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address most of paulromano's comments
This commit is contained in:
yardasol
parent
61ad9c977c
commit
06f34b9f23
5 changed files with 159 additions and 189 deletions
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@ -113,7 +113,7 @@ should be, including indirect components. Some examples are provided below::
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fission_q = {"U235": 202e+6} # energy in eV
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# create a Model object
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model = openmc.model.Model(geometry, settings)
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model = openmc.Model(geometry, settings)
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# create a modified chain and write it to a new file
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chain = openmc.deplete.Chain.from_xml("chain.xml", fission_q)
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@ -180,7 +180,7 @@ across all material instances.
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number of tallies and material definitions.
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Transport-independent depletion
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-------------------------------
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===============================
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.. note::
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@ -188,10 +188,10 @@ Transport-independent depletion
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possible and likely in the near future.
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OpenMC supports running depletion calculations independent of the OpenMC
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transport solver using the :class:`FluxDepletionOperator` class. This class
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transport solver using the :class:`~openmc.deplete.FluxDepletionOperator` class. This class
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has two ways to initalize it; the default constructor accepts an
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:class:`openmc.Materials` object, a flux spectra, and one-group microscopic
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cross sections as a :class:`pandas.Dataframe`, while the `from_nuclides`
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cross sections as a :class:`pandas.DataFrame`, while the `from_nuclides`
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method accepts a volume and dictionary of nuclide concentrations in place of
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the :class:`openmc.Materials` object in addition to the other parameters.
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The class includes helper functions to construct the dataframe from a csv file
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@ -1,6 +1,6 @@
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"""Pure depletion operator
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This module implements a pure depletion operator that uses user provided fluxes
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This module implements a pure depletion operator that uses user- provided fluxes
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and one-group cross sections.
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"""
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@ -17,25 +17,23 @@ from uncertainties import ufloat
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import openmc
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from openmc.checkvalue import check_type, check_value, check_iterable_type
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from openmc.mpi import comm
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from .abc import TransportOperator, ReactionRateHelper, OperatorResult
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from .atom_number import AtomNumber
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from .abc import ReactionRateHelper, OperatorResult
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from .chain import REACTIONS
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from .openmc_operator import OpenMCOperator
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from .reaction_rates import ReactionRates
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from .helpers import ConstantFissionYieldHelper, SourceRateHelper
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valid_rxns = list(REACTIONS)
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valid_rxns.append('fission')
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_valid_rxns = list(REACTIONS)
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_valid_rxns.append('fission')
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class FluxDepletionOperator(OpenMCOperator):
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"""Depletion operator that uses a user-provided flux spectrum and one-group
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"""Depletion operator that uses a user-provided flux and one-group
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cross sections to calculate reaction rates.
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Instances of this class can be used to perform depletion using one group
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Instances of this class can be used to perform depletion using one-group
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cross sections and constant flux. Normally, a user needn't call methods of
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this class directly. Instead, an instance of this class is passed to an
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integrator class, such as :class:`openmc.deplete.CECMIntegrator`
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integrator class, such as :class:`openmc.deplete.CECMIntegrator`.
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Parameters
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----------
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@ -44,8 +42,8 @@ class FluxDepletionOperator(OpenMCOperator):
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micro_xs : pandas.DataFrame
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DataFrame with nuclides names as index and microscopic cross section
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data in the columns. Cross section units are [cm^-2].
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flux_spectra : float
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Flux spectrum [n cm^-2 s^-1]
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flux : float
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Neutron flux [n cm^-2 s^-1]
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chain_file : str
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Path to the depletion chain XML file.
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keff : 2-tuple of float, optional
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@ -63,16 +61,27 @@ class FluxDepletionOperator(OpenMCOperator):
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Depth of the search when reducing the depletion chain. Only used
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if ``reduce_chain`` evaluates to true. The default value of
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``None`` implies no limit on the depth.
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fission_yield_opts : dict of str to option, optional
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Optional arguments to pass to the `FissionYieldHelper`. Will be
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passed directly on to the helper. Passing a value of None will use
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the defaults for the associated helper.
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Attributes
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----------
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materials : openmc.Materials
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All materials present in the model
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cross_sections : pandas.DataFrame
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Object containing one-group cross-sections.
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dilute_initial : float
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Initial atom density [atoms/cm^3] to add for nuclides that
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are zero in initial condition to ensure they exist in the decay
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chain. Only done for nuclides with reaction rates.
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output_dir : pathlib.Path
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Path to output directory to save results.
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round_number : bool
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Whether or not to round output to OpenMC to 8 digits.
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Useful in testing, as OpenMC is incredibly sensitive to exact values.
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prev_res : Results or None
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Results from a previous depletion calculation. ``None`` if no
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results are to be used.
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number : openmc.deplete.AtomNumber
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Total number of atoms in simulation.
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nuclides_with_data : set of str
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@ -81,14 +90,55 @@ class FluxDepletionOperator(OpenMCOperator):
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The depletion chain information necessary to form matrices and tallies.
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reaction_rates : openmc.deplete.ReactionRates
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Reaction rates from the last operator step.
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burnable_mats : list of str
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All burnable material IDs
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heavy_metal : float
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Initial heavy metal inventory [g]
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local_mats : list of str
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All burnable material IDs being managed by a single process
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prev_res : Results or None
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Results from a previous depletion calculation. ``None`` if no
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results are to be used.
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"""
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"""
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def __init__(self,
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materials,
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micro_xs,
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flux,
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chain_file,
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keff=None,
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fission_q=None,
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prev_results=None,
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reduce_chain=False,
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reduce_chain_level=None,
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fission_yield_opts=None):
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# Validate micro-xs parameters
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check_type('materials', materials, openmc.Materials)
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check_type('micro_xs', micro_xs, pd.DataFrame)
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if keff is not None:
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check_type('keff', keff, tuple, float)
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keff = ufloat(*keff)
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self._keff = keff
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self.flux = flux
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helper_kwargs = dict()
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helper_kwargs = {'fission_yield_opts': fission_yield_opts}
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super().__init__(
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materials,
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micro_xs,
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chain_file,
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prev_results,
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fission_q=fission_q,
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helper_kwargs=helper_kwargs,
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reduce_chain=reduce_chain,
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reduce_chain_level=reduce_chain_level)
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@classmethod
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def from_nuclides(cls, volume, nuclides, micro_xs,
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flux_spectra,
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flux,
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chain_file,
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keff=None,
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fission_q=None,
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@ -107,8 +157,8 @@ class FluxDepletionOperator(OpenMCOperator):
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micro_xs : pandas.DataFrame
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DataFrame with nuclides names as index and microscopic cross section
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data in the columns. Cross section units are [cm^-2].
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flux_spectra : float
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Flux spectrum [n cm^-2 s^-1]
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flux : float
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Neutron flux [n cm^-2 s^-1]
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chain_file : str
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Path to the depletion chain XML file.
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keff : 2-tuple of float, optional
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@ -126,12 +176,17 @@ class FluxDepletionOperator(OpenMCOperator):
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Depth of the search when reducing the depletion chain. Only used
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if ``reduce_chain`` evaluates to true. The default value of
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``None`` implies no limit on the depth.
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fission_yield_opts : dict of str to option, optional
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Optional arguments to pass to the `FissionYieldHelper`. Will be
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passed directly on to the helper. Passing a value of None will use
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the defaults for the associated helper.
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"""
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check_type('nuclides', nuclides, dict, str)
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materials = cls._consolidate_nuclides_to_material(nuclides, volume)
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return cls(materials,
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micro_xs,
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flux_spectra,
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flux,
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chain_file,
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keff,
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fission_q,
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@ -140,42 +195,6 @@ class FluxDepletionOperator(OpenMCOperator):
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reduce_chain_level,
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fission_yield_opts)
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def __init__(self,
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materials,
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micro_xs,
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flux_spectra,
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chain_file,
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keff=None,
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fission_q=None,
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prev_results=None,
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reduce_chain=False,
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reduce_chain_level=None,
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fission_yield_opts=None):
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# Validate micro-xs parameters
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check_type('materials', materials, openmc.Materials)
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check_type('micro_xs', micro_xs, pd.DataFrame)
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if keff is not None:
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check_type('keff', keff, tuple, float)
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keff = ufloat(keff)
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self._keff = keff
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self.flux_spectra = flux_spectra
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diff_burnable_mats = False
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helper_kwargs = dict()
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helper_kwargs['fission_yield_opts'] = fission_yield_opts
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super().__init__(
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materials,
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micro_xs,
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chain_file,
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prev_results,
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diff_burnable_mats,
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fission_q,
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0.0,
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helper_kwargs,
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reduce_chain,
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reduce_chain_level)
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@staticmethod
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def _consolidate_nuclides_to_material(nuclides, volume):
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@ -185,27 +204,18 @@ class FluxDepletionOperator(OpenMCOperator):
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openmc.reset_auto_ids()
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mat = openmc.Material()
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for nuc, conc in nuclides.items():
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mat.add_nuclide(nuc, conc / 1e24) # convert to at/b-cm
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mat.add_nuclide(nuc, conc * 1e-24) # convert to at/b-cm
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mat.volume = volume
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mat.depleteable = True
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mat.depletable = True
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return openmc.Materials([mat])
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def _differentiate_burnable_mats():
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"""Assign distribmats for each burnable material"""
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pass
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def _load_previous_results():
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"""Load in results from a previous depletion calculation."""
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pass
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def _get_nuclides_with_data(self, cross_sections):
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"""Finds nuclides with cross section data
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"""
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"""Finds nuclides with cross section data"""
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return set(cross_sections.index)
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class FluxTimesXSHelper(ReactionRateHelper):
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class _FluxDepletionRateHelper(ReactionRateHelper):
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"""Class for generating one-group reaction rates with flux and
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one-group cross sections.
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@ -216,13 +226,14 @@ class FluxDepletionOperator(OpenMCOperator):
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Parameters
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----------
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outer : openmc.deplete.FluxDepletionOperator
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Reference to the object encapsulate FluxTimesXSHelper.
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We pass this so we don't have to duplicate the ``number`` object.
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n_nucs : int
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Number of burnable nuclides tracked by :class:`openmc.deplete.Operator`
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n_react : int
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Number of reactions tracked by :class:`openmc.deplete.Operator`
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op : openmc.deplete.FluxDepletionOperator
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Reference to the object encapsulate _FluxDepletionRateHelper.
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We pass this so we don't have to duplicate the ``number`` object.
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Attributes
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----------
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@ -233,14 +244,17 @@ class FluxDepletionOperator(OpenMCOperator):
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"""
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def __init__(self, n_nuc, n_react, outer):
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def __init__(self, n_nuc, n_react, op, nuc_ind_map. rxn_ind_map)
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super().__init__(n_nuc, n_react)
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self.outer = outer
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self.nuc_ind_map = None
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self.rxn_ind_map = None
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self._op = op
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rates = self.reaction_rates
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# Get classes to assit working with tallies
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self.nuc_ind_map = {ind: nuc for nuc, ind in rates.index_nuc.items()}
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self.rxn_ind_map = {ind: rxn for rxn, ind in rates.index_rx.items()}
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def generate_tallies(self, materials, scores):
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"""Unused in this case"""
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pass
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def get_material_rates(self, mat_id, nuc_index, react_index):
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"""Return 2D array of [nuclide, reaction] reaction rates
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@ -255,37 +269,35 @@ class FluxDepletionOperator(OpenMCOperator):
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Ordering of reactions
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"""
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self._results_cache.fill(0.0)
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for i, (i_nuc, i_react) in enumerate(
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product(nuc_index, react_index)):
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for i_nuc, i_react in product(nuc_index, react_index):
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nuc = self.nuc_ind_map[i_nuc]
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rxn = self.rxn_ind_map[i_react]
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density = self.outer.number.get_atom_density(mat_id, nuc)
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density = self._op.number.get_atom_density(mat_id, nuc)
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self._results_cache[i_nuc,
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i_react] = self.outer.cross_sections[rxn][nuc] * density
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i_react] = self._op.cross_sections[rxn][nuc] * density
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return self._results_cache
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def _get_helper_classes(self, helper_kwargs):
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"""Get helper classes for calculating reation rates and fission yields"""
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"""Get helper classes for calculating reation rates and fission yields
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fission_yield_opts = helper_kwargs['fission_yield_opts']
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Parameters
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----------
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helper_kwargs : dict
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Keyword arguments for helper classes
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rates = self.reaction_rates
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# Get classes to assit working with tallies
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nuc_ind_map = {ind: nuc for nuc, ind in rates.index_nuc.items()}
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rxn_ind_map = {ind: rxn for rxn, ind in rates.index_rx.items()}
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self._rate_helper = self.FluxTimesXSHelper(
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"""
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fission_yield_opts = helper_kwargs.get('fission_yield_opts', {})
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self._rate_helper = self._FluxDepletionRateHelper(
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self.reaction_rates.n_nuc, self.reaction_rates.n_react, self)
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self._rate_helper.nuc_ind_map = nuc_ind_map
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self._rate_helper.rxn_ind_map = rxn_ind_map
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self._normalization_helper = SourceRateHelper()
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# Select and create fission yield helper
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fission_helper = ConstantFissionYieldHelper
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fission_yield_opts = (
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{} if fission_yield_opts is None else fission_yield_opts)
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self._yield_helper = fission_helper.from_operator(
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self, **fission_yield_opts)
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@ -320,8 +332,8 @@ class FluxDepletionOperator(OpenMCOperator):
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self._update_materials_and_nuclides(vec)
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# Use the flux spectra as a "source rate"
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rates = self._calculate_reaction_rates(self.flux_spectra)
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# Use the flux as a "source rate"
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rates = self._calculate_reaction_rates(self.flux)
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keff = self._keff
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op_result = OperatorResult(keff, rates)
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@ -356,36 +368,14 @@ class FluxDepletionOperator(OpenMCOperator):
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# negative. CRAM does not guarantee positive
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# values.
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if val < -1.0e-21:
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print(
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"WARNING: nuclide ",
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nuc,
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" in material ",
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mat,
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" is negative (density = ",
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val,
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" at/barn-cm)")
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print(f'WARNING: nuclide {nuc} in material'
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f'{mat} is negative (density = {val}'
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' at/barn-cm)')
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number_i[mat, nuc] = 0.0
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# TODO Update densities on the Python side, otherwise the
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# summary.h5 file contains densities at the first time step
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def write_bos_data(self, step):
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"""Document beginning of step data for a given step
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Called at the beginning of a depletion step and at
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the final point in the simulation.
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Parameters
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----------
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step : int
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Current depletion step including restarts
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"""
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# Since we aren't running a transport simulation, we simply pass
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pass
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@staticmethod
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def create_micro_xs_from_data_array(
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nuclides, reactions, data, units='barn'):
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def create_micro_xs_from_data_array(nuclides, reactions, data, units='barn'):
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"""
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Creates a ``micro_xs`` parameter from a dictionary.
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@ -420,7 +410,7 @@ class FluxDepletionOperator(OpenMCOperator):
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# Convert to cm^2
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if units == 'barn':
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data /= 1e24
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data *= 1e-24
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return pd.DataFrame(index=nuclides, columns=reactions, data=data)
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@ -450,7 +440,7 @@ class FluxDepletionOperator(OpenMCOperator):
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micro_xs.to_numpy())
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if units == 'barn':
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micro_xs /= 1e24
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micro_xs *= 1e-24
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return micro_xs
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@ -461,4 +451,4 @@ class FluxDepletionOperator(OpenMCOperator):
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check_iterable_type('reactions', reactions, str)
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check_type('data', data, np.ndarray, expected_iter_type=float)
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for reaction in reactions:
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check_value('reactions', reaction, valid_rxns)
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check_value('reactions', reaction, _valid_rxns)
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|
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@ -66,16 +66,14 @@ class OpenMCOperator(TransportOperator):
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diff_burnable_mats : bool, optional
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Whether to differentiate burnable materials with multiple instances.
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Volumes are divided equally from the original material volume.
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Default: False.
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fission_q : dict, optional
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Dictionary of nuclides and their fission Q values [eV].
|
||||
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 decay chain.
|
||||
Only done for nuclides with reaction rates.
|
||||
Defaults to 1.0e3.
|
||||
helper_kwargs : dict
|
||||
Arguments for helper classes
|
||||
Keyword arguments for helper classes
|
||||
reduce_chain : bool, optional
|
||||
If True, use :meth:`openmc.deplete.Chain.reduce` to reduce the
|
||||
depletion chain up to ``reduce_chain_level``. Default is False.
|
||||
|
|
@ -89,6 +87,10 @@ class OpenMCOperator(TransportOperator):
|
|||
----------
|
||||
materials : openmc.Materials
|
||||
All materials present in the model
|
||||
cross_sections : str or pandas.DataFrame
|
||||
Path to continuous energy cross section library, or object
|
||||
containing one-group cross-sections.
|
||||
|
||||
dilute_initial : float
|
||||
Initial atom density [atoms/cm^3] to add for nuclides that
|
||||
are zero in initial condition to ensure they exist in the decay
|
||||
|
|
@ -182,9 +184,9 @@ class OpenMCOperator(TransportOperator):
|
|||
|
||||
self._get_helper_classes(helper_kwargs)
|
||||
|
||||
@abstractmethod
|
||||
def _differentiate_burnable_mats(self):
|
||||
"""Assign distribmats for each burnable material"""
|
||||
pass
|
||||
|
||||
def _get_burnable_mats(self):
|
||||
"""Determine depletable materials, volumes, and nuclides
|
||||
|
|
@ -235,9 +237,9 @@ class OpenMCOperator(TransportOperator):
|
|||
|
||||
return burnable_mats, volume, nuclides
|
||||
|
||||
@abstractmethod
|
||||
def _load_previous_results(self):
|
||||
"""Load results from a previous depletion simulation"""
|
||||
pass
|
||||
|
||||
@abstractmethod
|
||||
def _get_nuclides_with_data(self, cross_sections):
|
||||
|
|
@ -271,9 +273,7 @@ class OpenMCOperator(TransportOperator):
|
|||
Results from a previous depletion calculation
|
||||
|
||||
"""
|
||||
self.number = AtomNumber(
|
||||
local_mats, all_nuclides, volume, len(
|
||||
self.chain))
|
||||
self.number = AtomNumber(local_mats, all_nuclides, volume, len(self.chain))
|
||||
|
||||
if self.dilute_initial != 0.0:
|
||||
for nuc in self._burnable_nucs:
|
||||
|
|
@ -349,20 +349,8 @@ class OpenMCOperator(TransportOperator):
|
|||
|
||||
Parameters
|
||||
----------
|
||||
reaction_rate_mode : str
|
||||
Indicates the subclass of :class:`ReactionRateHelper` to
|
||||
instantiate.
|
||||
normalization_mode : str
|
||||
Indicates the subclass of :class:`NormalizationHelper` to
|
||||
instantiate.
|
||||
fission_yield_mode : str
|
||||
Indicates the subclass of :class:`FissionYieldHelper` to instatiate.
|
||||
reaction_rate_opts : dict
|
||||
Keyword arguments that are passed to the :class:`ReactionRateHelper`
|
||||
subclass.
|
||||
fission_yield_opts : dict
|
||||
Keyword arguments that are passed to the :class:`FissionYieldHelper`
|
||||
subclass.
|
||||
helper_kwargs : dict
|
||||
Keyword arguments for helper classes
|
||||
|
||||
"""
|
||||
|
||||
|
|
@ -420,6 +408,20 @@ class OpenMCOperator(TransportOperator):
|
|||
def _update_materials(self):
|
||||
"""Updates material compositions in OpenMC on all processes."""
|
||||
|
||||
def write_bos_data(self, step):
|
||||
"""Document beginning of step data for a given step
|
||||
|
||||
Called at the beginning of a depletion step and at
|
||||
the final point in the simulation.
|
||||
|
||||
Parameters
|
||||
----------
|
||||
step : int
|
||||
Current depletion step including restarts
|
||||
"""
|
||||
# Since we aren't running a transport simulation, we simply pass
|
||||
pass
|
||||
|
||||
def _get_reaction_nuclides(self):
|
||||
"""Determine nuclides that should be tallied for reaction rates.
|
||||
|
||||
|
|
|
|||
|
|
@ -83,7 +83,6 @@ class Operator(OpenMCOperator):
|
|||
diff_burnable_mats : bool, optional
|
||||
Whether to differentiate burnable materials with multiple instances.
|
||||
Volumes are divided equally from the original material volume.
|
||||
Default: False.
|
||||
normalization_mode : {"energy-deposition", "fission-q", "source-rate"}
|
||||
Indicate how tally results should be normalized. ``"energy-deposition"``
|
||||
computes the total energy deposited in the system and uses the ratio of
|
||||
|
|
@ -99,7 +98,6 @@ class Operator(OpenMCOperator):
|
|||
Initial atom density [atoms/cm^3] to add for nuclides that are zero
|
||||
in initial condition to ensure they exist in the decay chain.
|
||||
Only done for nuclides with reaction rates.
|
||||
Defaults to 1.0e3.
|
||||
fission_yield_mode : {"constant", "cutoff", "average"}
|
||||
Key indicating what fission product yield scheme to use. The
|
||||
key determines what fission energy helper is used:
|
||||
|
|
@ -228,11 +226,13 @@ class Operator(OpenMCOperator):
|
|||
self.cleanup_when_done = True
|
||||
|
||||
helper_kwargs = dict()
|
||||
helper_kwargs['reaction_rate_mode'] = reaction_rate_mode
|
||||
helper_kwargs['normalization_mode'] = normalization_mode
|
||||
helper_kwargs['fission_yield_mode'] = fission_yield_mode
|
||||
helper_kwargs['reaction_rate_opts'] = reaction_rate_opts
|
||||
helper_kwargs['fission_yield_opts'] = fission_yield_opts
|
||||
helper_kwargs = {
|
||||
'reaction_rate_mode': reaction_rate_mode,
|
||||
'normalization_mode': normalization_mode,
|
||||
'fission_yield_mode': fission_yield_mode,
|
||||
'reaction_rate_opts': reaction_rate_opts,
|
||||
'fission_yield_opts': fission_yield_opts
|
||||
}
|
||||
|
||||
super().__init__(
|
||||
model.materials,
|
||||
|
|
@ -322,37 +322,21 @@ class Operator(OpenMCOperator):
|
|||
|
||||
Parameters
|
||||
----------
|
||||
**kwargs : ...
|
||||
reaction_rate_mode : str
|
||||
Indicates the subclass of :class:`ReactionRateHelper` to
|
||||
instantiate.
|
||||
normalization_mode : str
|
||||
Indicates the subclass of :class:`NormalizationHelper` to
|
||||
instatiate.
|
||||
fission_yield_mode : str
|
||||
Indicates the subclass of :class:`FissionYieldHelper` to instantiate.
|
||||
reaction_rate_opts : dict
|
||||
Keyword arguments that are passed to the :class:`ReactionRateHelper`
|
||||
subclass.
|
||||
fission_yield_opts : dict
|
||||
Keyword arguments that are passed to the :class:`FissionYieldHelper`
|
||||
subclass.
|
||||
helper_kwargs : dict
|
||||
Keyword arguments for helper classes
|
||||
|
||||
"""
|
||||
reaction_rate_mode = helper_kwargs['reaction_rate_mode']
|
||||
normalization_mode = helper_kwargs['normalization_mode']
|
||||
fission_yield_mode = helper_kwargs['fission_yield_mode']
|
||||
reaction_rate_opts = helper_kwargs['reaction_rate_opts']
|
||||
fission_yield_opts = helper_kwargs['fission_yield_opts']
|
||||
reaction_rate_opts = helper_kwargs.get('reaction_rate_opts', {})
|
||||
fission_yield_opts = helper_kwargs.get('fission_yield_opts', {})
|
||||
|
||||
# Get classes to assist working with tallies
|
||||
if reaction_rate_mode == "direct":
|
||||
self._rate_helper = DirectReactionRateHelper(
|
||||
self.reaction_rates.n_nuc, self.reaction_rates.n_react)
|
||||
elif reaction_rate_mode == "flux":
|
||||
if reaction_rate_opts is None:
|
||||
reaction_rate_opts = {}
|
||||
|
||||
# Ensure energy group boundaries were specified
|
||||
if 'energies' not in reaction_rate_opts:
|
||||
raise ValueError(
|
||||
|
|
@ -378,8 +362,6 @@ class Operator(OpenMCOperator):
|
|||
|
||||
# Select and create fission yield helper
|
||||
fission_helper = self._fission_helpers[fission_yield_mode]
|
||||
fission_yield_opts = (
|
||||
{} if fission_yield_opts is None else fission_yield_opts)
|
||||
self._yield_helper = fission_helper.from_operator(
|
||||
self, **fission_yield_opts)
|
||||
|
||||
|
|
@ -405,8 +387,7 @@ class Operator(OpenMCOperator):
|
|||
openmc.lib.init(intracomm=comm)
|
||||
|
||||
# Generate tallies in memory
|
||||
materials = [openmc.lib.materials[int(i)]
|
||||
for i in self.burnable_mats]
|
||||
materials = [openmc.lib.materials[int(i)] for i in self.burnable_mats]
|
||||
|
||||
return super().initial_condition(materials)
|
||||
|
||||
|
|
@ -500,15 +481,12 @@ class Operator(OpenMCOperator):
|
|||
# negative. CRAM does not guarantee positive
|
||||
# values.
|
||||
if val < -1.0e-21:
|
||||
print(
|
||||
"WARNING: nuclide ",
|
||||
nuc,
|
||||
" in material ",
|
||||
mat,
|
||||
" is negative (density = ",
|
||||
val,
|
||||
" at/barn-cm)")
|
||||
number_i[mat, nuc] = 0.0
|
||||
print(f'WARNING: nuclide {nuc} in material'
|
||||
f'{mat} is negative (density = {val}'
|
||||
|
||||
' at/barn-cm)')
|
||||
|
||||
number_i[mat, nuc] = 0.0
|
||||
|
||||
# Update densities on C API side
|
||||
mat_internal = openmc.lib.materials[int(mat)]
|
||||
|
|
|
|||
|
|
@ -11,7 +11,7 @@ from openmc.deplete.flux_operator import FluxDepletionOperator
|
|||
import pandas as pd
|
||||
import numpy as np
|
||||
|
||||
FLUX_SPECTRA = 5e16
|
||||
FLUX = 5e16
|
||||
CHAIN_PATH = Path(__file__).parents[1] / "chain_simple.xml"
|
||||
ONE_GROUP_XS = Path(__file__).parents[1] / "micro_xs_simple.csv"
|
||||
|
||||
|
|
|
|||
Loading…
Add table
Add a link
Reference in a new issue