Atomic Energy最新文献

An analysis of velocity and temperature distribution along the normal to the heat-transfer surfaces of in-reactor units was carried out under the conditions of natural convective heat and mass transfer. A comparison with the theoretical results, obtained by solving boundary-layer equations with the corresponding conditions, was made. The longitudinal velocity component and the temperature are presented in dimensionless form depending on the dimensionless coordinate η: 123 and (uptheta =(T-T_{mathrm{infty }})/(T_{w}-T)=f(upeta )). The reason for the insufficient coordination of the theoretical results with experimental data is explained along with the selection of a nondimensionalization scale for velocity and temperature. Simple generalizing dependencies on the coolant velocity and temperature distributions for natural convection conditions are presented based on the proposed approach for the selection of characteristic scales.

The Heavy Water Zero Power Reactor (HWZPR) is a 100‑W research reactor that uses natural uranium metallic fuel and heavy water as moderator. The HWZPR is designed for training and research in reactor physics fields. The main aim of this paper is to evaluate the variation of neutronic parameters of a HWZPR resulting from changes in the fuel lattice pitch using MCNPX code. In such a manner, neutronic parameters including effective multiplication factor (k_eff) for critical water level, effective delayed neutrons fraction (β_eff), prompt neutrons lifetime (l_p), and neutron flux distribution are investigated and benchmarked. The results illustrated that, as the lattice pitch is increased, more heavy water is needed for criticality condition according to the decrease in the number of fuel rods and fewer available fissile materials. Comparison of the results with the reactor Safety Analysis Report shows reasonable agreement.

In order to improve the safety of operating nuclear power plants with RBMK-1000 reactors, work is underway for developing and substantiating the control actions of personnel under severe accident conditions for all possible initial states of the power unit, including shutdown for decommissioning. An accident is considered using the STEPAN‑T 3D program developed specifically for simulating an accident with the complete blackout of a RBMK reactor. The time dependence of the temperature in the core, metal structures, and OR scheme is given. Quantitative estimates of the hydrogen formation during an accident are shown. The possibility of re-criticality and release of radioactive substances occurring during the accident are discussed. Possible measures to mitigate the consequences of the accident are given.

The separation of americium during the fractionation of a highly active raffinate obtained in the extraction processing of spent nuclear fuel represents an urgent task of the contemporary nuclear fuel cycle. The article discusses new approaches to this task. It is shown that a sodium bismuthate powder (NaBiO₃), upon contact with a solution of Am (III) and Cm (III), oxidizes Am (III) to Am (VI) and sorbs actinides. The addition of a (NH4)₂CO₃ solution results in a content of up to 91% of americium and about 2% of curium in the solution after desorption. The behavior of americium and curium in acidic and alkaline solutions of potassium hexacyanoferrate (III) was studied. In acidic solutions of HNO₃, americium and curium are precipitated, while praseodymium, comprising a lanthanide simulator, remains quantified in the supernatant. In alkaline solutions of potassium hexacyanoferrate (III), ~50% of Am (III) is shown to oxidize to Am (V). The obtained results can be used as a basis for a new technology of separating americium from curium and lanthanides for the purposes of americium transmutation.

The IGRIK‑2 solution pulse reactor is a new generation irradiation facility with a distinguishing feature of a large experimental channel of up to 39 cm in diameter. The characteristics of gamma-neutron radiation were justified at the stage of its development. The possibility of implementing the design characteristics is determined by the critical configuration of the core: its geometry and the composition of the fuel solution comprising a light water solution of uranyl sulfate. One of the tasks in the implementation of the critical configuration, including the assembly of the IGRIK‑2 reactor core, is the preparation of a fuel solution. The complexity of the task is firstly due to the use of fuel solutions from decommissioned IGRIK and ELIR reactors as the initial components, and secondly due to the requirements of ensuring a given geometry of the core while minimizing the concentration of cadmium contained in the fuel solution of the ELIR reactor. The paper describes an experimental calculation method for assembling the core of the IGRIK‑2 reactor. The preparation of the fuel solution and confirmation of the critical core configurations were carried out in four stages with a cycle of computational and experimental studies carried out at each stage. The results of each stage are presented. The achievement of the desired critical configuration is demonstrated.

The article confirms the possibility of using a two-component iron-phosphate glass composite material synthesized at 700 °C to immobilize a simulator of a highly-active spent chloride electrolyte, such as those obtained in the pyrochemical processing of the spent mixed uranium-plutonium fuel of a BREST-OD-300 reactor. The structure and phase composition of the material were studied using scanning electron microscopy with energy-dispersion X‑ray spectroscopy, X‑ray fluorescence analysis, and X-ray powder diffractometry. The waste components are shown to form stable pyrophosphate phases. The material leachability of waste components is defined according to the PCT standard. The glass composite material is highly water resistant. Thus, the prospects for the practical application of the studied material for the reliable immobilization of spent electrolyte materials are demonstrated.

Neutron production in high-energy-density plasma is an interesting topic for developing thermonuclear sources of neutrons and hybrid reactors. The promising candidate reactions for thermonuclear fusion (thermonuclear sources of neutrons) are T(d, n)⁴He and D(d, n)³He (D‑T and D‑D reactions). As T is a radioactive material, fusion experiments in all over the world deal with the D‑D reaction by using nuclear fuel in the form of plastic material such as CD or CD₂. Consequently, reactions of deuterium and carbon effectively participate in the neutron yield. In this work, the neutrons flux of the nuclear reactions D(d, n)³He and D(¹²c, n)¹³N were simulated using the MCUNED Code in high-energy-density plasma produced by an ultra-high-power laser. Neutrons have been produced using nuclear fuels D₂, CD, and CD₂. Neutron flux has been calculated per kW of the laser energy expended in the fuel. Deuteron and proton fluxes have been calculated under the same energy conditions. The obtained data for the reactions D(d, n)³He and D(¹²c, n)¹³N were compared with each other and with the available experimental data. A computer program, NJOY, has been used to calculate cross section in Acer format to be used by MCUNED. The neutrons flux peak was increased inside the source and magnet cells for CD and CD₂ due to adding carbon to the deuterium fuel. These values were decreased in the cells outside the source. The total neutron yields were approximately the same for D₂, CD, and CD₂ fuels in the air cells outside the source.

Hydrogen-air mixtures are highly flammable. The acceleration of the flame and subsequent deflagration to detonation transition (DDT) can cause enormous damage to hydrogen energy infrastructure. Since leakage of hydrogen and its subsequent stratification according to the height of a room or structure is the most likely process to cause emergencies that arise at hydrogen energy facilities, studies of combustion and detonation in stratified hydrogen-air mixtures are of particular interest. The paper presents the results of the estimated flame front velocity and maximum overpressure in experiments involving the deflagration of a hydrogen-air mixture for vertical gradients of the hydrogen volume fraction in a closed channel with an annular blockage. This horizontally oriented channel has a square section of 0.6 × 0.6 m and a length of 12 m. The average hydrogen content of the gas used in the experiments varied in the range of 9–15 vol %.

A mobile laser system is proposed for monitoring the atmosphere around the location of a NPP determining the concentration of molecular iodine at the level of 10⁸ cm⁻³ and assessing the volumetric activity at the level of 1 MBq/m³. Due to the high mobility of the system, which is ensured by the use of an unmanned aerial vehicle, it can be used in the event of an emergency at a nuclear power plant or similar enterprise to assess the radioactive contamination of various parts of the adjacent territory including the upper hemisphere to a radius of 5–10 km, which limitation is due to the flight range of an unmanned aerial vehicle.

A schematic thermal diagram of a turbine plant with a thermal gas boiler developed for nuclear power plants with BREST-300 reactors, which considers means for ensuring increasing power generation efficiency at power facilities, is presented. The heat produced in a gas boiler during the combustion of organic fuel is used for an initial and intermediate superheating of the working fluid upstream the turbine cylinders, as well as for preheating the air supplied to the boiler. The BREST-300 NPP discussed in the work is designated as an organic-nuclear power plant (ONPP). In the presented version of the thermal diagram for a turbine plant, the internal efficiency of the cycle is 54.12%.