Nuclear thermal hydraulic analysis using coupled CFD and system codes

Abstract

The thermal hydraulic analysis of nuclear reactors is largely performed by what are known as system codes. These codes predict the flows in the complex network of pipes, pumps, vessels and heat exchangers that together form the thermal hydraulic systems of a nuclear reactor. These codes have been used for many decades and are now very well established. Given this long process of refinement, they are able to produce remarkably accurate predictions of plant behaviour under both steady and transient conditions. Modern CFD is able to produce high quality predictions of flows in complex geometries, but only with the use of large computing resources. It would be impractical to build a CFD model of, for example, the entire primary circuit of a PWR. However, it is possible to model with adequate fidelity much of the primary circuit using a cheaper one-dimensional system code, and it may only be in a limited part of the circuit that full three-dimensional effects are important. A coupling scheme was developed to couple the CFD software STAR-CCM+ and the system code RELAP5-3D. The structure of the scheme is presented, together with validations for single phase flow in smooth pipes in both transient and steady state cases. Attention was also given to the problem of reconstructing the flow profile at the inlet of the CFD model under the hypothesis of fully developed flow. This problem arises when flow data has to be passed from the one-dimensional system code to the three-dimensional CFD software. The coupling scheme was then modified to be able to perform multiphase simulations. The PWR subchannel and bundle test (PSBT) benchmark was used to validate the multiphase coupling methodology.