Under the conditions of the actual operation of the nuclear power plant, especially in the drainage system of the heaters, water recycling system, and condensate pump recirculation system, there is a sudden pressure drop after the control valve in the pipeline, which will lead to serious vapor-liquid two-phase flow. Vapor liquid two phase flow can lead to accelerated corrosion in the pipeline. In order to study the influence of vapor liquid two phase flow on the accelerated corrosion of the pipeline, an experimental facility for simulating the operating conditions of the Second Loop was designed and completed. The experimental facility comprises a pressure regulator, high pressure water tank, circulating pump, power heater, low pressure water tank, condensing and cooling system, instrumentation and data acquisition system. The flow characteristics and the corrosion characteristics of the inner wall of the pipe in the flow channel after the control value were studied in the experiments. The flow characteristics include the distribution of water and vapor temperature, pressure drop distribution and void fraction along the flow channel. In the meanwhile, the corrosion rate and surface characteristics of inner wall at different locations were also studied by CCD and SEM. In addition, the influences of pipe material and pipeline geometry on the erosion-corrosion were also studied in the experiments.
Investigation on Two-Phase Flow Characteristics in Nuclear Power Equipment