openmc.data.MaxwellEnergy¶
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class
openmc.data.MaxwellEnergy(theta, u)[source]¶ Simple Maxwellian fission spectrum represented as
\[f(E \rightarrow E') = \frac{\sqrt{E'}}{I} e^{-E'/\theta(E)} \]Parameters: - theta (openmc.data.Tabulated1D) – Tabulated function of incident neutron energy
- u (float) – Constant introduced to define the proper upper limit for the final particle energy such that \(0 \le E' \le E - U\)
Variables: - theta (openmc.data.Tabulated1D) – Tabulated function of incident neutron energy
- u (float) – Constant introduced to define the proper upper limit for the final particle energy such that \(0 \le E' \le E - U\)
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classmethod
from_ace(ace, idx=0)[source]¶ Create a Maxwell distribution from an ACE table
Parameters: - ace (openmc.data.ace.Table) – An ACE table
- idx (int) – Offset to read from in XSS array (default of zero)
Returns: Maxwell distribution
Return type:
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classmethod
from_endf(file_obj, params)[source]¶ Generate Maxwell distribution from an ENDF evaluation
Parameters: - file_obj (file-like object) – ENDF file positioned at the start of a section for an energy distribution.
- params (list) – List of parameters at the start of the energy distribution that includes the LF value indicating what type of energy distribution is present.
Returns: Maxwell distribution
Return type:
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classmethod
from_hdf5(group)[source]¶ Generate Maxwell distribution from HDF5 data
Parameters: group (h5py.Group) – HDF5 group to read from Returns: Maxwell distribution Return type: openmc.data.MaxwellEnergy