bluuebunny/crossref_metadata_2025 huggingface dataset:
RESULTS OF “BENCHMARK ROSTOV 2” TEST TASK SIMULATION USING SAPFIR_95&RC_VVER AND KORSAR PROGRAM PACKAGE
Authors: Artemov V , Artemova L , Korotaev V , Kuznetsov A
Published: January 2022

This paper presents the results of “Benchmark Rostov 2” test task simulation using SAPFIR_95&RC_VVER program package; the test task is based on a full-scale experiment performed at the Rostov NPP, where the transition mode with boric acid dilution and a working group insertion to compensate the reactivity were implemented. The purpose of this task is to develop the methods of non-stationary computation of pin-by-pin power energy distribution in calculation programs for neutronic and thermohydraulic characteristics of VVER-type reactors. The SAPFIR_95&RC_VVER program package uses two calculation methods of pin-by-pin power energy distribution: micro- and macro-flux superposition method and combined fine-grid method, where the nodes of radial computational grid in the reactor core coincide with the centers of fuel elements in fuel rod assembly. The combined fine-grid method for calculating pin-by-pin power energy distribution is carried out with the known values of burn-up distributions, fuel temperature, and coolant density obtained from superposition method. Both approaches are used for solving stationary and non-stationary tasks. When simulating the non-stationary processes, the test task can be solved in conjunction with KORSAR thermohydraulic code. This paper describes neutronic models of VVER-1000 reactor core used in “Benchmark Rostov 2” test task for superposition method and combined method. Calculation results for power energy distribution fields are given.
INTRODUCTION ENERGY DEPOSITION OF DELAYED GAMMAS USING KERMA FACTORS IN MCU-PTR
Authors: Kuznetcov A , Kalugin M , Oleynik D , Shkarovsky D
Published: January 2022

The paper describes the realization of upgrade precision Monte Carlo program MCU-PTR for introduction energy deposition of delayed gammas in energy release calculations. For this, based on the next position approximations are applied: after some time of reactor operation delayed gammas spectrum reaches asymptotic behavior and then does not change. This modification allows to take into account both promptly and delayed photons in spectrum of generation photons in neutron-photon task. The application of this approach was verified by calculating the task 2b of the VERA benchmark and by comparing the results with similar Serpent code results, which showed their good coincidence in spite of the various approaches in accounting for delayed photon radiation. This modification of the MCU-PTR program allows one to obtain the full value of radiation energy release in a nuclear reactor as a result of a single neutron-photon calculation. Thus, labor costs for performing such calculations are reduced, since now it is not necessary to perform additional calculation of the photon problem with a source of delayed photons.
THE ANALYSIS OF MIR REACTOR FULFILLED FUEL HANDLING NUCLEAR SAFETY
Authors: Fraaz E , Akimov A , Malkov A
Published: January 2022

The main results of neutron and physical parameters calculations, which are important for nuclear safety during storage and transportation of irradiated working and loop fuel assemblies and irradiation devices of the MIR reactor, as well as working in radiation-protective research cameras for the assembling and disassembling of irradiation devices, are introduced in this report. The analysis was made in full accordance with requirements of the state normative research nuclear facilities nuclear safety documents. The results of effective neutron reproduction coefficient calculations received by the MCU-RFFI/A program were used for the analysis. The various emergency situations associated with personnel errors, falling of products during their transportation, pouring out the contents of covers, violation of the products placement order, changing in inhibitor (in particular water) density were considered during the analysis. The full unloading of the reactor core in case of an accident was also considered. As a result of the neutron and physical calculations analysis, safety of spent nuclear fuel storage, at all stages of its transportation, as well as working in radiation- protective research cameras subject to established restrictions was proved. Restrictions for the number of products and the geometry of their placement in the cooling pool, transport corridor, and radiation - protective cameras were set.
CALCULATION OF NON-STATIONARY THREE-DIMENSIONAL NEUTRON-PHYSICAL TESTS OF THE INTERNATIONAL BENCHMARK C5G7-TD USING THE CODE SUHAM-3D-TD
Authors: Boyarinov V
Published: January 2022

Three-dimensional tests of the international non-stationary benchmark C5G7-TD were calculated using the SUHAM-3D-TD code. Non-stationary three-dimensional finite-difference equations of the surface harmonics method with three transversal and two longitudinal trial matrices for reactor with square lattice are realized in the SUHAM-3D-TD code. This paper presents the results of calculations of five non-stationary tests associated with movement of groups of control rods (tests TD4-1, TD4-2, TD4-3, TD4-4 and TD4-5) and four non-stationary tests associated with the disturbance of the coolant density in the core of a model light water reactor of the PWR type, namely, tests TD5-1, TD5-2, TD5-3 and TD5-4. All tests of the C5G7-TD benchmark are “blind”, i.e. the numerical results of the calculations of these tests will be suitable for comparison only after a certain time, when most of the codes involved in the calculations will calculate these tests. As the results of the calculated tests, the distributions on time of the total power and Henry reactivity of the calculated object, the dependence on time of the total fuel assemblies power, the dependence on the longitudinal coordinate z of the power of the fuel assemblies for individual time points and the distribution of the local power integrated along the z axis as function of the x, y coordinates for individual time points are given.
SUBSTANTIATION OF TECHNICAL SOLUTIONS OF THE REACTOR CIRCUIT OF THE BRS-GPG UNITS OF SMALL AND MEDIUM POWER WITH A HEAVY LIQUID METAL COOLANT
Authors: Beznosov A , Bokova T , Bokov P , Marov A , L'vov A , Volkov N
Published: January 2022

The analysis and new scientific and technical solutions corresponding to the evolutionary development of low and medium power reactor plants with heavy liquid metal coolants are presented. The analysis was carried out on the basis of experience in the creation and operation of reactor installation with lead-bismuth coolant and research, primarily experimental, performed at the Nizhny Novgorod State Technical University (NSTU) in rationale of the small and medium-sized power plants developed at NNSTU reactor installation with horizontal steam generators (BRS-GPG). The paper substantiates the choice of coolant. With the proposed arrangement of the reactor loop, the maximum possible natural circulation is achieved, which significantly increases the safety of reactor installation. The results of studies (simulations) of one of the most potentially dangerous accidents in reactor installation with HLMC - ″large steam generator leak″ can qualitatively reduce the consequences of the accident when using the design of a horizontal steam generator in which the pipe system is placed with minimal burial of pipes under the level of HLMC, which eliminates the flow of water into the reactor core, re-pressurization of the reactor circuit, etc. It is probably advisable to consider the characteristics of the core, analog ary accepted for SVBR or BREST-OD-300. A system is proposed for cooling the reactor and providing the reactor installation shutdown modes, its characteristics are studied and tested at the NSTU stands.
PARALLELIZATION OF THE ODETTA CALCULATION CODE FOR SOLVING THE PROBLEMS OF NEUTRON AND GAMMA TRANSPORT
Authors: Shoshina A , Belousov V
Published: January 2022

The ODETTA calculation code and a comparative analysis of the program simulation with using of MPI principles as well as the possibility of applying MPI technology to speed up the calculation code are considered. The development of the ODETTA program algorithmic part with MPI using is described. Features and specifics of this program parallelizing are presented. The modification of the calculation code is given. The corresponding software algorithms are considered. The structure of ODETTA program modules with MPI using is displayed. The results of ODETTA serial and parallel versions in OS Linux with MPI using on the HPC cluster of National Research Nuclear University MEPhI are presented. In addition, the comparative analysis of two program implementations in terms of speed and accuracy of results by using of two different clusters and a different number of nodes on them is given.
MODELING OF HEAT TRANSFER IN ROUND CHANNELS OF CONTINUOUS CASTING MACHINES WITH PROMISING LEAD-BISMUTH CRYSTALLIZER
Authors: Ulyanov V , Koshelev M , Konovalov M , Kharchuk S
Published: January 2022

As part of the search for new areas of application of liquid metal coolants, their use for cooling crystallizers of steel in continuous casting machines instead of water has been proposed. This solution will reduce energy costs in the production of steel due to the implementation of the steam-power cycle with the generation of electricity. This article describes the results of the justification of the optimal liquid metal coolant and the modeling of cooling of a round crystallizer channel by it. It is shown that the eutectic alloy of lead with bismuth, which was successfully used as a coolant for ship reactors in the Soviet Union, is best suited to solve this problem. A comparative simulation of cooling of hardened steel in a round channel with a lead-bismuth eutectic with temperatures of 180 °C and 450 °C and water with a temperature of 20 °C revealed that replacing water with a liquid metal coolant practically does not reduce the efficiency of steel production. The greatest heat flux is removed from the steel crystallizer at a temperature of lead-bismuth eutectic at the inlet of the steel crystallizer equal to 180 °C. The temperature of the lead-bismuth eutectic at the outlet of the steel crystallizer reaches 400 °C, which is sufficient to generate electricity in the steam-power cycle. The choice in favor of modeling in comparison with the experiment was made in accordance with current trends in cost reduction due to the use of computer-based calculation tools. It is advisable to continue research to select the wall material and determine the optimal shape of the cooling channels of the crystallizer for maximum intensification of heat transfer between hardening steel and lead-bismuth coolant.
DEVELOPMENT AND SUBSTANTIATION OF MASS TRANSFER APPARATUS DESIGNS FOR REGULATING OF THE OXIDATIVE POTENTIAL IN LEAD-BISMUTH COOLANT OF NEW GENERATION REACTOR FACILITIES
Authors: Askhadullin R , Legkikh A
Published: January 2022

In a lead-bismuth heat carrier, to ensure the conditions for the formation and maintenance of protective oxide films on the surfaces of structural elements in contact with the heat carrier, it is necessary to regulate the dosage of dissolved oxygen into the heat carrier. In the absence of feeding the lead-bismuth heat carrier with dissolved oxygen, the oxidative potential of the heat carrier can decrease to values at which the development of corrosion and erosion processes begins. A solid - phase method for regulating the oxidizing potential in a lead-bismuth heat carrier, which is implemented using mass transfer devices, has been developed and justified at the SSC RF-FEI. The authors of this paper develop various structures of mass transfer devices for use in research installations with lead-bismuth coolant, as well as in advanced reactor installations of the new generation, in which the lead-bismuth melt is considered as a primary coolant. The article presents the results of the development and justification of mass transfer devices that can be used to regulate the oxide potential in the lead-bismuth coolant of a new generation of reactor plants.
MATHEMATICAL SIMULATION OF THE LASER RADIATION BEAMS DISTRIBUTION IN MOVING NON-EQUILIBRIUM IRRADIATED BY NEUTRONS DUSTY PLASMA CONTAINING URANIUM NANOPARTICLES
Authors: Budnik A , Sliuniaev M
Published: January 2022

The use of uranium nanoparticles dispersed in a laser-active medium instead of traditionally used methods of heterogeneous nuclear pumping of active gas medium can increase the fraction of energy carried by fission fragments from the condensed phase into the gas medium by more than an order of magnitude. This creates the conditions for increasing the efficiency of nuclear into optical radiation energy conversion. However, the scattering and absorption of laser radiation by an active medium containing uranium nanoparticles are important factors preventing the laser radiation generation in such a medium. Mathematical modeling of the generation and amplification of laser radiation in a moving spatially inhomogeneous nuclear-excited neutron-irradiated argon-xenon medium containing uranium nanoparticles injected into a cylindrical laser-active dusty element, taking into account optical inhomogeneities caused by the dust component. The process of amplification and intensity distribution of a master oscillator laser beam in a moving spatially inhomogeneous nuclear-excited argon-xenon neutron-irradiated medium containing uranium nanoparticles has been studied for the first time. Data were also obtained on the propagation of a light beam near a relatively sharp dust component concentration decrease in moving gas medium. As a brief result: moving spatially inhomogeneous nuclear-excited neutron-irradiated argon-xenon medium containing uranium nanoparticles does not have a pronounced destructive effect on the propagation of laser radiation in it, on condition that the light flux is not wider than the dust component.
INVESTIGATION OF GETTER PURIFICATION OF SODIUM FROM OXYGEN
Authors: Kuzina Yu , Alekseev V , Sorokin A , Voronin I , Konovalov M , Zykova R
Published: January 2022

The article focuses on a comprehensive analysis of the new concept of a combined sodium oxygen purification module. The combined purification module includes a combination of two modules: chemisorption purification with a soluble getter and high-temperature getter purification with a high-temperature getter. A special zirconium alloy in a finely dispersed granular form can be used as a high-temperature getter. Other materials (e.g. titanium and various alloys) can be used and must be tested accordingly. At the first stage of the work, research was carried out on the module for chemisorption purification of sodium from oxygen at the Protva-1 stand, which showed the effectiveness of purification using the proposed getter materials, at the second, research on the module for high-temperature getter purification. For 20 hours of operation of the chemisorption module, the oxygen concentration in sodium was reduced by 12 ppm (3 g oxygen), while 4.5 g of magnesium reacted (the initial mass of magnesium was 25 g). According to the results of experiments in the sodium loop, the decrease in the concentration of dissolved oxygen in sodium during purification with a zirconium getter for 20 h was about 50 ppm. Two promising options for the development of a combined module for the purification of sodium from oxygen are proposed. Experimental justifications of selected chemisorbent, getter and filtering material are carried out. The main part of the study is analysis of chemisorption efficiency, purification process dynamics and influence on it of temperature factor for selected chemisorbent.
TO THE QUESTION ABOUT THE INFLUENCE OF DISPERSED IRON OXIDES ON THERMODYNAMIC STATE OF HLMC UNDER NON-ISOTHERMAL CONDITIONS
Authors: Osipov A , Ivanov K
Published: January 2022

In the initial period of HLMT development, it was shown that ensuring the corrosion resistance of structural steels is associated with ensuring a certain oxygen regime of the coolant. To control it, oxygen TDA sensors were created later. It was assumed that the oxygen activity in HLMC in different temperature zones of the circulation loop obeys the so-called "isoconcentration" distribution, which, when formally recalculated the measured oxygen TDA values, gives a constant concentration value at different temperatures. However, later it turned out that such a distribution of TDA is not always realized. The observed character of deviations could be explained by the influence of iron impurity. At the same time, quantitative estimates of this effect were carried out under the assumption of the formation of stoichiometric magnetite under conditions of its thermodynamic stability in the entire range of temperatures and concentrations of the initial components. The limitation of this approach lies, first of all, in the fact that it does not take into account the processes of dissociation of solid-phase iron oxides, which can occur in the hot zone when appropriate conditions are created, which can be realized with a decrease in the content of dispersed iron oxides. The importance of taking this factor into account is due to the fact that, during corrosion testing of steels in HLMC, the processes of dissociation of magnetite are actually observed in practice. Within the framework of this work, a computational method for assessing the effect of filtration processes on the thermodynamic state of HLMC has been developed. Quantitative estimates of the effect of the content of dispersed oxides in HLMC on the thermodynamic state of the coolant under non-isothermal conditions in iso-concentration and non-iso-concentration modes have been obtained and which can be used in calculation codes and comparison of experimental results with calculations.
CONVERSION OF RBMK-1000 REACTORS TO REPROCESSED FUEL WITH AN INCREASED CONTENT OF EVEN URANIUM ISOTOPES
Authors: Alimov Yu , Galeyeva N , Davydov V , Zhirnov A , Kuznetsov P , Rozhdestvenskiy I
Published: January 2022

The structure of the Russian nuclear power industry includes reactors with different designed fuel enrichment. It is possible to mix, in certain proportions, nuclear fuel (NF) from various reactors, thus closing the nuclear fuel cycle. Reprocessed uranium is a product of radiochemical reprocessing of spent nuclear fuel (SNF) from NF with a high initial enrichment. Use of uranium-erbium fuel based on reprocessed uranium is planned for the RBMK-1000 reactor. Along with 235U and 238U, SNF contains non-fissionable ballast isotopes of uranium (232, 234, 236 U). The 232,234U isotopes have a relatively high radioactivity and the presence of these leads to an increased dose rate of ionizing radiation but, due to their small content in fuel, does not affect the neutron balance, the neutron multiplication factor, and the reactivity margin. A large presence of 236U requires additional enrichment with 235U due a greater probability of inefficient neutron absorption by the 236U nuclei. This absorption with no fission leads to a reduced neutron multiplication factor, a reduced reactivity margin in fresh fuel, and a smaller burn-up of unloaded fuel. Analyzing the effects the increased content of even uranium isotopes (IEI) has on the reactor’s neutronic performance and fuel burn-up makes it possible to determine the amount of additional 235U fuel enrichment to make up for the negative effects of 236U on the RBMK-1000 neutronic performance.
COMPARISON OF THE RESULTS OF NUMERICAL SIMULATION OF MEASUREMENT OF SCRAM-SYSTEM EFFICIENCY AND MEASUREMENT PERFORMED DURING THE REACTOR PHYSICAL STARTUP TESTS OF UNITS 1, 2 NOVOVORONEZH-2 NPP AND UNIT 1 LENINGRAD-2 NPP
Authors: Kulikov V , Zhylmaganbetov N , Popykin A , Smirnova A , Kryakvin L , Pitilimov V , Sedov O , Tereshonok V , Sizov R
Published: January 2022

The article represents the results of measurement and numerical simulation of measurement of scram-system efficiency of reactor plants performed during the reactor physical startup tests of Novovoronezh-2 units 1, 2 and Leningrad-2 unit 1. These power units are equipped with a reactor VVER-1200, therefore, one numerical simulation of measurement of scram-system efficiency on the basis of the reactor project was performed. The article represents comparisons of the calculated and the measured currents of the ionization chambers and the readings of reactimeter calculate for them during measurement of scram-system efficiency.
NON-STATIONARY THREE-DIMENSIONAL TEMPERATURE FIELD IN A MULTILAYER CYLINDER
Authors: Levchenko V , Kascheev M , Dorokhovich S , Zaytsev A
Published: January 2022

The problem of determining a non-stationary three-dimensional temperature field in a k-layer cylinder of length is solved. There is a symmetrically located cylindrical cavity in the center of this body. The absence of a cavity is a special case of the problem. In each layer, there are heat sources, depending on the coordinates and time. The initial temperature of the layers is a function of the coordinates. In the center of the body the symmetry condition is fulfilled. At the boundary of contact of the layers - ideal thermal contact: continuity of temperatures and heat flows. On the inner and outer side surfaces and ends, heat exchange occurs according to Newton's law with environments whose temperatures change over time according to an arbitrary law. The periodicity condition is set for the angle φ. The problem in this statement is solved for the first time. For the solution of the problem the following approach is used: by means of the method of finite integral transformations differential operations on longitudinal coordinate, angle and transverse coordinate are sequentially excluded, and the determination of time dependence of temperature is reduced to the solution of the ordinary differential equation of the first order.
COMPUTATIONAL AND THEORETICAL EVALUATION OF THE PARAMETERS RESPONSIBLE FOR THE COMPATIBILITY OF METALLIC MATERIALS WITH THE LIQUID SN-20% LI ALLOY
Authors: Krasin V , Soyustova S
Published: January 2022

The main features of the thermodynamic evaluation of the parameters responsible for compatibility of metal materials with liquid Sn-20%Li alloy are considered in the article. Interest in the study of the physicochemical properties of liquid lithium-tin alloys is associated with the prospects for their use in plasma facing components of tokamaks. The main advantages of capillary-porous systems with a liquid metal in comparison with solid materials are their resistance to degradation of properties under tokamak conditions and the ability to self-repair the surface. Due to the fact that liquid tin is a very corrosive metal with respect to many structural materials, the advancement of liquid Li-Sn alloys is largely constrained by the lack of systematic studies of the corrosion resistance of structural materials in contact with these liquid alloys. To calculate the temperature dependences of the solubility of metals in the liquid Sn-20% Li alloy, the method of thermodynamic modeling was used, which included the following steps: (1) selection of models for the Gibbs energy functions; (2) selection and evaluation of input data; (3) optimization of model parameters; (4) calculations and comparisons. Using information on the excess Gibbs energies of mixing for the liquid phase in the form of the Redlich-Kister polynomial decomposition for the corresponding binary systems, the temperature dependences of the solubility of nickel, iron, chromium, molybdenum, and tungsten in the liquid alloy Sn-20% Li were calculated by thermodynamic modeling.
CALCULATION OF GAS PARAMETERS IN A NPP CONTAINMENT TAKING INTO ACCOUNT THE ADIABATIC ISOCHORIC COMPLETE COMBUSTION OF HYDROGEN BY THE KUPOL-M CODE
Authors: Kryglikov Y , Supotnitskaya O , Yantseva L
Published: January 2022

An important factor for the analysis of hydrogen safety is to take into account the effect of possible hydrogen combustion on the parameters of NPP containment atmosphere. In order to evaluate this effect a model of adiabatic isochoric complete combustion (AICC) of hydrogen was developed and implemented in the KUPOL-M containment code. With the model of adiabatic isochoric complete combustion of hydrogen included in the KUPOL-M code, this code will become applicable for the VVER NPP safety analysis performed with the aim to estimate possible thermal mechanical loads on the containment in case of hydrogen burning. This model is simple, conservative and based on the laws of conservation of mass and energy. The following assumptions are made in the model of adiabatic isochoric complete combustion of hydrogen: the volume in which burning takes place is constant; there is no heat transfer; the limiting component (in most cases it is hydrogen) burns down completely; intermediate stages of chemical transformations are not considered. The test results of this model on experimental data are presented in the article. There is a good agreement between the data of calculations and experiments in a wide range of hydrogen concentrations in gas mixtures. Based on the test results it can be concluded that implementation of AICC model in the KUPOL-M code allows conservatively evaluating the parameters in the containment atmosphere without significantly increasing the calculation time.
THE COOLING SYSTEM OF THE PERIODIC PULSED REACTOR
Authors: Gribacheva V , Shcherbakov S
Published: January 2022

The paper proposes the configuration and composition equipment of the cooling system of the designed periodic pulsed reactor (PPR) of high power. The special features of the PPR are a small flow section, a large heating of the coolant in the power pulse and the impossibility of useful use of thermal energy in the periodic mode of operation. Liquid lithium is proposed as a coolant and heat is discharged through air heat exchangers (AHE). The goal was to achieve compactness and low power consumption, the ability to work with frequent stops and optimize the operation of equipment in pulse modes. For this purpose, high-temperature AHE with a small heat exchange surface and forced air cooling are used, the circulation circuit is divided into two parts - the reactor circuit and the AHE circuit with two circulation pumps and a common drain tank. The separation of the circuit allows to independently perform the operations of starting, stopping and heating the circuits in a periodic mode. The drain tank limits the composition of the equipment exposed to temperature pulses. Numerical studies of the temperature regime of the coolant in the equipment of the PDR cooling system are carried out. The calculations were performed using the TURBOFLOW code in two-dimensional terms for all the main elements of the equipment. Quasi-stationary (nominal and partial power levels) and pulse modes of operation are considered. Calculated characteristics for forced and natural air circulation are obtained. The limits of the air circulation modes under the conditions of non-freezing of the coolant are determined. The obtained values of the maximum temperatures of the coolant: in the pulsed mode is 750 °C, in the quasi-stationary mode - 490 °C with an average power of 15 MW, air flow of 150 m3/s and the size of the AHE in the plan of 5×5 m, 100 panels of 1.08×0.025×5 m.
DETERMINATION OF THE COMPLEX OF THERMOPHYSICAL PROPERTIES OF BORIC ACID WATER SOLUTIONS AT PARAMETERS CHARACTERISTIC FOR VVER EMERGENCY MODE
Authors: Shlepkin A , Morozov A , Sakhipgareev A , Kalyakin D
Published: January 2022

The article presents the results of experimental studies of the thermophysical (density, viscosity) and physicochemical (degree of acidity - pH) properties of water solutions of boric acid. A review of the available literature data on the effect of the properties of boric acid solutions on heat removal from the reactor core is presented. It has been established that the available information is very general and does not cover the entire range of parameters (temperature, pressure, acid concentration) characteristic of a possible emergency at a nuclear power plant with a VVER reactor. Methods and facilities for conducting experimental studies are described. The results of experimental studies are presented. The density of aqueous solutions of boric acid with a concentration of 2.5-450 g/kg H2O at a temperature of 25-130 °C was determined. The dependence of the investigated characteristic on temperature and concentration was also obtained. The results of an experimental study of the kinematic viscosity of water solutions of boric acid with a concentration of 2.5-200 g/kg H2O at a temperature of 25-90 °C were obtained. The total error in measuring the viscosity of aqueous solutions of boric acid did not exceed 2 %. The pH values of water solutions of boric acid in the temperature range 25-50 and concentrations of 2.5-450 g/kg H2O were determined. The dependence for calculating the degree of acidity of boric acid is obtained. Experimental data on the thermophysical and physicochemical properties of water solutions of boric acid can be used to refine the results of calculations of emergency heat removal processes in a reactor facility, carried out using both one-dimensional calculation programs and three-dimensional CFD codes.
APPLICATION OF KALMAN FILTER TO THE ESTIMATION OF 8-GROUP DE-LAYED NEUTRON SPECTRA FOR THE 235U THERMAL NEUTRON FISSION
Authors: Egorov A , Piksaikin V , Gremyachkin D , Mitrofanov K , Mitrofanov V
Published: January 2022

In the present work the estimation of the 8-group delayed neutron energy spectra emitted in the fission of 235U by thermal neutrons have been done. Aggregate delayed neutron spectra that were measured with high resolution in the different time intervals after the thermal neutron irradiation of the 235U sample presented in the IAEA reference database for beta-delayed neutron emission have been used for the estimation process. Due to the lack of the uncertainties for the spectra presented in this database and the necessity of this data in the process of estimation of delayed neutron group spectra at first the estimation of uncertainties of these spectra have been done. In the estimation process the following components of uncertainty have been considered: statistical; the component due to the neutron background; component due to the distinctive features of neutron spectra restoration process measured using spectrometer filled by helium-3 and also the component appearing due to introduction of the correction on the distortion of the neutron spectra as a result of passing through the lead sheilding. Estimation of spectra have been made using the technique based on the Kalman filter that allows to take into account the uncertainties of the input data. Estimation results are shown in the numerical and graphic form.
SURFACE TENSION OF Pb44,6Bi55,4 EUTECTIC MELT IN AIR ATMOSPHERE
Authors: Alchagirov B , Khibiev A
Published: January 2022

In the 1950s liquid lead and the lead-bismuth eutectic alloy (Pb45Bi55) were considered candidates for their use as coolants for nuclear power systems in the USSR and the USA. At the same time, the alloy (Pb45Bi55), first proposed by A.I. Leypunsky, was chosen as a coolant for the nuclear submarine “Alpha”, which was ahead of its time in terms of its tactical and technical data. However, in 1968, one of them suffered a severe radiation accident caused by the melting of fuel elements in the core of an onboard nuclear reactor due to the precipitation of oxides from the coolant and their accumulation, which blocked the pipeline cross-section and sharply worsened the cooling of the reactor, which led to its failure. In fact, the primary cause of the accident was a lack of knowledge about the physicochemical and technological properties of the lead-bismuth coolant. Thus, the main disadvantage of the Pb45Bi55 coolant is its corrosiveness to structural materials used in the nuclear power plant. But it has been found that corrosion by liquid lead alloys can be reduced by adjusting the oxygen level in the coolant. For example, the corrosion rate of martensitic steel at 770 K in a Pb45Bi55 flowing coolant without oxygen is about 1 mm per year, but it can be reduced to 0.01 mm per year, i.e. 100 times, if oxygen is dissolved in Pb45Bi55 coolant and its mass concentration is maintained at the level of 0.01 ppm. The observed effect is explained by the protection provided by the oxide layer formed on the steel surface of the pipeline. Thus, for a deeper understanding of the phenomena occurring at the boundaries of the liquid metal “coolants - gases” section, it remains relevant to study the processes of formation and destruction of the protective oxide layer and its behavior in coolant fluids, especially from the point of view of long-term operation of nuclear power plants. In this regard, data on the surface tension of the “coolant - gas” interphase boundaries are of great scientific and practical importance. In this connection, the present work sets the task of experimentally studying the influence of atmospheric air on the surface tension of Pb45Bi55 eutectic melt. Measurements of the surface tension of the eutectic melt Pb44,6Bi55,4 prepared by the authors were carried out in a non-stop mode sequentially, under static vacuum and atmospheric air, on the same surface. About three hundred experimental points obtained in this work made it possible to describe the dynamics of the surface tension changing process depending on the time of exposure of the coolant surface in vacuum and atmospheric air. It is shown that in comparison with the results obtained by the authors by the large lying drop method in a static vacuum, in the first 10 minutes from the beginning of the exposure of the eutectic melt in atmospheric air at a pressure of about 300 mmHg), the surface tension of the eutectic Pb44.6Bi55.4 decreases by 55 mN/m and more, which is an order of magnitude higher than the total error (2 %) of our measurements.
FEATURES OF THE FORMATION OF HYDRODYNAMIC LOADS ON THE FUEL RODS DEPENDING ON THE STRUCTURE OF THE COOLANT FLOW AT THE ENTRANCE TO THE FUEL ASSEMBLY WWER
Authors: Perevezentsev V
Published: January 2022

Experimental studies of hydrodynamic induced of fuel rods vibrations WWER-440 tweers using a full-scale mock-up with lead imitators of fuel tablets have been carried out. It is shown that the speed of the turbulent flow in the fuel rods beam cannot be the only hydrodynamic characteristic that determines the vibrational characteristics of the fuel rods beams. The presence of indignant flow of various elements of the tract of the coolant to the fuel rods beams inlet significantly affects the intensity of their vibrations. An energy model of the balance of pulsating energy of turbulent flow and expendable mechanical energy to the fuel rods beams has been developed. On the basis of the proposed model, a functional connection of the intensity of the vibrations of the fuel rods beams with the levels of pressure pulsations in the turbulent flow of the coolant at the entrance to the fuel rods beams is established. The modelling of Fuel Assembly (FA) as a mechanical vibrational system has received considerable attention. At the same time, the end-element methods, as the most universal for the sampling of spatial continual systems, have become widespread. However, the analysis of hydrodynamically excited vibrations is impossible without describing the processes of interaction of the flow with the streamlined surfaces of the mechanical vibration system. In most cases, such information can be obtained only through experimental research.
INVESTIGATION OF THE CHARACTERISTICS OF LEAD OXIDE GRANULES AFTER PROLONGED EXPOSURE IN LIQUID LEAD
Authors: Skobeev D , Legkikh A
Published: January 2022

In order to ensure the safe operation of reactor installations under development with heavy liquid metal coolants (HLMC), such as lead and lead-bismuth, it is necessary to address issues related to the control and regulation of the oxygen potential of the coolant. This is necessary to maintain the normalized range of dissolved oxygen concentrations to ensure the conditions for the formation and maintenance of the integrity of protective oxide coatings on structural materials, as well as to prevent the formation of oxide slags from the coolant during the operation of the reactor plant. Specialists of IPPE have developed a method and a means of solid-phase regulation of the thermo-dynamic activity of oxygen in HLMC. In mass transfer devices developed for the implementation of the solid-phase method of controlled feeding of a heavy liquid metal coolant with dissolved oxygen, lead oxide granules are used as a filler. One of the important issues in substantiating the reliability of mass transfer devices is the question of the constancy of the mechanical properties and chemical composition of lead oxide granules after their prolonged stay under the level of a heavy liquid metal coolant at operating temperature. The lead oxide granules were aged in a sealed container filled with lead. The tank was equipped with the necessary means to control the temperature of the lead and the pressure of the protective gas (argon). The article presents the results of experimental studies of lead oxide granules after their exposure in a lead melt at a temperature of (420±10) °C in a non-carbon mode for 6000 hours. To investigate the characteristics of the lead oxide pellets, a batch of pellets was excavated after 500, 750, 1000, 1750, 2500, 3000, 4500 and 6000 hours from the start of the tests. For the pellets from each recess, the following studies were performed: - pellet density measurements; - measurement of breaking forces of lead oxide granules; - determination of pellet impact strength; - chemical analysis of granule composition.
THE PHYSICAL MODEL OF MASS TRANSFER CORROSIVE HYDROGEN IN FAST REACTOR STEAM GENERATORS OF THE SODIUM-WATER TYPE
Authors: Smykov V , Kanukhina S , Legkikh K
Published: January 2022

For timely and prompt detection of the inter-circuit depressurization of the steam generator of the “sodium-water” type in the 2nd circuit, continuous monitoring of the content of hydrogen, as a product of the interaction of sodium with water, is carried out in all operating modes. However, in reality, the main source of hydrogen in the sodium of the 2nd circuit is the process of electrochemical corrosion of 10X2M steam generator steel in a steam-water medium, the speed of which is never zero and is controlled at the minimum achievable level by means of the water-chemical regime of the 3rd circuit. At the same time, during the operation of the steam generator, deposits of corrosion products accumulate on the surface of the structural steel from the side of the 3rd circuit. The removal of deposits is carried out with the help of operational chemical flushes (ECPs). However, during the ECP, the corrosion rate of steel inevitably increases significantly (by about three orders of magnitude), which leads to an increase in the concentration of hydrogen in water and in sodium. This phenomenon has previously caused concern to the operational services of nuclear power plants due to the likely through-corrosion during regular ECP (every year, one of the three steam generators is washed). However, it is unacceptable to operate a steam generator without ECP, and to reduce the corrosion rate of steel of steam generator pipes in a steam-water environment, after ECP, a chemical passivation stage is carried out with different formulations, the effectiveness of which is different. This article presents a physical model of the mass transfer of corrosive hydrogen in a steam generator during chemical washing, passivation and start-up of the steam generator. An operational criterion for evaluating the effectiveness of passivation by the rate of increase in the concentration of hydrogen in the sodium of the second circuit during the subsequent start-up of BN-600 steam generators is proposed: if the rate of increase in the concentration of hydrogen in the 2nd circuit is less than 0.05 ppm/h, then passivation is effective, if more than the passivation formula requires revision. For BN-800 steam generators, the technology of passivation after ECP has not yet been determined, the article proposes a calculated criterion for its efficiency of 0.015 ppm/h.
APPLICATION OF AEROSOL WATER SPRAY TO INCREASE THE EFFICIENCY OF COOLING OF FINNED TUBES
Authors: Shlepkin A , Morozov A , Sorokin A
Published: January 2022

The article presents an overview of works on the experimental and computational study of the processes of heat removal from heated surfaces using a water-air mixture. A sharp increase in the heat transfer coefficient is shown even when adding water with a mass content of 0.1 % to the air flow. The factors that determine the efficiency of the heat exchange process are listed: the shape of the water spray jet, the distance from the outlet point of the water-air flow to the heat exchange surface, the characteristics of the heat exchange surface, the method of jet formation, the size of droplets and the location of the outlet points of the gas-droplet flow. The weak applicability of the data available in the literature is shown for calculating the cooling of finned tubes of industrial heat exchangers using a water-air mixture. It is substantiated that in order to establish the most optimal cooling modes for each for a separate heat exchange surface, it is necessary to conduct experimental studies, due to the complexity of heat transfer processes and the presence of a large number of influencing factors. It is shown that the finned tubes of heat exchangers of the passive heat removal system of WWER-1200 have a number of important features that affect the efficiency of their cooling using a water-air mixture. An experimental setup has been developed and a technique has been proposed for performing experiments to study these processes as applied to heat exchangers of a passive WWER safety system.
DISSOCIATION OF IRON OXIDES IN MELTS OF HEAVY METALS
Authors: Osipov A , Ivanov K , Arnol'dov M
Published: January 2022

At present, the problem of providing corrosion resistance of steels in heavy liquid metal coolants is solved by forming and maintaining protective oxide films on the surfaces of structural steels. However, oxide films are not absolutely impermeable barriers for the components of steels (primarily iron), which, as a result of diffusion processes, inevitably enter the coolant and can interact with oxygen to form solid-phase oxides. Within the framework of this work, a method is considered for obtaining the numerical values of the quantities characterizing the processes of dissociation of iron oxides as a function of temperature and oxygen potential of HLMC. The performed calculations made it possible to obtain specific numerical values of the limiting solubilities of iron oxides as a function of the temperature and the initial state of the coolant with respect to iron and oxygen impurities, and also made it possible to obtain numerical values of other thermodynamic parameters characterizing the current and limiting state of HLMC. The considered approach is of a general nature and can be used in experimental studies of the kinetic and thermodynamic characteristics of the dissociation processes of compounds in liquid metals.
TRANSMUTATION OF MINOR ACTINIDES IN MOLTEN SALT BURNER REACTOR
Authors: Belonogov M , Volkov I , Modestov D , Simonenko V , Khmelnitsky D , Eliseev V
Published: January 2022

Reduction in the amount of accumulated minor actinides (Np, Am, and Cm isotopes), as well as, the long-lived fission products, to the extent possible, is a top-priority task for the nuclear power engineering. Transmutation of this waste in a special-purpose molten-salt burner reactor has been currently proposed as a possible solution for this problem. The objective of the present paper is to investigate basic regularities of Np, Am, and Cm transmutation in the molten salt burner reactor and to determine optimal operational conditions of the reactor. In the optimal equilibrium mode just fluorides of these elements are added into the fuel composition but the fission products are extracted from it. This mode are reached by maintaining a specified actinide concentration. In order to provide the reactor criticality with a lower than optimum actinide concentration, it is required to add plutonium in the feed fuel instead of a certain amount of minor actinides, thus impairing the transmutation efficiency. To keep the critical state in the equilibrium mode, in situations where the actinide concentration is higher than the optimum one, it is essential to extract a portion of fuel with high content of 238Pu. It has been shown that the reactor configuration is the primary factor responsible for optimum actinide concentration value and that this value varies little with the feed fuel composition, the type of salt dissolving agent, and the fuel reprocessing mode. Optimal concentration for the molten-salt burner reactor with the core volume ranging from 2 to 30 m3 is within 17…10 mole %.
HEAT PIPES IN NUCLEAR ENGINEERING
Authors: Vereshchagina T , Loginov N , Sorokin A
Published: January 2022

The paper provides an overview of technical solutions for using of heat pipes in nuclear power plants both developed and operating. The review based on the scientific, technical and patent literature shows wide application heat pipes as heat transfer devices. Using of them for small and super-small power plants seems to be especially effective, because of high specific cost of plants with circulating coolants. A heat pipe is a device transferrind the heat by means of evaporation and condensation of a coolant circulating automatically under the action of capillar or gravitation forces. Heat pipes are used rather widely, both abroad and in Russia. The first application of a heat pipe principle in nuclear power plants was published in 1957, even before the emergence of the term "heat pipe". Now, there are about 300 patents in the world related to heat pipes application in nuclear power plants. Theare are seweral thouthands articles on the development of nuclear reactors with heat pipes have been published in the scientific and technical literature. One should expect that fifth-generation nuclear reactors cooled by heat pipes without any mechanisms and machines for the circulation of the coolant, as well as without the consumption of mechanical and electrical energy, will be appeared in this decade.
SIMULATION OF THE THERMOMECHANICAL BEHAVIOR OF (U, Gd)O2 FUEL ELEMENTS IN THE RTOP-CA CODE
Authors: Likhanskii V , Sorokin A , Zborovskii V , Ulibyshev K , Strozhuk A , Chernov I
Published: January 2022

The paper presents a description of the RTOP-CA code models designed for simulating the behavior of (U, Gd)O2 fuel rods, presents the examples of program verification using experimental data obtained at the research reactor in Halden and in the post-irradiation examinations of WWER fuel. The experience of operating WWER fuel has shown that differences in the thermomechanical behavior of (U, Gd)O2 fuel elements and UO2 fuel rods during the first - second fuel cycles can be observed. In particular, an accelerated increase in the length of the (U, Gd)O2 fuel rod cladding as compared to UO2 fuel elements is possible. To simulate the behavior of (U, Gd)O2 fuel rods under irradiation in the RTOP-CA code, models that describe the features of uranium-gadolinium fuel were developed. To calculate the evolution of heat release in (U, Gd)O2 fuel rods in the RTOP-CA code, a model of burnup of gadolinium isotopes was implemented. The developed model makes it possible, without using of detailed neutron-physical calculations, to simulate the burnup of gadolinium and the evolution of heat generation along the radius of the pellets. The models take into account the effect of the gadolinium oxide content on the thermophysical and mechanical properties of the fuel. The calculations performed using the RTOP-CA code are in good agreement with the results obtained during the experiments in Halden on the dynamics of the central fuel temperature, gas pressure in the (U, Gd)O2 fuel rod, and the elongation of the fuel column and cladding. To simulate the accelerated elongation of (U, Gd)O2 fuel rods in comparison with standard WWER fuel elements at the initial stages of fuel operation at nuclear power plants in the RTOP-CA code, a model of mechanical interaction of pellets and cladding was developed taking into account the misaligned arrangement of pellets.