Filter material from (#2750)

Co-authored-by: Patrick Shriwise <pshriwise@gmail.com>
Co-authored-by: Paul Romano <paul.k.romano@gmail.com>

CMakeLists.txt

@@ -406,6 +406,7 @@ list(APPEND libopenmc_SOURCES
   src/tallies/filter_energyfunc.cpp
   src/tallies/filter_legendre.cpp
   src/tallies/filter_material.cpp
+  src/tallies/filter_materialfrom.cpp
   src/tallies/filter_mesh.cpp
   src/tallies/filter_meshsurface.cpp
   src/tallies/filter_mu.cpp

docs/source/pythonapi/base.rst

@@ -118,6 +118,7 @@ Constructing Tallies
    openmc.Filter
    openmc.UniverseFilter
    openmc.MaterialFilter
+   openmc.MaterialFromFilter
    openmc.CellFilter
    openmc.CellFromFilter
    openmc.CellBornFilter

include/openmc/particle_data.h

@@ -435,6 +435,7 @@ class ParticleData {
   int& material() { return material_; }
   const int& material() const { return material_; }
   int& material_last() { return material_last_; }
+  const int& material_last() const { return material_last_; }
 
   BoundaryInfo& boundary() { return boundary_; }
 

include/openmc/tallies/filter.h

@@ -31,6 +31,7 @@ enum class FilterType {
   ENERGY_OUT,
   LEGENDRE,
   MATERIAL,
+  MATERIALFROM,
   MESH,
   MESH_SURFACE,
   MU,

include/openmc/tallies/filter_material.h

@@ -46,7 +46,7 @@ class MaterialFilter : public Filter {
 
   void set_materials(gsl::span<const int32_t> materials);
 
-private:
+protected:
   //----------------------------------------------------------------------------
   // Data members
 

include/openmc/tallies/filter_materialfrom.h

@@ -0,0 +1,29 @@
+#ifndef OPENMC_TALLIES_FILTER_MATERIALFROM_H
+#define OPENMC_TALLIES_FILTER_MATERIALFROM_H
+
+#include <string>
+
+#include "openmc/tallies/filter_material.h"
+
+namespace openmc {
+
+//==============================================================================
+//! Specifies which material particles exit when crossing a surface.
+//==============================================================================
+
+class MaterialFromFilter : public MaterialFilter {
+public:
+  //----------------------------------------------------------------------------
+  // Methods
+
+  std::string type_str() const override { return "materialfrom"; }
+  FilterType type() const override { return FilterType::MATERIALFROM; }
+
+  void get_all_bins(const Particle& p, TallyEstimator estimator,
+    FilterMatch& match) const override;
+
+  std::string text_label(int bin) const override;
+};
+
+} // namespace openmc
+#endif // OPENMC_TALLIES_FILTER_MATERIALFROM_H

openmc/filter.py

@@ -22,7 +22,7 @@
 _FILTER_TYPES = (
     'universe', 'material', 'cell', 'cellborn', 'surface', 'mesh', 'energy',
     'energyout', 'mu', 'polar', 'azimuthal', 'distribcell', 'delayedgroup',
-    'energyfunction', 'cellfrom', 'legendre', 'spatiallegendre',
+    'energyfunction', 'cellfrom', 'materialfrom', 'legendre', 'spatiallegendre',
     'sphericalharmonics', 'zernike', 'zernikeradial', 'particle', 'cellinstance',
     'collision', 'time'
 )
@@ -451,7 +451,7 @@ class UniverseFilter(WithIDFilter):
     Parameters
     ----------
     bins : openmc.UniverseBase, int, or iterable thereof
-        The Universes to tally. Either openmc.UniverseBase objects or their
+        The Universes to tally. Either :class:`openmc.UniverseBase` objects or their
         Integral ID numbers can be used.
     filter_id : int
         Unique identifier for the filter
@@ -475,7 +475,31 @@ class MaterialFilter(WithIDFilter):
     Parameters
     ----------
     bins : openmc.Material, Integral, or iterable thereof
-        The Materials to tally. Either openmc.Material objects or their
+        The material(s) to tally. Either :class:`openmc.Material` objects or their
+        Integral ID numbers can be used.
+    filter_id : int
+        Unique identifier for the filter
+
+    Attributes
+    ----------
+    bins : Iterable of Integral
+        openmc.Material IDs.
+    id : int
+        Unique identifier for the filter
+    num_bins : Integral
+        The number of filter bins
+
+    """
+    expected_type = Material
+
+
+class MaterialFromFilter(WithIDFilter):
+    """Bins tally event locations based on the Material they occurred in.
+
+    Parameters
+    ----------
+    bins : openmc.Material, Integral, or iterable thereof
+        The material(s) to tally. Either :class:`openmc.Material` objects or their
         Integral ID numbers can be used.
     filter_id : int
         Unique identifier for the filter
@@ -499,7 +523,7 @@ class CellFilter(WithIDFilter):
     Parameters
     ----------
     bins : openmc.Cell, int, or iterable thereof
-        The cells to tally. Either openmc.Cell objects or their ID numbers can
+        The cells to tally. Either :class:`openmc.Cell` objects or their ID numbers can
         be used.
     filter_id : int
         Unique identifier for the filter

openmc/lib/filter.py

@@ -20,7 +20,7 @@
     'Filter', 'AzimuthalFilter', 'CellFilter', 'CellbornFilter', 'CellfromFilter',
     'CellInstanceFilter', 'CollisionFilter', 'DistribcellFilter', 'DelayedGroupFilter',
     'EnergyFilter', 'EnergyoutFilter', 'EnergyFunctionFilter', 'LegendreFilter',
-    'MaterialFilter', 'MeshFilter', 'MeshSurfaceFilter', 'MuFilter', 'ParticleFilter',
+    'MaterialFilter', 'MaterialFromFilter', 'MeshFilter', 'MeshSurfaceFilter', 'MuFilter', 'ParticleFilter',
     'PolarFilter', 'SphericalHarmonicsFilter', 'SpatialLegendreFilter', 'SurfaceFilter',
     'UniverseFilter', 'ZernikeFilter', 'ZernikeRadialFilter', 'filters'
 ]
@@ -342,6 +342,10 @@ def bins(self, materials):
         _dll.openmc_material_filter_set_bins(self._index, n, bins)
 
 
+class MaterialFromFilter(Filter):
+    filter_type = 'materialfrom'
+
+
 class MeshFilter(Filter):
     filter_type = 'mesh'
 
@@ -501,6 +505,7 @@ class ZernikeRadialFilter(ZernikeFilter):
     'energyfunction': EnergyFunctionFilter,
     'legendre': LegendreFilter,
     'material': MaterialFilter,
+    'materialfrom': MaterialFromFilter,
     'mesh': MeshFilter,
     'meshsurface': MeshSurfaceFilter,
     'mu': MuFilter,

src/tallies/filter.cpp

@@ -21,6 +21,7 @@
 #include "openmc/tallies/filter_energyfunc.h"
 #include "openmc/tallies/filter_legendre.h"
 #include "openmc/tallies/filter_material.h"
+#include "openmc/tallies/filter_materialfrom.h"
 #include "openmc/tallies/filter_mesh.h"
 #include "openmc/tallies/filter_meshsurface.h"
 #include "openmc/tallies/filter_mu.h"
@@ -121,6 +122,8 @@ Filter* Filter::create(const std::string& type, int32_t id)
     return Filter::create<LegendreFilter>(id);
   } else if (type == "material") {
     return Filter::create<MaterialFilter>(id);
+  } else if (type == "materialfrom") {
+    return Filter::create<MaterialFromFilter>(id);
   } else if (type == "mesh") {
     return Filter::create<MeshFilter>(id);
   } else if (type == "meshsurface") {

src/tallies/filter_materialfrom.cpp

@@ -0,0 +1,24 @@
+#include "openmc/tallies/filter_materialfrom.h"
+
+#include "openmc/cell.h"
+#include "openmc/material.h"
+
+namespace openmc {
+
+void MaterialFromFilter::get_all_bins(
+  const Particle& p, TallyEstimator estimator, FilterMatch& match) const
+{
+  auto search = map_.find(p.material_last());
+  if (search != map_.end()) {
+    match.bins_.push_back(search->second);
+    match.weights_.push_back(1.0);
+  }
+}
+
+std::string MaterialFromFilter::text_label(int bin) const
+{
+  return "Material from " +
+         std::to_string(model::materials[materials_[bin]]->id_);
+}
+
+} // namespace openmc

tests/regression_tests/surface_tally/inputs_true.dat

@@ -54,7 +54,7 @@
     <filter id="4" type="surface">
       <bins>1</bins>
     </filter>
-    <filter id="7" type="cellfrom">
+    <filter id="7" type="materialfrom">
       <bins>2</bins>
     </filter>
     <filter id="8" type="cell">

tests/regression_tests/surface_tally/test.py

@@ -106,19 +106,19 @@ def __init__(self, *args, **kwargs):
 
         # Create partial current tallies from water to fuel
         # Filters
-        cell_from_filter = openmc.CellFromFilter(water)
+        mat_from_filter = openmc.MaterialFromFilter(borated_water)
         cell_filter = openmc.CellFilter(fuel)
 
         # Cell to cell filters for partial current
         cell_to_cell_tally = openmc.Tally(name=str('water_to_fuel_1'))
-        cell_to_cell_tally.filters = [cell_from_filter, cell_filter, \
+        cell_to_cell_tally.filters = [mat_from_filter, cell_filter, \
              energy_filter, polar_filter, azimuthal_filter]
         cell_to_cell_tally.scores = ['current']
         tallies_file.append(cell_to_cell_tally)
 
         # Cell from + surface filters for partial current
         cell_to_cell_tally = openmc.Tally(name=str('water_to_fuel_2'))
-        cell_to_cell_tally.filters = [cell_from_filter, surface_filter, \
+        cell_to_cell_tally.filters = [mat_from_filter, surface_filter, \
              energy_filter, polar_filter, azimuthal_filter]
         cell_to_cell_tally.scores = ['current']
         tallies_file.append(cell_to_cell_tally)