ENDL Overview

ENDL Overview

The Evaluated Nuclear Data Library (ENDL) is a collection of flat file databases which represent

Binary collision cross sections
Particle number multiplicity distributions of reaction products
Energy and angular distributions of reaction products
Derived calculational constants

in a pointwise format as a function of incident projectile energy on stationary targets. To view ENDL data, visit the NADS graphical data plotter.

ASC and legacy simulation codes at LLNL use databases of calculational constants derived or copied from ENDL. Examples of derived calculational constants are group averaged cross sections and transfer matrices for P_N transport codes. There are four different derived databases used at LLNL:

TDF – Thermonuclear Reaction Rates and Secondary Particle Spectra
NDF – Nuclear Data for Deterministic Transport
MCF – Monte Carlo All Particle Method (i.e. Monte Carlo Collision Package)
Legacy TART – Group averaged and multi-band data tables for TART Monte Carlo transport routines

Contact:

Documentation:

None available yet.

UCRL-WEB-200195

mckinley9@llnl.gov

Updated: October 9, 2003