mcdc.MaterialMG¶

class mcdc.MaterialMG(name: str = '', capture: ndarray[tuple[Any, ...], dtype[float64]] | None = None, scatter: ndarray[tuple[Any, ...], dtype[float64]] | None = None, fission: ndarray[tuple[Any, ...], dtype[float64]] | None = None, nu_s: ndarray[tuple[Any, ...], dtype[float64]] | None = None, nu_p: ndarray[tuple[Any, ...], dtype[float64]] | None = None, nu_d: ndarray[tuple[Any, ...], dtype[float64]] | None = None, chi_p: ndarray[tuple[Any, ...], dtype[float64]] | None = None, chi_d: ndarray[tuple[Any, ...], dtype[float64]] | None = None, speed: ndarray[tuple[Any, ...], dtype[float64]] | None = None, decay_rate: ndarray[tuple[Any, ...], dtype[float64]] | None = None)¶

Define a multigroup material.

Cross-section arrays are provided as NumPy arrays of length G (number of energy groups). Scatter and fission matrices are (G, G).

Parameters:

name (str, optional) – User label.
capture (ndarray, optional) – Capture cross section for each group.
scatter (ndarray, optional) – Scattering matrix (G, G).
fission (ndarray, optional) – Fission cross section for each group.
nu_s (ndarray, optional) – Average scattering multiplicity.
nu_p (ndarray, optional) – Average prompt fission neutron yield.
nu_d (ndarray, optional) – Average delayed fission neutron yield.
chi_p (ndarray, optional) – Prompt fission spectrum.
chi_d (ndarray, optional) – Delayed fission spectrum.
speed (ndarray, optional) – Neutron speeds for each group (cm/s).
decay_rate (ndarray, optional) – Delayed neutron precursor decay rates (1/s).

Returns:

The multigroup material object.

Return type:

MaterialMG