openmc.data.IncidentNeutron¶
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class
openmc.data.
IncidentNeutron
(name, atomic_number, mass_number, metastable, atomic_weight_ratio, kTs)[source]¶ Continuous-energy neutron interaction data.
This class stores data derived from an ENDF-6 format neutron interaction sublibrary. Instances of this class are not normally instantiated by the user but rather created using the factory methods
IncidentNeutron.from_hdf5()
,IncidentNeutron.from_ace()
, andIncidentNeutron.from_endf()
.Parameters: - name (str) – Name of the nuclide using the GND naming convention
- atomic_number (int) – Number of protons in the target nucleus
- mass_number (int) – Number of nucleons in the target nucleus
- metastable (int) – Metastable state of the target nucleus. A value of zero indicates ground state.
- atomic_weight_ratio (float) – Atomic mass ratio of the target nuclide.
- kTs (Iterable of float) – List of temperatures of the target nuclide in the data set. The temperatures have units of eV.
Variables: - atomic_number (int) – Number of protons in the target nucleus
- atomic_symbol (str) – Atomic symbol of the nuclide, e.g., ‘Zr’
- atomic_weight_ratio (float) – Atomic weight ratio of the target nuclide.
- fission_energy (None or openmc.data.FissionEnergyRelease) – The energy released by fission, tabulated by component (e.g. prompt neutrons or beta particles) and dependent on incident neutron energy
- mass_number (int) – Number of nucleons in the target nucleus
- metastable (int) – Metastable state of the target nucleus. A value of zero indicates ground state.
- name (str) – Name of the nuclide using the GND naming convention
- reactions (collections.OrderedDict) – Contains the cross sections, secondary angle and energy distributions, and other associated data for each reaction. The keys are the MT values and the values are Reaction objects.
- resonances (openmc.data.Resonances or None) – Resonance parameters
- resonance_covariance (openmc.data.ResonanceCovariance or None) – Covariance for resonance parameters
- temperatures (list of str) – List of string representations the temperatures of the target nuclide in the data set. The temperatures are strings of the temperature, rounded to the nearest integer; e.g., ‘294K’
- kTs (Iterable of float) – List of temperatures of the target nuclide in the data set. The temperatures have units of eV.
- urr (dict) – Dictionary whose keys are temperatures (e.g., ‘294K’) and values are unresolved resonance region probability tables.
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add_elastic_0K_from_endf
(filename, overwrite=False)[source]¶ Append 0K elastic scattering cross section from an ENDF file.
Parameters: Raises: ValueError
– If 0 K data is already present and the overwrite parameter is False.
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add_temperature_from_ace
(ace_or_filename, metastable_scheme='nndc')[source]¶ Append data from an ACE file at a different temperature.
Parameters: - ace_or_filename (openmc.data.ace.Table or str) – ACE table to read from. If given as a string, it is assumed to be the filename for the ACE file.
- metastable_scheme ({'nndc', 'mcnp'}) – Determine how ZAID identifiers are to be interpreted in the case of a metastable nuclide. Because the normal ZAID (=1000*Z + A) does not encode metastable information, different conventions are used among different libraries. In MCNP libraries, the convention is to add 400 for a metastable nuclide except for Am242m, for which 95242 is metastable and 95642 (or 1095242 in newer libraries) is the ground state. For NNDC libraries, ZAID is given as 1000*Z + A + 100*m.
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export_to_hdf5
(path, mode='a', libver='earliest')[source]¶ Export incident neutron data to an HDF5 file.
Parameters: - path (str) – Path to write HDF5 file to
- mode ({'r', r+', 'w', 'x', 'a'}) – Mode that is used to open the HDF5 file. This is the second argument
to the
h5py.File
constructor. - libver ({'earliest', 'latest'}) – Compatibility mode for the HDF5 file. ‘latest’ will produce files that are less backwards compatible but have performance benefits.
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classmethod
from_ace
(ace_or_filename, metastable_scheme='nndc')[source]¶ Generate incident neutron continuous-energy data from an ACE table
Parameters: - ace_or_filename (openmc.data.ace.Table or str) – ACE table to read from. If the value is a string, it is assumed to be the filename for the ACE file.
- metastable_scheme ({'nndc', 'mcnp'}) – Determine how ZAID identifiers are to be interpreted in the case of a metastable nuclide. Because the normal ZAID (=1000*Z + A) does not encode metastable information, different conventions are used among different libraries. In MCNP libraries, the convention is to add 400 for a metastable nuclide except for Am242m, for which 95242 is metastable and 95642 (or 1095242 in newer libraries) is the ground state. For NNDC libraries, ZAID is given as 1000*Z + A + 100*m.
Returns: Incident neutron continuous-energy data
Return type:
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classmethod
from_endf
(ev_or_filename, covariance=False)[source]¶ Generate incident neutron continuous-energy data from an ENDF evaluation
Parameters: - ev_or_filename (openmc.data.endf.Evaluation or str) – ENDF evaluation to read from. If given as a string, it is assumed to be the filename for the ENDF file.
- covariance (bool) – Flag to indicate whether or not covariance data from File 32 should be retrieved
Returns: Incident neutron continuous-energy data
Return type:
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classmethod
from_hdf5
(group_or_filename)[source]¶ Generate continuous-energy neutron interaction data from HDF5 group
Parameters: group_or_filename (h5py.Group or str) – HDF5 group containing interaction data. If given as a string, it is assumed to be the filename for the HDF5 file, and the first group is used to read from. Returns: Continuous-energy neutron interaction data Return type: openmc.data.IncidentNeutron
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classmethod
from_njoy
(filename, temperatures=None, evaluation=None, **kwargs)[source]¶ Generate incident neutron data by running NJOY.
Parameters: - filename (str) – Path to ENDF file
- temperatures (iterable of float) – Temperatures in Kelvin to produce data at. If omitted, data is produced at room temperature (293.6 K)
- evaluation (openmc.data.endf.Evaluation, optional) – If the ENDF file contains multiple material evaluations, this argument indicates which evaluation to use.
- **kwargs – Keyword arguments passed to
openmc.data.njoy.make_ace()
Returns: data – Incident neutron continuous-energy data
Return type:
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get_reaction_components
(mt)[source]¶ Determine what reactions make up redundant reaction.
Parameters: mt (int) – ENDF MT number of the reaction to find components of. Returns: mts – ENDF MT numbers of reactions that make up the redundant reaction and have cross sections provided. Return type: list of int