[back to index] [experiment]



 1. Name of Experiment:
    FNG Benchmark Experiment on Silicon Carbide (SiC)

 2. Purpose and Phenomena Tested:
    The purpose is to validate the cross sections of Si and of C in the
    European Fusion File (EFF), as the SiC, in the form of ceramic matrix
    (SiC-fiber/SiC), is a candidate structural material for the fusion reactor
    and its development is pursued in the European Fusion Technology Program. 
    The experiment [1] - [4] was performed at the 14 MeV Frascati Neutron
    Generator (FNG) on a monolithic, sintered SiC block. 

 3. Description of Source and Experimental Configuration:
    The 14-MeV d-T Frascati Neutron Generator (FNG, [5]) was the neutron
    source. The angular dependence of the source intensity is presented in
    Table 1 and in Figure 1. The angular dependence of the source energy
    distribution is given in Table 2 and in Figure 2. The geometry of the
    mock-up is outlined in Figure 3. The experimental set-up consisted of
    a block of sintered SiC (45.72 cm x 45.72 cm, 71.12 cm in thickness), 
    located in front of the FNG target, 5.3 cm from the neutron source. The
    SiC block was assembled with a total of 116 bricks. Inside the block, four
    experimental positions at different penetration depths were available to
    locate detectors of various types (activation foils, TLD holders, active
    The average weight density of the used SiC was 3.158 g/cm3, major
    impurities were boron (0.19 % in weight), aluminium (0.79 wt%) and iron
    (0.14 wt%). 

 4. Measurement System:
    The following quantities are measured :

        a - Neutron reaction rates by activation foils
        b - Nuclear heating by thermo-luminescent detectors (GR-200A)

    Four different reactions: 197Au(n,g), 58Ni(n,p), 27Al(n,a) and 93Nb(n,2n)
    were used to derive the neutron flux, from thermal energy up to the fusion
    neutron peak. The reaction rates were measured at four experimental
    positions, 10.41 cm, 25.65 cm, 40.89 cm and 56.13 cm respectively from
    the block surface, using the radiometric techniques based upon the use of
    absolutely calibrated HPGe detectors. The overall contribution to the
    quoted uncertainty comes from the HPGe calibration (±2%), measured activity
    (<±3%) and total neutron yield (±3%). The neutron yield is measured using
    the associated alpha particle method. The experimental results are given in
    Table 3.

    Nuclear heating was measured at 14.99 cm, 30.23 cm, 45.47 cm and 60.71 cm 
    depth from the block surface using high sensitivity GR-200 (LiF:Mg,Cu,P)
    thermoluminescent dosimeters (TLD). Since the TLD technique is a relative
    method, TLDs were calibrated against a Co-60 g-ray source. The calibration
    uncertainty was within ±5%. The calibration was performed using a secondary
    standard. Four TLDs were located in each experimental position and the
    averaged values of the measured TL light was used as experimental data. 
    From this datum, after conversion via the calibration factor, the
    experimental total dose in TLDs was obtained. The total uncertainty is ±7%.
    This uncertainty includes the above mentioned calibration error, the error
    (standard deviation) on the averaged TL light value, the uncertainty on
    the FNG neutron yield (±3%), summed using the quadratic law.
    The experimental results are given in Table 4.

 5. Description of Results and Analysis:
    The experimental results (E) were analysed by the Monte Carlo code MCNP-4C [6]
    using point-wise cross sections derived from FENDL-2.0 [7], EFF-2.4 and also
    with the new evaluation for 28Si included in EFF-3.0. The calculations are
    presented in [1] and [2]. Activation reaction rates where calculated using
    tally f4 of MCNP, the dosimetric reactions needed for the calculation were
    taken from IRDF-90.2 library [8] (mcnp.inp). The calculated reaction rates (C)
    are given in Table 5 together with the statistical uncertainty.

    The gamma dose Dg  and the neutron dose Dn in TLDs was calculated with tally
    f6 of MCNP, using a very detailed model in MCNP of the dosimeters and of their
    plastic containers (mcnp_tld.inp). The dose in TLD was obtained from:

     DTLD = Kn*Dn + Dg

    where Kn is neutron sensitivity of the TL detector to neutrons and is usually
    depending upon the neutron energy. The Kn data were taken from [9] and were
    then weighted over the calculated neutron spectra at each experimental position.
    The resulting  Kn values are reported in Table 7. 
    Comparison between the measured dose in TLD and the same quantity calculated
    with EFF-2.4, EFF-3.0 and FENDL-2.0 nuclear data is also given in Tab. 7.  

    The transport and cross section sensitivity/uncertainty analyses were also
    performed using the deterministic codes DORT, TWODANT and SUSD3D, and are
    presented in [3] and [4]. Analytic uncollided and first collision source
    approach was used in order to mitigate the ray effects.
    Input data for the following codes are included:
    - TRANSX (cross section preparation),
    - GRTUNCL and DORT (uncollided/first collision source and discrete ordinates
      (SN) transport calculation)
    - TWODANT (SN neutron transport using first collision approach).

 6. Special Features:

 7. Author/Organizer:
    Experiment and analysis:
    P. Batistoni, M. Angelone, M. Pillon, L. Petrizzi 
    Centro Ricerche Energie Frascati
    UTS Fusione  
    Via E. Fermi 27
    C.P. 65
    I-00044 Frascati (Rome) 

    Compiler of data for Sinbad:
    P. Batistoni
    Centro Ricerche Energie Frascati
    UTS Fusione  
    Via E. Fermi 27
    C.P. 65
    I-00044 Frascati (Rome) 
    E-mail: batiston at efr406.frascati.enea.it

    Reviewer of compiled data:
    I. Kodeli
    OECD/NEA, 12 bd des Iles, 92130 Issy les Moulineaux, France
    e-mail: ivo.kodeli at oecd.org

    The experiment and the corresponding analysis was performed in the framework of the
    EFDA (European Fusion Development Agreement) Task (TTMN-002-2001).

 8. Availability:

 9. References:

   [1] P. Batistoni, M. Angelone, L. Petrizzi, M. Pillon, Measurements and
       Analysis of Reaction Rates and of Nuclear Heating in SiC, Dec. 2001
   [2] M. Angelone, P. Batistoni, I. Kodeli, L. Petrizzi, M. Pillon,
       “Benchmark analysis of neutronics performances of a SiC block
       irradiated with 14 Mev neutrons”, Fus. Eng. Design 63-64 (2002) 475
   [3] Y. Chen, U. Fischer, I. Kodeli, R. L. Perel, M. Angelone, P. Batistoni,
       L. Petrizzi, K. Seidel, S. Unholzer, “Sensitivity and uncertainty
       analyses of 14 Mev neutron benchmark experiment on Silicon Carbide”,
       22nd Symposium on Fusion Technology, Helsinki, Finland, 9-13 Sept 2002.
   [4] I. Kodeli, Deterministic Transport, Sensitivity and Uncertainty
       Analysis of SiC Benchmark Experiment Using EFF-3 and FENDL-2 Evaluations,
       EFFDOC-818, Dec. 2001.
   [5] M. Martone, M. Angelone, M. Pillon, The 14 MeV Frascati Neutron 
       Generator, Journal of Nuclear Materials 212-215 (1994) 1661-1664;
   [6] Briesmeister, J. F. (Ed.), MCNP - A general Monte Carlo n-particle
       transport code, version 4C, Report LA12625, Los Alamos, September 1999.
   [7] M. Herman, A. B. Pashchenko, Extension and improvement of the FENDL
       library for fusion applications (FENDL-2), Report INDC(NDS)-373, IAEA  
       Vienna, 1997.
   [8] N. P. Kocherov, P. K. McLaughlin, The International Reactor Dosimetry
       File (IRDF-90), Report IAEA-NDS-141, Rev. 2, Oct. 1993.
   [9] J.A. Gibson, The relative tissue-kerma sensitivity of thermoluminescent
       materials to neutrons, Report EUR 10105 EN (1985)

10. Data and Format:

  FILE     NAME        bytes   Content
  ---- ------------    ------   -------
   1   fngsic-a.htm    12,207  This information file 
   2   fngsic-e.htm    30,979  Description of Experiment 
   3   source.for      45,178  FORTRAN subroutine for MCNP source description
   4   mcnp.inp        47,410  3-D model for MCNP-4A calculations of neutron
                               activation reaction rates
   5   mcnp_tld.inp    53,655  3-D model for MCNP-4A calculations of nuclear
   6   trx-sic.inp        522  Input data for TRANSX cross-section preparation
   7   dort-sic.inp     8,085  Input data for GRTUNCL first collision source and
                               DORT transport codes
   8   2dant-si.inp     6,478  Input data for TWODANT transport code
   9   fig1.gif         5,242  Fig. 1: Angular dependence of the source 
  10   fig2.gif         9,505  Fig. 2: Energy/angular dependence of the source 
  11   fig3.gif        64,368  Fig. 3: Geometry of the experimental mock-up 
  12   fig4.gif         9,633  Fig. 4: Geometry of the source
  13   fig5.gif         5,165  Fig. 5: Geometry of TLD detector
  14   enea-sic.pdf   904,911  Reference 1
  15   fed2002.pdf    282,929  Reference 2
  16   soft-22.pdf    236,191  Reference 3
  17   eff-818.pdf    605,072  Reference 4

     File fngsic-e.htm contains the following tables:

     Tab. 1: Angular dependence of the source
     Tab. 2: Angular/energy dependence of the source energy distribution 
     Tab. 3: Measured neutron reaction rates
     Tab. 4: Measured absorbed dose in TLD
     Tab. 5: Calculated reaction rates
     Tab. 6: Comparison between calculated (C) and measured (E) reaction rates
            - C/E ratios
     Tab. 7: Calculated absorbed dose in TLD - C/E ratios

    The figures are included in gif format.

SINBAD Benchmark Generation Date: 4/2003
SINBAD Benchmark Last Update: 4/2003