Development Of Sub-channel Coupling Analysis Method And Its Applications For Lead Based Fast Reactors

As one of the six most potential reactor types proposed by The Generation IV International Forum(GIF),lead or lead alloy cooled fast reactor(LFR)has drawn much attention due to its capability of nuclear waste transmutation and fuel breeding,and the properties of good safety and economic practicability.The concept of lead alloy cooled fast reactor was put forward during the same period with sodium cooled fast reactor.In recent years,the technical innovations in anti-corrosion materials,oxygen concentration control and other key technologies,make lead alloy fast reactor more acceptable as commercial nuclear power plant,demonstration activities are underway and planned during GIF's time frame.Accurate predictions of thermal hydraulics field such as temperature distributions in the fuel assemblies and velocity fields in reactor vessel,are required in the design phase of a lead cooled fast reactor for safety reasons.Based on these requirements of thermal hydraulic characteristics and transient safety characteristics analysis for LFR,a new sub-channel analysis code was developed,a thermal hydraulics-neutronics coupled analysis code was also developed based on this new code,some multi-scale analysis methods have also been performed in this paper,work accomplished in this dissertation can be divided into the following parts:1.Develop a sub-channel analysis code for lead based fast reactors,KMC-Sub,the conservation equations of fuel rods and fluid are solved separately and empirical correlations are combined to simulate the transverse mass and momentum phenomenon between adjacent channels and rods.The structure of the program and correlations have been upgraded.2.Test the KMC-Sub code,by using liquid metal fuel rods bundle experiments data,the pressure drop correlations and heat transfer correlations have been validated separately,the overall validation of the code have been performed by using temperature distribution data.Applicability and accuracy of the models adopt by KMC-sub have been confirmed.3.Develop a thermal hydraulics-neutronics coupled analysis code,the point kinetics neutronics model has been combined with KMC-Sub model,the coupled simulation results have been compared with FLUENT/PK to test the accuracy of the code.4.Demonstrate the porous media simulating capability by a nuclear system using a combined CFD model of clear flow and porous media flow.Propose a hybrid approach to couple a CFD code with the sub-channel code,develop the coupled CFD/sub-channel approach,verify the coupled method using a simple reactor.Thus,by developing an new sub-channel thermal hydraulics analysis code,introducing the point kinetics neutron ics model into the sub-channel model,and combing the best features of a CFD code and the sub-channel code,a simulation capability has been developed to model multi physics,multi scale effects in lead based reactor system with existing computational resources.The utility of this capability has been demonstrated by application to calculation cases.These coupled sub-channel codes capability will be useful in developing better optimized reactor designs.