of HT oxidation (Dictra associated with the Zircomob database, EKINOX-Zr with the Zircobase database), modeling of the material mechanical properties and finite element simulation of mechanical tests. Various Zr alloys have been studied Zircaloy-4, M5, model alloys The influence of several parameters has been investigated pre-transient oxide layer and hydriding resulting from in-service corrosion, oxidation temperature (700-1400DC) and time (1 min up to a few hours), steam pressure (1-80 bar), cladding having burst or not, cooling scenario This work has contributed to a better understanding of oxidation and hydriding of Zr alloys at HT and to the development of rather predictive models. Nevertheless, there are still some pending issues and some of the models still need to be improved to become fully predictive. Thanks to its expertise and to the methodologies, devices, tools and models developed to study the behavior of Zr alloys under LOCA conditions, CEA is contributing for a few years to the development, for light water reactors, of enhanced accident tolerant fuel claddings (Cr-coated Zr based and sandwich SiC-SiC claddings), having in particular a higher resistance to oxidation at HT.
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Submitted on : Tuesday, January 14, 2020 - 11:01:30 AM
Last modification on : Tuesday, May 25, 2021 - 9:52:03 PM
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  • HAL Id : cea-02438347, version 1




J.-C. Brachet, M. Le Saux, C. Toffolon-Masclet. CEA studies on High temperature oxidation and hydriding of Zr based nuclear fuel claddings upon LOCA transients phenomenology, mechanisms and modelling => consequences on mechanical properties. Séminaire scientifique DEN sur la corrosion dans les REP - Corrosion in pressurized water reactors: phenomenology, mechanisms and modelling, Oct 2016, Saclay, France. ⟨cea-02438347⟩



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