A Nuclear Cross Section Data Handbook

and angular distributions of the induced photons are also carried on these files;
however, not all neutron cross-section sets include photon production data.
Each neutron cross-sect ion set is generated from an evaluated data set. An
evaluated set of cross sections is produced by analyzing experimentally measured
cross sections and combining the data with the predictions of nuclear model calculations
in an attempt to extract the most accurate cross-section information. In
an evaluated data set no ambiguity is allowed; this simply means that a decision
has been reached on what the cross section for each reaction should be and what
the secondary energy and angular distributions should be. The majority of the
data presented in this Handbook comes from Version V of the American national
ENDF/B (Evaluated Nuclear Data File/B) system. 2$3These cross sections are supplemented
by evaluations from two other sources: the Lawrence Livermore National
Laboratory’s Evaluated Nuclear Data Library (ENDL)4 and evaluations from the
Los Alamos Applied Nuclear Science Group T-2. Older evaluations accessible from
other sources have now essentially been phssed out.
From any one evaluated data set, several sets of cross sections can be generated.
For the Monte Carlo code MCNP the evaluated data are processed into ACE format
(ACE is an acronym for —— A Compact — ENDF) described in Ref. 1. Cross sections
in ACE format are given and used as cent inuous-energy functions, For use with
discrete ordinates transport codes like 0NEDANT5 and TWODANTG, muhigroup
scattering matrices must be calculated from the evaluated data. The format of the
multigroup data is described in Ref. 7. These multigroup scattering matrices for
the SN method are used as the source from which multigroup data as described
in Ref. 8 are generated for input to MCNP when it is run with the multigroup
option. At Los Alamos we have cross sections in both pointwise and multigroup
form available from the same evaluated data source; this makes possible meaningful
comparisons between the cent inuous-energy Monte Carlo, multigroup Monte Carlo,
and SN methods.
In ACE format the cross sections for all reactions are provided on a single
energy grid. The grid is sufficiently dense that linear-linear interpolat ion between
points reproduces the evaluated cross sections within a specified tolerance that is
generally one percent or less. All cross-section tabulations given in the evaluated
data with semi-log or log-log interpolation schemes are linearized. Depending on
the linearization tolerance, but primarily on the number of unresolved resonances
and the resonance reconstruction tolerance (generally 0.5$%0or better), the resulting
energy grid may contain as few as 250 points (e. g., H-1) or as many as 22,500
points (e. g., Au-197). The “original” ACE-format libraries - those most closely
representing the intentions of the evaluators - can be very long files. “Thinned” files
were prepared for the same evaluations in such a way as to preserve the flat-weighted
integral of the total cross section to within 0.59’0. There are also discrete-reaction
cross-section files in which the pointwise reaction cross sections have been averaged
over 262 energy groups using a flat weight function. Thus, for Monte Carlo cross
sections it is possible to generate many different sets dependent on the linearization,
resonance-reconstruction, and thinning tolerances, to say nothing of the range of
temperatures to which the cross sections can be broadened. With multigroup cross
2