In order to guarantee the integrity of nuclear fuel rod cladding, it is necessary for EDF to characterize the ductility of cladding after a Loss of Coolant Accident (LOCA). The thesis is about the characterization of the fracture behavior of cold-worked stress-relieved Zircaloy-4 claddings which have undergone LOCA conditions simulated in laboratory by a high temperature oxidation followed by a cooling. The high temperature oxidation is carried out at 1100°C and 1200°C with different times, which leads to different oxidation levels varying from 3% to 30% ECR (Equivalent Cladding Reacted). The high temperature oxidation is followed by two types of cooling: water quench and air cooling. The oxidized claddings contain two fragile layers - the ...
Ces travaux ont pour but d’établir une meilleure compréhension du comportement thermomécanique à la ...
During a Loss Of Coolant Accident (LOCA), an important safety requirement is that the reflooding of ...
International audienceThis work addressed fracture mechanisms of nuclear fuel claddings in hypotheti...
In order to guarantee the integrity of nuclear fuel rod cladding, it is necessary for EDF to charact...
Dans le cadre des études visant à garantir l'intégrité de la gaine du crayon combustible, EDF est am...
Le comportement des assemblages combustibles des Réacteurs Nucléaires à Eau Pressurisée (REP) doit ê...
During hypothetical Loss-Of-Coolant-Accident (LOCA) scenarios, zirconium alloy fuel cladding tubes a...
During the second stage of Loss Of Coolant Accident (LOCA) in Pressurized Water Reactors (PWR) zirco...
International audienceDuring reactivity initiated accident, the importance of cladding tube oxidatio...
This thesis deals with the study of the creep behavior of pre-oxidized Zy-4 cladding under simulated...
Lors d’un scénario hypothétique d’Accident par Perte de Réfrigérant Primaire (APRP), les gainages co...
En cas d'accident de réactivité, le comportement thermomécanique des gaines de crayons de combustibl...
Lors de certains scénarios accidentels, percement de cuve de réacteur avec entrée d’air, dénoyage de...
In nuclear plants, some accidental situations can result in air exposure of Pressurized Water Reacto...
The present work investigates the steady-state creep behavior of Stress Relieved Annealed Zircaloy-4...
Ces travaux ont pour but d’établir une meilleure compréhension du comportement thermomécanique à la ...
During a Loss Of Coolant Accident (LOCA), an important safety requirement is that the reflooding of ...
International audienceThis work addressed fracture mechanisms of nuclear fuel claddings in hypotheti...
In order to guarantee the integrity of nuclear fuel rod cladding, it is necessary for EDF to charact...
Dans le cadre des études visant à garantir l'intégrité de la gaine du crayon combustible, EDF est am...
Le comportement des assemblages combustibles des Réacteurs Nucléaires à Eau Pressurisée (REP) doit ê...
During hypothetical Loss-Of-Coolant-Accident (LOCA) scenarios, zirconium alloy fuel cladding tubes a...
During the second stage of Loss Of Coolant Accident (LOCA) in Pressurized Water Reactors (PWR) zirco...
International audienceDuring reactivity initiated accident, the importance of cladding tube oxidatio...
This thesis deals with the study of the creep behavior of pre-oxidized Zy-4 cladding under simulated...
Lors d’un scénario hypothétique d’Accident par Perte de Réfrigérant Primaire (APRP), les gainages co...
En cas d'accident de réactivité, le comportement thermomécanique des gaines de crayons de combustibl...
Lors de certains scénarios accidentels, percement de cuve de réacteur avec entrée d’air, dénoyage de...
In nuclear plants, some accidental situations can result in air exposure of Pressurized Water Reacto...
The present work investigates the steady-state creep behavior of Stress Relieved Annealed Zircaloy-4...
Ces travaux ont pour but d’établir une meilleure compréhension du comportement thermomécanique à la ...
During a Loss Of Coolant Accident (LOCA), an important safety requirement is that the reflooding of ...
International audienceThis work addressed fracture mechanisms of nuclear fuel claddings in hypotheti...