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 not used so far, e.g. data missing in EXFOR: Cross Sections, Neutron Sources Spectra or 235U(n,f)PFNS.
Main CRP OutputImproved, tested and validated International Reactor Dosimetry and Fusion File (IRDFF) with proper decay data and documentation.
Content of IRDFF1. IRDFF-1.05 (actual version since Oct 2014) contains 79 dosimetry and 3 absorption (cover materials) cross sections. It also includes total and elastic cross sections for materials or reactions with resonance structure for evaluation of the self-shielding effect.
IRDFF-1.05 cp. to IRDFF-1.04 has 3 new 28Si(n,p)28Al, 29Si(n,x)28Al, 113In(n,g)114mIn and update of 31P(n,p)31Si reactions (see K.Zolotarev et al. INDC(NDS)-0668)
Cross sections and uncertainties:
- List of reactions
- MAT, Lab, Date, Authors, ENDF zip-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-1.05 and IRDFF-1.04
- Plots of differences between IRDFF-1.05 and IRDF-2002
- Original IRDFF evaluated files without extending to 60 MeV by TENDL
- 209Bi(n,xn), x=3-10, E<100 MeV preliminary evaluation by V.Pronyaev
IRDFF-1.05 processed by NJOY-2012.42 (ACE files for 6Li(n,a) and 10B(n,a) were corrected) in ACE format: cross sections (300K) Type1 ASCII or Type2 binary and xsdir Type1 or Type2 (the order of Temperature is corrected as E-08).
ACE files were verified by comparison of SPA calculated by MCNP5 (see exapmle input) and RR_UNC codes with IRDFF.ENDF spectra: results for 252Cf(s.f.) and thermal Maxwellian spectra.
IRDFF-1.05 Cross-Sections and Covariences in the NJOY Plots
Previous IRDFF cros-sections versions:
IRDFF-1.04 contained 76 dosimetry and 3 absorption (cover materials) cross sections. It has minor format corrections cp. IRDFF-1.03.
IRDFF-1.04 processed by NJOY-2012.32 in ACE format: cross sections (300K) Type1 ASCII or Type2 binary and xsdir Type1 or Type2.
ACE files were verified by comparison of SPA calculated by MCNP5 and RR_UNC codes in IRDFF.ENDF spectra), see results for 252Cf(s.f.) and 235U(nth,f) spectra.
IRDFF-1.03 includes updates of 54Fe(n,p)54Mn, 58Ni(n,2n)57Ni, 93Nb(n,γ)94Nb, 115In(n,γ)116mIn and new 238U(n,2n)237U reaction (see K.Zolotarev et al. INDC(NDS)-0657).
IRDFF-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: cross sections (300K) Type1 ASCII or Type2 binary and xsdir Type1 or Type2.
2. Reference Decay data for reaction residuals and Isotope Abundancies:
- actual (Oct 2014) version includes 82 + new 6 Isotopes/Isomers: Actual status and Needs for updating
- actual IRDFF Decay Library converted in ENDF format: IRDFF2015.ENDF (source is ENSDF as March 2014 or new DDEP-Chechev evaluations)
- previous (2012) IRDFF Decay Library in ENDF format: irdf2012.endf (source was ENSDF evaluations as Dec 2011)
- Abundances of the Stable Isotopes: Table
3. 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
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
5. Photo-induced Reactions which produce the same residual isotope or fission product as neutrons do:
- (g,n) vs (n,2n): cross sections and contributions in the n-g mixed field, e.g. 238U, 23Na
- (g,f)FP vs (n,f)FP: cross sections, their contributions in the n-g mixed field
- Photonuclear Reaction Libraries: IAEA or others
- Photo-Induced Fission Product Yields: no Evaluations (?), only Measurements (? - see PhysRev C91(2014)034603, Eur.Phys.J. A51(2015):150)
(see IAEA CRP on Photonuclear Data)
IRDFF: Needs for measurements, updates or new evaluations, data formats ...
- Proposals for new measurements for IRDFF community and HPRL: Reactions to Measure
- The list of reactions recommended for updating or new evaluation and inclusion in IRDFF : Reactions to Update/Evaluate
- Energy group structure 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".
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) (Spectra and C/E and ENDF formated data )
- research reactors and critical assemblies (e.g., ICSBEP collection: Spectra and SPA and Indices)
- 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 experimental (n,γ) MACS at 30keV (and C/E) for IRDFF reactions (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-1.03 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)
(where cross sections are still scarce and uncertain:
89Y(n,2-4n) + (n,p),
see also this).
- quasi-monoenergetic p-7Li sources
- white spectra like p(113MeV)-Al/U, ...
- participants: NPI/Rez, RCNP, PTB+iThembaLAB, PNNL
- examples of (n,xn) reactions use for unfolding of Spallation neutron spectra up to 150MeV:
at AGS/Brookhaven, p(24GeV)-Hg: Y.Kasugai et al. ASTM STP 1398, 2001, 207 and F.Maekawa et al. NSE 150(2005)99,
at SNS/OakRidge, p(2GeV)-Hg: N.Luciano' M.S.Thesis (2012),
at FZK/Karlsruhe, d(55MeV)-Li: F.Maekawa et al. ASTM STP 1398, 2001, p.417 and INDC(JPN)-0185/U, 2000, p.226.
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
Interaction of IRDFF with other neutron reaction data libraries: