openmc.deplete.ResultsList

class openmc.deplete.ResultsList

A list of openmc.deplete.Results objects

It is recommended to use from_hdf5() over direct creation.

classmethod from_hdf5(filename)

Load in depletion results from a previous file

Parameters:filename (str) – Path to depletion result file Returns:new – New instance of depletion results Return type:ResultsList

get_atoms(mat, nuc, nuc_units='atoms', time_units='s')

Get number of nuclides over time from a single material

Note

Initial values for some isotopes that do not appear in initial concentrations may be non-zero, depending on the value of openmc.deplete.Operator dilute_initial. The openmc.deplete.Operator adds isotopes according to this setting, which can be set to zero.

Parameters:

• mat (str) – Material name to evaluate
• nuc (str) – Nuclide name to evaluate
• nuc_units ({"atoms", "atom/b-cm", "atom/cm3"}, optional) –

  Units for the returned concentration. Default is "atoms"

  New in version 0.12.

• time_units ({"s", "min", "h", "d"}, optional) –

  Units for the returned time array. Default is "s" to return the value in seconds.

  New in version 0.12.

Returns:

• times (numpy.ndarray) – Array of times in units of time_units
• concentrations (numpy.ndarray) – Concentration of specified nuclide in units of nuc_units

get_depletion_time()

Return an array of the average time to deplete a material

Note

Will have one fewer row than number of other methods, like get_eigenvalues(), because no depletion is performed at the final transport stage

Returns:times – Vector of average time to deplete a single material across all processes and materials. Return type:numpy.ndarray

get_eigenvalue()

Evaluates the eigenvalue from a results list.

Returns:

• times (numpy.ndarray) – Array of times in [s]
• eigenvalues (numpy.ndarray) – k-eigenvalue at each time. Column 0 contains the eigenvalue, while column 1 contains the associated uncertainty

get_reaction_rate(mat, nuc, rx)

Get reaction rate in a single material/nuclide over time

Note

Initial values for some isotopes that do not appear in initial concentrations may be non-zero, depending on the value of openmc.deplete.Operator dilute_initial The openmc.deplete.Operator adds isotopes according to this setting, which can be set to zero.

Parameters:

• mat (str) – Material name to evaluate
• nuc (str) – Nuclide name to evaluate
• rx (str) – Reaction rate to evaluate

Returns:

• times (numpy.ndarray) – Array of times in [s]
• rates (numpy.ndarray) – Array of reaction rates