Pub Date : 2021-06-26DOI: 10.55176/2414-1038-2021-2-174-180
V. Blokhin, V. Borisov, V. Zhmurin, I. Zazorin, A. Kamayev, I. Pakhomov
Solid oxide electrolyte based on zirconium dioxide stabilized with calcium oxide or yttrium oxide is the most studied. It’s currently widely used to control oxygen in gas, in the metallurgical industry production of steel, non-ferrous metals, operation of nuclear power plants with a heavy coolant, and therefore it’s interest to use it to control the oxygen content in alkaline coolants, for example, in sodium. Sodium is an extremely reducing agent for metal oxides. There are practically no literature data on the limiting value of the partial oxygen pressure and temperature for an electrolyte based on zirconium dioxide stabilized with yttrium oxide. This work presents experimental studies of the applicability of solid polycrystalline oxide electrolyte 0.85ZrO2•0.15Y2O3 for determining the oxygen content in sodium at a temperature of (400 ± 5) °C. Studies of the electrolyte 0.85ZrO2•0.15Y2O3 were carried out in the working section, which is a galvanic concentration cell (GCC). The electrolyte in the form of a pellet with a diameter of 4 mm and a length of 5-7 mm is hermetically inserted into an insulator made of alumina-magnesia spinel with the addition of magnesium oxide, which is reinforced with EI-852 steel. The reference electrode was placed in an insulator made of magnesia-alumina spinel with the addition of magnesium oxide and was hermetically sealed from the environment by a sealed lead. A weighed portion of sodium was placed in a small tank made of nickel. To change the concentration of oxygen in sodium, weighed portions of a deoxidizer were introduced into it. Lithium was used as a deoxidizer. EMF of GCC was measured by a ph-meter - ionometer “Expert 001” combined with a computer. The kinetics of the change in the each lithium sample input EMF of the GCC is presented. The weighed portions of lithium were injected until the EMF of the GCC changes with the last injection of the subsequent weighed portion. This value of the EMF of the GCC will be the limit of the applicability of a solid electrolyte to control and dose oxygen into sodium. From the measured value of the EMF GCC obtained after introducing weighed portions of lithium, the lower limit of applicability of the electrolyte was calculated from the partial pressure of oxygen over sodium and the lower limit of applicability of the electrolyte was determined from the oxygen content in sodium using the Nodena formula for the oxygen solubility in sodium. It is shown that the lower limit of applicability of solid polycrystalline oxide electrolyte 0.85ZrO2•0.15Y2O3 for monitoring the oxygen content in sodium at a temperature of (400 ± 5) °C is ~7•10-5 ppm, and for the partial pressure of oxygen over sodium - 4.6•10-59 Pa.
{"title":"EXPERIMENTAL STUDY OF THE APPLICABILITY OF SOLID OXIDE ELECTROLYTE TO DETERMINE THE LOWER LIMIT OF OXYGEN CONTROL IN SODIUM","authors":"V. Blokhin, V. Borisov, V. Zhmurin, I. Zazorin, A. Kamayev, I. Pakhomov","doi":"10.55176/2414-1038-2021-2-174-180","DOIUrl":"https://doi.org/10.55176/2414-1038-2021-2-174-180","url":null,"abstract":"Solid oxide electrolyte based on zirconium dioxide stabilized with calcium oxide or yttrium oxide is the most studied. It’s currently widely used to control oxygen in gas, in the metallurgical industry production of steel, non-ferrous metals, operation of nuclear power plants with a heavy coolant, and therefore it’s interest to use it to control the oxygen content in alkaline coolants, for example, in sodium. Sodium is an extremely reducing agent for metal oxides. There are practically no literature data on the limiting value of the partial oxygen pressure and temperature for an electrolyte based on zirconium dioxide stabilized with yttrium oxide. This work presents experimental studies of the applicability of solid polycrystalline oxide electrolyte 0.85ZrO2•0.15Y2O3 for determining the oxygen content in sodium at a temperature of (400 ± 5) °C. Studies of the electrolyte 0.85ZrO2•0.15Y2O3 were carried out in the working section, which is a galvanic concentration cell (GCC). The electrolyte in the form of a pellet with a diameter of 4 mm and a length of 5-7 mm is hermetically inserted into an insulator made of alumina-magnesia spinel with the addition of magnesium oxide, which is reinforced with EI-852 steel. The reference electrode was placed in an insulator made of magnesia-alumina spinel with the addition of magnesium oxide and was hermetically sealed from the environment by a sealed lead. A weighed portion of sodium was placed in a small tank made of nickel. To change the concentration of oxygen in sodium, weighed portions of a deoxidizer were introduced into it. Lithium was used as a deoxidizer. EMF of GCC was measured by a ph-meter - ionometer “Expert 001” combined with a computer. The kinetics of the change in the each lithium sample input EMF of the GCC is presented. The weighed portions of lithium were injected until the EMF of the GCC changes with the last injection of the subsequent weighed portion. This value of the EMF of the GCC will be the limit of the applicability of a solid electrolyte to control and dose oxygen into sodium. From the measured value of the EMF GCC obtained after introducing weighed portions of lithium, the lower limit of applicability of the electrolyte was calculated from the partial pressure of oxygen over sodium and the lower limit of applicability of the electrolyte was determined from the oxygen content in sodium using the Nodena formula for the oxygen solubility in sodium. It is shown that the lower limit of applicability of solid polycrystalline oxide electrolyte 0.85ZrO2•0.15Y2O3 for monitoring the oxygen content in sodium at a temperature of (400 ± 5) °C is ~7•10-5 ppm, and for the partial pressure of oxygen over sodium - 4.6•10-59 Pa.","PeriodicalId":20426,"journal":{"name":"PROBLEMS OF ATOMIC SCIENCE AND TECHNOLOGY. SERIES: NUCLEAR AND REACTOR CONSTANTS","volume":null,"pages":null},"PeriodicalIF":0.0,"publicationDate":"2021-06-26","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"80627002","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 : 2021-06-26DOI: 10.55176/2414-1038-2021-2-139-166
A. Sorokin, J. Kuzina, N. Denisova
The article describes the factors under the influence of which the formation of thermohydraulic characteristics occurs in the fuel assemblies of the core of fast reactors with liquid metal cooling. It is shown that one of the most important factors is a complex multiply connected geometry of a stochastic nature, subject to deformation during the campaign under the influence of temperature irregularities and radiation effects. The paper presents and analyzes the results of experimental and computational studies of the velocity field and shear stress, the microstructure of turbulence, momentum transfer in the central and peripheral regions of fuel assemblies without and with displacers, as well as in the case of deformation of the lattice of rods. The intensification of turbulent momentum transfer in the channels in the azimuthal direction in the area of the gaps between the rods is demonstrated. The anisotropy coefficient of turbulent momentum transfer reaches 30-40 units. The performed analysis indicated a significant difference in the calculated in the framework of semi-empirical models of turbulent transfer and experimental dependences of the coefficients of turbulent transfer of momentum in the radial and azimuthal directions and the coefficients of anisotropy of turbulent transfer of momentum in rod bundles. The results of an open benchmark on the thermohydraulics of fuel assemblies showed that common commercial computational thermohydraulic codes only approximately describe the experimental data. It is shown that the intensification of turbulent momentum transfer in the channels of rod assemblies is due to the appearance of large-scale turbulent momentum transfer (secondary flows). The contribution of large-scale turbulent momentum transfer to the kinetic energy of turbulent pulsations, azimuthal turbulent shear stresses, and turbulent momentum transfer coefficients in rod assemblies is calculated. An empirical dependence of the coefficient of interchannel turbulent impulse exchange in bundles of smooth rods is obtained, on the basis of a semi-empirical model, data on interchannel turbulent impulse exchange in assemblies of smooth rods are generalized, and the intensification of interchannel turbulent exchange in close lattices of rods is explained. Data on hydraulic resistance in bundles of smooth rods are analyzed. The tasks of further research are discussed.
{"title":"HYDRODYNAMICS OF TURBULENT FLOW IN A FAST REACTOR FUEL ASSEMBLIES (VELOCITY FIELD AND MICROSTRUCTURE OF TURBULENCE)","authors":"A. Sorokin, J. Kuzina, N. Denisova","doi":"10.55176/2414-1038-2021-2-139-166","DOIUrl":"https://doi.org/10.55176/2414-1038-2021-2-139-166","url":null,"abstract":"The article describes the factors under the influence of which the formation of thermohydraulic characteristics occurs in the fuel assemblies of the core of fast reactors with liquid metal cooling. It is shown that one of the most important factors is a complex multiply connected geometry of a stochastic nature, subject to deformation during the campaign under the influence of temperature irregularities and radiation effects. The paper presents and analyzes the results of experimental and computational studies of the velocity field and shear stress, the microstructure of turbulence, momentum transfer in the central and peripheral regions of fuel assemblies without and with displacers, as well as in the case of deformation of the lattice of rods. The intensification of turbulent momentum transfer in the channels in the azimuthal direction in the area of the gaps between the rods is demonstrated. The anisotropy coefficient of turbulent momentum transfer reaches 30-40 units. The performed analysis indicated a significant difference in the calculated in the framework of semi-empirical models of turbulent transfer and experimental dependences of the coefficients of turbulent transfer of momentum in the radial and azimuthal directions and the coefficients of anisotropy of turbulent transfer of momentum in rod bundles. The results of an open benchmark on the thermohydraulics of fuel assemblies showed that common commercial computational thermohydraulic codes only approximately describe the experimental data. It is shown that the intensification of turbulent momentum transfer in the channels of rod assemblies is due to the appearance of large-scale turbulent momentum transfer (secondary flows). The contribution of large-scale turbulent momentum transfer to the kinetic energy of turbulent pulsations, azimuthal turbulent shear stresses, and turbulent momentum transfer coefficients in rod assemblies is calculated. An empirical dependence of the coefficient of interchannel turbulent impulse exchange in bundles of smooth rods is obtained, on the basis of a semi-empirical model, data on interchannel turbulent impulse exchange in assemblies of smooth rods are generalized, and the intensification of interchannel turbulent exchange in close lattices of rods is explained. Data on hydraulic resistance in bundles of smooth rods are analyzed. The tasks of further research are discussed.","PeriodicalId":20426,"journal":{"name":"PROBLEMS OF ATOMIC SCIENCE AND TECHNOLOGY. SERIES: NUCLEAR AND REACTOR CONSTANTS","volume":null,"pages":null},"PeriodicalIF":0.0,"publicationDate":"2021-06-26","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"78106937","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 : 2021-06-26DOI: 10.55176/2414-1038-2021-2-236-255
J. Kuzina, M. Arnoldov, Yu. I. Orlov, A. Sorokin
The article presents the main research results of thermal physicists of the IPPE from its inception to the present time. Research results in the areas of heat and mass transfer and hydrodynamics of coolants (liquid metals, water), physical chemistry and technology of liquid metal coolants for nuclear power plants for various purposes (nuclear power plants, nuclear submarines, space nuclear power plants), development codes, innovative projects, non-nuclear technologies for the use of liquid metals, heat pipes, analysis and generalization of thermophysical data are considered in the article. As a result of a large complex of experimental and computational studies, the fundamental physicochemical and thermohydraulic regularities of the coolant - impurities - structural materials - protective gas have been studied, scientific foundations have been created for the use of liquid metal coolants in nuclear power. Studies have been carried out to substantiate the technical and economic characteristics of nuclear fuel for operating, under construction and future NPPs of VVER RP, design solutions for passive safety, technical solutions and hydrogen safety devices, heat removal from the reactor through a steam generator and PHRS in case of beyond design basis accidents. As well as design solutions and safety for NPP designs with BN-1200 reactor with sodium coolant, BREST-OD-300 reactor with lead coolant, SVBR-100 reactor with lead-bismuth alloy, MBIR research reactor. The results of these studies made it possible, together with institutes and design organizations, to scientifically substantiate thermal-hydraulic parameters and highly efficient technological processes, develop and practically implement devices and systems that ensure the successful operation of fundamentally new nuclear power plants cooled by water and liquid metals, with original scientific and technical solutions that had no analogue in world practice. R&D works were carried out to substantiate the innovative project VVER with supercritical pressure, the concept of an electro-nuclear subcritical blanket based on the modular principle of constructing an core with liquid-salt melts of fissile materials, studies of thermal hydraulics, mass transfer of high-temperature sodium and the development of a combined coolant purification system to justify the BN-HT reactor with temperature sodium ~900 °C for hydrogen production. The directions of investigations at the present stage are discussed.
{"title":"THERMOPHYSICAL INVESTIGATIONS: FROM THE FIRST TO STAND LARGE-SCALE NUCLEAR ENERGY","authors":"J. Kuzina, M. Arnoldov, Yu. I. Orlov, A. Sorokin","doi":"10.55176/2414-1038-2021-2-236-255","DOIUrl":"https://doi.org/10.55176/2414-1038-2021-2-236-255","url":null,"abstract":"The article presents the main research results of thermal physicists of the IPPE from its inception to the present time. Research results in the areas of heat and mass transfer and hydrodynamics of coolants (liquid metals, water), physical chemistry and technology of liquid metal coolants for nuclear power plants for various purposes (nuclear power plants, nuclear submarines, space nuclear power plants), development codes, innovative projects, non-nuclear technologies for the use of liquid metals, heat pipes, analysis and generalization of thermophysical data are considered in the article. As a result of a large complex of experimental and computational studies, the fundamental physicochemical and thermohydraulic regularities of the coolant - impurities - structural materials - protective gas have been studied, scientific foundations have been created for the use of liquid metal coolants in nuclear power. Studies have been carried out to substantiate the technical and economic characteristics of nuclear fuel for operating, under construction and future NPPs of VVER RP, design solutions for passive safety, technical solutions and hydrogen safety devices, heat removal from the reactor through a steam generator and PHRS in case of beyond design basis accidents. As well as design solutions and safety for NPP designs with BN-1200 reactor with sodium coolant, BREST-OD-300 reactor with lead coolant, SVBR-100 reactor with lead-bismuth alloy, MBIR research reactor. The results of these studies made it possible, together with institutes and design organizations, to scientifically substantiate thermal-hydraulic parameters and highly efficient technological processes, develop and practically implement devices and systems that ensure the successful operation of fundamentally new nuclear power plants cooled by water and liquid metals, with original scientific and technical solutions that had no analogue in world practice. R&D works were carried out to substantiate the innovative project VVER with supercritical pressure, the concept of an electro-nuclear subcritical blanket based on the modular principle of constructing an core with liquid-salt melts of fissile materials, studies of thermal hydraulics, mass transfer of high-temperature sodium and the development of a combined coolant purification system to justify the BN-HT reactor with temperature sodium ~900 °C for hydrogen production. The directions of investigations at the present stage are discussed.","PeriodicalId":20426,"journal":{"name":"PROBLEMS OF ATOMIC SCIENCE AND TECHNOLOGY. SERIES: NUCLEAR AND REACTOR CONSTANTS","volume":null,"pages":null},"PeriodicalIF":0.0,"publicationDate":"2021-06-26","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"90868799","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 : 2021-06-26DOI: 10.55176/2414-1038-2021-2-25-33
V. Dekusar, O. Gurskaya
A possible approach to accounting for the specific present value of plutonium produced in fast reactors of a two-component nuclear power system (NPS) with thermal and fast reactors is described. The approach is based on taking into account the additional income that can be obtained by selling at the market price the natural uranium released when thermal reactors are replaced with fast reactors with MOX-fuel based on plutonium produced in NPS. At the same time, along with the sale of natural uranium, the sale at market value of other products made on its basis, for example, enriched uranium or fuel assemblies for a thermal reactors, can be considered. Relations between the main fuel characteristics of the considered nuclear reactors and the economic parameters characterizing the efficiency of nuclear reactors and the fuel cycle of a NPS are obtained. Using the methodology described in this paper, a computational study of the specific present value of plutonium in a two-product nuclear power model with commercial sodium high-power fast reactor and VVER-1200 reactors was carried out. The calculation results in all considered cases indicate very significant present value of plutonium. Comparison of the obtained cost of plutonium, which is ultimately based on the energy equivalent of plutonium and uranium, and the cost of plutonium, determined on the basis of the costs of the back-end of the fuel cycle (plutonium extraction from SNF), show the economic efficiency of closing the fuel cycle even at existing uranium prices.
{"title":"ON THE ISSUE OF PLUTONIUM COST IN A TWO-COMPONENT NUCLEAR POWER SYSTEM","authors":"V. Dekusar, O. Gurskaya","doi":"10.55176/2414-1038-2021-2-25-33","DOIUrl":"https://doi.org/10.55176/2414-1038-2021-2-25-33","url":null,"abstract":"A possible approach to accounting for the specific present value of plutonium produced in fast reactors of a two-component nuclear power system (NPS) with thermal and fast reactors is described. The approach is based on taking into account the additional income that can be obtained by selling at the market price the natural uranium released when thermal reactors are replaced with fast reactors with MOX-fuel based on plutonium produced in NPS. At the same time, along with the sale of natural uranium, the sale at market value of other products made on its basis, for example, enriched uranium or fuel assemblies for a thermal reactors, can be considered. Relations between the main fuel characteristics of the considered nuclear reactors and the economic parameters characterizing the efficiency of nuclear reactors and the fuel cycle of a NPS are obtained. Using the methodology described in this paper, a computational study of the specific present value of plutonium in a two-product nuclear power model with commercial sodium high-power fast reactor and VVER-1200 reactors was carried out. The calculation results in all considered cases indicate very significant present value of plutonium. Comparison of the obtained cost of plutonium, which is ultimately based on the energy equivalent of plutonium and uranium, and the cost of plutonium, determined on the basis of the costs of the back-end of the fuel cycle (plutonium extraction from SNF), show the economic efficiency of closing the fuel cycle even at existing uranium prices.","PeriodicalId":20426,"journal":{"name":"PROBLEMS OF ATOMIC SCIENCE AND TECHNOLOGY. SERIES: NUCLEAR AND REACTOR CONSTANTS","volume":null,"pages":null},"PeriodicalIF":0.0,"publicationDate":"2021-06-26","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"75296857","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 : 2021-03-26DOI: 10.55176/2414-1038-2021-1-48-54
O. Ashraf, G. Tikhomirov
The current work introduces a reliable safety system based on control rods in addition to the online feed system reactivity control in the Single-fluid Double-zone Thorium-based Molten Salt Reactor (SD-TMSR). The reactivity of the SD-TMSR core is controlled through two systems of control assemblies: (1) the Control Safety Devices (CSD) and (2) the Diverse Safety Devices (DSD). In the present work, the control rods are natural B4C and B4C-90 (with 90% weight content of the main absorbing 10B isotope). Since the numbers and distribution of control assemblies in SD-TMSR have not been studied previously, we proposed a unique distribution as a starting point for this analysis. The distribution of these 25 fuel assemblies with control rods in the SD-TMSR core and their numbering scheme were presented in this paper. Additionally, excess reactivity, control rod worth, and shutdown margin were calculated using Monte Carlo code Serpent2. Analysis results showed that the proposed placement of the control rods will make it possible to compensate for excess reactivity during fuel burnout and emergency shutdown of the reactor.
{"title":"NEW CONTROL AND PROTECTION SYSTEM FOR SD-TMSR","authors":"O. Ashraf, G. Tikhomirov","doi":"10.55176/2414-1038-2021-1-48-54","DOIUrl":"https://doi.org/10.55176/2414-1038-2021-1-48-54","url":null,"abstract":"The current work introduces a reliable safety system based on control rods in addition to the online feed system reactivity control in the Single-fluid Double-zone Thorium-based Molten Salt Reactor (SD-TMSR). The reactivity of the SD-TMSR core is controlled through two systems of control assemblies: (1) the Control Safety Devices (CSD) and (2) the Diverse Safety Devices (DSD). In the present work, the control rods are natural B4C and B4C-90 (with 90% weight content of the main absorbing 10B isotope). Since the numbers and distribution of control assemblies in SD-TMSR have not been studied previously, we proposed a unique distribution as a starting point for this analysis. The distribution of these 25 fuel assemblies with control rods in the SD-TMSR core and their numbering scheme were presented in this paper. Additionally, excess reactivity, control rod worth, and shutdown margin were calculated using Monte Carlo code Serpent2. Analysis results showed that the proposed placement of the control rods will make it possible to compensate for excess reactivity during fuel burnout and emergency shutdown of the reactor.","PeriodicalId":20426,"journal":{"name":"PROBLEMS OF ATOMIC SCIENCE AND TECHNOLOGY. SERIES: NUCLEAR AND REACTOR CONSTANTS","volume":null,"pages":null},"PeriodicalIF":0.0,"publicationDate":"2021-03-26","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"85589250","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 : 2021-03-26DOI: 10.55176/2414-1038-2021-1-86-96
V. Krasin, S. Soyustova
The main features of the thermodynamic evaluation of the parameters responsible for compatibility of metal materials with liquid Sn-20%Li alloy are considered in the article. Interest in the study of the physicochemical properties of liquid lithium-tin alloys is associated with the prospects for their use in plasma facing components of tokamaks. The main advantages of capillary-porous systems with a liquid metal in comparison with solid materials are their resistance to degradation of properties under tokamak conditions and the ability to self-repair the surface. Due to the fact that liquid tin is a very corrosive metal with respect to many structural materials, the advancement of liquid Li-Sn alloys is largely constrained by the lack of systematic studies of the corrosion resistance of structural materials in contact with these liquid alloys. To calculate the temperature dependences of the solubility of metals in the liquid Sn-20% Li alloy, the method of thermodynamic modeling was used, which included the following steps: (1) selection of models for the Gibbs energy functions; (2) selection and evaluation of input data; (3) optimization of model parameters; (4) calculations and comparisons. Using information on the excess Gibbs energies of mixing for the liquid phase in the form of the Redlich-Kister polynomial decomposition for the corresponding binary systems, the temperature dependences of the solubility of nickel, iron, chromium, molybdenum, and tungsten in the liquid alloy Sn-20% Li were calculated by thermodynamic modeling.
{"title":"COMPUTATIONAL AND THEORETICAL EVALUATION OF THE PARAMETERS RESPONSIBLE FOR THE COMPATIBILITY OF METALLIC MATERIALS WITH THE LIQUID SN-20% LI ALLOY","authors":"V. Krasin, S. Soyustova","doi":"10.55176/2414-1038-2021-1-86-96","DOIUrl":"https://doi.org/10.55176/2414-1038-2021-1-86-96","url":null,"abstract":"The main features of the thermodynamic evaluation of the parameters responsible for compatibility of metal materials with liquid Sn-20%Li alloy are considered in the article. Interest in the study of the physicochemical properties of liquid lithium-tin alloys is associated with the prospects for their use in plasma facing components of tokamaks. The main advantages of capillary-porous systems with a liquid metal in comparison with solid materials are their resistance to degradation of properties under tokamak conditions and the ability to self-repair the surface. Due to the fact that liquid tin is a very corrosive metal with respect to many structural materials, the advancement of liquid Li-Sn alloys is largely constrained by the lack of systematic studies of the corrosion resistance of structural materials in contact with these liquid alloys. To calculate the temperature dependences of the solubility of metals in the liquid Sn-20% Li alloy, the method of thermodynamic modeling was used, which included the following steps: (1) selection of models for the Gibbs energy functions; (2) selection and evaluation of input data; (3) optimization of model parameters; (4) calculations and comparisons. Using information on the excess Gibbs energies of mixing for the liquid phase in the form of the Redlich-Kister polynomial decomposition for the corresponding binary systems, the temperature dependences of the solubility of nickel, iron, chromium, molybdenum, and tungsten in the liquid alloy Sn-20% Li were calculated by thermodynamic modeling.","PeriodicalId":20426,"journal":{"name":"PROBLEMS OF ATOMIC SCIENCE AND TECHNOLOGY. SERIES: NUCLEAR AND REACTOR CONSTANTS","volume":null,"pages":null},"PeriodicalIF":0.0,"publicationDate":"2021-03-26","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"85185519","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 : 2021-03-26DOI: 10.55176/2414-1038-2021-1-117-123
V. Levchenko, M. Kascheev, S. Dorokhovich, A. Zaytsev
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.
{"title":"CHARACTERISTICS OF A STRAIGHT FIN WITH ENERGY RELEASE","authors":"V. Levchenko, M. Kascheev, S. Dorokhovich, A. Zaytsev","doi":"10.55176/2414-1038-2021-1-117-123","DOIUrl":"https://doi.org/10.55176/2414-1038-2021-1-117-123","url":null,"abstract":"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.","PeriodicalId":20426,"journal":{"name":"PROBLEMS OF ATOMIC SCIENCE AND TECHNOLOGY. SERIES: NUCLEAR AND REACTOR CONSTANTS","volume":null,"pages":null},"PeriodicalIF":0.0,"publicationDate":"2021-03-26","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"89528266","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 : 2021-03-26DOI: 10.55176/2414-1038-2021-1-97-107
Vladimir N. Sergeev
The paper considers methodological aspects during the development of thermo-hydro-dynamic of numerical calculation models for spent nuclear fuel pool (SNFP) on the example of Bilibino NPP by using international industry codes. The purpose of these models development is the fast simulation of thermal and humidity operational modes of spent fuel pool during periods of filling, "wet" and dry storage with an interval of up to 50 years, taking into account passive and forced convection heat removal systems. The following methodological aspects are considered in detail: 1. Use of isotope kinetics codes for calculating of the heat power dynamics for separate spent fuel assembly. 2. Calculation method of the heat removal power from the evaporation mirror of SNFP during “wet” storage, including evaporation power calculation and exhaust ventilation operation. Using the law of similarity of heat transfer and mass transfer processes (Lewis' law) for evaporation calculations. 3. Methods of accelerated computational forecasting of the dynamics of the thermal regimes of the SNFP during “wet” storage. 4. Condensation of atmospheric moisture at the bottom of the SNFP after “dry” storage and methods for its removal. It is shown that TRAC (TRACE) code with a 3D porous body model and complete evaporation-condensation models is the most suitable for solving the problems under consideration among the system thermo-hydraulic two-phase codes for nuclear energy.
{"title":"SPECIFIC FEATURES OF NUMERICAL SIMULATION OF THERMAL OPERATION MODE THE SPENT FUEL POOLS OF BILIBINO NPP","authors":"Vladimir N. Sergeev","doi":"10.55176/2414-1038-2021-1-97-107","DOIUrl":"https://doi.org/10.55176/2414-1038-2021-1-97-107","url":null,"abstract":"The paper considers methodological aspects during the development of thermo-hydro-dynamic of numerical calculation models for spent nuclear fuel pool (SNFP) on the example of Bilibino NPP by using international industry codes. The purpose of these models development is the fast simulation of thermal and humidity operational modes of spent fuel pool during periods of filling, \"wet\" and dry storage with an interval of up to 50 years, taking into account passive and forced convection heat removal systems. The following methodological aspects are considered in detail: 1. Use of isotope kinetics codes for calculating of the heat power dynamics for separate spent fuel assembly. 2. Calculation method of the heat removal power from the evaporation mirror of SNFP during “wet” storage, including evaporation power calculation and exhaust ventilation operation. Using the law of similarity of heat transfer and mass transfer processes (Lewis' law) for evaporation calculations. 3. Methods of accelerated computational forecasting of the dynamics of the thermal regimes of the SNFP during “wet” storage. 4. Condensation of atmospheric moisture at the bottom of the SNFP after “dry” storage and methods for its removal. It is shown that TRAC (TRACE) code with a 3D porous body model and complete evaporation-condensation models is the most suitable for solving the problems under consideration among the system thermo-hydraulic two-phase codes for nuclear energy.","PeriodicalId":20426,"journal":{"name":"PROBLEMS OF ATOMIC SCIENCE AND TECHNOLOGY. SERIES: NUCLEAR AND REACTOR CONSTANTS","volume":null,"pages":null},"PeriodicalIF":0.0,"publicationDate":"2021-03-26","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"77837397","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 : 2021-03-26DOI: 10.55176/2414-1038-2021-1-124-134
T. Vereshchagina, V. Lemekhov, M. Morkin
A 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 into coolant and the defective fuel assembly localization. The transit time of gaseous fission products from the defect to the place of activity measurement depends on the flow rate of the coolant through the gas-lift probe. Since most fission products have a short half-life period, their delivery time to the measuring vessel should be minimal. Therefore, the calculation of the flow rate of the coolant, as well as the transit time of gaseous fission products in the lifting path of the gas-lift probe is an actual task. A computational methodology of the hydraulics of the gas-lift probe with a lead coolant is presented in this work. The calculation results of two-phase flow characteristics in a reactor gas-lift probe and in the tested model in NIKIET experimental setup are presented. It is obtained the significant difference between the coolant flow rate in the tested model and in the reactor probe at the same gas flow rate. The reasons for these differences are defined.
{"title":"HYDRODYNAMICS OF GAS-LIFT PUMP WITH LEAD COOLANT","authors":"T. Vereshchagina, V. Lemekhov, M. Morkin","doi":"10.55176/2414-1038-2021-1-124-134","DOIUrl":"https://doi.org/10.55176/2414-1038-2021-1-124-134","url":null,"abstract":"A 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 into coolant and the defective fuel assembly localization. The transit time of gaseous fission products from the defect to the place of activity measurement depends on the flow rate of the coolant through the gas-lift probe. Since most fission products have a short half-life period, their delivery time to the measuring vessel should be minimal. Therefore, the calculation of the flow rate of the coolant, as well as the transit time of gaseous fission products in the lifting path of the gas-lift probe is an actual task. A computational methodology of the hydraulics of the gas-lift probe with a lead coolant is presented in this work. The calculation results of two-phase flow characteristics in a reactor gas-lift probe and in the tested model in NIKIET experimental setup are presented. It is obtained the significant difference between the coolant flow rate in the tested model and in the reactor probe at the same gas flow rate. The reasons for these differences are defined.","PeriodicalId":20426,"journal":{"name":"PROBLEMS OF ATOMIC SCIENCE AND TECHNOLOGY. SERIES: NUCLEAR AND REACTOR CONSTANTS","volume":null,"pages":null},"PeriodicalIF":0.0,"publicationDate":"2021-03-26","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"91014001","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 : 2021-03-26DOI: 10.55176/2414-1038-2021-1-145-151
E. Avdeev, V. Smirnova
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.
{"title":"DEVELOPMENT AND AUTOMATION OF MEASUREMENT AND REGULATION OF FLOW IN THE HEAT-RELEASING ELEMENT CELL OF FUEL ASSEMBLIES","authors":"E. Avdeev, V. Smirnova","doi":"10.55176/2414-1038-2021-1-145-151","DOIUrl":"https://doi.org/10.55176/2414-1038-2021-1-145-151","url":null,"abstract":"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.","PeriodicalId":20426,"journal":{"name":"PROBLEMS OF ATOMIC SCIENCE AND TECHNOLOGY. SERIES: NUCLEAR AND REACTOR CONSTANTS","volume":null,"pages":null},"PeriodicalIF":0.0,"publicationDate":"2021-03-26","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"80426380","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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