Development of Auto-Modeling Tool for Neutron Transport Simulation
The leading method for studying radiation transport and interaction continues to be the use of Monte Carlo simulations. However, it requires the geometry be provided in a unique manner, without providing robust or user-friendly tools. One often has to “manually translate a shape of a target into the Monte Carlo codes; it is very difficult to create the actual detailed geometry. The neutron transport software has been programed on GEANT4 framework, it simulates the ignition process of neutron. It is necessary for high accuracy calculation results to create an actual pulse reactor model. A more efficient way is to make use of available CAD geometry data for Monte Carlo simulations. The neutron transport program uses the GDML file for importing geometry, but CAD systems use the Boundary representation (BREP) to define solids and use special file format for data exchange. So this requires a suitable interface between CAD systems and MC codes which achieved data exchange from CAD data into a representation appropriate for MC codes. The author presents the application of the interface program for the automatic generation of the pulse reactor model for the MC code. The pulse reactor is transfered to the neutorn transport program successfully.
Yan Ma Ping Jun Xu Gang Xiao
The Institute of Applied Physics and Computational Mathematics, China
国际会议
上海
英文
623-627
2010-10-20（万方平台首次上网日期，不代表论文的发表时间）