ASSESSMENT OF THE 3-D THERMAL-HYDRAULIC NUCLEAR CORE COMPUTER CODE FLICA-IV ON ROD BUNDLE EXPERIMENTS

摘要

The physical validation compared with the hydrau- lic and two-phase flow experiments of the thermal- hydraulic FLICA-IV nuclear core computer code, in the case of a pressurized water reactor is presented. This three-dimensional two-phase flow code is devoted to steady state and transient thermal-hydraulic analysis of nuclear reactor cores. The four balance equations used by the code and the closure relationships are first pre- sented.