Pub Date : 2021-12-26DOI: 10.55176/2414-1038-2021-4-18-27
M. Belonogov, I. Volkov, D. Modestov, V. Simonenko, D. Khmelnitsky, V. Eliseev
Reduction in the amount of accumulated minor actinides (Np, Am, and Cm isotopes), as well as, the long-lived fission products, to the extent possible, is a top-priority task for the nuclear power engineering. Transmutation of this waste in a special-purpose molten-salt burner reactor has been currently proposed as a possible solution for this problem. The objective of the present paper is to investigate basic regularities of Np, Am, and Cm transmutation in the molten salt burner reactor and to determine optimal operational conditions of the reactor. In the optimal equilibrium mode just fluorides of these elements are added into the fuel composition but the fission products are extracted from it. This mode are reached by maintaining a specified actinide concentration. In order to provide the reactor criticality with a lower than optimum actinide concentration, it is required to add plutonium in the feed fuel instead of a certain amount of minor actinides, thus impairing the transmutation efficiency. To keep the critical state in the equilibrium mode, in situations where the actinide concentration is higher than the optimum one, it is essential to extract a portion of fuel with high content of 238Pu. It has been shown that the reactor configuration is the primary factor responsible for optimum actinide concentration value and that this value varies little with the feed fuel composition, the type of salt dissolving agent, and the fuel reprocessing mode. Optimal concentration for the molten-salt burner reactor with the core volume ranging from 2 to 30 m3 is within 17…10 mole %.
{"title":"TRANSMUTATION OF MINOR ACTINIDES IN MOLTEN SALT BURNER REACTOR","authors":"M. Belonogov, I. Volkov, D. Modestov, V. Simonenko, D. Khmelnitsky, V. Eliseev","doi":"10.55176/2414-1038-2021-4-18-27","DOIUrl":"https://doi.org/10.55176/2414-1038-2021-4-18-27","url":null,"abstract":"Reduction in the amount of accumulated minor actinides (Np, Am, and Cm isotopes), as well as, the long-lived fission products, to the extent possible, is a top-priority task for the nuclear power engineering. Transmutation of this waste in a special-purpose molten-salt burner reactor has been currently proposed as a possible solution for this problem. The objective of the present paper is to investigate basic regularities of Np, Am, and Cm transmutation in the molten salt burner reactor and to determine optimal operational conditions of the reactor. In the optimal equilibrium mode just fluorides of these elements are added into the fuel composition but the fission products are extracted from it. This mode are reached by maintaining a specified actinide concentration. In order to provide the reactor criticality with a lower than optimum actinide concentration, it is required to add plutonium in the feed fuel instead of a certain amount of minor actinides, thus impairing the transmutation efficiency. To keep the critical state in the equilibrium mode, in situations where the actinide concentration is higher than the optimum one, it is essential to extract a portion of fuel with high content of 238Pu. It has been shown that the reactor configuration is the primary factor responsible for optimum actinide concentration value and that this value varies little with the feed fuel composition, the type of salt dissolving agent, and the fuel reprocessing mode. Optimal concentration for the molten-salt burner reactor with the core volume ranging from 2 to 30 m3 is within 17…10 mole %.","PeriodicalId":20426,"journal":{"name":"PROBLEMS OF ATOMIC SCIENCE AND TECHNOLOGY. SERIES: NUCLEAR AND REACTOR CONSTANTS","volume":null,"pages":null},"PeriodicalIF":0.0,"publicationDate":"2021-12-26","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"85071343","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2021-12-26DOI: 10.55176/2414-1038-2021-4-154-161
Y. Zagorulko, N. Ganichev, M. Konovalov, V. Alekseev
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.
{"title":"BEHAVIOR OF CHEMICALLY ACTIVE ORGANIC IMPURITIES IN SODIUM CIRCUITS","authors":"Y. Zagorulko, N. Ganichev, M. Konovalov, V. Alekseev","doi":"10.55176/2414-1038-2021-4-154-161","DOIUrl":"https://doi.org/10.55176/2414-1038-2021-4-154-161","url":null,"abstract":"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.","PeriodicalId":20426,"journal":{"name":"PROBLEMS OF ATOMIC SCIENCE AND TECHNOLOGY. SERIES: NUCLEAR AND REACTOR CONSTANTS","volume":null,"pages":null},"PeriodicalIF":0.0,"publicationDate":"2021-12-26","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"83636232","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2021-12-26DOI: 10.55176/2414-1038-2021-4-98-105
V. Levchenko, M. Kascheev, S. Dorokhovich, A. Zaytsev, A. Sorokin
The heat conduction equation for an annular 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. As boundary conditions, the temperature of the base of the fin is set, and at the end of the fin, heat exchange occurs according to the Newton - Richmann law with the environment. The equation for the fin of a rectangular profile is an inhomogeneous modified Bessel equation. Its solution contains the Bessel functions of the imaginary argument of the first and second kind of zero order. The efficiency of the fin and the heat flow through the base of the fin are determined. The energy release in the fin increases its efficiency compared to the efficiency of the fin in the absence of energy release, and also reduces the heat flow. The restriction by 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. In the article, an expression is obtained for the surface build-up coefficient kh. When calculating the heating (cooling) of a body with a finned surface, the heat transfer coefficient should be increased by kh times.
{"title":"CHARACTERISTICS OF AN ANNULAR FIN OF A RECTANGULAR PROFILE WITH ENERGY RELEASE","authors":"V. Levchenko, M. Kascheev, S. Dorokhovich, A. Zaytsev, A. Sorokin","doi":"10.55176/2414-1038-2021-4-98-105","DOIUrl":"https://doi.org/10.55176/2414-1038-2021-4-98-105","url":null,"abstract":"The heat conduction equation for an annular 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. As boundary conditions, the temperature of the base of the fin is set, and at the end of the fin, heat exchange occurs according to the Newton - Richmann law with the environment. The equation for the fin of a rectangular profile is an inhomogeneous modified Bessel equation. Its solution contains the Bessel functions of the imaginary argument of the first and second kind of zero order. The efficiency of the fin and the heat flow through the base of the fin are determined. The energy release in the fin increases its efficiency compared to the efficiency of the fin in the absence of energy release, and also reduces the heat flow. The restriction by 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. In the article, an expression is obtained for the surface build-up coefficient kh. When calculating the heating (cooling) of a body with a finned surface, the heat transfer coefficient should be increased by kh times.","PeriodicalId":20426,"journal":{"name":"PROBLEMS OF ATOMIC SCIENCE AND TECHNOLOGY. SERIES: NUCLEAR AND REACTOR CONSTANTS","volume":null,"pages":null},"PeriodicalIF":0.0,"publicationDate":"2021-12-26","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"90900500","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2021-12-26DOI: 10.55176/2414-1038-2021-4-5-17
T. Bobrovsky, P. Prusachenko, V. Khryachkov, P. D’yachenko
Machine learning is one of the leading directions in digital signal processing. For example, in neutron spectrometry, artificial neural networks are actively used to suppress gamma background when analyzing signals from scintillation detectors. This article describes a method for determining the quality of n/γ-separation by an artificial neural network. The efficiency of the method is demonstrated by analyzing the signals obtained by measuring the prompt neutron spectrum of 252Cf spontaneous fission using a scintillation detector based on a stilbene crystal. The essence of the method is to determine the proportion of falsely identified events for each of the analyzed signal classes using a known reference method. An exemplary gamma-ray source was used to determine the false count of recoil protons. This approach made it possible to estimate the fraction of events from electrons identified as recoil protons and the fraction of recoil protons perceived as electrons, depending on the light yield of the scintillation signal. This, in turn, made it possible to reconstruct the true energy spectra for different types of particles, including for the region of low signal amplitudes, where the separation quality is usually poor. The reconstructing error was less than 8 % for the light yield region of less than 120 keVee.
{"title":"ALGORITHM FOR ESTIMATING THE EFFICIENCY OF NEURAL NETWORKS FOR n/γ-SEPARATION IN ORGANIC SCINTILLATORS","authors":"T. Bobrovsky, P. Prusachenko, V. Khryachkov, P. D’yachenko","doi":"10.55176/2414-1038-2021-4-5-17","DOIUrl":"https://doi.org/10.55176/2414-1038-2021-4-5-17","url":null,"abstract":"Machine learning is one of the leading directions in digital signal processing. For example, in neutron spectrometry, artificial neural networks are actively used to suppress gamma background when analyzing signals from scintillation detectors. This article describes a method for determining the quality of n/γ-separation by an artificial neural network. The efficiency of the method is demonstrated by analyzing the signals obtained by measuring the prompt neutron spectrum of 252Cf spontaneous fission using a scintillation detector based on a stilbene crystal. The essence of the method is to determine the proportion of falsely identified events for each of the analyzed signal classes using a known reference method. An exemplary gamma-ray source was used to determine the false count of recoil protons. This approach made it possible to estimate the fraction of events from electrons identified as recoil protons and the fraction of recoil protons perceived as electrons, depending on the light yield of the scintillation signal. This, in turn, made it possible to reconstruct the true energy spectra for different types of particles, including for the region of low signal amplitudes, where the separation quality is usually poor. The reconstructing error was less than 8 % for the light yield region of less than 120 keVee.","PeriodicalId":20426,"journal":{"name":"PROBLEMS OF ATOMIC SCIENCE AND TECHNOLOGY. SERIES: NUCLEAR AND REACTOR CONSTANTS","volume":null,"pages":null},"PeriodicalIF":0.0,"publicationDate":"2021-12-26","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"79788967","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2021-12-26DOI: 10.55176/2414-1038-2021-4-72-81
O. Mitrofanova, A. Bayramukov, O. Ivlev, D. Urtenov, A. Fedorinov, A. Sorokin
The mechanisms of generation of stable large-scale eddies and flow swirling in complex channels of the 1st and 2nd circuits of transport nuclear power plants (NPP) are revealed. An analysis is given of the influence of vortex formation and flow swirling processes on failures in NPP systems. It is shown that with a complex geometry of the channels of the pipe systems of the thermohydraulic tract and the reactor plant (RP), the unauthorized swirl arising in certain dynamic modes can lead to the effect of blocking the flow caused by the swirl flow crisis. It has been established that the phenomenon of the crisis is accompanied by the generation of low-frequency acoustic vibrations and large pressure losses required to form a recirculation flow zone, which prevents the flow of fluids in complex channels of the nuclear power plant. In the presented work, the manifestation of a swirling flow crisis is considered by examples of modeling the processes of hydrodynamics and heat transfer in the channels of steam generators and the pressure compensation system of transport nuclear power plants. It is shown that the generation of stable vortex formations in the elements of the NPP equipment of the integral type can be directly related to the mechanism of excitation of hazardous vibroresonance effects. Methods for improving the design of pipe systems of steam generating plants are proposed to increase their thermal-hydraulic efficiency and vibration resistance.
{"title":"RESEARCH OF THE VORTEX FLOWS INFLUENCE ON THE NPP ELEMENTS","authors":"O. Mitrofanova, A. Bayramukov, O. Ivlev, D. Urtenov, A. Fedorinov, A. Sorokin","doi":"10.55176/2414-1038-2021-4-72-81","DOIUrl":"https://doi.org/10.55176/2414-1038-2021-4-72-81","url":null,"abstract":"The mechanisms of generation of stable large-scale eddies and flow swirling in complex channels of the 1st and 2nd circuits of transport nuclear power plants (NPP) are revealed. An analysis is given of the influence of vortex formation and flow swirling processes on failures in NPP systems. It is shown that with a complex geometry of the channels of the pipe systems of the thermohydraulic tract and the reactor plant (RP), the unauthorized swirl arising in certain dynamic modes can lead to the effect of blocking the flow caused by the swirl flow crisis. It has been established that the phenomenon of the crisis is accompanied by the generation of low-frequency acoustic vibrations and large pressure losses required to form a recirculation flow zone, which prevents the flow of fluids in complex channels of the nuclear power plant. In the presented work, the manifestation of a swirling flow crisis is considered by examples of modeling the processes of hydrodynamics and heat transfer in the channels of steam generators and the pressure compensation system of transport nuclear power plants. It is shown that the generation of stable vortex formations in the elements of the NPP equipment of the integral type can be directly related to the mechanism of excitation of hazardous vibroresonance effects. Methods for improving the design of pipe systems of steam generating plants are proposed to increase their thermal-hydraulic efficiency and vibration resistance.","PeriodicalId":20426,"journal":{"name":"PROBLEMS OF ATOMIC SCIENCE AND TECHNOLOGY. SERIES: NUCLEAR AND REACTOR CONSTANTS","volume":null,"pages":null},"PeriodicalIF":0.0,"publicationDate":"2021-12-26","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"91481854","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2021-12-26DOI: 10.55176/2414-1038-2021-4-82-97
E. Belavina, I. Belyaev, N. Pyatnitskaya, A. Sorokin
The possibility of using molten salts based on fluorides as a coolant in promising developments in nuclear power systems requires solving several key scientific and technical problems. These problems are associated with development of reliable structural materials and study of thermophysical characteristics of coolant. Solution to the latter problem was constrained by lack of reliable systematic data on physical and chemical properties, specificity of heat transfer processes, and operation technology of promising compositions of fluoride molten salt. In this regard, a comprehensive study, consisting of studying properties of promising fluoride salt melts, as well as creating an experimental base for verification of engineering calculations, is of particular interest for design of reactor facilities. In course of the work, specifics of interaction of electromagnetic forces and buoyancy forces during MHD heat transfer of molten salt, which can manifest itself in the formation of local overheating points or quasi-stationary flow regimes accompanied by low-frequency temperature fluctuations, were experimentally investigated. For first time, dependences of heat transfer coefficients in a stream under influence of a magnetic field in an extended range of operating parameters were experimentally obtained. The results are compared with known patterns and numerical modeling.
{"title":"SPECIFICS OF HEAT TRANSFER AND HYDRODYNAMICS DURING MOLTEN SALT IMITATOR FLOW IN A DUCT","authors":"E. Belavina, I. Belyaev, N. Pyatnitskaya, A. Sorokin","doi":"10.55176/2414-1038-2021-4-82-97","DOIUrl":"https://doi.org/10.55176/2414-1038-2021-4-82-97","url":null,"abstract":"The possibility of using molten salts based on fluorides as a coolant in promising developments in nuclear power systems requires solving several key scientific and technical problems. These problems are associated with development of reliable structural materials and study of thermophysical characteristics of coolant. Solution to the latter problem was constrained by lack of reliable systematic data on physical and chemical properties, specificity of heat transfer processes, and operation technology of promising compositions of fluoride molten salt. In this regard, a comprehensive study, consisting of studying properties of promising fluoride salt melts, as well as creating an experimental base for verification of engineering calculations, is of particular interest for design of reactor facilities. In course of the work, specifics of interaction of electromagnetic forces and buoyancy forces during MHD heat transfer of molten salt, which can manifest itself in the formation of local overheating points or quasi-stationary flow regimes accompanied by low-frequency temperature fluctuations, were experimentally investigated. For first time, dependences of heat transfer coefficients in a stream under influence of a magnetic field in an extended range of operating parameters were experimentally obtained. The results are compared with known patterns and numerical modeling.","PeriodicalId":20426,"journal":{"name":"PROBLEMS OF ATOMIC SCIENCE AND TECHNOLOGY. SERIES: NUCLEAR AND REACTOR CONSTANTS","volume":null,"pages":null},"PeriodicalIF":0.0,"publicationDate":"2021-12-26","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"78559402","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2021-12-26DOI: 10.55176/2414-1038-2021-4-35-46
V. Likhanskii, A. Sorokin, V. Zborovskii, K. Ulibyshev, A. Strozhuk, I. Chernov
The paper presents a description of the RTOP-CA code models designed for simulating the behavior of (U, Gd)O2 fuel rods, presents the examples of program verification using experimental data obtained at the research reactor in Halden and in the post-irradiation examinations of WWER fuel. The experience of operating WWER fuel has shown that differences in the thermomechanical behavior of (U, Gd)O2 fuel elements and UO2 fuel rods during the first - second fuel cycles can be observed. In particular, an accelerated increase in the length of the (U, Gd)O2 fuel rod cladding as compared to UO2 fuel elements is possible. To simulate the behavior of (U, Gd)O2 fuel rods under irradiation in the RTOP-CA code, models that describe the features of uranium-gadolinium fuel were developed. To calculate the evolution of heat release in (U, Gd)O2 fuel rods in the RTOP-CA code, a model of burnup of gadolinium isotopes was implemented. The developed model makes it possible, without using of detailed neutron-physical calculations, to simulate the burnup of gadolinium and the evolution of heat generation along the radius of the pellets. The models take into account the effect of the gadolinium oxide content on the thermophysical and mechanical properties of the fuel. The calculations performed using the RTOP-CA code are in good agreement with the results obtained during the experiments in Halden on the dynamics of the central fuel temperature, gas pressure in the (U, Gd)O2 fuel rod, and the elongation of the fuel column and cladding. To simulate the accelerated elongation of (U, Gd)O2 fuel rods in comparison with standard WWER fuel elements at the initial stages of fuel operation at nuclear power plants in the RTOP-CA code, a model of mechanical interaction of pellets and cladding was developed taking into account the misaligned arrangement of pellets.
{"title":"SIMULATION OF THE THERMOMECHANICAL BEHAVIOR OF (U, Gd)O2 FUEL ELEMENTS IN THE RTOP-CA CODE","authors":"V. Likhanskii, A. Sorokin, V. Zborovskii, K. Ulibyshev, A. Strozhuk, I. Chernov","doi":"10.55176/2414-1038-2021-4-35-46","DOIUrl":"https://doi.org/10.55176/2414-1038-2021-4-35-46","url":null,"abstract":"The paper presents a description of the RTOP-CA code models designed for simulating the behavior of (U, Gd)O2 fuel rods, presents the examples of program verification using experimental data obtained at the research reactor in Halden and in the post-irradiation examinations of WWER fuel. The experience of operating WWER fuel has shown that differences in the thermomechanical behavior of (U, Gd)O2 fuel elements and UO2 fuel rods during the first - second fuel cycles can be observed. In particular, an accelerated increase in the length of the (U, Gd)O2 fuel rod cladding as compared to UO2 fuel elements is possible. To simulate the behavior of (U, Gd)O2 fuel rods under irradiation in the RTOP-CA code, models that describe the features of uranium-gadolinium fuel were developed. To calculate the evolution of heat release in (U, Gd)O2 fuel rods in the RTOP-CA code, a model of burnup of gadolinium isotopes was implemented. The developed model makes it possible, without using of detailed neutron-physical calculations, to simulate the burnup of gadolinium and the evolution of heat generation along the radius of the pellets. The models take into account the effect of the gadolinium oxide content on the thermophysical and mechanical properties of the fuel. The calculations performed using the RTOP-CA code are in good agreement with the results obtained during the experiments in Halden on the dynamics of the central fuel temperature, gas pressure in the (U, Gd)O2 fuel rod, and the elongation of the fuel column and cladding. To simulate the accelerated elongation of (U, Gd)O2 fuel rods in comparison with standard WWER fuel elements at the initial stages of fuel operation at nuclear power plants in the RTOP-CA code, a model of mechanical interaction of pellets and cladding was developed taking into account the misaligned arrangement of pellets.","PeriodicalId":20426,"journal":{"name":"PROBLEMS OF ATOMIC SCIENCE AND TECHNOLOGY. SERIES: NUCLEAR AND REACTOR CONSTANTS","volume":null,"pages":null},"PeriodicalIF":0.0,"publicationDate":"2021-12-26","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"86983418","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2021-12-26DOI: 10.55176/2414-1038-2021-4-131-146
V. Zborovskii, O. Khoruzhii, V. Likhanskii, N. Elkin, M. Chernetskii, V. Mahin
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.
{"title":"SOFTWARE MODULE FOR CALCULATING THERMOPHYSICAL PARAMETERS IN HYDRAULIC CHANNELS OF SUPERCRITICAL WATER REACTOR","authors":"V. Zborovskii, O. Khoruzhii, V. Likhanskii, N. Elkin, M. Chernetskii, V. Mahin","doi":"10.55176/2414-1038-2021-4-131-146","DOIUrl":"https://doi.org/10.55176/2414-1038-2021-4-131-146","url":null,"abstract":"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.","PeriodicalId":20426,"journal":{"name":"PROBLEMS OF ATOMIC SCIENCE AND TECHNOLOGY. SERIES: NUCLEAR AND REACTOR CONSTANTS","volume":null,"pages":null},"PeriodicalIF":0.0,"publicationDate":"2021-12-26","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"89003855","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2021-12-26DOI: 10.55176/2414-1038-2021-4-106-120
Yu. A. Kuzina, A. Sorokin, N. Denisova, M. Kascheev
The results of experimental studies of heat transfer carried out at the State Research Center RF - IPPE on models of fuel assemblies in cores and screens of fast reactors BOR-60, BN-350, BN-600, BN-800 (the relative pitch of the fuel element simulators varied in the range 1.04≤s/d≤1.24) with natural and mixed convection of liquid metal coolants, which are the basis for the development of design codes. Model fuel assemblies of fast reactors contain 37 fuel elements (fuel rod simulators) located in a triangular lattice and enclosed in a hexagonal cover. The elements were heated with nichrome heaters, which ensure a constant heat flux on the inner surface of the elements and along the length of the heat generation. The central, lateral and corner elements are rotary. On their surface, 12 chromel-alumel thermocouples each in covers made of X18H9T steel (diameter of the cover 0.8-0.5 mm, diameter of thermoelectrodes 0.2 mm) are embedded, measuring the coolant temperature in the collectors of the models, as well as in each cell at the exit from the bundles. The spacing of the elements is carried out using wire coils; variants of smooth fuel rod simulators are also used. Experimental studies on model fuel assemblies revealed regularities in the formation of temperature fields in fuel elements and coolant. Natural convection manifests itself in the region of low velocities, promotes fluid mixing between channels, leveling the uneven heating of the coolant in the cross section of the fuel assembly, and reduces the azimuthal non-uniformity of the temperature of the wall fuel elements in the fuel assembly. The results of experimental studies show that the effect of natural convection manifests itself in the range of Pe<10 (Re<2000) numbers in gratings with relative steps (s/d<1.05) and in wide bundles (s/d<1.3 to a greater extent in bundles of smooth fuel rods in comparison with bundles of ribbed fuel rods. The introduction of displacers into the peripheral channels of a fuel assembly does not fundamentally change the nature of the temperature field in a fuel assembly as compared to the version of the geometry of a fuel assembly without displacers. An attempt was made to generalize the experimental data using the Gr*Pr criterion, where Gr* is the modified Grashof number calculated from the local heat flux on the heat exchange surface and the axial coordinate measured from the beginning of heat generation. In the dependences DTwmax = f(Gr*Pr2) at Pe<100 or ribbed side elements and Pe<10 for smooth side elements, two regions of DTwmax change in the Gr*Pr2 growth function are observed - first, DTwmax ncreases to a certain “limiting” value characteristic for fixed Pe, and then - a drop in DTwmax at large values of the parameter Gr*Pr2. As the Peclet number increases, the “limiting” value of DTwmax hifts to the region of larger values of Gr*Pr2; and at Pe=370 or ribbed elements and Pe=26.5; 100 or smooth elements, the limiting value of DTwmax is not achieved in
{"title":"EXPERIMENTAL MODELING OF HEAT EXCHANGE IN HEAT GENERATING ASSEMBLIES OF FAST REACTORS IN ACCIDENT REGIMES UNDER NATURAL CONVECTION CONDITIONS","authors":"Yu. A. Kuzina, A. Sorokin, N. Denisova, M. Kascheev","doi":"10.55176/2414-1038-2021-4-106-120","DOIUrl":"https://doi.org/10.55176/2414-1038-2021-4-106-120","url":null,"abstract":"The results of experimental studies of heat transfer carried out at the State Research Center RF - IPPE on models of fuel assemblies in cores and screens of fast reactors BOR-60, BN-350, BN-600, BN-800 (the relative pitch of the fuel element simulators varied in the range 1.04≤s/d≤1.24) with natural and mixed convection of liquid metal coolants, which are the basis for the development of design codes. Model fuel assemblies of fast reactors contain 37 fuel elements (fuel rod simulators) located in a triangular lattice and enclosed in a hexagonal cover. The elements were heated with nichrome heaters, which ensure a constant heat flux on the inner surface of the elements and along the length of the heat generation. The central, lateral and corner elements are rotary. On their surface, 12 chromel-alumel thermocouples each in covers made of X18H9T steel (diameter of the cover 0.8-0.5 mm, diameter of thermoelectrodes 0.2 mm) are embedded, measuring the coolant temperature in the collectors of the models, as well as in each cell at the exit from the bundles. The spacing of the elements is carried out using wire coils; variants of smooth fuel rod simulators are also used. Experimental studies on model fuel assemblies revealed regularities in the formation of temperature fields in fuel elements and coolant. Natural convection manifests itself in the region of low velocities, promotes fluid mixing between channels, leveling the uneven heating of the coolant in the cross section of the fuel assembly, and reduces the azimuthal non-uniformity of the temperature of the wall fuel elements in the fuel assembly. The results of experimental studies show that the effect of natural convection manifests itself in the range of Pe<10 (Re<2000) numbers in gratings with relative steps (s/d<1.05) and in wide bundles (s/d<1.3 to a greater extent in bundles of smooth fuel rods in comparison with bundles of ribbed fuel rods. The introduction of displacers into the peripheral channels of a fuel assembly does not fundamentally change the nature of the temperature field in a fuel assembly as compared to the version of the geometry of a fuel assembly without displacers. An attempt was made to generalize the experimental data using the Gr*Pr criterion, where Gr* is the modified Grashof number calculated from the local heat flux on the heat exchange surface and the axial coordinate measured from the beginning of heat generation. In the dependences DTwmax = f(Gr*Pr2) at Pe<100 or ribbed side elements and Pe<10 for smooth side elements, two regions of DTwmax change in the Gr*Pr2 growth function are observed - first, DTwmax ncreases to a certain “limiting” value characteristic for fixed Pe, and then - a drop in DTwmax at large values of the parameter Gr*Pr2. As the Peclet number increases, the “limiting” value of DTwmax hifts to the region of larger values of Gr*Pr2; and at Pe=370 or ribbed elements and Pe=26.5; 100 or smooth elements, the limiting value of DTwmax is not achieved in ","PeriodicalId":20426,"journal":{"name":"PROBLEMS OF ATOMIC SCIENCE AND TECHNOLOGY. SERIES: NUCLEAR AND REACTOR CONSTANTS","volume":null,"pages":null},"PeriodicalIF":0.0,"publicationDate":"2021-12-26","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"88882125","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2021-09-26DOI: 10.55176/2414-1038-2021-3-206-212
V. Smykov, S. Kanukhina, K. Legkikh
For timely and prompt detection of the inter-circuit depressurization of the steam generator of the “sodium-water” type in the 2nd circuit, continuous monitoring of the content of hydrogen, as a product of the interaction of sodium with water, is carried out in all operating modes. However, in reality, the main source of hydrogen in the sodium of the 2nd circuit is the process of electrochemical corrosion of 10X2M steam generator steel in a steam-water medium, the speed of which is never zero and is controlled at the minimum achievable level by means of the water-chemical regime of the 3rd circuit. At the same time, during the operation of the steam generator, deposits of corrosion products accumulate on the surface of the structural steel from the side of the 3rd circuit. The removal of deposits is carried out with the help of operational chemical flushes (ECPs). However, during the ECP, the corrosion rate of steel inevitably increases significantly (by about three orders of magnitude), which leads to an increase in the concentration of hydrogen in water and in sodium. This phenomenon has previously caused concern to the operational services of nuclear power plants due to the likely through-corrosion during regular ECP (every year, one of the three steam generators is washed). However, it is unacceptable to operate a steam generator without ECP, and to reduce the corrosion rate of steel of steam generator pipes in a steam-water environment, after ECP, a chemical passivation stage is carried out with different formulations, the effectiveness of which is different. This article presents a physical model of the mass transfer of corrosive hydrogen in a steam generator during chemical washing, passivation and start-up of the steam generator. An operational criterion for evaluating the effectiveness of passivation by the rate of increase in the concentration of hydrogen in the sodium of the second circuit during the subsequent start-up of BN-600 steam generators is proposed: if the rate of increase in the concentration of hydrogen in the 2nd circuit is less than 0.05 ppm/h, then passivation is effective, if more than the passivation formula requires revision. For BN-800 steam generators, the technology of passivation after ECP has not yet been determined, the article proposes a calculated criterion for its efficiency of 0.015 ppm/h.
{"title":"THE PHYSICAL MODEL OF MASS TRANSFER CORROSIVE HYDROGEN IN FAST REACTOR STEAM GENERATORS OF THE SODIUM-WATER TYPE","authors":"V. Smykov, S. Kanukhina, K. Legkikh","doi":"10.55176/2414-1038-2021-3-206-212","DOIUrl":"https://doi.org/10.55176/2414-1038-2021-3-206-212","url":null,"abstract":"For timely and prompt detection of the inter-circuit depressurization of the steam generator of the “sodium-water” type in the 2nd circuit, continuous monitoring of the content of hydrogen, as a product of the interaction of sodium with water, is carried out in all operating modes. However, in reality, the main source of hydrogen in the sodium of the 2nd circuit is the process of electrochemical corrosion of 10X2M steam generator steel in a steam-water medium, the speed of which is never zero and is controlled at the minimum achievable level by means of the water-chemical regime of the 3rd circuit. At the same time, during the operation of the steam generator, deposits of corrosion products accumulate on the surface of the structural steel from the side of the 3rd circuit. The removal of deposits is carried out with the help of operational chemical flushes (ECPs). However, during the ECP, the corrosion rate of steel inevitably increases significantly (by about three orders of magnitude), which leads to an increase in the concentration of hydrogen in water and in sodium. This phenomenon has previously caused concern to the operational services of nuclear power plants due to the likely through-corrosion during regular ECP (every year, one of the three steam generators is washed). However, it is unacceptable to operate a steam generator without ECP, and to reduce the corrosion rate of steel of steam generator pipes in a steam-water environment, after ECP, a chemical passivation stage is carried out with different formulations, the effectiveness of which is different. This article presents a physical model of the mass transfer of corrosive hydrogen in a steam generator during chemical washing, passivation and start-up of the steam generator. An operational criterion for evaluating the effectiveness of passivation by the rate of increase in the concentration of hydrogen in the sodium of the second circuit during the subsequent start-up of BN-600 steam generators is proposed: if the rate of increase in the concentration of hydrogen in the 2nd circuit is less than 0.05 ppm/h, then passivation is effective, if more than the passivation formula requires revision. For BN-800 steam generators, the technology of passivation after ECP has not yet been determined, the article proposes a calculated criterion for its efficiency of 0.015 ppm/h.","PeriodicalId":20426,"journal":{"name":"PROBLEMS OF ATOMIC SCIENCE AND TECHNOLOGY. SERIES: NUCLEAR AND REACTOR CONSTANTS","volume":null,"pages":null},"PeriodicalIF":0.0,"publicationDate":"2021-09-26","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"74590536","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
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