# openmc.deplete.abc.TransportOperator¶

class openmc.deplete.abc.TransportOperator(chain_file, fission_q=None, dilute_initial=1000.0, prev_results=None)[source]

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.CoupledOperator or openmc.deplete.IndependentOperator rather than this class.

Parameters
• 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 chain_file.

• 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 (Results, optional) – Results from a previous depletion calculation.

Variables
• 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.

• output_dir (pathlib.Path) – Path to output directory to save results.

• prev_res (Results or None) – Results from a previous depletion calculation. None if no results are to be used.

• chain (openmc.deplete.Chain) – The depletion chain information necessary to form matrices and tallies.

abstract __call__(vec, source_rate)[source]

Runs a simulation.

Parameters
• vec (list of numpy.ndarray) – Total atoms to be used in function.

• source_rate (float) – Power in [W] or source rate in [neutron/sec]

Returns

Eigenvalue and reaction rates resulting from transport operator

Return type

openmc.deplete.OperatorResult

property dilute_initial

Initial atom density for nuclides with zero initial concentration

abstract get_results_info()[source]

Returns volume list, cell lists, and nuc lists.

Returns

• 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()[source]

Performs final setup and returns initial condition.

Returns

Total density for initial conditions.

Return type

list of numpy.ndarray

abstract write_bos_data(step)[source]

Document beginning of step data for a given step

Called at the beginning of a depletion step and at the final point in the simulation.

Parameters

step (int) – Current depletion step including restarts