mcdc.material

mcdc.material(nuclides=None, capture=None, scatter=None, fission=None, nu_s=None, nu_p=None, nu_d=None, chi_p=None, chi_d=None, speed=None, decay=None, name='P', sensitivity=False, dsm_Np=1.0)

Create a material card.

The material card is defined either as a collection of nuclides or directly by its macroscopic constants.

Parameters

nuclideslist of tuple of (dictionary, float), optional

List of pairs of nuclide card and its density [/barn-cm].

capturenumpy.ndarray (1D), optional

Capture macroscopic cross-section [/cm].

scatternumpy.ndarray (2D), optional

Differential scattering macroscopic cross-section [gout, gin] [/cm].

fissionnumpy.ndarray (1D), optional

Fission macroscopic cross-section [/cm].

nu_snumpy.ndarray (1D), optional

Scattering multiplication.

nu_pnumpy.ndarray (1D), optional

Prompt fission neutron yield.

nu_dnumpy.ndarray (2D), optional

Delayed neutron precursor yield [dg, gin].

chi_pnumpy.ndarray (2D), optional

Prompt fission spectrum [gout, gin].

chi_dnumpy.ndarray (2D), optional

Delayed neutron spectrum [gout, dg].

speednumpy.ndarray (1D), optional

Energy group speed [cm/s].

decaynumpy.ndarray (1D), optional

Precursor group decay constant [/s].

sensitivitybool, optional

Set to True to calculate sensitivities to the material (only relevant for single-nuclide material).

dsm_Npfloat

Average number of derivative particles produced at each sensitivity material collision (only relevant for single_nuclide material).

Returns

dictionary

A material card.

See also

mcdc.nuclide : A material can be defined as a collection of nuclides.