diff --git a/openmc/deplete/abc.py b/openmc/deplete/abc.py
index fb638e5c21..3d4c93ac83 100644
--- a/openmc/deplete/abc.py
+++ b/openmc/deplete/abc.py
@@ -5,6 +5,7 @@ to run a full depletion simulation.
 """
 
 from collections import namedtuple
+from collections import defaultdict
 from collections.abc import Iterable
 import os
 from pathlib import Path
@@ -376,14 +377,17 @@ class FissionYieldHelper(ABC):
 
     Attributes
     ----------
-    constant_yields : dict of str to :class:`openmc.deplete.FissionYield`
+    constant_yields : collections.defaultdict
         Fission yields for all nuclides that only have one set of
-        fission yield data. Can be accessed as ``{parent: {product: yield}}``
+        fission yield data. Dictionary of form ``{str: {str: float}}``
+        representing yields for ``{parent: {product: yield}}``. Default
+        return object is an empty dictionary
+
     """
 
     def __init__(self, chain_nuclides):
         self._chain_nuclides = {}
-        self._constant_yields = {}
+        self._constant_yields = defaultdict(dict)
 
         # Get all nuclides with fission yield data
         for nuc in chain_nuclides:
@@ -407,14 +411,16 @@ class FissionYieldHelper(ABC):
         Parameters
         ----------
         local_mat_index : int
-            Index for material tracked on this process that
-            exists in :attr:`local_mat_index` and fits within
-            the first axis in :attr:`results`
+            Index for the material with requested fission yields.
+            Should correspond to the material represented in
+            ``mat_indexes[local_mat_index]`` during
+            :meth:`generate_tallies`.
 
         Returns
         -------
-        library : dict
-            Dictionary of ``{parent: {product: fyield}}``
+        library : collections.abc.Mapping
+            Dictionary-like object mapping ``{str: {str: float}``.
+            This reflects fission yields for ``{parent: {product: fyield}}``.
         """
 
     @staticmethod
diff --git a/openmc/deplete/helpers.py b/openmc/deplete/helpers.py
index 56644a3bba..f6e6bb04da 100644
--- a/openmc/deplete/helpers.py
+++ b/openmc/deplete/helpers.py
@@ -5,6 +5,7 @@ from copy import deepcopy
 from itertools import product
 from numbers import Real
 import bisect
+from collections import defaultdict
 
 from numpy import dot, zeros, newaxis
 
@@ -179,9 +180,11 @@ class ConstantFissionYieldHelper(FissionYieldHelper):
 
     Attributes
     ----------
-    constant_yields : dict of str to :class:`openmc.deplete.FissionYield`
+    constant_yields : collections.defaultdict
         Fission yields for all nuclides that only have one set of
-        fission yield data. Can be accessed as ``{parent: {product: yield}}``
+        fission yield data. Dictionary of form ``{str: {str: float}}``
+        representing yields for ``{parent: {product: yield}}``. Default
+        return object is an empty dictionary
     energy : float
         Energy of fission yield libraries.
     """
@@ -237,7 +240,7 @@ class ConstantFissionYieldHelper(FissionYieldHelper):
 
         Returns
         -------
-        library : dict
+        library : collections.defaultdict
             Dictionary of ``{parent: {product: fyield}}``
         """
         return self.constant_yields
@@ -282,6 +285,11 @@ class FissionYieldCutoffHelper(TalliedFissionYieldHelper):
     fast_yields : dict
         Dictionary of the form ``{parent: {product: yield}}``
         with fast yields
+    constant_yields : collections.defaultdict
+        Fission yields for all nuclides that only have one set of
+        fission yield data. Dictionary of form ``{str: {str: float}}``
+        representing yields for ``{parent: {product: yield}}``. Default
+        return object is an empty dictionary
     results : numpy.ndarray
         Array of fission rate fractions with shape
         ``(n_mats, 2, n_nucs)``. ``results[:, 0]``
@@ -417,7 +425,7 @@ class FissionYieldCutoffHelper(TalliedFissionYieldHelper):
 
         Returns
         -------
-        library : dict
+        library : collections.defaultdict
             Dictionary of ``{parent: {product: fyield}}``
         """
         yields = self.constant_yields
@@ -469,9 +477,11 @@ class AveragedFissionYieldHelper(TalliedFissionYieldHelper):
 
     Attributes
     ----------
-    constant_yields : dict of str to :class:`openmc.deplete.FissionYield`
+    constant_yields : collections.defaultdict
         Fission yields for all nuclides that only have one set of
-        fission yield data. Can be accessed as ``{parent: {product: yield}}``
+        fission yield data. Dictionary of form ``{str: {str: float}}``
+        representing yields for ``{parent: {product: yield}}``. Default
+        return object is an empty dictionary
     results : None or numpy.ndarray
         If tallies have been generated and unpacked, then the array will
         have shape ``(n_mats, n_tnucs)``, where ``n_mats`` is the number
@@ -569,12 +579,13 @@ class AveragedFissionYieldHelper(TalliedFissionYieldHelper):
 
         Returns
         -------
-        library : dict
-            Dictionary of ``{parent: {product: fyield}}``
+        library : collections.defaultdict
+            Dictionary of ``{parent: {product: fyield}}``. Default return
+            value is an empty dictionary
         """
         if not self._tally_nucs:
             return self.constant_yields
-        mat_yields = {}
+        mat_yields = defaultdict(dict)
         average_energies = self.results[local_mat_index]
         for avg_e, nuc in zip(average_energies, self._tally_nucs):
             nuc_energies = nuc.yield_energies