# openmc.data.IncidentNeutron¶

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(), and IncidentNeutron.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. 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.
add_elastic_0K_from_endf(filename, overwrite=False)[source]

Append 0K elastic scattering cross section from an ENDF file.

Parameters: filename (str) – Path to ENDF file overwrite (bool) – If existing 0 K data is present, this flag can be used to indicate that it should be overwritten. Otherwise, an exception will be thrown. ValueError – If 0 K data is already present and the overwrite parameter is False.
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.
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.
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. Incident neutron continuous-energy data openmc.data.IncidentNeutron
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 Incident neutron continuous-energy data openmc.data.IncidentNeutron
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. Continuous-energy neutron interaction data openmc.data.IncidentNeutron
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() data – Incident neutron continuous-energy data openmc.data.IncidentNeutron
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. mts – ENDF MT numbers of reactions that make up the redundant reaction and have cross sections provided. list of int