Application of computational fluid dynamics in nuclear reactor safety analysis research

The last decade has seen an increasing use of three-dimensional CFD codes to predict steady state and transient flows in nuclear reactors because a number of important phenomena such as pressurized thermal shocks, coolant mixing, and thermal striping cannot be predicted by traditional one-dimensional system codes with the required accuracy and spatial resolution. CFD codes contain models for simulating turbulence, heat transfer, multi-phase flows, and chemical reactions. For this reason the long-term objective of the activities of the Helmholtz-Zentrum Dresden-Rossendorf Germany (HZDR) R&D program lies in the development of theoretical models for basic phenomena of transient, three-dimensional single and multiphase systems. Such models must be validated before they can be used with sufficient confidence in nuclear reactor safety (NRS) applications. The necessary validation is performed by comparing model results against measured data. However, in order to obtain a reliable model assessment, CFD simulations for validation purposes must satisfy strict quality criteria given in the Best Practice Guidelines (BPG). The following topical issues which are related to PWR, where CFD calculations have been performed, will be briefly discussed: Coolant mixing of the primary circuit, horizontal stratified flow phenomena in the hot leg and fibre material transport in a core under loss of coolant conditions.