Institute for Neutron Physics and Reactor Technology (INR)

Liquid Metal Fluid-dynamics and Effect on Materials

The experimental liquid metal loops hosted within the Karlsruhe Sodium laboratory (KASOLA) comprise a set of liquid metal (LM) facilities to study LM flows for various type of energy applications ranging from room temperature for education and training and fundamental research up to challenges posed by multi-physics problems such as material-fluid interactions. Extreme conditions such as sodium boiling relevant to thermo-electric conversion or fast reactor safety are covered. The complete application range is complemented by CFD model development and validation allowing for a transfer not only on component but also on system scale. The largest facility named also KASOLA is designed to cover several thermal-hydraulic aspects in different geometries and is capable to host test sections up to 5 m with temperatures up to 550 °C. It provides a flow rate of up to 150 m³/h and allows for heat transfer experiments with an extraction of up to  400 kW provided by a sodium air heat exchanger installed in the upper part of the loop.

For education and training purposes as well as the development/qualification of novel/advanced measurement techniques or high precision heat transfer measurements in low Prandtl fluid dynamics, the DITEFA facility operated with eutectic GaInSn is available. The KASOLA framework includes three other sodium 8-shaped sodium loops (SOLTEC-1/-3), operating up to 750°C for long-term investigations of materials, sensors, welding/soldering and high temperature pumps. Prototypical heat sources as created by nuclear heating or within neutron producing accelerator targets can be replicated using IR-laser heating in pulsed and/or CW-mode, to allow for an immediate transfer to application including sodium boiling as demonstrated in the KARIFA test device to support the EU-program ESFR-SMART.

 

Publications

Oder J, Tiselj I, Jäger W, et al. Thermal fluctuations in low-Prandtl number fluid flows over a backward facing step. Nuclear Engineering and Design (2020) DOI: doi.org/10.1016/j.nucengdes.2019.110460

Jaeger W, Hering W. Investigation of Local Heat Transfer Enhancement in Generic Liquid Metal–Cooled Fuel Assemblies with Empirical Models. Nuclear Science and Engineering (2019) 193(1-2) 160-170 DOI: 10.1080/00295639.2018.1493855

Onea, A., Perez-Martin, S., Jaeger, W., Hering, W., Stieglitz, R. Advances in New Heat Transfer Fluids: From Numerical to Experimental Techniques. Liquid metals as heat transfer fluids for science and technology. Chapter 12. 305-376 2017 doi.org/10.1201/9781315368184

Bubelis E., Schikorr M.  Revised review and revised proposal for best fit of wire-wrapped fuel bundle friction factor and pressure drop predictions using various existing correlations. KIT Scientific Working Paper No. 46 (URN: nbn-resolving.org/urn:nbn:de:swb:90-534226). – KIT, Karlsruhe, Germany. 2016.

E. Bubelis, S. Perez-Martin, S. Passerini, et al. IAEA NAPRO Coordinated Research Project: Heat Transfer and Pressure Drop Correlations for Sodium Cooled Systems. Proceedings of the 2016 International Congress on Advances in Nuclear Power Plants – ICAPP 2016, 17-20 April, 2016. – San Francisco, CA, USA. 2016. P. 1608-1617.

Schikorr, M., Bubelis, E.et al. Proposal for pressure drop prediction for a fuel bundle with grid spacers using Rehme pressure drop correlations. Nuclear Engineering and Design Volume 240, Issue 7, July 2010, Pages 1830-1842  doi.org/10.1016/j.nucengdes.2010.03.039

Bubelis, E., Schikorr, M. Review and proposal for best fit of wire-wrapped fuel bundle friction factor and pressure drop predictions using various existing correlations. Nuclear Engineering and Design Volume 238, Issue 12, December 2008, Pages 3299-3320 doi.org/10.1016/j.nucengdes.2008.06.024