# 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.

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 of 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.

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.

Raises

ValueError – If 0 K data is already present and the overwrite parameter is False.

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+', '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.

Returns

Incident neutron continuous-energy data

Return type

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

Returns

Incident neutron continuous-energy data

Return type

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.

Returns

Continuous-energy neutron interaction data

Return type

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()

Returns

data – Incident neutron continuous-energy data

Return type

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.

Returns

mts – ENDF MT numbers of reactions that make up the redundant reaction and have cross sections provided.

Return type

list of int