Ignoring nuclides with no xs data when writing material xml files

openmc/deplete/coupled_operator.py

@@ -83,7 +83,6 @@ def _get_nuclides_with_data(cross_sections):
 
     return nuclides
 
-
 class CoupledOperator(OpenMCOperator):
     """Transport-coupled transport operator.
 
@@ -300,22 +299,6 @@ def _load_previous_results(self):
                 new_res = res_obj.distribute(self.local_mats, mat_indexes)
                 self.prev_res.append(new_res)
 
-    def _get_nuclides_with_data(self, cross_sections):
-        """Loads cross_sections.xml file to find nuclides with neutron data
-
-        Parameters
-        ----------
-        cross_sections : str
-            Path to cross_sections.xml file
-
-        Returns
-        -------
-        nuclides : set of str
-            Set of nuclide names that have cross secton data
-
-        """
-        return _get_nuclides_with_data(cross_sections)
-
     def _get_helper_classes(self, helper_kwargs):
         """Create the ``_rate_helper``, ``_normalization_helper``, and
         ``_yield_helper`` objects.
@@ -404,7 +387,7 @@ def _generate_materials_xml(self):
         for mat in self.materials:
             mat._nuclides.sort(key=lambda x: nuclides.index(x[0]))
 
-        self.materials.export_to_xml(nuclides_to_ignore=self._decay_nucs)
+        self.materials.export_to_xml(ignore_phantom_nuclides=True)
 
     def __call__(self, vec, source_rate):
         """Runs a simulation.

openmc/deplete/independent_operator.py

@@ -283,10 +283,6 @@ def _load_previous_results(self):
                 self.prev_res.append(new_res)
 
 
-    def _get_nuclides_with_data(self, cross_sections: List[MicroXS]) -> Set[str]:
-        """Finds nuclides with cross section data"""
-        return set(cross_sections[0].nuclides)
-
     class _IndependentRateHelper(ReactionRateHelper):
         """Class for generating one-group reaction rates with flux and
         one-group cross sections.

openmc/deplete/microxs.py

@@ -13,12 +13,12 @@
 
 from openmc.checkvalue import check_type, check_value, check_iterable_type, PathLike
 from openmc.exceptions import DataError
-from openmc import StatePoint
 from openmc.mgxs import GROUP_STRUCTURES
 from openmc.data import REACTION_MT
 import openmc
 from .abc import change_directory
 from .chain import Chain, REACTIONS
+from .coupled_operator import _find_cross_sections
 from .coupled_operator import _find_cross_sections, _get_nuclides_with_data
 import openmc.lib
 
@@ -91,7 +91,7 @@ def get_microxs_and_flux(
         reactions = chain.reactions
     if not nuclides:
         cross_sections = _find_cross_sections(model)
-        nuclides_with_data = _get_nuclides_with_data(cross_sections)
+        nuclides_with_data = get_nuclides_with_data(cross_sections)
         nuclides = [nuc.name for nuc in chain.nuclides
                     if nuc.name in nuclides_with_data]
 

openmc/deplete/openmc_operator.py

@@ -15,6 +15,7 @@
 from openmc.checkvalue import check_value
 from openmc.exceptions import DataError
 from openmc.mpi import comm
+from openmc import get_nuclides_with_data
 from .abc import TransportOperator, OperatorResult
 from .atom_number import AtomNumber
 from .reaction_rates import ReactionRates
@@ -148,16 +149,12 @@ def __init__(
         # for which there is an entry in the micro_xs parameter
         openmc.reset_auto_ids()
 
-        self.nuclides_with_data = self._get_nuclides_with_data(
-            self.cross_sections)
+        self.nuclides_with_data = get_nuclides_with_data(self.cross_sections)
 
         # Select nuclides with data that are also in the chain
         self._burnable_nucs = [nuc.name for nuc in self.chain.nuclides
                                if nuc.name in self.nuclides_with_data]
 
-        # Select nuclides without data that are also in the chain
-        self._decay_nucs = [nuc.name for nuc in self.chain.nuclides
-                            if nuc.name not in self.nuclides_with_data]
 
         self.burnable_mats, volumes, all_nuclides = self._get_burnable_mats()
         self.local_mats = _distribute(self.burnable_mats)
@@ -207,10 +204,10 @@ def _get_burnable_mats(self) -> Tuple[List[str], Dict[str, float], List[str]]:
         # Iterate once through the geometry to get dictionaries
         for mat in self.materials:
             for nuclide in mat.get_nuclides():
-                if nuclide in self.nuclides_with_data or self._decay_nucs:
+                if nuclide in self.nuclides_with_data or self.chain.nuclides:
                     model_nuclides.add(nuclide)
                 else:
-                    msg = (f"Nuclilde {nuclide} in material {mat.id} is not "
+                    msg = (f"Nuclide {nuclide} in material {mat.id} is not "
                            "present in the depletion chain and has no cross "
                            "section data.")
                     raise warn(msg)
@@ -247,23 +244,6 @@ def _load_previous_results(self):
         """Load results from a previous depletion simulation"""
         pass
 
-    @abstractmethod
-    def _get_nuclides_with_data(self, cross_sections):
-        """Find nuclides with cross section data
-
-        Parameters
-        ----------
-        cross_sections : str or pandas.DataFrame
-            Path to continuous energy cross section library, or object
-            containing one-group cross-sections.
-
-        Returns
-        -------
-        nuclides : set of str
-            Set of nuclide names that have cross secton data
-
-        """
-
     def _extract_number(self, local_mats, volume, all_nuclides, prev_res=None):
         """Construct AtomNumber using geometry
 

openmc/material.py

@@ -16,6 +16,7 @@
 import openmc
 import openmc.data
 import openmc.checkvalue as cv
+from openmc.data import DataLibrary
 from ._xml import clean_indentation, reorder_attributes
 from .mixin import IDManagerMixin
 from openmc.checkvalue import PathLike
@@ -1314,7 +1315,7 @@ def _get_macroscopic_xml(self, macroscopic: str) -> ET.Element:
 
     def _get_nuclides_xml(
             self, nuclides: typing.Iterable[NuclideTuple],
-            nuclides_to_ignore: Optional[typing.Iterable[str]] = None)-> List[ET.Element]:
+            nuclides_to_ignore: Optional[typing.Iterable[str]] = None) -> List[ET.Element]:
         xml_elements = []
 
         # Remove any nuclides to ignore from the XML export
@@ -1340,6 +1341,16 @@ def to_xml_element(
             XML element containing material data
 
         """
+        if nuclides_to_ignore is not None:
+            nuclides = [nuc.name for nuc in self.nuclides]
+            phantoms = [nuc for nuc in nuclides_to_ignore if nuc in nuclides]
+            phantom_masses = sum([self.get_mass_density(nuc) for nuc in phantoms])
+            mass_loss_ratio = phantom_masses / self.get_mass_density()
+            msg = f"By ignoring {phantoms} in material {self.id} for transport calculation, " \
+                  f"{mass_loss_ratio:.2%} of the material total mass density in lost."
+            openmc.warnings.warn(msg)
+        else:
+            phantoms = None
 
         # Create Material XML element
         element = ET.Element("material")
@@ -1379,7 +1390,7 @@ def to_xml_element(
         if self._macroscopic is None:
             # Create nuclide XML subelements
             subelements = self._get_nuclides_xml(self._nuclides,
-                                                 nuclides_to_ignore=nuclides_to_ignore)
+                                                 nuclides_to_ignore=phantoms)
             for subelement in subelements:
                 element.append(subelement)
         else:
@@ -1636,7 +1647,7 @@ def make_isotropic_in_lab(self):
             material.make_isotropic_in_lab()
 
     def _write_xml(self, file, header=True, level=0, spaces_per_level=2,
-                   trailing_indent=True, nuclides_to_ignore=None):
+                   trailing_indent=True, nuclides_to_ignore=None, ignore_phantom_nuclides=False):
         """Writes XML content of the materials to an open file handle.
 
         Parameters
@@ -1653,8 +1664,23 @@ def _write_xml(self, file, header=True, level=0, spaces_per_level=2,
             Whether or not to write a trailing indentation for the materials element
         nuclides_to_ignore : list of str
             Nuclides to ignore when exporting to XML.
+        ignore_phantom_nuclides: bool
+            If True, nuclides present in the materials but with no available cross-sections
+            will be ignored when writing to xml.
 
         """
+        if ignore_phantom_nuclides:
+            xml_path = openmc.config.get("cross_sections", None)
+            if xml_path is None:
+                raise ValueError("A cross_sections.xml file must be set to identify phantom nuclides.")
+            available = get_nuclides_with_data(xml_path)
+            candidates = list(set([nuc.name for m in self for nuc in m.nuclides]))
+            phantoms = [nuc for nuc in candidates if nuc not in available]
+            if nuclides_to_ignore is None:
+                nuclides_to_ignore = phantoms
+            else:
+                nuclides_to_ignore += phantoms
+
         indentation = level*spaces_per_level*' '
         # Write the header and the opening tag for the root element.
         if header:
@@ -1689,7 +1715,8 @@ def _write_xml(self, file, header=True, level=0, spaces_per_level=2,
             file.write(indentation)
 
     def export_to_xml(self, path: PathLike = 'materials.xml',
-                      nuclides_to_ignore: Optional[typing.Iterable[str]] = None):
+                      nuclides_to_ignore: Optional[typing.Iterable[str]] = None,
+                      ignore_phantom_nuclides: bool = False):
         """Export material collection to an XML file.
 
         Parameters
@@ -1698,6 +1725,9 @@ def export_to_xml(self, path: PathLike = 'materials.xml',
             Path to file to write. Defaults to 'materials.xml'.
         nuclides_to_ignore : list of str
             Nuclides to ignore when exporting to XML.
+        ignore_phantom_nuclides: bool
+            If True, nuclides present in the materials but with no available cross-sections
+            will be ignored when writing to xml.
 
         """
         # Check if path is a directory
@@ -1710,7 +1740,9 @@ def export_to_xml(self, path: PathLike = 'materials.xml',
         # one go.
         with open(str(p), 'w', encoding='utf-8',
                   errors='xmlcharrefreplace') as fh:
-            self._write_xml(fh, nuclides_to_ignore=nuclides_to_ignore)
+            self._write_xml(fh,
+                            ignore_phantom_nuclides=ignore_phantom_nuclides,
+                            nuclides_to_ignore=nuclides_to_ignore)
 
     @classmethod
     def from_xml_element(cls, elem) -> Materials:
@@ -1758,3 +1790,32 @@ def from_xml(cls, path: PathLike = 'materials.xml') -> Materials:
         root = tree.getroot()
 
         return cls.from_xml_element(root)
+
+
+def get_nuclides_with_data(cross_sections):
+    """Loads cross_sections.xml file to find nuclides with neutron data or
+    finds nuclides with cross section data from iterable of MicroXs objects.
+
+    Parameters
+    ----------
+    cross_sections : patlib.Path or iterable of MicroXs
+        Path to cross_sections.xml file or iterable of MicroXs objects.
+
+    Returns
+    -------
+    nuclides : set of str
+        Set of nuclide names that have cross secton data
+
+    """
+    if isinstance(cross_sections, Path):
+        nuclides = set()
+        data_lib = DataLibrary.from_xml(cross_sections)
+        for library in data_lib.libraries:
+            if library['type'] != 'neutron':
+                continue
+            for name in library['materials']:
+                if name not in nuclides:
+                    nuclides.add(name)
+        return nuclides
+    else:
+        return set(cross_sections[0].nuclides)

openmc/model/model.py

@@ -431,7 +431,7 @@ def deplete(self, timesteps, method='cecm', final_step=True,
                 depletion_operator.cleanup_when_done = True
                 depletion_operator.finalize()
 
-    def export_to_xml(self, directory='.', remove_surfs=False):
+    def export_to_xml(self, directory='.', remove_surfs=False, ignore_phantom_nuclides=False):
         """Export model to separate XML files.
 
         Parameters
@@ -442,6 +442,9 @@ def export_to_xml(self, directory='.', remove_surfs=False):
         remove_surfs : bool
             Whether or not to remove redundant surfaces from the geometry when
             exporting.
+        ignore_phantom_nuclides: bool
+            If True, nuclides present in the materials but with no available cross-sections
+            will be ignored when writing to xml.
 
             .. versionadded:: 0.13.1
         """
@@ -457,18 +460,18 @@ def export_to_xml(self, directory='.', remove_surfs=False):
         # for all materials in the geometry and use that to automatically build
         # a collection.
         if self.materials:
-            self.materials.export_to_xml(d)
+            self.materials.export_to_xml(d, ignore_phantom_nuclides=ignore_phantom_nuclides)
         else:
             materials = openmc.Materials(self.geometry.get_all_materials()
                                          .values())
-            materials.export_to_xml(d)
+            materials.export_to_xml(d, ignore_phantom_nuclides=ignore_phantom_nuclides)
 
         if self.tallies:
             self.tallies.export_to_xml(d)
         if self.plots:
             self.plots.export_to_xml(d)
 
-    def export_to_model_xml(self, path='model.xml', remove_surfs=False):
+    def export_to_model_xml(self, path='model.xml', remove_surfs=False, ignore_phantom_nuclides=False):
         """Export model to a single XML file.
 
         .. versionadded:: 0.13.3
@@ -481,6 +484,9 @@ def export_to_model_xml(self, path='model.xml', remove_surfs=False):
         remove_surfs : bool
             Whether or not to remove redundant surfaces from the geometry when
             exporting.
+        ignore_phantom_nuclides: bool
+            If True, nuclides present in the materials but with no available cross-sections
+            will be ignored when writing to xml.
 
         """
         xml_path = Path(path)
@@ -523,7 +529,7 @@ def export_to_model_xml(self, path='model.xml', remove_surfs=False):
             fh.write("<model>\n")
             # Write the materials collection to the open XML file first.
             # This will write the XML header also
-            materials._write_xml(fh, False, level=1)
+            materials._write_xml(fh, False, level=1, ignore_phantom_nuclides=ignore_phantom_nuclides)
             # Write remaining elements as a tree
             fh.write(ET.tostring(geometry_element, encoding="unicode"))
             fh.write(ET.tostring(settings_element, encoding="unicode"))