The sub-channel analysis code MATRA is descended from COBRA, and was originally developed by Korean Atomic Energy Research Institute (KAERI) for PWRs. In the frame of a bilateral collaboration between the Forschungszentrum Karlsruhe and KAERI, the MATRA code is being further modified for both heavy liquid metal and supercritical water cooled reactors.
- Lumped parameter approach
- Steady-state & transient analysis
- Capability for single and two-phase flows
- Applicability to various fluids
- Design of fuel assemblies of LM cooled reactors and water-cooled reactors
- Safety analysis of nuclear reactor cores
- Development of advanced nuclear reactor cores
- Under preparation
(1) Yoo, Y. J., Hwang, D. H. and Sohn, D. S., “Development of a Subchannal Analysis Code MATRA Applicable to PWRs and ALMRs,” J. of Korean Nuclear Society , Vol. 31, p. 314, 1999.
(2) Wheeler, C. L., et al., “COBRA-IV-i : An Interim Version of COBRA for Thermal-Hydraulic Analysis of Rod Bundle Nuclear Fuel Elements and Cores,” BNWL-1662, 1976.