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SINBAD ABSTRACT NEA-1553/58

FNG-ITER STREAMING EXPERIMENT



 1. Name of Experiment:
    ------------------
    FNG-ITER NEUTRON STREAMING EXPERIMENT (1997-1998)   

 2. Purpose and Phenomena Tested:
    ----------------------------
    Determination of neutron reaction rates and of nuclear heating in a
    neutronic  mock-up of the International Thermonuclear Experimental
    Reactor (ITER) shielding system irradiated with 14-MeV neutrons, in
    presence of a streaming path.

 3. Description of Source and Experimental Configuration:
    ----------------------------------------------------
    The 14-MeV d-T Frascati Neutron Generator (FNG, [1]) was the neutron
    source. The angular dependence of the source intensity is presented in
    Figure 1. The angular dependence of the source energy distribution is
    shown in Figure 2.
    The x-y view of the geometry of the mock-up is outlined in Figure 3. It
    consists of a combination of slabs made of the water equivalent material 
    Perspex and the stainless steel AISI-316 (simulating shield-blanket and
    vacuum vessel) and has a front cross-section area of 100 cm x 100 cm. The
    total thickness of the assembly is 94.26 cm including a 1 cm thick Cu 
    layer in front (simulating first wall). The assembly is provided with a
    channel with high aspect ratio (inner diameter a=28 mm, length l = 39.07 cm, 
    channel wall thickness t = 1 mm stainless steel AISI316 ), (see Fig.3). 
    At the end of the channel, a cavity is also realised. A parallelepipedal
    stainless steel box fitting exactly the cavity is inserted to locate the 
    detectors in their positions. The inner size of the box is 52 mm (z) x
    148 mm (x) x 48 mm (in the deuterium beam direction-y). The cavity is
    symmetrically located with respect to the channel axis. 

    Behind this assembly a block of Cu and SS316 plates was arranged (simulating
    the coils for the toroidal magnetic field of the TOKAMAK; dimensions: depth 
    30.8 cm, area 47 cm x 47 cm). The rear part of the assembly was surrounded
    with a polythene shield covering also the last 30 cm of the Perspex/AISI316 
    block in order to reduce room-return background.
        
    The following quantities are measured :

    a - Measurements in the channel : 
    Neutron reaction rates by activation foils to monitor the flux gradient 
    in the channel

    b - Measurements in the cavity : 
    Neutron reaction rates by activation foils

    c - Measurements behind the channel : 
    Neutron reaction rates by activation foils
    Nuclear heating

    d - Measurements in the SC magnet : Nuclear heating

    The detectors were placed on the axis of the d-beam of the neutron generator.
    All  measurements (a,b,c and d) were performed with the neutron source in 
    axis with the channel/cavity, at 5.3 cm distance from the shielding block 
    surface (ON-AXIS set-up). The activation foils measurements in the channel 
    and in the cavity (a and b) were performed  also with the neutron source 
    shifted with respect to the channel axis to simulate the effect of the extended 
    neutron source from the plasma (OFF-AXIS set-up). The source lateral shift is 
    5.3 cm, i.e. equal to the distance between the target and  the mock-up
    surface. In this case the channel mouth is located at an angle of p/4 with
    respect to beam direction (see Fig.4). 
    
 4. Measurement System:
    ------------------
    The following activation reactions were selected to measure the neutron flux:

    Reaction	  Effective Threshold      
                        (MeV)                         
    93Nb(n,2n)92Nb	 10.8	       
    27Al(n,a)24Na	  8.5	     	 
    58Ni(n,p)58Co	  2.9	      	  
    197Au(n,g)198Au	   -	   	  

    All foils had a diameter of 18 mm. 1-3 mm thick foils were used for Nb, Al and
    Ni detectors (1 mm up to the depth of 38.65 cm and in the cavity, 2 mm behind
    the cavity between 46.35 and 73.90 cm, and 3 mm in the last 3 positions). For
    197Au(n,g)198Au reaction 0.05 mm thick foils were used.

    Inside the channel the foils were located in axis with the channel, at y = 0.25, 
    12.95, 25.95, 38.65 cm  (foil centers) from the shielding block surface. Eleven 
    activation foils were located inside the cavity to monitor the neutron flux 
    gradient and the effect of the void channel. The foils were located in the 
    positions shown in Fig. 4, between y = 39.12 cm and y = 43.82 mm depth (foils
    centers). Eight activation foils are located behind the cavity, at depths:
    y = 46.35, 53.30, 60.05, 66.90, 73.90, 80.60, 87.25, 91.65 cm (foil centers).
     
    Nuclear heating was measured in the shielding assembly behind the      
    channel/cavity, in the same positions as for activation foils, and inside the 
    superconducting magnet, using CaF2:Tm thermo-luminescent detectors (TLD-300) of 
    size 0.32 x 0.32 x 0.09 cm3 provided by Harshaw Company. The measured absorbed 
    dose in TLD-300, Q-TLD (E) is obtained from the peak 3 response in the glow 
    curve relative to Co-60 calibration. The air kerma was converted into absorbed
    dose in TLD-300 using the photon energy attenuation coefficient from [2].
    
    
 5. Description of Results and Analysis:
    -----------------------------------
    Measured neutron reaction rates are given in Tables 5-7. They  were compared
    with the same quantities calculated with the  same tools used in the ITER
    design, i.e. MCNP 4A/B [10] code and FENDL-1/FENDL-2 [11,12] nuclear data
    libraries. Analysis was performed also using the EFF-3 [13] library. Dosimetric 
    cross sections were taken from IRDF-90 [14] to compute the nutron reaction
    rates in the MCNP run. The geometry model for MCNP-4B used in the calculation
    of reaction rates, including neutron source backing and the experimental
    environment (walls, floor, racks, ...) is given in mcnpfoil.inp.
    C/E values were provided for all measured reaction rates, they are reported 
    in Tables 9-11.
    
    Measured absorbed dose in TLD-300 is given in Table 8. It is compared 
    with the same quantity calculated with  MCNP code and FENDL-1/FENDL-2, EFF-3.1 
    nuclear data libraries. 
    The geometrical model for MCNP used in the calculation of nuclear heating, 
    is given in mcnp_nh.inp. The comparison with TLD measurements requires, however, 
    the calculation of the absorbed dose in TLD, taking into account 
    the effects of the electron transport at the interface between the TLD and the 
    surrounding material. The absorbed dose in TLD (QTLD) is related to total dose 
    in material Q according to the following expressions:
    
                          Q = Qn+Qp
                          (QTLD ) =  Cp Qp + Ce Cn Qn     (1)
    
    where QTLD is the absorbed dose in TLD, 
    Qn, Qp is the absorbed dose due to neutrons, photons in material (i.e. steel or
           copper), 
    Cn is the ratio of the TLD/materialabsorbed neutron dose,
    Cp is the ratio of the TLD/material absorbed photon dose,
    Ce is the TLD neutron dose efficiency with respect to the photon dose efficiency, 
    taken from published data and weighted over the neutron spectra. 
    
    Simplified models have been adopted representing a smaller region limited to the 
    TLD volume and to a sufficient amount of surrounding material, to calculate Qn, Qp, 
    Cn and Cp. A neutron or photon surface source is applied to the boundary of the 
    limited region in the simplified models, with energy spectra recorded from the 
    mcnp_nh.inp run. These models are given in mcnp_hss.inp, mcnp_hcu.inp and
    mcnp_tld.inp for steel, copper and TLD material respectively. 
    
    The Cn, Cp, Ce factors and Qn, Qp calculated using FENDL-1, FENDL-2 and EFF-3.1  
    are used to derive the calculated dose in TLD according to Eq.1. For all calculated 
    absorbed doses the fractional standard deviation is about 5% or less. C/E values 
    are given in Tables 12.
    
    The calculations performed using the 2D discrete ordinates transport method is
    described in [5] and [15]. The input data are also provided, including inputs for
    the codes TRANSX, GIP, GRTUNCL and DORT.

 6. Special Features:
    ----------------
    None
    
 7. Author/Organizer:
    ----------------
    Experiment and analysis:
    P. Batistoni, M. Angelone, M. Pillon, L. Petrizzi 
    ENEA
    Centro Ricerche Energie Frascati
    UTS Fusione  
    Via E. Fermi 27
    C.P. 65
    I-00044 Frascati (Rome) 
    Italy

    Compiler of data for Sinbad:
    P. Batistoni
    ENEA
    Centro Ricerche Energie Frascati
    UTS Fusione  
    Via E. Fermi 27
    C.P. 65
    I-00044 Frascati (Rome) 
    Italy  
    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


    Acknowledgement
    ---------------
    The experiment and the corresponding analysis was performed in the framework of
    the EFDA (European Fusion Development Agreement) ITER Task (T-362-1997).


 8. Availability:
    ------------
    Unrestricted

 9. References:
    ----------

    [1] M. Martone, M. Angelone, M. Pillon, The 14 MeV Frascati Neutron
        Generator, Journal of Nuclear Materials 212-215 (1994) 1661-1664;
    [2] J.H. Hubble, Photon mass attenuation and energy-absorbtion
        coefficients from 1 keV up to 20 MeV, Int. J. Appl. Rad. Isot. 33
        (1982) 1269 
    [3] P. Batistoni, M. Angelone, L. Petrizzi, M. Pillon, “Neutron streaming
        experiment for ITER bulk shield at the Frascati 14-MeV neutron
        generator”, Proceedings of the 20th Symposium on Fusion Technology,
        Marseille, France, 7-11 September 1998, Edited by B. Beaumont, P.
        Libeyre, B. de Gentile, G. Tonon, CEA Cadarache 1998, Vol.2, pag. 1417
    [4] M. Angelone, P. Batistoni, L. Petrizzi, M. Pillon, “Neutron streaming
        Experiment at FNG :results and analysis”, Fusion Engineering and
        Design 51-52, (2000) 653-661
    [5] L. Petrizzi, P. Batistoni, I. Kodeli, “Sensitivity and uncertainty
        analysis performed on a 14-MeV neutron streaming experiment”, Fusion
        Engineering and Design 51-52, (2000) 843-848
    [6] K. Seidel, M. Angelone, P. Batistoni et al., “Investigation of
        neutron and photon flux spectra in a streaming mock-up for ITER”,
        Fusion Engineering and Design 51-52, (2000) 855-861
    [7] Nucl. Sci. Eng, 126, 176-186 (1997)
    [8] P. Batistoni et al, ITER Task T.362: Neutron Streaming Experiment - Final
        Report, EFF-Doc-639, July 1998  
    [9] P. Batistoni, L. Petrizzi, Analysis of the Neutron Streaming Experiment
        using FENDL-1.0/2.0 and EFF-3.0/3.1 nuclear data libraries, EFF-DOC-673
   [10] J. F. Briesmeister (Ed.), MCNP - A General Monte Carlo N-Particle
        Transport Code, Version 4B, Report, Los Alamos National Laboratory,
        LA-12625-M, March 1997.
   [11] S. Ganesan and P. K. McLaughlin, FENDL/E - evaluated nuclear data
        library of neutron interaction cross-sections and photon production
        cross-sections and photon-atom interaction cross-sections for fusion
        applications, version 1.0, Report IAEA-NDS-128, Vienna, May 1994.
   [12] M. Herman, A. B. Pashchenko, Extension and improvement of the FENDL
        library for fusion applications (FENDL-2), Report INDC(NDS)-373, IAEA
        Vienna, 1997.
   [13] A. J. Koning, H. Gruppelaar, A. Hogenbirk, Fusion Eng. Des. 37 (1997)
        211-216.
   [14] N. P. Kocherov, P. K. McLaughlin, The International Reactor Dosimetry
        File (IRDF-90), Report IAEA-NDS-141, Rev. 2, Oct. 1993.
   [15] I. Kodeli, Report on the 1999 Activity on ND-1.2.1 Subtask: Processed
        Multigroup Covariance Files with extended options for EFF-3, EFFDOC-698


10. Data and Format:
    ---------------

  FILE     NAME      bytes   Content
  ---- -----------  ------   -------
   1   fngstr-a.htm  17,611  This information file 
   2   fngstr-e.htm  58,935  Description of Experiment 
   3   mcnpfoil.inp  73,096  3-D model for MCNP-4A calculations of neutron
                             activation reaction rates
   4   mcnp_nh.inp   86,865  3-D model for MCNP-4A calculations of nuclear
                             heating
   5   mcnp_hss.inp  11,787  3-D simplified model for MCNP-4A calculations
                             of nuclear heating in stainless steel 
   6   mcnp_hcu.inp  11,837  3-D simplified model for MCNP-4A calculations
                             of nuclear heating in Copper
   7   mcnp_tld.inp  12,110  3-D simplified model for MCNP-4A calculations
                             of nuclear heating in TLD-300                  
   8   source.for    45,178  FORTRAN subroutine for MCNP source description
   9   trx-fng.inp    1.777  Input data for TRANSX cross-section preparation
  10   dort-fng.inp  10,880  Input data for GIP cross-section mixing, GRTUNCL
                             first collision source and DORT transport codes
  11   fig1.gif       5.242  Fig. 1: Angular dependence of the source 
  12   fig2.gif       9.505  Fig. 2: Energy/angular dependence of the source 
  13   fig3.gif      20,911  Fig. 3: Geometry of the experimental mock-up 
  14   fig4.gif      14,342  Fig. 4: Activation foils position in the
                                     channel and in the cavity  
  15   fig5.gif       9,633  Fig. 5: Geometry of the source
  16   fed2000.pdf  455,809  Reference 4
  17   fed2000a.pdf 699,972  Reference 5
  18   eff-639.pdf  700,204  Reference 8
  19   eff-673.pdf   66,110  Reference 9
  19   eff-698.pdf  336.644  Reference 15


    File fngstr-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: Geometrical arrangement of the bulk shielding assembly
    Tab. 4: Chemical composition of stainless steel AISI316
    Tab. 5: Experimental results (E) of reaction rates measurements along the
            central mock-up axis (ON-AXIS)
    Tab. 6: Experimental results (E) of reaction rates measurements inside 
            the cavity (ON-AXIS)
    Tab. 7: Experimental results (E) of reaction rates measurements in the
            channel and in the cavity (OFF-AXIS)
    Tab. 8: Measured dose in TDL-300, QTLD(E)
    Tab. 9: Calculated reaction rates (C) along the central mock-up axis. 
            Comparison between calculated and measured values (C/E ratios)
            (ON-AXIS) 
   Tab. 10: Calculated reaction rates (C) in the cavity. Comparison between
            calculated and measured values (C/E ratios) (ON-AXIS)
   Tab. 11: Calculated reaction rates (C) in the channel and in the cavity.
            Comparison between calculated and measured values (C/E ratios).
            (OFF-AXIS)
   Tab. 12: Calculated reaction rates (C) in the channel and in the cavity.
            Comparison between calculated and measured values (C/E ratios).


    The figures are included in gif format.

SINBAD Benchmark Generation Date: 6/2003
SINBAD Benchmark Last Update: 6/2003