Multi-group neutron cross section generation capability for FRENDY nuclear data processing code

Yamamoto, Akio*; Tada, Kenichi   ; Chiba, Go*; Endo, Tomohiro*

The multi-group cross section generation capability for neutrons is implemented in the FRENDY nuclear data processing code. ACE-formatted files are used as the source of nuclear data instead of ENDF-formatted files since FRENDY already has the capability to generate pointwise cross sections in the ACE format. Verification calculations of the newly implemented capability are carried out through the comparison with the NJOY nuclear data processing code. Cross section generations for all nuclides in JENDL-4.0, -4.0u, -54, ENDF/B-VII.1, -VIII.0, JEFF-3.3, and TENDL-2019 are carried out without unexpected processing issue, except for Pu-238 of TENDL-2019 that includes inconsistent data. The verification results indicate that the multi-group cross sections generated by FRENDY are consistent with those generated by NJOY or the calculation results by MCNP.