Design study of an advanced loop type sodium cooled fast reactor (JSFR) has been performed in Japan. A column type upper internal structure (UIS) is installed in the upper plenum of reactor vessel in JSFR. This UIS is composed of control rod guide tubes and several perforated horizontal plates. Further, each horizontal plate has a radial slit for operation of the fuel handling system. This unique structure of the UIS permits existence of the fluids from the fuel assembly outlets in the UIS. The maximum temperature difference between cold fluid and hot fluid is approximately 100 K as a tentative condition. Therefore, high cycle thermal fatigue may occur at the bottom plate (CIP) of the UIS where the hot sodium from the fuel subassembly can mix with the cold sodium from the control rod channel and the blanket fuel subassembly. We have been conducted a water experiment using a reactor upper plenum model to grasp the thermal-hydraulic phenomena around control rod (CR) channels, and radial blanket assemblies and to obtain countermeasures for significant temperature fluctuation on the CIP. The experimental apparatus has 1/3 scale and 60 degree sector model of the reactor upper plenum. The temperatures around the CR channels and the blanket assemblies were measured with thermocouples. By the experiment, characteristics of fluid temperature fluctuation between the handling head of the assemblies and the CIP are measured and countermeasure, for the significant temperature fluctuation generation will be discussed on the influence of the distance from the handling head outlet to the lower surface of the CIP. It was confirmed that temperature fluctuations could be mitigated by expanding a distance between a handling head and CIP.
