Application of CUPID for subchannel-scale thermal-hydraulic analysis of pressurized water reactor core under single-phase conditions

• Published in: Nuclear engineering and technology Vol. 50; no. 1; pp. 54 - 67
• Main Authors: Yoon, Seok Jong, Kim, Seul Been, Park, Goon Cherl, Yoon, Han Young, Cho, Hyoung Kyu
• Format: Journal Article
• Language: Korean
• Published: 2018
• Subjects: Validation; Reactor Core Thermale-Hydraulic Analysis; Analysis; Subchannel-Scale Thermale-Hydraulic; Advanced Power Reactor 1400; CUPID
• ISSN: 1738-5733; 2234-358X

There have been recent efforts to establish methods for high-fidelity and multi-physics simulation with coupled thermal-hydraulic (T/H) and neutronics codes for the entire core of a light water reactor under accident conditions. Considering the computing power necessary for a pin-by-pin analysis of the entire core, subchannel-scale T/H analysis is considered appropriate to achieve acceptable accuracy in an optimal computational time. In the present study, the applicability of in-house code CUPID of the Korea Atomic Energy Research Institute was extended to the subchannel-scale T/H analysis. CUPID is a component-scale T/H analysis code, which uses three-dimensional two-fluid models with various closure models and incorporates a highly parallelized numerical solver. In this study, key models required for a subchannel-scale T/H analysis were implemented in CUPID. Afterward, the code was validated against four subchannel experiments under unheated and heated single-phase incompressible flow conditions. Thereafter, a subchannel-scale T/H analysis of the entire core for an Advanced Power Reactor 1400 reactor core was carried out. For the high-fidelity simulation, detailed geometrical features and individual rod power distributions were considered in this demonstration. In this study, CUPID shows its capability of reproducing key phenomena in a subchannel and dealing with the subchannel-scale whole core T/H analysis.