Thermal Hydraulics’ Study in the Original Design of TRIGA 2000 Reactor
National energy policy which aims at new energy exploitation, such as nuclear energy is including many efforts to increase the safety reactor core condition and optimize the related aspects and the ability to build new research reactor with properly design. One of safety factor that need to be analyzed is thermal hydraulics characteristics for example velocity and temperature arround reactor core. Due to existance of masive components in the reactor core, the complexity of natural convection fluid flow characteristic in reactor core was difficult to be done in experimental analysis. Developing technology has modified the difficulties in analysis method become eassier. According to the problem, analysis had been carried out using the software. Simulation and numerical calculation were done in the Computational Fluid Dynamics package.This analysis shows the fluid flow vector movement, number of fluid flow velocity and its variety effect to different area in TRIFA 2000 reactor. Result of this analysis can be used for measuring data and simulation collection of coolant flow in TRIGA 2000 reactor.
thermal hydraulics fluid flow convection computational fluid dynamics
Rosalina Fiantini Efrizon Umar Damawidjaya Biksono
Graduate Student of Jenderal Achmad Yani University Jl.Ters.Jend.Sudirman PO BOX 148, Cimahi, Indone National Nuclear Energy Agency of Indonesia Jalan Tamansari 71, Bandung 40132, Indonesia Jenderal Achmad Yani University Jl.Ters.Jend.Sudirman PO BOX 148, Cimahi, Indonesia
国际会议
上海
英文
1977-1986
2010-10-10（万方平台首次上网日期，不代表论文的发表时间）