- class openmc.deplete.abc.TransportOperator(chain_file, fission_q=None, dilute_initial=1000.0, prev_results=None)¶
Abstract class defining a transport operator
Each depletion integrator is written to work with a generic transport operator that takes a vector of material compositions and returns an eigenvalue and reaction rates. This abstract class sets the requirements for such a transport operator. Users should instantiate
openmc.deplete.Operatorrather than this class.
chain_file (str) – Path to the depletion chain XML file
fission_q (dict, optional) – Dictionary of nuclides and their fission Q values [eV]. If not given, values will be pulled from the
dilute_initial (float, optional) – Initial atom density [atoms/cm^3] to add for nuclides that are zero in initial condition to ensure they exist in the decay chain. Only done for nuclides with reaction rates. Defaults to 1.0e3.
prev_results (ResultsList, optional) – Results from a previous depletion calculation.
dilute_initial (float) – Initial atom density [atoms/cm^3] to add for nuclides that are zero in initial condition to ensure they exist in the decay chain. Only done for nuclides with reaction rates.
- abstract __call__(vec, source_rate)¶
Runs a simulation.
- property dilute_initial¶
Initial atom density for nuclides with zero initial concentration
- abstract get_results_info()¶
Returns volume list, cell lists, and nuc lists.
volume (dict of str to float) – Volumes corresponding to materials in burn_list
nuc_list (list of str) – A list of all nuclide names. Used for sorting the simulation.
burn_list (list of int) – A list of all cell IDs to be burned. Used for sorting the simulation.
full_burn_list (list of int) – All burnable materials in the geometry.
- abstract initial_condition()¶
Performs final setup and returns initial condition.
Total density for initial conditions.
- Return type
list of numpy.ndarray