Atomic Energy最新文献

This paper assesses the time to radiation (radiotoxicity) and radiological (radiation risk) equivalence between natural uranium and radioactive waste of thermal and fast reactors at an increasing content of Np, Am, and Cm in radioactive waste. We calculate radiation risk by adapting models of the ICRP and other international organizations for the Russian Federation, taking into account background epidemiological indicators including cancer incidence, cancer mortality, and overall mortality. The radiation risk of internal exposure considers both the time after radionuclide intake and equivalent dose dynamics in human organs and tissues. To determine the time to radiological equivalence, we estimate the lifetime attributable risk of a single intake of natural uranium radionuclides and radioactive waste. An increase in the Np, Am, and Cm content of radioactive waste within 0.1–0.4% causes no effect on the radiological equivalence time, equal to ~100 years. However, a further increase from 0.5–0.8% prolongs the time to 300 years. A problem for optimizing the time to radiological equivalence is posed providing for waste storage costs and content of Np, Am, and Cm in long-lived radioactive waste.

Radiation safety of high temperature gas-cooled reactors (HTGRs) largely depends on the leakage of fission products from the fuel under both normal and emergency modes of operation. Within the development of a pilot industrial fuel technology for an HTGR of the nuclear power engineering plant, we conducted pre-reactor tests of laboratory manufactured fuel to determine the leakage of fission products, as well as to assess the contamination of structural materials with uranium. Bulk TRISO particles, TRISO particle compacts, and blank graphite compacts were studied. The conducted tests used a weak irradiation technique developed by the NRC “Kurchatov Institute”. The technique is based on measuring the leakage of gaseous fission products from weakly irradiated samples during their annealing at temperatures close to the core temperature. A small dose of radiation causes no effect on the basic initial physical and chemical properties of fuel (density, porosity, etc.), as well as the ability to retain fission products. We analyzed the quality of laboratory-manufactured HTGR fuel by comparing the measured relative leakage of reference radionuclides with the established permissible limits, including individual technological stages of production. The study highlights the importance of the technique for quality control of HTGR fuel.

The paper considers temperature and humidity simulation of the primary circuit pipeline rooms located at power units No. 3 and 4 of the Kola NPP upon the occurrence of equipment leakage. Simulation results were used to describe the dependence of the leakage on the temperature and relative humidity recorded by the sensors. Thus, advanced methods of automated coolant leak detection were developed; existing methods were adjusted.

The paper provides an analytical expression to determine the optimal rotor length of a gas centrifuge. The optimum is defined as the maximum separation power per unit volume occupied by the centrifuge. Long centrifuge rotors, typical for European centrifuges, appear economically inappropriate.

The paper presents the results of simulating circulation of a lead-bismuth coolant in a hydraulic circuit using the CONV‑3D code. The temperature head and flow rate of the coolant are simulated for various heating power values. The developed code satisfactorily simulates the natural convective flow of a lead-bismuth coolant in a hydraulic circuit, thus being appropriate for analyzing these flow types in various applications.

The article presents the results of experiments conducted by the IPPE JSC on heat exchange and circulation stability during boiling of alkaline liquid metals in the models of single and parallel fuel assemblies (FAs) under natural circulation conditions. It is shown that the complex structure of liquid metal boiling in FAs is formed under the influence of various factors and is characterized by both stable and oscillatory modes, as well as possible transition to departure from nucleate boiling. The influence of the fuel rod surface roughness on the heat exchange and flow modes of a two-phase liquid metal flow in bundles is shown. The presented results of experiments on the heat exchange during sodium boiling in an FA model with a sodium cavity above the reactor core demonstrate the possibility of long-term cooling of fuel rod simulators in FAs. The dependence on the heat transfer during boiling and a map of the modes for a two-phase liquid metal flow in the FA are presented. The agreement between the results of computational simulation and experiments on alkaline metal boiling in a single model FA and a system of parallel FAs is shown.

In order to structurally diversify the energy sector and transit to carbon-free development, R&D of nuclear hydrogen technologies for the large-scale production and consumption of hydrogen is being carried out including a project of a nuclear power engineering plant with a high-temperature gas-cooled reactor (HTGR NPEP) for obtaining hydrogen and hydrogen-containing products. Based on domestic and foreign experience, steam reforming of methane was selected as the method of the NPEP hydrogen production. A description of the NPEP and its technical characteristics is provided. The main technological processes implemented in the chemical engineering plant are described. A proposal for licensing a NPEP project providing for a single-site location of a nuclear power facility and a hazardous industrial facility with different regulatory requirements is considered. A list of R&D activities, which is necessary to confirm technical solutions and develop materials for licensing the NPEP project in regulatory authorities, is provided.

In the course of the study, various designs of experimental fuel rods with mixed uranium-plutonium nitride fuel and claddings made of various materials were tested as part of the dismountable irradiation devices of the BOR-60 reactor. The experimental fuel rods represent the mock-ups of the BREST-OD-300 and BN-1200 reactor fuel rods, which differ from the real ones in their shorter length, smaller gas collector volume, and fuel column height. Reactor tests and post-reactor studies of 2020–2023 provided new data on the state of fuel, various types of fuel rod claddings, and fuel assembly design elements.

The paper provides the results of determining forms and compounds of radioactive iodine isotopes in emissions of a controlled source during normal operation of the VVR-ts research reactor and production of radiopharmaceuticals. The determined average daily volumetric activity of four iodine isotopes ¹³¹I, ¹³²I, ¹³³I, and ¹³⁵I equals to 119, 126, 36, and 21 Bq/m³, respectively. A 97-% fraction of the first three isotopes has a gaseous form in the gas-air environment of the emission source. Weakly adsorbed compounds predominate among these gases. The isotope of ¹³⁵I was identified in a single experiment, which hampers drawing unambiguous conclusions about its forms and compounds in the emissions.

Being put into operation in December 1969, the BOR-60 reactor has been operating successfully for more than 54 years at design parameters without replacements of the main process equipment and sodium of the primary and secondary cooling circuits. Its operating experience was used in the development of BN-350 and -600 reactors. The experience of the BOR-60 operation and experimental studies was used to develop a concept of a MBIR multi-purpose research fast reactor currently being built at the site of the JSC “SSC RIAR”. The article presents the main results of the BOR-60 operation and experimental studies.