openmc.data.IncidentNeutron

class openmc.data.IncidentNeutron(name, atomic_number, mass_number, metastable, atomic_weight_ratio, kTs)[source]

Continuous-energy neutron interaction data.

Instances of this class are not normally instantiated by the user but rather created using the factory methods IncidentNeutron.from_hdf5() and IncidentNeutron.from_ace().

Parameters:
  • name (str) – Name of the nuclide using the GND naming convention
  • atomic_number (int) – Number of protons in the nucleus
  • mass_number (int) – Number of nucleons in the nucleus
  • metastable (int) – Metastable state of the 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 nucleus
  • atomic_symbol (str) – Atomic symbol of the nuclide, e.g., ‘Zr’
  • atomic_weight_ratio (float) – Atomic weight ratio of the target nuclide.
  • energy (dict of numpy.ndarray) – The energy values (eV) at which reaction cross-sections are tabulated. They keys of the dict are the temperature string (‘294K’) for each set of energies
  • 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 nucleus
  • metastable (int) – Metastable state of the 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
  • summed_reactions (collections.OrderedDict) – Contains summed cross sections, e.g., the total cross section. The keys are the MT values and the values are Reaction objects.
  • 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.
Raises:

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.
Returns:

Incident neutron continuous-energy data

Return type:

openmc.data.IncidentNeutron

classmethod from_endf(ev_or_filename)[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.
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, **kwargs)[source]

Generate incident neutron data by running NJOY.

Parameters:
  • filename (str) – Path to ENDF evaluation
  • temperatures (iterable of float) – Temperatures in Kelvin to produce data at. If omitted, data is produced at room temperature (293.6 K)
  • **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 summed 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 summed reaction and have cross sections provided.
Return type:list of int