CODE	NUCLIDE	ORIGINAL DATA TAKEN FROM	MODIFICATION	ADDITIONAL FILE MODIFICATIONS	ORIGINAL DATA TAKEN FROM	MODIFICATION	ADDITIONAL FILE MODIFICATIONS	COMMENTS
1006000	006_C_000	ENDF/B-VII	Modification to MF3 MT102 Capture cross sections up to 20 were replaced by the evaluation in JENDL-4. The total and non-elastic cross sections were adjusted by adding the differences in the capture. AFCI MF33 covariance data moved into these materials where there had been no covariance data. Covariances were estimated at LANL based on a brief analysis of experimental data and their agreement with the ENDF/B-VII.0 central values. An exception is elastic scattering which serves as neutron cross section standard from 10 eV up to 1.8 MeV, with uncertainties below 1%.		ENDF/B-VI MOD 3	Evaluation, June 1996, M.B. Chadwick and P.G. Young (LANL) INCIDENT NEUTRON ENERGIES < 20 MeV Below 20 MeV the evaluation is based completely on the ENDF/B- VI.1 (Release 1) evaluation by Fu [Fu90]. The following minor modifications were made to the ENDF/B-VI.1 evaluation: 1. The energy range from En = 20 MeV to 32 MeV was replaced by the LANL evaluation (described below); 2. The elastic, nonelastic, and total cross sections from 19 to 20 MeV were varied to join smoothly with the higher energy values at 20 MeV. 3. The new flag, LTT=3, is used in MF=4,MT=2 to indicate that Legendre polynomials are used below 20 MeV and probability tabulations at higher energies. A small discontinuity exists for MF=4,MT=2 at 20 MeV where the two different representations join. The higher energy evaluation utilizes a tabulation in order to overcome the inaccuracies caused by the ENDF-6 limitation of 20 for the maximum number of Legendre coefficients. INCIDENT NEUTRON ENERGIES > 20 MeV (12C analysis) The evaluation above 20 MeV utilizes MF=6, MT=5 to represent all reaction data. Production cross sections and emission spectra are given for neutrons, protons, deuterons, alpha particles, gamma rays, and all residual nuclides produced (A>5) in the reaction chains.