Overview

The new International Reactor Dosimetry and Fusion File (IRDFF-II) addresses neutron dosimetry needs for fission and fusion applications for incident neutron energies from 0 to 60 MeV. The library entries, enumerated in the Table I, include 119 metrology reactions with covariance information and corresponding decay data. The library also includes 4 cover cross sections of B, B-10, Cd and Gd used to support self-shielding corrections, 5 metrology metrics used by the dosimetry community, and 7 cumulative fission products yields. Several reference neutron fields for library validation are also provided. Finally, recommended radionuclide masses and elemental abundances to be used for dosimetry applications are also included.
The dosimetry library can be used in a broad range of applications from lifetime management and assessments of nuclear power reactors to other neutron metrology applications such as boron neutron capture therapy, therapeutic use of medical isotopes, nuclear physics measurements, and reactor safety applications. Library evaluations are based mainly on comprehensive experimental data, therefore the reaction library also represents an ideal benchmark collection for validation and improvement of theoretical nuclear reaction modelling.

IRDFF-II cross-section and decay data files (updated on December 9,2020)

The neutron dosimetry cross-section data for individual monitor and cover materials and corresponding decay data files can be accessed from the ENDF interface of the IAEA-NDS web page under the "Special Libraries" tab as IRDFF-II (Dosimetry), IAEA 2019, cross sections. All individual reactions used in the IRDFF-II evaluations of reactions on natural targets are also available as IRDFF-II (auxiliary files), IAEA 2019. Q-values of the MF40 covariances for reactions on natural targets were corrected for consistency on December 9 update.
Complete cross-section library in various forms can be downloaded from the links below:

• (updated) IRDFF-II dosimetry cross sections in 4-column format (Energy in eV, cross section in barn, absol. uncert. in barn, rel. uncert. in %, compressed).
• (updated) IRDFF-II dosimetry cross sections in pointwise ENDF-6 format (compressed).
• (updated) IRDFF-II dosimetry group cross sections in ENDF-6 format, 640-groups (compressed).
• (updated) IRDFF-II dosimetry group cross sections in ENDF-6 format, extending from 1.E-5 eV up to 60 MeV, 725-groups (compressed).
• (updated) IRDFF-II dosimetry cross sections in ACE format (compressed).
  A summary of contents is given in the ACE list file. Activation cross sections for the excitation of isomeric states are identified by ACE reaction numbers MT*=(10+LFS)*1000+MT, where LFS is the final state (LFS=0 is the ground state, LFS=1 is the first isomeric state, etc.). Cross sections for the production of radionuclides are identified by ACE reaction numbers MT*=(50+LFS)*1000000+ZA where ZA identifies the product (ZA=1000*Z+A).
  
• Recommended decay data are listed in Tables 7 and 8; these data are consistent with the cross-section evaluations. Latest decay data evaluations in ENDF-6 format are also available; these evaluations are consistent with recommended decay data for all nuclides but the half-life of Na-22. For more details please refer to the full IRDFF-II library documentation.

Metrology (dosimetry) metrics: Damage cross sections (updated on June 19,2020)

A dosimetry metric is the result of folding a calculated dosimetry-related energy-dependent response function with the incident neutron energy-dependent fluence. Only dosimetry metrics that have been endorsed by a national nuclear regulator and/or by an international standards organization have been included. Table I lists the six dosimetry metrics included in the IRDFF-II library (note the two metrics listed for JEFF-3.3):
Metrology metrics can be downloaded from the links below:

• natSi(n,1-MeV) ASTM E722-14 1-MeV(Si)-equivalent response function, ENDF-6 format (tabulated file=ASTM E722-14 standard, 1.E-10 to 20 MeV, README).
• natFe(n,X)dpa ASTM E693-17 displacement cross sections, NRT model, ENDF-6 format (tabulated file= ASTM E693-17 standard = IRDF-2002 MT900, 1.E-10 to 20 MeV, README).
• natFe(n,X)dpa EURATOM displacement cross sections NRT model, ENDF-6 format (tabulated file= IRDF-2002 library MT901, 1.E-10 to 20 MeV).
• natFe(n,X)dpa JEFF-3.3 damage cross sections, NRT model, ENDF-6 format (tabulated file = JEFF-3.3, 2017, ENDF-6 file, MT901, 1.E-10 to 200 MeV).
• natFe(n,X)dpa JEFF-3.3 damage cross sections, Arc model, ENDF-6 format (tabulated file = JEFF-3.3, 2017, ENDF-6 file, MT900; 1.E-10 to 200 MeV).
• GaAs(n,1-MeV) ASTM E722-14 1-MeV(GaAs)-equivalent response function, ENDF-6 format (tabulated file = ASTM E722-14 standard, 1.E-10 to 20 MeV, README).

Fission product yield data

The recommended set of cumulative fission product yield data relevant for neutron dosimetry are taken from the JEFF-33 and ENDF/B-VIII.0 libraries. Since the yields are given for selected isotopes only, no ENDF-formatted data are given.

IRDFF-II neutron reference spectra

Codes

The following codes may be useful in connection with the IRDFF library:
• RR_UNC Calculates uncertainties in reaction rates and cross sections.
• COVEIG Calculates eigenvalues of covariance matrices in an ENDF file.
• CODES for radiation damage calculations and neutron spectrum adjustment.