Apply suggestions from code review

Co-Authored-By: Paul Romano <paul.k.romano@gmail.com>

openmc/deplete/chain.py

@@ -127,7 +127,7 @@ class Chain(object):
     fission_yields : None or iterable of dict
         List of effective fission yields for materials. Each dictionary
         should be of the form ``{parent: {product: yield}}`` with
-        types ``{str: {str: Real}}``, where ``yield`` is the fission product yield
+        types ``{str: {str: float}}``, where ``yield`` is the fission product yield
         for isotope ``parent`` producing isotope ``product``.
         A single entry indicates yields are constant across all materials.
         Otherwise, an entry can be added for each material to be burned.
@@ -395,14 +395,14 @@ class Chain(object):
         Returns
         -------
         fission_yields : dict
-            Dictionary of ``{parent : {product : f_yield}}``
+            Dictionary of ``{parent: {product: f_yield}}``
             where ``parent`` and ``product`` are both string
             names of nuclides with yield data and ``f_yield``
             is a float for the fission yield.
         """
         out = {}
         for nuc in self.nuclides:
-            if len(nuc.yield_data) == 0:
+            if not nuc.yield_data:
                 continue
             yield_obj = nuc.yield_data[min(nuc.yield_energies)]
             out[nuc.name] = dict(yield_obj)

openmc/deplete/helpers.py

@@ -317,7 +317,7 @@ class FissionYieldCutoffHelper(TalliedFissionYieldHelper):
             if fast is None:
                 # find first index <= cutoff
                 rev_ix = self._find_fallback_energy(
-                    name, list(reversed(energies)), cutoff, False)
+                    name, reversed(energies), cutoff, False)
                 fast = yields[energies[-rev_ix]]
             self._thermal_yields[name] = thermal
             self._fast_yields[name] = fast
@@ -518,7 +518,7 @@ class AveragedFissionYieldHelper(TalliedFissionYieldHelper):
         fission_tally = self._fission_rate_tally
 
         weighted_tally = Tally()
-        weighted_tally.filters = fission_tally.filters.copy()
+        weighted_tally.filters = fission_tally.filters
         weighted_tally.scores = ['fission']
 
         ene_bin = EnergyFilter()

tests/unit_tests/test_deplete_chain.py

@@ -129,7 +129,7 @@ def test_from_xml(simple_chain):
     assert [r.branching_ratio for r in nuc.reactions] == [1.0, 0.7, 0.3]
 
     # Yield tests
-    assert nuc.yield_energies == (0.0253, )
+    assert nuc.yield_energies == (0.0253,)
     assert list(nuc.yield_data) == [0.0253]
     assert nuc.yield_data[0.0253].products == ("A", "B")
     assert (nuc.yield_data[0.0253].yields == [0.0292737, 0.002566345]).all()

tests/unit_tests/test_deplete_fission_yields.py

@@ -91,7 +91,7 @@ def test_cutoff_failure(key):
         FissionYieldCutoffHelper(None, None, **{key: -1})
 
 
-class ProxyMixin(object):
+class ProxyMixin:
     """Mixing that overloads the tally generation"""
     def generate_tallies(self, materials, mat_indexes):
         self._fission_rate_tally = Mock()

tests/unit_tests/test_deplete_nuclide.py

@@ -76,7 +76,7 @@ def test_from_xml():
         0.0253: {"Xe138": 0.0481413, "Zr100": 0.0497641, "Te134": 0.062155}})
     assert u235.yield_data == expected_yield_data
     # test accessing the yield energies through the FissionYieldDistribution
-    assert u235.yield_energies == (0.0253, )
+    assert u235.yield_energies == (0.0253,)
     assert u235.yield_energies is u235.yield_data.energies
     with pytest.raises(AttributeError):  # not settable
         u235.yield_energies = [0.0253, 5e5]