[en] We provide a summary report that accompanies the LA-UR-12-25154 oral presentation made at the October 2012 IAEA Technical Meeting on 'Primary Radiation Damage: From Nuclear Reactions to Point Defects' held in Vienna, Austria. The NJOY Nuclear Data Processing System is used worldwide to create application specific nuclear data libraries from Evaluated Nuclear Data File (ENDF) formatted files. In this summary report we briefly review the ENDF system, then discuss use of NJOY to calculate 'radiation damage'. ENDF is the United States' national nuclear data file. The ENDF system was initially developed in the 1960s, and currently is in its seventh generation. Candidate data files are reviewed by a standing committee known as the Cross Section Evaluation Working Group (CSEWG) which advises Brookhaven National Laboratory's (BNL) National Nuclear Data Center (NNDC) as to ENDF's recommended content. This file is formally known as ENDF/B. A lesser known file, ENDF/A, contains files that contain candidate evaluations for a future ENDF/B release, or partial evaluations that remain under development. As a product of the 1960's, the ENDF format is product constrained by the limits of computer card technology. The data (resonance parameters, cross sections, angular distributions, secondary distributions, and more) are defined to fit within a fixed 80 character per record format. Specifically, numerical data and ascii text strings occur in the first 66 columns and the final 14 columns contain a set of four control integers; matn(i4), mf(i2), mt(i3) and ns(i5). Matn, or material number, is a flag associated with a specific nuclide. Mf, or file number, is used to identify a specific type of data. The 66 column data region consists of 6 eleven column fields containing a mix of real and integer data. Most relevant to this topic are the data given in mf=3 (neutron cross sections), mf=4 (outgoing neutron angular distributions), mf=5 (secondary neutron energy distributions), and mf=6 for coupled energy-angle distributions for multiple emitted particles. Mt, or section number, identifies a specific reaction. Examples are mt = 1 for the total cross section, mt = 2 for elastic scattering, mt's 51 to 89 for discrete level inelastic scattering, and mt's 102 through 107 for the (n,γ), (n,p), (n,d), (n,t), (n,³He) and (n,α) reactions, respectively. (author)