Atomic Energy最新文献

Fire protection is one of the most important issues to ensure safety and reduce risks of nuclear power plants. The main control room (MCR) is one of the parts of a nuclear power plant that is often identified with a high risk of fire in safety assessments. Although robust programs to shut down commercial reactors in any fires have been successfully maintained, the purpose of this paper is to simulate the fire in the MCR of Bushehr nuclear power plant unit‑1 (BNPP-1) using CFAST software. CFAST is a two-zone fire model used to calculate the evolving distribution of smoke, fire gases, and temperature throughout the compartments of a building during a fire. The data required for simulation have been extracted from final safety analysis reports and operating documents of BNPP‑1 and NUREG6850 reports. The results showed that the highest recorded temperature was related to the upper part of the standing cabinet about 1013 °C, which is the result of flames as well as heat transfer through smoke. Also, the other highest recorded temperature was in the lower part of the standing cabinet about 700 °C. Finally, the first ventilation system was closed for about 350 s and the second ventilation system was closed for about 860 s after the fire started.

A Lead-Cooled Fast Reactor (LFR) is one of the six advanced fourth-generation reactors that many countries have been interested in designing, constructing, or develop them in the world. ALFRED reactor design investigated in this paper has 125 (MW) electric power and is being cooled by lead. The main purpose of the ALFRED project is to design and construct a reactor with high safety performance and reliability in all operation modes. In the present study, Monte Carlo codes MCNPX and SuperMC are used to calculate the neutronic parameters of the ALFRED reactor core including effective multiplication factor, kinetic parameters, neutron flux, and power distribution based on the latest design data. According to the results, the insertion of the control rods into the core can move the maximum flux height of about 40% to 50% to the end. Also, comparing codes’ results show an appropriate consistency.

The paper presents the results of experimental studies investigating the peculiarities of coolant movement at the border and in the bordering fuel rods of jacketless fuel assemblies with various fuel rod diameters and type of spacing grids (honeycomb or plate). The study focuses at the effect of the grid type and spacing of fuel rods in neighboring fuel assemblies on the hydrodynamics and mass transfer in the coolant flow. A series of experiments was performed using the created scale model of the boundary area between jacketless fuel assemblies with fuel rods of various diameters. The studies were carried out using the pneumometric method for measuring the static and dynamic flow characteristics, as well as thermocouple measurements.

In this paper, the optimization of the square and squared-off cascades for the separation of stable isotopes with a certain number of gas centrifuges (GCs) using particle swarm optimization (PSO) and the whale optimization algorithm (WOA) was studied. For this purpose, the “SC-PSO” and “SC-WOA” codes for the square cascade (SC), and the “SOC-PSO” and “SOC-WOA” codes for the squared-off cascade (SOC) were developed. The performance of the PSO and WOA algorithms in the optimization of the square and squared-off cascades as an example for the separation of ¹²⁴Xe from nine stable isotopes of xenon up to an enrichment of 85% was compared. The recovery coefficient of ¹²⁴Xe isotope by the codes “SC-PSO”, “SC-WOA”, “SOC-PSO”, and “SOC-WOA” was obtained 86.14, 78.11, 92.84, and 83.41% respectively. The performance of the algorithms indicates the higher yield of the PSO algorithm in the optimization of SC and SOC. The efficiency of the PSO has been demonstrated by the Griewank test function compared with the WOA, the sine cosine algorithm (SCA), and the dragon-fly algorithm (DA) (PSO>WOA>SCA>DA). Furthermore, the results show that SOC performs better than SC.

The paper considers the KIR-TG software package, which is designed for the interconnected neutron-physical and thermohydraulic test calculations of the infinite grid of VVER-SKD fuel elements. As part of the work, a software program written in the Python and Lua dynamic programming languages for linking between the module for calculating neutron physics (KIR computer program) and the TK-SKD thermohydraulic module was created and applied. The use of precision calculation performed by the Monte Carlo method in the neutron-physical part avoids the error of deterministic approximation methods in the calculation results. The problem statement, as well as the nature and results of the work, revealed the advantages of this approach to solving problems related to thermal-hydraulic calculations: parallel neutronic and sequential thermal-hydraulic calculations do not affect the overall speed of the associated calculation, the calculation codes are developed in parallel to save time and resources, and errors are tracked separately.

During the production or storage of a hydrogen gas in a confined space, the formation of an explosive mixture can occur in the event of an accidental leakage. Therefore, in order to ensure safety, it is necessary to create a room ventilation system that prevents the formation hazardous zones with an elevated content of combustible gases that can lead to a local explosion. The paper presents the results of calculations and experiments on hydrogen leakage with a constant flow rate into a ventilated room with a volume of 8 m³. In the experiments, the emergence of depressurized electrolyzer were simulated with various leakage rates. The dependence of the hydrogen volumetric content on the time in various points of the room was obtained. In order to test the prognostic capabilities of the calculation methodology, preliminary calculations were performed. Several approaches to the application of boundary conditions were investigated. The proposed calculation methodology is in the agreement with the experimental data.

The article presents the results of a radiation survey carried out on the liquid radioactive waste disposal system of decommissioned MR and RFT research reactors. The system is located in the buildings adjacent to the dismantled reactor building on the territory of the Kurchatov Institute national research center. The results of the visual and radiation survey carried out on the equipment and premises of the disposal system for liquid radioactive waste will inform the development of design documentation for its elimination. Following the removal of liquid radioactive waste from storage tanks, the equipment and building structures at this location should be dismantled. The carrying out of recommended works will complete the decommissioning of MR and RFT research reactors. Further, the site of the reactor buildings and their auxiliary systems will be rehabilitated to achieve a radiation situation that meets the sanitary and hygienic standards for the group B personnel.

Abstract

A new promising type of portable pulsed neutron generator is based on a laser-plasma ion diode with magnetic insulation. At the present stage of its evolution, such a generator requires the development of a magnetic insulation system capable of blocking secondary electrons in a laser-plasma diode. The article considers the development of such a magnetic insulation system, during which the effect of the magnetic field distribution on the shape of the ion current pulse through the diode was estimated using the KARAT electromagnetic code to significantly enhance the theoretical validity of the magnetic system design. The final design of the magnetic system should provide the complete insulation of secondary electrons, as well as significantly increase the fraction of ions deposited on the neutron target, as well as the electrical strength of the accelerating gap. Following was fabrication of the developed magnetic system, measurements of the magnetic field distribution within the accelerating gap showed the sufficient accuracy of its implementation. The average difference with the model is ~3%.

Abstract

During the storage or production of a hydrogen gas within a limited space, a combustible mixture may form in the event of an emergency leakage. In order to ensure safety, it is necessary to develop a room ventilation system for preventing the formation of areas in which there is an increased content of combustible substances. The present article considers experiments on the leakage of a hydrogen gas in a ventilated room with a volume of 8 m³. Emergencies connected with the depressurization of electrolyzer or a high-pressure hydrogen vessel during the storage were simulated. The time-dependencies of the hydrogen content at various points of the room were obtained. Data on the combustion of the hydrogen-air mixture in the room with relief valves were obtained from experiments carried out with the ignition of the mixture.

Accidents involving the release of large amounts of combustible gases, such as hydrogen, methane, carbon monoxide, etc., are possible at nuclear power plants, as well as in chemical production facilities. Currently, calculation tools are used for substantiate explosion safety. To permit their validation, experimental data obtained from a prototype facility are required. RFNC–VNIITF has developed a two-chamber BM‑U facility disposing a total volume of 156 m³ for simulating emergency processes of leakages and combustion of hydrogen-containing vapor-gas mixtures. Inside this facility, equipment prototypes characteristic of the simulated object can be installed to account for their influence on the overall level of explosion safety. The BM‑U facility is additionally equipped with the necessary state-of-the-art measuring instruments. When the facility was put into operation, an experiment with the ignition of an air mixture with 6 vol % of hydrogen was conducted.