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.
The present paper considers the computational simulation of experiments performed on the BFS-88 critical assembly, which is a full-scale model of the BREST-OD-300 reactor core. Experimental research included the study of neutronic performance core characteristics, such as criticality and reactivity worth of the control and protection and passive feedback systems, as well as the measurement of the spatial distribution of power density and reactivity effects. Radiation characteristics and neutron kinetics of the assembly were examined. The computational simulation of the measurements substantiated the neutronic performance characteristics of the BREST-OD-300 core and verified the design codes.
The development of advanced structural materials is a prerequisite for the feasibility of next-generation reactor projects. Current widely used steels and alloys fail to fully meet the requirements of these projects, while most innovative solutions are limited by scalability and manufacturability. This paper presents an approach of the NIKIET JSC to the development of structural materials for advanced reactor systems. Using this approach, EK199-VI nickel alloy was developed as the main structural material of the molten-salt research reactor, along with advanced EK205-Sh high-strength austenitic steel for operation in heavy liquid-metal coolants. The results of testing confirm the materials meeting specified requirements and demonstrate their technological feasibility.
The paper reviews the status of work by the NIKIET JSC (Moscow, Russian Federation) in the area of developing fuel assemblies (FAs) for various types of research reactors under construction and operation in our country and abroad. For the PIK high-flux reactor, advanced fuel and burnable absorber rods, as well as FAs were designed and justified. Their application will extend the operation of the reactor at rated power with the required process characteristics. An advanced modification of the VVR-M2 FA was developed and justified for the RB-01 reactor. The set of computational and experimental justifications confirms the performance of developed FAs in relation to its operating conditions. Advanced modifications of IRT-4M and VVR-M2 FAs were confirmed feasible for use in operated research reactors. The main results of the work on the performance justification of an advanced FA design with plate fuel elements are presented.
The NIKIET JSC (Moscow, Russian Federation) develops a blanket design for producing nuclear fuel in a hybrid fusion-fission plant. The present paper reviews the issues and challenges solved within the framework of these activities. A set of issues, both physical and engineering, connected with the approaches to the design, organization, and functioning of isotope-producing blanket is analyzed. Main conclusions on the required design approaches to the maximum use of the hybrid fusion-fission plant potential are provided. The estimated rates of producing 239Pu and 233U in a blanket based on 238U and 232Th raw isotopes, respectively, are given. In addition, the paper presents considerations regarding the strategic development line of hybrid fusion-fission, leading to the development of an industrial nuclear fuel production facility, including the stages of experimental development of the relevant technologies based on experimental and pilot-industrial plants.
The article substantiates the need for the development of a small modular reactor (SMR) plant for decentralized power supply to remote regions. Key 2022–2024 R D results on the development of the engineering design of the SHELF‑M reactor system for the SMR plant are presented. A computational and experimental substantiation of the design solutions adopted for the reactor plant, systems and equipment, and structural materials was carried out. The results of the coupled neutronics and thermal-hydraulic calculations are provided. The layout solutions of the experimental SMR plant site and key areas for increasing its competitiveness, e.g. the use of robotic complexes and AI-based remote-control systems, are presented.
The construction of a nuclear facility with a multipurpose fast research reactor assumes engineering challenges that require prompt and effective solutions. A wide range of work was carried out to ensure the MBIR project timelines: criterial dependencies for thermal-hydraulic processes in the reverse steam generator were experimentally obtained; compliance of the reactor plant with the requirements of federal regulations was justified in the course of a deterministic and probabilistic safety analysis; a program for trial assembly of the reactor was developed. Successful implementation of these activities will allow the plant to be commissioned within the established deadlines.
The application of nuclear energy for hydrogen production is considered in strategic documents of both Russia and other countries. A comparative study of scenarios for deployment of nuclear-hydrogen production allows its potential to be analyzed and preferrable options to be selected. To test the methodology for selecting the most preferred scenarios for the deployment of nuclear-hydrogen production. Modeling of scenarios for deployment of nuclear-hydrogen production in Russia up to 2050 and their comparative assessment were carried out taking into account the varying importance of the evaluation criteria. The realized methodology for selecting scenarios is based on a number of multi-criteria analysis methods and involves preparing a set of evaluation criteria, forming several sets of criterion weights, calculating aggregated indicators, step-by-step selection of scenarios, analyzing the uncertainty of criterion values and sensitivity of the final selection of scenarios. 30 scenarios were considered, modeled on the basis of combinations of two nuclear energy sources (electricity from a nuclear power plant, thermal energy from a high-temperature gas-cooled reactor (HTGR)) and three hydrogen production methods (water electrolysis, methane pyrolysis, steam methane reforming). A quantitative assessment of the scenarios was carried out according to 14 criteria. The conducted modeling of scenarios allows the potential of nuclear-hydrogen production in the energy system to be analyzed, taking into account its structure, volumes, resource needs, etc. A comparative assessment of the scenarios will be presented in the next publication, “Modeling and comparative assessment of scenarios for deployment of nuclear-hydrogen production in Russia. Part 2”.
The global consumption of electric energy is expected to exceed 3
Safe and efficient decommissioning of nuclear facilities represents an important strategic task. It requires a unified approach to developing and implementing innovative technologies, including robotic complexes, as well as systems for big data, machine learning, and artificial intelligence. To conduct an analysis and propose a set of standard technologies, equipment, methods, and approaches independent of reactor design and operating characteristics for developing a standard technology for serial decommissioning of nuclear facilities. The present paper considers the methodology for the development of a unified line of domestic robotic systems (RS) for standardized operations used in the decommissioning of nuclear facilities. A set of tasks forming the requirements for a unified line of RS is defined; limiting factors, process features, and requirements to RS are established. As applied to decommissioning and handling radioactive waste from nuclear facilities, such robots can perform a quality examination of facilities, limit the spread of contamination, perform remote operations with high accuracy, and reduce the impact of human factors. Contemporary domestic software and hardware solutions increase the resilience of robotic systems to external impacts and their ability to work in the face of constraints, e.g. increased radiation background, tight and difficult-to-reach places. A unified line of RS will increase the safety of completing a known set of repeatable tasks during the decommissioning of nuclear facilities, reduce the labor and dose costs of personnel, accelerate the transition to robotic systems, and implement the strategy for scientific and technical development of the Russian Federation. The methodology for the development of robotic solutions in complex tasks, such as handling of in-vessel devices of VVER and graphite stack of RBMK reactors, as well as the results of analyzing these solutions, will be presented in the article “Specialized robotic technologies for high-activity equipment and waste handling in nuclear facilities decommissioning”.
Over the next decade, the Russian Federation is planning to construct a research molten salt reactor. The purpose of this reactor is to demonstrate the feasibility of burning minor actinides with molten salt nuclear fuel. The advancement of innovative nuclear technologies requires the concomitant development of a regulatory framework. To ascertain the principal specific features that must be considered when enhancing approaches for the safety regulation of molten-salt reactors, in order to provide scientific and technical support to Rostechnadzor. The Scientific and Engineering Centre for Nuclear and Radiation Safety analyses international approaches to safety regulation and evaluates the suitability of current federal rules and regulations in the field of atomic energy for the safety regulation of molten-salt reactors (MSRs). It has been determined that the use of molten nuclear fuel endows the MSR with specific features in terms of physical barriers to the spread of radioactive substances. Thus, the molten salt nuclear fuel circuit acts as a second physical barrier, similar to fuel rods in solid-fuel reactors. This makes it necessary to determine alternative approaches to establishing and justifying limits for nuclear fuel damage in molten salt reactors. Furthermore, in MSRs, the intermediate circuit for cooling the nuclear fuel serves as a third physical barrier and must meet requirements similar to those established for the primary circuit in solid-fuel reactors. When developing approaches to safety regulation of MSRs, it is necessary to take into account the unique technical solutions implemented in modern designs, such as the possibility of emergency drainage of the fuel circuit and adjustment of the chemical composition of nuclear fuel during operation, taking into account the limitations on the solubility of components in the carrier salt, the high interdependence of the neutron-physical and thermohydraulic characteristics of the MSR, etc. The unique technical solutions implemented in MSR designs and the specific physical effects caused by the liquid state of nuclear fuel require principally new regulatory approaches to MSR safety regulation.
The safety of nuclear facilities (NFs) should be justified using qualified software tools. However, the procedure for validating thermal-hydraulic computer codes requires assessing the uncertainty and sensitivity of calculation results. To develop a technique for the correct and effective sensitivity analysis during the validation of thermal-hydraulic computer codes. For calculations, the KORSAR thermal-hydraulic code developed by the Alexandrov Research Institute of Technology (Sosnovy Bor, Russian Federation) was used; data of experiments on the true volumetric void fraction of subcooled pipe boiling (experiments of G.G. Bartolomei) were used for validation. The test calculations showed and justified the following necessary condition for the correctness of sensitivity analysis results: the input uncertainties set in the analysis of statistical characteristics should be equal. The performed study demonstrated the admissibility of excluding the input uncertain factors with the Spearman’s rank correlation coefficient less than |0.2| from the uncertainty analysis. This solution reduces labor costs for analyzing uncertainty. The efficiency of validating thermal-hydraulic computer codes can be increased by conducting a preliminary sensitivity analysis before the formation of validation matrices, as well as by excluding minor input uncertainties from the detailed quantitative assessment of statistical characteristics.
Improving the safety of nuclear power plants with water-water (VVER) reactors requires the development of passive heat removal systems (PHRSs), especially for complete blackout conditions. An important aspect of the research is the simulation of PHRS operation at different values of a decay heat flux. Integration of absorption refrigerators (ARs) into the PHRS is promising for increasing the efficiency of heat removal. To simulate the PHRS operating in integration with ARs and to assess the increase in the energy efficiency and safety of the VVER-1000 reactor in emergency conditions. The object of study is the PHRS: the number of its channels (two 8 MW each) is selected based on the single failure criterion for assessing the system performance in the conditions of partial degradation. The mathematical simulation of the PHRS under severe conditions including the failure of active safety systems, which corresponds to an accident with the loss of most energy sources, the equations of core heat balance (core), used natural circulation of the coolant and decay heat. Initial conditions for calculations included the coolant temperature of 60–200 °C and pressure of 6.0 MPa. The basic PHRS configuration (2×8 MW) fails to provide sufficient heat removal at a decay heat of more than 50 MW: in the first 2 h, a shortage of cooling to 40 MW is observed due to the inertia of starting natural circulation and a limited heat exchange area of 800 m2, which leads to an increase in the fuel temperature to 900 °C in 1.5 h. Insufficient PHRS power was confirmed. Additional safety systems are required for VVER reactors, especially under blackout. The simulation of AR operation in the PHRS and the impact of ARs on the system performance, accident development time, and core heating dynamics will be presented in the publication “Simulation of an absorption refrigerator operated in the passive heat removal system of a nuclear reactor. Part 2”.
Decontamination of concrete structures from radionuclides represents the most important task of decommissioning nuclear hazardous facilities. Relevant decontamination results obtained preferably without using radionuclide-contaminated building materials are needed to assess the effectiveness of developed decontamination methods. To develop a technique for simulating concrete surface layers contaminated with radioactive elements, including the subsequent assessment of element penetration depth. The object of the study involves concrete blocks saturated with specially selected radioactive simulants using an aqueous suspension at different durations of exposure. The research method is X‑ray fluorescence spectrometry, which is applicable to quantify the content of radioactive simulants in concrete layers depending on the distance to the surface treated with a suspension. The selected simulants suitable for aqueous suspension include SrO, Nd2O3, ZrO2, and Cs2MoO4; the exposure time of 1 h required to contaminate the concrete surface to a depth of 3 cm was established sufficient for decontamination studies. The developed technique can be used to simulate the concrete surface layers contaminated with radioactive elements, followed by the analysis of the composition of these layers and assessment of the element penetration depth for assessing the effectiveness of decontamination methods.
Background: In the Russian Federation, immediate dismantling was adopted as the main option for decommissioning nuclear power plants (NPPs) with RBMK-1000 reactors. Units No. 1 and 2 of the Leningrad NPP (Sosnovy Bor, Russian Federation) are the first of 11 units with RBMK-1000 reactors to be decommissioned in the coming decades. Aim: To summarize and present the results of developing the technological solutions for handling the dismantled equipment of RBMK-1000 reactors at Leningrad NPP units No. 1 and 2 decommissioned by immediate dismantling in two stages: pre-dismantling and dismantling. Materials and methods: A total of 3312 t of waste from one unit is planned to be dismantled and sent for processing and disposal in 8 years of the pre-dismantling stage, followed by 17,106 t, including 3078 and 483 t of intermediate- and high-level waste, respectively, in the next 20 years at the stage of dismantling. Technological solutions for handling dismantled equipment were developed using automated remote control and robotics complexes. Results: The safety of personnel handling materials of intermediate- and high-activity levels is ensured using remotely controlled robotic sites for fragmentation of equipment, placement of fragments in containers, decontamination of formed packages, and measurement of their characteristics before shipment. Estimated total capacity of equipment fragmentation and packaging sites is 5.4 t per shift; decontamination and container fingerprint assume a capacity of up to 12 t per shift. Conclusion: A project for decommissioning RBMK-1000 NPPs by immediate dismantling is being developed for the first time. Project implementation requires a significant number of robotics. Taking into account the analysis of domestic and foreign experience, as well as the calculated justification for the safe handling of the removed equipment, the project provides for the use of remotely controlled equipment. The publication “Development of technological solutions for decommissioning equipment and systems of separate rooms in the units No. 1 and 2 of the Leningrad NPP. Part 2” will provide a summary of the results on the development of technological solutions for dismantling the main equipment of the RBMK-1000 reactor plant, including the reactor, steam drums, and water pipelines.
The main requirements for small accelerator tubes include their service life and stability of operation. The magnetic field of the Penning ion source is one of the parameters that affect the characteristics of the tube. To study the operating modes of a Penning ion source at various parameters of the magnetic field for increasing the service life and stability of small accelerator tubes. Numerical simulation was carried out using the validated Magnetic Fields module of the COMSOL Multiphysics software and experimental verification of the calculated results on a vacuum bench with ion current photo recording. The magnetic system with the maximum cathode values of magnetic field induction, decreasing towards the anticathode, has the most beneficial operating modes of the Penning ion source: the set current level is provided at a lower gas pressure in the accelerator tube without compromising discharge current stability. The performed research proposes options of magnetic systems on the accelerator tube without ignition of a side discharge outside the discharge cell, which contributes to increasing the service life of a Penning ion source. The practical testing of the proposed magnetic system options will be presented in the following publication “Increasing the current characteristics of the Penning ion source in a small accelerator tube by changing the magnetic field. Part 2”.
