Atomic Energy最新文献

The DUM‑R salinometer is developed for in-circuit measurement of water salt content and monitoring of neutral, uncorrected water chemistry conditions in the secondary circuit of reactor plants for nuclear vessels and floating power units. The salinometer incorporates the latest advances in instrumentation and long-term operation experience of secondary circuits in marine reactor plants. The presented salinometer has an increased accuracy and reliability, compared to its analogues. The DUM‑R salinometer used in the secondary circuit systems of reactor plants for the nuclear icebreaker fleet and floating power units can increase the reliability and trouble-free operation of reactor plants, as it provides independent monitoring of water chemistry conditions, prompt receipt of information on changes in the water salt content and disruptions in maintaining the water chemistry conditions (salinization of the secondary circuit water), as well as timely and comprehensive implementation of compensatory measures aimed at eliminating failures and malfunctions of equipment and pipelines. DUM‑R salinometers appear to be promising for use in the secondary circuit, as well as the auxiliary circuit of a small modular reactor plant with the RITM-200N reactor.

The present paper considers methods for profiling the feeding stream of stages in a cascade of gas centrifuges with variable stage separation coefficients and a given concentration of the target component in the outlet streams. An advanced method is proposed to increase the efficiency of a gas centrifuge cascade with a given layout. A computational experiment is conducted to enrich reprocessed uranium hexafluoride and separate germanium tetrafluoride. The methods are compared; the advantages of the proposed method are demonstrated.

The article examines the composition and behavior of corrosion products, γ‑emitting radionuclides of corrosion origin, and α‑emitting radionuclides in the primary coolant of a light-water propulsion reactor in ammonia water chemistry conditions. Activated corrosion products in the coolant are represented by zirconium, chromium, manganese, cobalt, and iron. The total specific α‑activity of the coolant is determined by plutonium, americium, and curium. No increase in the activity of radionuclides in the coolant of a reactor operating at power was observed during the study. The stability of forms and content of impurities in the coolant means that ammonia water chemistry in the primary coolant of small modular reactors can be maintained without dosing under long-time operation conditions.

The paper presents the results of adapting the ROM code constant base and models for calculating the population exposure dose, taking into account the IAEA recommended radiation exposure pathways with a transition to the determination of the effective dose by ICRP Publication No. 103 and introduction of constants for calculating the absorbed and RBE-weighted dose. The considered example of a severe accident demonstrates that the replacement of the upper estimate based on equivalent and effective doses with special constants for assessing the acute exposure dose reduces the conservatism of skin, thyroid gland, red bone marrow, and lung exposure forecast by 60%, up to 5, 3, and 2 times, respectively. The adjusted determination of the effective dose shows a 10% change in the total radiation exposure dose: inhalation and external exposure dose are 20 and 1%, respectively, at a comparable contribution of both radiation exposure pathways to the total dose. Additionally considered exposure pathways recommended by EPR-NPP-OILs (IAEA) can contribute up to 25% to the dose depending on the specific dose value.

The paper describes a method for calculating a cascade with the loss of working substance and equal number of gas centrifuges in stages at variable stage separation coefficients and a given concentration of the target component in the outlet streams. The proposed method minimizes the loss of working substance in a cascade by determining the optimal number of stages and feeding stream of the cascade at the same feeding streams of stages. A computational experiment on the enrichment of cadmium dimethyl with ¹¹⁶Cd was carried out to demonstrate the features of applying the described method.

The present study aims to determine the thermophysical properties of industrial uranium-erbium oxide pellets with an erbium content of 1.2 wt % to substantiate the performance and safety of fuel for VVER reactor plants. The performed study yields experimental dependences of thermal expansion, specific heat capacity at a constant pressure, thermal diffusivity, as well as the calculated dependence of thermal conductivity on the temperature up to 1900 K.

Further improvement of industrial centrifuges requires a clear understanding on the interdependence of their optimal geometric, fluid dynamics, and separation parameters. The present article considers the theoretical optimization of contemporary high-speed countercurrent gas centrifuges for the separation of heavy isotopes. Within the framework of the refined radial averaging method, optimization problems for gas centrifuges with constant and axially profiled countercurrent are solved. The most optimal coordinate of the feeding point was determined offset from the known position to the waste.

The paper presents the results of neutron-physical and thermal gas dynamics studies of the core of a compact high-temperature gas-cooled reactor with fuel particles to justify the development of a small modular reactor concept. The performed calculations assume the conditions of stability and self-similarity for heat removal in nominal and variable modes. A heuristic kinematic image of the coolant flow in permeable channels is used to derive a one-dimensional equation of movement with strictly distinguished inertial and dissipative effects. The derived equation is important for determining the profiles of channels and fuel layer ends, which ensure the correspondence of coolant distribution and power density. The results of basic engineering for the core for a 5 MW reactor cooled with a helium-xenon mixture are presented.

To ensure timely detection of depressurized fuel rods, the BREST-OD-300 has a system for failure detection of their cladding. The present article describes the subsystems of fuel cladding failure detection operating according to different physical principles. The subsystems inform personnel about the state of the core by accumulating information on the concentration of reference radionuclides in process environments. The considered subsystems perform online detection fuel cladding failure based on delayed neutrons, as well as gas and coolant activity; moreover, a monitoring system for a shutdown reactor during core refueling, and equipment for periodic sampling and analysis of lead coolant. The experiments conducted by the NIKIET JSC to substantiate the possibility of using the described subsystems as part of the BREST-OD-300 fuel cladding failure detection are considered.

The primary task of development of the Far Eastern Federal District, Russian Federation involves the solution to the problems of shortage and reliability of electric power supply through large-scale modernization of the energy sector. The inclusion of reliable and safe floating nuclear power units in the district power systems based on the developed and ready-to-implement RITM-200 power plants will increase the reliability and sustainability of energy systems and security of energy supply for the region. This decision will facilitate the expansion of nuclear energy production at NPPs and will strengthen the position of nuclear industry in the energy market.