Make n_bmat required only for FissionYieldCutoffHelper

Base class TalliedFissionYieldHelper and AveragedFissionYieldHelper
instances do not need to know the number of burnable materials,
and thus it is no longer a required input argument.

Changes propagated into the from_operator methods and
tests/unit_tests/test_deplete_fission_yields.py

Removed explicit TalliedFissionYieldHelper.from_operator
method as it is identical and now inherited from the
base FissionYieldHelper.from_operator method

openmc/deplete/abc.py

@@ -497,13 +497,9 @@ class TalliedFissionYieldHelper(FissionYieldHelper):
     chain_nuclides : iterable of openmc.deplete.Nuclide
         Nuclides tracked in the depletion chain. Not necessary
         that all have yield data.
-    n_bmats : int
-        Number of burnable materials tracked in the problem.
 
     Attributes
     ----------
-    n_bmats : int
-        Number of burnable materials tracked in the problem.
     constant_yields : dict of str to :class:`openmc.deplete.FissionYield`
         Fission yields for all nuclides that only have one set of
         fission yield data. Can be accessed as ``{parent: {product: yield}}``
@@ -513,9 +509,8 @@ class TalliedFissionYieldHelper(FissionYieldHelper):
 
     _upper_energy = 20.0e6  # upper energy for tallies
 
-    def __init__(self, chain_nuclides, n_bmats):
+    def __init__(self, chain_nuclides):
         super().__init__(chain_nuclides)
-        self.n_bmats = n_bmats
         self._local_indexes = None
         self._fission_rate_tally = None
         self._tally_nucs = []
@@ -578,24 +573,6 @@ class TalliedFissionYieldHelper(FissionYieldHelper):
         tally data.
         """
 
-    @classmethod
-    def from_operator(cls, operator, **kwargs):
-        """Create a new instance by pulling data from the operator
-
-        Require the more concrete :class:`openmc.deplete.Operator`
-        because this needs knowledge of burnable materials
-
-        Parameters
-        ----------
-        operator : openmc.deplete.TransportOperator
-            Operator with a depletion chain
-        kwargs: optional
-            Optional keyword arguments to be passed to the underlying
-            ``__init__`` method
-        """
-        return cls(operator.chain.nuclides, len(operator.burnable_mats),
-                   **kwargs)
-
 
 class Integrator(ABC):
     """Abstract class for solving the time-integration for depletion

openmc/deplete/helpers.py

@@ -299,7 +299,8 @@ class FissionYieldCutoffHelper(TalliedFissionYieldHelper):
         check_greater_than("thermal_energy", thermal_energy, 0.0, equality=True)
         check_greater_than("cutoff", cutoff, thermal_energy, equality=False)
         check_greater_than("fast_energy", fast_energy, cutoff, equality=False)
-        super().__init__(chain_nuclides, n_bmats)
+        self.n_bmats = n_bmats
+        super().__init__(chain_nuclides)
         self._cutoff = cutoff
         self._thermal_yields = {}
         self._fast_yields = {}
@@ -477,13 +478,9 @@ class AveragedFissionYieldHelper(TalliedFissionYieldHelper):
     chain_nuclides : iterable of openmc.deplete.Nuclide
         Nuclides tracked in the depletion chain. Not necessary
         that all have yield data.
-    n_bmats : int
-        Number of burnable materials tracked in the problem.
 
     Attributes
     ----------
-    n_bmats : int
-        Number of burnable materials tracked in the problem.
     constant_yields : dict of str to :class:`openmc.deplete.FissionYield`
         Fission yields for all nuclides that only have one set of
         fission yield data. Can be accessed as ``{parent: {product: yield}}``
@@ -494,8 +491,8 @@ class AveragedFissionYieldHelper(TalliedFissionYieldHelper):
         is the number of nuclides with multiple sets of fission yields.
     """
 
-    def __init__(self, chain_nuclides, n_bmats):
-        super().__init__(chain_nuclides, n_bmats)
+    def __init__(self, chain_nuclides):
+        super().__init__(chain_nuclides)
         self._weighted_tally = None
 
     def generate_tallies(self, materials, mat_indexes):
@@ -588,11 +585,11 @@ class AveragedFissionYieldHelper(TalliedFissionYieldHelper):
 
         Parameters
         ----------
-        operator : openmc.deplete.Operator
+        operator : openmc.deplete.TransportOperator
             Operator with a depletion chain
 
         Returns
         -------
         AveragedFissionYieldHelper
         """
-        return cls(operator.chain.nuclides, len(operator.burnable_mats))
+        return cls(operator.chain.nuclides)

tests/unit_tests/test_deplete_fission_yields.py

@@ -142,7 +142,7 @@ class AverageProxy(ProxyMixin, AveragedFissionYieldHelper):
 
 @pytest.mark.parametrize("avg_energy", (0.01, 100, 15e6))
 def test_averaged_helper(nuclide_bundle, avg_energy):
-    proxy = AverageProxy(nuclide_bundle, len(MATERIALS))
+    proxy = AverageProxy(nuclide_bundle)
     proxy.generate_tallies(MATERIALS, [0])
     tallied_nucs = proxy.update_nuclides_from_operator(
         [n.name for n in nuclide_bundle])