Abstract
Thermal stratification after a reactor scram may cause significant thermal stress in the reactor vessel and components. Water experiment using an 1/10^<th> scaled upper plenum model was carried out to investigate thermal hydraulic issues in an advanced loop-type sodium cooled fast reactor, which was designed by Japan Atomic Energy Agency. A permeable upper inner structure (UIS) is adopted, which has a radial slit to simplify the fuel handling system. This slit also allows high velocity flow through the UIS. The experiments showed steep temperature gradient and large temperature fluctuation at the stratification interface near the UIS slit, which was caused by local impingement of the jet through the UIS slit. Parameter experiments for core flow rate and temperature difference indicated that the rising speed of the stratification interface was dependent on Richardson number and the temperature gradient of the stratification interface was also influenced. Configuration, i.e., a cylindrical plug in front of the slit and a perforated outer shell of the UIS were examined to mitigate the thermal stress at the stratification interface. The temperature gradient was reduced greatly in a case where the plug was located at a lower position near the core in the upper plenum.
