Fix formatting issues in test_void.py

Author: GuySten

tests/unit_tests/test_void.py

@@ -1,43 +1,43 @@
-import numpy as np
-import openmc
-import pytest
-
-
-@pytest.fixture
-def empty_sphere():
-    openmc.reset_auto_ids()
-    model = openmc.model.Model()
-    surf = openmc.Sphere(r=10, boundary_type = 'vacuum')
-    cell = openmc.Cell(region=-surf)
-    model.geometry = openmc.Geometry([cell])
-
-    model.settings.run_mode = 'fixed source'
-    model.settings.batches = 3
-    model.settings.particles = 1000
-    model.settings.source = openmc.IndependentSource(space=openmc.stats.Point())
-
-    tally = openmc.Tally()
-    tally.scores = ['total', elastic]
-    tally.nuclides = ['U235']
+import numpy as np
+import openmc
+import pytest
+
+
+@pytest.fixture
+def empty_sphere():
+    openmc.reset_auto_ids()
+    model = openmc.model.Model()
+    surf = openmc.Sphere(r=10, boundary_type = 'vacuum')
+    cell = openmc.Cell(region=-surf)
+    model.geometry = openmc.Geometry([cell])
+
+    model.settings.run_mode = 'fixed source'
+    model.settings.batches = 3
+    model.settings.particles = 1000
+    model.settings.source = openmc.IndependentSource(space=openmc.stats.Point())
+
+    tally = openmc.Tally()
+    tally.scores = ['total', 'elastic']
+    tally.nuclides = ['U235']
     tally.multiply_density = False
-    model.tallies.append(tally)
-
-    return model
-
-def test_equivalent_microxs(empty_sphere, run_in_tmpdir):
-    sp_file = empty_sphere.run()
-    with openmc.StatePoint(sp_file) as sp:
-        tally1 = sp.tallies[1]
-    
-    mat = openmc.Material()
-    mat.add_nuclide('H1', 1e-16) 
-     
-    empty_sphere.geometry.get_all_cells()[1].fill = mat
-    empty_sphere.materials.append(mat)
-    
-    sp_file = empty_sphere.run()
-    with openmc.StatePoint(sp_file) as sp:
-        tally2 = sp.tallies[1]
-    
+    model.tallies.append(tally)
+
+    return model
+
+def test_equivalent_microxs(empty_sphere, run_in_tmpdir):
+    sp_file = empty_sphere.run()
+    with openmc.StatePoint(sp_file) as sp:
+        tally1 = sp.tallies[1]
+    
+    mat = openmc.Material()
+    mat.add_nuclide('H1', 1e-16) 
+     
+    empty_sphere.geometry.get_all_cells()[1].fill = mat
+    empty_sphere.materials.append(mat)
+    
+    sp_file = empty_sphere.run()
+    with openmc.StatePoint(sp_file) as sp:
+        tally2 = sp.tallies[1]
+    
     assert np.isclose(tally1.mean.sum(),tally2.mean.sum(), rtol=1e-10, atol=0)
     assert tally1.mean.sum() > 0