Coupled CFD-STH analysis of liquid metal flows

Generation IV pool-type lead-cooled fast nuclear reactors are characterized by multi-scale/multi-dimensional flow and heat transfer phenomena. Commonly, the design and safety analysis of nuclear systems relies on fast-running 1D system thermal-hydraulics (STH) codes that have limited physics and geometry modeling capabilities. High-fidelity flexible 3D computational fluid dynamics (CFD) codes, on the other hand, are computationally too expensive to resolve very large systems with necessary resolution. Code coupling makes it possible to is improve the efficiency of such calculations while maintaining necessary accuracy.

This presentation describes the development and qualification of coupled RELAP5 and STAR-CCM+®. An overview of the implementation of a semi-implicit domain overlapping coupling scheme is presented. Selection of exchanged variables, coupling interfaces, timing and other relevant specifics are discussed in details.

Application of verification, validation and uncertainty quantification (VVUQ) is a pre-requisite for the numerical tools to be used in nuclear reactor safety field. In the presentation, VVUQ of standalone STAR-CCM+ and coupled models is carried out using experimental results from TALL-3D facility – a liquid Lead-Bismuth Eutectic (LBE) loop designed to produce exhaustive high-quality data for the calibration and validation of coupled STH-CFD codes.

Improvement of accuracy of coupled simulation results compared to standalone STH modeling is demonstrated.

Session Time Slot(s): 
06/03/2017 - 5:05pm-06/03/2017 - 5:30pm
Room II
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KTH Royal Institute of Technology
First Name: 
Last Name: 
Company / Institution: 
KTH Royal Institute of Technology