Clean up internals for AveragedFissionYieldHelper

Respond to some other reviewer comments for #1313

openmc/deplete/chain.py

@@ -731,6 +731,7 @@ class Chain(object):
                 yields = [yields]
             check_type("fission_yields", yields, Iterable, Mapping)
         self._fission_yields = yields
+
     def validate(self, strict=True, quiet=False, tolerance=1e-4):
         """Search for possible inconsistencies
 

openmc/deplete/helpers.py

@@ -425,11 +425,9 @@ class FissionYieldCutoffHelper(TalliedFissionYieldHelper):
             return yields
         rates = self.results[local_mat_index]
         # iterate over thermal then fast yields, prefer __mul__ to __rmul__
-        for therm_frac, nuc in zip(rates[0], self._tally_nucs):
-            yields[nuc.name] = self._thermal_yields[nuc.name] * therm_frac
-
-        for fast_frac, nuc in zip(rates[1], self._tally_nucs):
-            yields[nuc.name] += self._fast_yields[nuc.name] * fast_frac
+        for therm_frac, fast_frac, nuc in zip(rates[0], rates[1], self._tally_nucs):
+            yields[nuc.name] = (self._thermal_yields[nuc.name] * therm_frac
+                                + self._fast_yields[nuc.name] * fast_frac)
         return yields
 
     @property
@@ -444,9 +442,7 @@ class FissionYieldCutoffHelper(TalliedFissionYieldHelper):
 class AveragedFissionYieldHelper(TalliedFissionYieldHelper):
     r"""Class that computes fission yields based on average fission energy
 
-    Computes average energy at which fission events occured
-    reactions for all nuclides with multiple sets of fission yields
-    by
+    Computes average energy at which fission events occured with
 
     .. math::
 
@@ -481,6 +477,8 @@ class AveragedFissionYieldHelper(TalliedFissionYieldHelper):
         have shape ``(n_mats, n_tnucs)``, where ``n_mats`` is the number
         of materials where fission reactions were tallied and ``n_tnucs``
         is the number of nuclides with multiple sets of fission yields.
+        Data in the array are the average energy of fission events for
+        tallied nuclides across burnable materials.
     """
 
     def __init__(self, chain_nuclides):

openmc/deplete/operator.py

@@ -86,7 +86,7 @@ class Operator(TransportOperator):
         in initial condition to ensure they exist in the decay chain.
         Only done for nuclides with reaction rates.
         Defaults to 1.0e3.
-    fission_yield_mode : ("constant", "cutoff", "average")
+    fission_yield_mode : {"constant", "cutoff", "average"}
         Key indicating what fission product yield scheme to use. The
         key determines what fission energy helper is used:
 

tests/unit_tests/test_deplete_fission_yields.py

@@ -205,7 +205,7 @@ def test_cutoff_helper(materials, nuclide_bundle, therm_frac):
 
     non_zero_nucs = [n.name for n in nuclide_bundle]
     tally_nucs = helper.update_tally_nuclides(non_zero_nucs)
-    assert tally_nucs == ["Pu239", "U235",]
+    assert tally_nucs == ["Pu239", "U235"]
 
     # Check tallies
     fission_tally = helper._fission_rate_tally