The Westinghouse 4-Loop PWR is a 3411MWth Nuclear Power Plant (NPP). The reactor core consists of 193 fuel assemblies within the core shroud. Each fuel assembly is arranged in 17x17 arrays and includes 264 fuel rods, 24 control rod guide tubes and one instrument tube. The objective of thermal and hydrodynamic design is to safely remove of the generated heat in the fuel without producing excessive fuel temperatures or steam void formations and without approaching the critical heat flux under steady-state operating conditions. This paper presents reactor core and fuel assembly modeling of the Westinghouse 4-Loop NPP using the thermo hydraulic subchannel analysis COBRA-EN code. The results of this modeling are compared with the VIPRE-01 thermal hydraulic code. The study involves the determination of the departure from nucleate boiling ratio (DNBR) in the hot channel of the reactor core, the temperature profiles, heat flux and pressure drop across the hottest channel of the hot assemblies. The obtained results shows that the good agreements are exist between the COBRA-EN and VIPRE-01 thermal hydraulic codes.
Thermo-Hydraulic Design of Westinghouse 4-Loop Reactor Core by Cobra-EN Code
D'AURIA, FRANCESCO SAVERIO
2009-01-01
Abstract
The Westinghouse 4-Loop PWR is a 3411MWth Nuclear Power Plant (NPP). The reactor core consists of 193 fuel assemblies within the core shroud. Each fuel assembly is arranged in 17x17 arrays and includes 264 fuel rods, 24 control rod guide tubes and one instrument tube. The objective of thermal and hydrodynamic design is to safely remove of the generated heat in the fuel without producing excessive fuel temperatures or steam void formations and without approaching the critical heat flux under steady-state operating conditions. This paper presents reactor core and fuel assembly modeling of the Westinghouse 4-Loop NPP using the thermo hydraulic subchannel analysis COBRA-EN code. The results of this modeling are compared with the VIPRE-01 thermal hydraulic code. The study involves the determination of the departure from nucleate boiling ratio (DNBR) in the hot channel of the reactor core, the temperature profiles, heat flux and pressure drop across the hottest channel of the hot assemblies. The obtained results shows that the good agreements are exist between the COBRA-EN and VIPRE-01 thermal hydraulic codes.I documenti in IRIS sono protetti da copyright e tutti i diritti sono riservati, salvo diversa indicazione.
