Study on reactor vessel coolability of sodium-cooled fast reactor through water experiments

Sasaki, Keisuke; Aizawa, Kosuke ; Ezure, Toshiki ; Tanaka, Masaaki  

To enhance safety of sodium-cooled fast reactor, water experiments with the 1/10 scaled test section of reactor vessel (RV) equipped several-type core coolers were done in order to clarify characteristic thermal-hydraulics phenomena during operation of the decay heat removal system using the core cooler at a severe accident condition. Parameter was the power conditions using the electric heaters located at the core and the bottom of RV simulating the condition after core disruptive accident. Core coolers: dipped-type direct heat exchanger (DHX), penetrated-type DHX, and reactor vessel wall cooler, were individually tested in the experiment. Temperature distribution was measured to investigate the flow pass in the RV in each cooler operation and decay heat removal performance by each core cooler.