INTERNATIONAL REACTOR DOSIMETRY FILE: IRDF-2002

A Data Development Project of the IAEA Nuclear Data Section

New

A library in ACE-dosimetry format for the MCNP family of codes has been generated from the pointwise IRDF-2002 data. The ACE reaction MT* numbers are related to the ENDF MT numbers as
MT* = MT +1000*(10+LFS) where LFS is the metastable state designator of the reaction product.

Objectives

The objective of the project was to prepare and distribute a standardised, updated and benchmarked evaluated cross section library of neutron dosimetry reactions with uncertainty information (IRDF-2002) for use in lifetime management assessments of nuclear power reactors and other applications.

Within the scope of the project the following tasks were considered:

  • Intercomparison of reactor dosimetry cross-section data and their uncertainties in various libraries including IRDF-90.2, JENDL/D-99 and RRDF-98.


  • Reaction rates in a standard neutron field were compared. Select best data based on the above comparison of data for IRDF-2002.


  • Evaluate and test new reaction cross sections requested by the reactor dosimetry community for extension of the database. Include evaluated decay radiation characteristics in the files.


  • Comparison of experimental and calculated fission and thermal spectrum averaged cross sections.
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    Data files and codes

    The following data files and code packages are available:

  • Damage cross sections

  • Decay data

  • Standard spectra

  • Dosimetry cross sections in pointwise ENDF-6 format (compressed)

  • Dosimetry cross sections in groupwise ENDF-6 format

  • Dosimetry cross sections in ACE format for the MCNP code (compressed)

  • Dosimetry cross sections in metrology format

  • Codes for damage parameter and spectral adjustment calculations

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