CFD analysis on the test section of SCWR‐FQT demo loop, Numerical Analysis on the Effect of Wrapped Wire Spacers on Thermal Hydraulics in a Four Rod Fuel Bundle

B. Mervay (Mr), BME NTI

26th Symposium of AER on VVER Reactor Physics and Reactor Safety (2016, Helsinki, Finland)
Nuclear applications of computational fluid dynamics

Abstract

CFD ANALYSIS ON THE TEST SECTION OF SCWR-FQT DEMO LOOPBence Mervay, Attila KissBudapest University of Technology and Economics, Institute of Nuclear Techniques, 1111 Budapest, Muegyetem rkp. 9. R.317/7a, kissa@reak.bme.huABSTRACTThe CFD analysis presented in this paper is connected to the SCWR-FQT project funded by the EU that was finished in 2014 and aimed at the design and the licensing of a test loop that demonstrates the viability of the fuel bundle model of the European SCWR in real (radioactive) environment. The goal of the demonstration was to prove the compliance of the material of the fuel rod’s cladding (whether it is able to withstand the chemically aggressive, high pressure and temperature supercritical water) and the thermohydraulic adequacy of the fuel bundle geometry with helical spacers. The BME NTI participated in the thermohydraulic design of the test rod bundle. One of the Institute’s tasks was to participate in a blind CFD benchmark for which the measurement results were provided by one of the Chinese partners. The measurements were conducted in the previously built SWAMUP test loop in which the test section was placed, that was actually the SCWR-FQT fuel bundle geometry, scaled up by a factor of 1.25. Due to the relatively complex geometry the numerical grid is mainly unstructured tetrahedral grid. The heat conduction was modelled in the solid domains. A mesh sensitivity study was conducted as well as a turbulence model sensitivity study and a boundary layer sensitivity study.After the completion of the project and thus the benchmark the Institute carried on dealing with the problem, we tried to improve our model. We also examined different geometries that differed in the number of revolution of the helical spacers. In this way we could examine the thermohydraulic differences between the geometries with different number of revolution. 26th Symposium of AER on VVER Reactor Physics and Reactor Safety 49 10 – 14 October 2016, Helsinki, Finland

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