The paper discusses the development and implementation of the algorithm for calculation of the three-phase equilibrium in the hydrocarbon/aqueous system. The work is done as a part of PVT module improvement for the “d-Flow” hydraulic simulator which is under development at Novosibirsk R&D Center LLC. The numerical methods used in the presented three-phase algorithm are simple and available in the literature, however these methods are often need to be improved and tuned before an actual application. In order to increase phase composition calculation stability, a modification for stationary point selection algorithm is proposed. Moreover, unexpected behavior of the algorithm may be revealed during implementation, which is not described in the original articles. This work attempts to present the three-phase equilibrium calculation algorithm ready for direct use in the designed hydraulic simulator and provide the detailed description for different aspects and difficulties of the implementation. The numerical schemes for the phase stability test and the calculation of equilibrium compositions are described, and the general equilibrium search algorithm is given. To test the algorithm, simulations of different mixtures are done and phase diagrams are presented for several mixtures. The solubility of hydrocarbons in water for binary mixtures is also calculated and compared with experimental data from the literature.

—This paper studies the development of approaches to expand the method for calculating the properties of liquid sodium in the region of overheating and at high temperatures. The expansion is carried out using the hypothesis of the linearity of isochores and interpolation of the properties of sodium on the saturation line using the theory of critical exponents. Using modified techniques, the position of the spinodal is calculated and an expression is given to estimate its position. A comparison is made with the available published data. Satisfactory agreement is obtained.

Abstract—An analysis of the relationships for calculating the thermal properties of liquid lead (hereinafter referred to as lead) was carried out, and the method for determining its heat capacity over a wide range of temperatures, including at high values, was chosen. This is especially important for numerical studies to justify the safety of designed reactor installations with liquid metal coolants, such as BREST-OD-300 and BR-1200. Measuring the properties of lead at temperatures close to the boiling point is often difficult due to the lack of reliable methods and materials that can withstand temperatures above 2273 K. At present, theoretical approaches to calculating the properties of simple liquids based on phonon theory are being actively developed. Such approaches can be used to derive semiempirical relations for the heat capacity of liquid lead that would allow physically correct extrapolation of the data to the high-temperature region. In this regard, the aim of this work is to obtain a relationship for calculating the heat capacity of liquid lead from its melting point to its boiling point based on modern theoretical approaches. To achieve the set goal, the following tasks were solved. Firstly, an analysis of the works of various authors was carried out and empirical formulas were selected that make it possible to reliably calculate the heat capacity at a constant volume cv (isochoric heat capacity) for a lead coolant from the melting point to 1500 K. Secondly, based on them, using phonon theory, an approximating formula was constructed, thanks to which it is possible to physically correctly extrapolate the properties of lead to the boiling point (2022 K).

Justification of fusion reactor safety is impossible without multiphysics codes that make it possible to model many coupled physical and chemical phenomena. Currently, multimodule integral codes for modeling fusion reactors are actively developed abroad, but there are no such codes in Russia. In this article, the requirements for the EUCLID-F integral code, its modular structure, and the list of modeled phenomena are formulated. The code is developed at the Nuclear Safety Institute of the Russian Academy of Sciences in order to analyze accidents at fusion reactors with magnetic plasma confinement. The design features, hazards, and emergency modes of fusion reactors are briefly described, and a comparison of the EUCLID-F code with foreign counterparts is given. In addition, the requirements for the unified database and high-fidelity codes necessary for development and performance of calculations by the integral code are formulated. Taking into account the described requirements when developing the integral code will make it possible to use it for analyzing the majority of design and beyond design basis accident scenarios. The code developed in accordance with the above requirements will significantly surpass foreign counterparts in terms of the number and variety of calculation modules and simulated phenomena. The availability of this code in Russia will significantly contribute to the development of fusion technologies.

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.

In order to ensure efficient and safe operation of oil and gas fields, a tool is required to design production wells and surface infrastructure. It is important to take into account the physical processes that occur during production, as well as fluid properties and phase transitions. The "d-Flow" software package enables the creation of a comprehensive field model based on geological and field data to calculate hydraulic losses of pipelines and forecast hydrocarbon production. The fluid flow modelling is based on the calculation of multiphase friction with the wall of a well or a pipeline. The article discusses the implementation of four friction models: the Beggs-Brill model, the Gray model and its modification, and the Mukherjee-Brill model. The purpose of this work is to compare the predictions of friction models with the results obtained using commercial realizations of the same models. The results of the Schlumberger PIPESIM hydraulic simulator calculations were used as a benchmark for comparison with the "d-Flow" models. Numerical experiments were conducted to investigate two-phase flow under varying well geometry and flow regimes. The models used were compared, and the results showed that the predicted liquid holdup had an average relative error of 0.06%. In some cases, the error was as low as 0.02%. The predicted cumulative pressure drop in the well did not exceed 0.34% for all considered models. Based on the comparison results, we conclude that the "d-Flow" software package is suitable for calculating pressure drop in wells of different geometry and surface networks.

This paper describes testing of the INSIM-FT proxy simulation method (interwell-numerical-simulation model improved with front-tracking method) to assess the dependencies between production and injection wells, as well as to assess the forecast of oil/liquid production by wells depending on their operation parameters. The paper proposes the approach of taking into account the influence of various production enhancement operations. The method was tested on a synthetic hydrodynamic model and on a sector of a real field. The results show a good match between historical data and simulation results and indicate significant computational efficiency compared to classical reservoir simulators.

Continuous monitoring and control of hydrocarbon flow is not a new task. Today, there are many engineering solutions in flow measurement that are used in commercial applications. However, the search for other solutions continues and is associated with the emergence of new technical challenges of the oil and gas industry. There is also a constant request for optimization of already working flow-metering solutions. In the present work we propose a methodology to calculate the composition of a mixture using gamma densitometry methods with direct and scattered radiation tracking. Experiments were carried out on reference samples and ways of practical implementation of the proposed methodology are shown.

— Application of computation tools resting on contemporary physical and mathematical models for substantiating the design solutions adopted for various heat-transfer equipment components helps save time, manpower, and financial resources of design institutions. The variety of both existing reactors and those being designed, which differ from one another both in design and type of coolants calls for the availability of a versatile thermal hydraulic computer code suited for a wide range of applications. The new-generation HYDRA-IBRAE/LM thermal hydraulic module of the EUCLID integrated code, which has been developed as part of the Proryv (Breakthrough) Project, meets these requirements. The operation of this thermal hydraulic module as part of the integrated code opens the possibility to simulate an essentially wider range of reactor plant operation modes and, as a consequence, those of individual heat-transfer equipment components. The developed thermal hydraulic module, which has been certified at the Scientific and Engineering Center for Nuclear and Radiation Safety (SEC NRS), offers the possibility to analyze the thermal hydraulics of sodium, lead, lead–bismuth, gas, and water coolants in various NPP equipment items. Reactor plant steam generators (SGs) belong to the category of equipment components most complex for modeling since they may contain two types of coolants. The article presents study results demonstrating the code’s abilities to analyze in a correct way the processes in the steam generators of only sodium cooled reactor plants, because these plants exist and are actively operated in Russia and around the world. The data presented in the article allow a conclusion to be drawn that the thermal hydraulic module developed at IBRAE RAS is an efficient tool for numerically analyzing complex heat-transfer processes in reactor plants. By using an extended system of closing correlations implemented in the module, it is possible to perform substantiation of design thermal engineering solutions as applied to individual heat-transfer equipment components.

— For numerically simulating the melt behavior in the core catcher of a sodium-cooled reactor, the HEFEST-FR module—a software tool based on the SAFR computer code elaborated at the Nuclear Safety Institute, Russian Academy of Sciences, for simulating the meltdown and destruction of liquid metal-cooled fast reactor core components—has been developed and incorporated into the EUCLID/V2 integrated computer code. This module is intended for numerically simulating the melt retention and cooling-down processes in the reactor vessel with taking into account the heat transfer from the vessel’s internal structures to the coolant. For this purpose, a 2D-problem (implying that there is no dependence of temperature on the azimuthal angle) of unsteady heat conduction for materials located in the reactor’s core catcher tray is solved in a cylindrical coordinate system. The heat-conduction equation coefficients depend on time, coordinates, and temperature, the latter being the solution of equation. Boundary conditions of the first, second, and third kind are used, and heat loss by radiation at the boundary is specified. The total or volumetric decay heat power is given inside of the melt. For numerically solving the 2D heat-conductivity problem, an enthalpy approach based method is applied. The formulations used in the method make it possible to overcome, in a natural manner, the problem relating to discontinuity of material specific melting point enthalpy in solving the heat-conduction problem with taking phase transitions into account. The solution yields the temperature field in melting/freezing of materials (steels of various grades and fuel) in the fast reactor’s core catcher tray. The results of verifying the HEFEST-FR module against the solution of an analytical problem have been demonstrated. Using the module, methodical computations of the fuel and fuel pin cladding melt behavior in the sodium-cooled reactor’s core catcher have been carried out.

В работе представлены подходы, которые могут быть использованы для анализа поведения ТВЭЛов со смешанным нитридным уран-плутониевым топливом во время аварийных ситуаций, сопровождающихся термическим разрушением ТВЭЛов. Приведены результаты валидации на доступных в настоящее время данных. Оценена погрешность расчетов по результатам валидации. На базе валидированной модели представлены результаты численного исследования особенностей разрушения ТВЭЛов со смешанным нитридным уран-плутониевым топливом для условий аварий, характерных для реакторных установок с натриевым и свинцовым теплоносителями.

The paper presents approaches that can be used to analyze the behavior of mixed nitride uranium–plutonium fuel rods during emergency situations accompanied by thermal destruction of fuel rods. The validation results based on currently available data are presented. The calculation error was estimated from the validation results. Based on the validated model, the results of a numerical study of the specific features of the breakdown of mixed nitride uranium–plutonium fuel rods are presented for accident conditions typical of reactor plants with sodium and lead coolants.

The development of computer codes for modeling accidents in a reactor unit requires validation of the models built into these codes. In this work, the EUCLID/V2 integrated code developed at IBRAE RAS was validated as applied to the simulation of severe accidents with a failure of the core of liquid-metal cooled fast breeder reactors (LMFBR), against experiments on melting of the cladding of fuel-rod simulators carried out at the Institute of Thermophysics, Siberian Branch, Russian Academy of Sciences (IT SB RAS), and SCARABEE BE + 3 experiments performed at the Commissariat à l’Energie Atomique (CEA) in France. The investigations performed at IT SB RAS included measurements of the cladding surface temperatures without liquid-metal cooling of the fuel-rod simulator, which is typical for accidents involving an instantaneous blockage of the flow section in the fuel assembly (FA) or with loss-of-coolant for type BN-1200M reactor units (RUs). To create such conditions, experiments with fuel rods were carried out in an argon atmosphere at room temperature (25°C) and a pressure of approximately 10 5 Pa, and the surface temperature of the fuel-rod simulator was recorded with a pyrometer. In France, the SCARABEE BE + 3 series experiments were carried out in the SCARABEE reactor to study the consequences of a hypothetical accident with a complete instantaneous blockage of the flow cross-section in a sodium-cooled fast reactor. To determine the effect of uncertainty in the initial data, diversified calculations were made. The validation was done by comparing the predictions with the experimental values of temperatures in the range between 500 to 1800 K (experiments of IT SB RAS). The maximum calculation error did not exceed 200 K. For the experiments in the SCARABEE reactor, it was not greater than 88 K for the fuel-rod claddings and 100 K the coolant. The obtained data will be used to estimate uncertainty in the predictions by the models of severe accidents with thermal destruction of fuel rods in fast reactors.

Models of physicochemical processes in a lead coolant, which can be used as part of the CFD code to justify the coolant technology, and as part of an integral code to describe the state of the reactor plant at the time of the start of the simulated dynamic regime are presented. The described models are integrated into the OXID module of the EUCLID/V2 integral code developed at the Institute for Problems of Safe Development of Nuclear Power Engineering, Russian Academy of Sciences. The OXID module makes it possible to calculate, on the basis of a two-layer model, the growth of an oxide film in a lead coolant on the surface of a metal that is in contact with liquid lead, the crystallization of dissolved oxides on the channel walls, the formation of lead oxide particles (nucleation) in a coolant flow, the coagulation of solid oxide particles in a lead coolant, the deposition of oxide particles on the channel walls, the crystallization of dissolved oxides on particles and dissolution of oxide particles in the lead coolant, and the transfer of formed impurities in the coolant. A system of relationships and equations is given that allows modeling of these processes as part of a software module. By integrating this module into the EUCLID/V2 code, it is possible to carry out connected calculations of physical and chemical processes in a lead coolant, which are typical for various operating conditions of a BREST-OD-300 reactor plant. To demonstrate the performance of the implemented models, the results of verification based on analytical tests and validation based on experimental data on the growth of an oxide film on the surface of stainless steel at different concentrations of oxygen in lead, obtained at the AO SRC RF-IPPE on the CM-2 circulation stand, are presented. The validation made it possible to estimate the errors in the calculation of individual parameters important for safety.