openmc.deplete.abc.TransportOperator
- class openmc.deplete.abc.TransportOperator(chain_file=None, fission_q=None, 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.CoupledOperatororopenmc.deplete.IndependentOperatorrather than this class.- Parameters:
chain_file (PathLike or Chain) – Path to the depletion chain XML file or instance of openmc.deplete.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.prev_results (Results, optional) – Results from a previous depletion calculation.
- Variables:
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.
- abstractmethod __call__(vec, source_rate)[source]
Runs a simulation.
- Parameters:
vec (list of numpy.ndarray) – Total atoms to be used in function.
source_rate (float) – Power in [W] or source rate in [neutron/sec]
- Returns:
Eigenvalue and reaction rates resulting from transport operator
- Return type:
- abstractmethod get_results_info()[source]
Returns volume list, cell lists, and nuc lists.
- Returns:
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.