# openmc.deplete.abc.EnergyHelper¶

class openmc.deplete.abc.EnergyHelper[source]

Abstract class for obtaining energy produced

The ultimate goal of this helper is to provide instances of openmc.deplete.Operator with the total energy produced in a transport simulation. This information, provided with the power requested by the user and reaction rates from a ReactionRateHelper will scale reaction rates to the correct values.

Variables: nuclides (list of str) – All nuclides with desired reaction rates. Ordered to be consistent with openmc.deplete.Operator energy (float) – Total energy [J/s/source neutron] produced in a transport simulation. Updated in the material iteration with update().
nuclides

List of nuclides with requested reaction rates

prepare(chain_nucs, rate_index, materials)[source]

Perform work needed to obtain energy produced

This method is called prior to the transport simulations in openmc.deplete.Operator.initial_condition().

Parameters: chain_nucs (list of str) – All nuclides to be tracked in this problem rate_index (dict of str to int) – Mapping from nuclide name to index in the fission_rates for update(). materials (list of str) – All materials tracked on the operator helped by this object. Should correspond to openmc.deplete.Operator.burnable_materials
reset()[source]

Reset energy produced prior to unpacking tallies

update(fission_rates, mat_index)[source]

Update the energy produced

Parameters: fission_rates (numpy.ndarray) – fission reaction rate for each isotope in the specified material. Should be ordered corresponding to initial rate_index used in prepare() mat_index (int) – Index for the specific material in the list of all burnable materials.