Framatome ANP develops a new boiling water reactor called SWR 1000. For the case of a hypothetical core melt accident it is designed in such a way that the core melt is retained in the Reactor Pressure Vessel (RPV) at low pressure owing to cooling of the RPV exterior and high reliable depressurization devices. Framatome ANP performs — in co-operation with VTT — tests to quantify the safety margins of the exterior cooling concept for the SWR 1000, for determining the limits to avoid the critical heat fluxes (CHFs). The three step procedure has been set up to investigate the phenomenon: 1. Water-air study for a 1:10 scaled global model, with the aim to investigate the global flow conditions. 2. Water-air study for a 1:10 scaled, 10% sector model, with the aim to find a flow sector with almost similar flow conditions as in the global model. 3. Final CHF experiments for a 1:1-scaled, 10% sector., the boarders of this model have been selected based on the first two steps. The instrumentation for the water/air experiments included velocity profiles, the vertically averaged average void fraction and void fraction profiles in selected positions. The experimental results from the air-water experiments have been analyzed at VTT using the Fluent-4.5.2 code with its Eulerian multiphase flow modeling capability. The aim of the calculations was to learn how to model complex two-phase flow conditions. The structural mesh required by Fluent-4 is a strong limitation in the complex geometry, but modeling of the 1/4 sector from the facility was possible, when the GAMBIT pre-processor was used for the mesh generation. The experiments were analyzed with the 150 × 150 × 18 grid for the geometry. In the analysis the fluid viscosity was the main dials for adjusting the vertical liquid velocity profiles and the bubble diameter for adjusting the phase separation. The viscosity ranged between 1 to 10000 times the molecular viscosity, and bubble diameter between 3 to 100 mm, when the calculation results were adjusted for a good agreement with the experimental data. The analysis results were very valuable for designing the final water/steam facility for final CHF tests. The validation against data from the air-water experiments proved that the present CFD codes approach to the state where they can be used for simulating such two-phase experiments, where the fraction of both phases is essential and the flow is strongly affected by the density differences. It is still too early to predict, if the CFD calculation of the 1:1 scale critical heat flux experiments is successful, could the result be used for formulating a new type of a critical heat flux correlation, where the effects of CRD’s on the flow patterns and gap dimensions are model parameters.
- Nuclear Engineering Division
CFD Analyses for Water-Air Flow With the Euler-Euler Two-Phase Model in the Fluent4 CFD Code
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Miettinen, J, & Schmidt, H. "CFD Analyses for Water-Air Flow With the Euler-Euler Two-Phase Model in the Fluent4 CFD Code." Proceedings of the 10th International Conference on Nuclear Engineering. 10th International Conference on Nuclear Engineering, Volume 3. Arlington, Virginia, USA. April 14–18, 2002. pp. 653-663. ASME. https://doi.org/10.1115/ICONE10-22419
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