MCNPX-PoliMi for Nuclear Nonproliferation Applications

In the past few years, efforts to develop new measurement systems to support nuclear nonproliferation and homeland security have increased substantially. Monte Carlo radiation transport is one of the simulation methods of choice for the analysis of data from existing systems and for the design of new measurement systems; it allows for accurate description of geometries, detailed modeling of particle-nucleus interactions, and event-by-event detection analysis. This paper describes the use of the Monte Carlo code MCNPX-PoliMi for nuclear nonproliferation applications, with particular emphasis on the simulation of spontaneous and neutron-induced nuclear fission. In fact, of all possible neutron-nucleus interactions, neutron-induced fission is the most defining characteristic of special nuclear material (such as U-235 and Pu-239), which is the material of interest in nuclear-nonproliferation applications. The MCNP-PoliMi code was originally released from the Radiation Safety Shielding Center (RSSIC) at Oak Ridge National Laboratory in 2003; the MCNPX-PoliMi code contains many enhancements and is based on MCNPX ver. 2.7.0. MCNPX-PoliMi ver. 2.0 was released through RSICC in 2012 as a patch to MCNPX ver. 2.7.0 and as an executable.

POZZI Sara;  CLARKE Shaun D.;  WALSH W;  MILLER Eric;  DOLAN Jennifer;  FLASKA M.;  WIEGER B.;  ENQVIST Andreas;  PADOVANI Enrico;  MATTINGLY J.K.;  CHICHESTER David;  PEERANI Paolo; 

2012-10-15

ELSEVIER SCIENCE BV

JRC70879

0168-9002,   

http://www.sciencedirect.com/science/article/pii/S0168900212008224,    https://publications.jrc.ec.europa.eu/repository/handle/JRC70879,   

10.1016/j.nima.2012.07.040,