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Allow decay only nuclides to be specified in Material for a depletion simulation (#2616)

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Co-authored-by: Jonathan Shimwell <drshimwell@gmail.com>
Co-authored-by: Paul Romano <paul.k.romano@gmail.com>
This commit is contained in:
Olek 2023-09-13 20:00:52 -05:00 • committed by GitHub
parent 0cf29f1459
commit bbd7756930
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GPG key ID: 4AEE18F83AFDEB23
7 changed files with 233 additions and 20 deletions
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2
openmc/deplete/coupled_operator.py
View file

@ -416,7 +416,7 @@ class CoupledOperator(OpenMCOperator):
for mat in self.materials:
mat._nuclides.sort(key=lambda x: nuclides.index(x[0]))
self.materials.export_to_xml()
self.materials.export_to_xml(nuclides_to_ignore=self._decay_nucs)
def __call__(self, vec, source_rate):
"""Runs a simulation.

30
openmc/deplete/openmc_operator.py
View file

@ -6,6 +6,7 @@ transport-independent transport operators.
"""
from abc import abstractmethod
from warnings import warn
from typing import List, Tuple, Dict
import numpy as np
@ -133,6 +134,18 @@ class OpenMCOperator(TransportOperator):
# This nuclides variables contains every nuclides
# for which there is an entry in the micro_xs parameter
openmc.reset_auto_ids()
self.nuclides_with_data = self._get_nuclides_with_data(
self.cross_sections)
# Select nuclides with data that are also in the chain
self._burnable_nucs = [nuc.name for nuc in self.chain.nuclides
if nuc.name in self.nuclides_with_data]
# Select nuclides without data that are also in the chain
self._decay_nucs = [nuc.name for nuc in self.chain.nuclides
if nuc.name not in self.nuclides_with_data]
self.burnable_mats, volumes, all_nuclides = self._get_burnable_mats()
self.local_mats = _distribute(self.burnable_mats)
@ -142,13 +155,6 @@ class OpenMCOperator(TransportOperator):
if self.prev_res is not None:
self._load_previous_results()
self.nuclides_with_data = self._get_nuclides_with_data(
self.cross_sections)
# Select nuclides with data that are also in the chain
self._burnable_nucs = [nuc.name for nuc in self.chain.nuclides
if nuc.name in self.nuclides_with_data]
# Extract number densities from the geometry / previous depletion run
self._extract_number(self.local_mats,
volumes,
@ -171,7 +177,7 @@ class OpenMCOperator(TransportOperator):
Returns
-------
burnable_mats : list of str
List of burnable material IDs
list of burnable material IDs
volume : dict of str to float
Volume of each material in [cm^3]
nuclides : list of str
@ -188,7 +194,13 @@ class OpenMCOperator(TransportOperator):
# Iterate once through the geometry to get dictionaries
for mat in self.materials:
for nuclide in mat.get_nuclides():
model_nuclides.add(nuclide)
if nuclide in self.nuclides_with_data or self._decay_nucs:
model_nuclides.add(nuclide)
else:
msg = (f"Nuclilde {nuclide} in material {mat.id} is not "
"present in the depletion chain and has no cross "
"section data.")
raise warn(msg)
if mat.depletable:
burnable_mats.add(str(mat.id))
if mat.volume is None:

40
openmc/material.py
View file

@ -1250,7 +1250,7 @@ class Material(IDManagerMixin):
return memo[self]
def _get_nuclide_xml(self, nuclide: str) -> ET.Element:
def _get_nuclide_xml(self, nuclide: NuclideTuple) -> ET.Element:
xml_element = ET.Element("nuclide")
xml_element.set("name", nuclide.name)
@ -1267,15 +1267,28 @@ class Material(IDManagerMixin):
return xml_element
def _get_nuclides_xml(self, nuclides: typing.Iterable[str]) -> List[ET.Element]:
def _get_nuclides_xml(
self, nuclides: typing.Iterable[NuclideTuple],
nuclides_to_ignore: Optional[typing.Iterable[str]] = None)-> List[ET.Element]:
xml_elements = []
for nuclide in nuclides:
xml_elements.append(self._get_nuclide_xml(nuclide))
# Remove any nuclides to ignore from the XML export
if nuclides_to_ignore:
nuclides = [nuclide for nuclide in nuclides if nuclide.name not in nuclides_to_ignore]
xml_elements = [self._get_nuclide_xml(nuclide) for nuclide in nuclides]
return xml_elements
def to_xml_element(self) -> ET.Element:
def to_xml_element(
self, nuclides_to_ignore: Optional[typing.Iterable[str]] = None) -> ET.Element:
"""Return XML representation of the material
Parameters
----------
nuclides_to_ignore : list of str
Nuclides to ignore when exporting to XML.
Returns
-------
element : lxml.etree._Element
@ -1320,7 +1333,8 @@ class Material(IDManagerMixin):
if self._macroscopic is None:
# Create nuclide XML subelements
subelements = self._get_nuclides_xml(self._nuclides)
subelements = self._get_nuclides_xml(self._nuclides,
nuclides_to_ignore=nuclides_to_ignore)
for subelement in subelements:
element.append(subelement)
else:
@ -1576,7 +1590,8 @@ class Materials(cv.CheckedList):
for material in self:
material.make_isotropic_in_lab()
def _write_xml(self, file, header=True, level=0, spaces_per_level=2, trailing_indent=True):
def _write_xml(self, file, header=True, level=0, spaces_per_level=2,
trailing_indent=True, nuclides_to_ignore=None):
"""Writes XML content of the materials to an open file handle.
Parameters
@ -1591,6 +1606,8 @@ class Materials(cv.CheckedList):
Number of spaces per indentation
trailing_indentation : bool
Whether or not to write a trailing indentation for the materials element
nuclides_to_ignore : list of str
Nuclides to ignore when exporting to XML.
"""
indentation = level*spaces_per_level*' '
@ -1611,7 +1628,7 @@ class Materials(cv.CheckedList):
# Write the <material> elements.
for material in sorted(self, key=lambda x: x.id):
element = material.to_xml_element()
element = material.to_xml_element(nuclides_to_ignore=nuclides_to_ignore)
clean_indentation(element, level=level+1)
element.tail = element.tail.strip(' ')
file.write((level+1)*spaces_per_level*' ')
@ -1626,13 +1643,16 @@ class Materials(cv.CheckedList):
if trailing_indent:
file.write(indentation)
def export_to_xml(self, path: PathLike = 'materials.xml'):
def export_to_xml(self, path: PathLike = 'materials.xml',
nuclides_to_ignore: Optional[typing.Iterable[str]] = None):
"""Export material collection to an XML file.
Parameters
----------
path : str
Path to file to write. Defaults to 'materials.xml'.
nuclides_to_ignore : list of str
Nuclides to ignore when exporting to XML.
"""
# Check if path is a directory
@ -1645,7 +1665,7 @@ class Materials(cv.CheckedList):
# one go.
with open(str(p), 'w', encoding='utf-8',
errors='xmlcharrefreplace') as fh:
self._write_xml(fh)
self._write_xml(fh, nuclides_to_ignore=nuclides_to_ignore)
@classmethod
def from_xml_element(cls, elem) -> Material:

66
tests/chain_simple_decay.xml Normal file
View file

@ -0,0 +1,66 @@
<?xml version="1.0"?>
<depletion_chain>
<nuclide name="I135" decay_modes="1" reactions="1" half_life="2.36520E+04" decay_energy="1916827.5">
<decay type="beta" target="Xe135" branching_ratio="1.0" />
<reaction type="(n,gamma)" Q="0.0" target="Xe136" /> <!-- Not precisely true, but whatever -->
<source type="discrete" particle="photon">
<parameters>3696.125 4095.822 4477.27 5097.122 29452.1 29781.3 33566.5 33629.4 33865.1 33878.5 34395.3 34408.0 34486.2 34488.2 112780.0 113150.0 162650.0 165740.0 184490.0 197190.0 220502.0 229720.0 247500.0 254740.0 264260.0 288451.0 290270.0 304910.0 305830.0 326000.0 333600.0 342520.0 361850.0 403030.0 414830.0 417633.0 429930.0 433741.0 451630.0 530800.0 546557.0 575970.0 588280.0 616900.0 649850.0 656090.0 679220.0 684600.0 690130.0 707920.0 785480.0 795500.0 797710.0 807200.0 836804.0 960290.0 961430.0 971960.0 972620.0 995090.0 1038760.0 1096860.0 1101580.0 1124000.0 1131511.0 1151510.0 1159900.0 1169040.0 1225600.0 1240470.0 1254800.0 1260409.0 1315770.0 1334800.0 1343660.0 1367890.0 1441800.0 1448350.0 1457560.0 1502790.0 1521990.0 1543700.0 1566410.0 1678027.0 1706459.0 1791196.0 1830690.0 1845300.0 1927300.0 1948490.0 2045880.0 2112400.0 2151500.0 2189400.0 2255457.0 2408650.0 2466070.0 2477100.0 9.714352819815078e-10 7.460941651551526e-09 6.047882056201745e-09 8.510389107205747e-10 3.7979729633684727e-08 7.033747061198817e-08 6.602815946851458e-09 1.2800909198245071e-08 6.513060244589454e-11 8.894667887850525e-11 1.3843783302275766e-09 2.700005498135763e-09 1.2141115256573815e-11 1.654893875618862e-11 3.7007705885806645e-09 2.0186021392258173e-09 2.859686363903241e-09 9.16781804898392e-09 6.896890642354875e-09 9.588360161322631e-09 5.130613770532286e-07 7.06510748729036e-08 8.410842246774237e-09 6.7286737974193905e-09 5.3829390379355124e-08 9.083709626516178e-07 8.915492781580693e-08 9.251926471451662e-09 2.783988783682273e-08 6.728673797419391e-10 1.093409492080651e-08 2.5232526740322717e-10 5.467047460403255e-08 6.812782219887132e-08 8.83138435911295e-08 1.0345335963532313e-06 8.915492781580693e-08 1.623292553627428e-07 9.251926471451663e-08 9.251926471451662e-09 2.0942997194467853e-06 3.784879011048407e-08 1.513951604419363e-08 1.093409492080651e-08 1.337323917237104e-07 2.186818984161302e-08 1.5980600268871052e-08 6.7286737974193905e-09 3.784879011048407e-08 1.9344937167580748e-07 4.457746390790346e-08 6.7286737974193905e-09 5.0465053480645434e-08 1.3457347594838781e-08 1.9597262434983976e-06 1.0093010696129087e-08 4.289529545854862e-08 2.607361096500014e-07 3.53255374364518e-07 4.541854813258089e-08 2.329803302356464e-06 2.607361096500014e-08 4.7100716581935733e-07 1.059766123093554e-06 6.6193328482113254e-06 4.205421123387119e-10 3.027903208838726e-08 2.565306885266143e-07 1.2616263370161358e-08 2.649415307733885e-07 3.3643368987096953e-09 8.410842246774237e-06 1.934493716758075e-08 9.251926471451662e-09 2.2709274066290446e-08 1.7830985563161385e-07 5.046505348064544e-09 9.251926471451663e-08 2.5400743585258203e-06 3.154065842540339e-07 1.093409492080651e-08 7.569758022096815e-09 3.784879011048407e-07 2.8008104681758214e-06 1.2027504412887161e-06 2.26251656438227e-06 1.6989901338483962e-07 1.6821684493548476e-09 8.663167514177466e-08 1.8503852942903323e-08 2.5568960430193683e-07 2.0186021392258175e-08 6.560456952483906e-09 3.784879011048407e-09 1.7999202408096872e-07 2.800810468175822e-07 2.1027105616935597e-08 4.205421123387119e-10</parameters>
</source>
</nuclide>
<nuclide name="Xe135" reactions="1" decay_energy="567890.1">
<reaction type="(n,gamma)" Q="0.0" target="Xe136" />
<source type="tabular" interpolation="histogram" particle="photon">
<parameters>0.0 10000.0 20000.0 50000.0 100000.0 200000.0 300000.0 400000.0 600000.0 800000.0 1000000.0 1220000.0 1.1612176249914943e-11 0.0 3.1976524142990123e-11 0.0 6.418466676129901e-13 1.894551923065874e-10 5.099949011192198e-13 3.170644340540729e-13 2.756798470867507e-12 6.67627508515641e-14 3.711746246444946e-15 0.0</parameters>
</source>
</nuclide>
<nuclide name="Xe135_m1" half_life="917.4" decay_modes="1" decay_energy="526729.1900000001" reactions="0">
<decay type="IT" target="Xe135" branching_ratio="1"/>
</nuclide>
<nuclide name="Xe136" decay_modes="0" reactions="0" />
<nuclide name="Cs135" decay_modes="0" reactions="0" />
<nuclide name="Cs135_m1" half_life="3180.0" decay_modes="1" decay_energy="1633299.89" reactions="0">
<decay type="IT" target="Cs135" branching_ratio="1.0"/>
</nuclide>
<nuclide name="Gd157" decay_modes="0" reactions="1" >
<reaction type="(n,gamma)" Q="0.0" target="Nothing" />
</nuclide>
<nuclide name="Gd156" decay_modes="0" reactions="1">
<reaction type="(n,gamma)" Q="0.0" target="Gd157" />
</nuclide>
<nuclide name="U234" decay_modes="0" reactions="1">
<reaction type="fission" Q="191840000."/>
<neutron_fission_yields>
<energies>2.53000e-02</energies>
<fission_yields energy="2.53000e-02">
<products>Gd157 Gd156 I135 Xe135 Xe136 Cs135</products>
<data>1.093250e-04 2.087260e-04 2.780820e-02 6.759540e-03 2.392300e-02 4.356330e-05</data>
</fission_yields>
</neutron_fission_yields>
</nuclide>
<nuclide name="U235" decay_modes="0" reactions="1">
<reaction type="fission" Q="193410000."/>
<source type="discrete" particle="photon">
<parameters>3065.349 12960.11 13197.49 16125.43 19185.05 19590.0 31600.0 34700.0 41400.0 41960.0 51220.0 54100.0 54250.0 64350.0 72700.0 75020.0 76198.0 90330.0 93795.0 96090.0 105278.0 106074.0 106608.0 106771.0 108948.0 109154.0 109160.0 109395.0 109433.0 115450.0 120350.0 136550.0 140760.0 142400.0 143760.0 150930.0 163330.0 173300.0 182610.0 185715.0 194940.0 198900.0 202110.0 205311.0 215280.0 221380.0 228780.0 233500.0 240870.0 246840.0 266450.0 275129.0 275430.0 281420.0 282920.0 289560.0 291650.0 301700.0 317100.0 343500.0 345900.0 356030.0 387820.0 410290.0 428710.0 448400.0 1.4211820389290126e-18 3.695705148646686e-18 4.752472005970374e-19 4.0825252294602915e-18 8.998127221991115e-19 1.9035285506819052e-21 5.305446177665699e-21 1.1547147563154756e-20 9.362552078233586e-21 1.86835843588509e-20 1.0610892355331398e-20 2.736290732002608e-22 4.776811520523089e-21 3.977552295559136e-21 3.432935762018982e-20 1.872510415646717e-20 2.4966805541956234e-21 1.0265454754703584e-18 1.7575878976520235e-18 2.839974130397521e-20 2.1057468800839102e-19 4.1342252420362975e-19 7.098565272539688e-21 8.20050332279016e-21 5.276915360632628e-20 1.0602918581811435e-19 4.806110066826575e-19 2.0521431485853304e-21 2.375669880669523e-21 9.362552078233586e-21 8.267152210184412e-21 3.745020831293435e-21 6.865871524037964e-20 1.5604253463722645e-21 3.420452359248004e-18 2.4966805541956232e-20 1.5853921519142206e-18 1.8725104156467174e-21 1.0610892355331397e-19 1.7851265962498703e-17 1.966135936429053e-19 1.3107572909527022e-20 3.370518748164091e-19 1.5635461970650089e-18 9.050467008959134e-21 3.745020831293434e-20 2.18459548492117e-21 9.050467008959134e-21 2.3406380195583967e-20 1.6540508671546e-20 1.8725104156467174e-21 1.622842360227155e-20 2.18459548492117e-21 1.8725104156467174e-21 1.8725104156467174e-21 2.18459548492117e-21 1.2483402770978116e-20 1.5604253463722645e-21 3.120850692744529e-22 9.362552078233587e-22 1.2483402770978116e-20 1.5604253463722645e-21 1.2483402770978116e-20 9.362552078233587e-22 3.120850692744529e-22 3.120850692744529e-22</parameters>
</source>
<neutron_fission_yields>
<energies>2.53000e-02</energies>
<fission_yields energy="2.53000e-02">
<products>Gd157 Gd156 I135 Xe135 Xe136 Cs135</products>
<data>6.142710e-5 1.483250e-04 0.0292737 0.002566345 0.0219242 4.9097e-6</data>
</fission_yields>
</neutron_fission_yields>
</nuclide>
<nuclide name="U238" decay_modes="0" reactions="1">
<reaction type="fission" Q="197790000."/>
<source type="discrete" particle="photon">
<parameters>3061.32 12959.8 13440.07 16150.05 19148.83 49550.0 90330.0 93795.0 105278.0 106074.0 106608.0 106771.0 108948.0 109154.0 109395.0 109433.0 113500.0 5.3130501476983077e-20 1.4936931167066328e-19 1.943509007980312e-20 1.8224649880365157e-19 4.0452400597373603e-20 3.146222282132249e-21 3.300026341588747e-23 5.444173877278485e-23 6.3486292118621375e-24 1.2400176384733938e-23 2.124537722121886e-25 2.4542156071495764e-25 1.5792541400719878e-24 3.1731991718126067e-24 6.141568457919309e-26 7.109808533300877e-26 5.014291762148272e-22</parameters>
</source>
<neutron_fission_yields>
<energies>2.53000e-02</energies>
<fission_yields energy="2.53000e-02">
<products>Gd157 Gd156 I135 Xe135 Xe136 Cs135</products>
<data>4.141120e-04 7.605360e-04 0.0135457 0.00026864 0.0024432 3.7100E-07</data>
</fission_yields>
</neutron_fission_yields>
</nuclide>
</depletion_chain>

0
tests/regression_tests/deplete_decay_only/__init__.py Normal file
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103
tests/regression_tests/deplete_decay_only/test.py Normal file
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@ -0,0 +1,103 @@
""" Transport-free depletion test suite """
from pathlib import Path
import shutil
import numpy as np
import pytest
import openmc
import openmc.deplete
from openmc.deplete import CoupledOperator, IndependentOperator, MicroXS
@pytest.fixture(scope="module")
def model():
fuel = openmc.Material(name="uo2")
fuel.add_element("U", 1, percent_type="ao", enrichment=4.25)
fuel.add_element("O", 2)
fuel.add_nuclide("Xe135_m1", 1)
fuel.add_nuclide("Cs135_m1", 1)
fuel.set_density("g/cc", 10.4)
clad = openmc.Material(name="clad")
clad.add_element("Zr", 1)
clad.set_density("g/cc", 6)
water = openmc.Material(name="water")
water.add_element("O", 1)
water.add_element("H", 2)
water.set_density("g/cc", 1.0)
water.add_s_alpha_beta("c_H_in_H2O")
radii = [0.42, 0.45]
fuel.volume = np.pi * radii[0] ** 2
materials = openmc.Materials([fuel, clad, water])
pin_surfaces = [openmc.ZCylinder(r=r) for r in radii]
pin_univ = openmc.model.pin(pin_surfaces, materials)
bound_box = openmc.rectangular_prism(1.24, 1.24, boundary_type="reflective")
root_cell = openmc.Cell(fill=pin_univ, region=bound_box)
geometry = openmc.Geometry([root_cell])
settings = openmc.Settings()
settings.particles = 1000
settings.inactive = 5
settings.batches = 10
return openmc.Model(geometry, materials, settings)
@pytest.fixture(scope="module")
def micro_xs():
micro_xs_file = Path(__file__).parents[2] / 'micro_xs_simple.csv'
return MicroXS.from_csv(micro_xs_file)
@pytest.fixture(scope="module")
def chain_file():
return Path(__file__).parents[2] / 'chain_simple_decay.xml'
@pytest.mark.parametrize("operator_type", ["coupled", "independent"])
def test_decay_only(run_in_tmpdir, operator_type, model, micro_xs, chain_file):
"""Transport free system test suite.
"""
# Create operator
if operator_type == "coupled":
op = CoupledOperator(model, chain_file=chain_file)
else:
op = IndependentOperator(openmc.Materials([model.materials[0]]),
[1e15],
[micro_xs],
chain_file)
# Power and timesteps
dt = [917.4, 2262.6] # one Xe135_m1 half life and one Cs135_m1 half life
# Perform simulation using the predictor algorithm
openmc.deplete.PredictorIntegrator(op,
dt,
power=0.0,
timestep_units='s').integrate()
# Get path to test and reference results
path_test = op.output_dir / 'depletion_results.h5'
# Load the reference/test results
res_test = openmc.deplete.Results(path_test)
_, xe135m1_atoms = res_test.get_atoms('1', 'Xe135_m1')
_, xe135_atoms = res_test.get_atoms('1', 'Xe135')
_, cs135m1_atoms = res_test.get_atoms('1', 'Cs135_m1')
_, cs135_atoms = res_test.get_atoms('1', 'Cs135')
tol = 1.0e-14
assert xe135m1_atoms[0] == pytest.approx(xe135m1_atoms[1] * 2, rel=tol)
# WARNING: this is generally not true as Xe135_m1 has two
# decay modes, and Xe135 will also decay, but we've modified the depletion chain so
# that Xe135_m1 only decays to Xe135, and that Xe135 has has no decay modes
assert xe135_atoms[1] == pytest.approx(xe135m1_atoms[1], rel=tol)
assert cs135m1_atoms[0] == pytest.approx(cs135m1_atoms[2] * 2, rel=tol)
assert cs135_atoms[2] == pytest.approx(cs135m1_atoms[2], rel=tol)

12
tests/unit_tests/test_material.py
View file

@ -74,6 +74,18 @@ def test_add_components():
with pytest.raises(ValueError):
m.add_components({'H1': 1.0}, percent_type = 'oa')
def test_nuclides_to_ignore(run_in_tmpdir):
"""Test nuclides_to_ignore when exporting a material to XML"""
m = openmc.Material()
m.add_nuclide('U235', 1.0)
m.add_nuclide('H1', 1.0)
m.add_nuclide('O16', 1.0)
mats = openmc.Materials([m])
mats.export_to_xml(nuclides_to_ignore=['H1'])
test_mats = openmc.Materials.from_xml()
assert 'H1' not in test_mats[0].get_nuclides()
def test_remove_nuclide():
"""Test removing nuclides."""