- class openmc.deplete.openmc_operator.OpenMCOperator(materials=None, cross_sections=None, chain_file=None, prev_results=None, diff_burnable_mats=False, fission_q=None, dilute_initial=0.0, helper_kwargs=None, reduce_chain=False, reduce_chain_level=None)¶
Abstract class holding OpenMC-specific functions for running depletion calculations.
Specific classes for running transport-coupled or transport-independent depletion calculations are implemented as subclasses of OpenMCOperator.
materials (openmc.Materials) – List of all materials in the model
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.
fission_q (dict, optional) – Dictionary of nuclides and their fission Q values [eV].
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.
helper_kwargs (dict) – Keyword arguments for helper classes
reduce_chain_level (int, optional) – Depth of the search when reducing the depletion chain. Only used if
reduce_chainevaluates to true. The default value of
Noneimplies no limit on the depth.
materials (openmc.Materials) – All materials present in the model
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.
Noneif no results are to be used.
Returns volume list, material lists, and nuc lists.
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
Performs final setup and returns initial condition.
materials (list of str) – list of material IDs
Total density for initial conditions.
- Return type
list of numpy.ndarray