OpenMC/openmc/geometry.py

from xml.etree import ElementTree as ET

import openmc
from openmc.clean_xml import *
from openmc.checkvalue import check_type

def reset_auto_ids():
    openmc.reset_auto_material_id()
    openmc.reset_auto_surface_id()
    openmc.reset_auto_cell_id()
    openmc.reset_auto_universe_id()


class Geometry(object):
    """Geometry representing a collection of surfaces, cells, and universes.

    Attributes
    ----------
    root_universe : openmc.universe.Universe
        Root universe which contains all others

    """

    def __init__(self):
        # Initialize Geometry class attributes
        self._root_universe = None
        self._offsets = {}

    @property
    def root_universe(self):
        return self._root_universe

    @root_universe.setter
    def root_universe(self, root_universe):
        check_type('root universe', root_universe, openmc.Universe)
        if root_universe._id != 0:
            msg = 'Unable to add root Universe "{0}" to Geometry since ' \
                  'it has ID="{1}" instead of ' \
                  'ID=0'.format(root_universe, root_universe._id)
            raise ValueError(msg)

        self._root_universe = root_universe

    def get_offset(self, path, filter_offset):
        """Returns the corresponding location in the results array for a given path and
        filter number. This is primarily intended to post-processing result when
        a distribcell filter is used.

        Parameters
        ----------
        path : list
            A list of IDs that form the path to the target. It should begin with
            0 for the base universe, and should cover every universe, cell, and
            lattice passed through. For the case of the lattice, a tuple should
            be provided to indicate which coordinates in the lattice should be
            entered. This should be in the form: (lat_id, i_x, i_y, i_z)
        filter_offset : int
            An integer that specifies which offset map the filter is using

        Returns
        -------
        offset : int
            Location in the results array for the path and filter

        """

        # Return memoize'd offset if possible
        if (path, filter_offset) in self._offsets:
            offset = self._offsets[(path, filter_offset)]

        # Begin recursive call to compute offset starting with the base Universe
        else:
            offset = self._root_universe.get_offset(path, filter_offset)
            self._offsets[(path, filter_offset)] = offset

        # Return the final offset
        return offset

    def get_all_cells(self):
        """Return all cells defined

        Returns
        -------
        list of openmc.universe.Cell
            Cells in the geometry

        """

        return self._root_universe.get_all_cells()

    def get_all_universes(self):
        """Return all universes defined

        Returns
        -------
        list of openmc.universe.Universe
            Universes in the geometry

        """

        return self._root_universe.get_all_universes()

    def get_all_nuclides(self):
        """Return all nuclides assigned to a material in the geometry

        Returns
        -------
        list of openmc.nuclide.Nuclide
            Nuclides in the geometry

        """

        nuclides = {}
        materials = self.get_all_materials()

        for material in materials:
            nuclides.update(material.get_all_nuclides())

        return nuclides

    def get_all_materials(self):
        """Return all materials assigned to a cell

        Returns
        -------
        list of openmc.material.Material
            Materials in the geometry

        """

        material_cells = self.get_all_material_cells()
        materials = set()

        for cell in material_cells:
            materials.add(cell._fill)

        return list(materials)

    def get_all_material_cells(self):
        all_cells = self.get_all_cells()
        material_cells = set()

        for cell_id, cell in all_cells.items():
            if cell._type == 'normal':
                material_cells.add(cell)

        return list(material_cells)

    def get_all_material_universes(self):
        """Return all universes composed of at least one non-fill cell

        Returns
        -------
        list of openmc.universe.Universe
            Universes with non-fill cells

        """

        all_universes = self.get_all_universes()
        material_universes = set()

        for universe_id, universe in all_universes.items():
            cells = universe._cells
            for cell_id, cell in cells.items():
                if cell._type == 'normal':
                    material_universes.add(universe)

        return list(material_universes)


class GeometryFile(object):
    """Geometry file used for an OpenMC simulation. Corresponds directly to the
    geometry.xml input file.

    Attributes
    ----------
    geometry : Geometry
        The geometry to be used

    """

    def __init__(self):
        # Initialize GeometryFile class attributes
        self._geometry = None
        self._geometry_file = ET.Element("geometry")

    @property
    def geometry(self):
        return self._geometry

    @geometry.setter
    def geometry(self, geometry):
        check_type('the geometry', geometry, Geometry)
        self._geometry = geometry

    def export_to_xml(self):
        """Create a geometry.xml file that can be used for a simulation.

        """

        root_universe = self._geometry._root_universe
        root_universe.create_xml_subelement(self._geometry_file)

        # Clean the indentation in the file to be user-readable
        sort_xml_elements(self._geometry_file)
        clean_xml_indentation(self._geometry_file)

        # Write the XML Tree to the geometry.xml file
        tree = ET.ElementTree(self._geometry_file)
        tree.write("geometry.xml", xml_declaration=True,
                             encoding='utf-8', method="xml")