SINBAD ABSTRACT NEA-1553/40
MEPhI Empty Slits Streaming Experiment
1. Name of Experiment:
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MEPhI empty slits streaming experiment (1995-1998)
2. Purpose and Phenomena Tested:
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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
slits.
3. The maximum position of the slits in the installation, when
the path of source neutrons is less than a distance between
slits.
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:
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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:
where:
- 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:
where
- 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:
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None
7. Author/Organizer:
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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
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The experiment and the corresponding analysis was
performed in the framework of ISTC project No. 180.
8. Availability:
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Unrestricted
9. References:
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[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:
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FILE NAME Bytes Content
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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
