Development of the Monju core safety analysis numerical models by Super-COPD code

Yamada, Fumiaki; Minami, Masaki*

Japan Atomic Energy Agency constructed a computational model for safety analysis of Monju reactor core to be built into a modularized plant dynamics analysis code Super-COPD code, for the purpose of heat removal capability evaluation at the in total 21 defined transients in the annex to the construction permit application. The applicability of this model to core heat removal capability evaluation has been estimated by back to back result comparisons of the constituent models with conventionally applied codes and by application of the unified model.