Investigation of Transient Flow and Heat Transfer for Passive Nuclear Reactor Direct Safety Injection

This paper describes a transient flow and heat transfer characteristics for a 1400MW passive pressurized-water reactor (PWR) direct vessel injection (DVI) system in different accident transient processes. The study components include reactor pressure vessel and a series reactor internal such as core barrel and radiation surveillance capsules, the flow channel include downcomer and lower plenum. Furthermore, the inject device is designed with special structures: first, a venturi type tube nozzle is connected to pressure vessel, second, a flow deflector is arranged in the downcomer which is facing the nozzle. This special structures will make the flow mixing and heat transfer very complicate and hard to predict. This study considers characteristics of the loops temperature and flow rate for both injection loop and reactor cold leg loop which are continuous change and long duration. Computational fluid dynamics (CFD) method is used in this study. Before this study, the physical model and numerical method are verified by an independent scaled model experiment. In this real reactor scale study, two typical accident transient processes are analyzed in this study, and temperature distribution on both reactor vessel and reactor internals are obtained. According to results analysis, the characteristics of heat distribution in downcomer were obtained: The injection fluid which is supposed to flow to core barrel is driven to the side of reactor vessel by the reflector. With the injection fluid flows in downcomer, the injection flow shape comes to a triangle. In addition, the transient results show that correlation degree of temperature distribution and injection flow character is gradually decreased with the increase of time history for passive injection. Overall, the exercise complements the activities in reactor safety analysis areas in understanding the origins of thermal load in reactor vessel, and being able to quantify them. Results of this study can be directly used by analyzing of reactor fatigue mechanics. (CSPE)