Atomic Energy最新文献

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.

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%.

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.

Fuel elements with mixed uranium-plutonium nitride fuel were tested in an IGR reactor to justify their application in a BREST-OD-300 reactor. The mid-radial enthalpy limit for the fresh mixed nitride fuel as experimentally determined during IGR launching with fast reactivity input amounted to 167 cal/g. The maximum temperature of 1000 °C and its maintenance for 100 s, comprising one of the design limits for the fuel cladding temperature, was experimentally confirmed. The main results of the performed experiments and post-irradiation studies are analyzed.

The main approaches to americium transmutation in BN-1200M are considered: homogeneous transmutation in the core and heterogeneous transmutation with a moderator in the radial blanket. It is stated that the main limiting factor of transmutation is the residual heat release of the unloaded FAs. Within the design constraints, the annual transmutation rates are approximately the same in both cases. It is proposed to combine the advantages of both transmutational methods so as to ensure the homogeneous transmutation of americium in the core at the level of an equilibrium content and the heterogeneous transmutation of external americium in the radial blanket. On the basis of a simplified economic model, a comparison of various approaches was made and the corresponding technical and economic estimates were given.

When designing steam-generating devices, it is necessary to consider the issue of determining the limits of their stable operation. Such stable operation limits can be determined by testing a model that reflects the main design features of a full-scale steam generator. For this purpose, JSC OKBM Afrikantov has developed and tested a steam generator model. The results obtained in the tests for determining the limits of hydrodynamic stability can be used to determine the permitted range of operating loads for steam generators.

The article studies the possibility of using a matrix of gibbsite and kuzelite for the preservation of liquid radioactive wastes consisting of aqueous NaNO₃ concentrates. The matrix is formed during the solidification of these wastes with cement, consisting of calcium aluminate and gypsum. The requirements for cemented solid waste are met by a matrix obtained from a mixture of the following composition: 45.2%-NaNO₃ solution 53.04%; gypsum—5.4%; calcium aluminate—36.16%; diatomite powder treated with polydiallyldimethylammonium chloride (polyDADMAC)—5.4%. The matrix has the following characteristics: ratio of matrix and cemented solution volumes—1.45; average ¹³⁷Cs leaching rate for 90 days—1.4∙10⁻⁴ g/(cm²∙day); strength—15.2 MPa, including after the tests for water resistance, 30 freezing-thawing cycles, and irradiation to a dose of 1 MGy—6.8, 15.4, and 9.6 MPa, respectively.

The paper considers the optimization of a liquid system for compensating excess reactivity by varying the number of gadolinium-containing fuel elements in a VVER fuel assembly and gadolinium mass loading into the fuel elements. Using a simplified refueling model, an ideal dependence of the neutron multiplication factor with burnable absorbers is obtained with reactor criticality, which is maintained using only burnable absorbers. The calculations were carried out using the Serpent neutron physics program code, version 2.1.32 (Finland). A significant reduction in uncompensated reactivity and low-level liquid radioactive waste generated using a liquid compensation system is estimated for real options similar to the ideal one.