1. ENDF/B-VII.0: Next Generation Evaluated Nuclear Data Library for Nuclear Science and Technology;Chadwick;Nuclear Data Sheets,2006
2. ENDF/B-VII.1 Nuclear Data for Science and Technology: Cross Sections, Covariances, Fission Product Yields and Decay Data;Chadwick;Nuclear Data Sheets,2011
3. ENDF/A, National Nuclear Data Center, Brookhaven National Laboratory. See: http://www.nndc.bnl.gov.
4. Proceedings of the Eighth International Topical Meeting on Nuclear Applications and Utilization of Accelerators (AccAppʼ07),2007
5. D.L. Smith, “A Unified Monte Carlo Approach to Fast Neutron Cross Section Data Evaluation”, Report ANL/NDM-166, Argonne National Laboratory (2008). See also: D.L. Smith, Ref. [4], p. 736.