Analysis of core heat removal capability under DLOFC accidents for HTGRs

Sato, Hiroyuki  ; Ohashi, Hirofumi ; Tachibana, Yukio ; Kunitomi, Kazuhiko ; Ogawa, Masuro

Design envelope of prismatic High Temperature Gas-cooled Reactors in terms of core heat removal capability under depressurized loss-of-forced-circulation accidents without operating active or passive decay heat removal systems are investigated. Lumped element models consist of core, reactor pressure vessel and cavity wall are presented in order to evaluate transient response of core temperature. Parametric calculations changing the core height, initial core temperature, thickness of side reflector, cavity size and peaking factor are performed. A series of calculation provides relationships of core radius to average power density and reactor thermal power which can remove the heat in core without reliance on specific design features. The results clarified the design envelope for the Naturally Safe HTGR in terms of core decay heat removal.