Remove unused arguments on NormalizationHelper classes

openmc/deplete/abc.py

@@ -335,7 +335,7 @@ class NormalizationHelper(ABC):
         self._energy = 0.0
 
     @abstractmethod
-    def prepare(self, chain_nucs, rate_index, materials):
+    def prepare(self, chain_nucs, rate_index):
         """Perform work needed to obtain energy produced
 
         This method is called prior to the transport simulations
@@ -348,13 +348,9 @@ class NormalizationHelper(ABC):
         rate_index : dict of str to int
             Mapping from nuclide name to index in the
             `fission_rates` for :meth:`update`.
-        materials : list of str
-            All materials tracked on the operator helped by this
-            object. Should correspond to
-            :attr:`openmc.deplete.Operator.burnable_materials`
         """
 
-    def update(self, fission_rates, mat_index):
+    def update(self, fission_rates):
         """Update the energy produced
 
         Parameters
@@ -363,9 +359,6 @@ class NormalizationHelper(ABC):
             fission reaction rate for each isotope in the specified
             material. Should be ordered corresponding to initial
             ``rate_index`` used in :meth:`prepare`
-        mat_index : int
-            Index for the specific material in the list of all burnable
-            materials.
         """
 
     @property

openmc/deplete/helpers.py

@@ -113,7 +113,7 @@ class ChainFissionHelper(NormalizationHelper):
         super().__init__()
         self._fission_q_vector = None
 
-    def prepare(self, chain_nucs, rate_index, _materials):
+    def prepare(self, chain_nucs, rate_index):
         """Populate the fission Q value vector from a chain.
 
         Parameters
@@ -125,8 +125,6 @@ class ChainFissionHelper(NormalizationHelper):
             Dictionary mapping names of nuclides, e.g. ``"U235"``,
             to a corresponding index in the desired fission Q
             vector.
-        _materials : list of str
-            Unused. Materials to be tracked for this helper.
         """
         if (self._fission_q_vector is not None
                 and self._fission_q_vector.shape == (len(rate_index),)):
@@ -143,7 +141,7 @@ class ChainFissionHelper(NormalizationHelper):
 
         self._fission_q_vector = fission_qs
 
-    def update(self, fission_rates, _mat_index):
+    def update(self, fission_rates):
         """Update energy produced with fission rates in a material
 
         Parameters
@@ -152,9 +150,6 @@ class ChainFissionHelper(NormalizationHelper):
             fission reaction rate for each isotope in the specified
             material. Should be ordered corresponding to initial
             ``rate_index`` used in :meth:`prepare`
-        _mat_index : int
-            index for the material requested. Unused, as identical
-            isotopes in all materials have the same Q value.
         """
         self._energy += dot(fission_rates, self._fission_q_vector)
 

openmc/deplete/operator.py

@@ -492,7 +492,7 @@ class Operator(TransportOperator):
                      for i in self.burnable_mats]
         self._rate_helper.generate_tallies(materials, self.chain.reactions)
         self._normalization_helper.prepare(
-            self.chain.nuclides, self.reaction_rates.index_nuc, materials)
+            self.chain.nuclides, self.reaction_rates.index_nuc)
         # Tell fission yield helper what materials this process is
         # responsible for
         self._yield_helper.generate_tallies(
@@ -672,7 +672,7 @@ class Operator(TransportOperator):
             fission_yields.append(self._yield_helper.weighted_yields(i))
 
             # Accumulate energy from fission
-            self._normalization_helper.update(tally_rates[:, fission_ind], mat_index)
+            self._normalization_helper.update(tally_rates[:, fission_ind])
 
             # Divide by total number and store
             rates[i] = self._rate_helper.divide_by_adens(number)