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Adding material depletion function (#3420)
Co-authored-by: Jon Shimwell <jon@proximafusion.com> Co-authored-by: Micah Gale <mgale@fastmail.com> Co-authored-by: Paul Romano <paul.k.romano@gmail.com>
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Jonathan Shimwell
GitHub
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659e43af7d
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8be65513b7
4 changed files with 281 additions and 1 deletions
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@ -273,7 +273,7 @@ class IndependentOperator(OpenMCOperator):
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Returns
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-------
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nuclides : set of str
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Set of nuclide names that have cross secton data
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Set of nuclide names that have cross section data
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"""
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return set(cross_sections[0].nuclides)
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@ -6,6 +6,8 @@ from numbers import Real
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from pathlib import Path
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import re
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import sys
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import tempfile
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from typing import Sequence, Dict
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import warnings
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import lxml.etree as ET
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@ -1717,6 +1719,67 @@ class Material(IDManagerMixin):
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return mat
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def deplete(
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self,
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multigroup_flux: Sequence[float],
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energy_group_structure: Sequence[float] | str,
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timesteps: Sequence[float] | Sequence[tuple[float, str]],
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source_rates: float | Sequence[float],
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timestep_units: str = 's',
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chain_file: cv.PathLike | "openmc.deplete.Chain" | None = None,
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reactions: Sequence[str] | None = None,
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) -> list[openmc.Material]:
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"""Depletes that material, evolving the nuclide densities
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.. versionadded:: 0.15.3
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Parameters
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----------
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multigroup_flux: Sequence[float]
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Energy-dependent multigroup flux values, where each sublist corresponds
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to a specific material. Will be normalized so that it sums to 1.
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energy_group_structure : Sequence[float] | str
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Energy group boundaries in [eV] or the name of the group structure.
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timesteps : iterable of float or iterable of tuple
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Array of timesteps. Note that values are not cumulative. The units are
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specified by the `timestep_units` argument when `timesteps` is an
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iterable of float. Alternatively, units can be specified for each step
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by passing an iterable of (value, unit) tuples.
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source_rates : float or iterable of float, optional
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Source rate in [neutron/sec] or neutron flux in [neutron/s-cm^2] for
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each interval in :attr:`timesteps`
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timestep_units : {'s', 'min', 'h', 'd', 'a', 'MWd/kg'}
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Units for values specified in the `timesteps` argument. 's' means
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seconds, 'min' means minutes, 'h' means hours, 'a' means Julian years
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and 'MWd/kg' indicates that the values are given in burnup (MW-d of
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energy deposited per kilogram of initial heavy metal).
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chain_file : PathLike or Chain
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Path to the depletion chain XML file or instance of openmc.deplete.Chain.
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Defaults to ``openmc.config['chain_file']``.
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reactions : list of str, optional
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Reactions to get cross sections for. If not specified, all neutron
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reactions listed in the depletion chain file are used.
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Returns
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-------
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list of openmc.Material, one for each timestep
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"""
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materials = openmc.Materials([self])
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depleted_materials_dict = materials.deplete(
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multigroup_fluxes=[multigroup_flux],
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energy_group_structures=[energy_group_structure],
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timesteps=timesteps,
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source_rates=source_rates,
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timestep_units=timestep_units,
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chain_file=chain_file,
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reactions=reactions,
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)
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return depleted_materials_dict[self.id]
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def mean_free_path(self, energy: float) -> float:
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"""Calculate the mean free path of neutrons in the material at a given
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@ -1947,3 +2010,114 @@ class Materials(cv.CheckedList):
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root = tree.getroot()
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return cls.from_xml_element(root)
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def deplete(
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self,
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multigroup_fluxes: Sequence[Sequence[float]],
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energy_group_structures: Sequence[Sequence[float] | str],
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timesteps: Sequence[float] | Sequence[tuple[float, str]],
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source_rates: float | Sequence[float],
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timestep_units: str = 's',
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chain_file: cv.PathLike | "openmc.deplete.Chain" | None = None,
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reactions: Sequence[str] | None = None,
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) -> Dict[int, list[openmc.Material]]:
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"""Depletes that material, evolving the nuclide densities
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.. versionadded:: 0.15.3
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Parameters
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----------
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multigroup_fluxes: Sequence[Sequence[float]]
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Energy-dependent multigroup flux values, where each sublist corresponds
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to a specific material. Will be normalized so that it sums to 1.
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energy_group_structures': Sequence[Sequence[float] | str]
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Energy group boundaries in [eV] or the name of the group structure.
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timesteps : iterable of float or iterable of tuple
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Array of timesteps. Note that values are not cumulative. The units are
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specified by the `timestep_units` argument when `timesteps` is an
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iterable of float. Alternatively, units can be specified for each step
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by passing an iterable of (value, unit) tuples.
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source_rates : float or iterable of float, optional
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Source rate in [neutron/sec] or neutron flux in [neutron/s-cm^2] for
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each interval in :attr:`timesteps`
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timestep_units : {'s', 'min', 'h', 'd', 'a', 'MWd/kg'}
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Units for values specified in the `timesteps` argument. 's' means
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seconds, 'min' means minutes, 'h' means hours, 'a' means Julian years
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and 'MWd/kg' indicates that the values are given in burnup (MW-d of
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energy deposited per kilogram of initial heavy metal).
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chain_file : PathLike or Chain
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Path to the depletion chain XML file or instance of openmc.deplete.Chain.
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Defaults to ``openmc.config['chain_file']``.
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reactions : list of str, optional
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Reactions to get cross sections for. If not specified, all neutron
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reactions listed in the depletion chain file are used.
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Returns
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-------
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list of openmc.Material, one for each timestep
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"""
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import openmc.deplete
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from .deplete.chain import _get_chain
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# setting all materials to be depletable
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for mat in self:
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mat.depletable = True
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chain = _get_chain(chain_file)
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# Create MicroXS objects for all materials
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micros = []
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fluxes = []
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with openmc.lib.TemporarySession():
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for material, flux, energy in zip(
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self, multigroup_fluxes, energy_group_structures
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):
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temperature = material.temperature or 293.6
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micro_xs = openmc.deplete.MicroXS.from_multigroup_flux(
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energies=energy,
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multigroup_flux=flux,
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chain_file=chain,
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temperature=temperature,
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reactions=reactions,
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)
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micros.append(micro_xs)
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fluxes.append(material.volume)
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# Create a single operator for all materials
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operator = openmc.deplete.IndependentOperator(
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materials=self,
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fluxes=fluxes,
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micros=micros,
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normalization_mode="source-rate",
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chain_file=chain,
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)
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integrator = openmc.deplete.PredictorIntegrator(
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operator=operator,
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timesteps=timesteps,
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source_rates=source_rates,
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timestep_units=timestep_units,
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)
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with tempfile.TemporaryDirectory() as tmpdir:
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# Run integrator
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results_path = Path(tmpdir) / "depletion_results.h5"
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integrator.integrate(path=results_path)
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# Load depletion results
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results = openmc.deplete.Results(results_path)
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# For each material, get activated composition at each timestep
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all_depleted_materials = {
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material.id: [
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result.get_material(str(material.id))
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for result in results
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]
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for material in self
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}
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return all_depleted_materials
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@ -3,8 +3,11 @@ from pathlib import Path
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import pytest
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import numpy as np
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import openmc
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from openmc.data import decay_photon_energy
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from openmc.deplete import Chain
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import openmc.examples
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import openmc.model
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import openmc.stats
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@ -712,6 +715,46 @@ def test_avoid_subnormal(run_in_tmpdir):
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assert mats[0].get_nuclide_atom_densities()['H2'] == 0.0
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def test_material_deplete():
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pristine_material = openmc.Material()
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pristine_material.add_nuclide("Ni58", 1.0)
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pristine_material.set_density("g/cm3", 7.87)
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pristine_material.depletable = True
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pristine_material.temperature = 293.6
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pristine_material.volume = 1.
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mg_flux = [0.5e11] * 42
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chain = Chain.from_xml(
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Path(__file__).parents[1] / "chain_ni.xml"
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)
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depleted_material = pristine_material.deplete(
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multigroup_flux=mg_flux,
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energy_group_structure="VITAMIN-J-42",
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timesteps=[10, 70.86],
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source_rates=[1e19, 0.0],
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timestep_units="d",
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chain_file=chain,
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)
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for material in depleted_material:
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assert isinstance(material, openmc.Material)
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assert len(material.get_nuclides()) > len(pristine_material.get_nuclides())
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Co58_mat_1_step_0 = depleted_material[0].get_nuclide_atom_densities("Co58")["Co58"]
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Co58_mat_1_step_1 = depleted_material[1].get_nuclide_atom_densities("Co58")["Co58"]
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Co58_mat_1_step_2 = depleted_material[2].get_nuclide_atom_densities("Co58")["Co58"]
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assert Co58_mat_1_step_0 == 0.0
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# Check that Co58 is produced in the first step
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assert Co58_mat_1_step_1 > 0.0
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# Check that Co58 is halved in the second step which is one halflife later
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assert np.allclose(Co58_mat_1_step_1 * 0.5, Co58_mat_1_step_2)
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def test_mean_free_path():
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mat1 = openmc.Material()
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63
tests/unit_tests/test_materials.py
Normal file
63
tests/unit_tests/test_materials.py
Normal file
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@ -0,0 +1,63 @@
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from pathlib import Path
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import openmc
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from openmc.deplete import Chain
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def test_materials_deplete():
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pristine_material_1 = openmc.Material()
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pristine_material_1.add_nuclide("Ni58", 1.)
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pristine_material_1.set_density("g/cm3", 7.87)
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pristine_material_1.depletable = True
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pristine_material_1.temperature = 293.6
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pristine_material_1.volume = 1.
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pristine_material_2 = openmc.Material()
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pristine_material_2.add_nuclide("Ni60", 1.)
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pristine_material_2.set_density("g/cm3", 7.87)
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pristine_material_2.depletable = True
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pristine_material_2.temperature = 293.6
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pristine_material_2.volume = 1.
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pristine_materials = openmc.Materials([pristine_material_1, pristine_material_2])
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mg_flux = [0.5e11] * 42
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chain = Chain.from_xml(
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Path(__file__).parents[1] / "chain_ni.xml"
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)
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depleted_material = pristine_materials.deplete(
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multigroup_fluxes=[mg_flux, mg_flux],
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energy_group_structures=["VITAMIN-J-42", "VITAMIN-J-42"],
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timesteps=[100, 100],
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source_rates=[1e19, 0.0],
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timestep_units="d",
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chain_file=chain,
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)
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assert list(depleted_material.keys()) == [pristine_material_1.id, pristine_material_2.id]
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for mat_id, materials in depleted_material.items():
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for material in materials:
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assert isinstance(material, openmc.Material)
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assert len(material.get_nuclides()) > 1
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assert mat_id == material.id
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mats = depleted_material[pristine_material_1.id]
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Co58_mat_1_step_0 = mats[0].get_nuclide_atom_densities("Co58")["Co58"]
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Co58_mat_1_step_1 = mats[1].get_nuclide_atom_densities("Co58")["Co58"]
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Co58_mat_1_step_2 = mats[2].get_nuclide_atom_densities("Co58")["Co58"]
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assert Co58_mat_1_step_0 == 0.0
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# Co58 is the main activation product of Ni58 in the first irradiation step.
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# It then decays in the second cooling step (flux = 0)
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assert Co58_mat_1_step_1 > 0.0 and Co58_mat_1_step_1 > Co58_mat_1_step_2
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Ni59_mat_1_step_0 = mats[0].get_nuclide_atom_densities("Ni59")["Ni59"]
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Ni59_mat_1_step_1 = mats[1].get_nuclide_atom_densities("Ni59")["Ni59"]
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Ni59_mat_1_step_2 = mats[2].get_nuclide_atom_densities("Ni59")["Ni59"]
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assert Ni59_mat_1_step_0 == 0.0
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# Ni59 is one of the main activation product of Ni60 in the first irradiation
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# step. It then decays in the second cooling step (flux = 0)
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assert Ni59_mat_1_step_1 > 0.0 and Ni59_mat_1_step_1 > Ni59_mat_1_step_2
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