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. IRDFF (actual version 1.02) contains 75 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 (Table 4 from INDC(NDS)-0616, with noted comments/corrections for the next release)
- MAT, Lab, Date, Authors ... 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-1.02 and IRDF-2002
Following existing IRDFF cross sections will be updated soon: 54Fe(n,p)54Mn, 58Ni(n,2n)57Ni, 93Nb(n,γ)94Nb and 115In(n,γ)116mIn.
Further list of reactions recommended for inclusion in IRDFF, see Summary Report of the 1st RCM INDC(NDS)-0639
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).
Recommended energy group structure for extention above 20 MeV 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"
Processing (LANL + NDS) of IRDFF 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.
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:
- the latest coversion (Oct 2012) of IRDFF Decay Library in ENDF format (the source was ENSDF as Dec 2011)
- currently includes 82 Isotopes and Isomers (see Summary and Needs for updating or extension)
- 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
Reactor driven and spontaneous or induced fission spectra (thermal, fast)
- Spectrum averaged cross sections (SPA) in standard and reference fields
- 252Cf(s.f.): Measured, Calculated with Standard spectrum and C/E Ratio plots
- 235U(nth,f): Measured, Calculated with ENDF/B-VII.1 or JENDL-4.0 spectra and C/E Ratio plots
- Thermal: Experimental σ and RI (Atlas in EXFOR: Z=1-50, Z=51-100 or Z=1-100) and Calculated with Maxwellian (25.3meV) spectrum
- SPA for high threshold reactions not measured yet in 252Cf(s.f.) and 235U(nth,f) fields
- IRDF-2002 collection of standard and reference reactor spectra (however without uncertainties) (Fig. and ENDF data )
- research reactors and critical assemblies (e.g., ICSBEP spectra)
- k0-database for Neutron Activation Analysis
- participants: CEA, IPPE, JSI-Ljubljana, LANL, SNL
Maxwellian Averaged Cross Sections (5-500 keV)
(example - MACS spectra and IRDFF XS)
- p-7Li, p-3H, p-18O ... (thick target) sources
- recommended (n,γ) MACS at 30keV and found C/E for several IRDFF materials (except 109Ag, 235,238U and 232Th) are available in the
Kadonis or EXFOR databases
(can such neutron fields be used to measure other low threshold reactions such as (n,n'), (n,p) or (n,α) to verify the large cross section changes noted near threshold: IRDFF vs. IRDF-2002 ?)
- participants: LENOS, IRMM and facilities under constructions SARAF, FRANZ (?)
Fusion Energies (14 MeV)
(example - status of D-T spectra and IRDFF)
- bare quasi-monoenergetic D-T sources like used for EAF validation
- mixed spectra (e.g., D-T neutrons scattered in the assemblies where spectra could be well charachterised)
- paticipants: FNS, FNG, KIT
Medium Energies (5 - 50 MeV)
(example - status of d-Be spectra and 59Co(n,x))
- quasi-monoenergetic p-7Li sources
- white spectra like d-Be, d-Al...
- participants: Ohio University, PNNL, PTB
High Energies (15/20 - 60/200 MeV)
(here cross sections are very scarce and uncertain, e.g. 59Co(n,3n),
- quasi-monoenergetic p-7Li sources
- white spectra like p(113MeV)-Al/U, ...
- participants: NPI/Rez, RCNP, PTB+iThembaLAB, PNNL
In the case of validation of dosimetry cross sections, that should have high accuracy as references, specific attention should be paid to:
- characterisation of the neutron spectrum together with uncertainties and correlations - usually spectrum is determined by a combination of experimental (TOF, proton recoils, Bonner spheres, foil activation ...) and calculation (Monte-Carlo simulation ...) methods; it is important to quantify the uncertainty and strong energy-energy correlations (for more guidance see ASTM Guide);
- possible strong dosimetry reaction-reaction cross sections correlations - these may result from (i) the use of dosimetry cross sections (sensors) for characterisation of the neutron spectrum that is later used to validate another dosimetry cross sections or (ii) from the dosimetry cross section evaluation in joint analysis with other (standard) cross sections; Often experimentalists experience difficulties in construction of covariance matrices from uncertainties. As a help NDS has developed EXFOR format to store (un)correlated uncertainties and on-line tool to construct covarience martices from partial uncerttainties (see example);
- establishing of the reference decay data for the dosimetry reaction residuals for consistent use in evaluation and applications:
- processing of the IRDFF cross sections and uncertainties in formats that can be used by different codes
- when neutron environment is simulated (by MCNP) the dosimetry cross section uncertainty should be propagated in the observed activity