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Adam Nelson
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1 changed files with 133 additions and 2 deletions
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@ -8,6 +8,7 @@ import h5py
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import numpy as np
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import openmc
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from openmc.data import REACTION_MT
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import openmc.checkvalue as cv
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from ..tallies import ESTIMATOR_TYPES
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from . import EnergyGroups
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@ -698,7 +699,7 @@ class MGXS:
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Parameters
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----------
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mgxs_type : {'total', 'transport', 'nu-transport', 'absorption', 'capture', 'fission', 'nu-fission', 'kappa-fission', 'scatter', 'nu-scatter', 'scatter matrix', 'nu-scatter matrix', 'multiplicity matrix', 'nu-fission matrix', 'chi', 'chi-prompt', 'inverse-velocity', 'prompt-nu-fission', 'prompt-nu-fission matrix', 'current', 'diffusion-coefficient', 'nu-diffusion-coefficient'}
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mgxs_type : {'total', 'transport', 'nu-transport', 'absorption', 'capture', 'fission', 'nu-fission', 'kappa-fission', 'scatter', 'nu-scatter', 'scatter matrix', 'nu-scatter matrix', 'multiplicity matrix', 'nu-fission matrix', 'chi', 'chi-prompt', 'inverse-velocity', 'prompt-nu-fission', 'prompt-nu-fission matrix', 'current', 'diffusion-coefficient', 'nu-diffusion-coefficient', mt}
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The type of multi-group cross section object to return
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domain : openmc.Material or openmc.Cell or openmc.Universe or openmc.RegularMesh
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The domain for spatial homogenization
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@ -782,7 +783,18 @@ class MGXS:
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mgxs = DiffusionCoefficient(domain, domain_type, energy_groups)
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elif mgxs_type == 'nu-diffusion-coefficient':
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mgxs = DiffusionCoefficient(domain, domain_type, energy_groups, nu=True)
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else:
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if mgxs_type in REACTION_MT.keys() + REACTION_MT.values():
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# Then it is a reaction not covered by the above that is
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# supported by the ArbitraryMTXS Class
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mgxs = ArbitraryMTXS(mgxs_type, domain, domain_type,
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energy_groups)
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else:
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# This is an invalid type so let the user know.
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msg = 'Unable to set "{0}" to "{1}"'.format("mgxs_type",
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mgxs_type) + \
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" as it is not an accepted MGXS type."
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raise ValueError(msg)
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mgxs.by_nuclide = by_nuclide
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mgxs.name = name
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mgxs.num_polar = num_polar
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@ -3850,6 +3862,125 @@ class ScatterXS(MGXS):
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self._valid_estimators = ['analog']
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class ArbitraryMTXS(MGXS):
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r"""A multi-group cross section for an arbitrary reaction type.
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This class can be used for both OpenMC input generation and tally data
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post-processing to compute spatially-homogenized and energy-integrated
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multi-group total cross sections for multi-group neutronics calculations. At
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a minimum, one needs to set the :attr:`TotalXS.energy_groups` and
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:attr:`TotalXS.domain` properties. Tallies for the flux and appropriate
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reaction rates over the specified domain are generated automatically via the
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:attr:`TotalXS.tallies` property, which can then be appended to a
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:class:`openmc.Tallies` instance.
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For post-processing, the :meth:`MGXS.load_from_statepoint` will pull in the
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necessary data to compute multi-group cross sections from a
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:class:`openmc.StatePoint` instance. The derived multi-group cross section
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can then be obtained from the :attr:`TotalXS.xs_tally` property.
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For a spatial domain :math:`V` and energy group :math:`[E_g,E_{g-1}]`, the
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total cross section is calculated as:
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.. math::
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\frac{\int_{r \in V} dr \int_{4\pi} d\Omega \int_{E_g}^{E_{g-1}} dE \;
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\sigma_t (r, E) \psi (r, E, \Omega)}{\int_{r \in V} dr \int_{4\pi}
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d\Omega \int_{E_g}^{E_{g-1}} dE \; \psi (r, E, \Omega)}.
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Parameters
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----------
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domain : openmc.Material or openmc.Cell or openmc.Universe or openmc.RegularMesh
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The domain for spatial homogenization
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domain_type : {'material', 'cell', 'distribcell', 'universe', 'mesh'}
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The domain type for spatial homogenization
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groups : openmc.mgxs.EnergyGroups
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The energy group structure for energy condensation
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by_nuclide : bool
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If true, computes cross sections for each nuclide in domain
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name : str, optional
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Name of the multi-group cross section. Used as a label to identify
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tallies in OpenMC 'tallies.xml' file.
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num_polar : Integral, optional
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Number of equi-width polar angle bins for angle discretization;
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defaults to one bin
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num_azimuthal : Integral, optional
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Number of equi-width azimuthal angle bins for angle discretization;
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defaults to one bin
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Attributes
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----------
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name : str, optional
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Name of the multi-group cross section
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rxn_type : str
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Reaction type (e.g., 'total', 'nu-fission', etc.)
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by_nuclide : bool
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If true, computes cross sections for each nuclide in domain
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domain : openmc.Material or openmc.Cell or openmc.Universe or openmc.RegularMesh
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Domain for spatial homogenization
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domain_type : {'material', 'cell', 'distribcell', 'universe', 'mesh'}
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Domain type for spatial homogenization
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energy_groups : openmc.mgxs.EnergyGroups
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Energy group structure for energy condensation
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num_polar : Integral
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Number of equi-width polar angle bins for angle discretization
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num_azimuthal : Integral
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Number of equi-width azimuthal angle bins for angle discretization
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tally_trigger : openmc.Trigger
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An (optional) tally precision trigger given to each tally used to
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compute the cross section
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scores : list of str
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The scores in each tally used to compute the multi-group cross section
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filters : list of openmc.Filter
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The filters in each tally used to compute the multi-group cross section
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tally_keys : list of str
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The keys into the tallies dictionary for each tally used to compute
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the multi-group cross section
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estimator : {'tracklength', 'collision', 'analog'}
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The tally estimator used to compute the multi-group cross section
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tallies : collections.OrderedDict
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OpenMC tallies needed to compute the multi-group cross section. The keys
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are strings listed in the :attr:`TotalXS.tally_keys` property and values
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are instances of :class:`openmc.Tally`.
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rxn_rate_tally : openmc.Tally
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Derived tally for the reaction rate tally used in the numerator to
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compute the multi-group cross section. This attribute is None
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unless the multi-group cross section has been computed.
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xs_tally : openmc.Tally
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Derived tally for the multi-group cross section. This attribute
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is None unless the multi-group cross section has been computed.
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num_subdomains : int
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The number of subdomains is unity for 'material', 'cell' and 'universe'
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domain types. This is equal to the number of cell instances
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for 'distribcell' domain types (it is equal to unity prior to loading
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tally data from a statepoint file).
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num_nuclides : int
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The number of nuclides for which the multi-group cross section is
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being tracked. This is unity if the by_nuclide attribute is False.
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nuclides : Iterable of str or 'sum'
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The optional user-specified nuclides for which to compute cross
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sections (e.g., 'U238', 'O16'). If by_nuclide is True but nuclides
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are not specified by the user, all nuclides in the spatial domain
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are included. This attribute is 'sum' if by_nuclide is false.
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sparse : bool
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Whether or not the MGXS' tallies use SciPy's LIL sparse matrix format
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for compressed data storage
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loaded_sp : bool
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Whether or not a statepoint file has been loaded with tally data
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derived : bool
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Whether or not the MGXS is merged from one or more other MGXS
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hdf5_key : str
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The key used to index multi-group cross sections in an HDF5 data store
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"""
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def __init__(self, rxn_type, domain=None, domain_type=None, groups=None,
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by_nuclide=False, name='', num_polar=1, num_azimuthal=1):
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super().__init__(domain, domain_type, groups, by_nuclide, name,
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num_polar, num_azimuthal)
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self._rxn_type = rxn_type
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class ScatterMatrixXS(MatrixMGXS):
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r"""A scattering matrix multi-group cross section with the cosine of the
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change-in-angle represented as one or more Legendre moments or a histogram.
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