INTERIM REPORT C03RA-III:A DIGITAL COMPUTER PROGRAM FOR STEADY STATE ANO TRANSIENT THERMAL-HYDRAULIC ANALYSIS OF ROD BUNDLE NUCLEAR FUEL ELEMENTS

摘要

The COBRA-III computer program computes the flow and enthalpy in rod-bundle nuclear fuel element subchannels during both steady state and tran¬sient conditions. It uses a mathematical model that considers both turbulent and diversion crossflow mixing between adjacent subchannels. Each subchannel is assumed to contain one-dimansional, two-phase slip-flow.nThe features of COBRA-III can be summarized as follows:nIt contains all the analysis capability of the COBRA-II program.nIt can consider transients of fast-to-intermediate speed--no sonic velocity propagation effects are considered.nThe numerical scheme performs a boundary value solution to more properly include the effects of crossflow resistance.nThe numerical solution has no stability limitation on space or time steps.