Sergej Vladimirovich Bedenko

The study focuses on modeling a device designed for conducting experiments under low reactor neutron fluence rate conditions. The relevance of the work is driven by the challenges of accurately measuring low reactor neutron fluence rates, which are critical for the safety and efficiency of nuclear reactors and for conducting research in various scientific fields. To create a directed neutron flux, a horizontal collimation channel of the Prizm-AN irradiation device with a radioisotope capsule neutron source of the internal bounded neutron (IBN) type is being investigated. The paper presents the results of particle transport modeling using the PHITS and SOURCES-4C codes, which allowed for the evaluation of the generated neutron fluence rate parameters and the effective dose rate of radiation. The obtained results confirm the functionality of the proposed device and its compliance with radiation safety requirements, making it suitable for conducting experiments similar to those performed on reactor installations.

 Abstract. The VVER-1200 (V-491) reactor is a water-cooled power reactor, the design of which provides for higher fuel and coolant operating parameters compared to the VVER-1000 (V-320) reactor. For long-term and trouble-free operation of the reactor, the fuel is modified by adding various homogeneous compounds and heterogeneous inclusions. Uranium-gadolinium fuel in a homogeneous design with axial profiling of fuel elements has received practical application. The possibility of heterogeneous use of Gd₂O₃, ZrB₂, Am₂O₃ and other burnable and alloying additives is being investigated. Such additives make it possible to maintain the thermal conductivity of the fuel at the level of conventional oxide fuel. The studied modifications show satisfactory behavior under irradiation at extremely high temperatures and burnup. However, the issues of radiation safety when handling both fresh and spent fuel remain less studied. In this work, a computational assessment of the neutron component of the radiation characteristics of a UO₂ composition with a heterogeneous variant of the localization of ^natGd₂O₃ and Am₂O₃ microcapsules was carried out. This design option does not impair the thermal conductivity of the fuel and has a positive effect on the nuclear physical and thermophysical properties of the fuel. Americium has been studied not only as a possible alternative to Gd, but also from the perspective of its possible utilization in thermal reactors. The influence of Am on the photon component of the radiation characteristics of fresh fuel is considered. It is concluded that the radiation safety of fresh and irradiated products containing Am should be achieved primarily by solving problems of protection from photon radiation. The research is carried out to develop procedures and regulations for handling new fuel during its manufacture and after irradiation in the reactor. The studies were carried out using verified calculation codes of the MCNP 6.2 and Nedis 2m programs.

For long-term and safe operation of the reactor, nuclear fuel is modified by doping with various homogeneous compounds and heterogeneous inclusions. Such modified fuel has demonstrated satisfactory performance when irradiated at elevated temperatures and high burnup. However, the issues of radiation safety when handling modified fuel remain less studied. Elements of low and medium atomic mass are often targets for (α, n) reactions, so their use as alloying additives, as well as the use of α-emitting additives, can complicate the radiation situation at the stages of the nuclear fuel cycle. In this work, the neutron components of the radiation characteristics of UO₂ with Gd₂O₃ and AmO₂ additives were analyzed. Americium has been investigated as a possible alternative to gadolinium. In fuel containing AmO₂, the neutron yield is higher compared to Am-free fuel and is formed mainly by (α, n) reactions in AmO₂ in fresh fuel and spontaneous fission of ²⁴⁴Cm nuclides in spent fuel. The research was carried out with the aim of developing procedures and regulations for handling new fuel during its manufacture and after irradiation in the reactor. This work contributes to the study of the neutronic and radiation characteristics of Am-containing fuel, which has the potential for use in modern reactors. Calculations were performed using verified computational codes SOURCES-4C and WIMS-D5B.

The current state of research in the field of nuclear and thermonuclear power aimed at creating power generation plants makes it possible to predict the further development of modern power industry in the direction hybrid reactor power plants. Such hybrid systems include a tokamak with reactor technologies, worked out in detail in Russia, and systems with an additional source of neutrons. Power generation plants using tokamaks and accelerators with the required level of proton energy will be of exceptionally large size and power, which will postpone their construction on an industrial scale to the distant future. The ongoing research is aimed at the development of small generation and has the prospect of entering the field of energy use in a shorter period. The hybrid reactor facility under study consists of an axisymmetric assembly of fuel blocks of a high-temperature gas-cooled reactor and a linear plasma source of additional neutrons. The paper demonstrates the results of optimization plasma-physical, thermophysical and gas-dynamic studies, the purpose of which is to level the distortions of the power density field, which are formed in the volume of the multiplicating part of the facility due to the pulsed operation of the plasma source of D-T-neutrons. The studies on increasing the “brightness” of the source and modeling its operating modes were carried out using the DOL and PRIZMA programs. The thermophysical optimization and gas-dynamic calculations were performed using the verified SERPENT and FloEFD software codes. The calculations were made on a high-performance cluster of the Tomsk Polytechnic University.