Sodium-cooled fast-neutron reactors (SFR) are currently considered to be the most mature type of reactor able to optimize uranium ore usage and reduce nuclear waste produced from Generation II and III reactors. CEA led studies up to 2019 on the features of a 600MWe reactor within the frame of the Advanced Sodium Technological Reactor for Industrial Demonstration (ASTRID) project. The chosen pool-type design offers the advantage of containing the primary sodium within a single vessel, ensuring safer operations by transferring heat to the secondary sodium circuit via Intermediate Heat Exchangers (IHX). This design eliminates the risk of water/primary sodium interaction. A tertiary loop then generates steam for power conversion. Given the safety implications of the design, careful study of the vessel's geometry is essential, particularly the IHX, which plays a critical role in heat exchange. To investigate the flow dynamics within the vessel, a scaled-down model of the ASTRID reactor was constructed. Using a similarity approach water was used as a simulant fluid due to the complexity and cost of sodium-based experiments. This model allows for adjustments in IHX geometry to conduct parametric studies on flow behavior. Particle Image Velocimetry (PIV) was employed to measure velocity near the IHX inlet across different configurations. The results align with previous studies, indicating that, whatever the configuration, flow is concentrated in the lower section of the IHX, offering valuable insights for future design improvements.
Guenadou et al. (2026) studied this question.