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). 

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
[PARENTDIR] Parent Directory - [DIR] fendlds/ 1996-04-14 22:00 - [DIR] pointwise/ 1997-12-16 23:00 -