Study on reactor vessel coolability of sodium-cooled fast reactor under severe accident condition; Water experiments using a scale model

Ono, Ayako ; Kurihara, Akikazu ; Tanaka, Masaaki  ; Ohshima, Hiroyuki; Kamide, Hideki ; Miyake, Yasuhiro*; Ito, Masami*; Nakane, Shigeru*

The water experiment apparatus simulating the thermal hydraulics in a reactor vessel under operating the decay heat removal systems (DHRSs) was fabricated. The theoretical evaluation for similarity and results of basic experiments show applicability for a scale model experiment of a sodium-cooled fast reactor. This paper, moreover, describes the results of flow visualization experiment under operating a dipped-type passive DHX, which is planned to be installed in both a loop type reactor and pool type reactor, and the calculation results using FLUENT comparing with the result of water experiment.