The thermal-hydraulic performance of a nuclear reactor fuel assembly grid spacer is predicted using computational fluid dynamics. The modeled flow domain exploits the periodicity of the spacer and is separated into a bare bundle and grid region to maintain a manageable model size. An iterative process is used to couple the segregated flow domains to arrive at a converged solution. The grid spacer is a 7 × 7 mixing vane grid representative of an actual pressurized water reactor grid. Pressure drop and rod wall temperature predictions for steady-state operation are computed. The results show excellent agreement with experimental data. The agreement in these results demonstrates the usefulness of the method presented as a design tool for nuclear fuel manufacturers and as a prediction tool for off-design operating conditions such as simulated accident scenarios.
All Science Journal Classification (ASJC) codes
- Nuclear and High Energy Physics
- Nuclear Energy and Engineering
- Condensed Matter Physics