Abstract
A Monte Carlo code system, MCNP-ANT, has been developed to simulate the transport of charged particles (<4MeV) and the neutron production in an accelerator neutron target as well as the neutron transport. The code system consists of a newly developed Monte Carlo code for charged particle transport and neutron production, and an existing neutron transport code, MCNP-4B. The available reactions in the code system are 2H(d,n)3He, 3H(d,n)4He, 3H(p,n)3He, 7Li(p,n)7Be and 45Sc(p,n)45Ti as main neutron sources. The calculated neutron spectra by MCNP-ANT were compared with the experimental ones in typical monoenergetic neutron fields.
Full Text
Published Version
Talk to us
Join us for a 30 min session where you can share your feedback and ask us any queries you have
Schedule a call