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MCNPX Benchmark Experiment Number 3 Rev. 0.0.1
Transmission of Medium Energy Neutrons Through Concrete Shields (1991)

 1. Name of Experiment:

    Measurements of the transmission of secondary neutrons through concrete shields, involved a 75-MeV proton beam incident on a 
    stopping-range copper assembly, performed at the AVF Cyclotron Facility at Osaka University.

 2. Purpose and Phenomena Tested:

    A series of experiments, performed at the AVF Cyclotron Facility at Osaka University, involved a 75-MeV proton beam incident 
    on a stopping-range copper assembly to study the transmission of secondary neutrons through concrete shields.  The data were 
    used by the experimenters to benchmark the MORSE Monte Carlo code with the DLC-87 multigroup cross section library for neutron
    penetration calculations.
 3. Description of Source and Experimental Configuration:

    Protons were accelerated to 75 MeV in the AVF cyclotron and transported to a 1-cm-thick (stopping range) Cu target. Neutrons 
    generated in the forward direction were passed to an experimental room through a 7.5-cm-inner-diameter iron-lined concrete 
    collimator 50 cm in length (the thickness of the lining was not given).  Concrete shields of 40 x 40 cm2 area and 20, 50, 
    and 100 cm thickness were placed 7 cm from the collimator exit. Figure 1 illustrates the experimental arrangement[1].  

 4. Measurement System:

    Neutrons penetrating the shield were measured by a 7.6-cm-diameter x 7.6-cm-long NE-213 scintillator.
    Background data were obtained for this configuration with the collimator blocked by an iron plug.  The source spectrum measurement 
    was made in a similar configuration, the only differences being that the concrete shields were removed and the detector was located 
    778 cm from the target.
 5. Description of Results and Analysis:

   The atom densities (atoms/cm3) for concrete were given in Reference 1. All atomic weights were taken from Reference 3.  For conversion 
   of atom densities to atom and weight percents performed here,  Avogadro's number was taken to be 6.022E23. The material isotopic compositions 
   are given in Table 1.  No information was given on the compositions of Fe and Cu. 
   Figure 2 illustrates two experimentally obtained source spectra:  one from the unfolding of the pulse height spectrum, and the other converted 
   from the time-of-flight (TOF) spectrum.  The pulse height spectrum was unfolded using the FERDO-U[2] method.  The TOF measurement was only 
   possible for neutrons above 25 MeV, because the repetition rate (1/55 ns-1) of the proton beam was not variable.  Contributions to the TOF spectrum 
   from lower energy neutrons were removed by discriminating against signals corresponding to the pulse height for recoil protons of energy less than 
   25 MeV.  The secondary neutron emission spectrum from the target at 0° from the beam axis was calculated by the experimenters using the ARIES 
   code system.  Results from this calculation are given with the experimental results in Figure 2[1].
   Figure 3 illustrates the spectra of neutrons transmitted through the concrete shields.  Also shown here are results from calculations of 
   transmission spectra performed by the experimenters with the MORSE Monte Carlo code with the DLC-87 cross section library. The data from Figure 3 
   has been digitized and is given in tabular format in Tables 3, 4, and 5 for the 20 cm, 50 cm, and 100 cm thick concrete shields respectively.
   The error bars given on the measured values in Figure 3 above are from the output of the FERDO-U code.  Fluctuations in the measured spectra at 
   20 - 40 MeV arise for two reasons.  First, the transmitted neutron fluence for the 100-cm-thick shield was on the same order of magnitude as the 
   background neutron fluence, and so subtracting the background spectrum increases the uncertainty in the transmission spectrum.  Second, the response 
   matrix used for unfolding was constructed based on measured response data, and since these data did not cover the entire range of interest, data 
   were interpolated and extrapolated where necessary.

   The neutron source for this benchmark was the measured neutron source spectrum given in Reference 1. All neutron transmission calculations
   were performed by emitting the measured neutron source spectrum in a cone of 3.14E-3 steradians. The measured neutron spectrum is given in Table 6
   and Figure 4.
   The neutron source spectrum was also calculated using MCNPX with 75-MeV protons incident on a copper target. The calculated neutron source was
   calculated by performing a tally over all neutrons emitted from the target in an angle of 3.5°. The calculated neutron spectrum is given in 
   Table 7 and Figure 5.  The MCNPX model used to calculate the neutron source spectrum is given in Appendix A.
   The concrete shields were placed 385 cm from the neutron source.  These concrete shields were modeled as slabs 40 cm by 40 cm in cross section and
   20, 50, and 100 cm thick.  The shield assemblies were broken into subsections to facilitate importance weighting.
   The transmitted neutron spectrum was tallied using the F4 tally over a cell of the same size as the scintillator that was used in the experiment.
   The MCNPX model used to calculate the transmitted neutron spectra is given in Appendix B.
   The composition of the shield assemblies used in the calculations were the same as the measured shield compositions, with exception of those trace
   elements that are not contained within the LA150 cross section set. The composition of the shield assemblies is given in Table 8.
   For both the calculation of the neutron source and the neutron transmission calculations the default physics options were used[4].
   The 150-MeV cross section set, LA-150, created for the Accelerator Production of Tritium Project was used for these calculations[5]. The LAHET/MCNP 
   cross over value was set to 150-MeV so that tabular neutron cross section data was used over the whole range of neutron energies.
   Importance regions in the shield assemblies were used to force neutrons towards the detector. The importances were varied to maintain the neutron
   population constant in all regions in the shield assembly.
   The results of the calculations of neutrons transmitted through the concrete shields are listed in Table 9 and Figure 6.

 6. Special Features:

 7. Author/Organizer:
    Experiment and analysis:

    Compiler of data for Sinbad:
    B. Maidou
    URANUS, 78 470 St Rémy les Chevreuse, France

    Screened by:
    I. Kodeli
    OECD/NEA, 12 bd des Iles, 92130 Issy les Moulineaux, France

 8. Availability:

 9. References:
    [1] Shin K., Ishii Y.,  Uwamino Y., Sakai H. and Numata S.: "Transmission of Medium Energy Neutrons Through Concrete Shields,"
    Radiation Protection Dosimetry, Vol. 37, No. 3, 175-178 (1991).

    [2] K. Shin et al., Nucl. Technol., 53, 78 (1981).

    [3] Knolls Atomic Power Laboratory, Chart of the Nuclides, 12th Edition (1977)

    [4] H. Grady Hughes, Richard E. Prael, and Robert C. Little.: "MCNPX - The LAHET/MCNP Code Merger." Los Alamos National Laboratory 
    Memorandum XTM-RN(U) 97-012, April 22, 1997.

    [5] M.B. Chadwick, P.G. Young, S. Chiba, S.C. Frankle, G.M. Hale, H.G. Hughes, A.J. Koning, R.C. Little, R.E. MacFarlane, R.E. Prael, 
    and L.S. Waters. Cross Section Evaluations to 150 MeV for Accelerator-driven Systems and Implementation in MCNPX. Nucl. Sci. and Eng., 
    131(3):293, March 1999.

    [6] Georgia Tech MCNPX Benchmarkng Homepage: http://epicws.epm.ornl.gov/pending_benchmarks/GTECH_ACCELERATOR/index.html

10. Data and Format:

    Filename          Size[bytes] Content
    ----------------- ----------- ----------------
  1 AVF75-a.html        18,331 This information file
  2 AVF75-e.html          6,082 Tables with numerical data
  3 Figure_1.html           211 Fig. 1: Experimental Setup
  4 Figure_2.html           220 Fig. 2: Source Neutron Spectra
  5 Figure_3.html           225 Fig. 3: Transmitted Neutron Spectra
  6 Figure_4.html           229 Fig. 4: Measured Neutron Spectrum
  7 Figure_5.html           235 Fig. 5: Calculated Neutron Spectrum
  8 appendix_A.html       3,320 Appendix A: The MCNPX model used to calculate source
  9 appendix_B.html       4,057 Appendix B: The MCNPX model
 10 Figure_6.html           199 Fig. 6: Calculated Transmission Neutron Spectrum behind Concrete Shields

    File AVF75-e.html contains the following table:

     Table 1: Material Isotopic Compositions.
     Table 2: Measured Neutron Source Spectrum.
     Table 3: Measured Transmission Neutron Spectra behind a 20cm Concrete Shield.
     Table 4: Measured Transmission Neutron Spectra behind a 50cm Concrete Shield.
     Table 5: Measured Transmission Neutron Spectra behind a 100cm Concrete Shield.

    Figures are included in the HTML format.
SINBAD Benchmark Generation Date: 11/2008 SINBAD Benchmark Last Update: 11/2008