endf-python.

• What is the cross section for a specific reaction on a nuclide at a given energy in a particular evaluated library?
• Which reaction channels and secondary products does an evaluation define for a target nuclide?
• What does an ACE table used by a transport code contain, and does it match the evaluation it was processed from?

• Reading ENDF-6 files (any MF/MT section) through a low-level material interface
• Reading ACE continuous-energy tables and locating them through an xsdir
• Interpreted incident-neutron data: reactions, cross sections, Q-values, products and yields
• Tabulated 1-D functions with interpolation at arbitrary energies

Limitations

A reader and interpreter, not a processing code: no resonance reconstruction, Doppler broadening or group collapse. The high-level interface covers incident-neutron data; other sublibraries are reachable only through the raw interface. The package is marked alpha, development is concentrated in one person, and there is no formal citation.

A general nuclear-data reader, not a fusion code. Fusion neutronics and activation work run on ENDF-6 libraries and ACE tables, and this package lets an engineer inspect what a library says about a blanket, first-wall or shield material near 14 MeV before feeding it to a transport or activation code.

Published by the project itself. These are the project's own tutorials, not Morton Labs runs.

• Detailed usage: low-level interface — Reads a U-235 file, pulls an (n,2n) section and plots the cross section.
• Detailed usage: ACE files — Opens an ACE table and interprets it as incident-neutron data.

No reference workflow documents this tool yet. Its evidence status is “Listed”: A reviewed catalog entry with upstream links.

Roles name a specific relationship with evidence. Authorship of a paper is not maintainership, and a listed person is not a Morton Labs partner or endorser.

• Awesome Nuclear — Curated by Paul Romano and contributors