[en] Some new nonstandard techniques and means for detecting hidden charges (explosives) are described. (author). 4 refs., 6 tabs., 9 figs

[en] Short communication

[en] We discuss the diffusion equation as an asymptotic limit of the neutron transport equation for large scattering cross sections. We show that the classical asymptotic expansion procedure does not lead to the diffusion equation and present two modified approaches to overcome this difficulty. The effect of the initial layer is also discussed. (authors). 9 refs

[en] CNSKINO - a user friendly code for the quick assessment of nuclear reactor kinetics problems. The calculation of reactor power is based on effective solving point-kinetics equations by using third-order Hermite polynomial approximations of reactor power and analytical integration of delayed group precursor equations at every finite time step of the process. Problems of fast transient processes caused by reactivity accidents and slow transients for different type of reactors can be successfully evaluated using CNSKINO code. The CNSKINO is written in the algorithmic Turbo Pascal for Windows language. (author). 4 refs., 9 figs., 1 tabs

[en] The fast neutron fields of the RB reactor are presented in this paper. The activation method for spectrum determination is described and explained. The obtained results for intermediate and fast spectrum are given and discussed. (authors). 7 refs., 3 tabs

[en] Three areas where nuclear based techniques have significant are briefly described. These are: Nuclear material control and non-proliferation, on-line elemental analysis of coal and minerals, and non- detection of explosives and other contraband. The nuclear physics and the role of reactor physics methods are highlighted. (author). 5 refs., 10 figs., 5 tabs

[en] This document provides the calculation technique for the fuel elements which consists of the one substance as a matrix and the other substance as the corn embedded in it. This technique can be used in the neutron flux density calculation by the universal Monte Carlo code. The estimation of accuracy is presented too. (authors). 6 refs., 1 fig

[en] The delayed neutron parameters and methods used for calculation in reactor safety studies are verified by measurement of the effective neutron fraction in the coupled fast-thermal system 'HERBE'. (authors). 13 refs., 1 tabs., 3 figs

[en] The conventional type of the WWER-440 reactors have a very efficient in-core measurement system. The accuracy of the calculated power and flux distribution can be improved using of the signals of the in-core system. In this paper a method is presented for the determination of more exact flux or power distribution inside the reactor core than the simple calculation. We used the general perturbation theory to improve these distributions. The application of our method to get of more perfect relative axial fast flux is shown in this paper. (author). 4 refs., 2 figs

[en] The present work describes the activity performed in the framework of the validation of both calculation tools and nuclear data to use for simulation of the Los Alamos Actinide Transmutation Waste Project (ATW). The calculations have shown a reasonable agreement with available reference data which make us confident in the adequacy of the selected code chain and associated cross section file in simulating the main phenomena of the transmutation process. On the other hand, such validation results have highlighted some code inaccuracy and cross section data unrealistic behaviour which need of further investigations and updating activities. (authors). 14 refs