Abstract
Many research reactors with plate-type fuel are operated worldwide and the thermal hydraulic core analysis is done with dedicated codes or with 1D system thermal hydraulic codes. At the Karlsruhe Institute of Technology (KIT), investigations are underway to develop advanced coupled neutronic/thermal hydraulic simulation tools for the Safety analysis of research reactors with plate-type fuel. For this purpose, the subchannel thermal hydraulic code SubChanFlow was extended and validated using relevant tests. The SubChanFlow extension consists of the addition of a heat conduction module for a thin plate and the implementation of heat transfer correlations for rectangular channels. For the validation of the modified SubChanFlow code, temperature data from RA-6 experiment obtained in a Reynolds range of 7×103<Re<1.4×104 were used.The RA-6 experiment was carried out at the Bariloche Atomic Center to support the development of MTR research reactors. Different heat transfer correlations e.g., the Colburn, Dittus-Boelter, Gnielinski, Sieder-Tate, and Y-Sudo were used to simulate the RA-6 tests. The comparison of the plate temperatures measured along the active height with the predicted values using the extended SubChanFlow code shows a deviation ranging from ± 4 °C to ± 18 °C for the Colburn and Y-Sudo correlations, respectively, which is promising compared to results of other codes.
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