class openmc.deplete.CoupledOperator(model, chain_file=None, prev_results=None, diff_burnable_mats=False, normalization_mode='fission-q', fission_q=None, dilute_initial=1000.0, fission_yield_mode='constant', fission_yield_opts=None, reaction_rate_mode='direct', reaction_rate_opts=None, reduce_chain=False, reduce_chain_level=None)[source]

Transport-coupled transport operator.

Instances of this class can be used to perform transport-coupled depletion using OpenMC’s transport solver. Normally, a user needn’t call methods of this class directly. Instead, an instance of this class is passed to an integrator class, such as openmc.deplete.CECMIntegrator.

Changed in version 0.13.0: The geometry and settings parameters have been replaced with a model parameter that takes a Model object

Changed in version 0.13.1: Name changed from Operator to CoupledOperator

  • model (openmc.model.Model) – OpenMC model object

  • chain_file (str, optional) – Path to the depletion chain XML file. Defaults to openmc.config['chain_file'].

  • prev_results (Results, optional) – Results from a previous depletion calculation. If this argument is specified, the depletion calculation will start from the latest state in the previous results.

  • diff_burnable_mats (bool, optional) – Whether to differentiate burnable materials with multiple instances. Volumes are divided equally from the original material volume.

  • normalization_mode ({"energy-deposition", "fission-q", "source-rate"}) – Indicate how tally results should be normalized. "energy-deposition" computes the total energy deposited in the system and uses the ratio of the power to the energy produced as a normalization factor. "fission-q" uses the fission Q values from the depletion chain to compute the total energy deposited. "source-rate" normalizes tallies based on the source rate (for fixed source calculations).

  • fission_q (dict, optional) – Dictionary of nuclides and their fission Q values [eV]. If not given, values will be pulled from the chain_file. Only applicable if "normalization_mode" == "fission-q"

  • 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.

  • fission_yield_mode ({"constant", "cutoff", "average"}) –

    Key indicating what fission product yield scheme to use. The key determines what fission energy helper is used:

    The documentation on these classes describe their methodology and differences. Default: "constant"

  • fission_yield_opts (dict of str to option, optional) – Optional arguments to pass to the helper determined by fission_yield_mode. Will be passed directly on to the helper. Passing a value of None will use the defaults for the associated helper.

  • reaction_rate_mode ({"direct", "flux"}, optional) –

    Indicate how one-group reaction rates should be calculated. The “direct” method tallies transmutation reaction rates directly. The “flux” method tallies a multigroup flux spectrum and then collapses one-group reaction rates after a transport solve (with an option to tally some reaction rates directly).

    New in version 0.12.1.

  • reaction_rate_opts (dict, optional) –

    Keyword arguments that are passed to the reaction rate helper class. When reaction_rate_mode is set to “flux”, energy group boundaries can be set using the “energies” key. See the FluxCollapseHelper class for all options.

    New in version 0.12.1.

  • reduce_chain (bool, optional) –

    If True, use openmc.deplete.Chain.reduce() to reduce the depletion chain up to reduce_chain_level.

    New in version 0.12.

  • reduce_chain_level (int, optional) –

    Depth of the search when reducing the depletion chain. Only used if reduce_chain evaluates to true. The default value of None implies no limit on the depth.

    New in version 0.12.

  • model (openmc.model.Model) – OpenMC model object

  • geometry (openmc.Geometry) – OpenMC geometry object

  • settings (openmc.Settings) – OpenMC settings object

  • 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.

  • round_number (bool) – Whether or not to round output to OpenMC to 8 digits. Useful in testing, as OpenMC is incredibly sensitive to exact values.

  • number (openmc.deplete.AtomNumber) – Total number of atoms in simulation.

  • nuclides_with_data (set of str) – A set listing all unique nuclides available from cross_sections.xml.

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

  • reaction_rates (openmc.deplete.ReactionRates) – Reaction rates from the last operator step.

  • burnable_mats (list of str) – All burnable material IDs

  • heavy_metal (float) – Initial heavy metal inventory [g]

  • local_mats (list of str) – All burnable material IDs being managed by a single process

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

  • cleanup_when_done (bool) – Whether to finalize and clear the shared library memory when the depletion operation is complete. Defaults to clearing the library.

__call__(vec, source_rate)[source]

Runs a simulation.

Simulation will abort under the following circumstances:

  1. No energy is computed using OpenMC tallies.

  • 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

Return type



Finalize a depletion simulation and release resources.


Performs final setup and returns initial condition.


Total density for initial conditions.

Return type

list of numpy.ndarray

static write_bos_data(step)[source]

Write a state-point file with beginning of step data


step (int) – Current depletion step including restarts