ICONE19-43572 AN ANALYSIS OF A SUBCOOLED BOILING FLOW TEST BY USING THE CUPID CODE

Abstract

A thermal-hydraulic code, named CUPID, has been being developed for the realistic analysis of transient two-phase flows in nuclear reactor components. The CUPID code is a meso-scale code, and it is designed to have features for both open and porous media, distinguishing the CUPID from the three-dimensional components of system codes. This paper presents the implementation of the subcooled boiling model into the CUPID code and some assessment results. Thus, it becomes a first step is for the expansion of the application area of the CUPID code to the finer scaled two phase flow.