BAGIRA - a thermal-hydraulic program complex was primarily developed for using it in nuclear power plant simulator models, but is also used as a best-estimate analytical tool for modeling two-phase mixture flows. The code models allow consideration of phase transients and the treatment of the hydrodynamic behavior of boiling and pressurized water reactor circuits. It provides the capability to explicitly model three-dimensional flow regimes in various regions of the primary and secondary circuits such as, the mixing regions, circular downcomer, pressurizer, reactor core, main primary loops, the steam generators, the separator-reheaters. In addition, it is coupled to a severe-accident module allowing the analysis of core degradation and fuel...
One of the key safety issues currently addressed by the Pressurized Water Reactor (PWR) industry is ...
The present work deals with the application of single-phase and two-phase Computational Fluid Dynami...
This thesis focuses on the validation of the coupled codes developed in Finland for the safety analy...
In this paper we present recent results of the application of the thermal-hydraulic code BAGIRA to t...
Angra 2, the second Brazilian nuclear power plant, began the commercial operation in 2001. The plant...
The evaluation model and computational capabilities required for engineering design and safety analy...
Thermal hydraulic modelling is needed for describing processes that use water for cooling. Simulator...
Three-dimensional computer simulation and analyses of the horizontal steam generator thermal-hydraul...
Thermal hydraulics in nuclear engineering including multifluid, multiphase flow phenomena is indispe...
Recently, three-dimensional neutron-kinetics core models have been coupled to advanced thermal-hydra...
The design and analysis of the thermal/hydraulic systems of nuclear power plants necessitates system...
The reactor dynamics codes for transient and accident analyses inherently include the coupling of ne...
An attempt is made in the Book to provide a vision of nuclear thermal-hydraulics as it appears follo...
In the last four decades, large efforts have been undertaken to provide reliable thermal-hydraulic s...
PORFLO is a 3D two-phase flow simulation code which utilizes porous medium approach in the modeling ...
One of the key safety issues currently addressed by the Pressurized Water Reactor (PWR) industry is ...
The present work deals with the application of single-phase and two-phase Computational Fluid Dynami...
This thesis focuses on the validation of the coupled codes developed in Finland for the safety analy...
In this paper we present recent results of the application of the thermal-hydraulic code BAGIRA to t...
Angra 2, the second Brazilian nuclear power plant, began the commercial operation in 2001. The plant...
The evaluation model and computational capabilities required for engineering design and safety analy...
Thermal hydraulic modelling is needed for describing processes that use water for cooling. Simulator...
Three-dimensional computer simulation and analyses of the horizontal steam generator thermal-hydraul...
Thermal hydraulics in nuclear engineering including multifluid, multiphase flow phenomena is indispe...
Recently, three-dimensional neutron-kinetics core models have been coupled to advanced thermal-hydra...
The design and analysis of the thermal/hydraulic systems of nuclear power plants necessitates system...
The reactor dynamics codes for transient and accident analyses inherently include the coupling of ne...
An attempt is made in the Book to provide a vision of nuclear thermal-hydraulics as it appears follo...
In the last four decades, large efforts have been undertaken to provide reliable thermal-hydraulic s...
PORFLO is a 3D two-phase flow simulation code which utilizes porous medium approach in the modeling ...
One of the key safety issues currently addressed by the Pressurized Water Reactor (PWR) industry is ...
The present work deals with the application of single-phase and two-phase Computational Fluid Dynami...
This thesis focuses on the validation of the coupled codes developed in Finland for the safety analy...