Research On Three-dimensional Hexagonal Node Code Verification And Cross Section Processing For Fast Reactor

With the introduction of the concept of the fourth generation advanced nuclear energy system,fast reactor emerges in the fourth generation advanced nuclear energy system with the advantages of inherent safety,proliferative nuclear fuel and long-life nuclides in transmutation and waste.Countries have gradually begun to attach importance to the development of fast reactor technology.Compared with the traditional finite difference method,the nodal method has great advantages in computing speed and is widely used in fuel management or parameter design.Different from rectangular fuel assemblies used in traditional PWR,hexagonal fuel assemblies are often used in fast reactors.Therefore,the study of hexagonal nodal method is of great significance to the current research of fast reactor and to the promotion of the autonomy of domestic nuclear power related design procedures.In this paper,based on the transverse integration method,the high-order polynomial is used to describe the distribution of neutron flux density in the hexagonal node to solve the neutron diffusion equation,and the three-dimensional hexagonal nodal method program HEXA3D is developed independently.The radial and axial distribution of one-dimensional neutron flux density is approximated by the fourth-order and third-order polynomials respectively by the transverse integration method.The flux moment equation and the response equation are established by using the weighted residual method,and the effective multiplication factor and the average neutron flux density in each node are obtained.On this basis,the two-dimensional and three-dimensional minority group benchmarks such as VVER-440,VVER-1000 and IAEA bare core are verified.In order to further validate HEXA3D's computing ability for complex problems,a unique set of processes including cross-section processing and core calculation has been established by combining MGGC,a multi-group cross-section processing program independently developed by the team.Thirty-three groups of neutron reaction cross-sections have been produced from ENDF/B-?.1 evaluation nuclear database for the international lead-based fast reactor RBEC-M benchmark problem,and an interface has been developed.Triso,the I/O module of MGGC,is used to generate cross-section files in specific formats for calculation.The final calculation results show that in the two-dimensional and three-dimensional minority group problems,the calculation time is only about one second.Keff error is less than several hundred PCM compared with that calculated by the finely partitioned finite difference program in the literature,and the maximum relative deviation of power distribution is 8.75%.The calculation accuracy is comparable to that of some mature nodal method programs.For 33-group RBEC-M benchmarks,using the same cross-section,the maximum relative deviation of power distribution of each node is-4.243%,Keff error is 343pcm compared with the reference program,and the calculation time is 17.4 seconds.In summary,the self-developed hexagonal nodal method program HEXA3D has high computational efficiency and sufficient computational accuracy,and can be combined with the existing multi-group cross-section processing program of the team to form a calculation system.In the future,more benchmark questions will be used to verify the engineering applicability of the program.