Testing and Improving the International Reactor Dosimetry and Fusion File (IRDFF)


Coordinated Research Project (CRP) - approved on 30 October 2012, CRP code F41031
duration period: 4 years, mid 2013 (1st RCM) - 2017


Motivation/Purposes

   The Nuclear Data Section of IAEA, in accordance with the recommendation of the International Nuclear Data Committee (INDC(NDS)-0619), has initiated a Coordinated Research Project (CRP) with the main goal to test, validate and improve the IRDFF library.
   The International Reactor Dosimetry and Fusion File (IRDFF) (for more information see IRDFF release page) is an extension of the International Reactor Dosimetry File (IRDF-2002) to cover fission, fusion and accelerator driven applications.
   This extension includes 4 new reactions (67Zn(n,p)67Cu, 113In(n,n')113mIn, 169Tm(n,3n)167Tm, 209Bi(n,3n)207Bi), 32 updated evaluations and increases the end-point energy of the library from 20 to 60 MeV.
   The energy extrapolation has been made in a formal way by using the TENDL-2010 cross sections (and covariance matrices) after matching to the IRDF-2002 cross section values at the extension point (typically 20 MeV).
   In despite of the current IRDFF end-point energy 60 MeV (however already with several exceptions: 186W(n,γ) - 150 MeV; 31P(n,p)31Si, 92Mo(n,p)92mNb, 235,238U(n,f) and (n,γ), 239Pu(n,f) - 200 MeV), CRP will strive to evaluate and eventually add to the library the high threshold reactions with cross section peaks located between 20 and 100 MeV to meet requirements of the higher energy nuclear installations such as ADS. Often this is a set of several reactions (n,3-6n) on one of such isotope: 197Au, 169Tm, 209Bi, 59Co, 63Cu, 89Y, 93Nb. The set of such reactions are very convenient for neutron fluence monitoring and spectrum unfolding at the high energy accelerator driven neutron sources.
   The CRP will strive to stimulate new energy integrated (integral) and point energy (differential) cross section measurements and collect all other experimental information suitable for validation but which has not been used before: Cross Sections and Neutron Sources Spectra which are still missed in EXFOR (please report us about such data).

Main CRP Output

    Improved, tested and validated International Reactor Dosimetry and Fusion File (IRDFF) with proper decay data and documentation.

Content of IRDFF

   1.1. IRDFF (actual version 1.03 since March 2014) contains 76 dosimetry and 3 absorption (cover materials) cross sections and also includes total and elastic cross sections for materials or reactions with resonance structure for evaluation of the self-shielding effect.
    Cross sections and uncertainties:
    - List of reactions
    - MAT, Lab, Date, Authors, ENDF files ... of evaluations
    - Retrieval page (use "Quick plot" option for plotting IRDFF or "Universal plot" to include EXFOR data,
                            it also allows automatic correction of cross sections and uncertainties, select example #5)
    - Plots of differences between IRDFF-1.03 and IRDFF-1.02
    - Plots of differences between IRDFF-1.03 and IRDF-2002

    IRDFF-1.03 includes updates of 54Fe(n,p)54Mn, 58Ni(n,2n)57Ni, 93Nb(n,γ)94Nb, 115In(n,γ)116mIn and one new 238U(n,2n)237U reactions (for more information see INDC(NDS)-0657 or IRDFF release page).
   Full list of additional reactions recommended for inclusion in IRDFF - see Summary Report of RCM-1 INDC(NDS)-0639

   Energy group structure (above 20 MeV recommended by 1st RCM): "640 groups below 20 MeV, 0.5 MeV steps from 20 to 30 MeV, 1 MeV steps from 30 to 40 MeV, and 2 MeV steps from 40 to 100 MeV,and 5 MeV above 100 MeV"

   IRDFF-1.03 processed by NJOY-99.396 in ACE format for use with MCNP: cross sections (300K) files Type1 ASCII or Type2 binary and corresponding xsdir directories: Type1 or Type2.
   Prepared files were verified by the MCNP5 calculations of the averaged cross sections in the Cf-252 spectrum (IRDFF spectra: ENDF format) against RR_UNC code, see results of comparison for 252Cf(s.f.) and 235U(nth,f) spectra.

   1.2. IRDFF (previous version 1.02) contains 75 dosimetry and 3 absorption (cover materials) cross sections
    - MAT, Lab, Date, Authors, ENDF files ... of evaluations
    - Retrieval page
    IRDFF-1.02 processed by NJOY-99.396 in ACE format for use with MCNP: cross sections (300K) files Type1 ASCII or Type2 binary and corresponding xsdir directories: Type1 or Type2.

   1.3. Following IRDFF dosimetry cross sections were incorporated from Standards: 6Li(n,t)4He (below 2.8 MeV), 10B(n,α0)7Li and 10B(n,α1)7Li (below 1.0 MeV), 197Au(n,γ)198Au (from 4.8 keV to 2.6 MeV), 235U(n,f) (from 25 keV to 200 MeV) and 238U(n,f) (from 1 MeV to 200 MeV).

   2. Standard and Reference Spectra (interactive plots and numerical data in different formats):
    - 252Cf(s.f.): interactive plot for Spectrum/Covariance; original ENDF file and processed in 640 groups
    - 235U(nth,f): interactive plot for Spectrum/Covariance; original ENDF file

   3. Reference Decay database for the dosimetry reaction residuals:
   - currently includes 82+6 Isotopes and Isomers: Actual status and needs for updating
   - actual converted IRDFF Decay Library, IRDFF2014.ENDF, in ENDF format (the source is ENSDF as Mar 2013 or new Chechev evaluations)
   - previous converted IRDFF Decay Library, irdf2012.endf, in ENDF format (the source was ENSDF evaluations as Dec 2011)
   - Abundances of the Stable Isotopes: see Table

   4. Fission Yields data for fissile isotopes used in dosimetry (from JEFF-3.1 as a candidate reference source):
    - Th-232, U-235, U-238, Np-237, Pu-239, Am-241,

Energy domains, typical fields, facilities and data status

Actual collected Proposals for HPRL

In the case of validation of dosimetry cross sections, that should have high accuracy as references, specific attention should be paid to:

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