Add section in depletion user's guide about comparing to other codes (#3955)

docs/source/usersguide/depletion.rst

@@ -449,3 +449,42 @@ to transfer xenon from one material to another, you'd use::
     ...
 
     integrator.add_transfer_rate(mat1, ['Xe'], 0.1, destination_material=mat2)
+
+Comparing to Other Codes
+========================
+
+Comparing depletion results from OpenMC with those from another code, such as
+MCNP or Serpent, requires more than constructing equivalent transport models.
+At each depletion step, differences in the transport solution, nuclear data,
+reaction rate normalization, and numerical integration can all affect the
+result. Small differences can also accumulate over successive depletion steps.
+
+For a meaningful comparison, align as many of the following inputs and methods
+as possible:
+
+- Geometry and material definitions and associated physical properties such as
+  temperature
+- Neutron cross section library (e.g., ENDF/B-VIII.0)
+- Treatment of thermal scattering and unresolved resonance probability tables
+- Neutron reactions accounted for in the depletion chain
+- Decay data in the depletion chain
+- Isomeric branching ratios for reactions in the depletion chain
+- Fission product yields in the depletion chain
+- Fission product yield interpolation method
+  (``CoupledOperator(fission_yield_mode=...)``)
+- Reaction rate normalization, including fission Q values
+  (``CoupledOperator(fission_q=...)``)
+- Depletion integration method (``PredictorIntegrator``, ``CECMIntegrator``,
+  etc.) and time-step sizes
+
+When comparing to codes that use ACE format cross sections, it is recommended to
+directly convert the ACE files to HDF5 format using functionality from the
+:mod:`openmc.data` module (see :ref:`create_xs_library`). Some of the
+LANL-distributed ACE libraries used with MCNP have also been converted to HDF5
+format and are available for download at https://openmc.org/data.
+
+Even after these choices have been aligned, exact agreement should not be
+expected. Codes may use different approximations or numerical methods that
+cannot be configured identically. When investigating a discrepancy, first
+compare transport results and one-group reaction rates at the initial time, then
+compare changes over subsequent timesteps.

openmc/model/model.py

@@ -1750,8 +1750,8 @@ class Model:
     def _auto_generate_mgxs_lib(
         model: openmc.model.model,
         groups: openmc.mgxs.EnergyGroups,
-        correction: str | none,
-        directory: pathlike,
+        correction: str | None,
+        directory: PathLike,
     ) -> openmc.mgxs.Library:
         """
         Automatically generate a multi-group cross section libray from a model
@@ -1958,7 +1958,7 @@ class Model:
 
         # Set materials on the model
         model.materials = [material]
-        if temperature != None:
+        if temperature is not None:
           model.materials[-1].temperature = temperature
 
         # Settings
@@ -1985,7 +1985,7 @@ class Model:
         mgxs_lib = Model._auto_generate_mgxs_lib(
                 model, groups, correction, directory)
 
-        if temperature != None:
+        if temperature is not None:
             return mgxs_lib.get_xsdata(domain=material, xsdata_name=name,
                                        temperature=temperature)
         else:
@@ -2058,12 +2058,12 @@ class Model:
         )
 
         temp_settings = {}
-        if temperature_settings == None:
+        if temperature_settings is None:
             temp_settings = self.settings.temperature
         else:
             temp_settings = temperature_settings
 
-        if temperatures == None:
+        if temperatures is None:
             mgxs_sets = []
             for material in self.materials:
                 xs_data = Model._isothermal_infinite_media_mgxs(
@@ -2236,7 +2236,7 @@ class Model:
         model = openmc.Model()
         model.geometry = stoch_geom
 
-        if temperature != None:
+        if temperature is not None:
             for material in model.geometry.get_all_materials().values():
                 material.temperature = temperature
 
@@ -2260,7 +2260,7 @@ class Model:
                 model, groups, correction, directory)
 
         # Fetch all of the isothermal results.
-        if temperature != None:
+        if temperature is not None:
             return {
                 mat.name : mgxs_lib.get_xsdata(domain=mat, xsdata_name=mat.name,
                                                temperature=temperature)
@@ -2346,12 +2346,12 @@ class Model:
         )
 
         temp_settings = {}
-        if temperature_settings == None:
+        if temperature_settings is None:
             temp_settings = self.settings.temperature
         else:
             temp_settings = temperature_settings
 
-        if temperatures == None:
+        if temperatures is None:
             mgxs_sets = Model._isothermal_stochastic_slab_mgxs(
                 geo,
                 groups,
@@ -2444,7 +2444,7 @@ class Model:
         model = copy.deepcopy(input_model)
         model.tallies = openmc.Tallies()
 
-        if temperature != None:
+        if temperature is not None:
             for material in model.geometry.get_all_materials().values():
                 material.temperature = temperature
 
@@ -2460,7 +2460,7 @@ class Model:
                 model, groups, correction, directory)
 
         # Fetch all of the isothermal results.
-        if temperature != None:
+        if temperature is not None:
             return {
                 mat.name : mgxs_lib.get_xsdata(domain=mat, xsdata_name=mat.name,
                                                temperature=temperature)
@@ -2515,12 +2515,12 @@ class Model:
             entries in openmc.Settings.temperature_settings.
         """
         temp_settings = {}
-        if temperature_settings == None:
+        if temperature_settings is None:
             temp_settings = self.settings.temperature
         else:
             temp_settings = temperature_settings
 
-        if temperatures == None:
+        if temperatures is None:
             mgxs_sets = Model._isothermal_materialwise_mgxs(
                 self,
                 groups,

tests/unit_tests/test_lib.py

@@ -251,9 +251,9 @@ def test_material(lib_init):
     m.name = "Not hot borated water"
     assert m.name == "Not hot borated water"
 
-    assert m.depletable == False
+    assert not m.depletable
     m.depletable = True
-    assert m.depletable == True
+    assert m.depletable
 
 
 def test_properties_density(lib_init):