Simulations of neutron transport at low energy: A comparison between GEANT and MCNP

N. Colonna, S. Altieri

Research output: Contribution to journalArticlepeer-review

15 Citations (Scopus)

Abstract

The use of the simulation tool GEANT for neutron transport at energies below 20 MeV is discussed, in particular with regard to shielding and dose calculations. The reliability of the GEANT/MICAP package for neutron transport in a wide energy range has been verified by comparing the results of simulations performed with this package in a wide energy range with the prediction of MCNP-4B, a code commonly used for neutron transport at low energy. A reasonable agreement between the results of the two codes is found for the neutron flux through a slab of material (iron and ordinary concrete), as well as for the dose released in soft tissue by neutrons. These results justify the use of the GEANT/MICAP code for neutron transport in a wide range of applications, including health physics problems.

Original languageEnglish
Pages (from-to)840-846
Number of pages7
JournalHealth Physics
Volume82
Issue number6
DOIs
Publication statusPublished - 2002
Externally publishedYes

Keywords

  • Attenuation
  • Dose
  • Monte Carlo
  • Neutrons

ASJC Scopus subject areas

  • Epidemiology
  • Radiology Nuclear Medicine and imaging
  • Health, Toxicology and Mutagenesis

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