Development of Reactor Vessel Thermal-hydraulic Analysis Method in Natural Circulation Conditions

-Investigation of Thermal-hydraulic Analysis Model for Interwrapper Gap between Assemblies-

Abstract

To enhance a safety of sodium-cooled fast reactors (SFRs), decay heat removal systems under natural circulation (NC) with a dipped-type direct heat exchanger (D-DHX) installed in an upper plenum of a reactor vessel (RV) have been investigated. During the D-DHX operation, the flow of the coolant at low temperature from the D-DHX into assemblies and an interwrapper gap (IWG) between them, and the radial heat transfer through a wrapper tube and the IWG among assemblies occur. Such phenomena in the core can remove the decay heat without external electric power supply. In terms of the design study, modeling of an RV using a CFD code (RV-CFD) with a coarse mesh arrangement has an advantage in a reduced computational cost. In this study, focused on the modeling of the IWG, to achieve a lower computational cost while maintaining the prediction accuracy, an influence of combination of the mesh number in the IWG and the pressure loss correlation on the core temperature distribution was investigated through the numerical analysis of a sodium experimental apparatus named PLANDTL-1. The result shows the coarse mesh with correlation reduced the IWF, or a circulation flow with the upward and downward flow in the IWG, and shifted the temperature distribution in the core to the high-temperature side.