Development of a computer program named KAIS-KORI to simulate a three-dimensional PWR core

Abstract

A computer code, named KAIS-KORI for simulating the reactor core, has been developed on the basis of the transient code TWINKLE. KAIS-KORI solves the three-dimensional two-group time dependent diffusion equations using the implicit finite-difference technique. The code containes a detailed six region fuel-clad-coolant transient heat transfer model at each spatial mesh for calculating doppler and moderator feedback effects. The mesh scheme is structured in KAIS-KORI with one mesh per assembly and 12 axial mesh for fuel region in octant core. KAIS-KORI colculates assembly-wise power, axial offset, point-wise temperature distribution for the steady state and/or transient state. Moderator and doppler coefficients, control rod and boron worths can also be predicted with the present code. Comparison between the measured values and the predicted values of the present computer code for cycle-1 of KNU1 shows that there is relatively a large difference depending on the local region. However, the predicted results are in good agreement with the general trend of the transient phenomena of the core.