Adding tally filter type option to statepoint get_tally (#3584)

Co-authored-by: Jon Shimwell <jon@proximafusion.com>
Co-authored-by: GuySten <62616591+GuySten@users.noreply.github.com>

openmc/statepoint.py

@@ -536,7 +536,7 @@ class StatePoint:
     def get_tally(self, scores=[], filters=[], nuclides=[],
                   name=None, id=None, estimator=None, exact_filters=False,
                   exact_nuclides=False, exact_scores=False,
-                  multiply_density=None, derivative=None):
+                  multiply_density=None, derivative=None, filter_type=None):
         """Finds and returns a Tally object with certain properties.
 
         This routine searches the list of Tallies and returns the first Tally
@@ -580,6 +580,9 @@ class StatePoint:
             to the same value as this parameter.
         derivative : openmc.TallyDerivative, optional
             TallyDerivative object to match.
+        filter_type : type, optional
+            If not None, the Tally must have at least one Filter that is an
+            instance of this type. For example `openmc.MeshFilter`.
 
         Returns
         -------
@@ -653,6 +656,10 @@ class StatePoint:
                 if not contains_filters:
                     continue
 
+            if filter_type is not None:
+                if not any(isinstance(f, filter_type) for f in test_tally.filters):
+                    continue
+
             # Determine if Tally has the queried Nuclide(s)
             if nuclides:
                 if not all(nuclide in test_tally.nuclides for nuclide in nuclides):

tests/unit_tests/test_statepoint.py

@@ -0,0 +1,65 @@
+import openmc
+
+
+def test_get_tally_filter_type(run_in_tmpdir):
+    """Test various ways of retrieving tallies from a StatePoint object."""
+
+    mat = openmc.Material()
+    mat.add_nuclide("H1", 1.0)
+    mat.set_density("g/cm3", 10.0)
+
+    sphere = openmc.Sphere(r=10.0, boundary_type="vacuum")
+    cell = openmc.Cell(fill=mat, region=-sphere)
+    geometry = openmc.Geometry([cell])
+
+    settings = openmc.Settings()
+    settings.particles = 10
+    settings.batches = 2
+    settings.run_mode = "fixed source"
+
+    reg_mesh = openmc.RegularMesh().from_domain(cell)
+    tally1 = openmc.Tally(tally_id=1)
+    mesh_filter = openmc.MeshFilter(reg_mesh)
+    tally1.filters = [mesh_filter]
+    tally1.scores = ["flux"]
+
+    tally2 = openmc.Tally(tally_id=2, name="heating tally")
+    cell_filter = openmc.CellFilter(cell)
+    tally2.filters = [cell_filter]
+    tally2.scores = ["heating"]
+
+    tallies = openmc.Tallies([tally1, tally2])
+    model = openmc.Model(
+        geometry=geometry, materials=[mat], settings=settings, tallies=tallies
+    )
+
+    sp_filename = model.run()
+
+    sp = openmc.StatePoint(sp_filename)
+
+    tally_found = sp.get_tally(filter_type=openmc.MeshFilter)
+    assert tally_found.id == 1
+
+    tally_found = sp.get_tally(filter_type=openmc.CellFilter)
+    assert tally_found.id == 2
+
+    tally_found = sp.get_tally(filters=[mesh_filter])
+    assert tally_found.id == 1
+
+    tally_found = sp.get_tally(filters=[cell_filter])
+    assert tally_found.id == 2
+
+    tally_found = sp.get_tally(scores=["heating"])
+    assert tally_found.id == 2
+
+    tally_found = sp.get_tally(name="heating tally")
+    assert tally_found.id == 2
+
+    tally_found = sp.get_tally(name=None)
+    assert tally_found.id == 1
+
+    tally_found = sp.get_tally(id=1)
+    assert tally_found.id == 1
+
+    tally_found = sp.get_tally(id=2)
+    assert tally_found.id == 2