Simulation of an ICF reaction as a source in an MCNPX 2.6.0 simulation for a fission-fusion hybrid reactor
Document Type : Research Paper
Authors
Abstract
In this work, we had examined 7 different types of neutron sources -An isotropic point source, sending 14.1(MeV) monoenergetic neutrons ,An spherical source with 30 micron radius sending 14.1 (MeV) monoenergetic neutrons, and A 30 micron radius spherical source having Gaussian fusion energy spectrum from a d-t fusion, with different bandwidths -1,10,50,100 and 1000 KeV and investigated 5 different important factors for a reactor –neutron surface current, neutron surface flux, neutron volume flux, energies deposited in cells and deposited fission energies in the fission cell, using most current Monte Carlo Calculation Code, MCNPX 2.6.0. The results show that we can simulate an ICF table as an isotropic point source sending monoenergetic neutrons, with no essential errors.
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