Study On Key Parameters In Reactor Particle Transport Based On Monte Carlo Method

Monte Carlo method is good at accurate simulation and flexible geometric description, so it is widely used in particle transport simulation. Since the mesh tally can support accurate statistics of reactor physics parameters, most international particle transport code have implemented this function in recent years. And as the basis for reactor safety and thermal analysis, the capability of reactor dynamic parameters calculation is essential for Monte Carlo particle transport simulation codes.In this study, with extensive investigation of various mesh tally techniques and reactor dynamic parameters calculation of methods, a mesh tally and a reactor dynamic parameters calculation method were studied and implemented based on SuperMC code (Super Monte Carlo Calculation Program for Nuclear and Radiation Process)and hybrid evaluation nuclear data library HENDL.The mesh tally of static parameters(i.e., flux, reaction rate and power) based on rectangular coordinate system and cylindrical coordinate system was implemented. At the same time, we dealt with the inaccuracy of traditional mesh tally when the mesh crosses the multiple material boundaries. For the dynamic parameters, we studied and compared different methods and chose the iterated fission probability method to calculate the effective delayed neutron fraction, prompt neutron lifetime, neutron generation time and reactor cycle.A series of examples represented by TCA benchmark are selected to verify the correctness of the mesh tally. And28cases were chosen form ICSBEP (International Criticality Safety Benchmark Evaluation Project) to verify the correctness of the dynamic parameters calculation method. The chosen benchmarks involved main fissionable nuclides, most energy spectrum and geometric models with different complexity. The results showed its validation and correctness.