Investigation into Thermal-hydraulic Behavior in the KALIMER-600 Pool in a Steady State

The conceptual design of KALIMER-600 (Korea Advanced Liquid MEtal Reactor) has been developed under the Mid- and Long-term Nuclear R and D Program. In the design of those nuclear reactors securing a reliable decay heat removal system and evaluating its capability is an important safety criterion. Especially various studies are necessary in order to assess the major parameters affecting the transient decay heat removal performances. One of the approaches for those studies is numerical analysis based on 3D thermal hydraulic codes such as COMMIX. COMMIX-1AR/P code is a three-dimensional transient single-phase computer program for a thermal-hydraulic analysis. Especially the porous-medium formulation which has been implemented in the COMMIX code allows the user to simulate a complex geometrical arrangement as well as a simple one. As the first step of an investigation into the transient behavior in KALIMER-600 after the shut down, the modeling approaches were introduced and the thermal-hydraulic behavior of KALIMER-600 in a steady state condition was investigated briefly. This study will provide the basic design information for modeling KALIMER-600 and the thermal-hydraulic behavior of it in a steady state condition