Nuclear-excited source of coherent and incoherent radiation with direct nuclear pumping.

Uranium fission fragments, as well as the products of He(n,p)H and B(n,α)Li nuclear reactions were utilized in the nuclear reactor for gas ionization and excitation. However, the Li(n,α)H nuclear reaction was less examined. The use of lithium-6 as a surface source of excitation of the gas medium, due to the long path length of tritium nuclei in the gas, allows to excite large volumes of gas as opposed to using U or B.

Calculation of Temperature Fields in a Lithium Ceramic Pebble Bed during Reactor Irradiation in a Vacuum.

Two-phase lithium ceramic LiTiO-LiSiO is considered as a tritium multiplier for use in the solid blanket of fusion reactors. To date, the most accurate understanding of the processes of tritium and helium production and release occurring in the breeder blanket materials under neutron irradiation can only be obtained from experiments in fission research reactors. At that, irradiations in vacuum give the possibility to register even very fast gas release processes (bursts) from the ceramics' voids and pores, although it reduces the thermal conductivity of the pebble bed.

Features of Helium and Tritium Release from LiTiO Ceramic Pebbles under Neutron Irradiation.

The operation of fusion reactors is based on the reaction that occurs when two heavy hydrogen isotopes, deuterium and tritium, combine to form helium and a neutron with an energy of 14.1 MeV D + T → He + n. For this reaction to occur, it is necessary to produce tritium in the facility itself, as tritium is not common in nature.

Study of Morphological, Structural, and Strength Properties of Model Prototypes of New Generation TRISO Fuels.

The purpose of this work is to characterize the morphological, structural, and strength properties of model prototypes of new-generation TRi-structural ISOtropic particle fuel (TRISO) designed for Generation IV high-temperature gas reactors (HTGR-type). The choice of model structures consisting of inner pyrolytic carbon (I-PyC), silicon carbide (SiC), and outer pyrolytic carbon (O-PyC) as objects of research is motivated by their potential use in creating a new generation of fuel for high-temperature nuclear reactors. To fully assess their full functional value, it is necessary to understand the mechanisms of resistance to external influences, including mechanical, as in the process of operation there may be external factors associated with deformation and leading to the destruction of the surface of fuel structures, which will critically affect the service life.