Atomic Energy最新文献

The article presents the results of an experimental study and numerical simulation of argon ion flow formation in the inter-electrode gap of the ion-optical system used in a miniature linear accelerator. In the experimental study, an ion source with a Penning discharge was used to generate charged particles. The experimental results of measuring the beam emittance at the output of an ion source were used as initial data to numerically simulate the dynamics of charged-particle flows. The simulation results are consistent with those obtained via an experimental study on the trajectories and parameters of the ion flow local spot on the collector of the accelerating system.

The article presents the results of the study on the amplitude-time characteristics of pulsed discharges and ion flux formation processes in spark-arc ion sources used in small vacuum neutron tubes. The obtained data show that the initial phase of the discharge ignition is accompanied by the development of instabilities that affect the formation of ion fluxes. An interruption of the vacuum arc at the moment of a change in the discharge current polarity results in a sharp change in the energy characteristics of the discharge and the neutron yield of the tube. Based on the results of the study, options are proposed for constructing pulsed power supply systems based on ion sources that increase the operational stability of neutron generators.

The work describes features of the procedure for obtaining a response matrix to neutron radiation with energies from 1 to 19 MeV for a single-crystal stilbene scintillation detector. This response matrix is subsequently used in solving the problem of unfolding the neutron energy distribution at the detector location. The procedure for obtaining the matrix includes both experimental work and simulations of nuclear physics processes. In order to carry out the simulations, the Geant4 library package with the connected optical physics accounting module was used. The resulting matrix can be used as part of the amplitude method for unfolding the spectra of neutrons with energies in the range of 1–19 MeV with a step of 0.1 MeV. The obtained response matrix takes into account the main physical processes and instrumental effects of detectors based on organic scintillators, including the dependence of the light yield in the scintillator on the type and energy of charged particles, as well as the energy resolution of the detector depending on the energy of detected particles. In comparison with earlier results obtained using response matrices that do not take into account the above formation features of hardware spectra, it was possible to reduce the lower limit of the detected neutron energy from 1.5 to 1.0 MeV, as well as to increase the reliability of unfolded neutron energy distributions at the measurement point. The error in determining the neutron energy in the range of 1–15 MeV was not more than 200 keV.

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.