saltproc.openmc_depcode module

class saltproc.openmc_depcode.OpenMCDepcode(exec_path='openmc_deplete.py', template_input_file_path={'chain_file': './chain_simple.xml', 'geometry': './geometry.xml', 'settings': './settings.xml'}, geo_files=None, npop=50, active_cycles=20, inactive_cycles=20)

Bases: Depcode

Class contains information about input, output, geometry, and template files for running OpenMC depletion simulations. Also contains neutrons population, active, and inactive cycles. Contains methods to read template and output files, write new input files for OpenMC.

param

Holds depletion step parameter information. Parameter names are keys and parameter values are values.

Type

dict of str to type

sim_info

Holds simulation settings information. Setting names are keys and setting values are values.

Type

dict of str to type

iter_inputfile

Paths to OpenMC input files for OpenMC rerunning.

Type

dict of str to str

iter_matfile

Path to iterative, rewritable material file for OpenMC rerunning. This file is modified during the simulation.

Type

str

__init__(exec_path='openmc_deplete.py', template_input_file_path={'chain_file': './chain_simple.xml', 'geometry': './geometry.xml', 'settings': './settings.xml'}, geo_files=None, npop=50, active_cycles=20, inactive_cycles=20)

Initializes the OpenMCDepcode object.

Parameters

• 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_files (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.

• npop (int, optional) – Size of neutron population per cycle for Monte Carlo.

• active_cycles (int, optional) – Number of active cycles.

• inactive_cycles (int, optional) – Number of inactive cycles.

read_dep_comp(read_at_end=False)

Reads the depleted material data from the OpenMC depletion simulation and returns a dictionary with a Materialflow object for each burnable material.

Parameters

read_at_end (bool, optional) – Controls at which moment in the depletion step to read the data. 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

mats – Dictionary that contains Materialflow objects.

key

Name of burnable material.

value

Materialflow object holding composition and properties.

Return type

dict of str to Materialflow

read_depcode_info()

Parses initial OpenMC simulation info from the OpenMC output files and stores it in the Depcode object’s sim_info attribute.

read_depcode_step_param()

Parses data from OpenMC depletion output for each step and stores it in Depcode object’s param attributes.

run_depcode(cores, nodes)

Runs OpenMC depletion simulation as a subprocess with the given parameters.

Parameters

• cores (int) – Number of cores to use for depletion code run.

• nodes (int) – Number of nodes to use for depletion code run.

switch_to_next_geometry()

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

write_depcode_input(reactor, dep_step, restart)

Writes prepared data into OpenMC input file(s).

Parameters

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

• dep_step (int) – Current depletion time step.

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

write_depletion_settings(reactor, current_depstep_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.

• current_depstep_idx (int) – Current depletion step.

write_mat_file(dep_dict, dep_end_time)

Writes the iteration input file containing the burnable materials composition used in OpenMC depletion runs and updated after each depletion step.

Parameters

• dep_dict (dict of str to Materialflow) –

  Dictionary that contains Materialflow objects.

  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_saltproc_openmc_tallies(materials, geometry)

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