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Remove use of mgxs module in MicroXS.from_model #2572

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Merged
yardasol merged 2 commits into from
Jun 25, 2023
Merged

Remove use of mgxs module in MicroXS.from_model #2572

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4ed3cc6
Remove use of mgxs module in MicroXS.from_model
paulromano Jun 19, 2023
b87a9a9
Update exception message per @yardasol suggestion
paulromano Jun 22, 2023
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65 changes: 41 additions & 24 deletions openmc/deplete/microxs.py
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Original file line number Diff line number Diff line change
Expand Up @@ -7,13 +7,12 @@
import tempfile
from copy import deepcopy

from pandas import DataFrame, read_csv
from pandas import DataFrame, read_csv, Series
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 import Tallies, StatePoint, Materials
from openmc import StatePoint, Materials
import openmc
from .chain import Chain, REACTIONS
from .coupled_operator import _find_cross_sections, _get_nuclides_with_data
Expand Down Expand Up @@ -67,26 +66,36 @@ def from_model(cls,
Cross section data in [b]

"""
groups = EnergyGroups(energy_bounds)

# Set up the reaction tallies
original_tallies = model.tallies
original_materials = deepcopy(model.materials)
tallies = Tallies()
xs = {}
reactions, diluted_materials = cls._add_dilute_nuclides(chain_file,
model,
dilute_initial)
reactions, burnable_nucs, diluted_materials = cls._add_dilute_nuclides(
chain_file, model, dilute_initial)
model.materials = diluted_materials

for rx in reactions:
if rx == 'fission':
xs[rx] = FissionXS(domain=reaction_domain,
energy_groups=groups, by_nuclide=True)
else:
xs[rx] = ArbitraryXS(rx, domain=reaction_domain,
energy_groups=groups, by_nuclide=True)
tallies += xs[rx].tallies.values()
energy_filter = openmc.EnergyFilter(energy_bounds)
if isinstance(reaction_domain, openmc.Material):
domain_filter = openmc.MaterialFilter([reaction_domain])
elif isinstance(reaction_domain, openmc.Cell):
domain_filter = openmc.CellFilter([reaction_domain])
elif isinstance(reaction_domain, openmc.Universe):
domain_filter = openmc.UniverseFilter([reaction_domain])
else:
raise ValueError(f"Unsupported domain type: {type(reaction_domain)}")

# TODO: Right now, we use all nuclides from the material but it probably
# should be based on the burnable nuclides
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paulromano marked this conversation as resolved.
Resolved
rr_tally = openmc.Tally(name='MicroXS RR')
rr_tally.filters = [domain_filter, energy_filter]
rr_tally.nuclides = reaction_domain.get_nuclides()
rr_tally.multiply_density = False
rr_tally.scores = reactions

flux_tally = openmc.Tally(name='MicroXS flux')
flux_tally.filters = [domain_filter, energy_filter]
flux_tally.scores = ['flux']
tallies = openmc.Tallies([rr_tally, flux_tally])

model.tallies = tallies

Expand All @@ -98,14 +107,22 @@ def from_model(cls,
statepoint_path = model.run(**run_kwargs)

with StatePoint(statepoint_path) as sp:
for rx in xs:
xs[rx].load_from_statepoint(sp)
rr_tally = sp.tallies[rr_tally.id]
rr_tally._read_results()
flux_tally = sp.tallies[flux_tally.id]
flux_tally._read_results()

# Get reaction rates and flux values
reaction_rates = rr_tally.mean.sum(axis=0) # (nuclides, reactions)
flux = flux_tally.mean[0, 0, 0]

# Divide RR by flux to get microscopic cross sections
xs = reaction_rates / flux

# Build the DataFrame
# Build Series objects
series = {}
for rx in xs:
df = xs[rx].get_pandas_dataframe(xs_type='micro')
series[rx] = df.set_index('nuclide')['mean']
for i, rx in enumerate(reactions):
series[rx] = Series(xs[..., i], index=rr_tally.nuclides)

# Revert to the original tallies and materials
model.tallies = original_tallies
Expand Down Expand Up @@ -170,7 +187,7 @@ def _add_dilute_nuclides(cls, chain_file, model, dilute_initial):
dilute_density)
diluted_materials.append(material)

return reactions, diluted_materials
return reactions, burnable_nucs, diluted_materials

@classmethod
def from_array(cls, nuclides, reactions, data):
Expand Down