The prestressed concrete confinement vessel is the third and last barrier in French Nuclear Power Plants (NPP). In case of a s vere accident (loss of cooling agent of the reactor for instance), pressure and temperature will increase in the nuclear vessel (+0,5 MPa and 180 °C during 2 weeks). Due to elevated temperatures, the evolution of basic creep will be accelerated. High compressive stresses would appear and induce higher delayed strains and damage. The modelling of basic creep and its couplings with temperature is then very important for the safety of the structure (tightness of the concrete vessel). Here, we present a model considering the following elements: a coupling between creep and damage is introduced, kinetics of basic creep i...
This thesis deals with the problem of creep in concrete and its influence on thermal stress developm...
The paper is given of an extended model for concrete in compression at elevated temperature that inc...
This report is concerned with the analysis and its' review of a prestresied concrete reactor vessel ...
The prestressed concrete confinement vessel is the third and last barrier in French Nuclear Power Pl...
This study is a part of a French national project dealing with the mechanical behaviour of the conta...
WOS:000313369400002International audienceThe present study concerns the uniaxial compressive creep o...
WOS:000356421200006International audienceThis research concerns the uniaxial compressive basic creep...
International audienceThe objective of this research is to understand the behavior of concrete subje...
A numerical model has been developed to predict early-age cracking for massive concrete structures. ...
Based on the existing limited test data, it is possible to set up an approximate constitutive model ...
A constitutive model for the analysis of deformations of concrete subject to transient temperature a...
International audienceDifferent factors related to construction site conditions must be accounted fo...
This research has been conducted within the framework of the study of concrete behaviour at high tem...
This paper takes the experimental strain behaviour of three nuclear reactor type plain concretes exp...
This thesis describes an investigation of the effect of elevated temperatures upon the properties of...
This thesis deals with the problem of creep in concrete and its influence on thermal stress developm...
The paper is given of an extended model for concrete in compression at elevated temperature that inc...
This report is concerned with the analysis and its' review of a prestresied concrete reactor vessel ...
The prestressed concrete confinement vessel is the third and last barrier in French Nuclear Power Pl...
This study is a part of a French national project dealing with the mechanical behaviour of the conta...
WOS:000313369400002International audienceThe present study concerns the uniaxial compressive creep o...
WOS:000356421200006International audienceThis research concerns the uniaxial compressive basic creep...
International audienceThe objective of this research is to understand the behavior of concrete subje...
A numerical model has been developed to predict early-age cracking for massive concrete structures. ...
Based on the existing limited test data, it is possible to set up an approximate constitutive model ...
A constitutive model for the analysis of deformations of concrete subject to transient temperature a...
International audienceDifferent factors related to construction site conditions must be accounted fo...
This research has been conducted within the framework of the study of concrete behaviour at high tem...
This paper takes the experimental strain behaviour of three nuclear reactor type plain concretes exp...
This thesis describes an investigation of the effect of elevated temperatures upon the properties of...
This thesis deals with the problem of creep in concrete and its influence on thermal stress developm...
The paper is given of an extended model for concrete in compression at elevated temperature that inc...
This report is concerned with the analysis and its' review of a prestresied concrete reactor vessel ...