SOME ISSUES OF MODELING OF DESIGN TECHNOLOGY OF NEW REACTORS BASED ON THE PHYSICOCHEMICAL PROPERTIES OF MOX-FUEL

To the issue of modeling of design technology of new reactors based on the physicochemical properties of MOX-fuel, the authors considered the existing mathematical models for estimation of neutronics and radiation characteristics of the main experimental volumes using the example of the designed fast research reactor. The efficient use of MOX-fuel is achieved when combusting it in fast reactors; and its production is possible by processing irradiated fuel of power reactors. The use of MOX-fuel in existing reactors requires significant alterations (introduction of more control rods), but it will be possible in full in specially designed MBIR reactor. One of the attractive properties of MOX-fuel is that during its production it is possible to dispose the extra amount of weapon grade plutonium, which otherwise would be the radioactive wastes. For central serpentine passage, the authors get the estimations of average and maximum neutron flux density, axial distribution of neutron flux density that is actual for the issue of radiation resistance of the materials used in modern reactor engineering. The design model is developed on the basis of MCU applied software package (MCU-RR2 version) implementing the paradigm of Monte-Carlo method when drawing the traces of neutrons and gamma-quanta in 3D geometry for mutual simulation of neutrons and photons flux functionality in the research nuclear reactors, basing on the estimated nuclear data. 

For vertical experimental channel, the authors determined the dependences of specific induced activity and induced activity of commercial nitrogen gas on the operation time of reactor in order to ensure the cost-effectiveness of the materials used for the cooling of channel supposed for the nuclear doping of silicon. Simulation of the cooling environment activation was carried out by means of mathematical modeling of kinetics of nuclide transformations according to the UPM-PREPRO_2007-FENDL-2.0-ENDF/B-VII.0 software complex, where PREPRO_2007 is the package of utilities for the preprocessing of nuclear data in ENDF/B format.