Remove FissionYieldHelper.libraries; Return from compute_yields

Previously the yields were appended into the libraries
attribute. Now, the yields are returned directly and not
appended at all. It is up to the caller to place these yields
in the correct location.

The Operator maintains a list of of the libraries that is updated
through the unpacking method, and then set onto the Chain
after working through all local materials

openmc/deplete/helpers.py

@@ -163,7 +163,7 @@ class FissionYieldHelper(object):
 
     Parameters
     ----------
-    chain_nuclides : iterable of openmc.deplete.nuclide
+    chain_nuclides : iterable of openmc.deplete.Nuclide
         Nuclides tracked in the depletion chain. Not necessary
         that all have yield data.
     n_bmats : int
@@ -176,15 +176,9 @@ class FissionYieldHelper(object):
     results : numpy.ndarray
         Array of tally results for this process with shape
         ``(n_local_mat, n_energy, n_nucs)``
-    libraries : list of dict
-        List of fission yield dictionaries of the form
-        ``{parent: {product: yield}}``. Populated in
-        :meth:`compute_yields` and reset during
-        :meth:`unpack`.
     """
 
     def __init__(self, chain_nuclides, n_bmats):
-        self.libraries = []
         self._chain_nuclides = {}
         self._chain_set = set()
         self._tally_map = {}
@@ -245,7 +239,7 @@ class FissionYieldHelper(object):
         Parameters
         ----------
         nuclides : iterable of str
-            nuclides with non-zero densities that are candidates
+            Nuclides with non-zero densities that are candidates
             for the fission tally. Not necessary that all are nuclides
             with fission yields, but at least one must be
 
@@ -275,9 +269,6 @@ class FissionYieldHelper(object):
 
     def unpack(self):
         """Unpack fission rate tallies to produce :attr:`results`
-
-        Resets :attr:`libraries` under the assumption this is called
-        during the :class:`openmc.deplete.Operator` unpacking process
         """
         # if this process is not responsible for depleting anything
         # [more processes than burnable materials]
@@ -285,9 +276,6 @@ class FissionYieldHelper(object):
         if self.local_indexes.size == 0:
             return
 
-        # clear old libraries
-        self.libraries = []
-
         # get view into tally results
         # new shape: [material, energy, parent nuclide]
         result_view = self._reaction_tally.results[..., 1].reshape(
@@ -312,7 +300,7 @@ class FissionYieldHelper(object):
     def compute_yields(self, local_mat_index):
         """Compute single fission yields using :attr:`results`
 
-        Produces a new library in :attr:`self.libraries`
+        Produces a new library in :attr:`libraries`
 
         Parameters
         ----------
@@ -320,6 +308,11 @@ class FissionYieldHelper(object):
             Index for material tracked on this process that
             exists in :attr:`local_mat_index` and fits within
             the first axis in :attr:`results`
+
+        Returns
+        -------
+        library : dict
+            Dictionary of ``{parent: {product: fyield}}``
         """
         tally_results = self.results[local_mat_index]
 
@@ -337,9 +330,8 @@ class FissionYieldHelper(object):
                 initial_library[parent] += yield_data * fiss_frac
 
         # convert to dictionary that can be passed to Chain.form_matrix
-        # {parent: {product: yield}}
         library = {}
         for k, yield_obj in initial_library.items():
             library[k.name] = dict(zip(yield_obj.products, yield_obj.yields))
 
-        self.libraries.append(library)
+        return library

openmc/deplete/operator.py

@@ -565,6 +565,9 @@ class Operator(TransportOperator):
         self._energy_helper.reset()
         self._fsn_yield_helper.unpack()
 
+        # Store fission yield dictionaries
+        fission_yields = []
+
         # Create arrays to store fission Q values, reaction rates, and nuclide
         # numbers, zeroed out in material iteration
         number = np.empty(rates.n_nuc)
@@ -587,7 +590,7 @@ class Operator(TransportOperator):
                 mat_index, nuc_ind, react_ind)
 
             # Compute fission yields for this material
-            self._fsn_yield_helper.compute_yields(i)
+            fission_yields.append(self._fsn_yield_helper.compute_yields(i))
 
             # Accumulate energy from fission
             self._energy_helper.update(tally_rates[:, fission_ind], mat_index)
@@ -603,7 +606,7 @@ class Operator(TransportOperator):
         rates *= power / energy
 
         # Store new fission yields on the chain
-        self.chain.fission_yields = self._fsn_yield_helper.libraries
+        self.chain.fission_yields = fission_yields
 
         return OperatorResult(k_combined, rates)
 

tests/unit_tests/test_deplete_helpers.py

@@ -91,9 +91,6 @@ def test_fission_yield_helper():
         }
     }
 
-    assert len(helper.libraries) == 0
-    helper.compute_yields(0)
-    assert len(helper.libraries) == 1
-    act_library = helper.libraries[0]
+    act_library = helper.compute_yields(0)
     for parent, sub_yields in exp_lib.items():
         assert act_library[parent] == pytest.approx(sub_yields)