Numerical Simulation Of Flow Boiling And Critical Heat Flux In A PWR Fuel Assembly

The critical heat flux(CHF)is essential to the operational safety of nuclear power plants,and it plays a vital role on the structural design and the research of thermal hydraulic phenomena of a fuel assembly in the reactor.To improve the heat transfer efficiency of the reactor fuel assembly,it is necessary to accurately calculate the twophase flow boiling characteristics and the CHF value in the fuel assembly.Compared with single-phase flow,mass,momentum,and energy transfer process occur at the interface of two-phase flow.In addition to the conventional turbulence model and grid model,the generation,detachment,mergence and condensation of bubbles are also needed to be modeled.The interactions between liquid and vapor phases need to be characterized by various interphase forces.Furthermore,the wall boiling model is needed to describe the distribution of heat flux at the heating surface,which includes several empirical auxiliary models.Accurate simulation of flow boiling phenomenon requires the further study of many sub-models in numerical calculations.In this paper,the research status at home and abroad was reviewed according to the complexity of geometric structure,the presence or absence of phase transition,and whether or not to predict the critical heat flux.On the basis of literature research and analysis,aiming at the deficiency of existing studies,the research was carried out to accurately predict the critical heat flux in bundle channel of PWR fuel assembly.Firstly,in view of the above research objectives,numerical simulation of subcooled flow boiling in a vertical tube was carried out.According to the distribution of wall temperature,mainstream temperature and void fraction,the sensitivity analysis of the parameter settings of the grid models,turbulence models,boiling models and interphase force models were analyzed.The recommended model and parameter setting for CFD simulation of two-phase flow boiling was given.Later,with the increase of heat flux,it is not clear whether the model will still be usable when the flow reaches the boiling crisis.Therefore,two phase flow boiling simulation of water and R134 a was carried out in a vertical round tube and a single rod channel with spacer grids separately under high pressure close to the conditions of PWR to verify the applicability of the numerical model at boiling critical state,and then predict the CHF value under various conditions.The results demonstrated that the average error of the CHF prediction for a vertical round tube with water was less than 11% and the average error for a single rod channel with R134 a was less than 8%.The auxiliary model combination and the method based on boiling curves to predict CHF showed a great application prospect for the CHF prediction of complex structures.Finally,the critical phenomena of two-phase flow in a 5 × 5 fuel rod bundle were investigated.In addition to predicting the CHF value under various working conditions,a method to determine the CHF position was also proposed.The simulation results show that the average deviation of CHF was less than 16%,and the average deviation of DNB location was less than 10%.The numerical model used in this paper can better predict the CHF value and its location in complex structures.The research in this paper provides technical support for thermal hydraulic characteristics analysis of reactor fuel assembly.