JAERI- M 94- 014 Gamma-ray Energy Spectra Emitted from Spheres with 14 MeV Neutron Source Authors J. Yamamoto, T. Kanaoka, I. Murata, A. Takahashi and K. Sumita Organization Department of Nuclear Engineering, Faculty of Engineering, Osaka University 2-1 Yamada-oka, Suita, Osaka 565, Japan Facility OKTAVIAN, Osaka University Date From June, 1987 to September, 1989 Measured Quantity Energy spectra of gamma-rays were measured using a 14 MeV neutron source. The gamma-rays were produced from (n,xgamma) reactions in spheres and emitted from the spherical samples. The measured quantity was a "leakage current spectrum." Experimental Method Gamma-rays were detected with a cylindrical NaI crystal and the energy spectra were obtained from the unfolding process of the gamma-ray pulse-height spectra, using a response matrix of the NaI detector. The detector was located at 5.8 m distance from the neutron source and counted the gamma-rays emitted from the sphere. Time spectra of neutrons and gamma-rays from the sphere were measured simultaneously with the pulse-height spectra by means of a TOF technique. OKTAVIAN was run in the pulsed mode with a repetition frequency of 500 kHz. The pulse width was 3 ns in FWHM and the difference in flight times between the 14 MeV neutrons and the prompt gamma-rays was about 90 ns from the sphere to the detector. Since those were enough to separate the gamma-rays from the neutron background in the TOF spectra, the desired gamma-rays could be discriminated from a neutron background. The emission spectra were dominated by the gamma-rays from (n,n') and (n,2n) reactions rather than the gamma-rays from (n,xgamma) reaction. The data are therefore available to the assessment in the nuclear data for energy distributions of gamma-rays from non-elastic scattering by high energy neutrons. Neutron Source Characteristics Pulses of 14 MeV neutrons were generated by D-T reaction. A 370 GBq TiT target was bombarded with a pulsed D+ beam of 243 keV. The following information is available: Energy spectrum of emission neutrons, neutron yield as a function of emission angle and production of gamma-rays at the source. The neutron energy spectrum was same as that given in Ref. (1). The other information about the emission neutrons is given elsewhere 2). The energy spectrum of gamma-rays at the source is shown in Table 3.1 and Fig. 3.1. Materials / Geometry / Configuration The samples in use were A1, Si, Ti, Cr, Mn, Co, Cu, Nb, Mo, W, Pb, LiF and CF2 (Teflon). All samples except Pb were same as those for neutron spectra measurement shown in Table 4.1. As for Pb, inner and outer diameters are 10 and 20 cm, and the pile has an opening of 10 cm diameter for the beam duct. The pile is made of natural lead and its density is 11.34 g/cm3. Numerical values of the spectra are given in Tables 3.2 - 3.15. Error Assessment The following error sources were included in the errors. (a) Uncertainty in monitoring absolute fluxes of the source neutrons (b) Errors of the response matrix (c) Statistical deviation (lcs) Examples of Experimental Analysis MCNP + JENDL-3 Some results are compared with the experimental data in the figures. Input data for a graphite sphere of 30 cm in diameter are shown in Fig. 3.20. Configuration is shown in Fig. 3.4 and the result is shown in Fig. 3.5 in comparison with the measured energy spectrum. Comments and So Forth Both the leakage spectra of neutron and gamma-ray can be compared with transport calculations, because of the same condition with regard to the spherical samples and the neutron source. (See Ref. 1 ) In the present experiment the period to measure the prompt gamma-rays from the sphere was 60 to 80 ns after the source neutrons generating, so that it was necessary to compute the gamma-ray fluxes in the spheres by using time-dependent transport calculations. However, steady state transport calculations were applicable under the compatible condition with the experimental one by a simple method, in which the neutron-mean-emission times from the sphere were investigated as a function of neutron energy to estimate neutron slowing down times in the spheres. The periods of measurement of the prompt gamma-ray spectrum after the source neutrons generation are i) 56 ns for Cr and W, ii) 65 ns for Mn, Cu, Nb, Mo and LiF, iii) 69 ns for A1 and iv) 70 ns for Si and Ti. As for Co and CF2, the periods are not given. Typical period, i. e., 65 ns, can be used instead because the measured spectra are less sensitive to the small difference of the period. In numerical tables of the experimental data, each energy spectrum was normalized in a leakage current flux from a whole surface of the sphere per a source neutron. References (1) Ichihara C., et al.: "Leakage Neutron Spectra from Various Sphere Piles with 14 MeV Neutrons", chapter 4 in this issue. (2) Yamamoto J., et al.: "Numerical Tables and Graphs of Leaka~ge Neutron Spectra from Slabs of Typical Shielding Materials with D-T Neutron Source", OKTAVIAN-Report A-8305, Dept. of Nuclear Eng., Osaka University (1983). (3) Yamamoto J., et al.: "Gamma-Ray Emission Spectra from Spheres with 14 MeV Neutron Source", JAERI-M 89-026, 232 (1989). (4) Yamamoto J.: "Integral Experiment on Gamma-Ray Production at OKTAVIAN", JAERI-M 91-062, 118 (1991). List of Tables and figure Table 3.1: Energy spectrum of gamma-rays at the source 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. Example of the input data for the MCNP calculation. Table 3.1 Target gamma-ray spectrum emitted from the target. ----------------------------- Upper Photon Photons/Group Energy [MeV] /Source ----------------------------- 0.6 0.000e+00 0.7 2.961e-03 0.8 4.793e-03 0.9 9.615e-03 1.0 4.543e-03 1.1 4.082e-03 1.2 5.335e-03 1.3 6.609e-03 1.4 4.793e-03 1.5 3.476e-03 1.6 2.035e-03 1.7 1.733e-03 1.8 1.733e-03 1.9 1.557e-03 2.0 1.326e-03 2.1 1.257e-03 2.2 1.191e-03 2.3 1.070e-03 2.4 1.014e-03 2.5 1.070e-03 2.6 1.129e-03 2.7 1.014e-03 2.8 8.638e-04 2.9 7.761e-04 3.0 8.188e-04 3.1 8.638e-04 3.2 8.638e-04 3.3 8.188e-04 3.4 7.761e-04 3.5 7.356e-04 3.6 7.356e-04 3.7 7.356e-04 3.8 6.973e-04 3.9 6.264e-04 4.0 5.628e-04 4.1 5.335e-04 4.2 5.335e-04 4.3 5.628e-04 4.4 5.938e-04 4.5 6.264e-04 4.6 5.938e-04 4.7 5.057e-04 4.8 4.082e-04 4.9 3.476e-04 5.0 3.295e-04 5.1 3.476e-04 5.2 3.667e-04 5.3 4.082e-04 5.4 4.082e-04 5.5 3.869e-04 5.6 3.667e-04 5.7 3.295e-04 5.8 2.960e-04 5.9 2.806e-04 6.0 2.806e-04 6.1 2.960e-04 6.2 2.960e-04 6.3 2.960e-04 6.4 2.806e-04 6.5 2.659e-04 6.6 2.521e-04 6.7 2.389e-04 6.8 2.389e-04 6.9 2.521e-04 7.0 2.521e-04 7.1 2.389e-04 7.2 1.929e-04 7.3 1.642e-04 7.4 1.476e-04 7.5 1.399e-04 7.6 1.326e-04 7.7 1.326e-04 7.8 1.326e-04 7.9 1.326e-04 8.0 1.326e-04 8.1 1.326e-04 8.2 1.257e-04 8.3 1.070e-04 8.4 9.113e-05 8.5 8.188e-05 8.6 6.973e-05 8.7 6.609e-05 8.8 6.264e-05 8.9 5.938e-05 9.0 5.938e-05 9.1 5.628e-05 9.2 5.628e-05 9.3 5.335e-05 9.4 5.057e-05 9.5 4.306e-05 9.6 3.667e-05 9.7 2.960e-05 9.8 2.659e-05 9.9 2.659e-05 10.0 2.806e-05 ----------------------------- Total number of photons per source neutron: 0.0862 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 2.044OE-02 1.0320E-03 4.0884E+00 4.5184E+00 2.2130E-02 1.0450E-03 3.6993E+00 4.0884E+00 2.433OE-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 l.OO90E-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 8.943OE-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+O0 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 6.213OE+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+O1 l.9110E-02 5.7860E-04 8.233OE+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 4.9936Ee00 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 5.123OE-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 8.233OE+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 S.Ol95E-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-OS 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-O1 O.OOOOE+OO 1.2781E-02 9.7125E-02 1.0417E-01 2.0059E-02 1.6848E-Ol ========================================================== An example of input data of MCNP for a graphite sphere of 30 cm in diameter. TALLY CARD FOR BE EXPERIMET ANALYSIS BY MCNP JENDL3 (1991,9) C *-*t* CELL CARDS '** 1 0 -2 2 1 -1.800 -1 2 3 0 -3 1 4 0 3 C ***** SURFACE CARDS ***** 1 50 15.00 2 SO 4.0 3 SO 621.35 C ***** DATA CARDS ***** MODE N P SDEF ERG=D1 POS=0.0 0.0 0.0 IMP:N 1 1 1 0 IMP:P 1 1 1 0 C *** ENERGY BIN FOR SOURCE NEUTRON *** SI1 H 0.10580E+01 0.11620E+01 0.12750E+01 0.14000E+01 0.15420E+01 0.16980E+01 0.18710E+01 0.20610E+01 0.22700E+01 0.25000E+01 0.27040E+01 0.29240E+01 0.31620E+01 0.34190E+01 0.36990E+01 0.40000E+01 0.41650E+01 0.43370E+01 0.45160E+01 0.47030E+01 0.48970E+01 0.50990E+01 0.53100E+01 0.55290E+01 0.57570E+01 0.59950E+01 0.62420E+01 0.65000E+01 0.67650E+01 0.70410E+01 0.73270E+01 0.76270E+01 0.79380E+01 0.82610E+01 0.85980E+01 0.89490E+01 0.93140E+01 0.96930E+01 0.10089E+02 0.10500E+02 0.10817E+02 0.11143E+02 0.11479E+02 0.11825E+02 0.12182E+02 0.12549E+02 0.12775E+02 0.13005E+02 0.13239E+02 0.13477E+02 0.13720E+02 0.13967E+02 0.14218E+02 0.14474E+02 0.14735E+02 0.15000E+02 0.15270E+02 0.15545E+02 0.15825E+02 0.16110E+02 0.16399E+02 C ****'** SOURCE DISTRIBUTION '''''''*******'************-^* SP1 D 0.00000E+00 0.43294E-01 0.32540E-01 0.42945E-01 0.40272E-01 0.27926E-01 0.34186E-01 0.34042E-01 0.23955E-01 0.31569E-01 0.43807E-01 0.41738E-01 0.17916E-01 0.17044E-01 0.13307E-01 0.13417E-01 0.11811E-01 0.11526E-01 0.97991E-02 0.12098E-01 0.91486E-02 0.87084E-02 0.64650E-02 0.68528E-02 0.59667E-02 0.62186E-02 0.64940E-02 0.62448E-02 0.55632E-02 0.62477E-02 0.53803E-02 0.51710E-02 0.55622E-02 0.53794E-02 0.61662E-02 0.61352E-02 0.68313E-02 0.63205E-02 0.68856E-02 0.80788E-02 0.76653E-02 0.87169E-02 0.10680E-01 0.10213E-01 0.12114E-01 0.15459E-01 0.14916E-01 0.14518E-01 0.22867E-01 0.22260E-01 0.85022E-01 0.82752E-01 0.80572E-01 0.25612E+00 0.24935E+00 0.24271E+00 0.40968E+00 0.39870E+00 0.38819E+00 0.38627E+00 0.37601E+00 C ***** MATERIAL CARDS *-*** Z M1 6012.33 1 C '*'** TALLY CARDS ***** F21:P 3 C '''' TIME DISTRIBUTION ( UPPER BOUND ) 1=10 nsec '**''' T21 10.0889 1.0E4 C ***'' ENERGY BIN ***** E21 0.05 0.10 0.15 0.20 0.25 0.30 0.35 0.40 0.45 0.50 0.55 0.60 0.65 0.70 0.75 0.80 0.85 0.90 0.95 1.00 1.05 1.10 1.15 1.20 1.25 1.30 1.35 1.40 1.45 1.50 1.55 1.60 1.65 1.70 1.75 1.80 1.85 1.90 1.95 2.00 2.05 2.10 2.15 2.20 2.25 2.30 2.35 2.40 2.45 2.50 2.55 2.60 2.65 2.70 2.75 2.80 2.85 2.90 2.95 3.00 3.05 3.10 3.15 3.20 3.25 3.30 3.35 3.40 3.45 3.50 3.55 3.60 3.65 3.70 3.75 3.80 3.85 3.90 3.95 4.00 4.05 4.10 4.15 4.20 4.25 4.30 4.35 4.40 4.45 4.50 4.55 4.60 4.65 4.70 4.75 4.80 4.85 4.90 4.95 5.00 5.05 5.10 5.15 5.20 5.25 5.30 5.35 5.40 5.45 5.50 5.55 5.60 5.65 5.70 5.75 5.80 5.85 5.90 5.95 6.00 6.05 6.10 6.15 6.20 6.25 6.30 6.35 6.40 6.45 6.50 6.55 6.60 6.65 6.70 6.75 6.80 6.85 6.90 6.95 7.00 7.05 7.10 7.15 7.20 7.25 7.30 7.35 7.40 7.45 7.50 7.55 7.60 7.65 7.70 7.75 7.80 7.85 7.90 7.95 8.00 8.05 8.10 8.15 8.20 8.25 8.30 8.35 8.40 8.45 8.50 8.55 8.60 8.65 8.70 8.75 8.80 8.85 8.90 8.95 9.00 9.05 9.10 9.15 9.20 9.25 9.30 9.35 9.40 9.45 9.50 9.55 9.60 9.65 9.70 9.75 9.80 9.85 9.90 9.95 10.0 C ***** CUT OFF CARD ***** CUT:N 1.0E+4 1.0E-8 0.01 CUT:P 1.0E+4 1.0E-2 0.01 C ***** NEUTRON HISTORY ***** NPS 1600000 PRINT