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Session Chair: Tomasz Kwiatkowski, National Centre for Nuclear Research NCBJ, Poland
Facilities to investigate sodium and materials behavior up sodium boiling
Wolfgang Hering, Alexandru Onea, Angela Jianu, Joachim Fuchs, Th. Schaub, Alfons Weisenburger, Swen Ulrich, Robert Stieglitz
Karlsruhe Institute of Technology (KIT)
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.
All experiments provide data for qualification of accident tolerant instrumentation and/or passive shut down systems usable for sodium fast reactors.
Towards the accurate prediction of flow and heat transfer in a tightly spaced bare rod bundle configuration
Tomasz Kwiatkowski1, Afaque Shams2
1National Centre for Nuclear Research (NCBJ); 2Nuclear Research and Consultancy Group (NRG)
The understanding of flow behaviour and temperature distribution in rod bundles during operation is of great importance for the safety and economical design of existing nuclear reactors and as well as new nuclear technologies. In order to study the thermal-hydraulics phenomena in the fuel assembly geometries, researchers all over the world are increasingly using Computational Fluid Dynamics (CFD) as a reliable research tool. In this regard, an extensive effort, between NCBJ and NRG, has been put forward to assess the prediction capabilities of available Reynolds Average Navier-Stokes (RANS) models. Accordingly, a research program has been formed and is based on three main steps. As a first step, a wide range of RANS and unsteady RANS computations has been performed to design a numerical experiment for a closely-spaced bare rod bundle in order to perform a direct numerical simulation (DNS), which will serve as a reference for the validation purpose. As a second step, DNS of this bare rod bundle has been performed for three different Prandtl number (Pr) fluids, i.e. Pr = 2, 1 and 0.025. This DNS has been performed using a higher-order spectral element code and consists of 660 million grid points. Accordingly, an extensive database has been generated for the validation purpose. Finally, this database is used to assess the prediction capabilities of different linear and non-linear RANS models. This article will provide a good overview of all three steps and will also highlight the main lessons learned from this research.
CFD validation of heavy liquid metal thermal-hydraulic in bare rod bundles
Abdalla Batta, Andreas Class
Karlsruher Institut für Technologie (KIT)
Computational Fluid Dynamic (CFD) validation of thermal hydraulic in rod bundles is essential for the design of fuel assemblies and for the safety analysis of nuclear reactors. One of the main objectives of European SESAME project (Thermal-hydraulic Simulation and Experiments for the Safety Assessment of MEtal cooled reactors) is the development and validation of advanced numerical approaches for liquid metal fast reactors. In this study, work related to bare bundles in the SESAME work package II is presented. The selected fuel pin bundle configuration is relevant for the ALFRED’s core thermal-hydraulic design. ALFRED is an Advanced Lead Fast Reactor European Demonstrator with a flexible fast spectrum of 300 MWth.
Experimental data and direct numerical simulation (DNS) results are used for the validation of RANS Model. The experimental data for fuel pin simulator test section are generated by using the facility NACIA-UP (heavy liquid metal cooled loop) located at ENEA, Italy. Several cases and flow conditions are considered. DNS data are generated for three cases by UNIMORE in Italy. At KIT (Karlsruhe Institute of Technology) RANS data are generated for the nominal cases selected from the DNS study by the University of Modena (UNIMORE). The considered cases represent flow of liquid metal in infinite heated FPBS (Fuel Pin Bundle Simulator). The comparisons of computed Nusselt number from RANS results to the experimental and DNS results show a very good agreement.
SBLOCA analysis of SMART with Trans-Critical CO2 power conversion system for Maritime Propulsion
Bong Seong Oh, Jeong Ik Lee
Korea Advanced Institute of Science and Technology
The Northern Sea Route is a new way to connect Asia and Europe with reduced sailing distance. In order to utilize the sea route merchant ships are required to have icebreaking capability. Since conventional fossil fuel powered engines has limitations to be used in an icebreaking merchant ship propulsion, nuclear power having long refueling period for icebreaking merchant ship is proposed. To evaluate technical feasibility of the nuclear power for icebreaking ship’s engine 330MWth SMART reactor is selected for the reference reactor core, which is proven to achieve high degree of safety. The power conversion system for the SMART reactor core is substituted with the trans-critical CO2 recompression cycle that has advantages in low temperature heat sink as in Northern Sea Route and low temperature heat source as in Pressurized Water Reactor conditions instead of the conventional steam Rankine cycle. To guarantee the safety of SMART with T-CO2 cycle at hypothetical accident condition, a passive residual heat removal system that is driven by CO2 natural circulation is designed in this work. Then, SBLOCA of SMART with T-CO2 is modeled to demonstrate whether the designed CO2 PRHRS mitigates the initial event.