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MEPhI Empty Slits Streaming Experiment

 1. Name of Experiment:
    MEPhI empty slits streaming experiment (1995-1998)

 2. Purpose and Phenomena Tested:
    Measurements of neutron reaction rates, heating rates, neutron
    and photons spectra in iron shielding models with empty slits
    irradiated by 14-MeV neutrons (see Figure 4).

 3. Description of Source and Experimental Configuration:
    The 14-MeV DT MEPhI neutron generator was the neutron source. The 
    maximum neutron output equal 2·1011 neutron per second by 200 keV 
    accelerating potential. The angular and energy dependence of the source
    neutron output is presented in Table 1 and Figure 1.
    The calculation method corresponds to work [1].
    The geometry of target units is given in Figure 2. The base geometry
    of researched schielding compositions are presented in Figure 3,
    Figure 4 and Figure 5. The following possible cases of a mutual
    position of two slits were considered during designing mock-ups 
    with hollow slits in protective compositions:

    1. The source neutrons can intersect two slits.
    2. The neutrons of a source can intersect only one slit at 
       minimum possible distance between the axis of the displaced 
    3. The maximum position of the slits in the installation, when 
       the path of source neutrons is less than a distance between 

    In these cases at the 500 mm thickness of a mock-up and the 100 mm
    distance of a neutron source from a front surface the axis 
    displacement of slits was 32.5, 70 and 120 mm. The models of 
    shieldings were assembled from the steel milling blocks of the 
    25x200x500 mm3 and 50x200x500 mm3 sizes so that to avoid a direct 
    run of neutrons on the touch surfaces of blocks. For this purpose 
    the figured subsidiary blocks of a 1000 mm length, on which the 
    upper part of a mock-up of a shield was installed, were made. The 
    horizontal slit gap of a 20 mm size was made by erection of 
    subsidiary blocks on two steel stops having the 20x25x500 mm3 
    sizes. The distance between the stops was selected from 
    conditions of installation strength and their minimum influence 
    to a condition of infinity of a slit size in a horizontal 
    direction. For measurements of neutron and photon spectra, the
    horizontal channels of a 40 mm diameter each were arranged. The  
    channels had a steel stub removing under insertion into the channel 
    of a probe of a scintillation spectrometer. The shielding material 
    is steel, having impurity content presented in Table 2. 

 4. Measurement System:
    The following nuclear reactions were chosen for the measurements:
    115In(n,n’), 64Zn(n,p), 27Al(n,alpha), 56Fe(n,p), 19F(n,2n) (see 
    Table 3) and 233U(n,f). In defining detector mass and sizes the 
    following peculiarities of measurements in slit shield compositions 
    were taken into account:
    - considerable neutron field gradients near slit boundaries 
      perpendicular to their directions,
    - increase of reaction rate gradients near the slits with their 
      energy threshold increase,
    - insignificant neutron field gradients along the direction of 
      the slits,
    - decrease of reaction rates by a factor of 2-3 with a high-energy 
      threshold in moving away from the slit edge.

    Detector masses necessary for reaction rates measurements with a 
    statistical error in the range of (3-5)% as well as their form 
    and sizes needed for obtaining necessary spatial resolution of 
    the measured functionals were determined on the basis of the
    calculations. To reduce the number of detectors, modules in the 
    form of a parallelepiped with edge sizes equal to 2x5x30 mm3 were 
    manufactured. In the experiments detectors were put at the 
    back surface of the shielding compositions. The error of 
    coordinate setting of the detector didn’t exceed 0.2 mm. The 
    detectors length is equal to 60 mm, their height being 5 mm. The 
    detectors were arranged at the back surface of the assembly in 
    such a way that 60x5 mm2 faces were parallel to the slit surface. 
    The measurements along X and Y axis were made. All detectors co-
    coordinates are given in Tables 4 and 5.

    To measure heating rate in an iron composition material a technique
    based on application of thermoluminescent detectors was used.
    Harshaw-2080 which included a heating device, a photoelectric
    detector and necessary electronics circuits was used to process TLD. 

    The heating device provides uniform volume heating with a 
    constant linear velocity in the temperature range of (50-400)°C. 
    Detectors were some powder poured into an glass ample with inner 
    and outside diameters equal to 1 mm and 2 mm correspondingly, the 
    height being equal to (15-20) mm. Contribution of a gamma-
    radiation, arising in a target unit, was accounted in heating 
    rates for a mock-up of a uniform iron shield.

    Two stilbene crystals of cylindrical shape with the sizes of 
    Ø30x30 mm2 and Ø40x40 mm2 are used in the basic version of a 
    spectrometer for measurements in bulk composition. For 
    measurements on the back surface the stilbene crystal of the 
    size Ø 50x40 mm3 was used [2].

    All measured neutron and photon functionals were normalized to 
    one neutron source and one nucleus (or gramme).

 5. Description of Results and Analysis:
    Calculation and experimental studies showed that there was some
    dependence of nuclear reaction rates on the size of an ion spot
    which defined the neutron source. Mostly this was observed for 
    reaction rates with high energy thresholds. To avoid this effect 
    high voltage stabilization at a level less than 0.5% at the 
    neutron generator in MEPhI was required. It is a complicated 
    technical problem this is the reason why all the slits in target 
    structures were located horizontally and a diaphragm was introduced 
    into the target block. It was proved experimentally that in case 
    of 500 mm thick shield and 20 mm transverse slit the largest width 
    of a square diaphragm limiting falling ion flux down the target was 
    equal to 13.8 mm. Thus there is a compromise between the size of 
    a neutron source and its yield decrease.

    The measurement of an absolute neutron yield is based on counting
    of alpha particles accompanying neutrons when the latter are
    produced. Alpha particle radiometers are placed in the ion beam
    conduit and have two independent channels with different diaphragm
    diameters. 0.3 mm thick scintillation crystals CsJ(Tl) are placed
    behind these diaphragms. The accuracy of the neutron output 
    corresponds to 2.5% for one σ. The nuclear reaction rate measurement
    system includes eight gamma-spectrometric tracts on NaJ(Tl) crystals
    with Ø63x63 mm2 sizes that are connected with an amplitude 
    analyzer plate entered into a personal computer.

    Heating rate in shield compositions is determined by an absorbed 
    dose of ionizing radiations in its material components. In the
    present experiments the shield material is steel, the main 
    component of which is iron. The method of measuring the absorbed 
    dose should meet the following demands:
    - high sensitivity;
    - small distortion of neutron and gamma-fields;
    - high spatial resolution;
    - wide linear range of absorbed dose measurements;
    - error in measuring absolute value of the absorbed dose should 
      be at (8-10)% level.

    At present TLD irradiation in shield mock-ups is considered to be 
    the most suitable for these purposes. Bragg-Grey relations are 
    taken to be a methodical base because they correlate with the 
    dose absorbed in the detector and its environment. TLD used in the
    present experiments basic characteristics makes it possible to obtain
    values of the absorbed dose from gamma-quanta in the material of shield
    compositions by an effective atomic number by means of interpolation.
    It should be pointed out that light output of TLD which is being 
    measured after irradiating the latter in shield compositions, 
    will correspond to the absorbed dose of not only gamma but also 
    neutron radiation. This is the reason why the dose absorbed in TLD is 
    taken to be determined by the sum of two constituents:


        - Dn - is a neutron constituent of the absorbed dose and Dg
        corresponds to its gamma component. As the methodical base of 
        energy release determination refers to a gamma component, a 
        neutron component should be subtracted from the total value of 
        the absorbed dose. According to the assumptions adopted in the 
        work [3] the neutron dose absorbed in TLD can be presented by 
        the following:


     - F(En)- an energy distribution of neutron flux density 
       (spectrum) normalized to one source neutron 
     - G(En) - a function of TLD neutron response of a certain type. 

    Measured neutron reaction rates are given in Tables 6-14. Basic properties
    of the used thermoluminescent detectors CaSO4(Dy) and SrSO4(Tb) type are 
    given in Table 15. Measured heating rates and neutron spectra, are given in 
    Tables 16-18 and 19, 20, 21, 22. Physics methods of streaming experiments
    analysis are given in works [4], [5], [6,] [7], [8], [9]. For MCNP-4c2 
    calculation the MCNP-4c2 input files are provided. The co-ordinate axis in
    calculation and experiment have not mutual correlation.

 6. Special Features:

 7. Author/Organizer:
    Experiment and analysis:

    Romodanov V.L.* - science lieder and physics method creation

Andreev M.I.* - heating rates measurements

Afanasiev V.V.* - reaction rates measurements

Belevitin A.G.* - reaction rates measurements

Sacharov V.K.* - MCNP-4c2 calculation models

Trykov L.A.** - neutron and photons spectra measurements *Moscow Engineering-Physics Institute, State University Moscow, 115409, Kashirskoe Shosse 31, Russia **State Research Center “Institute of Nuclear Power Engineering”, Obninsk, Kaluga region, Bondarenko Square 2, Russia. Compiler of data for Sinbad: Romodanov V.L. Sacharov V.K. Moscow Engineering-Physics Institute (State University), Moscow, 115409, Kashirskoe Shosse 31, Russia Reviewer of compiled data: I. Kodeli OECD/NEA, 12 bd des Iles, 92130 Issy les Moulineaux, France Acknowledgement --------------- The experiment and the corresponding analysis was performed in the framework of ISTC project No. 180. 8. Availability: ------------ Unrestricted 9. References: ---------- [1] Fewell T.R., On evaluation of the alpha counting technique for determining 14 MeV neutron yield, NIM, Amsterdam, V.61 (1), pp. 61-71. 1968. [2] Trikov L.A., Colevatov Yu. I., Volkov V.S., Methods of spec- trometer adjustment by means of radionuclide sources of neutrons, Preprint of INPE No. 1730, Obninsk, INPE, 1985 (in Russian). [3] H.Hashikura et al., Calculation of neutron response of thermo- luminiscent dosimeters, J.of Faculty of Eng., University of Tokio, Vol. XXXIX, No1 pp7-16, 1987. [4] Afanasiev V.V. Andreev M.I., Belevitin A.G., Romodanov V.L., Benchmark-Experiments and Analyses on Streaming of 14-MeV Neutrons in Iron and Iron-Water Radiation Shielding Mock-Ups ith Slits, Preprint 003-94, MEPhI, Moscow, 1994. [5] V.V. Afanasiev, Andreev M.I., Belevitin A.G., Romodanov V.L., at al., Benchmark Experiments and Analyses on Streaming of 14-MeV Neutrons in Iron and Iron-Water Radiation Shielding Mock-Ups with Slits, Topical Meeting, Radiation Protection and Shieldings. No. Falmouth, April 21-25, Vol.2, p. 687. 1996. [6] Afanasiev V.V., Andreev M.I., Belevitin A.G., Dmitriev D.M., Romodanov V.L., Experimental and calculational studies of non- uniform shieldings of fusion reactors, Proc. ISFNT-4, April 1997, Japan, p.271. [7] V.V. Afanasiev, Andreev M.I., Belevitin A.G., Romodanov V.L., Dmitriev D.M at al., Benchmark experiments on non-uniform iron shielding compositions (ISTC project No. 180), Proc. ICENES98 - 9th International Conference on Emerging Nuclear Energy Systems, Tel-Aviv, Israel, June 28-July 2, 1998 pp.407-413. [8] Afanasiev V.V., Andreev M.I., Belevitin A.G., Dmitriev D.M., Romodanov V.L., Experimental and calculational studies of non-uniform shieldings of fusion reactor, Fusion Engineering and Design, 42, 1998, 261-266. [9] Andreev M.I., Afanasiev V.V., Belevitin A.G., A.V.Karaulov, Romodanov V.L. et al., Set of benchmark experiments on slit shielding compositions of thermonuclear reactors, Fusion Engineering and Design, 55, 2001, 373-385. 10. Data and Format: --------------- FILE NAME Bytes Content --------------------------------------------------------------------- 1 MEPhIstr-a.htm 19159 This information file 2 MEPhIstr.htm 41583 Description of Experiment 3 mcnp1Y.inp 4878 Model for Y axis MCNP-4c2 calculations of 1 mock-up 4 mcnp1Z.inp 4525 Model for Z MCNP-4c2 calculations of 1 mock-up 5 mcnp2Z.inp 4922 Model for Z MCNP-4c2 calculations of 2 mock-up 6 mcnp3Z.inp 4894 Model for Z MCNP-4c2 calculations of 3 mock-up 7 mcnp4Z.inp 4915 Model for Z MCNP-4c2 calculations of 4 mock-up 8 mcnp5Z.inp 4889 Model for Z MCNP-4c2 calculations of 5 mock-up 9 mcnp6Z.inp 5813 Model for Z MCNP-4c2 calculations of 6 mock-up 10 mcnp7Z.inp 5799 Model for Z MCNP-4c2 calculations of 7 mock-up 11 mcnp8Z.inp 5799 Model for Z MCNP-4c2 calculations 12 spectra.inp 5158 Model for MCNP-4c2 spectra calculation 13 fig1.gif 43754 The angular and energy dependence of the source neutron output 14 fig2.gif 17859 The geometry of target units 15 fig3.gif 15215 General view on mock-up 16 fig4.gif 68162 View on mock-up 17 >fig5.gif 43537 View on mock-up 18 table1.xls 31993 The angular and energy dependence of the source neutron output 19 table2.xls 17408 The material of shielding composition 20 table3.xls 9475 Set of threshold nuclear reactions 21 table4.xls 11376 X detectors co-ordinate 22 table5.xls 28401 Y detectors co-ordinate 23 table6-14.xls 156571 Reaction rates in mock-ups 24 table15.xls 10892 Thermoluminescent detectors 25 table16-18.xls 50043 Heating rates in mock-ups 26 table19.xls 27204 Basic neutron and photon spectra 27 table20.xls 48650 Basic neutron and photon spectra behind the model 28 table21.xls 63097 Spectra of neutrons in different X coordinates mock-up 1 29 table22.xls 30998 Spectra of photons in different X coordinates mock-up 1 30 table23.xls 17920 Error source of absolute value of neutron yield 31 table24-26.xls 19968 Error source components 32 MEPhI-str1.pdf 372136 Reference 8 33 MEPhI-str2.pdf 1630465 Reference 9 Figures are included in gif format. SINBAD Benchmark Generation Date: 12/2005 SINBAD Benchmark Last Update: 12/2005