JAERI- M 94- 014 Leakage Neutron Spectra from Various Sphere Piles with 14 MeV Neutrons Authors: Chihiro Ichihara I), Katsuhei Kobayashi I), Shu A. Hayashi II), Itsuro Kimura III), Junji Yamamoto IV) and Akito Takahashi IV). Organizations: I) Research Reactor Institute, Kyoto University Noda, Sennan-gun, Osaka 590-04, Japan II) Institute for Atomic Energy, Rikkyo University 2-5-1 Nagasaka, Yokosukas Kanagawa 240-01, Japan III) Department of Nuclear engineering, Faculty of Engineering, Kyoto University Yoshida-honmachi, Sakyo-ku, Kyoto 606, Japan IV) Department of Nuclear Engineering, Faculty of Engineering, Osaka University 2-1, Yamada-oka, Suita, Osaka 565, Japan Facility OKTAVIAN, Osaka University Date From 1984 to 1988 Measured Quantity The leakage current spectrum from the outer surface of the sphere pile with 14 MeV neutrons normalized per source neutron was measured for each sample. See Ref. 1 and 2. Experimental Method The experiment was performed with the time-of-flight (TOF) technique using the intense 14 MeV neutron source facility OKTAVIAN [Ref.3] at Osaka University. A tritium neutron producing target was placed at the center of the pile. A cylindrical liquid organic scintillator NE-218 (12.7 cm-diam - 5.1 cm-long) was used as a neutron detector, which was located about 11 m from the tritium target, and 55 deg. with respect to the deuteron beam axis. A pre-collimator made of polyethylene-iron multi-layers was set between the pile and the detector in order to reduce the background neutrons. The aperture size of this collimator was deterrnined so that the whole surface of the piles facing the detector could be viewed. The detector efficiency was determined by combining, 1) the Monte Carlo calculation, 2) the measured efficiency derived from the TOF measurement of Cf-252 spontaneous fission spectrum and the Watt's spectrum, and 3) the measured efficiency from the leakage spectrum from a graphite sphere, 30 cm in diameter with the similar detection system. To monitor the absolute number of the spectrum per source neutron, a cylindrical niobium foil was set in front of the tritium target and irradiated during the TOF experiment. From the gamma-ray intensity of the induced activity,Nb-92m and the integrated counts of the source neutron spectrum, the absolute neutron leakage spectrum can be obtained. The formulation of this procedure is stated elsewhere4). Neutron Source Characteristics The pulsed beam line of OKTAVIAN was used. Neutrons were produced by bombarding a 370 GBq tritium target with 250 keV deuteron beam. The energy spectrum of the neutron source is measured by using the same detection system as the each leakage spectrum measurement. The spatial distribution of the emitted neutrons were measured for each target assembly. The theoretical calculations were performed assuming the isotropical neutron source distribution. The neutron source spectra are given in Tables 4.2 to 4.6. Material / Geometry / Configuration Sample Piles Sample piles were made by filling spherical vessels with sample powder or flakes. Four different types of vessels were used, as the followings. 1) 61 cm diameter shell (Type-I) This type of vessels are used for the LiF, Mn, Cu, Zr and Mo piles. These are made of stainless steel (JIS SUS-304) for LiF and Zr pile, and Soft steel (JIS SS-41) for the Mn, Cu and Mo piles. Inner diameter, wall thickness of the vessels, are 60 cm and 0.5 cm, respectively. A reentrant hole for a beam duct is equipped, the diameter of which is 5.1 cm up to the center of the vessel. 2) 40 cm diameter vessel (Type-II) This stainless steel (JIS SUS-304) vessels are used for the TEFLON, A1,S3 Cr, Co, As, Se and W piles. This is equipped with a 20 cm diameter void at its center and a 11 cm diameter reentrant hole for the target beam duct. The thickness is 0.2 cm everywhere. 3) 60 cm diameter vessel (Type-III) This is made of 0.5 cm thick stainless steel (JIS SUS-304) and is used only for silicon pile. This vessel has a void and a re-entrant hole of same size as the 40 cm vessel. 4) 28 cm diameter vessel (Type-IV) This is made of 0.3 cm thick stainless steel (JIS SUS-304) and is used for Nb pile only. Similar reentrant hole is equipped as the type-I vessel. Table 4.1 shows the diameters, sarnple weight, apparent densities, sample thickness in units of centimeters and mean free paths for 14 MeV neutrons. Detailed descriptions of each pile are given in the various experimental data (*.dat) files. Experimental Data with Errors Error Assessment The experimental errors include only statistical deviation (1C) in the measurement of neutrons. The relative error to measure the niobium activation foils is less than 1 % (0.4 to 1 %), which is not included here. Example of Experimental Analysis An example of the input data for the MCNP calculation are given in Fig. 4.21. References 1) Ichihara C., et al.: Proc. Int. Conf. on Nucl. Data for Sci. and Technol., Mito, Japan, pp. 319-322 (1988). 2) Ichihara C., et al.: Proc. Second Specialists' Meeting on Nucl. Data for Fusion Reactors (1991), JAERI-M 91-062 (1991). 3) Sumita K., et al.: Proc. 12th SOFT, Vol. 1 (1982) 4) Takahashi A., et el.: OKTAVIAN Report, C-83-02 (1983). List of Tables and figure Table 4.1 Characteristic parameters of the sample piles. Table 4.2 Neutron source spectrum for LiF, Mn, Cu. Mo and W. Table 4.3 Neutron source spectrum for TEFLON, Si and Co. Table 4.4 Neutron source spectrum for Al. Table 4.5 Neutron source spectrum for Ti, As, Se and Zr. Table 4.6 Neutron source spectrum for Cr and Nb. Fig 4.21 Example of the input data for the MCNP calculation. Table 4.1 Characteristic parameters of the sample piles Sample Apparent Thickness Pile Diam. Weight Density (g/ (cm) (kg) cm**3) (cm) (MFPs) LiF 61 198.0 1.79 27.5 3.5 A1 40 32.8 1.22 9.8 0.5 Si 60 138.05 1.29 20.0 1.1 Ti 40 41.20 1.54 9.8 0.5 Cr 40 99.7 3.72 9.8 0.7 Mn 61 480.0 4.37 27.5 3.4 Co 40 52.0 1.94 9.8 0.5 Cu 61 675.0 6.23 27.5 4.7 Zr 61 311.9 2.84 27.5 2.0 Nb 28 47.7 4.39 11.2 1.1 Mo 61 236.0 2.15 27.5 1.5 W 40 118.6 4.43 9.8 0.8 Table 4.2 Neutron source spectrum for LiF, Mn, Cu, Mo and W. ============================================================= Lower Energy Upper Energy Lethargy Flux Error (MeV) (MeV) (l/MeV/n) ============================================================= 1.8323E+01 2.0250E+01 8.0240E-04 5.4400E-09 1.6579E+01 1.8323E+01 3.2070E-03 6.7210E-04 1.5002E+01 1.6579E+01 2.4850E+00 1 2160E-02 1.3574E+01 1.5002E+01 6.5260E+00 1.9780E-02 1.2282E+01 1.3574E+01 1.3510E-01 3.9130E-03 1.1113E+01 1.2282E+01 5.0880E-02 1.8430E-03 1.0056E+01 1.1113E+01 3.0860E-02 1.4510E-03 9.0989E+00 1.0056E+01 2.3540E-02 1.2570E-03 8.2330E+00 9.0989E+00 1.7990E-02 1.1080E-03 7.4496E+00 8.2330E+00 1.5640E-02 1.0220E-03 6.7406E~00 7.4496E+00 1.5780E-02 1.0060E-03 6.0992E+00 6.7406E+00 1.7340E-02 1.0260E-03 5.5188E+00 6.0992E+00 1.7930E-02 1.0250E-03 4.9936E+00 5.5188E+00 1.8360E-02 1.0130E-03 4.5184E+00 4.9936E+00 2.0440E-02 1.0320E-03 4.0884E+00 4.5184E+00 2.2130E-02 1.0450E-03 3.6993E+00 4.0884E+00 2.4330E-02 1.0680E-03 3.3473E+00 3.6993E+00 2.5480E-02 1.0610E-03 3.0288E+00 3.3473E+00 2.6990E-02 1.0580E-03 2.7405E+00 3.0288E+00 3.2490E-02 1.1180E-03 2.4797E+00 2.7405E+00 3.4730E-02 1.1290E-03 2.2438E+00 2.4797E+00 2.8860E-02 1.0320E-03 2.0302E+00 2.2438E+00 3.1340E-02 1.0590E-03 1.8370E+00 2.0302E+00 2.9550E-02 1.0270E-03 1.6622E+00 1.8370E+00 2.9740E-02 1.0240E-03 1.5040E+00 1.6622E+00 3.0440E-02 1.0300E-03 1.3609E+00 1.5040E+00 3.0190E-02 1.0250E-03 1.2314E+00 1.3609E+00 2.7970E-02 9.9390E-04 1.1142E+00 1.2314E+00 2.9060E-02 1.0090E-03 1.0082E+00 1.1142E+00 2.7260E-02 9.8140E-04 9.1225E-01 1.0082E+00 2.4160E-02 9.3480E-04 8.2544E-01 9.1225E-01 2.2850E-02 9.1430E-04 7.4689E-01 8.2544E-01 2.1490E-02 8.9430E-04 6.7581E-01 7.4689E-01 1.9240E-02 8.7000E-04 6.1150E-01 6.7581E-01 1.8130E-02 8.6600E-04 5.5331E-01 6.1150E-01 1.7090E-02 8.6300E-04 5.0065E-01 5.5331E-01 1.4640E-02 8.3790E-04 4.5301E-01 5.0065E-01 1.3790E-02 8.3610E-04 4.0990E-01 4.5301E-01 8.6970E-03 7.6210E-04 3.7089E-01 4.0990E-01 9.5660E-03 7.8980E-04 3.3560E-01 3.7089E-01 7.8440E-03 7.9370E-04 3.0366E-01 3.3560E-01 6.6080E-03 8.2520E-04 2.7476E-01 3.0366E-01 6.1700E-03 8.7260E-04 2.4862E-01 2.7476E-01 6.7080E-03 9.3870E-04 2.2496E-01 2.4862E-01 3.7490E-03 9.9020E-04 2.0355E-01 2.2496E-01 2.7650E-03 1.1400E-03 1.8418E-01 2.0355E-01 5.4770E-03 1.3620E-03 1.6665E-01 1.8418E-01 2.1220E-03 1.4790E-03 1.5079E-01 1.6665E-01 1.4210E-03 1.7600E-03 1.3644E-01 1.5079E-01 1.9360E-03 2.2820E-03 ============================================================= Table 4.3 Neutron source spectrum for Si and Co. ============================================================= Lower Energy Upper Energy Lethargy Flux Error (MeV) (MeV) (l/MeV/n) ============================================================= 1.9072E+01 2.0660E+01 5.9638E-04 4.3050E-04 1.7605E+01 1.9072E+01 1.0422E-03 4.6837E-04 1.6252E+01 1.7605E+01 7.9903E-02 2.2684E-03 1.5002E+01 1.6252E+01 3.8974E+00 1.5597E-02 1.3849E+01 1.5002E+01 7.5085E+00 2.1739E-02 1.2784E+01 1.3849E+01 1.9762E-01 3.5648E-03 1.1801E+01 1.2784E+01 7.4357E-02 2.2188E-03 1.0894E+01 1.1801E+01 4.2723E-02 1.7033E-03 1.0056E+01 1.0894E+01 3.1142E-02 1.4372E-03 9.2831E+00 1.0056E+01 2.1824E-02 1.1971E-03 8.5694E+00 9.2831E+00 1.9816E-02 1.1336E-03 7.9106E+00 8.5694E+00 1.7296E-02 1.0440E-03 7.3024E+00 7.9106E+00 1.5981E-02 1.0079E-03 6.7410E+00 7.3024E+00 1.3074E-02 9.2335E-04 6.2227E+00 6.7410E+00 1.3690E-02 9.3339E-04 5.7443E+00 6.2227E+00 1.4827E-02 9.5018E-04 5.3026E+00 5.7443E+00 1.2886E-02 8.7326E-04 4.8949E+00 5.3026E+00 1.4827E-02 9.0197E-04 4.5186E+00 4.8949E+00 1.5829E-02 9.0793E-04 4.1712E+00 4.5186E+00 1.7314E-02 9.2590E-04 3.8505E+00 4.1712E+00 1.6312E-02 8.9241E-04 3.5545E+00 3.8505E+00 1.8050E-02 9.2711E-04 3.2812E+00 3.5545E+00 1.7735E-02 9.1735E-04 3.0289E+00 3.2812E+00 1.8735E-02 9.1496E-04 2.7960E+00 3.0289E+00 3.5940E-02 1.1845E-03 2.5811E+00 2.7960E+00 3.8688E-02 1.1967E-03 2.3826E+00 2.5811E+00 2.0972E-02 8.9446E-04 2.1994E+00 2.3826E+00 2.0388E-02 8.6898E-04 2.0303E+00 2.1994E+00 2.0197E-02 8.5990E-04 1.8742E+00 2.0303E+00 2.1026E-02 8.7188E-04 1.7301E+00 1.8742E+00 1.8760E-02 8.2791E-04 1.5971E+00 1.7301E+00 1.9198E-02 8.3130E-04 1.4743E+00 1.5971E+00 1.9903E-02 8.4003E-04 1.3610E+00 1.4743E+00 1.9635E-02 8.2933E-04 1.2563E+00 1.3610E+00 1.8923E-02 8.1136E-04 1.1598E+00 1.2563E+00 1.9196E-02 8.1314E-04 1.0706E+00 1.1598E+00 1.8680E-02 8.0529E-04 9.8827E-01 1.0706E+00 1.4834E-02 7.3749E-04 9.1229E-01 9.8827E-01 1.5870E-02 7.4464E-04 8.4215E-01 9.1229E-01 1.4000E-02 6.9678E-04 7.7740E-01 8.4215E-01 1.3629E-02 6.8183E-04 7.1763E-01 7.7740E-01 1.1392E-02 6.3905E-04 6.6246E-01 7.1763E-01 1.1344E-02 6.3850E-04 6.1153E-01 6.6246E-01 1.0160E-02 6.1814E-04 5.6451E-01 6.1153E-01 9.3554E-03 6.0566E-04 5.2111E-01 5.6451E-01 8.5601E-03 5.9818E-04 4.8104E-01 5.2111E-01 8.3229E-03 6.1382E-04 4.4406E-01 4.8104E-01 6.9568E-03 6.0548E-04 4.0992E-01 4.4406E-01 6.6226E-03 6.2084E-04 3.7840E-01 4.0992E-01 5.8809E-03 6.2829E-04 3.4931E-01 3.7840E-01 5.8653E-03 6.5494E-04 3.2245E-01 3.4931E-01 3.7316E-03 6.3750E-04 2.9766E-01 3.2245E-01 4.3191E-03 6.8646E-04 2.7478E-01 2.9766E-01 4.4708E-03 7.4560E-04 2.5365E-01 2.7478E-01 3.6779E-03 7.9344E-04 2.3415E-01 2.5365E-01 2.6244E-03 8.4902E-04 2.1615E-01 2.3415E-01 1.6672E-03 9.3685E-04 1.9953E-01 2.1615E-01 4.0529E-03 1.1263E-03 1.8419E-01 1.9953E-01 1.9073E-03 1.1995E-03 1.7003E-01 1.8419E-01 9.4240E-04 1.3108E-03 1.5696E-01 1.7003E-01 5.6324E-04 1.5061E-03 1.4489E-01 1.5696E-01 6.9063E-04 1.9711E-03 ============================================================= Table 4.4 Neutron source spectrum for Al. ============================================================= Lcwer Energy Upper Energy Lethargy Flux Error (MeV) (MeV) (l/MeV/n) ============================================================= 1.8323E+01 2.0250E+01 1.9260E-04 2.3920E-04 1.6579E+01 1.8323E+01 6.0260E-03 3.8700E-04 1.5002E+01 1.6579E+01 2.9560E+00 6.7470E-03 1.3574E+01 1.5002E+01 6.2130E+00 9.8250E-03 1.2282E+01 1.3574E+01 2.0720E-01 1.8200E-03 1.1113E+01 1.2282E+01 4.6410E-02 8.9260E-04 1.0056E+01 1.1113E+01 2.5470E-02 6.7210E-04 9.0989E+00 1.0056E+01 1.9110E-02 5.7860E-04 8.2330E+00 9.0989E+00 1.4960E-02 5.1850E-04 7.4496E+00 8.2330E+00 1.3600E-02 4.9350E-04 6.7406E+00 7.4496E+00 1.2540E-02 4.6740E-04 6.0992E+00 6.7406E+00 1.3790E-02 4.6950E-04 5.5188E+00 6.0992E+00 1.3060E-02 4.4880E-04 4.9936E+00 5.5188E+00 1.4050E-02 4.5230E-04 4.5184E+00 4.9936E+00 1.5430E-02 4.6750E-04 4.0884E+00 4.5184E+00 1.6440E-02 4.7830E-04 3.6993E+00 4.0884E+00 1.7800E-02 4.9030E-04 3.3473E+00 3.6993E+00 1.8240E-02 4.7360E-04 3.0288E+00 3.3473E+00 1.7690E-02 4.5140E-04 2.7405E+00 3.0288E+00 3.4900E-02 5.8170E-04 2.4797E+00 2.7405E+00 2.7990E-02 5.2580E-04 2.2438E+00 2.4797E+00 2.1380E-02 4.6850E-04 2.0302E+00 2.2438E+00 2.0870E-02 4.6190E-04 1.8370E+00 2.0302E+00 2.1140E-02 4.5790E-04 1.6622E+00 1.8370E+00 2.0290E-02 4.4550E-04 1.5040E+00 1.6622E+00 2.0180E-02 4.4130E-04 1.3609E+00 1.5040E+00 1.9440E-02 4.3850E-04 1.2314E+00 1.3609E+00 1.9430E-02 4.4190E-04 1.1142E+00 1.2314E+00 1.9780E-02 4.4790E-04 1.0082E+00 1.1142E+00 1.7880E-02 4.2730E-04 9.1225E-01 1.0082E+00 1.8240E-02 4.1830E-04 8.2544E-01 9.1225E-01 1.6270E-02 4.0760E-04 7.4689E-01 8.2544E-01 1.5300E-02 3.9840E-04 6.7581E-01 7.4689E-01 1.2930E-02 3.7740E-04 6.1150E-01 6.7581E-01 1.1500E-02 3.6490E-04 5.5331E-01 6.1150E-01 9.6970E-03 3.4920E-04 5.0065E-01 5.5331E-01 9.5770E-03 3.4950E-04 4.5301E-01 5.0065E-01 7.3460E-03 3.3120E-04 4.0990E-01 4.5301E-01 5.7080E-03 3.2660E-04 3.7089E-01 4.0990E-01 6.3080E-03 3.4440E-04 3.3560E-01 3.7089E-01 5.2100E-03 3.4610E-04 3.0366E-01 3.3560E-01 5.1230E-03 3.6370E-04 2.7476E-01 3.0366E-01 3.7550E-03 3.6880E-04 2.4862E-01 2.7476E-01 3.0070E-03 3.8260E-04 2.2496E-01 2.4862E-01 3.8170E-03 4.3270E-04 2.0355E-01 2.2496E-01 2.0560E-03 4.6180E-04 1.8418E-01 2.0355E-01 2.1010E-03 5.1720E-04 1.6665E-01 1.8418E-01 3.0630E-03 6.1650E-04 1.5079E-01 1.6665E-01 2.8080E-04 7.0060E-04 1.3644E-01 1.5079E-01 9.6510E-04 8.4910E-04 1.2346E-01 1.3644E-01 3.0100E-04 1.0580E-03 Table 4.5 Neutron source spectrum for Ti and Zr. ============================================================= Lower Energy Upper Energy Lethargy Flux Error (MeV) (MeV) (l/MeV/n) ============================================================= 1.3323E+01 2.0250E+01 8.5210E-04 4.9700E-04 1.6579E+01 1.8323E+01 7.3980E-03 7.5650E-04 1.5002E+01 1.6579E+01 3.0480E+00 1.2110E-02 1.3574E+01 1.5002E+01 5.2090E+00 1.5780E-02 1.2282E+01 1.3574E+01 3.1820E-01 3.8420E-03 1.1113E+01 1.2282E+01 5.7600E-02 1.6710E-03 1.0056E+01 1.1113E+01 3.3970E-02 1.3080E-03 9.0989E+00 1.0056E+01 2.7110E-02 1.1690E-03 8.2330E+00 9.0989E+00 2.1230E-02 1.0390E-03 7.4496E+00 8.2330E+00 1.9520E-02 9.8360E-04 6.7406E+00 7.4496E+00 1.8390E-02 9.4700E-04 6.0992E+00 6.7406E+00 1.8500E-02 9.3990E-04 5.5188E+00 6.0992E+00 1.9540E-02 9.5230E-04 4.9936E+00 5.5188E+00 1.9840E-02 9.4860E-04 4.5184E+00 4.9936E+00 2.1800E-02 9.5030E-04 4.0884E+00 4.5184E+00 2.2430E-02 9.3400E-04 3.6993E+00 4.0884E+00 2.3140E-02 9.2330E-04 3.3473E+00 3.6993E+00 2.5600E-02 9.3280E-04 3.0288E+00 3.3473E+00 2.4330E-02 8.9160E-04 2.7405E+00 3.0288E+00 4.8600E-02 1.1730E-03 2.4797E+00 2.7405E+00 3.4110E-02 9.9500E-04 2.2438E+00 2.4797E+00 2.6910E-02 8.9400E-04 2.0302E+00 2.2438E+00 2.6310E-02 8.7980E-04 1.8370E+00 2.0302E+00 2.6640E-02 8.7660E-04 1.6622E+00 1.8370E+00 2.5950E-02 8.6310E-04 1.5040E+00 1.6622E+00 2.7550E-02 8.7870E-04 1.3609E+00 1.5040E+00 2.5270E-02 8.3850E-04 1.2314E+00 1.3609E+00 2.6230E-02 8.4180E-04 1.1142E+00 1.2314E+00 2.5560E-02 8.2620E-04 1.0082E+00 1.1142E+00 2.2460E-02 7.8160E-04 9.1225E-01 1.0082E+00 2.1490E-02 7.6540E-04 8.2544E-01 9.1225E-01 2.0260E-02 7.4640E-04 7.4689E-01 8.2544E-01 1.8070E-02 7.1140E-04 6.7581E-01 7.4689E-01 1.7000E-02 6.9280E-04 6.1150E-01 6.7581E-01 1.4440E-02 6.5390E-04 5.5331E-01 6.1150E-01 1.2660E-02 6.1700E-04 5.0065E-01 5.5331E-01 1.0160E-02 5.7230E-04 4.5301E-01 5.0065E-01 9.6910E-03 5.6050E-04 4.0990E-01 4.5301E-01 7.8450E-03 5.4010E-04 3.7089E-01 4.0990E-01 7.9410E-03 5.4900E-04 3.3560E-01 3.7089E-01 5.8960E-03 5.2630E-04 3.0366E-01 3.3560E-01 5.8060E-03 5.3810E-04 2.7476E-01 3.0366E-01 5.7410E-03 5.5040E-04 2.4862E-01 2.7476E-01 3.5090E-03 5.3130E-04 2.2496E-01 2.4862E-01 3.6500E-03 5.6750E-04 2.0355E-01 2.2496E-01 4.5690E-03 6.1670E-04 1.8418E-01 2.0355E-01 2.9960E-03 6.3180E-04 1.6665E-01 1.8418E-01 3.0180E-04 6.9590E-04 1.5079E-01 1.6665E-01 3.9150E-03 7.7890E-04 1.3644E-01 1.5079E-01 4.0100E-04 8.0300E-04 1.2346E-01 1.3644E-01 3.2820E-03 9.3210E-04 1.1171E-01 1.2346E-01 1.3190E-03 9.9830E-04 1.0108E-01 1.1171E-01 2.7870E-03 1.1260E-03 9.1461E-02 1.0108E-01 8.6330E-05 1.2880E-03 8.2757E-02 9.1461E-02 1.6870E-03 1.5580E-03 7.4882E-02 8.2757E-02 2.2860E-03 1.8850E-03 6.7756E-02 7.4882E-02 4.4540E-03 2.3850E-03 6.1308E-02 6.7756E-02 4.3580E-04 2.9990E-03 5.5474E-02 6.1308E-02 4.1570E-03 4.0280E-03 5.0195E-02 5.5474E-02 2.3470E-04 5.4320E-03 ============================================================= Table 4.6 Neutron source spectrum for Cr and Nb. ============================================================= Lower Energy Upper Energy Lethargy Flux Error (MeV) (MeV) (l/MeV/n) ============================================================= 1.9851E+01 2.1290E+01 5.6528E-04 7.6934E-05 1.8509E+01 1.9851E+01 8.9187E-04 9.1009E-05 1.7257E+01 1.8509E+01 6.2716E-03 2.1460E-04 1.6091E+01 1.7257E+01 2.5845E-02 4.2836E-04 1.5003E+01 1.6091E+01 1.5107E+00 3.2555E-03 1.3989E+01 1.5003E+01 1.0173E+01 8.4874E-03 1.3043E+01 1.3989E+01 1.1774E+00 2.8681E-03 1.2161E+01 1.3043E+01 2.3692E-01 1.2753E-03 1.1339E+01 1.2161E+01 1.1093E-01 8.7561E-04 1.0572E+01 1.1339E+01 6.7508E-02 6.8220E-04 9.8576E+00 1.0572E+01 4.3770E-02 5.3107E-04 9.1911E+00 9.8576E+00 3.3211E-02 4.5984E-04 8.5697E+00 9.1911E+00 2.9003E-02 4.2457E-04 7.9904E+00 8.5697E+00 2.1512E-02 3.5464E-04 7.4502E+00 7.9904E+00 1.8452E-02 3.1954E-04 6.9465E+00 7.4502E+00 1.7686E-02 3.0679E-04 6.4769E+00 6.9465E+00 1.8390E-02 3.0820E-04 6.0390E+00 6.4769E+00 1.7823E-02 2.9963E-04 5.6307E+00 6.0390E+00 1.7977E-02 2.9699E-04 5.2501E+00 5.6307E+00 1.8432E-02 2.9710E-04 4.8951E+00 5.2501E+00 2.0260E-02 3.0781E-04 4.5642E+00 4.8951E+00 2.0650E-02 3.0109E-04 4.2556E+00 4.5642E+00 2.0475E-02 2.9066E-04 3.9679E+00 4.2556E+00 2.0855E-02 2.8544E-04 3.6997E+00 3.9679E+00 2.1243E-02 2.8255E-04 3.4495E+00 3.6997E+00 2.2556E-02 2.8604E-04 3.2163E+00 3.4495E+00 2.3450E-02 2.8686E-04 2.9989E+00 3.2163E+00 2.3214E-02 2.7990E-04 2.7961E+00 2.9989E+00 2.5287E-02 2.8682E-04 2.6071E+00 2.7961E+00 2.7205E-02 2.9285E-04 2.4308E+00 2.6071E+00 2.5341E-02 2.7960E-04 2.2665E+00 2.4308E+00 2.4825E-02 2.7385E-04 2.1133E+00 2.2665E+00 2.5031E-02 2.7199E-04 1.9704E+00 2.1133E+00 2.4625E-02 2.6646E-04 1.8372E+00 1.9704E+00 2.4426E-02 2.6272E-04 1.7130E+00 1.8372E+00 2.4443E-02 2.6130E-04 1.5972E+00 1.7130E+00 2.4628E-02 2.6167E-04 1.4892E+00 1.5972E+00 2.4120E-02 2.5849E-04 1.3885E+00 1.4892E+00 2.4115E-02 2.5755E-04 1.2947E+00 1.3885E+00 2.3317E-02 2.5222E-04 1.2071E+00 1.2947E+00 2.3842E-02 2.5406E-04 1.1255E+00 1.2071E+00 2.3898E-02 2.5274E-04 1.0494E+00 1.1255E+00 2.1836E-02 2.4054E-04 9.7847E-01 1.0494E+00 2.0025E-02 2.2935E-04 9.1232E-01 9.7847E-01 2.0373E-02 2.3143E-04 8.5064E-01 9.1232E-01 1.8970E-02 2.2381E-04 7.9314E-01 8.5064E-01 1.8197E-02 2.1994E-04 7.3952E-01 7.9314E-01 1.7442E-02 2.1856E-04 6.8952E-01 7.3952E-01 1.6948E-02 2.1850E-04 6.4290E-01 6.8952E-01 1.5096E-02 2.1026E-04 5.9944E-01 6.4290E-01 1.4809E-02 2.1399E-04 5.5891E-01 5.9944E-01 1.3402E-02 2.0943E-04 5.2113E-01 5.5891E-01 1.1994E-02 2.0688E-04 4.8590E-01 5.2113E-01 1.0816E-02 2.0815E-04 4.5305E-01 4.8590E-01 1.0554E-02 2.1788E-04 4.2242E-01 4.5305E-01 9.0655E-03 2.1817E-04 3.9386E-01 4.2242E-01 9.3138E-03 2.3884E-04 3.6723E-01 3.9386E-01 9.1826E-03 2.5881E-04 3.4241E-01 3.6723E-01 7.3661E-03 2.5891E-04 3.1926E-01 3.4241E-01 6.2065E-03 2.6745E-04 2.9767E-01 3.1926E-01 5.9811E-03 2.9443E-04 2.7755E-01 2.9767E-01 5.7234E-03 3.2673E-04 2.5878E-01 2.7755E-01 5.6127E-03 3.7221E-04 2.4129E-01 2.5878E-01 5.0132E-03 4.2506E-04 2.2498E-01 2.4129E-01 4.3433E-03 4.5321E-04 2.0977E-01 2.2498E-01 4.3453E-03 5.0127E-04 1.9559E-01 2.0977E-01 5.4237E-03 6.1372E-04 1.8236E-01 1.9559E-01 4.8982E-03 7.8557E-04 1.7003E-01 1.8236E-01 2.9865E-03 1.0623E-03 1.5854E-01 1.7003E-01 2.6089E-04 1.4566E-03 1.4782E-01 1.5854E-01 3.9903E-03 1.8749E-03 1.3783E-01 1.4782E-01 4.5304E-03 2.3566E-03 1.2851E-01 1.3783E-01 5.1923E-03 3.0776E-03 1.1982E-01 1.2851E-01 1.2950E-02 4.4153E-03 1.1172E-01 1.1982E-01 2.8607E-02 7.0492E-03 1.0417E-01 1.1172E-01 0.0000E+00 1.2781E-02 9.7125E-02 1.0417E-01 2.0059E-02 1.6848E-01 ========================================================== Fig. 21 Example of input data for MCNP calculations. LEAKAGE FROM ALUMINUM(40CM DIA) SPHERE 3-D SURFACE TALLY (JENDL-3) C **t** CELL CARDS ***** 1 0 (-3 -8):(8 -1 -6) 2 2 -7.824 (-4 3 -8):(8 -2 1 -6) 3 1 -1.223 (-5 4 -8):(-5 2 8) 4 2 -7.824 (-6 5 -8):(-6 5 8 2) 5 0 -7 6 C ***** SURFACE CARDS ***** 1 CX 5.55 2 CX 5.75 3 SO 10.0 4 SO 10.2 5 SO 19.75 6 SO 19.95 7 SO 100.0 8 PX 8.32 C ***** DATA CARDS ***** MODE n Sdef pos = 0.0 0.0 0.0 cel=1 erg = d1 Imp:n 1 1 1 1 0 C *** ENERGY BIN FOR SOURCE NEUTRON *** SI1 1.000E-01 1.120E-01 1.260E-01 1.410E-01 1.590E-01 1.780E-01 2.000E-01 2.240E-01 2.520E-01 2.830E-01 3.170E-01 3.560E-01 4.000E-01 4.490E-01 5.040E-01 5.660E-01 6.350E-01 7.130E-01 8.000E-01 8.780E-01 9.640E-01 1.058E+00 1.162E+00 1.275E+00 1.400E+00 1.542E+00 1.698E+00 1.871E+00 2.061E+00 2.270E+00 2.500E+00 2.704E+00 2.924E+00 3.162E+00 3.419E+00 3.699E+00 4.000E+00 4.165E+00 4.337E+00 4.516E+00 4.703E+00 4.897E+00 5.099E+00 5.310E+00 5.529E+00 5.757E+00 5.995E+00 6.242E+00 6.500E+00 6.765E+00 7.041E+00 7.327E+00 7.627E+00 7.938E+00 8.261E+00 8.598E+00 8.949E+00 9.314E+00 9.693E+00 1.009E+01 1.050E+01 1.082E+01 1.114E+01 1.148E+01 1.183E+01 1.218E+01 1.255E+01 1.277E+01 1.300E+01 1.324E+01 1.348E+01 1.372E+01 1.397E+01 1.422E+01 1.447E+01 1.474E+01 1.500E+01 1.527E+01 1.554E+01 1.583E+01 1.611E+01 1.640E+01 C *** SOURCE DISTRIBUTION *** SP1 0.000E-04 0.000E-00 0.000E-00 1.270E-04 5.774E-05 2.536E-04 2.722E-04 2.076E-04 4.366E-04 3.873E-04 4.756E-04 6.161E-04 6.727E-04 6.648E-04 8.581E-03 1.098E-03 1.184E-03 1.412E-03 1.616E-03 1.546E-03 1.556E-03 1.631E-03 1.771E-03 1.804E-03 1.823E-03 1.891E-03 1.935E-03 2.002E-03 2.052E-03 2.004E-03 2.068E-03 2.091E-03 3.354E-03 1.492E-03 1.322E-03 1.451E-03 1.401E-03 7.112E-04 6.423E-04 6.514E-04 6.195E-04 6.428E-04 6.209E-04 5.788E-04 5.227E-04 5.250E-04 5.456E-04 5.106E-04 5.789E-04 5.391E-04 4.998E-04 4.813E-04 5.300E-04 5.756E-04 5.230E-04 5.394E-04 6.256E-04 7.047E-04 7.729E-04 7.951E-04 8.659E-04 8.106E-04 8.923E-04 1.022E-03 1.281E-03 1.687E-03 2.286E-03 1.825E-03 2.479E-03 3.794E-03 7.010E-03 1.565E-02 3.634E-02 7.492E-02 1.279E-01 1.768E-01 1.916E-01 1.500E-01 8.676E-02 3.950E-02 1.430E-02 4.269E-03 C ***** MATERIAL CARDS ***** M1 13027 1 M2 24000 -0.185 26000 -0.704 28000 -0.111 C ***** TALLY CARDS ***** F21:n 6 C ***** ENERGY BIN ***** E21 1.000E-03 1.290E-03 1.670E-03 2.150E-03 2.780E-03 3.590E-03 4.640E-03 5.990E-03 7.740E-03 1.000E-02 1.290E-02 1.670E-02 2.150E-02 2.780E-02 3.590E-02 4.640E-02 5.990E-02 7.740E-02 1.000E-01 1.120E-01 1.260E-01 1.410E-01 1.590E-01 1.780E-01 2.000E-01 2.240E-01 2.520E-01 2.830E-01 3.170E-01 3.560E-01 4.000E-01 4.490E-01 5.040E-01 5.660E-01 6.350E-01 7.130E-01 8.000E-01 8.780E-01 9.640E-01 1.058E+00 1.162E+00 1.275E+00 1.400E+00 1.542E+00 1.698E+00 1.871E+00 2.061E+00 2.270E+00 2.500E+00 2.704E+00 2.924E+00 3.162E+00 3.419E+00 3.699E+00 4.000E+00 4.165E+00 4.337E+00 4.516E+00 4.703E+00 4.897E+00 5.099E+00 5.310E+00 5.529E+00 5.757E+00 5.995E+00 6.242E+00 6.500E+00 6.765E+00 7.041E+00 7.327E+00 7.627E+00 7.938E+00 8.261E+00 8.598E+00 8.949E+00 9.314E+00 9.693E+00 1.009E+01 1.050E+01 1.082E+01 1.114E+01 1.148E+01 1.183E+01 1.218E+01 1.255E+01 1.277E+01 1.300E+01 1.324E+01 1.348E+01 1.372E+01 1.397E+01 1.422E+01 1.447E+01 1.474E+01 1.500E+01 1.527E+01 1.554E+01 1.583E+01 1.611E+01 1.640E+01 C ***** CUT OFF CARD, WIN SHAKES E-L WEIGHT***** CUT:N 1.0E16 1.0E-3 0.01 C ***** NEUTRON HISTORY ***** NPS 1000000 PRINT