Research On Coupled Multi-scale Thermal-Hydraulic And Coupled Neutronic/Thermal-Hydraulic Methodology For Integral PWR

In nuclear reactor system research,coupled multi-scale thermal-hydraulic(T-H)and coupled 3-D neutron kinetic(N-K)and T-H method have been used to analyze coupling neutronic-thermohydraulic features of reactor core,which is one of the most concerned research areas,and it could help improve simulation fidelity and optimize nuclear reactor design.In this thesis,to further understand the operation strategy and safety features of Integral PWR-200(IP200),the coupled method combining RELAP5 and computational fluid dynamics(CFD)code Fluent,coupling RELAP5 and 3-D two-group N-K diffusion code,and coupling Fluent/N-K method have been developed.This thesis reported the detailed processing of coupling ways,time advancement scheme,spatial mesh mapping,synchronizing different time steps,way of exchanging data and convergence logic in the coupling code development procedure.Especially,function fitting method(FFM)for coupling parameter distribution has been first proposed for exchanging data on the coupling interface,which will eliminate error and numerical instability,meanwhile,also accelerate convergence speed and improve simulation accuracy.To verify and validate the above coupled codes,comparison to benchmark test results have been conducted.The results showed good agreement with Edwards blowdown test,VVER hexagonal fuel assembly benchmark and the Qinshan nuclear power plant(NPP)operation and experiment data.IP200's entire system and reactor core were modeled using the coupled code.The simulation has been done regarding to different operation strategies,including rated power condition,natural circulation(NC)and once-through steam generators'(OTSGs)group operation under low power conditions.For forced circulation(FC)operation,reactor power,coolant flow and temperature features were mainly influenced by fuel assembly enrichment,control rods configuration and thrust of main coolant pumps.Under 25%FP NC operation,the coolant flow rate has been determined by flow resistance,density difference and height difference in the loop.The different coolant flow features compared to FC operation further influence the different power distribution.Besides,25%FP OTSGs group operation transient behavior eliminated secondary-side flow instability,but also characterized by a strongly asymmetric behavior of the primary system which was caused by non-uniform coolant temperature distribution at the core inlet.To further study coolant flow features through downcomer and lower plenum in the reactor pressure vessel(RPV),in this thesis,coupled RELAP5/Fluent method has been used to investigate the coolant distribution by the effect of FMC.Velocity and temperature characteristics under 50%FP and 25%FP low power conditions have been analyzed.The results illustrate that the FMC can help improve the non-uniform coolant temperature distribution at the core inlet effectively compared to without FMC existence.Meanwhile,the FMC will induce more flow resistance in the downcomer and lower plenum.The coupled method considers system behavior that matters in CFD simulation,as a result,it will provide precise boundary conditions to the CFD portion.To analyze safety feature of IP200,a system analysis of the double-ended rupture of main steam line break(MSLB)accident,and single double-ended steam generator tube rupture(SGTR)accident occurred on IP200 has been conducted.The system and break simulation have been modeled using RELAP5 code.In this thesis,we present the main operation parameters variation under MSLB,and compare them with Organization for Economic Cooperation and Development(OECD)benchmark problem.The results illustrate that due to SMR's closedtype fuel assembly and OTSG inherent features,there is significant difference compared to benchmark results.However,the simulation demonstrates IP200 is safe enough under the most severe MSLB accident.To further investigate safety performance under MSLB and SGTR accidents,coupled codes using two-group neutron diffusion kinetics code and CFD solver Fluent has been introduced.The reactor core behaviour was analyzed in coupled code simulation,whose boundary conditions were derived from system code interface.The results presented transient three-dimensional distribution of the key operation parameters such as reactor power and coolant temperature,that also demonstrated the inherent safety features of IP200.The current work will bring about the ability to explore coupled multi-scale thermalhydraulic and coupled neutronic/thermal-hydraulic operation characteristics and transient safety features,which give more precise local parameters.Meanwhile,the coupled code could also fully consider operation characteristics and performance of the whole system.Furthermore,through the research and application of coupled methods,it can provide a better understanding on operation and safety features of integral pressurized water reactor,which will also be beneficial to improve best estimate ability in future work.