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FNS Integral Experiment on Graphite Cylindrical Assembly

1. Name of Experiment:
 Integral Experiment on a 60 cm-thick Graphite Cylindrical Assembly
 (FNS/JAERI clean benchmark) (1984)

2. Purpose and Phenomena Tested:
 A cylindrical experimental assembly made of graphite was placed in front of
 D-T neutron source of Fusion Neutronic Source (FNS) facility at JAERI.
 Detectors were inserted in the assembly at several positions along the
 central axis of the cylinder. Nuclear responses related to both neutron
 and gamma-ray were measured with various experimental techniques. The
 measured quantities were: fission rate, reaction rate, neutron spectra
 and dose rate.

3. Description of Source and Experimental Configuration:
 The mail sources of information used in this compilation were refs. [10],
 [2], [8] and [9].
 Reactor-grade graphite blocks were stacked in thin wall (2 mm) aluminum
 tubes to form a cylindrical slab in the same manner as the lithium-oxide
 assembly [1]. The size of the graphite assembly was 31.4 cm in equivalent
 radius and 61.0 cm in thickness. The blocks used in this assembly were four
 types except near the central region where there was an experimental channel.
 A sectional view of the assembly is shown in Figure 1. The graphite blocks
 used were selected from the inventory so as to have the density with the
 deviation within ±2%. The data of blocks are summarized in Table 1. The
 average density was (1.641±0.015) g/cm3.

 The experimental channel, a set of sheath and drawer, was made of the same
 grade graphite (100 x 100 x 1000 mm3 - 1.654±0.002 g/cm3) in order to save
 the changing time of detector position and for minimizing the personnel
 exposure for experimentalists. A sectional view of the sheath, drawer
 and spacers are also shown in Figure 2. The experimental channel was
 placed at the central axis of the assembly. Therefore, this experimental
 assembly consisted of a single element, i.e., carbon, except the aluminum
 framework. Graphite blocks with experimental hole of 21 mm diam. were
 made to allow insertion of a detector. Homogenized nuclide densities in
 each region are tabulated in Table 2.

 The 80-degree beam line in the first target room of the FNS facility was
 used. A high speed water-cooled target [3] was set at the end of the beam
 line. A 7.4 x 1011 Bq (20Ci) Ti-T target was mounted on the target assembly.
 Neutrons were generated at the distance of 20 cm from the assembly surface
 on its central axis. The setting accuracy is estimated to be within ±1 mm.
 A view of the experimental arrangement is shown in Figure 3. The layout in
 the first target room of 15 m x 15 m is illustrated in Figure 4. The
 distances from the target to the west and south walls are 5.5 m, and those
 to the ceiling, the grating floor and the basement floor are 7.9, 1.8 and
 3.8 m, respectively.

 Neutron yields were determined by means of the associated alpha-particle
 detection method [4]. A small silicon surface-barrier detector with an
 aperture of about 1 mm diam. was mounted inside the beam line to detect
 the alpha-particle of 3T(d,n)4He reaction. Source characteristics, that
 is, neutron yield, angular distribution and spectra of the target assembly
 were measured by the time-of-flight technique [5], foil activation and a
 NE213 spectrometer [6].

 A good agreement was obtained between neutron yield measured by different
 methods within the experimental error. An analysis by Monte Carlo
 calculation [7] also showed fairly good agreement with measured neutron
 energy spectra as well as angular distributions, the latter obtained by
 foil activation. Thus, the calculated source spectrum and the other
 characteristics were essentially confirmed and can be used as input
 information in the benchmark calculations. Source neutron spectrum is
 given in Table 3. It should be noticed that the number of neutrons
 emitted toward 0 degree with respect to the d+ beam must be normalized
 as 1.1767 per unit D-T reaction at the target.

4. Measurement System:
 Fission rate were measured by micro-fission chambers of 235U, 238U, 232Th and
 237Np, and fission track detectors.

 Reaction rate distribution along the central axis of the assembly were
 obtained by foil activation technique. The Al, Ni, Zr and Nb foils were
 10 mm in diameter and 0.5 mm in thickness; the In foils were 10 mm square
 and 0.1 mm thick; the Au foils were 10 mm square and 0.001 mm thick.

 The neutron spectra were measured by a small sphere (14 mm diam.) NE213
 organic scintillation counter at 8 positions given in Table 10. The
 proton recoil spectrum was unfolded by the FORIST code (see ref. [2] p.20
 for details).
 Gamma-ray dose rates were measured by thermoluminescence dosimeters (TLDs).

 Detailed description of the measurements is given in [2].

5. Description of Results and Analysis:
 Fission rates measured by the micro-fission chambers are shown in Table 4
 and Figure 5. Reaction rate distribution obtained by the foil activation
 technique is shown in Table 5 and Figures 6 and 7. Decay data needed for
 reaction rate calculation and error analysis for the reaction rate measurement
 are summarized in Table 6 and Table 7 respectively. Response rates of TLDs
 are shown in Table 8 and Figures 8, 9 and 10. Fission rates measured by the
 fission track detector method are shown in Table 9. Fission rates distribution
 in the graphite assembly measured by fission track detector method and those
 by micro-fission chambers are shown in Figure 11. Neutron spectra measured by
 the NE213 spectrometer at eight positions are given in Tables 12 to 19 and
 Figures 12, 13, 14, 15, 16, 17, 18 and 19. Detector loction and alpha count of
 the spectrum measurements are shown in Table 10.

 Error Assessment:
 Some descriptions about error assessment are given in the Tables.
 Major sources of error of fission chamber measurements were uncertainties
 in the neutron yield (2.1 %), effective fissile atom numbers (3.2 ~ 4.6 %)
 and positioning (1 %).
 The sources of error for the reaction rate measurements are listed in Table 7.
 The errors of the unfolded spectra include two contributions. One of them,
 the statistical error, is included in the tables together with the spectra.
 The other error is related to the response matrices which are used in the
 unfolding procedure. This is included in the systematic errors summarized in
 Table 11.
 More information about error assessment is given in [2].

 Example of Experiment Analysis:
 The sample input data for GRTUNCL and DOT codes taken from [8] are given in the
 file dot35.inp. The calculation model is shown in Figure 20. The sample input
 for the MCNP-4A code taken from [9] is given in the file mcnp4a.inp. Benchmark
 calculation with the MCNP-4A code are presented in ref. [8]. The calculations
 were performed using JENDL-3.2 and FENDL-1 nuclear data libraries. The results
 for the neutron spectrum calculation are compared with the measurements in the
 ref. [9] (Figs. 4.3.7 - 4.3.11, on pages 63 - 66, and Figs. 4.4.1 - 4.4.3, on
 pages 105 - 106).

6. Special Features:

7. Author/Organizer:
 Experiment and analysis:
 H. Maekawa, Y. Ikeda, Y. Oyama, S. Yamaguchi, K. Tsuda, T. Fukumoto,
 K. Kosako, M. Yoshizawa and T. Nakamura
 Japan Atomic Energy Research Institute
 Tokai-mura, Naka-gun, Ibaraki-ken, 319-1195 Japan
 Phone: 81-292-82-6809 (for Y.O.) or -6015 (for H.M.)
 Fax: 81-292-82-5996 (for Y.O.) or -6365 (for H.M.)
 e-mail: fujio@fnshp.tokai.jaeri.go.jp
 or oyama@cens.tokai.jaeri.go.jp

 Compiler of data for Sinbad:
 S. Kitsos
 OECD/NEA, 12 bd des Iles, 92130 Issy les Moulineaux, France
 e-mail: stavros.kitsos@free.fr

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

8. Availability:

9. References:

 [1] Maekawa H., et al.: "Fusion Blanket Benchmark Experiments on a 60 cm-Thick
 Lithium Oxide Cylindrical Assembly," JAERI-M 86-182 (1986).
 [2] Maekawa H., et al.: "Benchmark Experiments on a 60 cm-Thick Graphite
 Cylindrical Assembly," JAERI-M 88-034 (1988).
 [3] Seki M., et al.: J. Nucl. Sci. Technol., 16, 8238 (1979).
 [4] Maekawa H., et al.: "Neutron Yield Monitor for the Fusion Neutronics Source
 (FNS) For 80° Beam Line ---," JAERI-M 83-219 (1983).
 [5] Oyama Y. and Maekawa H.: Nucl. Instr. Meth., A245, 173 (1986).
 [6] Oyama Y., et al.: "Development of a Spherical NE213 Spectrometer with 14 mm
 Diameter," JAERI-M 84-124 (in Japanese, 1984);
 Nucl. Instr. Meth., A256, 333 (1987).
 [7] Seki Y., et al.: J. Nucl. Sci. Technol., 20, 686 (1983).
 [8] F. Maekawa, J. Yamamoto, C. Ichihara, K. Ueki, Y. Ikeda: "Sub Working Group
 of Fusion Reactor Physics Subcommittee: Collection of Experimental Data for
 Fusion Neutronics Benchmark", JAERI-M-94-014, Feb. 1994 (Ch. 1.6, p.217).
 [9] Fujio Maekawa, Masayuki Wada, Chihiro Ichihara, Yo Makita,
 Akito Takahashi, Yukio Oyama: "Compilation of Benchmark Results for Fusion
 Related Nuclear Data", JAERI-Data/Code 98-024, Nov. 1998.
 [10] Maekawa H., et al.: "Integral Experiment on Graphite Cylindrical Assembly",
 International Atomic Energy Agency, Nuclear Data Section: Compilation for
 FENDL benchmarks, available from:

10. Data and Format:

 Filename Size[bytes] Content
 ---------------- ----------- -------------
 1 fnsc-abs.htm 16.345 This information file
 2 fnsc-exp.htm 52,489 Description of experiment
 3 fnsc-f1.gif 7.304 Figure 1: Sectional view of the cylindrical assembly
 4 fnsc-f2.gif 12.350 Figure 2: Sectional view of the graphite sheath and drawer
 5 fnsc-f3.gif 5.630 Figure 3: Experimental layout
 6 fnsc-f4.gif 13.867 Figure 4: Layout of the FNS first target room
 7 fnsc-f5.gif 11.161 Figure 5: Fission rates measured by micro-fission chambers
 8 fnsc-f6.gif 9.480 Figure 6: Foil activation measured reaction rates
 9 fnsc-f7.gif 9.032 Figure 7: Foil activation measured reaction rates
10 fnsc-f8.gif 9.646 Figure 8: Graphite TLDs (TDL-600, 700, 100) response
11 fnsc-f9.gif 9.559 Figure 9: Graphite TLDs (MSO-S, SSO-S, BSO-S) response
12 fnsc-f10.gif 9.770 Figure 10: Graphite TLDs (UD-110S, 136N,137N) response
13 fnsc-f11.gif 8.422 Figure 11: Fission rates measured by FDT and 
   micro-fission chambers
14 fnsc-f12.gif 11.794 Figure 12: Measured neutron spectrum z=24.1 cm
14 fnsc-f13.gif 10.606 Figure 13: Measured neutron spectrum z=26.7 cm
15 fnsc-f14.gif 12.013 Figure 14: Measured neutron spectrum z=31.7 cm
15 fnsc-f15.gif 10.600 Figure 15: Measured neutron spectrum z=41.8 cm
16 fnsc-f16.gif 11.477 Figure 16: Measured neutron spectrum z=52.0 cm
16 fnsc-f17.gif 10.423 Figure 17: Measured neutron spectrum z=62.1 cm
17 fnsc-f18.gif 11.627 Figure 18: Measured neutron spectrum z=72.1 cm
17 fnsc-f19.gif 10.503 Figure 19: Measured neutron spectrum z=77.2 cm
18 fnsc-f20.gif 17.594 Figure 20: Calculation model for the graphite assembly
25 dot35.inp 12.793 Input data for GRTUNCL and DOT calculations
26 mcnp4a.inp 11.694 Input data for MCNP-4A calculations
27 j88-034.pdf 4,753,760 Reference
27 j94-014.pdf 34.067.821 Reference
28 j98-024.pdf 12.837.046 Reference

 File fnsc-exp.htm contains the following tables:

 Table 1: Data of graphite blocs
 Table 2: Homogenized characteristics for assembly
 Table 3: Neutron source distribution
 Table 4: Fission-rates (micro-fission chambers)
 Tables 5-7: Foil activation reactions rates, decay data and error analysis
 Table 8: TLD Measurements
 Table 9: Fission-rates (fission track detector)
 Tables 10-19: Neutron spectra, detector location and error analysis

 Figures are included in GIF format.

SINBAD Benchmark Generation Date: 12/2001
SINBAD Benchmark Last Update: 12/2001