Investigation of core deformation reactivity model improvement in plant dynamics analysis code during ULOHS Test of U.S. experimental fast reactor EBR-II

Yoshimura, Kazuo; Doda, Norihiro  ; Hamase, Erina ; Fujisaki, Tatsuya*; Igawa, Kenichi*; Tanaka, Masaaki  

Sodium-cooled fast reactors have intrinsic safety features decreasing reactor power during the increase of the core inlet temperature by the feedback reactivity of the radial expansion of the core support plate. It is necessary for the composition of the core highly of secure to understand the influence of the safety features with high accuracy. In this paper, first, the 1D-CFD coupling method with cold pool as CFD region which enables the plant dynamics analyses taking account of the thermal stratification in cold pool was applied to the ULOHS (Unprotected Loss Of Heat Sink) test performed in the experimental fast reactor U.S. EBR-II and the evaluation of the core inlet temperature could be improved. Secondly, the sensitivity analyses concerning the core bowing reactivity were carried out with the aim of improving the evaluations of the core deformation reactivity and the applicability of the core bowing reactivity model to the test could be indicated.