Allow default material cell offset generation to be turned off

docs/source/io_formats/settings.rst

@@ -195,6 +195,18 @@ based on the recommended value in LA-UR-14-24530_.
 
 .. _LA-UR-14-24530: https://laws.lanl.gov/vhosts/mcnp.lanl.gov/pdf_files/la-ur-14-24530.pdf
 
+---------------------------
+``<material_cell_offsets>``
+---------------------------
+
+By default, OpenMC will count the number of instances of each cell filled with a
+material and generate "offset tables" that are used for cell instance tallies.
+The ``<material_cell_offsets>`` element allows a user to override this default
+setting and turn off the generation of offset tables, if desired, by setting it
+to false.
+
+  *Default*: true
+
 ---------------------------
 ``<max_order>`` Element
 ---------------------------
@@ -423,11 +435,11 @@ attributes/sub-elements:
 
     :type:
       The type of spatial distribution. Valid options are "box", "fission",
-      "point", "cartesian", and "spherical". A "box" spatial distribution has 
+      "point", "cartesian", and "spherical". A "box" spatial distribution has
       coordinates sampled uniformly in a parallelepiped. A "fission" spatial
-      distribution samples locations from a "box" distribution but only 
-      locations in fissionable materials are accepted. A "point" spatial 
-      distribution has coordinates specified by a triplet. An "cartesian" 
+      distribution samples locations from a "box" distribution but only
+      locations in fissionable materials are accepted. A "point" spatial
+      distribution has coordinates specified by a triplet. An "cartesian"
       spatial distribution specifies independent distributions of x-, y-, and
       z-coordinates. A "spherical" spatial distribution specifies independent
       distributions of r-, theta-, and phi-coordinates where theta is the angle

include/openmc/settings.h

@@ -31,6 +31,7 @@ extern "C" bool cmfd_run;             //!< is a CMFD run?
 extern "C" bool dagmc;                //!< indicator of DAGMC geometry
 extern "C" bool entropy_on;           //!< calculate Shannon entropy?
 extern bool legendre_to_tabular;      //!< convert Legendre distributions to tabular?
+extern bool material_cell_offsets;    //!< create material cells offsets?
 extern "C" bool output_summary;       //!< write summary.h5?
 extern bool output_tallies;           //!< write tallies.out?
 extern bool particle_restart_run;     //!< particle restart run?

openmc/settings.py

@@ -61,6 +61,9 @@ class Settings(object):
         relative error used.
     log_grid_bins : int
         Number of bins for logarithmic energy grid search
+    material_cell_offsets : bool
+        Generate an "offset table" for material cells by default. These tables
+        are necessary when a particular instance of a cell needs to be tallied.
     max_order : None or int
         Maximum scattering order to apply globally when in multi-group mode.
     no_reduce : bool
@@ -216,6 +219,7 @@ class Settings(object):
             VolumeCalculation, 'volume calculations')
 
         self._create_fission_neutrons = None
+        self._material_cell_offsets = None
         self._log_grid_bins = None
 
         self._dagmc = False
@@ -352,6 +356,10 @@ class Settings(object):
     def create_fission_neutrons(self):
         return self._create_fission_neutrons
 
+    @property
+    def material_cell_offsets(self):
+        return self._material_cell_offsets
+
     @property
     def log_grid_bins(self):
         return self._log_grid_bins
@@ -683,6 +691,11 @@ class Settings(object):
                       create_fission_neutrons, bool)
         self._create_fission_neutrons = create_fission_neutrons
 
+    @material_cell_offsets.setter
+    def material_cell_offsets(self, value):
+        cv.check_type('material cell offsets', value, bool)
+        self._material_cell_offsets = value
+
     @log_grid_bins.setter
     def log_grid_bins(self, log_grid_bins):
         cv.check_type('log grid bins', log_grid_bins, Real)
@@ -917,6 +930,11 @@ class Settings(object):
             elem = ET.SubElement(root, "create_fission_neutrons")
             elem.text = str(self._create_fission_neutrons).lower()
 
+    def _create_material_cell_offsets_subelement(self, root):
+        if self._material_cell_offsets is not None:
+            elem = ET.SubElement(root, "material_cell_offsets")
+            elem.text = str(self._material_cell_offsets).lower()
+
     def _create_log_grid_bins_subelement(self, root):
         if self._log_grid_bins is not None:
             elem = ET.SubElement(root, "log_grid_bins")
@@ -1148,6 +1166,11 @@ class Settings(object):
         if text is not None:
             self.create_fission_neutrons = text in ('true', '1')
 
+    def _material_cell_offsets_from_xml_element(self, root):
+        text = get_text(root, 'material_cell_offsets')
+        if text is not None:
+            self.material_cell_offsets = text in ('true', '1')
+
     def _log_grid_bins_from_xml_element(self, root):
         text = get_text(root, 'log_grid_bins')
         if text is not None:
@@ -1202,6 +1225,7 @@ class Settings(object):
         self._create_resonance_scattering_subelement(root_element)
         self._create_volume_calcs_subelement(root_element)
         self._create_create_fission_neutrons_subelement(root_element)
+        self._create_material_cell_offsets_subelement(root_element)
         self._create_log_grid_bins_subelement(root_element)
         self._create_dagmc_subelement(root_element)
 
@@ -1267,6 +1291,7 @@ class Settings(object):
         settings._ufs_mesh_from_xml_element(root)
         settings._resonance_scattering_from_xml_element(root)
         settings._create_fission_neutrons_from_xml_element(root)
+        settings._material_cell_offsets_from_xml_element(root)
         settings._log_grid_bins_from_xml_element(root)
         settings._dagmc_from_xml_element(root)
 

src/finalize.cpp

@@ -73,6 +73,7 @@ int openmc_finalize()
   settings::gen_per_batch = 1;
   settings::legendre_to_tabular = true;
   settings::legendre_to_tabular_points = -1;
+  settings::material_cell_offsets = true;
   settings::n_particles = -1;
   settings::output_summary = true;
   settings::output_tallies = true;

src/geometry_aux.cpp

@@ -332,8 +332,10 @@ prepare_distribcell()
   }
 
   // By default, add material cells to the list of distributed cells
-  for (gsl::index i = 0; i < model::cells.size(); ++i) {
-    if (model::cells[i]->type_ == FILL_MATERIAL) distribcells.insert(i);
+  if (settings::material_cell_offsets) {
+    for (gsl::index i = 0; i < model::cells.size(); ++i) {
+      if (model::cells[i]->type_ == FILL_MATERIAL) distribcells.insert(i);
+    }
   }
 
   // Make sure that the number of materials/temperatures matches the number of

src/relaxng/settings.rnc

@@ -33,6 +33,8 @@ element settings {
 
   element log_grid_bins { xsd:positiveInteger }? &
 
+  element material_cell_offsets { xsd:boolean }? &
+
   element max_order { xsd:nonNegativeInteger }? &
 
   element mesh {

src/relaxng/settings.rng

@@ -156,6 +156,11 @@
         <data type="positiveInteger"/>
       </element>
     </optional>
+    <optional>
+      <element name="material_cell_offsets">
+        <data type="boolean"/>
+      </element>
+    </optional>
     <optional>
       <element name="max_order">
         <data type="nonNegativeInteger"/>

src/settings.cpp

@@ -46,6 +46,7 @@ bool create_fission_neutrons {true};
 bool dagmc                   {false};
 bool entropy_on              {false};
 bool legendre_to_tabular     {true};
+bool material_cell_offsets   {true};
 bool output_summary          {true};
 bool output_tallies          {true};
 bool particle_restart_run    {false};
@@ -776,6 +777,11 @@ void read_settings_xml()
       create_fission_neutrons = get_node_value_bool(root, "create_fission_neutrons");
     }
   }
+
+  // Check whether material cell offsets should be generated
+  if (check_for_node(root, "material_cell_offsets")) {
+    material_cell_offsets = get_node_value_bool(root, "material_cell_offsets");
+  }
 }
 
 void free_memory_settings() {