FENDL/DS-2.0 DOSIMETRY SUBLIBRARY

This directory contains neutron cross sections to be used for reactor 
neutron dosimetry by foil activation, radiation damage cross-sections, 
and benchmark neutron spectra. This sublibrary is identical to the 
International Reactor Dosimetry File (IRDF-90). 

Directories:
FENDLDS/   - IRDF-90 data. Neutron cross-section data processed into 
             SAND-II 640 multigroup structure.  
POINTWISE/ - pointwise data for 50 neutron induced reactions, for which 
             representation by the SAND-II 640 multigroup structure may 
             lead to inaccuracy. 

page modified: 13.3.1998
Icon  Name                    Last modified      Size  Description
[DIR] Parent Directory - [DIR] fendlds/ 14-Apr-1996 22:00 - [DIR] pointwise/ 16-Dec-1997 23:00 -