Thermal hydraulics for new nuclear systems

The nuclear reactor design has gone through four generations. Compared to the second generation (Gen II), Gen III ensures a higher safety level. However, Gen III nuclear reactors do not meet the requirements of a longer-term nuclear power development.

To power a safe and sustainable nuclear future, a fourth generation of reactors has been proposed that includes very high-temperature reactors, gas-cooled fast reactors and supercritical water-cooled reactors. A common component of Gen IV reactors is thermal hydraulics.

In spite of the difference in coolants and flow channel structures, the EU-funded project THINS (Thermal-hydraulics of innovative nuclear systems) identified five cross-cutting issues of thermal hydraulics. This collaboration, spanning 24 institutions, focused on core thermal hydraulics, single-phase mixed convection and turbulence, multiphase flow and code coupling.

The overall objective of the THINS project was the development and validation of computational and experimental methodologies for studying these thermal hydraulics phenomena. Efforts were also made to apply the scientific results for knowledge-sharing purposes.

Specifically, THINS partners verified simulation tools of coolant flows within the reactor core components and calculated pressure drop and heat transfer in tube bundles and spacer grids. Using turbulence models, they also modelled the lead-bismuth eutectic rod bundle flow.

Project researchers set up a comprehensive database containing the results of direct numerical simulations along with experimental data. This was used to characterise flow phenomena such as convection patterns, thermal stratification and fluid-structure thermal exchange in reactors.

Heat transfer and flow mixing were analysed in single-phase reactor cooling systems. THINS researchers developed new modelling approaches to describe the effects of buoyancy and non-isotropic turbulence accurately for a wide range of Prandtl numbers.

Existing models were further developed and validated for studying multiphase flows in innovative reactor systems. The flow phenomena studied were free surface flows in pool-type liquid-metal reactors and interactions between water and heavy liquid metal.

The THINS team developed and validated new code-coupling solutions for reliably predicting transient multi-scale thermal hydraulic phenomena in high- and very high-temperature reactor systems. In particular, they investigated graphite dust transported in the coolant loop.

Training of young nuclear engineers and researchers complemented the THINS project activities. Using the scientific results for teaching purposes proved highly effective in strengthening the basis for maintaining and extending know-how in the field.