openmc.Geometry

class openmc.Geometry(root: UniverseBase | Iterable[Cell] | None = None, merge_surfaces: bool = False, surface_precision: int = 10)[source]

Geometry representing a collection of surfaces, cells, and universes.

Parameters:

root (openmc.UniverseBase or Iterable of openmc.Cell, optional) – Root universe which contains all others, or an iterable of cells that should be used to create a root universe.

Variables:

root_universe (openmc.UniverseBase) – Root universe which contains all others
bounding_box (openmc.BoundingBox) – Lower-left and upper-right coordinates of an axis-aligned bounding box of the universe.
merge_surfaces (bool) – Whether to remove redundant surfaces when the geometry is exported.
surface_precision (int) – Number of decimal places to round to for comparing the coefficients of surfaces for considering them topologically equivalent.

add_volume_information(volume_calc)[source]

Add volume information from a stochastic volume calculation.

Parameters:: volume_calc (openmc.VolumeCalculation) – Results from a stochastic volume calculation

clone() → Geometry[source]: Create a copy of this geometry with new unique IDs for all of its enclosed materials, surfaces, cells, universes and lattices.

determine_paths(instances_only=False)[source]

Determine paths through CSG tree for cells and materials.

This method recursively traverses the CSG tree to determine each unique path that reaches every cell and material. The paths are stored in the Cell.paths and Material.paths attributes.

Parameters:: instances_only (bool, optional) – If true, this method will only determine the number of instances of each cell and material.

export_to_xml(path='geometry.xml', remove_surfs=False)[source]

Export geometry to an XML file.

Parameters:

path (str) – Path to file to write. Defaults to ‘geometry.xml’.
remove_surfs (bool) –
Whether or not to remove redundant surfaces from the geometry when exporting

Added in version 0.12.

find(point) → list[source]

Find cells/universes/lattices which contain a given point

Parameters:: point (3-tuple of float) – Cartesian coordinates of the point
Returns:: Sequence of universes, cells, and lattices which are traversed to find the given point
Return type:: list

classmethod from_xml(path: PathLike = 'geometry.xml', materials: PathLike | 'openmc.Materials' | None = 'materials.xml') → Geometry[source]

Generate geometry from XML file

Parameters:

path (PathLike, optional) – Path to geometry XML file
materials (openmc.Materials or PathLike) – Materials used to assign to cells. If PathLike, an attempt is made to generate materials from the provided xml file.

Returns:

Geometry object

Return type:

openmc.Geometry

classmethod from_xml_element(elem, materials=None) → Geometry[source]

Generate geometry from an XML element

Parameters:

elem (lxml.etree._Element) – XML element
materials (openmc.Materials or None) – Materials used to assign to cells. If None, an attempt is made to generate it from the materials.xml file.

Returns:

Geometry object

Return type:

openmc.Geometry

get_all_cells() → dict[int, Cell][source]

Return all cells in the geometry.

Returns:: Dictionary mapping cell IDs to openmc.Cell instances
Return type:: dict

get_all_lattices() → dict[int, Lattice][source]

Return all lattices defined

Returns:: Dictionary mapping lattice IDs to openmc.Lattice instances
Return type:: dict

get_all_material_cells() → dict[int, Cell][source]

Return all cells filled by a material

Returns:: Dictionary mapping cell IDs to openmc.Cell instances that are filled with materials or distributed materials.
Return type:: dict

get_all_material_universes() → dict[int, Universe][source]

Return all universes having at least one material-filled cell.

This method can be used to find universes that have at least one cell that is filled with a material or is void.

Returns:: Dictionary mapping universe IDs to openmc.Universe instances with at least one material-filled cell
Return type:: dict

get_all_materials() → dict[int, Material][source]

Return all materials within the geometry.

Returns:: Dictionary mapping material IDs to openmc.Material instances
Return type:: dict

get_all_nuclides() → list[str][source]

Return all nuclides within the geometry.

Returns:: Sorted list of all nuclides in materials appearing in the geometry
Return type:: list

get_all_surfaces() → dict[int, Surface][source]

Return all surfaces used in the geometry

Returns:: Dictionary mapping surface IDs to openmc.Surface instances
Return type:: dict

get_all_universes() → dict[int, Universe][source]

Return all universes in the geometry.

Returns:: Dictionary mapping universe IDs to openmc.Universe instances
Return type:: dict

get_cells_by_fill_name(name, case_sensitive=False, matching=False) → list[Cell][source]

Return a list of cells with fills with matching names.

Parameters:

name (str) – The name to match
case_sensitive (bool) – Whether to distinguish upper and lower case letters in each cell’s name (default is False)
matching (bool) – Whether the names must match completely (default is False)

Returns:

Cells with fills matching the queried name

Return type:

list of openmc.Cell

get_cells_by_name(name, case_sensitive=False, matching=False) → list[Cell][source]

Return a list of cells with matching names.

Parameters:

name (str) – The name to search match
case_sensitive (bool) – Whether to distinguish upper and lower case letters in each cell’s name (default is False)
matching (bool) – Whether the names must match completely (default is False)

Returns:

Cells matching the queried name

Return type:

list of openmc.Cell

get_instances(paths) → int | list[int][source]

Return the instance number(s) for a cell/material in a geometry path.

The instance numbers are used as indices into distributed material/temperature arrays and tally distribcell filter arrays.

Parameters:: paths (str or iterable of str) – The path traversed through the CSG tree to reach a cell or material instance. For example, ‘u0->c10->l20(2,2,1)->u5->c5’ would indicate the cell instance whose first level is universe 0 and cell 10, second level is lattice 20 position (2,2,1), and third level is universe 5 and cell 5.
Returns:: Instance number(s) for the given path(s)
Return type:: int or list of int

get_lattices_by_name(name, case_sensitive=False, matching=False) → list[Lattice][source]

Return a list of lattices with matching names.

Parameters:

name (str) – The name to match
case_sensitive (bool) – Whether to distinguish upper and lower case letters in each lattice’s name (default is False)
matching (bool) – Whether the names must match completely (default is False)

Returns:

Lattices matching the queried name

Return type:

list of openmc.Lattice

get_materials_by_name(name, case_sensitive=False, matching=False) → list[Material][source]

Return a list of materials with matching names.

Parameters:

name (str) – The name to match
case_sensitive (bool) – Whether to distinguish upper and lower case letters in each material’s name (default is False)
matching (bool) – Whether the names must match completely (default is False)

Returns:

Materials matching the queried name

Return type:

list of openmc.Material

get_surfaces_by_name(name, case_sensitive=False, matching=False) → list[Surface][source]

Return a list of surfaces with matching names.

Added in version 0.13.3.

Parameters:

name (str) – The name to search match
case_sensitive (bool) – Whether to distinguish upper and lower case letters in each surface’s name (default is False)
matching (bool) – Whether the names must match completely (default is False)

Returns:

Surfaces matching the queried name

Return type:

list of openmc.Surface

get_universes_by_name(name, case_sensitive=False, matching=False) → list[Universe][source]

Return a list of universes with matching names.

Parameters:

name (str) – The name to match
case_sensitive (bool) – Whether to distinguish upper and lower case letters in each universe’s name (default is False)
matching (bool) – Whether the names must match completely (default is False)

Returns:

Universes matching the queried name

Return type:

list of openmc.Universe

plot(*args, **kwargs)[source]

Display a slice plot of the geometry.

Added in version 0.14.0.

Parameters:

origin (iterable of float) – Coordinates at the origin of the plot. If left as None, the center of the bounding box will be used to attempt to ascertain the origin with infinite values being replaced by 0.
width (iterable of float) – Width of the plot in each basis direction. If left as none then the width of the bounding box will be used to attempt to ascertain the plot width. Defaults to (10, 10) if the bounding box contains inf values.
pixels (Iterable of int or int) – If an iterable of ints is provided then this directly sets the number of pixels to use in each basis direction. If a single int is provided then this sets the total number of pixels in the plot and the number of pixels in each basis direction is calculated from this total and the image aspect ratio based on the width argument.
basis ({'xy', 'xz', 'yz'}) – The basis directions for the plot
color_by ({'cell', 'material'}) – Indicate whether the plot should be colored by cell or by material
colors (dict) –
Assigns colors to specific materials or cells. Keys are instances of Cell or Material and values are RGB 3-tuples, RGBA 4-tuples, or strings indicating SVG color names. Red, green, blue, and alpha should all be floats in the range [0.0, 1.0], for example:
```
# Make water blue
water = openmc.Cell(fill=h2o)
universe.plot(..., colors={water: (0., 0., 1.))
```
seed (int) – Seed for the random number generator
openmc_exec (str) – Path to OpenMC executable.
axes (matplotlib.Axes) –
Axes to draw to

Added in version 0.13.1.
legend (bool) –
Whether a legend showing material or cell names should be drawn

Added in version 0.14.0.
axis_units ({'km', 'm', 'cm', 'mm'}) –
Units used on the plot axis

Added in version 0.14.0.
outline (bool or str) –
Whether outlines between color boundaries should be drawn. If set to ‘only’, only outlines will be drawn.

Added in version 0.14.0.
show_overlaps (bool) – Indicate whether or not overlapping regions are shown. Default is False.
overlap_color (Iterable of int or str) – Color to apply to overlapping regions. Default is red.
n_samples (int, optional) – The number of source particles to sample and add to plot. Defaults to None which doesn’t plot any particles on the plot.
plane_tolerance (float) – When plotting a plane the source locations within the plane +/- the plane_tolerance will be included and those outside of the plane_tolerance will not be shown
legend_kwargs (dict) –
Keyword arguments passed to matplotlib.pyplot.legend().

Added in version 0.14.0.
source_kwargs (dict, optional) – Keyword arguments passed to matplotlib.pyplot.scatter().
contour_kwargs (dict, optional) – Keyword arguments passed to matplotlib.pyplot.contour().
**kwargs – Keyword arguments passed to matplotlib.pyplot.imshow().

Returns:

Axes containing resulting image

Return type:

matplotlib.axes.Axes

remove_redundant_surfaces() → dict[int, Surface][source]

Remove and return all of the redundant surfaces.

Uses surface_precision attribute of Geometry instance for rounding and comparing surface coefficients.

Added in version 0.12.

Returns:: Dictionary whose keys are the ID of a redundant surface and whose values are the topologically equivalent openmc.Surface that should replace it.
Return type:: redundant_surfaces

to_xml_element(remove_surfs=False) → Element[source]

Creates a ‘geometry’ element to be written to an XML file.

Parameters:: remove_surfs (bool) – Whether or not to remove redundant surfaces from the geometry when exporting