openmc.data.Reaction¶
- class openmc.data.Reaction(mt)[source]¶
A nuclear reaction
A Reaction object represents a single reaction channel for a nuclide with an associated cross section and, if present, a secondary angle and energy distribution.
- Parameters
mt (int) – The ENDF MT number for this reaction.
- Variables
center_of_mass (bool) – Indicates whether scattering kinematics should be performed in the center-of-mass or laboratory reference frame. grid above the threshold value in barns.
redundant (bool) – Indicates whether or not this is a redundant reaction
mt (int) – The ENDF MT number for this reaction.
q_value (float) – The Q-value of this reaction in eV.
xs (dict of str to openmc.data.Function1D) – Microscopic cross section for this reaction as a function of incident energy; these cross sections are provided in a dictionary where the key is the temperature of the cross section set.
products (Iterable of openmc.data.Product) – Reaction products
derived_products (Iterable of openmc.data.Product) – Derived reaction products. Used for ‘total’ fission neutron data when prompt/delayed data also exists.
- classmethod from_endf(ev, mt)[source]¶
Generate a reaction from an ENDF evaluation
- Parameters
ev (openmc.data.endf.Evaluation) – ENDF evaluation
mt (int) – The MT value of the reaction to get data for
- Returns
rx – Reaction data
- Return type
- classmethod from_hdf5(group, energy)[source]¶
Generate reaction from an HDF5 group
- Parameters
group (h5py.Group) – HDF5 group to read from
energy (dict) – Dictionary whose keys are temperatures (e.g., ‘300K’) and values are arrays of energies at which cross sections are tabulated at.
- Returns
Reaction data
- Return type