[en] Semi-empirical methods based on what is called the removal cross-section are widely used for high-speed evaluation of fast neutron fluxes and the doses they deliver. Determination of this quantity is based on an approximation of the decay of the fast neutron flux beyond a layer of shielding material with an exponential function. In this study data were obtained for the removal cross-section and the relaxation length in the case of neutron flux from a 14 MeV one-directional source passing through layers of iron shielding. The numerical computations of the neutron distributions were performed by the discrete ordinate transport technique and the Monte Carlo method

[en] The calculations of neutron fluxes emitted by a 14-MeV source passing through lead sheets are analysed. The MORSE and BLANK programs are used with different input data and approximations. The aim of the study is to compare the results obtained by P-5 and P-8 approximations based on ENDL, the two libraries ENDL and ENDF in P-5 approximation, the programs MORSE and BLANK. 6 tabs., 2 figs., 3 refs

No abstract available

[en] The attenuation has been studied of the total 14 MeV neutron flux passing through an iron collimator and directly located behind lead slabs with thicknesses 2.5, 5.0, 7.5, 10.0 and 12.5 cm. A comparison is made between the calculated data and results from a benchmark experiment. The experiment is analysed with an one-dimensional model based on ANISN-code and a tree-dimentional model based on MORSE-code (Monte Carlo method). The energy dependences of the cross-sections and neutron fluxes are represented by a 25-group approximation based on the SUPERTOG program and the files ENDL and ENDF/B-4. The mean deviation of MORSE calculated data from experimental ones is 6.4% for ENDL-data and 8.5% for EDDF-data. ANISN-calculated data show practically full coincidence with experimental ones both for ENDL and ENDF

[en] For the calculation of the group-to-group transfer matrix with the SUPERTOG program it is necessary to specify the data for the anisotropy of the elastic scattering with the coefficients of expansion of the density function into Legendre polynomials. For all materials of the ENDL library and for some materials from FNDF/B-IV the information for the scattering anisotropy is specified as tabulated values of the density function itself. For these materials the existing versions of SUPERTOG are inapplicable. The modification of the SUPERTOG program prepared by the authors makes the calculation of the elastic scattering matrix possible using the two ways of anisotropy setting. A method based on approximation with orthonormal polynomials is used for computing the required coefficients within the process of the program itself. This procedure takes up less than 15% of the SUPERTOG machine time. (authors)

[en] A numerical method has been developed for obtaining of the Legendre polynomials expansion coefficients of the anisotropy density function of the elastic neutron scattering. The method uses natural cubic splines as a basis for representing of the anisotropy density function and the Legendre polynomials. The procedure is programmed and incorporated into the SUPERTOG package for preparation of multigroup neutron data

[en] The neutron transport and the spatial angular and energy distribution have been examined of the neutron flux at the outlet of an experimental iron collimator of the 14 MeV neutron generator D-150 kV SAMES. The MORSE Monte Carlo Transport Code System is used based on 25 group constants. The results show that in addition to unscattered 14 MeV neutrons the outlet neutron flux consists of substantial group of elastic scattered and (n,2n)-scattered neutrons. It was established that the outlet flux could be considered as 1.2% homogeneous, 14% monodirectional and 7.3% monochromatic. These errors have a slight effect when the detector is in the close proximity to the neutron source, but they are increasing when the detector is remoted from the source. On the base of 3 approximation models a comparison is made with data for unscattered and scattered flux when detector is located on 100 sm from the outlet. It is shown that the complete problem simulation (isotropic source, collimator, shield, air, detector) is in a closest correlation with the experimantal data

No abstract available