Initial Research Of Thermal Hydraulic Feedback In The Reactor Monte Carlo Analysis

The methods of solving the neutron transport problems are deterministic method and Monte Carlo method.Monte Carlo method has the advantages of realistic simulation of particle history and simulation of complex geometry.In the past,the Monte Carlo method was constrained by computer configuration and computational power,and was often used as a complement to deterministic methods.At present,with the rapid development of computer technology,Monte Carlo method has been widely used in critical safety analysis,transport calculation and full-core simulation of reactor.In the reactor calculation,the core physical parameters and thermal parameters affect and constraint each other,regulating the operation of the reactor,this phenomenon is known as the reactor thermal feedback effect,and is the basis for the analysis of core physical phenomena.Therefore,it is important to study the Monte Carlo method and Monte Carlo method and to study the thermal feedback effect of the reactor.The coupling of the reactor physic and thermal hydraulic is helpful to simulate the operating state of the reactor and reveal the physical properties of the reactor.In this paper,Monte Carlo methods and procedures are investigated and a thermal calculation model suitable for PWR is established.In the study of Monte Carlo method and procedure,the critical calculation process and,the calculation and statistical method of Keff and power distribution are mainly studied,and the cos RMC Monte Carlo program jointly developed by Beijing Software Center of State Nuclear Power Technology Company and Tsinghua University is selected.In the establishment of thermal calculation model,considering the purpose of PWR reactor and Monte Carlo program parameter feedback,a single-channel model is selected to establish the axial and lateral heat transfer of fuel,cladding,air gap and moder ator,and is studyed by selecting the appropriate thermal parameters and calculation formula using direct numerical simulation method.Also in this paper,the cross-section processing of temperature feedback has been studied,and the pseudo-material interpolation method has been selected and the ACE format cross-section library at different temperatures has been produced.Firstly,the method of cross-section temperature feedback is analyzed and compared.The interpolation method of pseudo-material is selected considering the calculation precision and calculation efficiency.Secondly,the ACE format section library is made with SIGACE program.By selected the cell benchmark,the effect of the effective value-added factor Keff in the use of different temperature section library of fuel is analyzed and calculated,and the change trend of the 235 U section at different temperatures is compared,and the reasonableness and correctness of the section library are proved.Then,the critical function of the Monte Carlo cos RMC is tested and verified,and thermal feedback module and coupling program are developed based on cos RMC critical function.First,cos RMC was used to calculate the BEAVRS benchmark,and based on the calculation and analysis of the critical eigenval ue,the control rod value and the whole reactor power distribution,the critical calculation function of cos RMC is preliminarily verified.Then,in the Monte Carlo program cos RMC on the basis of the use of C++ programming secondary development,the use of internal coupling method to modify the source code,add the thermal feedback calculation module and coupling program can achieve thermal parameters(fuel,Coolant temperature and coolant density)and the physical parameters(axial power distribution),whic h improves the critical calculation function of Monte Carlo program.Finally,this paper selects the PWR cell and the component calculation model respectively to test and validate the results of the coupling program.The effect of thermodynamic feedback effect on the critical settlement results of Monte Carlo program is verified by calculating iteration convergence,Keff and power distribution.At the same time,through the axial fuel temperature,clad temperature,and coolant temperature and density and other thermal parameters of the results of comparative analysis,verify the accuracy of thermal module calculation.The research of this paper will lay a foundation for the simulation of the reactor operation and the wider application space,and provide a reference for the Monte Carlo thermal feedback calculation and analysis.