ACE-FRENDY-CBZ: a new neutronics analysis sequence using multi-group neutron transport calculations

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

(2021) Akio Yamamoto et al.   JOURNAL OF NUCLEAR SCIENCE AND TECHNOLOGY

Numerical benchmark problem of solid-moderated enriched-uranium-loaded core at Kyoto university critical assembly

(2019) Go Chiba et al.   JOURNAL OF NUCLEAR SCIENCE AND TECHNOLOGY

Development of a fast reactor multigroup cross section generation code EXUS-F capable of direct processing of evaluated nuclear data files

(2018) Changhyun Lim et al.   Nuclear Engineering and Technology

Development and verification of a new nuclear data processing system FRENDY

(2017) Kenichi Tada et al.   JOURNAL OF NUCLEAR SCIENCE AND TECHNOLOGY

Impact of Incident Energy Dependence of Prompt Fission Neutron Spectra on Uncertainty Analyses

(2012) Go CHIBA et al.   JOURNAL OF NUCLEAR SCIENCE AND TECHNOLOGY

JENDL-4.0: A New Library for Nuclear Science and Engineering

(2012) Keiichi SHIBATA et al.   JOURNAL OF NUCLEAR SCIENCE AND TECHNOLOGY

Methods for Processing ENDF/B-VII with NJOY

(2010) R.E. MacFarlane et al.   NUCLEAR DATA SHEETS