Atomic Energy最新文献

The article considers an in-situ method potential for establishing the surface distribution of tritium in the first wall of a fusion reactor using neutron tomography. The method includes the formation of a pulse-periodic flux of accelerated deuterons bombarding the studied surface of the first reactor wall following its saturation with tritium, generation of a fast neutron field according to T (d, n)⁴He reaction, measurement of neutron fluence at given spatial points, and reconstruction of the spatial tritium distribution according to measurement data. An algorithm for such a reconstruction is presented based on an approximate solution to the inverse Fredholm problem.

The paper presents experimentally determined hydrodynamic characteristics of the coolant flow at the outlet of the RITM nuclear icebreaker reactor fuel assembly. The studies were carried out according to Pitot tube measurements based on the theory of hydrodynamic similarity using a model of the fuel assembly outlet with an air working medium. Friction factors were determined for several sections located at the outlet of the fuel assembly. The distribution of the hydrodynamic characteristics of the coolant flow is visualized by cartograms of the axial velocity at the outlet of the fuel rod bundle, in the wide and narrow parts of the nozzle, as well as in the fragments of drain windows in the upper base plate. The flow of the coolant through the central displacer and pipe for conducting the coolant to the resistance thermometer is determined. According to the performed studies, the asymmetric nozzle, coolant sampling pipe of the resistance thermometer, and upper base plate are the essential parameters affecting the formation of the field structure for the axial flow velocity.

The complex task of integrating various information, calculation, and simulation tools into a single system can be implemented using digital twins capable of optimizing design, engineering, and technological solutions, as well as supporting the ongoing operation of facilities. This article presents the results obtained within the development of integrating projects for the facilities of the “Proryv” project area, as well as discussing tools and solutions for creating digital twins, including integral mathematical models as a computational representation of the facility. The main directions for the further possible development of digital twins are considered relative to the construction of facilities, support for their operation, replication of digital twins themselves, and the search for innovative solutions that contribute to improvements in the quality both of the digital twins themselves and the actual created facilities.

The article presents the results of using a multifunctional laser module for remote separation cutting of MIR reactor loop channels. The working conditions and parameters for cutting radioactive equipment are described. Alternative methods, as well as the risks associated with the considered method are analyzed.

The article examines the effects of duration and movement mode of the control rod of the VVER-1200 control and protection system during its fast ejection from the core on the results of accident simulation. Two modes of movement are considered including uniform and equi-accelerated modes, in which the duration of the bottom-stop ejection varies from 0.01 to 5 s. The ejection of the control rod is simulated for reactor states of a full and minimum-controlled power, as well as with and without leakage for the first six 18-month VVER-1200 fuel campaigns. Calculations were carried out using the ATHLET/BIPR-VVER code (version 1.0). The performed studies demonstrated insignificant sensitivity of extreme fuel temperatures and departure-from-nucleate-boiling margin to ejection parameters. The approximated uniform movement of the control rod during 0.1 s of ejection is sufficiently conservative for use in the deterministic safety analysis of VVER nuclear power plants.

The paper considers the concentration of ^95–97Mo in a square cascade with large stage separation coefficients corresponding to gas centrifuges. Cascades were calculated using a method for varying the sections of partial stage streams and minimizing the deviation of calculated stage feeding streams from the specified one. A computational experiment on the multi-step separation of a molybdenum hexafluoride mixture was carried out. Isotopes of ^95–97Mo can be efficiently concentrated in a single square cascade. In streams depleted of these isotopes, isotopically modified molybdenum is accumulated for use as a structural material for fuel element claddings having improved thermophysical characteristics.

The paper describes the development and verification of the RING 2.0 software designed for thermal-hydraulic calculations of cores and fuel assemblies with annular fuel elements. Results from a set of experimental and theoretical studies carried out on the basis of a computational model embedded in code to verify the RING 2.0 software regarding the cores of research and isotope production reactors are presented.

Based on a conducted literature review, a training sample was compiled. The selected optimal parameters for training an artificial neural network included its structure, activation function, output layer transfer function, and the number of neurons in hidden layers. The results of calculations using the developed artificial neural network have an uncertainty of less than 1%, which confirms its suitability for creating a digital twin of a technological process.

The work was carried out as part of the computational and experimental justification of the operability of U and U‑Gd fuel rods of the VVER reactor in cycling power operating modes as part of the process of licensing domestically-produced fuel according to the requirements of Russian and foreign supervisory authorities. A methodology for forming the requirements for the rate of change in the local linear power of VVER reactors during tests in the research reactor simulating the cycling mode was developed.

The present article reviews the innovative MNT-CUDA software package designed for engineering calculations of the neutron-physical characteristics of RBMK-1000 reactors according to the multi-group Monte Carlo method. An essential feature of the software is the ability to conduct multi-group fuel element calculations of the entire reactor volume with the restoration of in-core sensor data using the Monte Carlo method. Thanks to NVIDIA CUDA GPU parallel computing technology, a counting speed of about 0.2–1 billion neutron events per minute is achieved on a personal computer equipped with a state-of-engineering video card. The history of creating the software package from the first version to the latest at the moment, as well as some results of verification and certification are given. A brief information on the methods used in the software package is provided. The concept of building its complex architecture, the prospects for development, and the results of testing in the calculations of fast and uranium-water systems are set out.