Atomic Energy最新文献

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.

Aspects of separation potentials for a multicomponent mixture of isotopes are considered. In order to separate isotopes of reprocessed uranium and germanium in cascades with high separation factors corresponding to gas centrifuges, a computational experiment was carried out. The application of potentials for calculating the separation power of elements, stages, and cascades is analyzed. Potentials that provide a satisfactory assessment of the actual separation power for the stages and the cascade, thus determining the costs of the separation work, are established. This characteristic, along with the separation power of the element, is virtually independent of the used potential type.

The paper presents the results of a calculation study connected with the composition of RBMK-1000 fuel having a reduced erbium content. The computational studies were carried out using SADCO software systems (developed by JSC “NIKIET”) and MCU-RBMK (developed by NRC “Kurchatov Institute”). The studies include calculations of a fuel cell with various compositions of uranium-erbium fuel and a RBMK-1000 reactor under fuel-loading conditions involving a reduced erbium content. The conversion of RBMK-1000 to a fuel having a reduced erbium content ensures the neutron-physical characteristics and volumes of accumulating commercial ⁶⁰Co to be maintained within operational limits under conditions of extending the reactor service life for more than 50 years.

The paper presents the results of critical, correlation, and spectral precision experiments carried out at the FKBN‑2 assembly machine using a cylindrical multiplying system made of low-background α‑phase plutonium. In the experiments, the critical state of the multiplying system was determined, the characteristics of non-stationary prompt neutron processes were investigated, and the number of nuclear reactions in the detectors was measured at a controlled power level of the multiplying system. After constructing a benchmark model for the multiplying system, a computational simulation of integral experiments was carried out. Based on the experimental results, the average lifetime of prompt neutrons and the effective fraction of delayed neutrons for the multiplying system were determined. The obtained results can be used to verify the codes of neutron-physics and refine neutron constants.

The paper addresses the development of prototype large-scale purified inert atmosphere cells for testing equipment used in the pyrochemical treatment of spent nuclear fuel from fast neutron reactors. In addition, design peculiarities of the cells are discussed along with a presentation of preliminary results of tests on the generation and maintenance of an inert atmosphere of required quality in them.