Numerical Simulation Of Steam Generator Heat Transfer Characteristics In VVER-1000 Asymmetric Operation

The pressurized water reactor with multiple loops may have abnormal working conditions with asymmetric boundary conditions. For nuclear power plants, this means turning on and turning off process of a loop. At present, there are great limitations on the traditional one-dimensional transient thermal hydraulic simulation codes, no detail information of the complex three-dimensional(3D) flow is provided. In this paper, the huge quantity cooling pipes are simplified by using the porous media model, and the dynamic b oundary conditions are developed and then embedded into the 3D numerical simulation software(FLUENT) code through utilizing the functionalization of user-defined function(UDF), and thus the dynamic numerical simulation of horizontal steam generator in the secondary side under changing conditions is achieved. So far, an effective, credible and practical numerical simulation method for four-loop PWR at asymmetric operation conditions is established.This paper introduced the models, governing equations and numerical methods used in numerical simulation for steam generator, the porous model theory and boiling phase change model are described in detail.Then, the heat exchange method between the primary side and the secondary side is settled, and the mass transfer coefficient for boiling is refined to describe the phase change process better. And by contrasting the results of 753 MW, 468 MW and 300 MW, find that as the bubbles generated have lower density, driven by buoyancy, a annular flow between the outer pipes and the shell. The annular flow improved the heat transfer efficiency in the outer pipes.After examining the effect of the Underwater Gas Board in VVER-1000, it indicates that the steam load average effect in the lower steam part is obvious.The dynamic boundary condition is compiled into FLUENT to simulate the thermal power changing process of the steam generator. For these cases, change of heat transfer coefficient, volume average temperature for different heat capacity, mass flow and volume fraction of steam in the pipes with time were obtained. Based on the numerical simulation of the secondary side in the thermal power changing process for the first time, the results provide theoretical basis for the analysis of flow induced vibration, damage mechanism of the pipes and the accident condition response measures. This research also made foundation for further study of the use on nuclear power ships and submarines.