Add cell_wise MGXS generation method

docs/source/usersguide/random_ray.rst

@@ -652,8 +652,8 @@ model to use these multigroup cross sections. An example is given below::
   )
 
 The most important parameter to set is the ``method`` parameter, which can be
-either "stochastic_slab", "material_wise", or "infinite_medium". An overview
-of these methods is given below:
+one of "material_wise", "cell_wise", "stochastic_slab", or
+"infinite_medium". An overview of these methods is given below:
 
 .. list-table:: Comparison of Automatic MGXS Generation Methods
    :header-rows: 1
@@ -673,6 +673,21 @@ of these methods is given below:
      - * Potentially slower as the full geometry must be run
        * If a material is only present far from the source and doesn't get tallied
          to in the CE simulation, the MGXS will be zero for that material.
+   * - ``cell_wise``
+     - * Highest Fidelity
+       * Like ``material_wise``, but clones the material in each cell so every
+         cell gets its own cross sections (each material-filled cell is assigned a
+         distinct macroscopic).
+     - * Resolves spatial variation between distinct cells that share a material,
+         which ``material_wise`` averages away (e.g. a shield wall modelled as
+         several cells of one material, each covering a different depth)
+       * Captures spatial self shielding between cells filled with the same material
+     - * Most expensive (one cross section set per cell) and a larger library
+       * Fidelity is limited by the cell definitions, not the source region mesh:
+         a single large cell uses one cross section set throughout, so a steep
+         gradient is only resolved if the geometry is split into several cells
+       * Same far-from-source limitation as ``material_wise``: a cell that is not
+         tallied to yields zero cross sections for that cell
    * - ``stochastic_slab``
      - * Medium Fidelity
        * Runs a CE simulation with a greatly simplified geometry, where materials
@@ -691,6 +706,13 @@ of these methods is given below:
          between materials)
        * May hang if a material has a k-infinity greater than 1.0
 
+.. note::
+    The ``cell_wise`` method generates one cross section set per cell definition,
+    not per cell instance. A cell that appears in several locations of a lattice
+    therefore shares a single cross section set across all of those locations. If
+    you need distinct cross sections per instance, subdivide the geometry into
+    separate cells.
+
 When selecting a non-default energy group structure, you can manually define
 group boundaries or specify the name of a known group structure (a list of which
 can be found at :data:`openmc.mgxs.GROUP_STRUCTURES`). The ``nparticles``

openmc/model/model.py

@@ -2721,8 +2721,12 @@ def convert_to_multigroup(
 
         Parameters
         ----------
-        method : {"material_wise", "stochastic_slab", "infinite_medium"}, optional
-            Method to generate the MGXS.
+        method : {"material_wise", "stochastic_slab", "infinite_medium", \
+                  "cell_wise"}, optional
+            Method to generate the MGXS. "cell_wise" is like
+            "material_wise" but gives each cell its own cross sections. The
+            material in each material-filled cell is cloned, so the per-material
+            generation produces one cross section set per cell.
         groups : openmc.mgxs.EnergyGroups, str, or sequence of float, optional
             Energy group structure for the MGXS. Can be an
             :class:`openmc.mgxs.EnergyGroups` object, a string name of a
@@ -2790,6 +2794,18 @@ def convert_to_multigroup(
                     self.settings.run_mode = original_run_mode
                     break
 
+            # For "cell_wise", give each cell its own cross sections by
+            # cloning the material in every material-filled cell. Each clone gets
+            # a unique id, so the per-material generation below produces (and
+            # assigns) one cross section set per cell.
+            if method == "cell_wise":
+                cell_materials = []
+                for cell in self.geometry.get_all_cells().values():
+                    if isinstance(cell.fill, openmc.Material):
+                        cell.fill = cell.fill.clone()
+                        cell_materials.append(cell.fill)
+                self.materials = openmc.Materials(cell_materials)
+
             # Temporarily replace each material's name with a unique, valid HDF5
             # dataset name (its name plus ID) for use as its MGXS library entry
             # and macroscopic. The ID keeps the name unique even when materials
@@ -2806,7 +2822,7 @@ def convert_to_multigroup(
                     self._generate_infinite_medium_mgxs(
                         groups, nparticles, mgxs_path, correction, tmpdir, source_energy,
                         temperatures, temperature_settings)
-                elif method == "material_wise":
+                elif method in ("material_wise", "cell_wise"):
                     self._generate_material_wise_mgxs(
                         groups, nparticles, mgxs_path, correction, tmpdir,
                         temperatures, temperature_settings)

tests/unit_tests/dagmc/test_convert_to_multigroup.py

@@ -51,3 +51,43 @@ def test_convert_to_multigroup_without_particles_batches(run_in_tmpdir):
 
     # Verify the model was converted successfully
     assert model.settings.energy_mode == 'multi-group'
+
+
+def test_convert_to_multigroup_cell_wise(run_in_tmpdir):
+    """cell_wise gives each DAGMC volume its own cross sections, so two
+    cells filled with the same material end up with distinct macroscopics."""
+    openmc.reset_auto_ids()
+
+    # dagmc.h5m has two fuel volumes (both "no-void fuel"), one water volume, a
+    # graveyard and an implicit complement.
+    u235 = openmc.Material(name="no-void fuel")
+    u235.add_nuclide("U235", 1.0)
+    u235.set_density("g/cm3", 11.0)
+    water = openmc.Material(name="water")
+    water.add_nuclide("H1", 2.0)
+    water.add_nuclide("O16", 1.0)
+    water.set_density("g/cm3", 1.0)
+    water.id = 41
+
+    dagmc_file = Path(__file__).parent / "dagmc.h5m"
+    model = openmc.Model()
+    model.materials = openmc.Materials([u235, water])
+    model.geometry = openmc.Geometry(openmc.DAGMCUniverse(dagmc_file))
+    model.settings = openmc.Settings()
+    model.settings.run_mode = "fixed source"
+    source = openmc.IndependentSource()
+    source.energy = openmc.stats.delta_function(2.0e6)
+    model.settings.source = source
+
+    # Pre-create the library so MGXS generation/transport is skipped; this
+    # exercises the per-cell material cloning for the DAGMC cells only.
+    Path("mgxs.h5").touch()
+    model.convert_to_multigroup(
+        method="cell_wise", groups="CASMO-2", mgxs_path="mgxs.h5")
+
+    # The three material-filled volumes (two fuel, one water) each get their own
+    # cloned material with a distinct macroscopic; the void cells are skipped.
+    assert model.settings.energy_mode == "multi-group"
+    assert len(model.materials) == 3
+    macros = [m._macroscopic for m in model.materials]
+    assert len(set(macros)) == 3

tests/unit_tests/test_model.py

@@ -1090,3 +1090,34 @@ def test_convert_to_multigroup_preserves_material_names(run_in_tmpdir):
     macro = [m._macroscopic for m in model.materials]
     assert macro == [f"Steel_Plate__1_{a.id}", f"Steel_Plate__1_{b.id}"]
     assert len(set(macro)) == 2
+
+
+def test_convert_to_multigroup_cell_wise(run_in_tmpdir):
+    """cell_wise clones the material in each cell so two cells sharing one
+    material end up with distinct (spatially-resolved) cross sections rather than a
+    single shared set."""
+    water = openmc.Material(name="water")
+    water.add_element("H", 2.0)
+    water.add_element("O", 1.0)
+    water.set_density("g/cm3", 1.0)
+
+    s1 = openmc.Sphere(r=1.0)
+    s2 = openmc.Sphere(r=2.0, boundary_type="vacuum")
+    c1 = openmc.Cell(fill=water, region=-s1)
+    c2 = openmc.Cell(fill=water, region=+s1 & -s2)  # same material, distinct cell
+    model = openmc.Model(openmc.Geometry([c1, c2]), openmc.Materials([water]))
+
+    # Pre-create the library so MGXS generation (and transport) is skipped; this
+    # exercises the per-cell material cloning and macroscopic assignment only.
+    Path("mgxs.h5").touch()
+    model.convert_to_multigroup(method="cell_wise", mgxs_path="mgxs.h5")
+
+    # Each cell now holds its own cloned material reading a distinct macroscopic.
+    assert len(model.materials) == 2
+    assert c1.fill is not c2.fill
+    assert c1.fill._macroscopic == f"water_{c1.fill.id}"
+    assert c2.fill._macroscopic == f"water_{c2.fill.id}"
+    assert c1.fill._macroscopic != c2.fill._macroscopic
+    # The user's original material object is left untouched.
+    assert water.name == "water"
+    assert model.settings.energy_mode == "multi-group"