MCNP Input Generation

Creation of an MCNP input file is based on a rectangular grid overlying the image.

image with a grid drawn on top
An image with a grid drawn on top

The fraction of each material in a grid cell is computed from the relative number of pixels in each partition range. Materials with fractions below an ignore threshold T I (set on the MCNP Options dialog) are discarded and the remaining isotope fractions are renormalized so that their sum is 1. The algorithm searches for another grid cell with the same material. Materials are considered to be the same if the difference between each isotope fraction in the g rid cells being compared is less than the same-as threshold T S. If the same material is not found in another grid cell, a new material is created. The isotopes of the new material are added with the isotope fraction multiplied by the fraction of the material in the grid cell. The density of the new material is the fraction weighted sum of the constituent material densities.

A material must be assigned to each partition and to the edge material (if used). If not, the conversion algorithm will either terminate or issue a number of error messages reading Material N Not Found where N is some not necessarily meaningful number.

Position and Orientation. Each grid cell becomes an MCNP cell or lattice element bounded by planes defined by the grid. The absolute position of the cell is determined from the image position (DICOM tag 0020,0032), image orientation (0020,0037), and pixel spacing (0028,0030) elements in the DICOM file. For a single image, the cell dimension perpendicular to the image is set to the slice thickness (0018,0050) centered on the image position. For a 3D set of images from multiple files, the perpendicular faces of the cell are halfway between neighboring images. When a 3D model is constructed from multiple frames in a single file, the slice thickness gives the distance between the image planes. The current version of the program can only handle images with orientations along the X, Y, or Z axes. If such an orientation does not obtain, the default orientation of rows parallel to the X axis, columns to Y, and Z as the perpendicular direction is used.

The model position and orientation can be changed using the Transformation dialog. The translation option moves the model by adding X, Y, and Z values to the surface coefficients. A transformation number can be entered that is used in the surface definitions. The user must supply the transformation (TR) definition. Orientations along the X, Y, and Z axes can be used instead of the values in the DICOM file.

Multiple Files. When using multiple files, the files must constitute a valid 3D data set. If they do not, a request to generate MCNP input will process only the image in the active window.

Cropping. When cropping is in effect, the grid size is adjusted to match the cropping rectangle. The grid may enclose more than the cropping region if the crop boundary falls within a grid cell. The grid shown on top of the image display is not adjusted for cropping. When a 3D data set is used, the crop region must be the same for all images in the set.

Geometry. The geometry can be written as individual cells or as a lattice. The lattice uses a smaller number of cells but adds somewhat to the MCNP execution time. Writing the geometry as cells together with the cell combination algorithm described below should be advantageous when there are large connected regions containing the same material; the reduction in the total number cells and consequent decrease in MCNP execution time can be significant. Lattice geometry is preferable for a very large number of voxels.

Cell Numbers. MCNP cell numbers cannot be greater than 99,999. If this limit is exceeded when using cell geometry, lattice geometry should be used. Cells are numbered sequentially starting at 1.

Lattice. When lattice geometry is written, a single spherical cell and universe is created for each material. These cells are centered on the unit lattice cell and are larger than the extent of the unit cell. The lattice elements are filled with one of these single cell universes according to the material at the lattice element position.

A lattice is contained within another cell that bounds the extent of the 3D model. MCNP versions prior to MCNP4B prohibited the planes of the container cell from being coincident with planes of the lattice (including translations of those planes to each lattice position). Although later versions of MCNP (and MCNPX) recognize such coincident planes and handle them correctly in most cases, some users have reported problems in these cases. The MCNP Options dialog contains an to permit Coincident Surfaces or not. When the option is not checked, the surfaces of the container cell are moved towards the center of the model by approximately 0.5% of the extent of a voxel cell in the respective direction.

Other Items. The region external to the grid is defined a single void (material 0) cell. The user may wish to modify this outside cell to add other geometry external to the scan volume. The user is also responsible to adding source and tally specifications and other MCNP data items such as the particle type (MODE) and number of histories (NPS) to run. A file containing MCNP data items can be copied to the end of the generated input. The file location and whether or not to include it is controlled on the Append File dialog.

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