openmc.deplete.helpers.ChainFissionHelper

class openmc.deplete.helpers.ChainFissionHelper

Computes energy using fission Q values from depletion chain

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().

prepare(chain_nucs, rate_index, _materials)

Populate the fission Q value vector from a chain.

Parameters:

• chain_nucs (iterable of openmc.deplete.Nuclide) – Nuclides used in this depletion chain. Do not need to be ordered
• rate_index (dict of str to int) – Dictionary mapping names of nuclides, e.g. "U235", to a corresponding index in the desired fission Q vector.
• _materials (list of str) – Unused. Materials to be tracked for this helper.

update(fission_rates, _mat_index)

Update energy produced with fission rates in a material

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 material requested. Unused, as identical isotopes in all materials have the same Q value.