The pool-type sodium-cooled fast reactor adopts a new type of decay heat removal method to discharge the core heat to ensure the safety and cool of the nuclear power plant in a long time under the station blackout accident. The natural circulation and forced circulation will lead to different thermal and hydraulic characteristics. In addition, the natural circulation produces different hydrothermal phenomena on the secondary side by operating Direct Reactor Air Cooling System (DRACS). The Plant Dynamics Test Loop (PLANDTL-DHX) experimental device can simulate the flow and heat transfer phenomenon in the core. In the present work, a “modular modeling and integrated calculation method” was developed to conduct three-dimensional numerical modeling and calculation for the PLANDTL-DHX experimental device. Meanwhile, the steady-state verification of PLANDTL-DHX was carried out and compared with the experimental data. On this basis, the transient operating conditions of DRACS were calculated, and the temperature, flow distribution and transient variation characteristics in the key positions such as center subassembly, upper plenum and Intermediate Heat Exchanger (IHX) were preliminarily obtained. Based on the experimental results, the numerical calculation is verified.

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