Total cross section measurements of nuclear graphite and high density polyethylene

The experimental thermal neutron cross sections are important in nuclear data validation. The Evaluated Nuclear Data File (ENDF) library is a comprehensive repository of evaluated nuclear reaction data, including thermal neutron cross sections (thermal scattering laws, TSLs), for a wide range of materials. ENDF libraries importance lies in providing critical data for nuclear science and engineering applications, such as reactor design, nuclear medicine, and safety analysis, where accurate thermal neutron cross sections are essential for modeling neutron interactions and behaviors in various materials and environments. There are some known issues with graphite and polytethylene TSLs in the ENDF/B-VIII.0 and ENDF/B-VIII.1 releases, hence we would like to perform some transmission (total cross section) measurements to rectify these issues. We would also like to use these measurements on graphite to characterize the impact of small angle neutron scattering on the total cross section, as well as to validate the models used to calculate this contribution.