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
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[DIR]fendlds/2020-03-05 13:17 -  
[DIR]pointwise/1997-12-16 23:00 -