bluuebunny/crossref_metadata_2025 huggingface dataset:
INFLUENCE OF OXYGEN IN SODIUM ON PERMEABILITY OF SODIUM IN STEEL
Authors: Orlova E , Orlov A , Solomatin A , Zasorin I
Published: January 2022

And widely used on the nuclear power plants (NPP) sodium is the most studied. The vital issue limiting its application is powerful exothermic interaction with water. Temperature of a torch of burning of sodium at contact with water vapor reaches 1600 K. The problem of protection of steel tubes against destruction at emission of water vapor in sodium is solved by timely detection of a small leak today. Generally in contact with sodium threefold oxides are stable. Penetration of sodium into steel leads to damage of structural material in the presence of chemical reaction with formation of ferro-chromates of sodium and is defined by oxygen level in sodium of existence of these compounds. Formation of compounds of sodium with iron and chrome in steel at its contact with sodium hydroxide submits to dependence on time in degree 0.5 that testifies to the diffusive mechanism of interaction. Penetration of sodium into steel 10X18H9, 10X18H10T with formation of the Na (FexCr1-x)O2 compound on a surface there were also zones with the volume structure aligned (α) in deeper layers as a result of contact with NaOH during 6 h at 1000 K made about 250 microns. At the same time, NaCrO2 on the surface of chrome for lack of iron does not exceed 30 microns and has anticorrosive properties and also high electrical resistance (about 1 megohm). It is shown that the distilled sodium gets in steel 10X2M and 10X18H10T at 923 and 1023 K to for 500 h on depth less than 2 microns, i.e. it is occluded in blankets and has no significant effect on properties of steel.
THE RESULTS OF THE STUDY OF THE INITIAL STAGE OF OXIDATION OF AUSTENITIC STEEL 12CR18NI10T IN WATER VAPOR
Authors: Ivanov K , Niyazov S , Gurbich A
Published: January 2022

The paper presents the results of the study of the initial stage of oxidation of austenitic steel 12X18H10T in water vapor, which were obtained using a method of research of thin surface layers of steel, based on the analysis of nuclear reaction products in the interaction of oxygen with accelerated ion beams. Austenitic 18 % chromium steel 12X18H10T is one of the structural steels used in reactor installations with heavy liquid metal coolant (HLMC) in the region is relatively high (T<350 °C) temperatures. In test benches and experimental installations the temperature area of use of this steel is much wider. Studies of this steel, especially in the field of low temperatures or short time intervals of exposure of samples, where there are significant methodological difficulties, are of some scientific interest. In addition, the study of the oxidation of structural steels directly in gaseous media, including water vapor, are of great practical importance. It is because these modes of oxidation are used as special technological operations of formation of oxide films on the individual elements and the equipment of the first circuits of nuclear power plant (NPP) with HLMC on the basis of lead and its alloys.
EXPERIMENTAL MODELING OF HYDRODYNAMICS AND HEAT TRANSFER IN REACTORS WITH HEAVY LIQUID METAL COOLANTS
Authors: Sorokin A , Kuzina Ju , Orlov A
Published: January 2022

The similarity analysis and the following criteria as well asymptotic solutions must be fully applied at planning and carry out the experimental study and generalizing its results. As academician S.S. Kutateladze notes "The external simplicity of the foundations of this analysis and the everincreasing multi-parametric nature of the problems for physical and mathematical modeling lead to many misunderstandings and direct mistakes”. Liquid metals form a special class of coolants characterized by a significant volumetric heat capacity and high thermal conductivity, which have the coefficient of kinematic viscosity much lower than the coefficient of thermal diffusivity, its Prandtl number is much less than one. The analysis results of the application of the similarity theory of the thermophysical processes to the modeling of hydrodynamics and heat transfer in liquid metals are presented in this report for: channels of complex shape, rod systems (reactor core and heat exchangers), temperature fields and coolant velocity in a hot chamber of a fast reactor in different operating modes. Practically the direct modeling can be applied unlimitedly only for processes in which determined similarity numbers are functions of only geometric simplices of the system and one defining criterion. For example the presence of two defining criteria such as the numbers Re and Pr during heat exchange makes modeling much more difficult. With three defining criteria the direct modeling is usually not feasible. In such cases it is necessary to caring out the systematic multivariate experiments. The purpose of such modeling experiments is the real detection of effects solved by a very general mathematical model but not impossible to carry out at the modern level of mathematical technologies, either analytically or in numerical studies.
ON THE ISSUE OF ACCUMULATION AND REACTOR UTILIZATION OF AMERICIUM IN THE NUCLEAR POWER
Authors: Dekusar V , Zrodnikov A , Eliseev V , Moseev A
Published: January 2022

Solving the postponed problems of spent nuclear fuel from thermal reactors, that is the main negative component of modern nuclear power in Russia, can be practically put into effect by the closure of the nuclear fuel cycle using fast reactors. Based on the considered nuclear energy systems, taking into account their structural organization and development scenarios, the work presents quantitative estimates and analysis of americium accumulation in the Russian nuclear power industry in the near future and describes the approaches to its utilization. Two alternative scenarios for the development of nuclear energy are considered. The numerical results of the comparison, analysis of the americium accumulation with a perspective up to 2050 are presented. The rate of formation of the Am-241 isotope in the separated plutonium and stored spent fuel of various types of nuclear reactors received a proper attention. Estimates show that the suppressive contribution to the accumulation of americium is due to the Pu-241 decay, as a result of which the accumulated americium is largely consists of isotope Am-241. Research results showed that without taking special measures to utilize americium in fast reactors, accumulation of americium in spent nuclear fuel of thermal reactors will reach a significant amount by the middle of this century, while the growth rate of its reserves will increase significantly. In addition, a significant amount of americium is likely to be contained in the spent nuclear fuel of foreign reactors of Russian design, which may be imported into Russia.
MODELING OF DETAILED NUCLEAR DATA AT ARBITRARY TEMPERATURET IN NEUTRON CALCULATIONS FOR MONTE-CARLO CODE
Authors: Koscheev V , Peregudov A , Rozikhin E , Semenov M , Yakunin A
Published: January 2022

In the practice of calculating the neutron characteristics of reactor applications, the relevant employees deal with computational models of various details. At the same time, a complex situation often arises before them, connected with the need to "calculate" models of RU with a large number of physical zones (~ 50,000), many of which have different temperatures. This complexity is due to the fact that the stage of preparation of nuclear data libraries is separated from the stage of their use in practical codes. For example, the code MMKC (or MCNP5) uses nuclear data libraries prepared for several "reference" temperatures. The stage of preparing nuclear data for a new set of temperatures can be simplified. Prepare a new library of nuclear data using the makxsf code, if there are ready-made sets of nuclear data for "reference" temperatures. However, and this simplified version of the preparation of a nuclear data library for an arbitrary temperature requires considerable time. In article a simple method of modeling nuclear data for an arbitrary temperature is proposed using proportional-temperature concentrations on the basis of data prepared once at "reference" temperatures. Proportional-temperature concentrations depend only on the concentration of the isotope, its temperature and the temperatures of neighboring "reference" values. When the temperature conditions in the calculation problem change, they can be easily recalculated. The use of the method of proportional-temperature concentrations is demonstrated on test models. These models are typical for both fast and thermal systems.
EQUILIBRIUM POTENTIALS OF LEAD OXIDE (II) IN AN EQUIMOLAR MIXTURE OF KCl-PbCl2
Authors: Pershin P , Batuhtin V , Tkacheva O , Arkhipov P , Zaikov Yu
Published: January 2022

The electromotive forces (EMF) of the electrochemical system (GC)Pb|(1-N)·KCl-PbCl2+N·PbO|ZrO2(Y2O3)|O2(Pt) in the concentration range of PbO 0.16-7.32 % mol. were measured at temperatures 776, 821 and 874 K. From the experimental values of EMF activities, coefficients of the activity and basic thermodynamic functions of lead oxide for dilute solutions were calculated. It is shown that PbO activities in the KCl-PbCl2 melt have moderate negative deviations from Raoul's law for ideal solutions. The activity coefficients of PbO at each temperature within the calculation errors are constant, that is, the diluted solutions of PbO in KCl-PbCl2 obey Henry's law. In the studied area of dilute solutions of PbO solvent activity coefficients are constant within the error, and their value is close to 1, as a consequence of the solvent activity within the calculation error will be equal to its molar fraction. This confirms the rule of Kubashevskii-Alcock about the obedience behavior of the solvent KCl-PbCl2 to Raoult's law for ideal solutions. Partial thermodynamic functions of the solvent - melt KCl-PbCl2 and integral thermodynamic functions of the KCl-PbCl2-PbO system are calculated using the standard equations. The interaction of lead oxide with chloride melts is caused by the formation of oxychloride compounds. According to the literature, there is a compound {[Pb2OCl]++Cl-} in the ion melt. The results of x-ray phase analysis show the presence of Pb2OCl2 compound in the frozen electrolyte melt. Using the device LECO ONH836 the concentration of oxygen in the metal lead was determined. It is shown that the oxygen concentration was 0.031 wt. % after the long-term exposure, which also exceeds the values in the Pb-O diagram almost twice.
FISSION MATRIX IN TDMCC CODE FOR LOOSELY COUPLED SYSTEMS CALCULATION
Authors: Semenova T , Mitenkova E , Solovyova E
Published: January 2022

In loosely coupled systems the criticality calculations using the traditional Monte Carlo power generation methods can lead to incorrect effective multiplication factor and neutron distribution functions, which is largely due to the steady-state neutron source in system. In reactor systems, the power generation method ensures the correct neutron source formation with a proper eigenfunction for any initial neutron distributions. A traditional power generation method does not guarantee the adequate source formation in loosely coupled systems even under very large calculation statistics. To count the neutron distribution functions in loosely coupled systems in TDMCC code is implemented algorithm based on fission matrix. Its capabilities are demonstrated to count the fission rate distributions in loosely coupled system such a slab with extended inhomogeneous fuel zones.
COMPARISON BETWEEN COMPUTER SIMULATIONS AND EXPERIMENTAL DATA FOR CARBON ION RANGES IN HOMOGENEOUS PHANTOMS USING THE RTS&T PARTICLE AND ION TRANSPORT CODE SYSTEM
Authors: Pryanichnikov A , Simakov A , Belikhin M , Degtyarev I , Novoskoltsev F , Altukhova E , Altukhov Yu , Sinyukov R
Published: January 2022

Lately, beams of heavy charged particles, e.g., protons and carbon ions, have found wide application in radiation therapy of oncological diseases owing to the fundamental possibility of qualitative im-provement of the spatial dose distributions when compared to sources of electrons and γ-rays conven-tionally used in radiation therapy, which makes it possible to radically decrease the radiation absorbed dose of the undamaged regions of the tissue adjacent to a tumor. In this paper a results of computer simulation are compared with experimental data for carbon ion carbon ion ranges in homogeneous phantoms using the RTS&T, FLUKA and MCNPX Monte Carlo multi-particle particle and ion transport code systems. Calculations of the main microdosimetric characteristics for cellular structures placed in homogeneous water phantoms are shown: average dose, linear energy transfer (LET), relative biological efficacy (RBE) and biological dose based on the Microbiological Kinetic Model (MKM) within software complex RTS&T. Calculations were made for the beam of 12C6+ ions with the energy of 454 MeV/u. Experimental data were obtained at the Temporary Radiobiological Stand of the U-70 accelerator complex at the Institute of High Energy Physics NRC "Kurchatov Institute", Protvino.
USING THE TECHNIQUE OF TEMPERATURE-PROPORTIONAL CONCENTRATION FOR CALCULATING MODELS OF FAST REACTOR
Authors: Koscheev V , Peregudov A , Rozikhin E , Semenov M , Yakunin A
Published: January 2022

In the practice of calculating the neutron-physical characteristics of fast reactors, codes are used that utilize various calculation methods, both engineering (implementing a group approach) and precision (using detailed modeling). The prepared library of constants is verified and then recommended for use in computational studies. A change in the temperature mode for a computational model often leads to the need to modify the nuclear data library, which must then be verified under new temperature conditions. In this work, we used the technique of temperature-proportional concentrations, with the help of which it is possible to avoid additional efforts to create nuclear data at new values of working temperatures. Four test models of large fast reactors are considered, which are representative enough to justify the reliability of the findings of the work performed. Calculations of the criticality of test models were performed with the MMKC code. Based on the ROSFOND-2010 library, nuclear data for calculations were prepared in two different ways. The first was using the NJOY code to directly prepare nuclear data in the ACE format for operating temperatures. The second method included the preparation of CE nuclear data in the ACE format at 6 “reference” values of temperatures, and further, for operating temperatures, the nuclear data were simulated using a temperature-proportional concentration technique. The results of the criticality and the Doppler Effect calculations are given. Differences of calculated values are analyzed depending on the method of nuclear data preparation.
BNcode - IMPROVED CODE FOR SCIENTIFIC SUPPORT OF EXISTING BN REACTORS
Authors: Peregudov A , Kryachko M , Koscheev V , Maslov P , Tormyshev I , Semenov M , Cunzio G , Gurskaya O , Ivanov A , Erpalov P
Published: January 2022

The relevance of this work is due to the need of the supervisor (IPPE) to perform work on the control and maintenance of the operation of fast sodium reactors. The purpose of this work is to develop an integrated code of a new generation for the scientific support of existing reactors BN-600 and BN-800 and in the future BN-1200. The code being developed for scientific support of the fast reactor operation serves for current and predictive calculations of reactor characteristics, safety justification during overload, justification of the micro-campaign reactivity balance, compliance with established performance criteria for fuel elements and other elements. Also, this code will be used as an independent calculation tool of increased accuracy for confirming and, if necessary, correcting the results of the operational support received at the station, as well as for analyzing non-standard settlement situations of increased complexity, such as the justification for entering BN-600 and BN-800 irradiating devices for producing isotopes (in particular, 60Co), irradiation of fuel assemblies with a nitride type of fuel.
SIMULATION OF NON-STATIONARY EXPERIMENTS AT THE ASTRA CRITICAL FACILITY USING THE SHIPR SOFTWARE-BASED SYSTEM
Authors: Volkov Yu , Kruglikov A , Zizin M , Boyarinov V , Nevinitsa V , Fomichenko P , Bobrov A
Published: January 2022

The paper presents the comparison of stationary and nonstationary experiments carried out at the ASTRA critical facility and calculation by SHIPR software-based system. The experiments to define of radial speeds distributions of 235U fission reactions rates were considered and simulated as stationary experiments. The results obtained in the course of calculations of these configurations can fully illustrate the effect of the presence of introduced control rod bodies and absorbing elements in the internal reflector on the accuracy of calculating the energy distribution in the core of the ring type. The revealed trends in the behavior of the distribution of 235U fission reaction rates when introducing control rods into the internal reflector emphasize the importance of a detailed analysis of the arrangement of control rods in the plane of the core and the algorithmization of their movement during the campaign. Experiments on the determination of kinetic parameters by the Simons-King pulse method were considered as nonstationary experiments. The results show good agreements with experiment in the calculations of both stationary and nonstationary experiments.
THERMAL PHYSICS OF ALKALINE LIQUID METAL COOLANTS. PART 2: PHYSICAL CHEMISTRY, TECHNOLOGY AND INNOVATIVE APPLICATIONS (RETROSPECTIVE-PERSPECTIVE VIEW)
Authors: Kuzina Ju , Sorokin A
Published: January 2022

The article presents the results of experimental and analytical studies carried out at the SSC RF - IPPE in the field of physical chemistry, mass transfer, technology and innovative applications of alkaline liquid metal coolants. An integrated approach to the study of coolants as a complex multi component heterogeneous system, the state of which is determined by the interaction of coolant - impurities - structural (technological) materials - protective gas, was substantiated. Thermodynamic analysis and experimental studies of the coolants properties and the kinetics of their interaction with air, water, hydrogen, structural and technological materials determined the sources of impurities and their intensity. It is shown that for temperature parameters of modern NPPs, the required concentration of impurities in sodium and sodium-potassium alloy, can be provided by purification the coolants by cold traps. The deeper purification of ciilants required for use in high-temperature installations of niobium, tungsten, vanadium, tantalum is achieved by getters (hot traps). This method can also be applied when limiting the weight and size characteristics, for example, for a space NPP. A fundamentally new combined purification system from impurities was proposed and substantiated for a high-temperature NPP for the production of hydrogen with a sodium coolant temperature of ~900 °C. As a result of the development of alkaline liquid metal coolants, liquid metal technology has been proposed in various technical applications: sodium at nuclear power plants with fast neutron reactors, sodium and sodium potassium in the metallurgical and chemical industries; sodium-potassium, cesium, lithium - in space installations; lithium - in fusion reactors - thermonuclear reactors.
USE OF DIRECT INTEGRATED TESTS WITH A STANDARDIZED SPECTRUM FOR TESTING THE CAPTURE CROSS SECTION OF FISSION PRODACT
Authors: Koscheev V , Rozikhin E , Yakunin A
Published: January 2022

The paper presents the results of validation for the capture cross section of the main fission product (FP) nuclides from different neutron data libraries on the results of measurements in integral experiments. A list of the main fission products for various types of fast reactors with various burnups of fuel is defined. When fissioning fuel nuclides in a nuclear reactor, numerous fission products are formed. The most important characteristic of fission products is the neutron capture cross section, as a result of which the valuable neutron for maintaining the fission chain reaction disappears. During long campaigns, the accumulated fission products can lead to a decrease in the criticality of the reactor to ~ 7 %. The evaluated neutron data for the main set of FP nuclides are contained in national libraries of evaluated neutron data. Evaluation of neutron data, as a rule, is carried out on the basis of experimental data from the international EXFOR database using various computational models of nuclear interactions. Often, the evaluated data is very different from each other. Evaluated data can be tested using integral experiments. To test the capture cross section for FP nuclides, direct integral experiments at the CFRMF reactor and experiments in which the neutron spectrum is close to the Maxwell spectrum with a spectrum temperature of kT = 30 keV were used. The first of the above experiments is a dosimetric benchmark experiment, given in the ENDF-202 benchmark experiment handbook. The second type of experiment underlies the recommendations for the MACS capture cross section for nuclides from the KADONiS astrophysical database.
DOMINANCE RATIO IN BN-TYPE REACTOR CORES
Authors: Gulevich A , Dolgikh V , Eliseev V , Peregudova O , Rozhikhin E , Semenov M , Stogov V , Tormyshev I
Published: January 2022

Maximum core height of BN-type reactors is limited by safety considerations. If the core height exceeds 85-90 cm, ULOF-type accident leads to the heavy core damage. Hence, increase of reactor power leads to increase of core diameter and D/H ratio. Neutron field stability in the large diameter cores should be considered. Ratio of the two first eigenvalues of static reactor balance equation (dominance ratio) widely used as neutron field stability criterion. The results of dominance ratio evaluation for large BN-type reactor are considered in the article. Dominance ratio of BN-type reactor compared with dominance ratios for several high power thermal reactors. It is shown that dominance ratio of BN-type high power reactor is still lower, than that of thermal reactors. Hence, instability of neutron field in BN-type reactors due to high dominance ratio seems to be unlikely.
NEUTRON-DOSIMETRIC SUPPORT OF CONSTRUCTION MATERIALS PROPERTIES RESEARCH EXPERIMENTS IN REACTOR CONDITIONS
Authors: Enin S , Belozerova A , Pavlov V , Chernysheva T
Published: January 2022

JSC "SSC RIAR" has accumulated a rich experience in the neutron-dosimetric support and neutron fields spectrometry using the activation method. Through the efforts of a specialists group, this direction continues to evolve, with advanced developments and modern high-precision equipment. The development of unified standards, computational and experimental methods of neutron-dosimetric support/monitoring to establish a one-to-one accordance between the observed changes in the properties of reactor materials and the conditions of reactor irradiation is topical. The application and distribution of neutron dosimetry using the activation method covers all the research reactors of the institute with all their various experimental conditions. The following examples of the practical neutron activation detectors usage in the following reactor materials science experiments are provided in this article.
MULTIDIMENSIONAL DATA ANALYSIS BASED ON THE RESULTS OF NUCLEAR TRANSMUTATION CALCULATIONS IN ZIRCONIUM ALLOYS
Authors: Belozerova A , Belozerov S , Shamardin V
Published: January 2022

The effects of nuclear transmutation in zirconium alloys are considered. The calculated scheme of transmutation of elements in alloys of the system [Zr + x% Nb + y% Mo] is determined. Numerical estimates of the molybdenum content in zirconium alloys of various compositions were carried out during irradiation in research (SM-3, BOR-60) and power (VVER-1000, PWR) reactors. The dynamics of the accumulation of molybdenum during irradiation in the SM reactor and the subsequent exposure of the samples for one year in standard zirconium alloys Zr + 1% Nb, Zr + 2.5% Nb and for the proposed initial compositions Zr + 1% Nb + 0.5% Mo were estimated Zr + 2.5% Nb + 0.5% Mo. Reactor irradiation leads to a significant accumulation of molybdenum in the Zr + 2.5% Nb alloy as a result of nuclear reactions: the molybdenum concentration increases to about 0.7 % with an increase in neutron fluence (E> 0.1 MeV) to 1×1023 cm-2 at irradiation in a research reactor SM. The dynamics of the effects of nuclear transmutation of zirconium upon irradiation in the active zone of research (SM-3, BOR-60) and power (VVER-1000, PWR) reactors is estimated. The Zr element transmutes no more than 10 % in 100 years of exposure. The accumulation of molybdenum over the same time can be up to 104 ppm (appm). The results of the intellectual analysis of the calculations performed with multidimensional analysis based on OLAP of the multidimensional data space of the database provide knowledge in the form of classifications and confirm the need to develop zirconium alloys with additional doping with molybdenum, which can provide optimal physicomechanical properties, in particular, radiation growth characteristics, as well as acceptable level of nuclear-physical and corrosive properties.
FRIGATE CODE TO THE COMPLEX SCENARIZED CORE AND SHIELDING CALCULATIONS IN GROUPWISED APPROXIMATION: BACKGROUND OF CREATION, ORGANIZATION STRUCTURE, FUNCTIONALITY
Authors: Nikolaev A
Published: January 2022

The short information about destination and abilities of program FRIGATE, structure description and program characteristics are presented. The FRIGATE program implements the methodology developed by the authors of a free programmable organization of complex interconnected multicomponent calculations to justify the design characteristics of reactors and protect reactor installations in a group approximation. The FRIGATE program within the framework of a single executable file allows the user to organize arbitrary chains of standard calculations to solve the main tasks of the design support of reactor installation design. The structure of the FRIGATE program is organized according to the principle of a set of interacting process units. Each program block performs a separate logically closed task. The program includes the following blocks: a geometric block, an SN block (based on the PMSNSYS-II program), a block group block preparation block, point kinetics parameter calculation block, a perturbation theory calculator of reactivity effects, a unit for working with ionizing radiation field functionals, isotope kinetics calculation unit, as well as other computational and / or service subroutine blocks. The program has no restrictions on the number of nuclides, the number of physical zones, cells, the number of nuclides in the zone, and the like. Parallelization of lengthy calculations in the current version of the program is organized using the OpenMP 3.0 standard. A set of software documents has been released for the program.
RESEARCHES OF THE POSSIBILITY TO BURN OUT MINOR ACTINIDES IN A FAST REACTOR WITH METAL FUEL ON THE BASIS OF ONLY MINOR ACTINIDES
Authors: Korobeinikov V , Kolesov V , Terekhova A , Karazhelevskaya Yu
Published: January 2022

Minor actinides (MA) - primarily, isotopes of neptunium, americium, and curium - attract particular attention due to their long-term radiotoxicity. Measures to reduce the increase in the number of minor actinides have not yet been taken their total number in the world is growing, and will reached more than 200 tons by 2020. Note that minor actinides, as well as nuclides formed as a result of neutron capture, have the ability to share, so it is necessary to investigate whether the nuclear reactor can work if only they are used as fuel? The paper examined the possibility of using fuels from MA alone in reactors without uranium or plutonium. From the results of comparing the fission and capture cross sections, it follows that fuel in the form of Am or Np-237 can only use a fast neutron reactor, since in the thermal and intermediate spectra the capture cross section significantly exceeds the fission cross section. The results of calculations of the active zones of a model fast reactor with fuel from one americium or neptunium-237 demonstrated the high speed of their transmutation and burning out.
MODELING OF HEAT AND MASS TRANSFER PROCESSES IN FUEL ASSEMBLIES OF FAST REACTORS AS PART OF THE CHANNEL-BY-CHANNEL CALCULATION METHOD. GENERALIZED EXCHANGE CHARACTERISTICS FOR SINGLE-PHASE FLOWS OF LIQUID METALS
Authors: Sorokin A , Kuzina Yu , Sorokin G , Denisova N
Published: January 2022

The results of experimental and computational-theoretical studies of the transverse interchannel molecular-turbulent and convective exchange of mass, momentum and energy are presented in the framework of the channel-by-channel model of thermohydraulic calculation of fuel assemblies of fuel elements of the fast reactors core with liquid metal coolants for nominal, non-nominal and accident conditions. A complete system of inter-channel exchange coefficients is obtained for closing the system of macro-transport equations in a wide range of parameters, taking into account the fuel assemblies deformation. It was shown that transverse convective exchange in fuel assemblies with spacing by wire wrapped is caused by a deficit of static pressure in the region behind winding at the surface of the rod, and the high intensity of molecular-turbulent exchange in tight bundles is explained by the appearance of secondary flows in cells. The heat exchange between cells due to the thermal conductivity of the fuel rods is determined by the thermal modeling parameter of the fuel rods for the first harmonic. The effect of wire wrapped on the flow can be modeled by periodically changing hydraulic resistance to the transverse coolant flow. The integral model of convective inter-channel exchange in fuel assemblies with distance wire wrap is substantiated. Taking into account the centrifugal effect and calculating the interchannel exchange coefficients taking into account the local distribution of parameters allows us to clarify the thermohydraulic characteristics in the shaped bundles. For the characteristic parameters of fuel assemblies of fast reactors (Pe ≥ 50; S/d ≥ 1,1; h/d ≤ 30; ε1 < 1) the predominant is convective inter-channel exchange due to the spacing of fuel rods by wire wrapped. For highly heat-conducting fuel compositions, a significant contribution to the inter-channel heat exchange can be made by the exchange due to the thermal conductivity of the fuel elements. At a low flow rate (Pe <50), the contribution of molecular-turbulent exchange and heat exchange increases due to the thermal conductivity of the fuel rods.
THE TWO-DIMENSIONAL PROBLEM OF NON-STATIONARY THERMAL CONDUCTIVITY IN A MULTILAYER PLATE AND CYLINDER
Authors: Levchenko V , Kascheev M , Dorokhovich S , Zaytsev A
Published: January 2022

The problem of determining a two-dimensional non-stationary temperature field in a k-layer cylinder and plate of length l is solved. There is a symmetrically located gap (plate) or cylindrical cavity (cylinder) in the center of these bodies. The absence of a gap or 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 bodies 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. With the help of the geometric parameter G in the mathematical formulation of the problem, one differential equation for both multilayer bodies is written. 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 and transverse coordinates 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.
EXPERIMENTAL AND COMPUTATIONAL STUDIES OF THE BOILING PROCESS OF LIQUID METAL DURING THE DEVELOPMENT OF AN ACCIDENT IN A FAST REACTOR: HEAT TRANSFER AND CIRCULATION STABILITY
Authors: Sorokin A , Ivanov Eu , Kuzina Ju , Denisova N , Nizovtsev A , Privezentsev V , Sorokin G
Published: January 2022

Studies of the boiling of alkali metals show that, compared with boiling water, the process of boiling liquid metals has significant features. Cooling of a fuel assembly in accident conditions when accident protection is triggered and circulation pumps are switched off (ULOF) leads to the study of fuel elements cooling at reduced coolant flow rates or even tipping circulation in fuel assemblies. There is only limited data on the boiling of sodium in fuel assemblies in these regimes. The results of a series of experiments on heat transfer and stability of circulation during boiling of a sodiumpotassium alloy on single fuel assembly models and in a parallel fuel assembly system with natural coolant circulation performed at JSC “SSC RF - IPPE” are presented. The results of comparing the data of the calculated and experimental studies are presented. The influence of the surface roughness of the fuel rods on the heat transfer and flow regimes during boiling of a liquid metal in bundles is demonstrated. The results of experimental studies of heat transfer during boiling of sodium in natural and forced convection regimes in a fuel assembly model with a “sodium cavity” located above the reactor core intended to compensate for the positive sodium void reactivity effect in emergency situations with sodium boiling are also presented. It is shown that it is possible to provide continuous sodium cooling of fuel element simulators in fuel assemblies under these conditions. The results of the generalization of data on heat transfer during boiling of liquid metals in bundles and a cartogram of the regimes of two-phase flow during boiling of liquid metals in bundles are presented. The objectives of further research are discussed.
RESULTS OF CALCULATION OF BENCHMARK MODELS ON RADIATION PROTECTION, INTENDED FOR RESEARCH OF TRANSMISSION OF IONIZING RADIATION THROUGH WATER AND STEEL LATTICES
Authors: Grabezhnoy V , Lomakov G
Published: January 2022

The work is devoted to evaluating the accuracy of using the BNAB-RF system of constants together with the discrete ordinate method and the KATRIN-2.5 and CASCAD codes (IPM RAS) for studying the passage of neutrons and photons through water and steel gratings. The results of calculating the benchmark models for radiation protection from the SINBAD database are presented, namely, the ASPIS Neutron / Gamma-Ray Transport Through Waiter / Steel Arrays, NESDIP-2 Benchmark Experiment and NESDIP-3 Benchmark Experiment models, which were located in experimental channel of the NESTOR reactor. As a neutron source, a disk made of an alloy of uranium and aluminum irradiated with a stream of thermal neutrons leaving the NESTOR reactor was used. Protective compositions consisted of steel containers with water and plates of various steels (stainless steel and mild steel). Measurements of the passage of ionizing radiation were carried out using foils placed between steel plates or immersed in water at different distances from the neutron source. The calculations of neutron and photon reactions were carried out using the discrete ordinate method in 3D and 2D geometries KATRIN-2.5 and CASCADE, respectively. The constant systems BNAB-93 and BNAB-RF were used as group libraries. We can especially note the calculated estimates for the passage of gamma radiation presented here, obtained from the reaction (n-gamma) in the materials of the ASPIS installation. The results of calculations showed a good agreement of the calculated values with the experimental data. Based on the obtained numerical results, the BNAB-RF constant system was verified as applied to the calculation of radiation protection from ionizing radiation passing through water and steel gratings.
HYDRODYNAMICS OF TYPICAL DISTRIBUTION HEADER SYSTEMS OF NUCLEAR POWER PLANTS: MODERN VIEWS AND RESEARCH PERSPECTIVES
Authors: Delnov V
Published: January 2022

The paper considers the issues of hydrodynamics of typical distribution header systems (DHS) in heat exchangers and reactors of nuclear power plants (NPP). The analysis of hydrodynamic properties of coolant flow in DHS has been carried out. The paper presents typical flow patterns of a coolant in the flow paths of the specified DHS. The classification of different types of DHS is presented. The coolant flow is shown to have a jet nature in a relatively confined and unconfined DHS. Consideration is given to several types of jets and processes of transformation of some types of jets into other types. The paper presents a brief description and formula of the fluid distribution regularity (registered as a scientific discovery) at the outlet from the flat and cylindrical DHS with central and side fluid supply. Consideration is given to the similarity property of hydrodynamics in the flat and cylindrical RHS with different points of fluid supply to the header. Advanced areas of research into revealing hydrodynamic effects in DHS and obtaining additional information on the influence of various factors on the fluid flow in them are proposed.
ESTIMATION OF THE LOW-OXYGEN BOUNDARY OF THE EXISTENCE OF PLUMBOFERRITE PHASES IN THE LEAD MELT
Authors: Osipov A , Askhadullin R , Lavrova O , Ivanov K , Niyazov S , Cheporov R
Published: January 2022

At present, installations with heavy liquid metal coolants (HLMC) based on lead melts are considered as promising nuclear power plants, since these coolants have a number of advantages over alkali metals and other coolants. A feature of heavy liquid metals is their rather high corrosivity towards structural materials; however, at present this problem has been partially solved due to the formation of protective oxide coatings on the surfaces of steels based on oxides of metal components of structural steels. This determines the main tasks of modern HLMC technology, which are to provide conditions for the formation and maintenance of oxide layers with optimal protective properties on the surfaces of structural steels. The success of solving these problems in the general case depends on an adequate representation and quantitative description of the properties of the system “liquid metal melt - oxide phase”. In this work, we consider the questions of the possibility of the formation of plumboferrite phases in liquid lead, their composition and thermodynamic properties.
SWELLING OF NEUTRON ABSORBER ON THE BASIS OF INDUSTRIAL AND RECONSTRUCTED QUALITIES B4C AFTER TWO-YEAR IRRADIATION IN THE EMERGENCY PROTECTION RODS
Authors: Kinev E , Ustinov A
Published: January 2022

The radiation swelling of hot-pressed B4C blocks of industrial and re-fabricated quality with the participation of absorber elements providing emergency protection, up to fast neutron fluence 7⋅1022 cm-2, was investigated. The reason of metal cladding machining deformation of neutron absorber during a two-year exposure period was swelling. The volumetric swelling and the porous parameters of the absorber are measured by hydrostatic weighing and digital analysis of electron microscopic fractograms of fresh splits. The pore allocation and destruction manner of B4C was studied by scanning electron microscopy. It was found that swelling for industrial and re-fabricated B4C, with maximum neutron fluence in the lower part of absorbing elements, culminates 12 and 18 % respectively. The drastic decrease in its swelling was caused by fluence reduction along of absorber rod. Volumetric changes are explained by growth of closed porosity. The maximum swelling of irradiated briquettes for industrial and re-manufactured B4C is 25 and 30 %, respectively, with the same initial values of about 19 and 23 %. The maximum value of closed porosity from industrial and re-manufactured B4C after irradiation is 14 and 21 %, respectively, against the initial values of 4 and 9 %. All detected pores settle at the grain border. The primary bulk of the pores is of technological origin. The size and concentration of pores correlate with an increase of neutron fluence and burnout during irradiation, as well as provoke to formation of microcracks between the crystals and grains of the boron carbide matrix. Nanoscale helium pores with a diameter of more than 100 nm were not detected in the grain body. The open porosity inside maximum swelling zone and mechanical interaction absorber with the cladding is reduced. The fracture nature of the swelling B4C changes from a mixture with a predominance of intragranular splitting to a purely intergranular type under increasing fluence.
SIMULATING OF DROP ENTRAINMENT IN THE JET-VORTEX CONDENSER OF THE VVER-440 CONFINEMENT SYSTEM
Authors: Kazantsev A , Supotnitskaya O , Sergeev V
Published: January 2022

The article presents the results of simulating the jet-vortex condenser operation. The jet-vortex condenser used as a part of the confinement system is designed to ensure the confinement integrity of an NPP with VVER-440 unit during LOCA accidents. To simulate the jet-vortex condenser operation it is important to keep a sufficient amount of water after a pressure peak was reached in the jet-vortex condenser hydraulic lock. The loss of water due to drop entrainment with steam-air mixture flow into the atmosphere stops when the velocity of drop sedimentation becomes higher than the velocity of drop entrainment with the mixture flow. The jet-vortex condenser model integrated into the KUPOL-M code was validated against the experimental data obtained on the VNIIAES test facility. To take into account drop entrainment with steam-air mixture flow the procedure of moisture separation and drop entrainment was used. A good agreement between calculated and experimental results was obtained when comparing the initial and final water levels in the hydraulic lock. The research results confirmed validity of the model of drop entrainment with steam-air mixture flow during the operation of the jet-vortex condenser and the preservation of water into the hydraulic lock during the accident.
CALCULATION ANALYSIS OF SIMULTANEOUSLY OCCURRING BEYOND DESIGN BASIS ACCIDENTS AT POWER GENERATING UNITS № 3 AND 4 OF THE BNPP
Authors: Anfimov A , Kirilov I , Kuznetsov D , Nikanorov O , Salyaev A
Published: January 2022

The paper presents the results of a computational analysis of simultaneously occurring beyond design basis accidents at power generating units No. 3 (BN-600) and No. 4 (BN-800) of the BNPP. Safety analysis for operating power generating units at the same time occurring beyond design basis accidents is carried out in accordance with the requirements of NP-001-15. Computational studies were performed with the SOCRAT-BN, ANSYS, ORIGEN2, and VIBROS 2.2 codes. There are considered beyond design basis accidents caused by seismic impact with intensity of 7 points on MSK-64 scale (exceeding the maximum allowed earthquake (MAE) for the site of power unit No. 3 and the corresponding MAE for the site of power unit No. 4). As a result of seismic impact at all the power generating units of the BNPP, a loss of system power supply occurs, which, together with additional failures of the main equipment, leads to beyond design basis accidents. The paper presents the results of computational studies of beyond design basis accidents in terms of reactor plant, assembly cooling pond and holding pond and radiation effects assessment. The results of a computation analysis of the beyond design basis accident showed that the release of radioactivity occurs only for the BN-600 reactor facility (where are breaking of the fuel elements cladding). At the same time, the predicted radiation doses for the first years after the accident are lower than the NRB-99/2009 criteria for making a decision on the evacuation and resettlement of inhabitants.
DYNAMICAL METHOD APPLICATION ANALYSIS FOR SAFETY ASSESSMENT CALCULATION OF VVER UNIT MANEUVERING REGIMES EXPERIMENT
Authors: Uvakin M , Nikolaev A , Makhin I , Sotskov E
Published: January 2022

Maneuvering operation test is actual problem of VVER reactors exploitation. Such tests are classified as nuclear hazardous procedures which are accompanied by power variations, regulators actions and continuous space power field fluctuations. As a result a lot of possible initial reactor plant conditions are occurred. This fact should be taken into account during experiment safety assessment. Current work presents calculation analysis results for high powered VVER unit safety assessment which was covered daily maneuvering experiment. All calculations are implemented by KORSAR/GP programming code with 3D-neutrion kinetic model. Quantitative and qualitative criteria show the successful solving of following problems: - Confirmation for normal operation limits adherence, especially for local power distribution parameters. - Reactor plant safety assessment for RIA accidents. Work contains analysis of developed method possibilities for VVER power unit calculation safety assessment during daily power maneuvering. Following criteria has been confirmed: - All conservative approach principles completely comply which has composed usual VVER safety assessment method for 3D neutron kinetic model. - All important features for reactor physics and dynamics during maneuvering have been taken into account. Maneuvering regimes calculations practical and methodical experience will be applied for the same tasks of VVER safety assessment.
EXPERIMENTS ON NEUTRONS TRANSMISSION THROUGH LITHIUM HYDRIDE LAYERS-7 IN URANIUM MULTIPLYING SYSTEM WITH NEUTRON ACTIVATION ANALYSIS
Authors: Vaivod A , Yudov A , Besov S , Andreev S
Published: January 2022

The paper provides the precision experimental results on fission spectrum neutrons transmission through lithium hydride layers-7 with neutron activation analysis. The experiments were performed in Zababakhin All-Russia Research Institute of Technical Physics” on a critical assembly stand FKBN 2. A cylindrical multiplying system (MS) made of high enriched uranium was used as a source of fission spectrum neutrons. Activation integrals of neutron activation detectors (NADs) located at different points of the composite lithium-7 hydride reflector face-mounted on the cylindrical uranium MS were determined in the experiments. NADs based on nickel, indium, titanium, aluminum and copper were used. NADs of different types were exposed to radiation at power ~25-30 W during 1-1,5 hours. Nickel NAD was used as a monitor. Activation integrals absolute measurement error was determined by a certified technique and ranged from ~4 to ~8 % (2σ). Relative measurement error was determined by statistical accuracy, instrument peaks software processing correctness and was found not to exceed 5,4 % (2σ). Using the data obtained, spectral indices (the ratio of the normalized values of the activation integrals of various detectors types to the one of nickel detector), which provide information on the neutron spectrum were determined.