The model subassembly of the BREST-type reactor core consists of a pin bundle of square arrangement. In this bundle there are two zones which differ in the pin diameter and heat production. The model pin bundle contains one spacer grid which is located near the mid-plane of the rod bundle geometry. The working is a eutectic alloy of 22% sodium (Na) plus 78% potassium (K). Three kinds of experiments were performed to observe the thermal and hydraulic behavior of the liquid metal coolant in the BREST core simulator. Results were obtained for the coolant exit temperature distribution, central measuring pin simulator external surface temperature distribution, and coolant velocity distribution over the perimeter of the measuring pin simulator. The experiments were performed five times with increasing pin power ratios. Analysis was performed on the model subassembly of the BREST-type reactor core using a subchannel analysis code MATRA and a computational fluid dynamics code CFX. Calculational results were compared against the experimental data. The experiment revealed that the temperature rise was strongly dependent upon the geometry of the pin simulator. In contrast to the experimental results, the MATRA results were mainly dependent upon the thermal and hydraulic conditions. It was concluded that MATRA requires modifications for the pressure drop correlations that were considered inappropriate for accurately simulating the coolant behavior near the BREST-type grid spacer. Hand calculations were additionally carried out under different assumptions to determine the coolant exit temperature distribution in the pin simulator. First, the hand calculation was performed to find the coolant exit temperature distribution assuming that there is no momentum or energy transfer between subchannels. Second, an assumption was made that the coolant mixing in the subchannel assembly took place instantaneously and the pressure was equilibrated at the channel exit. Since MATRA is based on a lumped parameter model, it only calculates the subchannel averaged velocity values. Here, CFX based on the finite volume method was utilized to calculate the velocity fields over the perimeter. Results from the experiment and CFX were averaged in each subchannel region so as to check on the tendency. The CFX analysis showed reasonable results which can be improved by imposing more detailed geometry accounting for the angle of the inclination of the grid spacer.
Solution to a Standard Thermal Hydraulics Problem in a Liquid Metal Subassembly
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Son, HM, & Suh, KY. "Solution to a Standard Thermal Hydraulics Problem in a Liquid Metal Subassembly." Proceedings of the 14th International Conference on Nuclear Engineering. Volume 2: Thermal Hydraulics. Miami, Florida, USA. July 17–20, 2006. pp. 159-167. ASME. https://doi.org/10.1115/ICONE14-89188
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