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Address comments by @paulromano
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2 changed files with 5 additions and 20 deletions
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@ -183,13 +183,13 @@ the :class:`openmc.Cell` class::
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fuel.fill = uo2
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fuel.region = pellet
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# This cell will be filled with void on export to XML
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gap = openmc.Cell(region=pellet_gap)
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In this example, an instance of :class:`openmc.Material` is assigned to the
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:attr:`Cell.fill` attribute. One can also fill a cell with a :ref:`universe
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<usersguide_universes>` or :ref:`lattice <usersguide_lattices>`. If no fill
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is provided to a cell, it will be filled with void by default.
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is provided to a cell, it will be filled with void by default::
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# This cell will be filled with void on export to XML
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gap = openmc.Cell(region=pellet_gap)
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The classes :class:`Halfspace`, :class:`Intersection`, :class:`Union`, and
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:class:`Complement` and all instances of :class:`openmc.Region` and can be
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@ -459,7 +459,7 @@ calculation::
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Once a volume is set, and a cell is filled with a material or distributed
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materials, it is possible to use the :func:`~openmc.Cell.atoms` method to obtain
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a dictionary of nuclides and their total number of atoms in all instances
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of a cell (e.g. ``{'H1':1.0e22, 'O16':0.5e22, ...}``)::
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of a cell (e.g. ``{'H1': 1.0e22, 'O16': 0.5e22, ...}``)::
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cell = openmc.Cell(fill = u02)
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cell.volume = 17.0
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@ -191,21 +191,6 @@ class Cell(IDManagerMixin):
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@property
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def atoms(self):
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""" Get total number of atoms of each nuclide in the cell
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Returns
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-------
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atoms: collections.OrderedDict
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Dictionary in which keys are nuclides and values are the number of
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atoms. For example, {'H1':1.0e22, 'O16':0.5e22, ...}
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Raises
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------
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ValueError
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If total volume of the cell is not set
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ValueError
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If cell is filled with Universe, Lattice or Void
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"""
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if self._atoms is None:
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if self._volume is None:
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msg = ('Cannot calculate atom content becouse no volume '
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