Dosimetry
dosimetry sublibrary: FENDL/DS-2.0

FENDL-2.0:
Activation
Decay
Dosimetry
Fusion
Transport
Benchmarks

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).

The FENDL/DS-2.0 contains:

• PROCESSED - IRDF-90 data. Neutron cross-section data processed into SAND-II 640 multigroup structure.

  xsections - recommended neutron cross-section data to be used for reactor neutron dosimetry by foil activation.

  damage - recommended values for radiation damage cross-sections.

  spectra - recommended values for benchmark neutron spectra.

  writeup - write-up of IRDF-90 Version 2

• POINTWISE - data for 50 neutron induced reactions, for which representation by the SAND-II 640 multigroup structure may lead to inaccuracy.