saltproc.OpenMCDepcode

class saltproc.OpenMCDepcode(output_path, exec_path, template_input_file_path, geo_file_paths, depletion_settings, chain_file_path)

Interface for running depletion steps in OpenMC, as well as obtaining depletion step results.

Parameters

• output_path (str) – Path to results storage directory.

• exec_path (str) – Path to OpenMC depletion script.

• template_input_file_path (dict of str to str) – Path to user input files (.xml file for geometry, material, and settings) for OpenMC. File type as strings are keys (e.g. ‘geometry’, ‘settings’, ‘material’), and file path as strings are values.

• geo_file_paths (str or list, optional) –

  Path to file that contains the reactor geometry.

  List of str if reactivity control by switching geometry is On or just str otherwise.

  depletion_settingsdict

  Keyword arguments to pass to openmc.model.deplete().

  chain_file_pathstr

  Path to depletion chain file

Variables

• neutronics_parameters (dict of str to type) – Holds OpenMC depletion step neutronics parameters. Parameter names are keys and parameter values are values.

• step_metadata (dict of str to type) – Holds OpenMC depletion step metadata. Metadata labels are keys and metadata values are values.

• depcode_metadata (dict of str to type) – Holds OpenMC simulation metadata. Metadata labels are keys and metadata values are values.

• runtime_inputfile (dict of str to str) – Paths to OpenMC input files used to run depletion step. Contains neutron settings and geometry.

• runtime_matfile (str) – Path to OpenMC material file containing materials used to run depletion step, and modified after fuel reprocessing.

• npop (int) – Size of neutron population per cycle

• active_cycles (int) – Number of active cycles.

• inactive_cycles (int) – Number of inactive cycles.

name_to_nuclide_code(nucname)

Converts depcode nuclide name to ZA nuclide code

Parameters

nucname (str) – Nuclide namce

Returns

nuc_name – ZA nuclide code

Return type

str

nuclide_code_to_name(nuc)

Converts depcode nuclide code to symbolic nuclide name.

Parameters

nuc_code (str) – Nuclide code

Returns

nuc_name – Symbolic nuclide name (Am242m1).

Return type

str

read_depcode_metadata()

Reads OpenMC’s depletion step metadata and stores it in the OpenMCDepcode object’s step_metadata attribute.

read_depleted_materials(read_at_end=False)

Reads depleted materials from OpenMC’s depletion_results.h5 file and returns a dictionary with a Materialflow object for each depleted material.

Parameters

read_at_end (bool, optional) – If True, the function reads data at the end of the depletion step. Otherwise, the function reads data at the beginning of the depletion step.

Returns

depleted_materials – Dictionary containing depleted materials.

key

Name of burnable material.

value

Materialflow object holding composition and properties.

Return type

dict of str to Materialflow

read_neutronics_parameters()

Reads OpenMC depletion step neutronics parameters and stores them in OpenMCDepcode object’s neutronics_parameters attribute.

read_step_metadata()

Reads OpenMC’s depletion step metadata and stores it in the OpenMCDepcode object’s step_metadata attribute.

run_depletion_step(mpi_args=None, threads=None)

Runs a depletion step in OpenMC as a subprocess

mpi_argslist of str

Arguments for running simulations on supercomputers using mpiexec or similar programs.

threadsint

Threads to use for shared-memory parallelism

switch_to_next_geometry()

Switches the geometry file for the OpenMC depletion simulation to the next geometry file in geo_file_paths.

update_depletable_materials(mats, dep_end_time)

Updates material file with reprocessed material compositions.

Parameters

• mats (dict of str to Materialflow) –

  Dictionary containing reprocessed material compositions

  key

  Name of burnable material.

  value

  Materialflow object holding composition and properties.

• dep_end_time (float) – Current time at the end of the depletion step (d).

write_depletion_settings(reactor, step_idx)

Write the depeletion settings for the openmc.deplete module.

Parameters

• reactor (Reactor) – Contains information about power load curve and cumulative depletion time for the integration test.

• step_idx (int) – Current depletion step.

write_runtime_input(reactor, depletion_step, restart)

Write OpenMC runtime input files for running depletion step.

Parameters

• reactor (Reactor) – Contains information about power load curve and cumulative depletion time for the integration test.

• depletion_step (int) – Current depletion time step.

• restart (bool) – Is the current simulation restarted?

write_saltproc_openmc_tallies(materials, geometry, energy_bounds, n_delayed_groups)

Write tallies for calculating burnup and delayed neutron parameters.

Parameters

• materials (openmc.Materials object) – The materials for the depletion simulation

• geometry (openmc.Geometry object) – The geometry for the depletion simulation

• energy_bounds (iterable of float) – Energy group boundaries for calculating \(\beta\), the delayed neutron fraction, and \(\lambda\), the decay rate for delayed neutron precursors.

• n_delayed_groups (int) – Number of delayed groups for calculating \(\beta\), the delayed neutron fraction, and \(\lambda\), the decay rate for delayed neutron precursors.