We describe a new program to read a sequence of tomographic scans and prepare the geometry and material sections of an MCNP input file. Image processing techniques include contrast controls and mapping of grey scales to colour. The user interface provides several tools with which the user can associate a range of image intensities to an MCNP material. Materials are loaded from a library. A separate material assignment can be made to a pixel intensity or range of intensities when that intensity dominates the image boundaries; this material is assigned to all pixels with that intensity contiguous with the boundary. Material fractions are computed in a user-specified voxel grid overlaying the scans. New materials are defined by mixing the library materials using the fractions. The geometry can be written as an MCNP lattice or as individual cells. A combination algorithm can be used to join neighbouring cells with the same material.
[1]
J. F. Briesmeister.
MCNP-A General Monte Carlo N-Particle Transport Code
,
1993
.
[2]
Kenneth A. Van Riper.
INTERACTIVE 3D DISPLAY OF MCNP GEOMETRY MODELS
,
2001
.
[3]
M G Stabin,et al.
Monte Carlo MCNP-4B-based absorbed dose distribution estimates for patient-specific dosimetry.
,
2001,
Journal of nuclear medicine : official publication, Society of Nuclear Medicine.
[4]
D Franck,et al.
A NEW GRAPHICAL USER INTERFACE FOR FAST CONSTRUCTION OF COMPUTATION PHANTOMS AND MCNP CALCULATIONS: APPLICATION TO CALIBRATION OF IN VIVO MEASUREMENT SYSTEMS
,
2002,
Health physics.
[5]
N Gupta,et al.
Absorbed dose estimates to structures of the brain and head using a high-resolution voxel-based head phantom.
,
2001,
Medical physics.