Atomic Energy最新文献

The paper presents an assessment of the effect caused by technological uncertainties on k_eff using the example of the test problem for a MET1000 fast sodium reactor with metal fuel. It is proposed to perform this assessment using nuclear sensitivity factors by random sampling or direct perturbation without requiring multiple calculations. The TSUNAMI-3D module of the SCALE program was used to analyze the sensitivity to nuclear data uncertainty. The obtained constant and technological uncertainties based on ENDF/B VII.1 equal 1.15 and 0.6%, respectively. The result was verified using direct perturbation. It is shown that a significant proportion of the technological uncertainties can be determined within the framework of the proposed evaluation method.

The annual consumption of tritium in a fusion reactor is estimated at ~56 kg per 1 GW. Its amount depends on the power of the reactor, plasma parameters, and the technologies used. Tritium can be produced in nuclear reactors to form an initial margin. The present article analyzes the demand for tritium during the future development of fusion energy. It is shown that tritium demand under various scenarios exceeds current production capabilities and margins. An overview of tritium production in PWRs (USA) is presented. Lithium absorbing rods and stages of their technological implementation are described. The production of tritium in a reactor limited to 1.5 kg/year is demonstrated to cause no effect on operation or safety. The introduction of such technologies, which have been worked out over a period of around 15 years, is possible in both VVER and RBMK reactors.

The study aims to determine the thermophysical properties of uranium-gadolinium oxide fuel pellets with 6, 7, 9, and 10 wt% of gadolinium oxide for the verification of the Start-4A calculation code. The determined properties will help to clarify calculations on the behavior of fuel elements in normal and emergency operational conditions during operability substantiation and licensing of the fuel for VVER reactor plants. Experimental dependencies of thermal expansion, specific thermal capacity at a constant pressure and thermal diffusivity were obtained along with the calculated dependence of thermal conductivity on the temperature for the fuels with various contents of Gd₂O₃. In comparison with uranium dioxide, the main contribution to an almost twofold decrease in thermal conductivity at an increase in the gadolinium content is made by the thermal diffusivity of uranium-gadolinium fuel.

A method for calculating a square cascade with an arbitrary number of external streams by varying the sections of partial-stage streams and minimizing the deviation of the calculated-stage feeding streams from the specified stream was developed. The method applies the bee colony algorithm, which minimizes the deviation of streams to ensure the required accuracy of determining cascade parameters. On the basis of the computational experiment on the separation of molybdenum hexafluoride, the method demonstrated high efficiency when calculating the concentration of isotopes with the lowest and highest mass numbers.

Calculations of severe accidents performed for level 2 probabilistic risk analysis (PRA-2) require the relevant leakage parameters such as location, area, flow rate, etc. Since the pressure of the medium at which the leakage occurs is probabilistic, a series of variant calculations of a severe accident is performed for the PRA‑2. However, when postulating the occurrence of a leakage, determining its size and location remains problematic. In this regard, an approach is proposed for reducing the number of variant calculations of severe accidents for the PRA‑2 to obtain its realistic results. According to conservative assumptions, a single through crack in the most stressed area of the lining is postulated, whose behavior will determine the ultimate state of the containment. The initial crack size corresponds to the leakage of the medium during the commissioning tests of the containment. Leakage parameters can thus be determined based on the assessment of the stress-strain state of the containment lining, while the ultimate state is established on the basis of the linear fracture mechanics.

The article presents the results of developing a fragment of a thermal displacement control system, writing the corresponding software for measuring the linear displacement of the monitored equipment and pipelines, and analyzing the obtained results for assessing the technical state and cyclic strength of the monitored equipment as a means of managing the resource characteristics of the controlled object. The measurement uncertainty was evaluated. Taking into account the influence of various factors, such as illumination, the degree of filling the image with the target, the distance from the video camera to the target, as well as the direction and dimensions of displacements and vibration, the volume of test checks was formed using 780 tests. In the course of the analysis, the relative error in measuring displacements taking vibrations into account was revealed to be no more than 5%.

The article considers the possibility of using high-temperature gas-cooled reactor (HTGR) technology for the needs of industrial heating, primarily for enterprises that require high-potential heat in technological processes and operate in a continuous cycle mode. The integral demand for the thermal power of HTGRs for the development of the metallurgical, petrochemical, chemical, and processing industries until 2100 is estimated. The potential of using thermal power for the needs of industrial hydrogen production by the method of the methane steam conversion is estimated. In addition to saving the natural resources of the hydrocarbon fuel currently used for heating, the introduction of HTGRs will ameliorate the environmental situation associated with emissions of pollutants and greenhouse gases, as well as reducing the evaporation of purified water.

The uniqueness of the nuclear fuel cycle facilities and their equipment, which must be adjusted to the specific conditions of manufacturing and (or) reprocessing the fuel of a particular reactor type in each case, complicates the application of traditional methods for assessing reliability and safety, thus relying on the availability of a significant operating experience. As a result, at the design stage, the reliability level of such equipment is determined by the available data on the reliability of analog elements. Following completion of commissioning of a nuclear fuel cycle facility, the regulatory requirements demand the updating of reliability and safety calculations to bring them in line with the actual state of the facility based on the results of the construction, commissioning, as well as pilot- and pilot-industrial operation. The article presents a universal method for accounting for the statistics of equipment operation at a high level of reliability in the absence of representative samples during the first years of its operation, which is compensated by the use of prior information within the framework of the Bayesian approach. The simulation of the time between failures is demonstrated using the synthesis of the prior information and the results of prototype bench tests for 5 years.

The present paper considers the processes occurring in a failed fuel element during fuel cladding failure detection in the mast sipping system of the refueling machine. The time taken for water and fission gas products to leak from a failed fuel element during its lifting was estimated depending on the size of the defect. A period of 10–12 min exposure prior to bubbling was demonstrated to be applicable for detecting defects with a size of about 30 μm. The sizes of gas bubbles leaving the fuel element during the failure detection were estimated along with their rate of ascent. An algorithm for conducting fuel element failure detection that increases the sensitivity of the method without extending the refueling is proposed.

The paper analyzes the assumptions relied upon when deriving the general form of the separation potential for multicomponent mixtures. The complexity of applying the general form for evaluating the number of separating elements in a centrifuge cascade without performing calculations is demonstrated. Its potential application for other methods characterized by high selectivity for individual components of the separated mixture is discussed.