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 aReactionRateHelper
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
-
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 inprepare()
- mat_index (int) – Index for the specific material in the list of all burnable materials.
- fission_rates (numpy.ndarray) – fission reaction rate for each isotope in the specified
material. Should be ordered corresponding to initial
- nuclides (list of str) – All nuclides with desired reaction rates. Ordered to be
consistent with