International Reactor Dosimetry and Fusion File
IRDFF v.1.05, 09 October, 2014
(Crosssection data supersede IRDF2002 and all previous versions of IRDFF)
IAEA Coordinator: A.Trkov
IRDFF REFERENCES:
R. Capote, K.I. Zolotarev, V.G. Pronyaev, and A. Trkov, J. ASTM International, Volume 9, Issue 4, April 2012, JAI104119
E.M.Zsolnay, R. Capote, H.K. Nolthenius, and A. Trkov, Technical report INDC(NDS)0616, IAEA, Vienna, 2012.
WARNING The IRDFFv1.05 is superseded by the new IRDFFII library.
Overview
The International Reactor Dosimetry and Fusion File (IRDFFv1.05) is a standardized evaluated cross section library of neutron dosimetry reactions and uncertainty information that supersedes the widely used IRDF2002 library. The new IRDFF library contains cross section data and related decay data for 79 dosimetry reactions, and absorption data for three cover materials B, Cd and Gd used for suppressing the themal neutron flux in the irradiation of some material samples. All reaction cross section data (except cover materials) include stateoftheart covariance information.
The library can be used in a broad range of applications from lifetime management
and assessments of nuclear power reactors and 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 data files (current version version 1.05, 08 October, 2014)
 IRDFF ver. 1.05 dosimetry cross sections in 3column format (E in eV, cross section in barn, absolute uncertainty in barn) (compressed)
 IRDFF ver. 1.05 dosimetry cross sections in pointwise ENDF6 format (compressed)
 IRDFF ver. 1.05 dosimetry cross sections in 640 groups ENDF6 format (compressed)
 IRDFF ver. 1.05 dosimetry cross sections in 725 groups ENDF6 format extending to 60 MeV (compressed)
 IRDFF ver. 1.05 dosimetry cross sections in ACE format (compressed).
A summary of contens is given in the list file. (WARNING: Formatting errors corrected on 6 and 13 January 2015 and 23 May 2017)
 IRDFF decay data library (ENDF6 format, unchanged from ver. 1.02)
 IRDFF decay data library documentation (provided by O. Bersillon, unchanged from ver. 1.02)
 IRDFF recommended isotopic abundances (unchanged from ver. 1.02)
 Listing from COMPLOT and comparison Plots of IRDFF v1.05 vs IRDF2002 cross sections
 Listing from COMPLOT and comparison Plots of IRDFF v1.05 vs IRDFF v1.04 cross sections
 Listing from COMPLOT and comparison Plots of IRDFF v1.05 vs ENDF/BVII.1 cross sections
 MAT = 9228: Prompt fission neutron spectrum for U235(n_th,f) (ENDFB/VII.1)
 MAT = 9861: Prompt fission neutron spectrum for Cf252(sf) (Reich,Mannhart,Englad evaluation for ENDF/BV, carried over to ENDF/BVII.1, interpolation law changed)
 MAT = 9862: Prompt fission neutron spectrum for U235(n_th,) (ENDFB/VII.0, interpolation law changed to log)
 MAT = 9865: Prompt fission neutron spectrum for U235(n_th,f) (JENDL4, interpolation law changed to log)
 MAT = 9901: Maxwellian at T=0.0253 eV (1.E5  0.55 eV)
 MAT = 9902: Epithermal pure 1/E spectrum (0.55 eV  2 MeV)
 MAT = 9903: Maxwellian at T=30 keV (MACS in astrophysics)
 Listing of thermal cross section values and uncertainties from IRDFFv1.05
 Listing and plots of IRDFF cumulative reaction rate integrals in a thermal neutron spectrum. Note that the spectrum is normalised to 2/sqrt(Pi) for comparison with the thermal value (see above). The remaining difference is due to the Westcott factor.
 Listing and plots of IRDFF cumulative reaction rate integrals in a 1/E neutron spectrum in the range 0.55 eV to 2 MeV
 Listing and plots of IRDFF cumulative reaction rate integrals in a 30 keV Maxwellian neutron spectrum
 Listing and plots of IRDFF cumulative reaction rate integrals in Cf252(sf) neutron spectrum
 The P31(n,p) reaction cross sections were updated.
 The Si28(n,p) reaction was added to the list of dosimetry reactions.
 The isomer production cross sections from the In113(n,g) reaction were added.
 The isomer production cross sections from the In115(n,g) reaction were added.
 The 725group data were added; the group structure covers the energy range from 1E5 eV to 60 MeV and can be used with the RR_UNC code to calculate the uncertainties.
 The error of missing ZAP designation of the product nuclides in MF10/MT5 of Ti47, Ti48 and Ti49 was corrected. No other changes to the data were made.
 New evaluations for Fe54(n,2n), In115(n,g)In116m, Nb93(n,g), Ni58(n,2n), U238(n,2n)
 Extension of evaluated Tm169(n,2n) cross sections to 60 MeV
 Change of interpolation flags in the Cf252 spontaneous fission neutron spectrum
 Correction of minor errors
 Tm169(n,2n) updated in the whole energy range
 Added reactions Li6(n,t); F19(n,2n); Ni60(n,p); Cu63(n,a); Cu63(n,2n); Cu65(n,2n); and Zn64(n,p) above 20 MeV
 CHECKR did not report any errors.
 FIZCON found two correlation coefficients in Cr52(n,2n) reaction that exceed unity by no more than 3E4.
 COVEIG checked the eigenvalues of the covariance matrices. The code has some limitations, which are reported in the listing (particularly related to crosscovariances between reactions and materials, but in the bulk of the data that were analysed, only three cases were found where the eigenvalues were very slightly negative, as shown in the summary.
 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.
IRDFF data web retrieval
IRDFF neutron spectra (Version 1.05)
The IRDFF library includes the neutron spectra listed below. The fission spectra taken from evaluated files were converted from MF5,MF35/MT18 to MF3,MF33/MT261 representation, respectively (ENDF file convention).Links to the spectra in 640group and 725group form are available from the shortcuts on the righthandside and include the dummy spectrum 9900, which simulates the thermal point. For 1/v absorbers the cross sections at 0.0253 eV are reconstructed exactly with this spectrum.
Plots of IRDFF cumulative reaction rate integrals in reference neutron spectra (IRDFF v1.03)
 Changes in v1.05:
 Changes in v1.04:
 Changes in v1.03:
 Changes in v1.02:
Verification

File integrity and consistency was checked with the standard ENDF Utility codes and the local COVEIG code
with the following results:
Validation
Preliminary validation of data is described in the IAEA Technical Report INDC(NDS)0616. For additional information follow the link to the new Coordinated Research Project "Testing and Improving the IAEA International Dosimetry Library for Fission and Fusion (IRDFF)".
Codes

The following codes may be useful in connection with the IRDFF library: