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

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39
docs/source/usersguide/depletion.rst
View file

@ -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.

44
openmc/model/model.py
View file

@ -265,22 +265,22 @@ class Model:
denom_tally = openmc.Tally(name='IFP denominator')
denom_tally.scores = ['ifp-denominator']
self.tallies.append(denom_tally)
# TODO: This should also be incorporated into lower-level calls in
# TODO: This should also be incorporated into lower-level calls in
# settings.py, but it requires information about the tallies currently
# on the active Model
def _assign_fw_cadis_tally_IDs(self):
# Verify that all tallies assigned as targets on WeightWindowGenerators
# exist within model.tallies. If this is the case, convert the .targets
# Verify that all tallies assigned as targets on WeightWindowGenerators
# exist within model.tallies. If this is the case, convert the .targets
# attribute of each WeightWindowGenerator to a sequence of tally IDs.
if len(self.settings.weight_window_generators) == 0:
return
# List of valid tally IDs
reference_tally_ids = np.asarray([tal.id for tal in self.tallies])
for wwg in self.settings.weight_window_generators:
# Only proceeds if the "targets" attribute is an openmc.Tallies,
# Only proceeds if the "targets" attribute is an openmc.Tallies,
# which means it hasn't been checked against model.tallies.
if isinstance(wwg.targets, openmc.Tallies):
id_vec = []
@ -291,7 +291,7 @@ class Model:
if tal == reference_tal:
id_next = reference_tal.id
break
if id_next == None:
raise RuntimeError(
f'Local FW-CADIS target tally {tal.id} not found on model.tallies!')
@ -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,

4
tests/unit_tests/test_lib.py
View file

@ -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):