Open AccessThe Use of the MCNP Code for Radiation Damage Calculations
Author(s) -
Hiwa Mohammad Qadr
Publication year - 2021
Publication title -
matematičeskaâ fizika i kompʹûternoe modelirovanie
Language(s) - English
Resource type - Journals
eISSN - 2587-6902
pISSN - 2587-6325
DOI - 10.15688/mpcm.jvolsu.2021.1.5
Subject(s) - monte carlo method , neutron , physics , neutron flux , nuclear physics , radiation , variance reduction , neutron transport , statistical physics , nuclear engineering , computational physics , mathematics , engineering , statistics
This work gives a detailed analysis of the result of Monte Carlo physics practical using MCNP. This paper describes basic concepts of the Monte Carlo theory of radiation transport calculation and also discusses the variance and the history method as used in Monte Carlo Problem solving. Therefore, in this exercise the MCNP code has been used to solve and estimate the number of neutron flux. The paper investigated the impact of the primary radiation damage in iron by the neutron energy irradiation. The established measurement of radiation damage is the displacements per atom (dpa) in matter as a function of neutron energy. The simulations were carried out to calculate the dpa cross section.