openmc.deplete.Operator

class openmc.deplete.Operator(geometry, settings, chain_file=None, prev_results=None, diff_burnable_mats=False, energy_mode='fission-q', fission_q=None, dilute_initial=1000.0, fission_yield_mode='constant', fission_yield_opts=None)[source]

OpenMC transport operator for depletion.

Instances of this class can be used to perform depletion using OpenMC as the transport operator. 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.

Parameters:
  • geometry (openmc.Geometry) – OpenMC geometry object
  • settings (openmc.Settings) – OpenMC Settings object
  • 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.
  • prev_results (ResultsList, 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. Default: False.
  • energy_mode ({"energy-deposition", "fission-q"}) – Indicator for computing system energy. "energy-deposition" will compute with a single energy deposition tally, taking fission energy release data and heating into consideration. "fission-q" will use the fission Q values from the depletion chain
  • 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 "energy_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. Defaults to 1.0e3.
  • 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.
Variables:
  • 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 (ResultsList or None) – Results from a previous depletion calculation. None if no results are to be used.
  • diff_burnable_mats (bool) – Whether to differentiate burnable materials with multiple instances
__call__(vec, power)[source]

Runs a simulation.

Simulation will abort under the following circumstances:

  1. No energy is computed using OpenMC tallies.
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
finalize()[source]

Finalize a depletion simulation and release resources.

get_results_info()[source]

Returns volume list, material lists, and nuc lists.

Returns:
  • volume (dict of str float) – Volumes corresponding to materials in full_burn_dict
  • 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 material IDs to be burned. Used for sorting the simulation.
  • full_burn_list (list) – List of all burnable material IDs
initial_condition()[source]

Performs final setup and returns initial condition.

Returns: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

Parameters:step (int) – Current depletion step including restarts