mcdc.MaterialMG

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

namestr, optional

User label.

capturendarray, optional

Capture cross section for each group.

scatterndarray, optional

Scattering matrix (G, G).

fissionndarray, optional

Fission cross section for each group.

nu_sndarray, optional

Average scattering multiplicity.

nu_pndarray, optional

Average prompt fission neutron yield.

nu_dndarray, optional

Average delayed fission neutron yield.

chi_pndarray, optional

Prompt fission spectrum.

chi_dndarray, optional

Delayed fission spectrum.

speedndarray, optional

Neutron speeds for each group (cm/s).

decay_ratendarray, optional

Delayed neutron precursor decay rates (1/s).

Returns

MaterialMG

The multigroup material object.

See Also

mcdc.Material : Creates a continuous-energy material.