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.

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