Study On Multi-Scale Thermal-Hydraulic Coupling Calculation Method For Reactor Core

Study on thermal hydraulic calculation is an important part of the safety analysis of pressurized water reactor,which is beneficial to improve the safety and economy of the reactor.At present,the thermal hydraulic calculation programs that have been applied to the nuclear reactor.According to the modeling scale,they can be divided into: system codes,sub-channel codes and CFD codes.The Codes of different scales have advantages in efficiency and precision.In order to achieve an appropriate balance between accuracy and efficiency in the integer simulation,it is necessary to carry out research work on multi-scale coupling calculation.In this paper,PWR was used as the object,and the commercial CFD code FLUENT was used to detailed calculate the downcomer and lowerplenum of the pressure vessel.At the same time,the sub-channel code COBRA-IIIC/MIT2 was used for the whole core thermal hydraulic analysis.Finally,combined with the lower plenum and the core domains,the thermal hydraulic multi-scale coupling calculation is carried out.Firstly,in the refined CFD calculation,the meshing scheme of the complex structure of the downcomer and lowerplenum was studied,the applicability of the turbulence model was studied by partition calculation.Based on the work above,Multiple RANS turbulence models computing scheme was designed.the applicability of the multi-turbulence model coupling calculation scheme in pressure vessel structure was validated by the experimental data,and compared with the traditional single model calculation scheme,they have achieved better economy on the premise of ensuring sufficient calculation accuracy.Secondly,the full core sub-channel model of KONVOI type reactor was established by fuel assembles,and the reasonableness of the model is proved by validation calculation.And the thermal-hydraulic phenomena of different channels were analyzed.The variation rules of key thermal-hydraulic parameters(pressure drop,flow rate,temperature,etc.)were summarized.The above work lays a foundation for the study of multi-scale thermal-hydraulic coupling calculation.Finally,on the basis of the independent verification of the two codes' calculation schemes the domain decomposition coupling method is adopted.Based on the secondary development function of FLUENT and the scalability of the COBRA-IIIC/MIT2.The FLUENT/COBRA Implicit coupling calculating program was developed by writing the data interaction programs and the logic master program.The interaction between the lowerplenum and the reactor core regions were analyzed through the steady state calculation.The sensitivity of time step and the transient response of the coupling program were studied through the transient simulation.The calculation and analysis of the partial-loop loss-of-coolant-accident were carried out,and the characterization of physical processes were summarized in the couple regions.Through the research of the above three aspects,the calculation method of the multi-scale coupling of the core thermal hydraulic analysis was mastered,that will provide a certain reference value for the optimization design and the safety analysis of the nuclear reactor.