title: Treating voxel geometries in radiation protection dosimetry with a patched version of the Monte Carlo codes MCNP and MCNPX.
abstract: The question of Monte Carlo simulation of radiation transport in voxel geometries is addressed. Patched versions of the MCNP and MCNPX codes are developed aimed at transporting radiation both in the standard geometry mode and in the voxel geometry treatment. The patched code reads an unformatted FORTRAN file derived from DICOM format data and uses special subroutines to handle voxel-to-voxel radiation transport. The various phases of the development of the methodology are discussed together with the new input options. Examples are given of employment of the code in internal and external dosimetry and comparisons with results from other groups are reported.
authors.affiliation: ENEA-Italian National Agency for Energy, New Technologies and the Environment, FIS-NUC, V.M.M. Sole 4, 40129 Bologna, Italy.
authors.forename: K W
authors.initials: KW
authors.lastname: Burn
authors.forename: C
authors.initials: C
authors.lastname: Daffara
authors.forename: G
authors.initials: G
authors.lastname: Gualdrini
authors.forename: M
authors.initials: M
authors.lastname: Pierantoni
authors.forename: P
authors.initials: P
authors.lastname: Ferrari
mesh_terms: D001822:Body Burden; D003198:Computer Simulation; D006801:Humans; D008954:Models, Biological; D009010:Monte Carlo Method; D011381:Programming Languages; D011829:Radiation Dosage; D011835:Radiation Protection / Q000379:methods*; D012062:Relative Biological Effectiveness; D012984:Software; D014915:Whole-Body Counting / Q000379:methods*