Abstract
The natural circulation process with a thermal hydraulic interaction between the core and the plenum in a reactor vessel is a significant issue for the evaluation of the feasibility of a decay heat removal system using dipped-type direct heat exchangers (D-DHXs) in the upper plenum. In the past years, various efforts have been dedicated to the development of numerical evaluation methods and the enhancement experimental knowledge on the thermal hydraulic interaction between the core and the plenum. This study performs a decay heat removal experiment operating a D-DHX in the sodium experimental facility, called PLANDTL-2. PLANDTL-2 has a simulated core and a reactor vessel. The core is modeled by hexagonal-shaped wrapper tubes and electric heater rods. The reactor vessel has a configuration similar to that of an advanced sodium-cooled fast reactor. In this experimental investigation, the temperature distributions under steady state conditions are obtained both in the core region and in the plenum. As a result, the thermal hydraulic behavior of the cold sodium penetration into multiple rows of the inter-wrapper region and the stratified temperature distribution in the upper plenum are investigated.
