mcdc.nuclide

mcdc.nuclide(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, sensitivity=False, dsm_Np=1.0)

Create a nuclide card.

Parameters

capturenumpy.ndarray (1D), optional

Capture microscopic cross-section [barn].

scatternumpy.ndarray (2D), optional

Differential scattering microscopic cross-section [gout, gin] [barn].

fissionnumpy.ndarray (1D), optional

Fission microscopic cross-section [barn].

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 nuclide.

dsm_Npfloat

Average number of derivative particles produced at each sensitivity nuclide collision.

Returns

dictionary

A nuclide card.

Notes

Parameters are set to zeros by default. Energy group size G is determined by the size of capture, scatter, or fission. Thus, at least capture, scatter, or fission needs to be provided. nu_p or nu_d is needed if fission is provided. chi_p and chi_d are needed if nu_p and nu_d are provided, respectively, and G > 1. Delayed neutron precursor group size J is determined by the size of nu_d; if nu_d is not given, J = 0.

See also

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