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Tanaka, Masaaki; Hiyama, Tomoyuki; Murakami, Satoshi*; Doda, Norihiro; Ohshima, Hiroyuki
no journal, ,
In order to improve the accuracy of the numerical estimation method for the thermal transient phenomena in the sodium-cooled fast reactor has been conducted by using code coupling technology with the system analysis code for plant dynamics analysis, the multi-dimensional code for analysis of thermal-hydraulics in the plenum, and numerical estimation code of structural integrity for the local region, in viewpoint of enhancement of safety measures in sodium-cooled fast reactor. In this presentation, outline of simulation method development is introduced.