Extension of the Component Thermal-Hydraulics Analysis code CUPID toward Sub-channel Scale Analysis of PWR Reactor Core

Subchannel scale T/H analysis;Reactor core T/H analysis;CUPID;Validation;APR1400;622

공과대학 에너지시스템공학부 ;

University:서울대학교 대학원

关键词: Subchannel scale T/H analysis;    Reactor core T/H analysis;    CUPID;    Validation;    APR1400;    622;   

Others  :  http://s-space.snu.ac.kr/bitstream/10371/123535/1/000000141829.pdf

美国|英语

来源: Seoul National University Open Repository

【 摘 要 】

In order to improve margin management, meet new safety requirements and secure operational flexibility, it is required to apply the latest codes and methods for safety demonstration. In this context, high-fidelity and multi-physics simulation for a full core of a light water reactor with coupled T/H (Thermal-Hydraulics) and neutronics code under steady and transient conditions become a critical issue in the nuclear reactor safety analysis. Considering the computational power necessary for a full core pin-by-pin analysis, a subchannel scale T/H analysis is desired to achieve required accuracy with an endurable computational time. In this study, KAERI (Korea Atomic Energy Research Institute)’s inhouse code CUPID was applied for the subchannel scale T/H analysis for PWR reactor core. The code adopts three-dimensional two-fluid model with various closure models and incorporates a highly parallelized numerical solver. These features of CUPID would be advantageous to extend its applicability for simulation of accident condition with full core pin-by-pin modeling.In this paper, implemented models required for a subchannel scale T/H analysis are introduced and the code validation results against various flow conditions are presented. Thereafter, preliminary calculation of simplified OPR1000 reactor core is conducted. From these calculation results, full core simulation capability and performance of parallel solver are evaluated. Finally, subchannel scale analysis for the APR1400 reactor core at cycle 1 hot full power steady state is performed. Specific geometry features of APR1400 including fuel assembly, water gap, shroud and guide tube are considered in the presented demonstration. Whole core pin power distributions from the calculation result of neutronics code nTRACER are applied to CUPID for high fidelity full core T/H simulation. Following the demonstration results, coolant temperature and velocity distributions are properly simulated with CUPID. Thereafter, MDNBR (Minimum Departure from Nucleate Boiling Ratio) analysis capability of CUPID using Biasi correlation is demonstrated.From this work, CUPID shows its capability of reproducing key phenomena in a reactor core and dealing with subchannel scale whole core T/H analysis.