Issue 1/2009

UDK 621.039.573
PULSE REACTORS OF URANIUM OR NEPTUNIUM ALLOYS WITH CENTRAL CAVITIES 20–30 CM IN DIAMETER / V. F. Kolesov, С. V. Vorontsov, V. KH. Khoruzhy, V. V. Evdokimov, A. A. Smolyakov // P. 3
There was realized a new series of refined calculation of fast pulse reactors with large cavity, namely, reactors with a core of uranium-molybdenum or neptunium-gallium alloys and axial cavities their diameter being from 20 to 30 cm. The problem of using a concept of mobile reflector blocks in neptunium-gallium reactor to decrease mechanical stresses was considered.
Key words: aperiodic pulse reactor, fission pulse parameters, uranium alloys, neptunium alloys, mobile reflector blocks, cavity for samples irradiation.

UDK 621.039.571;539.1.084
INFLUENCE OF THE SPECTRUM OF GAMMA-QUANTUMS ON SIZE OF THE ABSORBED DOZE IN MATERIALS OF VARIOUS ELEMENT STRUCTURE / A. S. Koshelev, V. Kh. Khoruzhy // P. 17
Results of calculations of quantitative influence of a spectrum γ-quantums on the absorbed doze for the elected spectral representations and materials of applied orientation are submitted.
Key words: the spectrum of the gamma-quantums, the absorbed doze, spectral variability of the doze

UDK 621.039.571;539.1.084
PRECISION PECULIARITIES OF REALIZATION OF THE ABSORBED DOSE POWER FROM THE DELAYING GAMMA QUANTA IN THE REACTOR БР-К1 SPECIALIZED GAMMA SOURCE FIELD / A. S. Koshelev, V. Kh. Khoruzhy // P. 28
Experimental and calculated data on the γ-radiation dose power in the specialized area in the reactor БР-К1 cavity are compared. Correction variants of calculation results are offered.
Key words: reactor БР-К1, specialized γ-source.

UDK 539.1
MODELING OF A MULTILAYER SHIELDING CONSTRUCTION FOR STORAGE/TRANSPORTATION OF RADIOACTIVE WASTES CONTAINING FISSION ELEMENTS / M. V. Alenina, V. P. Kolotov, L. I. Ivanov // P. 36
A multilayer construction of container for storage/transportation of radioactive wastes has been suggested. The construction consists of body and internal enclosure both made from cast iron (or steel). The space between body and enclosure is filled by layers of polyethylene and boron. Estimation of neutron flux attenuation by means both method of removal cross-sections and analysis of dynamics of flux density decreasing while passing of different materials layers has been carried out. The coefficient of neutron attenuation is no less than 10⁴.

UDK 550.06
EVALUATION OF SANDY-GEL MATERIAL EFFICIENCY AS AN ABSORBING BARRIER IN THE MIGRATION PATHWAYS OF NUCLIDES / V. I. Sergeev, N. N. Danchenko, M. L. Kuleshova, T. G. Shimko, Z. P. Malashenko, N. Yu. Stepanova // P. 42
A complex research was carried out to estimate the efficiency of sandy-gel material as an absorbing barrier in the migration pathways of nuclides. The current importance of such evaluation is based on its broad application for building horizontal and vertical protection barriers at industrial waste disposal areas.
Key words: migration of nuclides, absorbing barrier, sandy-gel material, prototype chemical solutions.