openmc.deplete.abc.TransportOperator

class openmc.deplete.abc.TransportOperator(chain_file=None, 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.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.

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

__call__(vec, power)

Runs a simulation.

Parameters:

• vec (list of numpy.ndarray) – Total atoms to be used in function.
• power (float) – Power of the reactor in [W]

Returns:

Eigenvalue and reaction rates resulting from transport operator

Return type:

openmc.deplete.OperatorResult

dilute_initial

Initial atom density for nuclides with zero initial concentration

get_results_info()

Returns volume list, cell lists, and nuc lists.

Returns:

• volume (list of 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()

Performs final setup and returns initial condition.

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

write_bos_data(step)

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