bluuebunny/crossref_metadata_2025 huggingface dataset:
INVESTIGATION OF HEAT EXCHANGE IN FUEL ROD BUNDLES AT SUPERCRITICAL WATER PRESSURE
Authors: Sorokin A , Kirillov P , Kuzina Ju , Grabezhnaya V , Loschinin V
Published: January 2022

Calculations to substantiate the characteristics of a reactor cooled by water with supercritical steam parameters and an adjustable neutron spectrum require clarification of data on heat transfer characteristics. The features and data on heat transfer in round pipes and rod bundles at supercritical pressures (SCP) are considered. In analysis of hydrodynamics and heat transfer dates with the flow of water of supercritical parameters in round pipes, an area of both more intense and degraded convective heat transfer was detected compared to subcritical pressures. It is assumed that one of the reasons for the deteriorated heat exchange is the presence of a near-wall layer consisting of a "gas" phase with low thermal conductivity and a central flow region in the form of a liquid-like phase having a lower temperature. A brief review of data on heat transfer in round tubes and rod bundles with SCP is presented. Since experiments on water are complex and expensive, many results, including the SSC RF - IPPE, were obtained on model media (CO2 and freon-12). Shown that despite the diversity of recommendations on the boundary regime with the deterioration of heat transfer, the proposed ratios give a different dependence of the heat flux density on the mass velocity. They do not consider the influence of such parameters as pressure, channel geometry, temperature of heat exchange surface, etc., they are obtained for a narrow range of parameters. And only extensive analytical studies of Yu.A. Zeigarnik et al. (JIHT RAS, NRU "MEI") allowed to summarize the data on heat transfer in round pipes at SCP. Only limited data was obtained by Russian, Ukrainian and Chinese specialists in heat transfer in bundles with a small number of pipes. The results of Chinese experts have shown that heat transfer in rod bundles at supercritical water pressures is higher and more stable than when water moves in pipes and annular channels. In close bundles, the deterioration of heat transfer occurred at low mass velocities and high heat fluxes, and in widespread deterioration of heat transfer was not observed. The growth of heat transfer in the rod bundles is promoted by mixing with spacer and mixing grids, which destroy the wall barrier layer, which prevents the transfer of heat under supercritical pressure from the wall to the center of the flow. The accumulation of new experimental data and additional analysis of research are necessary. A description and technical characteristics of the thermo-hydraulic facilities of the SCP on water (SVD-1 and SVD-2) and freon-12 (STF) available at SSC RF - IPPE is provided. The results of experiments at SSC RF - IPPE and technical approach and methodology of future experiments in hydrodynamics and heat transfer at SCP is presented and discussed.
THE SYNERGETIC OF FUEL ELEMENT
Authors: Orlova E
Published: January 2022

The fuel element is the complex system consisting of fuel, a sublayer, a cladding, protective coatings on a surface of a cladding and contacting to the cooler and material of a contour in general. The synergetic (self-coherence) of interaction of elements of this system has to be provided both concerning high heat conductivity of fuel element, and concerning corrosion compatibility. Use of a liquid metal sublayer (LMS) instead of gas, allows significantly (by hundreds of degrees) to reduce temperature in the center of fuel that increases safety at UTOP accidents (uncontrollable increase in power) at ULOF (loss of an expense of the cooler). Gap thickness when using LMS (unlike helium) can be significantly increased that will practically not affect thermal characteristics of fuel element, but will allow to distance considerably time of approach of direct contact of fuel with a cladding, having increased thereby depth of burning out of heavy atoms and having increased cost efficiency and competitiveness of fast reactors extension of the fuel elements resource. These principles when using heat-conducting nuclear fuel are especially effective (metal, nitride, carbide) and the lead heat carrier. Synergetic reasonable corrosion compatibility of a cladding of fuel element with LMS is confirmed with numerous settlement pilot studies by means of formation and self-curing of accidental damages of a protective coating of nitride of zirconium on the internal surface of steel in LMS of eutectic structure on the basis of lead with magnesium and zirconium. When using nitride fuel heat-conducting LMS with anticorrosive properties the resource of fuel element is limited not by swelling of fuel any more, and the dose damaging a cladding.
STUDY OF U-10%Zr METAL FUEL FOR POWER REACTOR FACILITIES
Authors: Kurina I , Riaby V , Frolova M , Plaxin O , Dvoryashin A
Published: January 2022

Currently, there is an increased interest in the development of uranium-zirconium metal fuel. Abroad, rods for fuel elements are manufactured by induction melting followed by injection casting. The casting of fuel rods is carried out in quartz test tubes with an inner coating of yttrium oxide. In the JSC “SSC RF - IPPE”, work was carried out on the manufacture and research of some basic properties of fuel slugs of the U-10%Zr alloy. Fuel slugs were made by melting of uranium and zirconium in an arc furnace, subsequent melting of ingots in quartz test tubes in a vacuum furnace and turning to the desired size on a lathe. Obtained slugs of U-10%Zr had a density of 16.35 g/cm3, lamellar structure consisting of the phases α-U and δ-UZr2. The results of studies on the interaction of fuel with materials of test tubes and crucibles have shown that yttrium oxide is the most suitable material for melting crucibles or coatings. It is also possible to use ZrO2-Y2O3, followed by grinding the fuel rods.
MODELLING OF NOISE EXPERIMENTS FOR VERIFICATION OF KINETIC PARAMETERS OBTAINED USING THE SAPFIR_RF CODE
Authors: Artemov V , Zinatullin R , Karpov A
Published: January 2022

New multi-group neutron data libraries and a new scheme of fuel burnup have been developed on the basis of the modern ROSFOND evaluated nuclear data files. The upgraded version has a working title - the SAPFIR_RF code. Noise experiments for defining the ratio of total delayed neutron yield to neutron lifetime, βeff/l, were performed to verify the computational scheme of kinetics parameters implemented in the SAPFIR_RF. These experiments have been performed using the homogeneous assemblies with various isotopic composition of uranium-plutonium fuel. In order to analyze and interpret the experimental data correctly, numerical investigations simulating the experimental conditions and features of measurement results processing were performed; these investigations highlighted the need for systematic correction to be made for measurement results. Taking into account the correction, the calculation results are qualitatively and quantitatively in agreement with measurement data within the measurement accuracy of ±8 %. This paper draws on the modern data of Russian neutron data library, ROSFOND, as well as on G.R. Keepin’s kinetic parameters and evaluated nuclear data of the foreign libraries, ENDF/B-VI and ENDF/B-VII. Kinetic parameters obtained by G.R. Keepin and ENDF/B-VI library give two extreme evaluations of βeff/l ratio; the difference between them averages ~4,5 %. It should be noted that the calculation results received using Keepin's parameters are closer to experimental data in the most cases.
ABOUT POWER ERROR ESTIMATION USING THE METHOD OF UNCERTAINTY AND SENSITIVITY ANALYSIS FOR VERIFICATION OF THE COUPLED NEUTRONIC AND THERMAL HYDRAULIC CODES
Authors: Artemov V , Artemova L , Mikheev P
Published: January 2022

This paper describes the capability of sensitivity and uncertainty analysis (SUA) for estimation of power calculation error during the computer code verification compared to the data received in experiments at operating power plants. Estimation of errors in code verification calculation against measurements at operating plants and assessment of conservatism in safety calculations are considered as two interrelated problems. A defining moment in this approach is that one or more test cases are selected; selected test cases are included in matrix of code verification assuming that these test cases involve the same physical phenomena as in simulated accident scenarios. The criterion of test case “representativeness” is similarity between correlation coefficients of model predictions in accident scenario and test case. The example demonstrates how the SUA approach is used in calculations with the VVER-1000 coupled neutronic and thermal-hydraulic simulation model based on the KORSAR/GP and SAPFIR_95&RC_VVER codes.
VERIFICATION OF KORSAR/GP PROGRAMMING CODE FOR CALCULATION WITH TIME CONSTANTS CHANGING FUNCTION FOR MODELING XENON TRANSIENT PROCESSES ON VVER REACTOR PLANT
Authors: Nikolaev A , Uvakin M
Published: January 2022

The verification of the method for acceleration the calculation of slow transition xenon processes based on the use of the similarity theorem were done in this paper. For this purpose experiments with spatial xenon oscillations were calculated by using this method. Two experiments were considered in this work: in the first experiment initiation of xenon oscillations was realized by injection of boron acid in primary circuit and following reactor plant power decreasing, in the second experiment initiation of xenon oscillations was realized by control rod group movement. The calculation results were compared with calculation results of analogical experiments without using calculation acceleration method and experimental data. Additionally calculations of daily maneuvering cycle with electrical power level changing on graph 100-70-100 % Nnom and using calculation acceleration method were performed. These calculations were done for AES-2006 project VVER reactor plant. The calculation results were compared with calculation results of analogical daily maneuvering cycle without using calculation acceleration method. Some conclusions on the possibility of using the calculation acceleration method for calculations of long-term xenon processes were made. The calculations were performed by using KORSAR/GP code.
DESIGN OF BIOLOGICAL SHIELD OF A TRANSPORT CONTAINER FOR IRRADIATED GRAPHITE MODERATOR OF NUCLEAR REACTOR OF RBMK-1000
Authors: Sidorov N , Ivshin A , Fedorovich E
Published: January 2022

In given paper is discussed a theme of the providing of the biological shield of a container which is intended for exportation from NPP spots the bloks of the irradiated graphite moderator of nuclear reactors RBMK-1000 at a stage of such reactors decommission. With usage of the computer program systems Microshield (MS) and MCC 3D has been performed a design of container’s biological shield. This container can be used as for graphite blocks transportation, as for their storage. Year 2018 was a final one for the first unit of Leyningradskay NPP with reactor RBMK-1000, which has been under operation from 1973. The stop of this unit laid the beginning of step by step process of decommissions of such type power reactors at our country territory. For power reactors, in which as moderator of neutrons is used graphite, at ending of their service time is necessary to remove it from NPP stop and to keep(store) it safely-before reprocessing on final disposal. For the transportation and for the safe storage of graphite blocks (GB) of irradiated graphite moderator (IGM) it is necessary to develop special container, which will provide proper level of a management with IGM at the spot of decommissioning unit, at a way to spot of IGM reprocessing or/and its final disposal. The structure of container should to satisfy to requirements of radiation protection, providing the necessary level of a biological shield and also should to have proper strength characteristics and ability to bear supposing dynamic loads without loss of integrity and without loss of hermetically. The practical meaning of given work consists in the possibility of usage of its results by designing and manufacturing of containers for transportation and storage of IGM. The usage of such containers is actual by decommission process of nuclear power units with reactor RBMK-1000, because such usage will allow to accelerate significantly this process and fulfill the requirements of the ecological safety. The usage in given work of the modern program systems of modeling of biological shield allows to evaluate numerically shielding possibility of the structure materials, which could be used for the biological shield of containers and allows to choose optimal option of loading of graphite blocks in container. In the work is presented information about positive properties of the high-strength cast iron with spherical graphite inclusion as a perspective structural material for container’s body, which able to provide high level of the ecological safety during the exploitation of containers for irradiated graphite blocks.
FUNDAMENTAL INVESTIGATONS IN THE FIELD OF NUCLEAR POWER FACILITIES THERMOPHYSICS: ACHIEVED RESULTS AND PROBLEMS FOR FURTHER INVESTIGATON
Authors: Kuzina Yu , Sorokin A
Published: January 2022

The results of experimental and theoretical calculations of fundamental research in the field of hydrodynamics, heat transfer, physical chemistry and nuclear power technology cooled by water and liquid metal coolants are analyzed, problems and tasks of further research are formulated. The effects and physical phenomena, laws and characteristics for a wide range of processes occurring in the first and second circuits of reactor facilities, including the flow part, the core, the heating equipment, the means of monitoring and cleaning coolants from impurities as applied to watercooled thermal neutron reactors, fast liquid metal reactors, fusion plants, etc. Particular attention in the field of hydrodynamics and heat transfer is given to studies of the heat transfer crisis in watercooled reactors, hydrodynamics of collector systems, stratification and mixing of jets, vibroacoustics, turbulent transfer in complex channels, heat transfer in the channels and assemblies of fast reactor fuel elements, contact thermal resistance, boiling in large volume and bundles of fuel rods, condensation of liquid metals. In the field of physical chemistry and the technology of liquid metal coolants, the types of impurities and the sources of their entry into the coolants for various nuclear power plants, the characteristics of the mass and heat transfer processes in the circuits and equipment of reactor facilities with liquid metal coolants (sodium, potassium, lithium, lead, bismuth lead), the efficiency of using various purification devices (cold and hot traps) and monitoring the state of impurities in coolants are analyzed.Problem questions and suggestions for further research are formulated. Information is given on the key thermophysical problems of basic research in relation to the development of innovative nuclear energy technologies: water-cooled reactors with supercritical water pressure, high-temperature fast reactors with sodium coolant for hydrogen production, thermonuclear installations, nuclear power plants with liquid-metal space coolants.
CALCULATION OF THERMAL NEUTRON FLUX IN RADIATION SHIELDING BY PROGRAM FRIGATE WITH USING A SUBMODELLING METHOD
Authors: Skobelev A , Nikolaev A
Published: January 2022

Using the example of calculating the thermal neutron flux density in an ionization chamber, where a methodically correct solution is required to be obtained only within a fragment of the computational model, and particle transport from the source to the studied region can be carried out in a more rough approximation, the use of DSN-DSN schemes discrete ordinate method on grids with a completely heterogeneous structure in combination with the submodeling method. The paper presents the results of calculating the thermal neutron flux density and the number of fission reactions in the sensitive part of an ionization chamber located in a concrete radiation protection shaft using the FRIGATE program (DSN method using the submodeling method) and cross-validation with the TDMCC program (Monte-Carlo method). The work continued the study, aimed at verifying the method of submodeling by spatial, angular, and energy variables, recommended for use in radiation protection calculations by deterministic codes. Cross-verification was performed using the FRIGATE DSN program and the TDMCC Monte Carlo code using the example of calculating the thermal neutron flux density and the number of fission reactions in the sensitive part of the fission ionization chamber. Verification results showed good agreement between the considered functionals (constant methodological discrepancy within 25 %). The obtained results confirm the effectiveness of using DSN programs in combination with irregular grids and the submodeling method in radiation protection calculations.
REFINEMENTS OF THE NEUTRON-PHYSICAL MODEL OF THE SM-3 REACTOR FOR CALCULATIONS USING IMCOR_SM SIMULATOR
Authors: Marikhin N , Valishin M
Published: January 2022

When calculating the maintenance of the operation of nuclear reactors, the most popular are calculations according to precision programs based on detailed core models, which make it possible to predict the characteristics of the upcoming reactor campaign or to calculate the conditions for irradiation of materials in it. Such calculations are of particular importance when justifying the modernization of a reactor or loading new irradiation devices. With small changes in the layout of the core, it is possible, on the basis of operational data, to estimate the magnitude of the systematic error of the design model arising from simplifications in the description of the core, and then make an appropriate correction later. When significant changes take place in the core, for example, during the modernization of the reactor, such corrections can no longer be used, since such an approach can lead to additional uncertainty in the calculations. Due to the fact that neutron-physical calculations in carrying out these works are the most important, the refinement of calculation models and finding the sources of systematic uncertainty is of great practical importance. The paper describes the refinements of the design model of the SM reactor. The division of beryllium into separate zones with the same nuclide composition has been clarified. An optimal list of nuclides for calculations was selected, and the effect of thermal expansion of materials on reactivity was investigated. The approach used by the authors is applicable to other models of research reactors.
MOVING IRRADIATED BY NEUTRONS CONTAINING URANIUM NANOPARTICLES GAS MEDIUM AMPLIFICATION PROPERTIES RESEARCH
Authors: Alekseeva I , Budnik A , Sliuniaev M
Published: January 2022

The search for promising ways of nuclear energy usage has led to the emergence of a whole line of research on the direct conversion of the nuclear reaction products energy into laser radiation. In the present work, the amplifying properties of a spatially inhomogeneous nuclear-excited containing uranium nanoparticles moving argon-xenon medium irradiated by an inhomogeneous neutron field were studied. This work purpose is to investigate the dependence of the laser active medium amplifying properties on the initial gas mixture velocity, neutron pulse duration, and the neutron flux space-time distribution. As a short result, we could present the radial dependence of the laser radiation intensity gain at different points in time showed on the picture below. Summarizing laser radiation intensity gain calculations results: an active laser- medium with a length of 1 m provides a significant amplification (10 or more times) of the laser radiation in one pass. And it seems to be reasonable to continue studying the amplifying properties of such laser-active medium, considering the laser radiation’s wave nature.
SYSTEM OF CODES FOR PHYSICAL DESIGN OF THE LEAD-COOLED FAST REACTOR
Authors: Balovnev A , Davidov V , Zhirnov A , Ivanyuta A , Moiseev A , Soldatov E , Yufereva V
Published: January 2022

One of the actual task at present is the substantiation of the project of the pilot demonstration reactor with a lead coolant BREST-OD-300. For the implementation of large-scale development of nuclear power, which meets modern requirements for new generation reactors, a competitive commercial power unit BR-1200 with an electric capacity of 1200 MW is being designed. To solve complex problems in the study for the optimal configurations of the core, it is required to develop a system of design codes that allows us to perform works on physical design and safety justification. System of codes includes the diffusion software package FACT-BR, the precision software package MCU-BR and the thermos-physical module IVIS-BR. The system is designed to conduct accurate studies of the physical characteristics during the design and computational support of a fast reactor with a lead coolant operating in a closed nuclear fuel cycle. The article provides a brief description of the system of codes. To demonstrate the capabilities of the system of codes, test calculation of the BR-1200 core was carried out.
VALIDATION OF THE GROUP PHYSICAL MODULE IN THE NEUTRON ENERGY CALCULATION PROGRAM KINETICS OF KIR2 BASED ON BENCHMARK TEST C5G7
Authors: Dyachkov I , Ioannisian M
Published: January 2022

Describes the software used in the calculation, according to the KIR2 program and the algorithms developed by the MAGMA group module (Group modulus taking into account anisotropic scattering). The results of the verification stage of the KIR2 program with a group physical module are presented. For verification, we used the C5G7 mathematical benchmark test, in which the energy field is divided into 7 groups. The results of comparison with the MCNP benchmark program on the effective breeding coefficient, confidence intervals and deviations from the benchmark program are presented. A comparison with other programs based on the Monte Carlo method (VIM and UNKMK and other independent calculations using the MCNP program) is also given. In addition, the use of the calculated data of the KIR2 program with high statistics as reference data for deterministic class codes (ATTILA, MCCG3D, CRX, PARTISN, UNKGRO) and programs based on the Monte Carlo method is presented. The analysis of the effectiveness of using comparison methods by deviations and confidence intervals is presented. The results of the verification phase of the KIR2 program with the MAGMA group module are summarized.
TEST CALCULATIONS OF HYDRODYNAMICS OF DISTRIBUTION HEADER SYSTEMS IN NPP HEAT EXCHANGERS AND REACTORS
Authors: Lunina S , Delnov V
Published: January 2022

The paper presents the results of test calculations of hydrodynamics for distribution header systems (DHS) in heat exchangers and reactors of nuclear power plants (NPP). The header system is a typical element of the flow path in the specified NPP equipment, their efficient and safe operation is a priority area for Rosatom State Corporation. The calculations have been performed with the use of computational fluid dynamics software systems known as CFD codes. The results of calculations are presented for the coolant turbulent flow in the flow paths of the flat and cylindrical DHS with different conditions for the coolant supply and removal. A comparative analysis of the calculation results and experimental data on hydrodynamics of DHS typical of the nuclear power plants has been carried out, their satisfactory agreement has been shown. An assumption is made of the legitimacy of using the CFD code for the calculation of hydrodynamics of various DHS types and their specific typical elements.
CROSSER - SOFTWARE MODULE FOR PREPARATION GROUP CONSTANTS FOR ENGINEERING CALCULATIONS FAST REACTORS
Authors: Koscheev V , Tormischev I , Mischin V , Peregudov A , Raskach K , Semenov M , Yakunin A
Published: January 2022

The purpose of this work was to test the CROSSER high-speed constant preparation module for use in the practice of engineering calculations of the neutron-physical characteristics of fast reactors. Computational models of fast reactors BN-600, BN-800 and BN-1200 were used as test tasks. The following neutron-physical parameters were calculated: the criticality of the reactor model, the efficiency of the CPS and the sodium void coefficient of reactivity. The main reactor functionals were calculated using the MMKK engineering program, the constants for which were prepared using the CROSSER module. The initial sets of nuclear data for the MMKK program were the ABBN-93 library of constants and the constants in the ABBN-RFE formats obtained on the basis of the ROSFOND-2010 library. The results obtained using the MMKC program with detailed tracking of the particle energy were used as the reference calculation results. Detailed data in ACE format obtained from the ROSFOND 2010 library were used as the initial nuclear data for this program.
SOME CALCULATION FEATURES OF TRANSIENT BEHAVIOR OF THE REACTOR
Authors: Seleznev E
Published: January 2022

To analyze the features of calculating the transient behavior of fast breeder reactors, both using the classical system of kinetic equations and using the system of partial neutron transfer equations, computational studies of the test model of the MET1000 reactor developed under the Generation-IV project were carried out. As transient processes, displacements (dumping) of control rods were simulated with the assessment of reactivity effects in the specified reactor through the solution of stationary problems, i. e. through the use of an asymptotic estimate of the reactivity obtained from solving stationary homogeneous neutron transfer equations and from processing by the Inverse Solution of the Kinetics Equation method of solving a transient problem. Test calculations were performed in the 26th and 28th group approximations using the ABBN-93 and ABBN-RF libraries and eight groups of delayed neutrons. It is shown that the features of the determination of both calculation and measurement of the effects of reactivity and efficiency of control rods in Fast Breeder Reactors are associated with the presence of both instantaneous and delayed neutrons in the reactor, with the working region of the neutron spectrum and with a change in the shape of the neutron flux in the process of introducing perturbation.
THERMAL AND HYDRAULIC STUDIES OF A HIGH-TEMPERATURE NPP FOR HYDROGEN PRODUCTION AND OTHER TECHNOLOGICAL APPLICATIONS
Authors: Sorokin A , Ivanov A , Kuzina Yu , Morozov A , Denisova N
Published: January 2022

An important problem determining the development of clean energy is the involvement of hydrogen in the fuel cycle. At present, the main method of hydrogen production is steam methane conversion. In the long term, large-scale hydrogen production, this method is not viable due to the consumption of non-renewable resources and the emission of greenhouse gases. Alternative methods of hydrogen production by water splitting methods using thermochemical or electrolysis processes require a high-temperature heat source. Nuclear reactors can serve as the most widely used high-temperature heat sources. The performed neutron-physical and thermophysical studies have shown that there is a fundamental possibility to provide the required parameters of a high-temperature (900-950 °C) with a 600 MW (thermal) fast neutron reactor with a sodium coolant for hydrogen production. It’s possible on the basis of one of the thermochemical cycles or high-temperature electrolysis with a high coefficient of thermal utilization of energy. It is shown that the temperature regime of core fuel elements is determined by a large number of parameters that have a regular and statistical nature. The developed methodology and numerical program allows to take into account, in the fuel assemblies shaped during the campaign, the effect on the temperature distribution of the fuel element cladding and temperature irregularities along the fuel element perimeter in the interchannel exchange fuel assembly, the random distribution of channel cross-sections and the heat generation of fuel elements using the Monte Carlo method, also other factors. For various reactor operating regimes, zones with stable temperature stratification with large gradients and temperature fluctuations have been identified. The results obtained make it possible to judge the amplitude and frequency characteristics of temperature pulsations in these potentially dangerous areas. The relative small size, the type of coolant, the choice of fissile material and structural materials make it possible to create a reactor with inherent properties that ensure increased nuclear and radiation safety.
CHARACTERISTICS OF A STRAIGHT FIN WITH ENERGY RELEASE
Authors: Levchenko V , Kascheev M , Dorokhovich S , Zaytsev A
Published: January 2022

The heat conduction equation for a straight fin with an arbitrary profile in the presence of energy release in the fin is obtained in the article. The resulting equation differs from the approximate equation given in the literature by the presence of energy release and a more accurate determination of the length of the arc element. The equation is solved for the fin of a rectangular profile with continuously operating heat sources. The efficiency of the fin and the heat flow through the base of the fin are determined. It is shown that energy release in the fin increases its efficiency in comparison with the efficiency of the fin in the absence of energy release. There is also a decrease in the heat flow in the presence of energy release in the fin. The restriction on the values of energy release in the fin is found as condition for the applicability of the finning. The fin efficiency must be less than one. If the efficiency exceeds one, the fin plays the opposite role: the flow is directed in the reverse side. To increase the build-up coefficient of the surface, tend to reduce the distance between the fins. There is a limit to such reduction. Theoretically, the distance between the fins should be at least double the maximum thickness of the boundary layer. As experience shows, this distance can be reduced to about one thickness. An approach to achieve the largest build-up coefficient at finning is described in the article.
A SENSITIVITY EVALUATION OF GAS-LIFT PROBE INCLUDED INTO CLADDING FAILURE DETECTION SYSTEM BY THE MODEL OF GASEOUS FISSION PRODUCT SOLUTION/DEGASSING INTO LEAD COOLANT
Authors: Gonchar N , Morkin M
Published: January 2022

Gas-lift probe is an element of cladding failure detection system of perspective lead cooled reactor. Its function is local measurement of gaseous fission product activity in the coolant and the most defected fuel assembly localization. In the coolant leaving the defected fuel assembly the specific activity of gaseous fission products is higher than the average one in the primary circuit. In the barbotage channel of gas-lift probe gaseous fission products diffuse through the bubble interface surface into the volume of the bubbles. The bubbles deliver gaseous fission product to interface surface in the separation volume. The gas enriched with radioactive gaseous fission product goes to measurement volume of the probe. The more significant the damage and the closer the defective fuel assembly is located to the probe input, the more gaseous fission product activity will be registered. The paper presents a model of gaseous activity transfer from cladding defect to probe measuring volume. The gaseous activity transfer is described on the basis of the inert gases dissolution/degassing processes in lead. The gas-lift probe sensitivity was estimated as the ratio of the entry velocity of gaseous activity into the measurement volume to the exit one into the coolant through fuel assemblies cladding defects. A gas-lift probe sensitivity for exposed fuel surface calculated as an example. Gaseous fission products with significant gamma radiation are considered. The calculation results are presented in the article.
DEVELOPMENT AND AUTOMATION OF MEASUREMENT AND REGULATION OF FLOW IN THE HEAT-RELEASING ELEMENT CELL OF FUEL ASSEMBLIES
Authors: Avdeev E , Smirnova V
Published: January 2022

The article presents the results of the development of a formula for determining the flow rate in the fuel element cells, taking into account the real velocities distribution in the flow cross section (average velocity deficit); the functional diagram of the system for monitoring and controlling the flow rates in the fuel element cell and the algorithm of the program of the flow control system at the experimental stands with spills or blowdowns of bundles of heat-releasing element cell simulators. Automation of flow control is carried out on the basis of the previously obtained semiempirical relationship between the volumetric flow rate of the coolant in the fuel element cell, the maximum flow rate in the flow on the cell axis and friction pressure losses along the length of the stabilized flow section; and also on the basis of the dependence of the average speed on the maximum speed and the coefficient of friction resistance. The formula used in the program code makes it possible to determine the coolant flow rate in a tightly packed fuel element cell with an accuracy of up to 2% of the nominal flow rate. The results presented in the article are of practical value for experimenters who investigate the heat transfer (or heat transfer) coefficients of model fuel assemblies, as well as fuel bundles with a limited number of fuel rod simulators, including life tests on full-scale fuel assemblies.
APPLICATION OF NON-ANALOG MONTE CARLO MODELING IN SHIELDING CALCULATIONS OF FAST REACTORS
Authors: Bogdanova E , Tikhomirov G , Suslov I , Homyakov Y
Published: January 2022

In the design and operation of nuclear power plants, one of the most important tasks is to assess the radiation protection of the reactor. Currently, the most widespread are deterministic (method of discrete ordinates) and stochastic computational methods for evaluating functionals. At large attenuations of the neutron flux (by 5-15 orders of magnitude) the deep penetration problems require large computational costs. The most accurate simulation of radiation transfer is achieved by using precision programs that implement the Monte Carlo method with a continuous energy dependence of the cross sections. A detailed description of the geometry and the use of continuous cross sections for particle interactions in calculations lead to high computational costs. To improve computational efficiency, there are variance reduction techniques (non-analog modeling). In this paper the possibility of using non-analog modeling in MCU-FR program by calculating the protection of the fast reactor full-scale model with a lead coolant is considered. The volume-integral neutron fluxes were estimated at points located in a long distance from the center of the reactor core. Analysis results were shown the significant reduction of the variance in the reactor shielding by using the non-analog Monte Carlo method.
EXPERIMENTS ON INVESTIGATION INTO PLUTONIUM TEMPERATURE EFFECT IN PLUTONIUM MULTIPLYING SYSTEMS WITHOUT REFLECTOR
Authors: Vaivod A , Besov S , Yudov A , Andreyev S
Published: January 2022

The article presents the investigation into reactivity and critical parameters affected by temperature changes in cylindrical and spherical space-effective multiplying systems (MS) of plutonium in α-phase without reflector during integral critical experiments. The experiments were performed in “Zababakhin All-Russia Research Institute of Technical Physics” on a critical assembly stand FKBN-2. Brief description of the experimental set up is provided. During the experiments, the multiplication factors and decay constant of the prompt neutrons were determined at a fixed gap between MS parts and different MS temperatures. Values of MS critical gap at different MS temperatures were established experimentally. The change in MS reactivity is initiated by energy release under the plutonium isotopes alpha-decay. As a result, the change in MS temperature during the experiments without its forced regulation may reach ~40 °C. It is proposed to consider this temperature effect when specifying experimental results including error estimation. The obtained experimental data were used to estimate the temperature coefficients of reactivity. The work results are applied to create benchmark models of the multiplying systems.
CURRENT STATE AND ISSUES OF THE HEAVY LIQUID METAL COOLANT TECHNOLOGY DEVELOPMENT (Pb, Pb-Bi)
Authors: Askhadullin R , Legkikh A , Ulyanov V , Voronin I
Published: January 2022

The technology of heavy liquid metal coolant (HLMC) is an important part of the safety system for the operation of reactor facilities with HLMC at all stages of their life cycle: from the preparation of a coolant and loading into the reactor to the decommissioning of the reactor facility (RF). The technology of heavy liquid metal coolant is a set of measures that allow: - to prepare the coolant for filling in the primary circuit of the RF; - to maintain the conditions in the coolant to ensure the corrosion resistance of structural steels; - to perform the coolant cleaning from solid-phase slags both based on lead oxides and based on oxides of structural steel components; - to clean the protective gas from coolant aerosols and corrosion products; - to perform the diagnostics of the contour state. HLMC technology includes both methods of performing technological operations, as well as special equipment, controls, and analysis methods. To date, a significant amount of research and development work has been carried out for the implementation of HLMC in the “BREST-OD-300” and special-purpose RF. The report paper presents the achievements in the field of HLMC technology in recent years of the work of the staff of IPPE JSC.
APPLICATION OF FLOW PATTERNS AND ACCOUNTING OF HYDRODYNAMIC EFFECTS IN THE COURSE OF UPGRADING DISTRIBUTION HEADER SYSTEMS OF NPP HEAT EXCHANGERS AND REACTORS
Authors: Delnov V
Published: January 2022

The article deals with application of flow patterns and hydrodynamic effects in the course of upgrading distribution header systems (DHS) used in heat exchangers and reactors of nuclear power plants (NPP). Consideration is given to four typical axisymmetric DHSs of flat and cylindrical types, which differ in the ways of fluid supply to the collector and its removal from it. The schematic diagrams of flow patterns are shown for different DHS designs of the cylindrical types with central and lateral fluid inlets of the collector. Besides, the water flow pattern is given for the flow path of various DHS designs of a flat type, also with central and lateral fluid inlets of the collector. Hydrodynamic features of the fluid flow in the DHS have been analyzed. It is shown that in a restricted and free DHS the coolant flow is of a jet nature. The fluid flow in the header is characterized by transformation of some types of jets into others, by the presence of eddy and stagnant zones. The designs of heat exchangers and reactors for nuclear power plants with the considered typical DHS characteristics implemented are presented. The DHS designs upgraded with the use of the indicated patterns and effects are demonstrated. The revealed regularity of liquid distribution at the outlet of the DHS flow paths and the hydrodynamics identity property for axially symmetric DHS with flow reversal are analyzed.
EXPERIMENTAL JUSTIFICATION OF DESIGN CHARACTERISTICS OF STEAM GENERATOR RP BREST-OD-300
Authors: Grabezhnaya V , Mikheyev A , Alekhin A , Kryukov A , Tikhomirov A
Published: January 2022

The project BREST-OD-300 reactor plant (RP) with a fast neutron reactor and a lead coolant in the primary circuit is being developed in NIKIET JSC. As a steam generator (SG), a helical-type steam generator with coiled tubes with subcritical pressure water in the second circuit is considered. To substantiate the design characteristics of the secondary coolant at the State Research Center of the Russian Federation - IPPE, thermohydraulic tests of various SG models were carried out at the SPRUT stand Initially, tests were carried out on a model of a coiled steam generator consisting of two three-tube modules with a longitudinal lead flow around a three-tube bundle of coiled tubes. The influence of operating parameters on thermohydraulic characteristics and hydrodynamic stability is shown in the case of operation of one module, as well as in the joint operation of two models in the investigated range of operating parameters. At the second stage, tests of a standard steam generator model were carried out with lead flowing around 18 heat exchange tubes. In the multitube model, the downward movement of the heating coolant took place with the flow around the bundle of heat transfer tubes close to the transverse flow. Data were obtained on the hydrodynamic stability of steam generating tubes and the entire model as a whole when operating in the entire range of changes in operating parameters, which are necessary for creating a databank and further verification of calculation codes describing the ongoing thermohydraulic processes. During the tests in both models of the steam generator, there were no noises inherent in unstable operating modes of the circuit. No pulsations of water and steam temperature were found, respectively, in the inlet and outlet collectors. At high lead temperatures, the temperature of the superheated steam was always close to the lead inlet temperature. A series of works devoted to the study of heat transfer from the side of a lead coolant with a transverse flow around a package of heat exchange tubes in normal heat transfer modes and with freezing of lead has been completed. Studies have been carried out on the effect of oxygen concentration in lead on heat transfer.
REVIEW OF METAL FUEL U-10 wt. % Zr STUDIES
Authors: Kurina I , Frolova M , Chesnokov E
Published: January 2022

The article provides a review of well-known foreign scientific publications devoted to the study of the properties of metallic nuclear fuel based on U-Zr, in composition close to U 10 wt. % Zr, which is widely used in reactors. Differences in the microstructure of fuel made by different methods: extrusion and casting - are considered. The effect of thermal annealing on the change in the microstructure of the alloy is shown. The photographs obtained using optical and electron microscopes are presented, as well as crystallographic data for two phases: α-U and δ-UZr2. The known literature data indicate that the density of uranium-rich U-Zr alloys corresponds to the rule of mixtures. The theoretical density of the alloy U-10 wt. % Zr (U-22.5 at. % Zr) should be taken as 16.2 g/cm3. The results of thermophysical studies of 10 wt. % Zr fuel using the method of differential scanning calorimetry (DSC) are presented. Data on measurements of thermal expansion of U-Zr alloys, as well as thermal conductivity are presented. Most of the thermal conductivity data are either calculated from the measured density, specific heat and thermal diffusivity, or obtained from simulations.
EXPERIMENTAL STUDY OF THE FEATURES OF PROCESS OF SOLUBILITY OF BORIC ACID IN STEAM DURING OF BOILING OF THE COOLANT OF THE WWER IN CASE OF ACCIDENT
Authors: Sakhipgareev A , Shlepkin A , Morozov A
Published: January 2022

The results of experimental studies of the solubility of boric acid in steam in the concentration range of 16-440 g/kg H2O and the pressure range of 0.1-0.3 MPa is considered in the article. An analysis of the literature data on the solubility of boric acid in steam showed that the results do not cover the entire range of parameters (temperature, pressure, concentration of boric acid) typical of a possible emergency situation at NNPs with the WWER. In paper the test facility is described. The obtained experimental results confirm the data available in the literature that the change in the concentration of boric acid in steam is described by a linear law. The range of application of the dependence of the distribution of H3BO3 between the liquid and steam phases of the boiling coolant has been expanded to a concentration of 440 g/kg H2O, which is close to the solubility limit of boric acid in water at a pressure of 0.3 MPa. The obtained experimental data can be used in the calculating of the emergency at nuclear power plants with WWER, taking into account the operating of passive safety systems.
SOFTWARE MODULE FOR CALCULATING THERMOPHYSICAL PARAMETERS IN HYDRAULIC CHANNELS OF SUPERCRITICAL WATER REACTOR
Authors: Zborovskii V , Khoruzhii O , Likhanskii V , Elkin N , Chernetskii M , Mahin V
Published: January 2022

The paper describes the software module FRC-SCP intended for thermohydraulic simulation of a coolant under supercritical pressure (SCP) and a cooled fuel rod. Several designs of supercritical water reactors utilize the transition of the coolant from the pseudoliquid state to pseudogas while it is heated in the reactor core. SCP coolant under pseudophase transition exhibits specific behavior as its density changes significantly. Furthermore, coolant thermophysical properties (density, heat capacity, viscosity, thermal conductivity) can also vary across the coolant channel affecting heat transfer from the fuel rod to the coolant and consequently the fuel temperature. Existing feedback dependencies on coolant density and fuel temperature are important for the nuclear safety analysis of the reactor. The paper considers the present version of the FRC-SCP module. It implements the steady-state thermohydraulic channel solver to calculate coolant parameters: temperatures of the flow core and a heater, as well as coolant pressures, densities etc. User-specified correlations define the heat transfer law under the normal conditions. The module solves the thermal problem for the fuel rod consistently with the channel thermohydraulic problem. It is also possible to couple the FRC-SCP module with the neutron physical codes. Thermohydraulic module is tested against experiments on the heat transfer to SCP water in heated tubes. We discuss the behavior of the fuel cladding under conditions imitating the deteriorated heat transfer modes and the effect of the nuclear fuel thermal conductivity.
BEHAVIOR OF CHEMICALLY ACTIVE ORGANIC IMPURITIES IN SODIUM CIRCUITS
Authors: Zagorulko Yu , Ganichev N , Konovalov M , Alekseev V
Published: January 2022

Mineral oils and some other organic substances are employed as cooling-agents and lubricators of reactor equipment with movable parts. At normal operation condition these substances are able to penetrate into reactor protective gas plenum in vapor and aerosols forms. Due to their thermolysis and interaction with sodium there are produced hydrogen and different hydrocarbons species, that are detected usually in cover gas. Kinetic of the above processes is dependent on temperature conditions and with temperature increase from 350 to 500 °C the evolution of some hydrocarbons could rise up to 1-2 orders of magnitude. There were published notes on accidental ingress of oil in the primary circuit of fast reactor sodium coolant. Investigation results of such accidents simulation on the experimental sodium circuit are presented in the report. There are discussed results on kinetic of volatile hydrocarbons evolution from oil-pyrolysis products depositions on the circuit pipes walls and gas plenum surfaces. As well, different brands of oil are compared on their behavior in sodium environment.