# openmc.deplete.abc.TransportOperator¶

class openmc.deplete.abc.TransportOperator(chain_file=None, 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.Operator rather than this class.

Parameters: chain_file (str, optional) – Path to the depletion chain XML file. Defaults to the file listed under depletion_chain in OPENMC_CROSS_SECTIONS environment variable. 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 (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. prev_res (ResultsList or None) – Results from a previous depletion calculation. None if no results are to be used.
__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] Eigenvalue and reaction rates resulting from transport operator openmc.deplete.OperatorResult
dilute_initial

Initial atom density for nuclides with zero initial concentration

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.
initial_condition()[source]

Performs final setup and returns initial condition.

Returns: Total density for initial conditions. list of numpy.ndarray
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