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, July 2013 (1st RCM) - March 2015 (2nd RCM) - March 2017 (3rd RCM)

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 not used so far, e.g. data missing in EXFOR: Cross Sections, Neutron Sources Spectra or ²³⁵U(n,f)PFNS.

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-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 example input) and RR_UNC codes with IRDFF 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(n_th,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(n_th,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)

   6. Preparation of the new REAL excercise for the international intercomparison of the modern neutron spectral adjustment codes employing the cross sections from the current version of IRDFF
   - for more information and actual status see REAL-201X (in preparation).

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(n_th,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(n_th,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: ²⁰⁹Bi(n,3-10n),  ⁸⁹Y(n,2-4n) + (n,p),  ⁵⁹Co(n,3-5n),  ⁹³Nb(n,3-4n),  ¹⁹⁷Au(n,3-5n),  ⁵⁴Fe(n,2n),  ¹⁶⁹Tm(n,2-3n),  ¹⁷⁵Lu(n,2-4n),  ¹⁰³Rh(n,4-8n),  ¹³⁹La(n,4-10n), 

• 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,
  at NPI/Rez, p(37MeV)-Li: S.Simakov, P.Bem et al. FED 82(2007)2510.

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:

following IRDFF cross sections were taken 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)

following IRDFF cross sections were accepted in the Medical database of Theurapeutic Radioisotope Production as clinically established: ⁹⁰Zr(n,p)⁹⁰Y and ³²S(n,p)³²P or emerging: ⁶⁴Zn(n,p)⁶⁴Cu

following IRDFF cross sections were accepted by ...