Research On Calculation Of Reactor Fine Power Distribution Based On MOC

Power distribution in the reactor core is one of the important parameters related to reactor safety and economy.But the increasing heterogeneity and complexity of the core design posts a challenge to the traditional power distribution and fine power distribution calculation.Meanwhile,along with the proposition of high-fidelity computational requirements for reactor,the core physics calculation method transits from traditional three-step method to the pin-by-pin two-step method and the whole core direct transport calculation approach.The method of characteristics?MOC?,which combines the merits of collision probability and SN method and integrally solves the neutron transport equation along the neutron flight trajectory,is from the characteristic line form of neutron transport equation.Its transport solving process is not restricted by the boundary conditions and geometry,which makes it applicable to solve the neutron transport equation with high precision and easy popularization in arbitrary complex geometry theoretically.Thus,MOC becomes one of the research hotspots and focuses currently in reactor physics computing field,and it has been widely used in transport calculation,especially for two-dimensional assembly transport problem.Besides,MOC is one of the crucial methods for future whole-core direct transport calculation.Through extensive research and further study of MOC,a program named FSMOC based on the MOC for neutron transport calculation is developed in order to realize the fine power distribution.The geometric processing in FSMOC adopts method of prefabricating geometry to describe the geometry.Taking into account that the actual computational domain of reactor engineering problem are often composed of cells or assemblies with same dimension and geometry style,we can just generate the characteristic lines within the repetitive structure?cell?and then extend it to the whole solution domain with certain connecting rules,realizing the cell modular ray tracing in the program.In order to facilitate debugging program and to intuitively observe the information of geometry and characteristic line,an interface module is developed to link AutoCAD.Transport equation solving module in FSMOC is developed by using the source iteration method on the condition of flat source approximation assumption.Finally,the program is verified by different benchmarks,which is issued by relevant international organizations,including typical PWR cell benchmark,BWR lattice benchmark with two gadolinium pins,UO₂ assembly of C5G7 benchmark and C5G7-MOX core benchmark.The numerical results demonstrate that the program can give excellent accuracy for both effective multiplication factor and relative flux distribution for neutron transport problem.In order to ensure sufficient accuracy of MOC,fine mesh and intensive characteristic lines are needed when ray tracing,thus resulting in slow computing speed and limiting wide application for large scale problem.The parallel design based on OpenMP is simply discussed in this paper for reducing the computing time.By adding OpenMP compile directive statement in the original program can realize parallel design.The use of OpenMP can obtain good acceleration effect,making the calculation time reduced.