Pub Date : 2021-12-26DOI: 10.55176/2414-1038-2021-4-162-171
A. Lubina, A. Sorokin
The paper considered the features of heat transfer of two fast sodium reactors of large and low power (1000 and 190 MW(e.)), which are developed for use in the production of secondary nuclear fuel. The design of a case fuel assembly for a higher power reactor is a combination of thin fuel elements (6.1 mm) with U-Pu fuel and a wide grating (relative pitch 1.39) with spacing of the gratings. For a low-power reactor, in this paper, three fuel assemblies (diameter of fuel rod 8.1 mm, relative pitch 1.16) were considered: without a cover, with a cover 1 mm apart from the peripheral row of fuel rods and with a cover spaced 0,5 mm from the peripheral row of fuel rods. Calculations were carried out using the COBRA-IV-I code. Data were obtained on the azimuthal temperature distributions on the claddings of the corner, peripheral and central fuel rods, temperatures on the surfaces of the covers, and calculations were performed to optimize the designs of fuel assemblies in order to reduce the temperature difference on the claddings of the peripheral fuel rods. For a large-power reactor, optimization of the configuration of the corner cell was proposed in order to reduce the azimuthal temperature difference at the corner and peripheral fuel rods. For a low-power reactor, optimization of the fuel assembly design is recommended by replacing the spacer grids with wire spacing and equalizing the temperature field by mixing the coolant, as well as increasing the relative pitch of the fuel element grid from 1.16 to 1.19.
{"title":"FEATURES OF THERMAL HYDRAULICS OF ACTIVE ZONES OF FAST LOW-POWER AND HIGH-POWER SODIUM PRODUCTION REACTORS FOR A CLOSED FUEL CYCLE SYSTEM","authors":"A. Lubina, A. Sorokin","doi":"10.55176/2414-1038-2021-4-162-171","DOIUrl":"https://doi.org/10.55176/2414-1038-2021-4-162-171","url":null,"abstract":"The paper considered the features of heat transfer of two fast sodium reactors of large and low power (1000 and 190 MW(e.)), which are developed for use in the production of secondary nuclear fuel. The design of a case fuel assembly for a higher power reactor is a combination of thin fuel elements (6.1 mm) with U-Pu fuel and a wide grating (relative pitch 1.39) with spacing of the gratings. For a low-power reactor, in this paper, three fuel assemblies (diameter of fuel rod 8.1 mm, relative pitch 1.16) were considered: without a cover, with a cover 1 mm apart from the peripheral row of fuel rods and with a cover spaced 0,5 mm from the peripheral row of fuel rods. Calculations were carried out using the COBRA-IV-I code. Data were obtained on the azimuthal temperature distributions on the claddings of the corner, peripheral and central fuel rods, temperatures on the surfaces of the covers, and calculations were performed to optimize the designs of fuel assemblies in order to reduce the temperature difference on the claddings of the peripheral fuel rods. For a large-power reactor, optimization of the configuration of the corner cell was proposed in order to reduce the azimuthal temperature difference at the corner and peripheral fuel rods. For a low-power reactor, optimization of the fuel assembly design is recommended by replacing the spacer grids with wire spacing and equalizing the temperature field by mixing the coolant, as well as increasing the relative pitch of the fuel element grid from 1.16 to 1.19.","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":"75577468","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: 10.55176/2414-1038-2021-4-213-233
T. Vereshchagina, N. Loginov, A. Sorokin
The paper provides an overview of technical solutions for using of heat pipes in nuclear power plants both developed and operating. The review based on the scientific, technical and patent literature shows wide application heat pipes as heat transfer devices. Using of them for small and super-small power plants seems to be especially effective, because of high specific cost of plants with circulating coolants. A heat pipe is a device transferrind the heat by means of evaporation and condensation of a coolant circulating automatically under the action of capillar or gravitation forces. Heat pipes are used rather widely, both abroad and in Russia. The first application of a heat pipe principle in nuclear power plants was published in 1957, even before the emergence of the term "heat pipe". Now, there are about 300 patents in the world related to heat pipes application in nuclear power plants. Theare are seweral thouthands articles on the development of nuclear reactors with heat pipes have been published in the scientific and technical literature. One should expect that fifth-generation nuclear reactors cooled by heat pipes without any mechanisms and machines for the circulation of the coolant, as well as without the consumption of mechanical and electrical energy, will be appeared in this decade.
{"title":"HEAT PIPES IN NUCLEAR ENGINEERING","authors":"T. Vereshchagina, N. Loginov, A. Sorokin","doi":"10.55176/2414-1038-2021-4-213-233","DOIUrl":"https://doi.org/10.55176/2414-1038-2021-4-213-233","url":null,"abstract":"The paper provides an overview of technical solutions for using of heat pipes in nuclear power plants both developed and operating. The review based on the scientific, technical and patent literature shows wide application heat pipes as heat transfer devices. Using of them for small and super-small power plants seems to be especially effective, because of high specific cost of plants with circulating coolants. A heat pipe is a device transferrind the heat by means of evaporation and condensation of a coolant circulating automatically under the action of capillar or gravitation forces. Heat pipes are used rather widely, both abroad and in Russia. The first application of a heat pipe principle in nuclear power plants was published in 1957, even before the emergence of the term \"heat pipe\". Now, there are about 300 patents in the world related to heat pipes application in nuclear power plants. Theare are seweral thouthands articles on the development of nuclear reactors with heat pipes have been published in the scientific and technical literature. One should expect that fifth-generation nuclear reactors cooled by heat pipes without any mechanisms and machines for the circulation of the coolant, as well as without the consumption of mechanical and electrical energy, will be appeared in this decade.","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":"76155305","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: 10.55176/2414-1038-2021-4-121-130
A. Shlepkin, A. Morozov, A. Sorokin
The article presents an overview of works on the experimental and computational study of the processes of heat removal from heated surfaces using a water-air mixture. A sharp increase in the heat transfer coefficient is shown even when adding water with a mass content of 0.1 % to the air flow. The factors that determine the efficiency of the heat exchange process are listed: the shape of the water spray jet, the distance from the outlet point of the water-air flow to the heat exchange surface, the characteristics of the heat exchange surface, the method of jet formation, the size of droplets and the location of the outlet points of the gas-droplet flow. The weak applicability of the data available in the literature is shown for calculating the cooling of finned tubes of industrial heat exchangers using a water-air mixture. It is substantiated that in order to establish the most optimal cooling modes for each for a separate heat exchange surface, it is necessary to conduct experimental studies, due to the complexity of heat transfer processes and the presence of a large number of influencing factors. It is shown that the finned tubes of heat exchangers of the passive heat removal system of WWER-1200 have a number of important features that affect the efficiency of their cooling using a water-air mixture. An experimental setup has been developed and a technique has been proposed for performing experiments to study these processes as applied to heat exchangers of a passive WWER safety system.
{"title":"APPLICATION OF AEROSOL WATER SPRAY TO INCREASE THE EFFICIENCY OF COOLING OF FINNED TUBES","authors":"A. Shlepkin, A. Morozov, A. Sorokin","doi":"10.55176/2414-1038-2021-4-121-130","DOIUrl":"https://doi.org/10.55176/2414-1038-2021-4-121-130","url":null,"abstract":"The article presents an overview of works on the experimental and computational study of the processes of heat removal from heated surfaces using a water-air mixture. A sharp increase in the heat transfer coefficient is shown even when adding water with a mass content of 0.1 % to the air flow. The factors that determine the efficiency of the heat exchange process are listed: the shape of the water spray jet, the distance from the outlet point of the water-air flow to the heat exchange surface, the characteristics of the heat exchange surface, the method of jet formation, the size of droplets and the location of the outlet points of the gas-droplet flow. The weak applicability of the data available in the literature is shown for calculating the cooling of finned tubes of industrial heat exchangers using a water-air mixture. It is substantiated that in order to establish the most optimal cooling modes for each for a separate heat exchange surface, it is necessary to conduct experimental studies, due to the complexity of heat transfer processes and the presence of a large number of influencing factors. It is shown that the finned tubes of heat exchangers of the passive heat removal system of WWER-1200 have a number of important features that affect the efficiency of their cooling using a water-air mixture. An experimental setup has been developed and a technique has been proposed for performing experiments to study these processes as applied to heat exchangers of a passive WWER safety system.","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":"91101512","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: 10.55176/2414-1038-2021-4-47-62
A. Belov, M. Kryachko, O. Chertovskikh, A. Ivanov
The calculation of activation inventories is one of the key aspects in nuclear plants simulation. Burnup calculations provide information not only about fuel actinides transmutation but also about impact of fuel/constructor materials/coolant impurities on the material. The inventory evolution also determines radiological response of a material by nuclides production/decay rates quantification. Results of the isotopic kinetics code BPSD for activity and afterheat calculation validation by means of direct comparison afterheat obtained by code BPSD with experimental data obtained due to steel SS-304 and steel SS-316 samples irradiation by fusion neutron spectra at the FNS facility and with results of FISPACT-II calculation are presented in this paper. BPSD code is intended for carrying of transmutation calculations of materials in fast neutron fluxes. Nuclides transformation chains are based on ROSFOND and ABBN-RF data. BPSD and FISPACT-II calculations results conform to each other. Differences in results could be explained by differences between transformation chains and by using different evaluation neutron data bases.
{"title":"CALCULATION OF THE STRUCTURAL MATERIALS ACTIVATION BY A FUSION NEUTRON FLUX WITH BPSD CODE","authors":"A. Belov, M. Kryachko, O. Chertovskikh, A. Ivanov","doi":"10.55176/2414-1038-2021-4-47-62","DOIUrl":"https://doi.org/10.55176/2414-1038-2021-4-47-62","url":null,"abstract":"The calculation of activation inventories is one of the key aspects in nuclear plants simulation. Burnup calculations provide information not only about fuel actinides transmutation but also about impact of fuel/constructor materials/coolant impurities on the material. The inventory evolution also determines radiological response of a material by nuclides production/decay rates quantification. Results of the isotopic kinetics code BPSD for activity and afterheat calculation validation by means of direct comparison afterheat obtained by code BPSD with experimental data obtained due to steel SS-304 and steel SS-316 samples irradiation by fusion neutron spectra at the FNS facility and with results of FISPACT-II calculation are presented in this paper. BPSD code is intended for carrying of transmutation calculations of materials in fast neutron fluxes. Nuclides transformation chains are based on ROSFOND and ABBN-RF data. BPSD and FISPACT-II calculations results conform to each other. Differences in results could be explained by differences between transformation chains and by using different evaluation neutron data bases.","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":"75750821","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: 10.55176/2414-1038-2021-4-147-153
A. Osipov, K. Ivanov, M. Arnol’dov
At present, the problem of providing corrosion resistance of steels in heavy liquid metal coolants is solved by forming and maintaining protective oxide films on the surfaces of structural steels. However, oxide films are not absolutely impermeable barriers for the components of steels (primarily iron), which, as a result of diffusion processes, inevitably enter the coolant and can interact with oxygen to form solid-phase oxides. Within the framework of this work, a method is considered for obtaining the numerical values of the quantities characterizing the processes of dissociation of iron oxides as a function of temperature and oxygen potential of HLMC. The performed calculations made it possible to obtain specific numerical values of the limiting solubilities of iron oxides as a function of the temperature and the initial state of the coolant with respect to iron and oxygen impurities, and also made it possible to obtain numerical values of other thermodynamic parameters characterizing the current and limiting state of HLMC. The considered approach is of a general nature and can be used in experimental studies of the kinetic and thermodynamic characteristics of the dissociation processes of compounds in liquid metals.
{"title":"DISSOCIATION OF IRON OXIDES IN MELTS OF HEAVY METALS","authors":"A. Osipov, K. Ivanov, M. Arnol’dov","doi":"10.55176/2414-1038-2021-4-147-153","DOIUrl":"https://doi.org/10.55176/2414-1038-2021-4-147-153","url":null,"abstract":"At present, the problem of providing corrosion resistance of steels in heavy liquid metal coolants is solved by forming and maintaining protective oxide films on the surfaces of structural steels. However, oxide films are not absolutely impermeable barriers for the components of steels (primarily iron), which, as a result of diffusion processes, inevitably enter the coolant and can interact with oxygen to form solid-phase oxides. Within the framework of this work, a method is considered for obtaining the numerical values of the quantities characterizing the processes of dissociation of iron oxides as a function of temperature and oxygen potential of HLMC. The performed calculations made it possible to obtain specific numerical values of the limiting solubilities of iron oxides as a function of the temperature and the initial state of the coolant with respect to iron and oxygen impurities, and also made it possible to obtain numerical values of other thermodynamic parameters characterizing the current and limiting state of HLMC. The considered approach is of a general nature and can be used in experimental studies of the kinetic and thermodynamic characteristics of the dissociation processes of compounds in liquid metals.","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":"84136281","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: 10.55176/2414-1038-2021-4-203-212
N. Matyukhin, A. Sorokin, N. Denisova, M. Kascheev
As a result of the processing and analysis of experimental data on natural convection of coolants for various forms of heat-transfer surfaces of in-reactor structures, a relation for the velocity profile along the normal to the heat exchange surface under conditions of natural turbulent convection in the coolant flow was proposed. By the example of a vertical isothermal surface, it is shown that for different values of the longitudinal coordinate there is a characteristic point at which the longitudinal velocity reaches its maximum value (umax) at the corresponding value of the transverse coordinate (ymax). This point divides the fluid flow along the heat exchange surface into two zones: an internal one adjacent to the wall and an external one located beyond the maximum value of the velocity. Taking for the characteristic scale umax and ymax and presenting experimental data in dimensionless form u/umax = f(y/ymax), generalization of experimental data obtained. The results of processing the experimental data of various authors are well generalized by the dependences obtained in the work for the velocity profile in the inner and outer zones of the coolant flow. The velocity profile at the horizontal, downward-facing heat transfer surface is characterized by the same regularities as for the vertical surface: the velocity in the near-wall region increases, reaches a maximum and then decreases. The results of processing the measured velocity profile for a horizontal cylinder according to the method proposed in the work show agreement with the generalized data for a vertical isothermal surface. The data of experimental studies of the velocity field around an isothermally heated sphere with free convection in water are also well generalized by the dependences proposed in this work. The analysis of the experimental data on the velocity profiles for various forms of heat-transfer surfaces under conditions of natural turbulent convection, carried out by the authors, shows that the velocity profile proposed by the authors along the normal to the heat exchange surface has a universal character.
{"title":"VELOCITY PROFILE UNDER THE CONDITIONS OF NATURAL TURBULENT CONVECTION ACCORDING TO THE HEAT SEPARATING SURFACES OF THE INTERNAL REACTOR DEVICES","authors":"N. Matyukhin, A. Sorokin, N. Denisova, M. Kascheev","doi":"10.55176/2414-1038-2021-4-203-212","DOIUrl":"https://doi.org/10.55176/2414-1038-2021-4-203-212","url":null,"abstract":"As a result of the processing and analysis of experimental data on natural convection of coolants for various forms of heat-transfer surfaces of in-reactor structures, a relation for the velocity profile along the normal to the heat exchange surface under conditions of natural turbulent convection in the coolant flow was proposed. By the example of a vertical isothermal surface, it is shown that for different values of the longitudinal coordinate there is a characteristic point at which the longitudinal velocity reaches its maximum value (umax) at the corresponding value of the transverse coordinate (ymax). This point divides the fluid flow along the heat exchange surface into two zones: an internal one adjacent to the wall and an external one located beyond the maximum value of the velocity. Taking for the characteristic scale umax and ymax and presenting experimental data in dimensionless form u/umax = f(y/ymax), generalization of experimental data obtained. The results of processing the experimental data of various authors are well generalized by the dependences obtained in the work for the velocity profile in the inner and outer zones of the coolant flow. The velocity profile at the horizontal, downward-facing heat transfer surface is characterized by the same regularities as for the vertical surface: the velocity in the near-wall region increases, reaches a maximum and then decreases. The results of processing the measured velocity profile for a horizontal cylinder according to the method proposed in the work show agreement with the generalized data for a vertical isothermal surface. The data of experimental studies of the velocity field around an isothermally heated sphere with free convection in water are also well generalized by the dependences proposed in this work. The analysis of the experimental data on the velocity profiles for various forms of heat-transfer surfaces under conditions of natural turbulent convection, carried out by the authors, shows that the velocity profile proposed by the authors along the normal to the heat exchange surface has a universal character.","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":"87597539","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: 10.55176/2414-1038-2021-4-195-202
I. Molotova, A. Zabirov, M. Vinogradov, V. Yagov, A. Sorokin
The process of cooling high-temperature bodies in liquids is an important physical process, in particular, for the safety of nuclear plants. After the accident at the Fukushima-1 nuclear power plant in 2011, large-scale research was launched to find a new accident tolerant fuel. Studying the effect of the properties of new materials on heat transfer during cooling in the case of repeated flooding of the active zone and the possibility of accurately predicting the transition temperature to the intensive cooling regime will allow substantiating the choice of a new type of tolerant fuel from the standpoint of thermophysics. The aim of this work was an experimental study of the cooling processes of high-temperature cylindrical bodies made of various metals in liquids, as well as determining the effect of coating properties on heat transfer during cooling. A large experimental data array was obtained on the cooling of cylindrical samples; experiments were carried out on smooth cylinders made of various metals, as well as on copper cylinders with various coatings and different degrees of roughness (gold coating, stainless steel coatings).
{"title":"EFFECT OF COATING PROPERTIES ON HEAT TRANSFER DURING COOLING OF HIGH-TEMPERATURE CYLINDRICAL BODIES","authors":"I. Molotova, A. Zabirov, M. Vinogradov, V. Yagov, A. Sorokin","doi":"10.55176/2414-1038-2021-4-195-202","DOIUrl":"https://doi.org/10.55176/2414-1038-2021-4-195-202","url":null,"abstract":"The process of cooling high-temperature bodies in liquids is an important physical process, in particular, for the safety of nuclear plants. After the accident at the Fukushima-1 nuclear power plant in 2011, large-scale research was launched to find a new accident tolerant fuel. Studying the effect of the properties of new materials on heat transfer during cooling in the case of repeated flooding of the active zone and the possibility of accurately predicting the transition temperature to the intensive cooling regime will allow substantiating the choice of a new type of tolerant fuel from the standpoint of thermophysics. The aim of this work was an experimental study of the cooling processes of high-temperature cylindrical bodies made of various metals in liquids, as well as determining the effect of coating properties on heat transfer during cooling. A large experimental data array was obtained on the cooling of cylindrical samples; experiments were carried out on smooth cylinders made of various metals, as well as on copper cylinders with various coatings and different degrees of roughness (gold coating, stainless steel coatings).","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":"84254219","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: 10.55176/2414-1038-2021-4-172-194
Yu. A. Kuzina, D. Klinov, G. Mikhailov, A. Sorokin, V. Alekseev
To substantiate the safety and characteristics of fast reactors with liquid metal coolants, a complex of more than 20 stands of various profiles and purposes, well equipped with modern measuring instruments, including hydrodynamic, thermohydraulic and technological stands, has been created at SSC RF - IPPE. In addition, JSC “SSC RF - IPPE” has a complex of fast physical stands, including two critical stands - BFS-1 and the world's largest physical stand BFS-2. The article presents the characteristics and the possibility of stands designed for research in the field of hydrodynamics, heat transfer and coolant technology in support of design solutions, safety improvement and testing of equipment elements and assemblies of operating and planned installations with fast reactors with sodium, lead and lead-bismuth coolants, as well as for accelerator-controlled systems and thermonuclear fusion, low-power nuclear power plants for space: - Hydrodynamic stands - “SGDI” (air), “V-2” (air), “SGI” (water), “V-200” (water), “GDK” (air). - Thermal-hydraulic liquid metal stands - “6B” (Na, Na-K), “AR-1” (Na, Na-K), “Pluton” (Na), “SPRUT” (Na, Na-K, Pb, Pb-Bi, water). - Technological liquid metal stands - “Protva-1” (Na), “Protva-2” (Na), “PUSHM” (Na), “Armatura” (Na), “IK-MT” (Na), “SID” (Na), “BTS” (Na), “TT-1M” (Pb), “TT-2M” (Pb-Bi), “LIS-M” (Li). A large-scale sodium test stands “SAZ” is under construction, which allows testing full-scale prototypes of equipment and its elements to substantiate existing and future projects of fast sodium reactors. The BFS complex of physical stands is the world's only experimental tool for full-scale modeling of the cores of nuclear reactors of various types (of any composition, geometry and configuration). The materials and construction of the stands allow simulating the core, breeding zones, reflectors and in-core shielding, as well as elements of fuel cycles and storage facilities for spent nuclear fuel and radioactive waste. Reactor materials of the stands (metallic plutonium, oxide and metallic highly enriched uranium with enrichment of 36% and 90% in uranium-235, hundreds of tons of fertile materials, construction materials, various coolants) make it possible to assemble both complex full-scale models of fast reactors, and benchmarks, experiments for which are carried out to correct neutron-physical constants and improve computational methods.
{"title":"COMPLEX OF EXPERIMENTAL FACILITIES FOR DESIGN AND SAFETY JUSTIFICATION OF FAST REACTORS WITH LIQUID METAL COOLANTS","authors":"Yu. A. Kuzina, D. Klinov, G. Mikhailov, A. Sorokin, V. Alekseev","doi":"10.55176/2414-1038-2021-4-172-194","DOIUrl":"https://doi.org/10.55176/2414-1038-2021-4-172-194","url":null,"abstract":"To substantiate the safety and characteristics of fast reactors with liquid metal coolants, a complex of more than 20 stands of various profiles and purposes, well equipped with modern measuring instruments, including hydrodynamic, thermohydraulic and technological stands, has been created at SSC RF - IPPE. In addition, JSC “SSC RF - IPPE” has a complex of fast physical stands, including two critical stands - BFS-1 and the world's largest physical stand BFS-2. The article presents the characteristics and the possibility of stands designed for research in the field of hydrodynamics, heat transfer and coolant technology in support of design solutions, safety improvement and testing of equipment elements and assemblies of operating and planned installations with fast reactors with sodium, lead and lead-bismuth coolants, as well as for accelerator-controlled systems and thermonuclear fusion, low-power nuclear power plants for space: - Hydrodynamic stands - “SGDI” (air), “V-2” (air), “SGI” (water), “V-200” (water), “GDK” (air). - Thermal-hydraulic liquid metal stands - “6B” (Na, Na-K), “AR-1” (Na, Na-K), “Pluton” (Na), “SPRUT” (Na, Na-K, Pb, Pb-Bi, water). - Technological liquid metal stands - “Protva-1” (Na), “Protva-2” (Na), “PUSHM” (Na), “Armatura” (Na), “IK-MT” (Na), “SID” (Na), “BTS” (Na), “TT-1M” (Pb), “TT-2M” (Pb-Bi), “LIS-M” (Li). A large-scale sodium test stands “SAZ” is under construction, which allows testing full-scale prototypes of equipment and its elements to substantiate existing and future projects of fast sodium reactors. The BFS complex of physical stands is the world's only experimental tool for full-scale modeling of the cores of nuclear reactors of various types (of any composition, geometry and configuration). The materials and construction of the stands allow simulating the core, breeding zones, reflectors and in-core shielding, as well as elements of fuel cycles and storage facilities for spent nuclear fuel and radioactive waste. Reactor materials of the stands (metallic plutonium, oxide and metallic highly enriched uranium with enrichment of 36% and 90% in uranium-235, hundreds of tons of fertile materials, construction materials, various coolants) make it possible to assemble both complex full-scale models of fast reactors, and benchmarks, experiments for which are carried out to correct neutron-physical constants and improve computational methods.","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":"77560159","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: 10.55176/2414-1038-2021-4-28-34
I. Kurina, M. Frolova, E. Chesnokov, O. Plaksin
It is known that the thermal shock resistance of ceramic crucibles is insufficient for melting uranium alloys in them. Typically, crucibles withstand one or more heats and then break down. The possibility of using nanostructured ceramic crucibles based on ZrO2-MgO-CaO to obtain a U-10 % Zr alloy in an induction electric furnace has been substantiated at IPPE. Crucibles were made in JSC “ONPP “Tekhnologiya” named after A.G. Romashin” from a mixture of zirconium dioxide powders, partially stabilized with nanocrystalline CaO and MgO in a ratio of 30 and 70 wt. %, using two-stage sintering. Such crucibles have a sufficiently high resistance to thermal shock in contact with melts of metals and alloys and withstand 1-3 melts without destruction. Crucibles with various densities (from 5.206 to 5.29 g/cm3) and porosity (from 5 to 9 %) were tested under conditions of heating the melt at a rate of 12 to 19 °C/min to a maximum temperature of 1455 to 1560 °C. The tested crucible ZrO2-MgO-CaO was inserted into a graphite crucible to prevent leakage of the melt in the electric furnace in case of destruction of the ceramic crucible. There was no complete destruction of the crucibles; some crucibles with small cracks could be reused. The best result (three melts) was obtained when using such a crucible with a porosity of about 5 % for melting a charge containing uranium and zirconium. In order to determine the degree of interaction of a uranium-zirconium melt with a crucible based on ZrO2-MgO-CaO, the microstructure and microhardness of the crucible surface in contact with the melt were studied at an elevated temperature of 1600 °C. The surface of all tested crucibles is not wetted by liquid uranium-zirconium melt, and there is no chemical interaction. Nanostructured ceramic crucibles based on ZrO2-MgO-CaO are suitable for melting uranium-containing materials in an electric induction furnace.
{"title":"POSSIBILITY OF APPLICATION OF ZrO2-MgO-CaO CRUCIBLES FOR PRODUCING ALLOY U-10 % Zr","authors":"I. Kurina, M. Frolova, E. Chesnokov, O. Plaksin","doi":"10.55176/2414-1038-2021-4-28-34","DOIUrl":"https://doi.org/10.55176/2414-1038-2021-4-28-34","url":null,"abstract":"It is known that the thermal shock resistance of ceramic crucibles is insufficient for melting uranium alloys in them. Typically, crucibles withstand one or more heats and then break down. The possibility of using nanostructured ceramic crucibles based on ZrO2-MgO-CaO to obtain a U-10 % Zr alloy in an induction electric furnace has been substantiated at IPPE. Crucibles were made in JSC “ONPP “Tekhnologiya” named after A.G. Romashin” from a mixture of zirconium dioxide powders, partially stabilized with nanocrystalline CaO and MgO in a ratio of 30 and 70 wt. %, using two-stage sintering. Such crucibles have a sufficiently high resistance to thermal shock in contact with melts of metals and alloys and withstand 1-3 melts without destruction. Crucibles with various densities (from 5.206 to 5.29 g/cm3) and porosity (from 5 to 9 %) were tested under conditions of heating the melt at a rate of 12 to 19 °C/min to a maximum temperature of 1455 to 1560 °C. The tested crucible ZrO2-MgO-CaO was inserted into a graphite crucible to prevent leakage of the melt in the electric furnace in case of destruction of the ceramic crucible. There was no complete destruction of the crucibles; some crucibles with small cracks could be reused. The best result (three melts) was obtained when using such a crucible with a porosity of about 5 % for melting a charge containing uranium and zirconium. In order to determine the degree of interaction of a uranium-zirconium melt with a crucible based on ZrO2-MgO-CaO, the microstructure and microhardness of the crucible surface in contact with the melt were studied at an elevated temperature of 1600 °C. The surface of all tested crucibles is not wetted by liquid uranium-zirconium melt, and there is no chemical interaction. Nanostructured ceramic crucibles based on ZrO2-MgO-CaO are suitable for melting uranium-containing materials in an electric induction furnace.","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":"73851038","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: 10.55176/2414-1038-2021-4-63-71
V. Trofimchuk, V. Nasonov, Y. Pesnya, K. Glyva, Yu Matveev
The paper considers the possibility of achieving the required neutron-physical parameters during irradiation of experimental fuel elements circumfluent with a SCW parameters coolant in the RR IR-8 in order to study the behavior of fuel element cladding under the influence of reactor radiation and coolant with SCW parameters under conditions closest to real operation in nuclear power plant. The possibility of carrying out such experiments in the RR IR-8 will make it possible to contribute to the development of new generation IV nuclear power plants, one of the areas of which is a reactor with super-critical parameters coolant. This direction is promising, as the use of the SCW coolant technology will increase the efficiency of the reactors by 10-12 %. Irradiation experiments in RR IR-8 are carried out using experimental ampoule rigs. During developing the design of an ampoule rig with experimental fuel elements, it is required to provide the necessary heat transfer from the fuel element to the water of the reactor pool, and to maintain the pressure of the SCW coolant by the structure of the ampoule rig vessel. The calculations of heat release in the experimental fuel elements and the accumulation of fluence in the cladding were carried out using the MCU-PTR code with the MDBPT-50 database, which implements the Monte Carlo method. The assessment of heat release in fuel elements during irradiation of an ampoule rig in the cells of the IR-8 core and reflector was carried out. The calculation results of heat release in experimental fuel elements and fast neutron fluxes in the cladding during irradiation in the central channels of fuel assemblies and special blocks made of aluminum, stainless steel, as well as water are presented. An assessment of the contribution of heat release from gamma heating of the ampoule rig structure elements was carried out, which showed that the amount of heat release can be comparable with the energy release due to the fuel fission reaction in an experimental fuel element, which must be taken into account when calculating the temperature irradiation regimes of experimental fuel elements. Obtained results showed the principle possibility of achievement the parameters of the SCW coolant in the ampoule rig on the outer surface of the experimental fuel elements cladding in the RR IR 8.
为了在最接近核电站实际运行的条件下研究燃料元件包壳在反应堆辐射和超临界水参数冷却剂的影响下的行为，考虑了在RR -8反应堆中使用超临界水参数冷却剂进行实验燃料元件循环辐照时达到所需中子物理参数的可能性。在RR IR-8上进行这种实验的可能性将有可能为新一代核电站的发展做出贡献，其中一个领域是使用超临界参数冷却剂的反应堆。这个方向是有希望的，因为使用SCW冷却剂技术将使反应堆的效率提高10- 12%。利用实验安瓿台进行了RR - IR-8的辐照实验。在开发装有实验燃料元件的安瓿装置的设计过程中，需要提供从燃料元件到反应堆池水的必要传热，并通过安瓿装置容器的结构来保持SCW冷却剂的压力。采用MDBPT-50数据库，采用蒙特卡罗方法，利用MCU-PTR代码对实验燃料元件的放热和包壳内能量积累进行了计算。对IR-8堆芯和反射器电池的安瓿装置辐照过程中燃料元件的放热进行了评估。给出了实验燃料元件放热和包层快中子通量在燃料组件中心通道和铝、不锈钢和水特制块体辐照过程中的计算结果。对安瓿钻机结构元件伽玛加热释放的热量贡献进行了评估，结果表明其释放的热量与实验燃料元件中燃料裂变反应释放的能量相当，在计算实验燃料元件的温度辐照制度时必须考虑到这一点。得到的结果表明，在RR - IR - 8试验燃料元件包壳的外表面上实现安瓿装置中SCW冷却剂参数的基本可能性。
{"title":"CALCULATION ESTIMATION OF THE REQUIRED HEAT GENERATION IN TESTING FUEL ELEMENTS TO ACHIEVE SUPERCRITICAL PARAMETERS OF THE COOLANT UNDER IRRADIATION IN THE RESEARCH NUCLEAR REACTOR","authors":"V. Trofimchuk, V. Nasonov, Y. Pesnya, K. Glyva, Yu Matveev","doi":"10.55176/2414-1038-2021-4-63-71","DOIUrl":"https://doi.org/10.55176/2414-1038-2021-4-63-71","url":null,"abstract":"The paper considers the possibility of achieving the required neutron-physical parameters during irradiation of experimental fuel elements circumfluent with a SCW parameters coolant in the RR IR-8 in order to study the behavior of fuel element cladding under the influence of reactor radiation and coolant with SCW parameters under conditions closest to real operation in nuclear power plant. The possibility of carrying out such experiments in the RR IR-8 will make it possible to contribute to the development of new generation IV nuclear power plants, one of the areas of which is a reactor with super-critical parameters coolant. This direction is promising, as the use of the SCW coolant technology will increase the efficiency of the reactors by 10-12 %. Irradiation experiments in RR IR-8 are carried out using experimental ampoule rigs. During developing the design of an ampoule rig with experimental fuel elements, it is required to provide the necessary heat transfer from the fuel element to the water of the reactor pool, and to maintain the pressure of the SCW coolant by the structure of the ampoule rig vessel. The calculations of heat release in the experimental fuel elements and the accumulation of fluence in the cladding were carried out using the MCU-PTR code with the MDBPT-50 database, which implements the Monte Carlo method. The assessment of heat release in fuel elements during irradiation of an ampoule rig in the cells of the IR-8 core and reflector was carried out. The calculation results of heat release in experimental fuel elements and fast neutron fluxes in the cladding during irradiation in the central channels of fuel assemblies and special blocks made of aluminum, stainless steel, as well as water are presented. An assessment of the contribution of heat release from gamma heating of the ampoule rig structure elements was carried out, which showed that the amount of heat release can be comparable with the energy release due to the fuel fission reaction in an experimental fuel element, which must be taken into account when calculating the temperature irradiation regimes of experimental fuel elements. Obtained results showed the principle possibility of achievement the parameters of the SCW coolant in the ampoule rig on the outer surface of the experimental fuel elements cladding in the RR IR 8.","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":"74542420","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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