Physics of Atomic Nuclei最新文献

A description of the PRIM-AES software package (NPP simulator program) for modeling neutronic and thermal-hydraulic processes in transient modes at power units with VVER reactors is presented. The package includes a dynamic-simulation environment for the implementation of complex algorithms for automatic control systems based on SimInTech CAD systems and the TIGR-M program for combined neutronic and thermal-hydraulic calculation of normal operation and emergency modes, which is a development of the TIGR-1 software package. The modernization of the thermal-hydraulic module of the TIGR-1 software package is described, which is carried out to extend the applicability of the code in the volume of the power unit. The structure of a new software-computing package with a description of its constituent modules and the interface for data exchange between them has been developed and presented. The goals and objectives of the application of the developed software package are determined, which allow assessing the relevant criteria of dynamic stability for stationary modes of normal operation and transient modes of normal operation, including power maneuvering modes, and operation limitation modes in the case of violation of safe operation conditions and emergency modes of the second category involving failures of the main equipment.

The applied technologies for handling tritium and the efficiency of its use in the fuel cycle of a fusion facility affect the start-up tritium inventory, the possibility of tritium self-sufficiency, etc. These parameters largely determine the cost of the facility, its efficiency, and safety during operation. Formation of the conceptual configuration of the fuel cycle, calculation of the tritium inventory in its systems, and determination of the scope for further development are important problems in facility design. In order to calculate the flows of fuel components and tritium inventory in the fuel cycle (FC) of a tokamak-based fusion neutron source, the FC-FNS code developed at the Kurchatov Institute has been used since 2013. It implements the possibility of calculating the number of hydrogen isotopes contained in different FC systems. The article describes the current state of research on modeling the flow of fuel components in FC systems of the tokamak with a tritium-breeding blanket, presents simplified schemes of FC systems, and describes the principles of their operation and the methodology for calculating the inventory of hydrogen isotopes in them. It is substantiated that the number of fuel components in the facility will be primarily determined by fuel cycle technologies, which, in turn, depend on plasma parameters and scenarios of facility operation. It is shown that, for the DEMO-FNS project with a fusion power of 40 MW, which corresponds to the combustion of 7 × 10^–5 g/s of tritium, the startup tritium inventory will be 400–430 g with a circulating fuel flow through the facility chamber of up to 0.1 g/s. Increasing fuel flows through the injection, hydrogen isotope separation, and some other DEMO-FNS systems, if convective edge-localized modes (ELMs) are taken into account, will lead to an increase in the operational tritium inventory in the fuel cycle up to 500 g. If there is a tritium-breeding blanket in the facility, tritium inventory in it (including the long-term storage) will be no more than 800 g.

The physical characteristics of stable formation of the limiting permissible thermo-electrodynamic states of superconductors with various types of nonlinearity in their current–voltage characteristics (CVC) are discussed. The analysis was based on describing the CVC using a model that assumes its continuous increase with varying degrees of smoothing (different intensities of increase at a fixed value of the critical current). A power-law equation was used to describe it. The results were compared with those from numerical experiments simulating thermo-electrodynamic states of superconductors with a zero-voltage subcritical region and an idealized CVC (nonsmoothed CVC), described by a piecewise continuous equation derived from the viscous flux-flow model. It is shown that an increase in the smoothing of the CVC of a superconductor, under otherwise equal conditions (at a fixed value of critical current density and cooling conditions), is accompanied by a decrease in its current-carrying capacity. Its degradation is caused by a corresponding increase in heat losses, which inevitably exist owing to the continuous increase in the CVC of the superconductor throughout the current injection process. As a consequence, the values of the limiting permissible currents stably flowing through the superconductor with a nonsmoothed CVC, under otherwise equal conditions, are higher than the corresponding values calculated for superconductors with a smoothed CVC. This feature is observed despite the different nature of current filling the cross section of the superconductor. It is proven that, for the correct determination of the current-carrying capacity of superconductors, the permissible values of temperature and electric field intensity preceding the onset of current instability cannot be predefined. They depend on the degree of smoothing of the CVC, the current injection rate, the transverse size of the superconductor, and the conditions of cooling. As a consequence, there is a nontrivial relationship between the maximum allowable losses and the maximum stable value of injected current. These features must be taken into account when experimentally measuring the CVCs of superconductors and their current-carrying capacity.

DEMO-FNS is a 40 MW tokamak-based hybrid (fusion–fission) facility being designed in Russia. To date, the architecture of the tritium fuel cycle (FC) has been formed, and the requirements for the main technological systems have been determined. The parameters of the FC systems have been obtained using the FC-FNS electronic code developed by the project team. In order to create the DEMO-FNS facility, it is necessary to determine the current level of development of FC technologies to assess the possibility of their application. The article continues the analysis of readiness of FC technologies of the DEMO-FNS facility in Russia started by the authors earlier. For the analysis, the technology readiness level (TRL) methodology is used, according to which technologies in the target application area correspond to different readiness levels from TRL1 (basic principles observed and reported) to TRL9 (technology is verified by successful operation in the target application area). The following technologies are considered in the article: pellet injection, neutral beam injection, gas injection, tritium breeding in the blanket, tritium extraction from the blanket, and sorption storage of hydrogen isotopes (HIs). It is shown that in Russia there is a significant potential (TRL4) in the field of technologies necessary for tritium breeding in the DEMO-FNS blanket, but at present the work is frozen. Injection technologies, especially pellet injection, are world-class (TRL4–5). HI sorption storage technologies are highly developed (TRL5–6) and are used in complex processes with large amounts of tritium. The readiness level of the listed technologies is insufficient for application in the FC of the DEMO-FNS facility. It is necessary to increase the level of readiness within research and development programs, to create specialized stands for testing and demonstration of technologies, and to create experimental fusion facilities for testing and integration of technologies.

In order to improve the accuracy of engineering calculations in the design of marine reactor installations with high energy resources, characterized by significant values of reactivity coefficients related to the coolant and axial offsets, a preliminary study of the method for refining the dependence of the reactivity density effect on the nature of the neutron field distribution has been carried out. The necessity of validating the methodology on the basis of comparing calculated and experimental natural temperature dependences is noted.

The article presents calculated and experimental assessments of the possibility of using Ni-rich RPV steels developed by the JSC NPO Central Research Institute of Machine Building Technology (CNIITMASh) as materials for RPVs of next generation reactors taking into account their structural-phase state, mechanical characteristics, and the possibility of manufacturing a shell with an industrially developed wall thickness. In order to improve service characteristics, it is necessary to apply a complex of measures: change the complex alloying of steels, conduct metallurgical purification from impurities (primarily from phosphorus), and optimize the grain size of castings. The calculation took into account the assumed operating temperature, coolant pressure, strength grade of candidate steels, and their thermal and radiation resistance.

The article presents an overview of the existing domestic and foreign practices of using artificial intelligence technologies for designing, safety assessment, and operation of nuclear facilities. The concept of artificial intelligence is interpreted in a general way, covering a whole range of information technologies and software and computational methods. Today, there is a growing interest around the world in the use of artificial intelligence (AI) technology in almost all technological areas. Nuclear energy, as an extremely science-intensive industry, has its own characteristics compared to the areas of mass application of AI (medicine, economics and finance, marketing, design, logistics, traffic analysis, etc.). The correct setting of tasks for the use of AI in the nuclear industry requires a clear definition of the possibilities and limitations of the use of AI. In this study, the authors analyze various aspects of the use of AI for designing, safety assessment, and operation of nuclear reactors. The main attention is paid to the opportunities for the development of the industry and improving the efficiency of technological processes. Examples of the development and testing of methods based on AI in the field of activity of OKB Gidropress JSC are given. Conclusions are drawn about the promising areas of using AI as a modern information technology, and as a development direction for the long term.

Investigation of discharges with a negative triangularity plasma configuration is carried out on various tokamaks: TCV, D-IIID, and ASDEX-U. Negative triangularity (NT) experiments in the TCV show a reduction in electron heat transport by a factor of two compared with the positive triangularity D-shaped configurations. Recent experiments on DIII-D with a NT configuration showed improved confinement compared to the positive triangularity (PT) D-shaped plasmas over a range of auxiliary heating powers and, in particular, for the case T_e ~ T_i. In addition, it was found that the NT-shaped plasma has the capability to achieve significant normalized β. The purpose of this study is a computational confirmation of the possibility of implementing scenarios with NT discharges in the ohmic heating mode in the T-15MD tokamak with its standard poloidal system. Simulation results show that the poloidal system of the T-15MD tokamak is rather flexible for study of scenarios with D-shaped NT. It is proposed to expand the research program in the T-15MD by including in it the study of scenarios with NT.

A method of taking in account heterogeneous and kinetic effects in the preparation of homogenized group neutronic constants for the diffusion model is described. The results of the application of the proposed technique in the preparation of homogenized group neutronic constants for a fast reactor of type BREST-OD-300 are presented. Consideration of heterogeneous and kinetic effects in the process of homogenization is shown to enhance the accuracy of calculating integral parameters and distributed characteristics that are of importance for safety.

Electron cyclotron (EC) resonance heating in the T-15MD tokamak is calculated using the GENRAY code. The injection schemes without current generation for heating plasma using an extraordinary wave with a frequency of 140 GHz at the third harmonic of the fundamental EC frequency are considered. The cases of the wave injected through vertical, upper inclined, and equatorial port-plugs for three scenarios of the T-15MD operation—the baseline configuration and two configurations with different moderate values of plasma elongation and triangularity—are analyzed. A comparison of calculations performed using the GENRAY code with the published calculations based on the OGRAY code shows that the efficiency of EC heating in single-pass absorption and the spatial position of the absorbed power maximum are in good agreement, while the shape of the spatial profile of the absorbed power can differ significantly.