In this study a code based method has been developed for calculation of integral and differential control rod worth in terms of burn-up for a WWER-1000 reactor. Parallel processing of WIMSD-5B, PARCS V2.7 and COBRA-EN has been used for this purpose. WIMSD-5B has been used for cell calculation and handling burn-up of core at different days. PARCS V2.7 has been used for neutronic calculation of core and critical boron concentration search. Thermal-hydraulic calculation has been performed by COBRA-EN. A Parallel processing algorithm has been developed by MATLAB to couple and transfer suitable data between these codes in each step. Steady-State Power Picking Factors (PPFs) of the core and Control rod worth have been calculated from Beginning Of...
International audienceThis paper discusses the calculation of the IAEA benchmark entitled In-core fu...
The rod ejection accident is defined as the postulated rupture of a control rod drive mechanism hous...
This paper describes the neutron transport calculations for the sodium-cooled fast reactors by using...
In this study a code based method has been developed for calculation of integral and differential co...
One of the main problems relating to operation of a nuclear reactor is its safety and controlling sy...
One of the main issues of a nuclear reactor is safety and controlling system. This system is designe...
A full scope thermal-neutronic couple has been developed to analyze WWER-1000 nuclear reactor core b...
A whole-core depletion analysis of OPR-1000 using MCS coupled with FRAPCON is presented in this pape...
The results of calculations of power densities and worths for control rods which use boron for neutr...
This thesis analyzes the behavior of the nuclear fuel of a WWER-1000 reactor during a Rod Ejection A...
An innovative method of combining the well-known computer codes WIMS and CITATION in an interactive ...
WOS:000346633500003This paper aims to investigate Th-232/U-233 fuel cycles in a VVER-1000 reactor th...
The developed model of the WWER-1000 reactor using MCNP6.2 (Monte Carlo N-Particle Transport Code) i...
International audienceNuclear reactors exhibit excess reactivity at start-up to ensure continuous op...
The reactor core system consists of many materials, involving multi-physics processes, and can be an...
International audienceThis paper discusses the calculation of the IAEA benchmark entitled In-core fu...
The rod ejection accident is defined as the postulated rupture of a control rod drive mechanism hous...
This paper describes the neutron transport calculations for the sodium-cooled fast reactors by using...
In this study a code based method has been developed for calculation of integral and differential co...
One of the main problems relating to operation of a nuclear reactor is its safety and controlling sy...
One of the main issues of a nuclear reactor is safety and controlling system. This system is designe...
A full scope thermal-neutronic couple has been developed to analyze WWER-1000 nuclear reactor core b...
A whole-core depletion analysis of OPR-1000 using MCS coupled with FRAPCON is presented in this pape...
The results of calculations of power densities and worths for control rods which use boron for neutr...
This thesis analyzes the behavior of the nuclear fuel of a WWER-1000 reactor during a Rod Ejection A...
An innovative method of combining the well-known computer codes WIMS and CITATION in an interactive ...
WOS:000346633500003This paper aims to investigate Th-232/U-233 fuel cycles in a VVER-1000 reactor th...
The developed model of the WWER-1000 reactor using MCNP6.2 (Monte Carlo N-Particle Transport Code) i...
International audienceNuclear reactors exhibit excess reactivity at start-up to ensure continuous op...
The reactor core system consists of many materials, involving multi-physics processes, and can be an...
International audienceThis paper discusses the calculation of the IAEA benchmark entitled In-core fu...
The rod ejection accident is defined as the postulated rupture of a control rod drive mechanism hous...
This paper describes the neutron transport calculations for the sodium-cooled fast reactors by using...