Radioactive isotopes of krypton and xenon comprise the bulk of the radiation dose from the molten-salt reactor, which is the closing link of the nuclear fuel cycle, intended for the transmutation of minor actinides. The sorption on modified coal is recognized as the most effective method for localizing krypton and xenon. To develop a unit for sorption recovery and activity suppression of radioactive inert gases in the gas treatment system of the research molten-salt reactor (RMSR); to determine the sorption capacity of various modified coal sorbents at various temperatures of the gas treatment system. The microstructural properties of coal were assessed using the method of low-temperature nitrogen adsorption-desorption according to the Brunauer–Emmett–Teller theory, as well as porosimetric measurements on the surface analyzer. A volumetric method was used to assess the sorption capacity of various modified coal sorbents for xenon and krypton of the natural isotope composition. The ability of all studied modified coal sorbents to sorb krypton and xenon is demonstrated exponentially increasing at decreasing temperature. The adsorption of gases, independent of the true density, only indirectly depends on the specific surface area of the sorbent. Henry solubility of sorbent is directly proportional to the specific surface area of its micropores. Among the examined modified coal sorbents, VSK‑5 and VSK-5IK are determined as the most promising for the adsorption of krypton and xenon in the gas treatment units of the molten-salt reactor. They have the highest values of Henry solubility, adsorption constants, and sorption capacity compared to other studied sorbents. The calculations of the technical parameters for the reactor gas treatment system will be presented in the following publication “Sorption recovery of xenon and krypton from the carrier gas in the gas treatment system of a molten-salt reactor. Part 2”.
Despite the successful operation of the BN-800 reactor during 2015–2025 and its transition to full core loading with MOX fuel pellets, no materials science data are available on the actual state of fuel assemblies and fuel rods irradiated in this reactor. Thus, it is advisable to use the data archive of the Institute of Reactor Materials JSC (Zarechny, Sverdlovsk Oblast, Russian Federation) on the post-reactor state of BN-800 fuel rod prototypes, which were operated in 2010–2012 under conditions of a lower power BN-600 reactor. To present the individual results of post-reactor studies of BN-600 model fuel rods (BN-800 fuel rod prototypes) that retained their integrity after irradiation in the current conditions of BN-800. BN-600 model fuel rods have the structural and geometric design of the BN-800 core with a cladding of advanced Cr16-Ni15 steel and a pellet core made of MOX fuel bushings. The maximum values of the damaging dose, burnup of materials, and the temperature of fuel claddings were 84.7 dpa, 11.6
The article covers the development and implementation of integrated systems of technical diagnostics and non-destructive testing set in the design of nuclear reactors, as applied to the projects of the NIKIET JSC (Moscow, Russian Federation). These systems can ensure the necessary level of safety and operational efficiency for nuclear power facilities. In addition, the article considers the condition monitoring system (CMS) developed by the NIKIET JSC. The system is based on a holistic approach to diagnostics, non-destructive testing, and endurance management of critical equipment and pipelines throughout their life cycle. The solutions developed within the framework of implementing the CMS concept are described to show the versatility both for innovative projects and modernization of existing nuclear power plant units.
The present article examines the development and justification of an innovative reactor technology based on the BEST-OD300 fast lead-cooled reactor with inherent safety and mixed uranium-plutonium nitride fuel for operation in a closed nuclear fuel cycle. The article provides the results of additional calculation and experimental studies performed for the reactor coolant pump set, steam generator, and steam generator leak containment system, as well as of neutronic performance studies on the BFS critical test bench and in the IGR reactor, long-term endurance tests of a full-scale fuel assembly model, in-pile tests of fuel rods, as well as systems of automatic instrumentation and control and cladding failure detection. A commercial BR-1200 reactor is demonstrated to successfully combine the BREST-OD-300 engineering solutions with some advances.
The paper considers the R D program to justify the molten-salt research reactor (MSRR), which is considered as a prototype of a reactor for burning long-lived actinides, mainly americium and curium, from spent fuel of power reactors. The selection of EK199- VI alloy as a structural material was justified; the technologies for the preparation of molten-salt media and physical properties of salt melts were determined; the initial data for the verification of calculation codes were formed; experimental facilities for the development of molten-salt technologies were constructed. The analysis of the main federal rules and regulations in the field of atomic energy use in relation to the MSRR facility was carried out with the formulation of the required changes to the regulatory documents. The results of R D indicate the fundamental possibility of substantiating design solutions and safety of the MSRR facility.
