EXPERIMENTAL THERMAL-HYDRAULIC STUDY OF A STEAM- DROPLETS FLOW INSIDE A VERTICAL PIPE -APPLICATION TO LOCA IN PWR REACTOR

International audienceDuring a Loss of Coolant Accident (LOCA) in a Pressurized Water Reactor (PWR), partial or even complete drying of fuel assemblies may occur. In these conditions, the fuel temperature can increase leading to significant deformation of the fuel rod cladding and a partial blockage of the fluid domain. These zones can thus affect the cooling of the nuclear core during the reflooding phase by the emergency core cooling systems (ECCS). Investigating the cooling capacity as well as the thermal-hydraulic characteristics of the flow in these zones is a crucial point to guarantee nuclear safety. In order to characterize the cooling capacity of these deformed zones by a two-phase flow, an experimental setup at sub-channel scale has been developed. The experimental setup consists of a tubular Venturi test section, a droplet supply system and a superheated steam supply system. The test section is heated up to about 700°C before starting the flow composed of superheated steam and dispersed droplets. During the cooling phase, optical measurements are carried out upstream and downstream the test section to measure the droplets diameter, velocity and temperature, while an IR camera records the transient of the tube temperature. This paper presents experimental results with three different test section diameters to analyze the influence of the blockage ratio on the cooling capacity. The droplets evaporation was evident with their reduction in size downstream the heated tube