Small Lead-Cooled Fast Reactor Core Thermal-Hydraulics And Neutronics Coupling Research And Application

As one of the six main types of Gen-IV nuclear reactors,the leadcooled fast reactor has outstanding advantages in terms of sustainability,safety,economy,and operating experience.In order to ensure the accuracy of lead-cooled fast reactor core design and safety analysis,the development of lead-cooled fast reactor analysis tools and neutronics and thermalhydraulics coupling research are urgently needed.This paper aims to integrate key models based on existing liquid metal experiments and redevelop a sub-channel code suitable for lead-cooled fast reactors.At present,the couplings of neutroncis and thermal-hydraulics mostly adopt traditional operator-split semi-implicit methods,but this method uses lagging parameters and cannot achieve synchronous convergence of various physical fields.The fixed-point implicit method can ensure synchronous convergence of parameters,but the convergence is slow or even failed.The approximate block Newton method has high calculation efficiency but the calculation performance in core transients is unknown.Regarding to neutron-physical and sub-channel code,the coupling codes based on the above methods have been developed.And the accuracy and computational efficiency of the three methods have been compared.Then based on the developed coupling method,a small lead-cooled fast reactor design has been proposed,and steady-state and transient analysis have been performed.The research route of the thesis can be summarized as follows: a)introducing the liquid metal models into the sub-channel code and completing validation through the experimental data;b)developing coupling code based on operator-split semi-implicit,fixed-point implicit and approximate block Newton coupling methods;c)validating the accuracy through the NEACRP PWR rod ejection accident benchmark;d)applicability evaluation in the lead-cooled fast reactor;e)application to the design and analysis of small transportable lead-cooled fast reactor.The main content of the paper includes:(1)Lead-cooled fast reactor analysis code development and evaluation:Firstly,The physical properties,pressure drop models,turbulence mixing models and heat transfer correlations of the liquid lead-based coolant has been summarized and introduced into the light water reactor sub-channel code COBRA-IV,redeveloping the sub-channel code COBRA-LM fit for lead cooled fast reactor.Then,the sub-channel code has been validated through the liquid lead-bismuth experiment KIT-KALLA and lead-cooled fast reactor SUPERSTAR,and the results showed that COBRA-LM has good accuracy and applicability in lead-cooled fast reactors.At last,the applicability of the neutron diffusion code in fast reactors has been evaluated through TAKEDA fast reactor benchmark.(2)Neutronics and thermal-hydraulics coupling codes development:Based on sub-channel code COBRA-LM and neutron-physics code SKETCH-N,operator-split semi-implicit,fixed-point implicit and approximate block Newton coupling codes have been developed via the parallel virtual machine PVM function library.The accuracy of the three methods has been validated by NEACRP PWR rod ejection benchmark,and the applicability of the coupling procedures has been evaluated by the simple-designed lead-cooled fast reactor core.The results show that the operator-split semi-implicit is fit for sharp transients,and the approximate block Newton method adopting bigger time step size in mild transient has the advantages of high accuracy and efficiency.(3)Lead-cooled fast reactor design: The sensitivity analysis among fuel rod numbers within a fuel asssembly,fuel rod diameters,pitch-diameter ratios,and outermost fuel rods distance to the inner wall of the assembly has been performed.Then the combination that meets the design criteria with the minimum equivalent diameter has been selected to carry out the sensitivity analysis of the proportion of plutonium dioxide.At last,a threezone core layout has been proposed,and the burnup,control system,reactivity coefficient,steady-state and transient neutroncis and thermalhydraulics coupling characteristics evaluation has been performed,indicating that the core can achieve the design goals and has inherent safety.This paper has completed the lead-cooled fast reactor sub-channel analysis tool redevelopment and the neutronics-thermal-hydraulics coupling method research.Based on the above tools,the design and the analysis of neutron-physics and thermal-hydraulics characteristics and the transient evaluation of the lead-cooled fast reactor have been performed,which provides analytical tools and methods with theoretical and practical significance to lead-cooled fat reactor design and research.