International audienceThis paper presents a unified phenomenological model to describe the anisotropic viscoplastic mechanical behavior of cold-worked stress relieved (CWSR) Zircaloy-4 fuel claddings submitted to reactivity initiated accident (RIA) loading conditions. The model relies on a multiplicative viscoplastic formulation and reproduces strain hardening, strain rate sensitivity and plastic anisotropy of the material. It includes temperature, fluence and irradiation conditions dependences within RIA typical ranges. Model parameters have been tuned using axial tensile, hoop tensile and closed-end internal pressurization tests results essentially obtained from the PROMETRA program, dedicated to the study of zirconium alloys under RIA lo...
The purpose of this study is to characterize and simulate the mechanical behaviour and failure of fr...
Abstract An anisotropic viscoplastic constitutive model was developed for the description of the hig...
International audienceDuring normal operating conditions, zirconium alloy nuclear fuel cladding tube...
This paper proposes a damaged viscoplastic model to simulate, for different isotherms (320, 350, 380...
International audienceThe research aims at characterizing the thermal-mechanical behavior of fresh Z...
International audienceThis paper presents new experimental thermo-mechanical tests recently develope...
Specimen geometries have been developed to determine the mechanical properties of irradiated Zircalo...
International audienceFuel cladding tubes, made of zirconium alloys are subjected, in-reactor, to a ...
The PROMETRA material testing program is a support program related to the study of high burnup fuel ...
International audienceThis paper presents an assessment of the thermo-mechanical behavior of non-irr...
Accurate prediction of cladding mechanical behavior is a key aspect of modeling nuclear fuel behavio...
International audienceZirconium alloys used as cladding tubes for the fuel of Pressurized Water Reac...
International audienceThis paper presents an assessment of the mechanical behavior of nuclear fuel c...
The purpose of this study is to characterize and simulate the mechanical behaviour and failure of fr...
Abstract An anisotropic viscoplastic constitutive model was developed for the description of the hig...
International audienceDuring normal operating conditions, zirconium alloy nuclear fuel cladding tube...
This paper proposes a damaged viscoplastic model to simulate, for different isotherms (320, 350, 380...
International audienceThe research aims at characterizing the thermal-mechanical behavior of fresh Z...
International audienceThis paper presents new experimental thermo-mechanical tests recently develope...
Specimen geometries have been developed to determine the mechanical properties of irradiated Zircalo...
International audienceFuel cladding tubes, made of zirconium alloys are subjected, in-reactor, to a ...
The PROMETRA material testing program is a support program related to the study of high burnup fuel ...
International audienceThis paper presents an assessment of the thermo-mechanical behavior of non-irr...
Accurate prediction of cladding mechanical behavior is a key aspect of modeling nuclear fuel behavio...
International audienceZirconium alloys used as cladding tubes for the fuel of Pressurized Water Reac...
International audienceThis paper presents an assessment of the mechanical behavior of nuclear fuel c...
The purpose of this study is to characterize and simulate the mechanical behaviour and failure of fr...
Abstract An anisotropic viscoplastic constitutive model was developed for the description of the hig...
International audienceDuring normal operating conditions, zirconium alloy nuclear fuel cladding tube...