| The reactor thermal hydraulic analysis code has different accuracy according to the different grid scale.The analytic code with large grid scale can calculate the whole system,but its resolution is low,so it can’t give detailed calculation result in local area.The small grid scale analysis code can compute the local area,but the simulation results are lack of the overall response of the system and has low applicability for some specific conditions.Nuclear power plants involve multiple physical fields interacting with each other,and if only the effects of their physical fields are taken into account,the computational results will be less responsive to other physical fields.Therefore,the reliability and accuracy of simulation results can be greatly improved by using multi-scale and multi-physical processes coupling method for reactor simulation calculation.Firstly,the primary coolant system of QINSHAN phase I nuclear power plant was selected as study project in this paper.The thermal hydraulic subchannel code COBRA(Coolant Boiling in Rod Arrays)and core neutronic code REMARK(Real-Time Multigroup Advanced Reactor Kinetics)were used to establish core thermal hydraulic model and core neutronic model,and loose coupling method and Picard iteration coupling method were used to develop the neutronics and thermal-hydraulics coupling code respectively,and the influence of different coupling methods in calculation is discussed.Meanwhile,the modular system code THEATRe~TMM was used to establish the primary coolant system of QINSHAN phase I nuclear power plant,and coupled with the developed neutronics and thermal-hydraulics coupling code,and obtain the multi scale and multiple physical process coupling code of the primary coolant system.The coupling code of primary coolant system was used to calculate the steady full power condition,the reactive insertion accident,pump outage accident,emergency shutdown accident and high power fast drop load condition are carried out,and the results of steady state calculation and transient calculation are analyzed.The result shows that the relative error of steady state calculation satisfies the requirement of simulation precision.The change trend of transient calculation is consistent with the actual process.It is proved that the code has the ability to simulate the primary coolant system.Secondly,in order to study the pin level neutronics and thermal-hydraulics coupling,the fuel assembly of Qinshan Phase I nuclear power Plant was selected as study project in this paper.Neutronic code which uses the method of characteristics to solve the three-dimensional neutron transport equation are used to establish the fine physical model of Qinshan Phase I fuel assembly,with the fine thermal hydraulic model build by subchannel code.A one by one corresponding grid mapping scheme and an integral average data transmission method based on pseudo legitimacy are proposed for fine coupling.Meanwhile,to develop the fine neutronics and thermal-hydraulics coupling code,the grid mapping,data transfer and convergence determination between the neutronic code and subchannal code are controlled by a external code,which is compiled by Python language.At the same time,the steady-state calculation of the 3×3 fuel assembly and Qinshan phase I fuel assembly is carried out by using the fine neutronics and thermal-hydraulics coupling code.By comparing the result with the reference value,it is proved that the developed refinement code satisfies the precision of the simulation requirement.Through the work of this paper,we have completed the development of multi-scale and multiple physical process coupling code and fine neutronics and thermal-hydraulics coupling code.,and proved that the developed code can provide reference and preview function in the field of reactor safety analysis and reactor fuel assembly design. |
