endf.Reaction#

class endf.Reaction(MT, xs=None, products=None, q_reaction=0.0, q_massdiff=0.0, redundant=False)[source]#

A nuclear reaction

This class represents a single reaction channel for a nuclide with an associated cross section and, if present, a secondary angle and energy distribution.

Parameters:

  • mt (int) – The ENDF MT number for this reaction.

  • xs (dict) – Microscopic cross section for this reaction as a function of incident energy; these cross sections are provided in a dictionary where the key is the temperature of the cross section set.

  • products (list) – Reaction products

  • q_reaction (float) – The reaction Q-value in [eV].

  • q_massdiff (float) – The mass-difference Q value in [eV].

  • redundant (bool) – Indicates whether or not this is a redundant reaction

MT#

The ENDF MT number for this reaction.

Type:

int

products#

Reaction products

Type:

list

q_reaction#

The reaction Q-value in [eV].

Type:

float

q_massdiff#

The mass-difference Q value in [eV].

Type:

float

redundant#

Indicates whether or not this is a redundant reaction

Type:

bool

xs#

Microscopic cross section for this reaction as a function of incident energy; these cross sections are provided in a dictionary where the key is the temperature of the cross section set.

Type:

dict

Methods

from_ace(table, i_reaction)

Generate incident neutron continuous-energy data from an ACE table

from_endf(MT, material)

Generate reaction from ENDF file