Research On Thermal-Hydraulic Analysis Method Of Nuclear Steam Generator Based On Porous Medium Theory

The steam generator is an important equipment to connect the primary and secondary loops of nuclear power plants.The three-dimensional thermohydraulics distribution of the secondary side,the flow distribution in the tube and the coupled heat transfer characteristics of the primary and secondary side are the basis of thermal design and structural optimization of the equipment,and are crucial to the economy and integrity of the equipment.The Computational Fluid Dynamics(CFD)method is used to study the thermohydraulics characteristics of nuclear steam generators in this paper.Firstly,based on ANSYS Fluent Euler two-fluid model,this paper introduces porous medium model to develop a set of three-dimensional two-phase thermal hydraulic characteristics analysis program of nuclear steam generator.Based on the two-fluid six-equation model in porous media,single-phase convective,subcooled boiling and saturated boiling models were established to predict the heat transfer process of the secondary side fluid.A multi-channel two-dimensional flow distribution and heat transfer model was established,which was coupled with the three-dimensional two-phase flow field outside the tube.The heat transfer difference caused by uneven flow distribution in different heat transfer tubes was well described.The distributed resistance model is used to calculate the resistance of tube bundle,flow distribution plate,support plate and separator.Secondly,the accuracy of the program is verified from three dimensions: unit,subsystem and whole system.In the dimension of unit problem,the resistance characteristics of single pipe flow and pebble bed are designed,and the correctness of the program distributed resistance model is verified by the analytical solutions.In the subsystem dimension,the boiling heat transfer vapor-liquid two-phase flow field verification program in FRIGG rod bundle experiment was used to predict the accuracy of two-phase flow heat transfer and phase field distribution in the tube channel.In the whole system dimension,MB-2 small steam generator experimental data were used to verify the program’s accuracy in predicting the coupled heat transfer of steam generator at the primary and secondary sides,and the internal outflow field and temperature field in the case of coupled heat transfer inside and outside the tube.Finally,the flow and heat transfer characteristics of ZH-65 steam generator simulator and AP1000 steam generator on the primary and secondary sides were studied based on the developed program,and the distribution rules of temperature,velocity,pressure,void fraction and heat transfer coefficient on the secondary side and the distribution rules of temperature,flow rate,energy and heat transfer coefficient on the primary side are obtained.The results show that there are thermodynamic and kinetic disequilibrium between the fluid phases on the secondary side.In addition,the secondary subcooled water of the large-diameter steam generator can reach saturation temperature above the tube plate,and the cross-flow in the support plate area of the small-diameter steam generator is more obvious.In this paper,according to the flow and heat transfer law of nuclear steam generator,a set of three-dimensional two-phase thermodynamic and hydraulic characteristics analysis program is developed and verified,which can be used to predict the thermal and hydraulic characteristics of the primary and secondary sides of small and large steam generators,and support the thermal and hydraulic design and performance check of nuclear steam generators.