[en] The paper presents Monte-Carlo simulations of characteristics of the scattered radiation produced in the target-moderator system to be used in the slow positron source when irradiated by synchrotron radiation beam from SPring-8 storage ring with superconducting 10 T wiggler

[en] Integral solution enthalpies of lithium perchlorate in diethyl ether, acetonitrile, acetone and DMSO at 298,2 K are determined calorimetrically; they are equal to -6.2, -9.1, -15.9 and -18.0 kcal/mol correspondingly. For diethyl ether dependence of integral solution enthalpy on salt concentration within the range from 3 x 10^-3 up to 5,6 mol/l is determined. Reducing exoeffect of salt dissolution in the ether from -6 up to -5 kcal/mol in the range of small concentrations (3 x 10^-3 - 3 x 10^-2 mol/l), the constancy of the effect in the 5 x 10^-2 - 1.3 mol/l range and following smooth reducing exoeffect up to -3.2 kcal/mol at salt concentration 5.6 mol/l is pointed out

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[en] We developed a full-scale 3D model of the initially fueled core of the VVER-1000 water-moderated power reactor for criticality calculations by the code PRIZMA. The core contains 163 hexagonal fuel assemblies of different types. The model includes blocks of in-core detectors which are arranged in the central pipes of 64 fuel assemblies in the entire core. Each block unites seven detectors equally spaced in the fuel assembly length. The sensing element of the detectors is rhodium emitter shaped as a slim cylinder. Conversion functions which relate the number of fission reactions in fuel elements closest to the central pipe and the number of neutron absorption reactions in the rhodium emitter were obtained through PRIZMA calculations where we estimated the rate of neutron adsorption reactions in the rhodium emitter and the rate of fission reactions in six fuel elements closest to the detector. The emitter's volume is negligible compared to the core volume and neutron flux in the detectors can hardly be estimated through analog Monte Carlo tracking. We used splitting and Russian roulette to make calculations more efficient. With these methods statistical uncertainties in the calculated rates of neutron absorption in rhodium were acceptable. In each of the 64 fuel assemblies with in-core detectors, we defined a set of cylindrical surfaces for splitting. Their axis coincided with that of the central pipe. When a particle crossed any of the surfaces to the cylinder axis, it was split into a number of identical particles. The Russian roulette was applied to particles which crossed the surfaces back. At collision points, particle weights were corrected in accord with a one-dimensional weight function defined in the entire volume of the fuel assemblies

[en] The SERA treatment planning system is being adapted to the NG-12I fast neutron facility used at Snezhinsk Neutron Therapy Center, Russia. This requires a computer model of the neutron facility that would match the SERA input format and provide adequately accurate results within a reasonable time. The detailed description of the facility cannot be used in SERA calculations because of its awkwardness, long time for dose calculation, and the requirement that the source must be defined as an energy-angle distribution of neutrons on a plane. First we made the detailed description of the neutron facility in the format of input data for the universal code PRIZMA which is a main tool for Monte-Carlo neutron transport simulations at RFNC-VNIITF. The detailed model of the generator was verified through comparison between calculated and experimental data. Then we considered a number of simplified models in which the neutron source was defined as a source on a plane. The source plane was defined in three positions: at the collimator's outlet, at the collimator's inlet, and near the real source. Only the third model provided a good agreement of results. Further analysis into the importance of neutrons in different regions of the third model showed the regions where they were of low importance that allowed us to limit the neutron tracking region by a truncated cone and significantly reduce the dose calculation time. Doses calculated along the neutron beam axis by PRIZMA, SERA and MNCP5 agree well. (authors)

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[en] Highlights: • PRIZMA is a Monte Carlo code for solving fixed source and criticality problems. • Particles: neutrons, gammas, electrons, positrons, ions. • Application fields: nuclear and radiation safety; pulsed, thermal and fast reactors. • PRIZMA is a part of package for precision core simulation through reactor campaign. • Special techniques for perturbation analysis. - Abstract: For more than thirty years the code PRIZMA has been used at RFNC-VNIITF for solving radiation transport problems with the Monte Carlo method. The code models the separate and coupled transport of neutrons, photons, electrons, positrons and ions in one-, two-, and three-dimensional geometry. For criticality calculations the code implements the method of generations with a constant number of fission sites in one generation. Now the code is extending its capabilities for nuclear reactor calculations. The paper describes the current status of the code and gives examples of its application to particle transport in nuclear reactors and other physical facilities

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No abstract available