Atomic Energy最新文献

The paper presents the results of validating the severe accident block of the EVKLID/V2 integral code used to calculate the processes of fission product release from the oxide fuel melt and dissociation of nitride fuel observed during the destruction of the core in a fast neutron reactor with liquid metal cooling. Based on the obtained results, the uncertainty of calculating individual parameters, including the fraction of released fission products and the loss of the fuel mass during dissociation, is presented.

The article presents a system for measuring the level and density of liquids in nuclear-safe devices, which is based on intelligent sensors and differential pressure measurements at several points of the process medium using capillary impulse lines. The presented system, which consists of a probe, differential pressure sensors, and a separating medium supply system, is operable when both air and liquid are used as a separating medium. The results of testing the system using a compressed air separating medium are presented.

The article describes a set of studies for estimating and substantiating the limit mass of graphite removed during the resource characteristic management of RBMK reactors. The limit mass is understood as the maximum mass of graphite that can be removed from the reactor core at the stages of resource characteristic management without violating the operational limits of neutron-physical characteristics during the operation of power units over a period of 45 years. The substantiation of the limit mass was made based on the detailed computational simulation of reactor operation for realistic operational scenarios of the third and fourth power units of the Leningrad NPP. Particular attention is paid to the change in the void coefficient for the predicted states of reactors. The uncertainty of the limit mass of removed graphite is analyzed. The trueness of the calculated predictive estimates of neutron-physical characteristics and the limit mass of removed graphite is confirmed by the measurement results. It is shown that, by taking into account the considered operational scenarios for the reactors of the third and fourth power units of the Leningrad NPP, nuclear safety is ensured.

The article presents an economic and analytical model that evaluates various criteria for the effectiveness of investments in spent nuclear fuel reprocessing plants at given capital, operating, and decommissioning costs, as well as expected project implementation deadlines. The model also solves the inverse problem of determining the requirements for plant parameters based on the criteria of investment effectiveness. The results of the calculation of the most important of these parameters, including the internal rate of return, the discounted payback period, and the levelized cost of reprocessing spent nuclear fuel, are presented. The mathematical definition of the net present value, representing the main criterion, depends on the plant capacity, which varies from 200 to 3000 t of fuel per year. The estimated conditions under which the net present value of reprocessing spent nuclear fuel does not exceed $1000/kg are presented.

The paper describes a methodology for calculating break probability and the applicability of a leak-before-break (LBB) concept for the secondary circuit pipelines at nuclear power plants with VVER reactors. As well as a calculation based on the LBB concept, an example of break elimination using probabilistic approaches is presented. The study was carried out on the feedwater and main steam pipelines of an AES-2006 project. As well as taking into account the known defectiveness, data setting used a calculation method in the absence of such information. An example of calculating secondary circuit pipelines using experimentally obtained defectiveness data is presented. The leak probability for various nominal crack diameters is provided taking into account the detection of cracks during non-destructive testing, as well as cases of non-detection.

The article proposes the use of permanent additional low-power steam turbines to back up the auxiliary units of nuclear power plants for removing decay heat from the reactor core. A study was carried out to develop a method of plant-wide multi-channel backup of nuclear power plant auxiliaries based on low-power steam turbines for a multi-unit station. In order to improve the efficiency of plant-wide auxiliary backup for a nuclear power plant having more than two power units, a method for controlled maintenance of the required reactor power at a lack of decay heat was developed based on the use of permanently operating low-power turbines for each power unit. Conditions for the removal of decay heat from one, two, and four reactors, which are demonstrated in the event of the nuclear power plant being disconnected from the power system, are based on the decay heat of one reactor and one low-power steam turbine. The study proved the possibility of creating a multi-channel plant-wide backup of nuclear power plant auxiliaries based on the equipping of each power unit with a low-power steam turbine.

A review of literature sources demonstrates the relevance of improving the accuracy of liquid flow measurements. To solve this problem, a neural-network model for liquid flow determination was developed and tested. The optimum structure and training parameters of an artificial neural network, such as the activation function, transfer function of the output layer, number of hidden layers and neurons in them, were selected. The training sample was generated using empirical expressions of GOST 8.586.1–2005 (ISO 5167–1:2022). The developed neural-network predictive model, which provides an uncertainty of calculations no greater than 0.32%, is intended for use as part of a software and hardware system for improving the accuracy of liquid flow measurements at nuclear industry enterprises.

Since 2021, a comprehensive program comprising five projects has been implemented by the Rosatom State Corporation in the Russian Federation. Of these, the project entitled “Development of controlled fusion and innovative plasma technologies” is scientifically supervised by NRC Kurchatov Institute. The program envisages a development of the concept of a hybrid reactor plant that combines fusion and fission technologies. As well as presenting an overview of the feasibility, functions, and parameters of planned fusion neutron sources, the paper outlines the development concept of the hybrid system and prospects for its implementation as a means of expanding the fuel base and recycling high-level wastes. An analysis of target indicators is provided.

The paper considers the processes involved in leaking fuel elements during the shutdown and preparation of a reactor for refueling in the case of a defect located at the bottom, in the middle, or at the top of the fuel column. The volume and activity of gaseous fission products released during fuel cladding failure detection were evaluated using the mast sipping system of the refueling machine. It is demonstrated that the sensitivity of the beta-radiometer, which starts at 37 kBq/m³, provides for the detection of leaking fuel elements at any power density located at various locations of the defect along its height. Neglecting to wash all fuel elements in the FA with bubble air may result in a failure to detect leaking fuel elements.

The concept of slowing down the kinetics of a fast reactor, according to which a rapid reactivity insertion is used to reduce the development rate of transients at initial events, is aimed at achieving a higher level of reactor self-protection. The paper presents a set of studies on the engineering and technical implementation of this concept on the example of a high-power fast lead-cooled reactor. Considering a hypothetical initial event with a rapid insertion of a complete reactivity margin, the stability of the reactor with the slowed down kinetics in relation to such an event was analyzed. An updated physical-mathematical apparatus is proposed for evaluating the slowed-down kinetics and calculating the dynamics of transient processes. The basic principles for the formation of a reflector for such a reactor based on formulated criteria for slowing down the kinetics include the use of ²⁰⁸Pb and weakly neutron-absorbing materials, as well as the optimum location of various structural elements in the reflector. The theoretical possibility of achieving a slowed-down kinetics in layouts that meet the developed principles is demonstrated.