Development of neutron coincidence counting simulation code based on open-source simulation code

Nuclear Material Control Center has used the open-source simulation code called MCNP (A General Monte Carlo N-Particle Transport Code) version 4B, which is developed by Los Alamos National Laboratory, to develop NDA detectors based on coincidence neutron counting and to have an initial estimation for their performance. However, the MCNP does not have the function to simulate coincidence neutrons because the code can not generate several number of neutron particles associated with a fission event, and also it can not simulate neutrons emitted with Poisson distribution like behavior in nature. For these reasons, new functions were added to the code to result in output of detection point, detection time, incident direction, and neutron energy detected. And also a calculation program was developed to estimate coincidence counts from these outputs. Simulation results with the developed code were compared with actual results to adjust bias. In the paper, new functions added to the code and comparison of calculation data with actual measurement data are presented. (author)