- class openmc.deplete.abc.TransportOperator(chain_file, 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.IndependentOperatorrather than this class.
chain_file (str) – Path to the depletion chain XML file
fission_q (dict, optional) – Dictionary of nuclides and their fission Q values [eV]. If not given, values will be pulled from the
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 (Results, 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.
output_dir (pathlib.Path) – Path to output directory to save results.
prev_res (Results or None) – Results from a previous depletion calculation.
Noneif no results are to be used.
chain (openmc.deplete.Chain) – The depletion chain information necessary to form matrices and tallies.
- abstract __call__(vec, source_rate)[source]¶
Runs a simulation.
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
- property dilute_initial¶
Initial atom density for nuclides with zero initial concentration
- abstract get_results_info()[source]¶
Returns volume list, cell lists, and nuc lists.
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.
- abstract initial_condition()[source]¶
Performs final setup and returns initial condition.
Total density for initial conditions.
- Return type
list of numpy.ndarray