In the short term period the use of advanced Small Modular Reactor (SMR) is one of the most promising options for the deployment of nuclear technology. The validation and assessment of the best estimate thermal hydraulic system code TRACE against SMR thermal hydraulic phenomena is a novel effort. In this framework the use of the natural circulation database developed at the OSU-MASLWR test facility, simulating the MASLWR reactor prototype, is of interest for analyses of the TRACE code capability in predicting natural circulation and primary/containment coupled behavior in SMR. The target of this paper is to analyze the TRACE V5 capability for the simulation of natural circulation phenomena, at different primary and secondary side conditions, and to simulate the primary/containment coupling behavior, typical of the MASLWR design in Beyond Design-Basis Accidents (BDBA), by using a 3D TRACE model of the containment. The results of the calculated data show that the TRACE code is able to predict, from a quantitative point of view, the primary natural circulation mass flow rate, and that a 3D TRACE model of the containment is able to predict the main thermal hydraulic parameters, characterizing the primary/containment coupled thermal-hydraulic behavior.

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