diff --git a/docs/source/pythonapi/capi.rst b/docs/source/pythonapi/capi.rst index 6e4e860fa..457f05320 100644 --- a/docs/source/pythonapi/capi.rst +++ b/docs/source/pythonapi/capi.rst @@ -33,7 +33,9 @@ Classes :nosignatures: :template: myclass.rst - openmc.capi.CellView - openmc.capi.MaterialView - openmc.capi.NuclideView - openmc.capi.TallyView + openmc.capi.Cell + openmc.capi.EnergyFilter + openmc.capi.MaterialFilter + openmc.capi.Material + openmc.capi.Nuclide + openmc.capi.Tally diff --git a/openmc/arithmetic.py b/openmc/arithmetic.py index 57afacac3..6a781cb5a 100644 --- a/openmc/arithmetic.py +++ b/openmc/arithmetic.py @@ -108,9 +108,9 @@ class CrossNuclide(object): Parameters ---------- - left_nuclide : Nuclide or CrossNuclide + left_nuclide : openmc.Nuclide or CrossNuclide The left nuclide in the outer product - right_nuclide : Nuclide or CrossNuclide + right_nuclide : openmc.Nuclide or CrossNuclide The right nuclide in the outer product binary_op : str The tally arithmetic binary operator (e.g., '+', '-', etc.) used to @@ -118,9 +118,9 @@ class CrossNuclide(object): Attributes ---------- - left_nuclide : Nuclide or CrossNuclide + left_nuclide : openmc.Nuclide or CrossNuclide The left nuclide in the outer product - right_nuclide : Nuclide or CrossNuclide + right_nuclide : openmc.Nuclide or CrossNuclide The right nuclide in the outer product binary_op : str The tally arithmetic binary operator (e.g., '+', '-', etc.) used to @@ -510,7 +510,7 @@ class AggregateNuclide(object): Parameters ---------- - nuclides : Iterable of str or Nuclide or CrossNuclide + nuclides : Iterable of str or openmc.Nuclide or CrossNuclide The nuclides included in the aggregation aggregate_op : str The tally aggregation operator (e.g., 'sum', 'avg', etc.) used @@ -518,7 +518,7 @@ class AggregateNuclide(object): Attributes ---------- - nuclides : Iterable of str or Nuclide or CrossNuclide + nuclides : Iterable of str or openmc.Nuclide or CrossNuclide The nuclides included in the aggregation aggregate_op : str The tally aggregation operator (e.g., 'sum', 'avg', etc.) used diff --git a/openmc/capi/__init__.py b/openmc/capi/__init__.py index e0f1c8b27..d302001c9 100644 --- a/openmc/capi/__init__.py +++ b/openmc/capi/__init__.py @@ -13,6 +13,7 @@ objects in the :mod:`openmc.capi` subpackage, for example: """ from ctypes import CDLL +import os import sys from warnings import warn @@ -25,10 +26,21 @@ if sys.platform == 'darwin': else: _suffix = 'so' -# Open shared library -_filename = pkg_resources.resource_filename( - __name__, 'libopenmc.{}'.format(_suffix)) -_dll = CDLL(_filename) +if os.environ.get('READTHEDOCS', None) != 'True': + # Open shared library + _filename = pkg_resources.resource_filename( + __name__, 'libopenmc.{}'.format(_suffix)) + _dll = CDLL(_filename) +else: + # For documentation builds, we don't actually have the shared library + # available. Instead, we create a mock object so that when the modules + # within the openmc.capi package try to configure arguments and return + # values for symbols, no errors occur + try: + from unittest.mock import Mock + except ImportError: + from mock import Mock + _dll = Mock() from .error import * from .core import * diff --git a/openmc/capi/cell.py b/openmc/capi/cell.py index 11b0e2199..62823c0ba 100644 --- a/openmc/capi/cell.py +++ b/openmc/capi/cell.py @@ -6,11 +6,11 @@ import numpy as np from numpy.ctypeslib import as_array from . import _dll -from .core import _View +from .core import _FortranObjectWithID from .error import _error_handler -from .material import MaterialView +from .material import Material -__all__ = ['CellView', 'cells'] +__all__ = ['Cell', 'cells'] # Cell functions _dll.openmc_cell_get_id.argtypes = [c_int32, POINTER(c_int32)] @@ -31,11 +31,11 @@ _dll.openmc_get_cell_index.restype = c_int _dll.openmc_get_cell_index.errcheck = _error_handler -class CellView(_View): - """View of a cell. +class Cell(_FortranObjectWithID): + """Cell stored internally. - This class exposes a cell that is stored internally in the OpenMC solver. To - obtain a view of a cell with a given ID, use the + This class exposes a cell that is stored internally in the OpenMC + library. To obtain a view of a cell with a given ID, use the :data:`openmc.capi.nuclides` mapping. Parameters @@ -72,9 +72,9 @@ class CellView(_View): if fill_type.value == 1: if n.value > 1: - return [MaterialView(i) for i in indices[:n.value]] + return [Material(index=i) for i in indices[:n.value]] else: - return MaterialView(indices[0]) + return Material(index=indices[0]) else: raise NotImplementedError @@ -84,7 +84,7 @@ class CellView(_View): n = len(fill) indices = (c_int*n)(*(m._index for m in fill)) _dll.openmc_cell_set_fill(self._index, 1, 1, indices) - elif isinstance(fill, MaterialView): + elif isinstance(fill, Material): materials = [fill] indices = (c_int*1)(fill._index) _dll.openmc_cell_set_fill(self._index, 1, 1, indices) @@ -108,12 +108,16 @@ class CellView(_View): class _CellMapping(Mapping): def __getitem__(self, key): index = c_int32() - _dll.openmc_get_cell_index(key, index) - return CellView(index.value) + try: + _dll.openmc_get_cell_index(key, index) + except (AllocationError, InvalidIDError) as e: + # __contains__ expects a KeyError to work correctly + raise KeyError(str(e)) + return Cell(index.value) def __iter__(self): for i in range(len(self)): - yield CellView(i + 1).id + yield Cell(i + 1).id def __len__(self): return c_int32.in_dll(_dll, 'n_cells').value diff --git a/openmc/capi/core.py b/openmc/capi/core.py index 3d74ad5cd..0d8ff5c43 100644 --- a/openmc/capi/core.py +++ b/openmc/capi/core.py @@ -172,12 +172,12 @@ class _DLLGlobal(object): self.ctype.in_dll(_dll, self.name).value = value -class _View(object): +class _FortranObject(object): def __repr__(self): return "{}[{}]".format(type(self).__name__, self._index) -class _ViewWithID(_View): +class _FortranObjectWithID(_FortranObject): def __init__(self, uid=None, new=True, index=None): # Creating the object has already been handled by __new__. In the # initializer, all we do is make sure that the object returned has an ID diff --git a/openmc/capi/error.py b/openmc/capi/error.py index 7ebf2cb64..98d43ae46 100644 --- a/openmc/capi/error.py +++ b/openmc/capi/error.py @@ -39,9 +39,6 @@ class InvalidTypeError(Error): """Tried to perform an operation on the wrong type.""" -_errmsg = (c_char*256).in_dll(_dll, 'openmc_err_msg') - - def _error_handler(err, func, args): """Raise exception according to error code.""" @@ -50,7 +47,8 @@ def _error_handler(err, func, args): return c_int.in_dll(_dll, s).value # Get error message set by OpenMC library - msg = _errmsg.value.decode() + errmsg = (c_char*256).in_dll(_dll, 'openmc_err_msg') + msg = errmsg.value.decode() # Raise exception type corresponding to error code if err == errcode('e_allocate'): diff --git a/openmc/capi/filter.py b/openmc/capi/filter.py index 1fce75702..ac9de8a43 100644 --- a/openmc/capi/filter.py +++ b/openmc/capi/filter.py @@ -7,17 +7,17 @@ import numpy as np from numpy.ctypeslib import as_array from . import _dll -from .core import _ViewWithID +from .core import _FortranObjectWithID from .error import _error_handler, AllocationError, InvalidIDError -from .material import MaterialView +from .material import Material -__all__ = ['FilterView', 'AzimuthalFilterView', 'CellFilterView', - 'CellbornFilterView', 'CellfromFilterView', 'DistribcellFilterView', - 'DelayedGroupFilterView', 'EnergyFilterView', 'EnergyoutFilterView', - 'EnergyFunctionFilterView', 'MaterialFilterView', 'MeshFilterView', - 'MuFilterView', 'PolarFilterView', 'SurfaceFilterView', - 'UniverseFilterView', 'filters'] +__all__ = ['Filter', 'AzimuthalFilter', 'CellFilter', + 'CellbornFilter', 'CellfromFilter', 'DistribcellFilter', + 'DelayedGroupFilter', 'EnergyFilter', 'EnergyoutFilter', + 'EnergyFunctionFilter', 'MaterialFilter', 'MeshFilter', + 'MuFilter', 'PolarFilter', 'SurfaceFilter', + 'UniverseFilter', 'filters'] # Tally functions _dll.openmc_energy_filter_get_bins.argtypes = [ @@ -57,7 +57,7 @@ _dll.openmc_mesh_filter_set_mesh.restype = c_int _dll.openmc_mesh_filter_set_mesh.errcheck = _error_handler -class FilterView(_ViewWithID): +class Filter(_FortranObjectWithID): __instances = WeakValueDictionary() def __new__(cls, filter_type, uid=None, new=True, index=None): @@ -102,7 +102,7 @@ class FilterView(_ViewWithID): _dll.openmc_filter_set_id(self._index, filter_id) -class EnergyFilterView(FilterView): +class EnergyFilter(Filter): def __new__(cls, bins=None, uid=None, new=True, index=None): return super().__new__(cls, b'energy', uid, new, index) @@ -123,47 +123,47 @@ class EnergyFilterView(FilterView): self._index, len(energies), energies_p) -class EnergyoutFilterView(FilterView): +class EnergyoutFilter(Filter): def __new__(cls, bins=None, uid=None, new=True, index=None): return super().__new__(cls, b'energyout', uid, new, index) -class AzimuthalFilterView(FilterView): +class AzimuthalFilter(Filter): def __new__(cls, bins=None, uid=None, new=True, index=None): return super().__new__(cls, b'azimuthal', uid, new, index) -class CellFilterView(FilterView): +class CellFilter(Filter): def __new__(cls, bins=None, uid=None, new=True, index=None): return super().__new__(cls, b'cell', uid, new, index) -class CellbornFilterView(FilterView): +class CellbornFilter(Filter): def __new__(cls, bins=None, uid=None, new=True, index=None): return super().__new__(cls, b'cellborn', uid, new, index) -class CellfromFilterView(FilterView): +class CellfromFilter(Filter): def __new__(cls, bins=None, uid=None, new=True, index=None): return super().__new__(cls, b'cellfrom', uid, new, index) -class DelayedGroupFilterView(FilterView): +class DelayedGroupFilter(Filter): def __new__(cls, bins=None, uid=None, new=True, index=None): return super().__new__(cls, b'delayedgroup', uid, new, index) -class DistribcellFilterView(FilterView): +class DistribcellFilter(Filter): def __new__(cls, bins=None, uid=None, new=True, index=None): return super().__new__(cls, b'distribcell', uid, new, index) -class EnergyFunctionFilterView(FilterView): +class EnergyFunctionFilter(Filter): def __new__(cls, bins=None, uid=None, new=True, index=None): return super().__new__(cls, b'energyfunction', uid, new, index) -class MaterialFilterView(FilterView): +class MaterialFilter(Filter): def __new__(cls, bins=None, uid=None, new=True, index=None): return super().__new__(cls, b'material', uid, new, index) @@ -177,7 +177,7 @@ class MaterialFilterView(FilterView): materials = POINTER(c_int32)() n = c_int32() _dll.openmc_material_filter_get_bins(self._index, materials, n) - return [MaterialView(index=materials[i]) for i in range(n.value)] + return [Material(index=materials[i]) for i in range(n.value)] @bins.setter def bins(self, materials): @@ -188,47 +188,47 @@ class MaterialFilterView(FilterView): _dll.openmc_material_filter_set_bins(self._index, n, bins) -class MeshFilterView(FilterView): +class MeshFilter(Filter): def __new__(cls, bins=None, uid=None, new=True, index=None): return super().__new__(cls, b'mesh', uid, new, index) -class MuFilterView(FilterView): +class MuFilter(Filter): def __new__(cls, bins=None, uid=None, new=True, index=None): return super().__new__(cls, b'mu', uid, new, index) -class PolarFilterView(FilterView): +class PolarFilter(Filter): def __new__(cls, bins=None, uid=None, new=True, index=None): return super().__new__(cls, b'polar', uid, new, index) -class SurfaceFilterView(FilterView): +class SurfaceFilter(Filter): def __new__(cls, bins=None, uid=None, new=True, index=None): return super().__new__(cls, b'surface', uid, new, index) -class UniverseFilterView(FilterView): +class UniverseFilter(Filter): def __new__(cls, bins=None, uid=None, new=True, index=None): return super().__new__(cls, b'universe', uid, new, index) _FILTER_TYPE_MAP = { - 'azimuthal': AzimuthalFilterView, - 'cell': CellFilterView, - 'cellborn': CellbornFilterView, - 'cellfrom': CellfromFilterView, - 'delayedgroup': DelayedGroupFilterView, - 'distribcell': DistribcellFilterView, - 'energy': EnergyFilterView, - 'energyout': EnergyoutFilterView, - 'energyfunction': EnergyFunctionFilterView, - 'material': MaterialFilterView, - 'mesh': MeshFilterView, - 'mu': MuFilterView, - 'polar': PolarFilterView, - 'surface': SurfaceFilterView, - 'universe': UniverseFilterView, + 'azimuthal': AzimuthalFilter, + 'cell': CellFilter, + 'cellborn': CellbornFilter, + 'cellfrom': CellfromFilter, + 'delayedgroup': DelayedGroupFilter, + 'distribcell': DistribcellFilter, + 'energy': EnergyFilter, + 'energyout': EnergyoutFilter, + 'energyfunction': EnergyFunctionFilter, + 'material': MaterialFilter, + 'mesh': MeshFilter, + 'mu': MuFilter, + 'polar': PolarFilter, + 'surface': SurfaceFilter, + 'universe': UniverseFilter, } diff --git a/openmc/capi/material.py b/openmc/capi/material.py index 2f2f7b3f8..a9cc9c148 100644 --- a/openmc/capi/material.py +++ b/openmc/capi/material.py @@ -5,12 +5,12 @@ from weakref import WeakValueDictionary import numpy as np from numpy.ctypeslib import as_array -from . import _dll, NuclideView -from .core import _ViewWithID +from . import _dll, Nuclide +from .core import _FortranObjectWithID from .error import _error_handler, AllocationError, InvalidIDError -__all__ = ['MaterialView', 'materials'] +__all__ = ['Material', 'materials'] # Material functions _dll.openmc_extend_materials.argtypes = [c_int32, POINTER(c_int32), POINTER(c_int32)] @@ -43,11 +43,11 @@ _dll.openmc_material_set_id.restype = c_int _dll.openmc_material_set_id.errcheck = _error_handler -class MaterialView(_ViewWithID): - """View of a material. +class Material(_FortranObjectWithID): + """Material stored internally. This class exposes a material that is stored internally in the OpenMC - solver. To obtain a view of a material with a given ID, use the + library. To obtain a view of a material with a given ID, use the :data:`openmc.capi.materials` mapping. Parameters @@ -141,7 +141,7 @@ class MaterialView(_ViewWithID): _dll.openmc_material_get_densities(self._index, nuclides, densities, n) # Convert to appropriate types and return - nuclide_list = [NuclideView(nuclides[i]).name for i in range(n.value)] + nuclide_list = [Nuclide(nuclides[i]).name for i in range(n.value)] density_array = as_array(densities, (n.value,)) return nuclide_list, density_array @@ -196,14 +196,14 @@ class _MaterialMapping(Mapping): index = c_int32() try: _dll.openmc_get_material_index(key, index) - except InvalidIDError as e: + except (AllocationError, InvalidIDError) as e: # __contains__ expects a KeyError to work correctly raise KeyError(str(e)) - return MaterialView(index=index.value) + return Material(index=index.value) def __iter__(self): for i in range(len(self)): - yield MaterialView(index=i + 1).id + yield Material(index=i + 1).id def __len__(self): return c_int32.in_dll(_dll, 'n_materials').value diff --git a/openmc/capi/nuclide.py b/openmc/capi/nuclide.py index c5b61a623..54d131498 100644 --- a/openmc/capi/nuclide.py +++ b/openmc/capi/nuclide.py @@ -6,11 +6,11 @@ import numpy as np from numpy.ctypeslib import as_array from . import _dll -from .core import _View +from .core import _FortranObject from .error import _error_handler, DataError, AllocationError -__all__ = ['NuclideView', 'nuclides', 'load_nuclide'] +__all__ = ['Nuclide', 'nuclides', 'load_nuclide'] # Nuclide functions _dll.openmc_get_nuclide_index.argtypes = [c_char_p, POINTER(c_int)] @@ -36,8 +36,8 @@ def load_nuclide(name): _dll.openmc_load_nuclide(name.encode()) -class NuclideView(_View): - """View of a nuclide. +class Nuclide(_FortranObject): + """Nuclide stored internally. This class exposes a nuclide that is stored internally in the OpenMC solver. To obtain a view of a nuclide with a given name, use the @@ -86,11 +86,11 @@ class _NuclideMapping(Mapping): except (DataError, AllocationError) as e: # __contains__ expects a KeyError to work correctly raise KeyError(str(e)) - return NuclideView(index.value) + return Nuclide(index.value) def __iter__(self): for i in range(len(self)): - yield NuclideView(i + 1).name + yield Nuclide(i + 1).name def __len__(self): return c_int.in_dll(_dll, 'n_nuclides').value diff --git a/openmc/capi/tally.py b/openmc/capi/tally.py index 89816e039..cb37e6971 100644 --- a/openmc/capi/tally.py +++ b/openmc/capi/tally.py @@ -4,13 +4,13 @@ from weakref import WeakValueDictionary from numpy.ctypeslib import as_array -from . import _dll, NuclideView -from .core import _ViewWithID +from . import _dll, Nuclide +from .core import _FortranObjectWithID from .error import _error_handler, AllocationError, InvalidIDError from .filter import _get_filter -__all__ = ['TallyView', 'tallies'] +__all__ = ['Tally', 'tallies'] # Tally functions _dll.openmc_get_tally_index.argtypes = [c_int32, POINTER(c_int32)] @@ -51,11 +51,11 @@ _dll.openmc_tally_set_type.restype = c_int _dll.openmc_tally_set_type.errcheck = _error_handler -class TallyView(_ViewWithID): - """View of a tally. +class Tally(_FortranObjectWithID): + """Tally stored internally. This class exposes a tally that is stored internally in the OpenMC - solver. To obtain a view of a tally with a given ID, use the + library. To obtain a view of a tally with a given ID, use the :data:`openmc.capi.tallies` mapping. Parameters @@ -64,7 +64,7 @@ class TallyView(_ViewWithID): Unique ID of the tally new : bool When `index` is None, this argument controls whether a new object is - created or a view to an existing object is returned. + created or a view of an existing object is returned. index : int or None Index in the `tallies` array. @@ -73,7 +73,7 @@ class TallyView(_ViewWithID): id : int ID of the tally filters : list - List of views to tally filters + List of tally filters nuclides : list of str List of nuclides to score results for results : numpy.ndarray @@ -135,7 +135,7 @@ class TallyView(_ViewWithID): nucs = POINTER(c_int)() n = c_int() _dll.openmc_tally_get_nuclides(self._index, nucs, n) - return [NuclideView(nucs[i]).name if nucs[i] > 0 else 'total' + return [Nuclide(nucs[i]).name if nucs[i] > 0 else 'total' for i in range(n.value)] @property @@ -191,14 +191,14 @@ class _TallyMapping(Mapping): index = c_int32() try: _dll.openmc_get_tally_index(key, index) - except InvalidIDError as e: + except (AllocationError, InvalidIDError) as e: # __contains__ expects a KeyError to work correctly raise KeyError(str(e)) - return TallyView(index=index.value) + return Tally(index=index.value) def __iter__(self): for i in range(len(self)): - yield TallyView(index=i + 1).id + yield Tally(index=i + 1).id def __len__(self): return c_int32.in_dll(_dll, 'n_tallies').value diff --git a/openmc/mgxs/library.py b/openmc/mgxs/library.py index 7ebe6cb8c..3cdd0fbf5 100644 --- a/openmc/mgxs/library.py +++ b/openmc/mgxs/library.py @@ -606,7 +606,7 @@ class Library(object): Parameters ---------- - domain : Material or Cell or Universe or Integral + domain : openmc.Material or openmc.Cell or openmc.Universe or openmc.Mesh or Integral The material, cell, or universe object of interest (or its ID) mgxs_type : {'total', 'transport', 'nu-transport', 'absorption', 'capture', 'fission', 'nu-fission', 'kappa-fission', 'scatter', 'nu-scatter', 'scatter matrix', 'nu-scatter matrix', 'multiplicity matrix', 'nu-fission matrix', chi', 'chi-prompt', 'inverse-velocity', 'prompt-nu-fission', 'prompt-nu-fission matrix', 'delayed-nu-fission', 'delayed-nu-fission matrix', 'chi-delayed', 'beta'} The type of multi-group cross section object to return @@ -668,7 +668,7 @@ class Library(object): Returns ------- - Library + openmc.mgxs.Library A new multi-group cross section library condensed to the group structure of interest @@ -880,7 +880,7 @@ class Library(object): Returns ------- - Library + openmc.mgxs.Library A Library object loaded from the pickle binary file See also diff --git a/openmc/mgxs/mdgxs.py b/openmc/mgxs/mdgxs.py index 055bf267d..d3f1a0646 100644 --- a/openmc/mgxs/mdgxs.py +++ b/openmc/mgxs/mdgxs.py @@ -71,7 +71,7 @@ class MDGXS(MGXS): Reaction type (e.g., 'chi-delayed', 'beta', etc.) by_nuclide : bool If true, computes cross sections for each nuclide in domain - domain : Material or Cell or Universe or Mesh + domain : openmc.Material or openmc.Cell or openmc.Universe or openmc.Mesh Domain for spatial homogenization domain_type : {'material', 'cell', 'distribcell', 'universe', 'mesh'} Domain type for spatial homogenization @@ -948,7 +948,7 @@ class ChiDelayed(MDGXS): Reaction type (e.g., 'total', 'nu-fission', etc.) by_nuclide : bool If true, computes cross sections for each nuclide in domain - domain : Material or Cell or Universe or Mesh + domain : openmc.Material or openmc.Cell or openmc.Universe or openmc.Mesh Domain for spatial homogenization domain_type : {'material', 'cell', 'distribcell', 'universe', 'mesh'} Domain type for spatial homogenization @@ -1462,7 +1462,7 @@ class DelayedNuFissionXS(MDGXS): Reaction type (e.g., 'total', 'nu-fission', etc.) by_nuclide : bool If true, computes cross sections for each nuclide in domain - domain : Material or Cell or Universe or Mesh + domain : openmc.Material or openmc.Cell or openmc.Universe or openmc.Mesh Domain for spatial homogenization domain_type : {'material', 'cell', 'distribcell', 'universe', 'mesh'} Domain type for spatial homogenization @@ -1598,7 +1598,7 @@ class Beta(MDGXS): Reaction type (e.g., 'total', 'nu-fission', etc.) by_nuclide : bool If true, computes cross sections for each nuclide in domain - domain : Material or Cell or Universe or Mesh + domain : openmc.Material or openmc.Cell or openmc.Universe or openmc.Mesh Domain for spatial homogenization domain_type : {'material', 'cell', 'distribcell', 'universe', 'mesh'} Domain type for spatial homogenization @@ -1782,7 +1782,7 @@ class DecayRate(MDGXS): Reaction type (e.g., 'total', 'nu-fission', etc.) by_nuclide : bool If true, computes cross sections for each nuclide in domain - domain : Material or Cell or Universe or Mesh + domain : openmc.Material or openmc.Cell or openmc.Universe or openmc.Mesh Domain for spatial homogenization domain_type : {'material', 'cell', 'distribcell', 'universe', 'mesh'} Domain type for spatial homogenization @@ -1959,7 +1959,7 @@ class MatrixMDGXS(MDGXS): Reaction type (e.g., 'total', 'nu-fission', etc.) by_nuclide : bool If true, computes cross sections for each nuclide in domain - domain : Material or Cell or Universe or Mesh + domain : openmc.Material or openmc.Cell or openmc.Universe or openmc.Mesh Domain for spatial homogenization domain_type : {'material', 'cell', 'distribcell', 'universe', 'mesh'} Domain type for spatial homogenization @@ -2551,7 +2551,7 @@ class DelayedNuFissionMatrixXS(MatrixMDGXS): Reaction type (e.g., 'total', 'nu-fission', etc.) by_nuclide : bool If true, computes cross sections for each nuclide in domain - domain : Material or Cell or Universe or Mesh + domain : openmc.Material or openmc.Cell or openmc.Universe or openmc.Mesh Domain for spatial homogenization domain_type : {'material', 'cell', 'distribcell', 'universe', 'mesh'} Domain type for spatial homogenization diff --git a/openmc/mgxs/mgxs.py b/openmc/mgxs/mgxs.py index 3325e0b15..192f484a5 100644 --- a/openmc/mgxs/mgxs.py +++ b/openmc/mgxs/mgxs.py @@ -154,7 +154,7 @@ class MGXS(object): Reaction type (e.g., 'total', 'nu-fission', etc.) by_nuclide : bool If true, computes cross sections for each nuclide in domain - domain : Material or Cell or Universe or Mesh + domain : openmc.Material or openmc.Cell or openmc.Universe or openmc.Mesh Domain for spatial homogenization domain_type : {'material', 'cell', 'distribcell', 'universe', 'mesh'} Domain type for spatial homogenization @@ -2018,7 +2018,7 @@ class MatrixMGXS(MGXS): Reaction type (e.g., 'total', 'nu-fission', etc.) by_nuclide : bool If true, computes cross sections for each nuclide in domain - domain : Material or Cell or Universe or Mesh + domain : openmc.Material or openmc.Cell or openmc.Universe or openmc.Mesh Domain for spatial homogenization domain_type : {'material', 'cell', 'distribcell', 'universe', 'mesh'} Domain type for spatial homogenization @@ -2512,7 +2512,7 @@ class TotalXS(MGXS): Reaction type (e.g., 'total', 'nu-fission', etc.) by_nuclide : bool If true, computes cross sections for each nuclide in domain - domain : Material or Cell or Universe or Mesh + domain : openmc.Material or openmc.Cell or openmc.Universe or openmc.Mesh Domain for spatial homogenization domain_type : {'material', 'cell', 'distribcell', 'universe', 'mesh'} Domain type for spatial homogenization @@ -2649,7 +2649,7 @@ class TransportXS(MGXS): If True, the cross section data will include neutron multiplication by_nuclide : bool If true, computes cross sections for each nuclide in domain - domain : Material or Cell or Universe or Mesh + domain : openmc.Material or openmc.Cell or openmc.Universe or openmc.Mesh Domain for spatial homogenization domain_type : {'material', 'cell', 'distribcell', 'universe', 'mesh'} Domain type for spatial homogenization @@ -2858,7 +2858,7 @@ class AbsorptionXS(MGXS): Reaction type (e.g., 'total', 'nu-fission', etc.) by_nuclide : bool If true, computes cross sections for each nuclide in domain - domain : Material or Cell or Universe or Mesh + domain : openmc.Material or openmc.Cell or openmc.Universe or openmc.Mesh Domain for spatial homogenization domain_type : {'material', 'cell', 'distribcell', 'universe', 'mesh'} Domain type for spatial homogenization @@ -2986,7 +2986,7 @@ class CaptureXS(MGXS): Reaction type (e.g., 'total', 'nu-fission', etc.) by_nuclide : bool If true, computes cross sections for each nuclide in domain - domain : Material or Cell or Universe or Mesh + domain : openmc.Material or openmc.Cell or openmc.Universe or openmc.Mesh Domain for spatial homogenization domain_type : {'material', 'cell', 'distribcell', 'universe', 'mesh'} Domain type for spatial homogenization @@ -3140,7 +3140,7 @@ class FissionXS(MGXS): If true, computes cross sections which only includes prompt neutrons by_nuclide : bool If true, computes cross sections for each nuclide in domain - domain : Material or Cell or Universe or Mesh + domain : openmc.Material or openmc.Cell or openmc.Universe or openmc.Mesh Domain for spatial homogenization domain_type : {'material', 'cell', 'distribcell', 'universe', 'mesh'} Domain type for spatial homogenization @@ -3309,7 +3309,7 @@ class KappaFissionXS(MGXS): Reaction type (e.g., 'total', 'nu-fission', etc.) by_nuclide : bool If true, computes cross sections for each nuclide in domain - domain : Material or Cell or Universe or Mesh + domain : openmc.Material or openmc.Cell or openmc.Universe or openmc.Mesh Domain for spatial homogenization domain_type : {'material', 'cell', 'distribcell', 'universe', 'mesh'} Domain type for spatial homogenization @@ -3441,7 +3441,7 @@ class ScatterXS(MGXS): If True, the cross section data will include neutron multiplication by_nuclide : bool If true, computes cross sections for each nuclide in domain - domain : Material or Cell or Universe or Mesh + domain : openmc.Material or openmc.Cell or openmc.Universe or openmc.Mesh Domain for spatial homogenization domain_type : {'material', 'cell', 'distribcell', 'universe', 'mesh'} Domain type for spatial homogenization @@ -3658,7 +3658,7 @@ class ScatterMatrixXS(MatrixMGXS): If True, the cross section data will include neutron multiplication by_nuclide : bool If true, computes cross sections for each nuclide in domain - domain : Material or Cell or Universe or Mesh + domain : openmc.Material or openmc.Cell or openmc.Universe or openmc.Mesh Domain for spatial homogenization domain_type : {'material', 'cell', 'distribcell', 'universe', 'mesh'} Domain type for spatial homogenization @@ -4020,7 +4020,7 @@ class ScatterMatrixXS(MatrixMGXS): # Override the nuclides for tally arithmetic correction.nuclides = scatter_p1.nuclides self._xs_tally -= correction - + self._compute_xs() return self._xs_tally @@ -4749,7 +4749,7 @@ class MultiplicityMatrixXS(MatrixMGXS): Reaction type (e.g., 'total', 'nu-fission', etc.) by_nuclide : bool If true, computes cross sections for each nuclide in domain - domain : Material or Cell or Universe or Mesh + domain : openmc.Material or openmc.Cell or openmc.Universe or openmc.Mesh Domain for spatial homogenization domain_type : {'material', 'cell', 'distribcell', 'universe', 'mesh'} Domain type for spatial homogenization @@ -4915,7 +4915,7 @@ class ScatterProbabilityMatrix(MatrixMGXS): Reaction type (e.g., 'total', 'nu-fission', etc.) by_nuclide : bool If true, computes cross sections for each nuclide in domain - domain : Material or Cell or Universe or Mesh + domain : openmc.Material or openmc.Cell or openmc.Universe or openmc.Mesh Domain for spatial homogenization domain_type : {'material', 'cell', 'distribcell', 'universe', 'mesh'} Domain type for spatial homogenization @@ -5088,7 +5088,7 @@ class NuFissionMatrixXS(MatrixMGXS): If true, computes cross sections which only includes prompt neutrons by_nuclide : bool If true, computes cross sections for each nuclide in domain - domain : Material or Cell or Universe or Mesh + domain : openmc.Material or openmc.Cell or openmc.Universe or openmc.Mesh Domain for spatial homogenization domain_type : {'material', 'cell', 'distribcell', 'universe', 'mesh'} Domain type for spatial homogenization @@ -5245,7 +5245,7 @@ class Chi(MGXS): If true, computes cross sections which only includes prompt neutrons by_nuclide : bool If true, computes cross sections for each nuclide in domain - domain : Material or Cell or Universe or Mesh + domain : openmc.Material or openmc.Cell or openmc.Universe or openmc.Mesh Domain for spatial homogenization domain_type : {'material', 'cell', 'distribcell', 'universe', 'mesh'} Domain type for spatial homogenization @@ -5823,7 +5823,7 @@ class InverseVelocity(MGXS): Reaction type (e.g., 'total', 'nu-fission', etc.) by_nuclide : bool If true, computes cross sections for each nuclide in domain - domain : Material or Cell or Universe or Mesh + domain : openmc.Material or openmc.Cell or openmc.Universe or openmc.Mesh Domain for spatial homogenization domain_type : {'material', 'cell', 'distribcell', 'universe', 'mesh'} Domain type for spatial homogenization diff --git a/openmc/tallies.py b/openmc/tallies.py index 6853db494..9c656d56d 100644 --- a/openmc/tallies.py +++ b/openmc/tallies.py @@ -502,7 +502,7 @@ class Tally(IDManagerMixin): Parameters ---------- - nuclide : str, Nuclide, CrossNuclide or AggregateNuclide + nuclide : str, openmc.Nuclide, CrossNuclide or AggregateNuclide Nuclide to add to the tally. The nuclide should be a Nuclide object when a user is adding nuclides to a Tally for input file generation. The nuclide is a str when a Tally is created from a StatePoint file