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Address @shimwell comments on #2211

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This commit is contained in:
Paul Romano 2022-09-21 07:26:27 -05:00
parent a95f4cc0d7
commit 3645a41861
3 changed files with 20 additions and 13 deletions
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2
openmc/deplete/coupled_operator.py
View file

@ -101,7 +101,7 @@ class CoupledOperator(OpenMCOperator):
model : openmc.model.Model
OpenMC model object
chain_file : str, optional
Path to the depletion chain XML file. Defaults to
Path to the depletion chain XML file. Defaults to
``openmc.config['chain_file']``.
prev_results : Results, optional
Results from a previous depletion calculation. If this argument is

5
openmc/deplete/independent_operator.py
View file

@ -40,7 +40,8 @@ class IndependentOperator(OpenMCOperator):
micro_xs : MicroXS
One-group microscopic cross sections in [b] .
chain_file : str
Path to the depletion chain XML file.
Path to the depletion chain XML file. Defaults to
``openmc.config['chain_file']``.
keff : 2-tuple of float, optional
keff eigenvalue and uncertainty from transport calculation.
Default is None.
@ -111,7 +112,7 @@ class IndependentOperator(OpenMCOperator):
def __init__(self,
materials,
micro_xs,
chain_file,
chain_file=None,
keff=None,
normalization_mode='fission-q',
fission_q=None,

26
openmc/deplete/microxs.py
View file

@ -5,18 +5,16 @@ nuclide names as row indices and reaction names as column indices.
"""
import tempfile
from pathlib import Path
from copy import deepcopy
from pandas import DataFrame, read_csv, concat
from pandas import DataFrame, read_csv
import numpy as np
from openmc.checkvalue import check_type, check_value, check_iterable_type
from openmc.exceptions import DataError
from openmc.mgxs import EnergyGroups, ArbitraryXS, FissionXS
from openmc.data import DataLibrary
from openmc import Tallies, StatePoint, Materials, Material
from openmc import Tallies, StatePoint, Materials
import openmc
from .chain import Chain, REACTIONS
from .coupled_operator import _find_cross_sections, _get_nuclides_with_data
@ -35,7 +33,7 @@ class MicroXS(DataFrame):
def from_model(cls,
model,
reaction_domain,
chain_file,
chain_file=None,
dilute_initial=1.0e3,
energy_bounds=(0, 20e6),
run_kwargs=None):
@ -48,11 +46,12 @@ class MicroXS(DataFrame):
OpenMC model object. Must contain geometry, materials, and settings.
reaction_domain : openmc.Material or openmc.Cell or openmc.Universe or openmc.RegularMesh
Domain in which to tally reaction rates.
chain_file : str
chain_file : str, optional
Path to the depletion chain XML file that will be used in depletion
simulation. Used to determine cross sections for materials not
present in the inital composition.
dilute_initial : float
present in the inital composition. Defaults to
``openmc.config['chain_file']``.
dilute_initial : float, optional
Initial atom density [atoms/cm^3] to add for nuclides that
are zero in initial condition to ensure they exist in the cross
section data. Only done for nuclides with reaction rates.
@ -144,6 +143,13 @@ class MicroXS(DataFrame):
:class:`openmc.Materials` object with nuclides added to burnable
materials.
"""
if chain_file is None:
chain_file = openmc.config.get('chain_file')
if chain_file is None:
raise DataError(
"No depletion chain specified and could not find depletion "
"chain in openmc.config['chain_file']"
)
chain = Chain.from_xml(chain_file)
reactions = chain.reactions
cross_sections = _find_cross_sections(model)