openmc.deplete.abc.TransportOperator¶
-
class
openmc.deplete.abc.
TransportOperator
(chain_file=None, 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.Operator
rather than this class.Parameters: - chain_file (str, optional) – Path to the depletion chain XML file. Defaults to the file
listed under
depletion_chain
inOPENMC_CROSS_SECTIONS
environment variable. - fission_q (dict, optional) – Dictionary of nuclides and their fission Q values [eV]. If not given,
values will be pulled from the
chain_file
. - 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 (ResultsList, optional) – Results from a previous depletion calculation.
Variables: - 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.
- prev_res (ResultsList or None) – Results from a previous depletion calculation.
None
if no results are to be used.
-
__call__
(vec, power)[source]¶ Runs a simulation.
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:
-
dilute_initial
¶ Initial atom density for nuclides with zero initial concentration
-
get_results_info
()[source]¶ Returns volume list, cell lists, and nuc lists.
Returns: - volume (list of 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.
- chain_file (str, optional) – Path to the depletion chain XML file. Defaults to the file
listed under