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Radiation Protection Dosimetry Advance Access originally published online on October 12, 2006
Radiation Protection Dosimetry 2007 123(3):345-353; doi:10.1093/rpd/ncl150
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© The Author 2006. Published by Oxford University Press. All rights reserved. For Permissions, please email: journals.permissions@oxfordjournals.org

Treating voxel geometries in radiation protection dosimetry with a patched version of the Monte Carlo codes MCNP and MCNPX

K. W. Burn1, C. Daffara2, G. Gualdrini3,*, M. Pierantoni4 and P. Ferrari3

1 ENEA—Italian National Agency for Energy, New Technologies and the Environment, FIS-NUC, V.M.M. Sole 4, 40129 Bologna, Italy
2 INOA—Italian National Institute for Applied Optics Largo E.Fermi 6, 50125 Firenze, Italy
3 ENEA—Italian National Agency for Energy, New Technologies and the Environment, BAS-ION-IRP, V. dei Colli 16, 40136 Bologna, Italy
4 IOR, Rizzoli Hospital, IT Service, Salita S.Benedetto 1, 40136 Bologna, Italy

* Corresponding author: guald{at}bologna.enea.it

Received May 5, 2006, amended August 31, 2006, accepted September 6, 2006


  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.


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