Thermohydraulic analyses with the AQUA code for basic design of FBR large-scale sodium test facility(I); Analyses for thermal stratification and reactor vessel cooling system

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Study of thermal hydraulics for Demonstration Fast Breeder Reactors (DFBRs) and Commercial FBRs has been planned to confirm the design and acceptability. A one-third sector model of the reactor vessel was proposed for basic design concept of the test facility. Numerical analyses using a multi-dimensional code AQUA were conducted to evaluate the simularity of the model on thermal stratification phenomena in the upper plenum during a reactor trip event through the comparison of numerical analyses with a full sector model. In addition, thermohydraulic numerical analyses for a reactor vessel cooling system were carried out to estimate the on-set conditions of natural convection in a vertical narrow annular gap in the system. [Analyses for Thermal Stratification in the Full and one-third Sector Reactor Models] (1)The comparison of the result of the one-third sector model with that of the full sector model showed the differences in temperature and velocity distributions at the period from the start of the transient to forty seconds. There were no differences after forty seconds because the velocity at the core outlet decreased. (2)After fortyseconds, there was no apparent effect of intake from the hot leg piping on temperature and velocity distributions in the upper plenum. (3)The results of (1) and (2) indicated that the one-third sector model is applicable for the thermal hydraulic tests instead of the full sector model. [Analyses for the reactor vessel cooling system] (1)No circumferential natural convection was found in the annulus under three different velocity conditions (1.53, 0.153 and 0.0153[m/s]). (2)Asymmetric velocity distribution at the inlet of the annulus also caused no circumferential natural convection, Thus, it was confirmed that there is little or no possibility of circumferential natural convection in the annulus of the reactor cooling system under the actual reactor conditions.