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Yoshimura, Kazuo; Doda, Norihiro; Fujisaki, Tatsuya*; Igawa, Kenichi*; Tanaka, Masaaki
no journal, ,
In the ULOHS tests performed in the experimental fast reactor U.S. EBR-II, the thermal stratification in the cold pool (CP) has influence on the whole plant behavior during the events because the secondary sodium pump tripped without scram nor tripping the primary pumps. In order to create the one-dimensional model for the CP of the plant dynamics analysis code, the multi-dimensional thermal hydraulics analyses using computational fluid dynamics (CFD) code were conducted to investigate the thermal hydraulics phenomena in the CP. It was found by comparison with the experimental data that the modeling of the detail sodium flow at the outlet of the intermediate heat exchanger, the leakage flow from the inner components to the cold pool, and the heat radiation from the CP to the atmosphere was important to the evaluation of the thermal stratification.
Yoshimura, Kazuo; Doda, Norihiro; Hamase, Erina; Fujisaki, Tatsuya*; Igawa, Kenichi*; Tanaka, Masaaki
no journal, ,
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.