Pub Date : 2021-06-26DOI: 10.55176/2414-1038-2021-2-5-24
G. Manturov, M. Nikolaev, V. Koshcheev
The work is devoted to one of the most important scientific and technical problems in reactor physics related to the development and verification of codes and nuclear data that provide reliable and highly accurate calculations of neutron-physical characteristics of fast reactors and radiation shielding, including nuclear fuel cycle, criticality and radiation safety parameters. The designed neutronics characteristics of fast reactors should be based on certified, qualified sets of codes and nuclear constants: the calculation tools should be related to the modern state of scientific knowledge and computational techniques, and used nuclear physics constants should be adequate to the most reliable evaluations of nuclear data. In connection with rapid development of the computing engineering and all greater introduction in practice of calculation Monte Carlo codes, the methodical constituent of calculation error falls down substantially. In these terms the nuclear constant’s constituent of error of calculations becomes fully qualificatory. A situation is intensifyed by the fall-off of financing of experimental works, why in this connection the amount of fast critical stands in the world diminishes sharply. The paper consider the state of art of the constant’s providing system CONSYST/ABBN, created on the basis of the national library of neutron data files ROSFOND and libraries of multigroup constants ABBN-93 and ABBN-RF. One of the most important problems under consideration here also is the methodical and software for estimating of errors of the calculated physical characteristics. System of codes and nuclear data for reactor neutronics calculations is based on the unified methodological basis that ensures the transparency of the procedure for obtaining the data used in the calculations, the reliability of their verification and the obtaining of guaranteed accuracy of the calculated physical reactor characteristics.
{"title":"NUCLEAR DATA FOR REACTOR NEUTRONICS CALCULATIONS - ROSFOND DATA LIBRARY AND ABBN-RF GROUP DATA SYSTEM","authors":"G. Manturov, M. Nikolaev, V. Koshcheev","doi":"10.55176/2414-1038-2021-2-5-24","DOIUrl":"https://doi.org/10.55176/2414-1038-2021-2-5-24","url":null,"abstract":"The work is devoted to one of the most important scientific and technical problems in reactor physics related to the development and verification of codes and nuclear data that provide reliable and highly accurate calculations of neutron-physical characteristics of fast reactors and radiation shielding, including nuclear fuel cycle, criticality and radiation safety parameters. The designed neutronics characteristics of fast reactors should be based on certified, qualified sets of codes and nuclear constants: the calculation tools should be related to the modern state of scientific knowledge and computational techniques, and used nuclear physics constants should be adequate to the most reliable evaluations of nuclear data. In connection with rapid development of the computing engineering and all greater introduction in practice of calculation Monte Carlo codes, the methodical constituent of calculation error falls down substantially. In these terms the nuclear constant’s constituent of error of calculations becomes fully qualificatory. A situation is intensifyed by the fall-off of financing of experimental works, why in this connection the amount of fast critical stands in the world diminishes sharply. The paper consider the state of art of the constant’s providing system CONSYST/ABBN, created on the basis of the national library of neutron data files ROSFOND and libraries of multigroup constants ABBN-93 and ABBN-RF. One of the most important problems under consideration here also is the methodical and software for estimating of errors of the calculated physical characteristics. System of codes and nuclear data for reactor neutronics calculations is based on the unified methodological basis that ensures the transparency of the procedure for obtaining the data used in the calculations, the reliability of their verification and the obtaining of guaranteed accuracy of the calculated physical reactor characteristics.","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":"86825477","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: 10.55176/2414-1038-2021-2-256-267
V. Gribacheva, S. Shcherbakov
The paper proposes the configuration and composition equipment of the cooling system of the designed periodic pulsed reactor (PPR) of high power. The special features of the PPR are a small flow section, a large heating of the coolant in the power pulse and the impossibility of useful use of thermal energy in the periodic mode of operation. Liquid lithium is proposed as a coolant and heat is discharged through air heat exchangers (AHE). The goal was to achieve compactness and low power consumption, the ability to work with frequent stops and optimize the operation of equipment in pulse modes. For this purpose, high-temperature AHE with a small heat exchange surface and forced air cooling are used, the circulation circuit is divided into two parts - the reactor circuit and the AHE circuit with two circulation pumps and a common drain tank. The separation of the circuit allows to independently perform the operations of starting, stopping and heating the circuits in a periodic mode. The drain tank limits the composition of the equipment exposed to temperature pulses. Numerical studies of the temperature regime of the coolant in the equipment of the PDR cooling system are carried out. The calculations were performed using the TURBOFLOW code in two-dimensional terms for all the main elements of the equipment. Quasi-stationary (nominal and partial power levels) and pulse modes of operation are considered. Calculated characteristics for forced and natural air circulation are obtained. The limits of the air circulation modes under the conditions of non-freezing of the coolant are determined. The obtained values of the maximum temperatures of the coolant: in the pulsed mode is 750 °C, in the quasi-stationary mode - 490 °C with an average power of 15 MW, air flow of 150 m3/s and the size of the AHE in the plan of 5×5 m, 100 panels of 1.08×0.025×5 m.
介绍了所设计的大功率周期脉冲堆(PPR)冷却系统的结构和组成设备。PPR的特点是流量小，在功率脉冲中冷却剂的热量大，并且在周期性运行模式下不可能有效地利用热能。液态锂作为冷却剂，热量通过空气热交换器(AHE)排出。目标是实现紧凑性和低功耗，能够频繁停止工作，并优化脉冲模式下设备的操作。为此，采用换热面小、强制风冷的高温AHE，循环回路分为反应器回路和AHE回路两部分，由两台循环泵和一个公共排水槽组成。电路的分离允许以周期性模式独立执行启动，停止和加热电路的操作。排液槽限制了暴露在温度脉冲下的设备的组成。对PDR冷却系统中冷却剂的温度状态进行了数值研究。使用TURBOFLOW代码对设备的所有主要部件进行二维计算。准平稳(标称和部分功率电平)和脉冲工作模式被考虑。得到了强制空气循环和自然空气循环的计算特性。确定了冷却剂不冻结条件下空气循环方式的极限。得到的冷却剂最高温度值:脉冲模式下为750℃，准平稳模式下为- 490℃，平均功率为15 MW，风量为150 m3/s, AHE尺寸为5×5 m, 100块面板为1.08×0.025×5 m。
{"title":"THE COOLING SYSTEM OF THE PERIODIC PULSED REACTOR","authors":"V. Gribacheva, S. Shcherbakov","doi":"10.55176/2414-1038-2021-2-256-267","DOIUrl":"https://doi.org/10.55176/2414-1038-2021-2-256-267","url":null,"abstract":"The paper proposes the configuration and composition equipment of the cooling system of the designed periodic pulsed reactor (PPR) of high power. The special features of the PPR are a small flow section, a large heating of the coolant in the power pulse and the impossibility of useful use of thermal energy in the periodic mode of operation. Liquid lithium is proposed as a coolant and heat is discharged through air heat exchangers (AHE). The goal was to achieve compactness and low power consumption, the ability to work with frequent stops and optimize the operation of equipment in pulse modes. For this purpose, high-temperature AHE with a small heat exchange surface and forced air cooling are used, the circulation circuit is divided into two parts - the reactor circuit and the AHE circuit with two circulation pumps and a common drain tank. The separation of the circuit allows to independently perform the operations of starting, stopping and heating the circuits in a periodic mode. The drain tank limits the composition of the equipment exposed to temperature pulses. Numerical studies of the temperature regime of the coolant in the equipment of the PDR cooling system are carried out. The calculations were performed using the TURBOFLOW code in two-dimensional terms for all the main elements of the equipment. Quasi-stationary (nominal and partial power levels) and pulse modes of operation are considered. Calculated characteristics for forced and natural air circulation are obtained. The limits of the air circulation modes under the conditions of non-freezing of the coolant are determined. The obtained values of the maximum temperatures of the coolant: in the pulsed mode is 750 °C, in the quasi-stationary mode - 490 °C with an average power of 15 MW, air flow of 150 m3/s and the size of the AHE in the plan of 5×5 m, 100 panels of 1.08×0.025×5 m.","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":"88922312","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: 10.55176/2414-1038-2021-2-105-115
R. Askhadullin, A. Legkikh, V. Ulyanov, I. Voronin
The technology of heavy liquid metal coolant (HLMC) is an important part of the safety system for the operation of reactor facilities with HLMC at all stages of their life cycle: from the preparation of a coolant and loading into the reactor to the decommissioning of the reactor facility (RF). The technology of heavy liquid metal coolant is a set of measures that allow: - to prepare the coolant for filling in the primary circuit of the RF; - to maintain the conditions in the coolant to ensure the corrosion resistance of structural steels; - to perform the coolant cleaning from solid-phase slags both based on lead oxides and based on oxides of structural steel components; - to clean the protective gas from coolant aerosols and corrosion products; - to perform the diagnostics of the contour state. HLMC technology includes both methods of performing technological operations, as well as special equipment, controls, and analysis methods. To date, a significant amount of research and development work has been carried out for the implementation of HLMC in the “BREST-OD-300” and special-purpose RF. The report paper presents the achievements in the field of HLMC technology in recent years of the work of the staff of IPPE JSC.
{"title":"CURRENT STATE AND ISSUES OF THE HEAVY LIQUID METAL COOLANT TECHNOLOGY DEVELOPMENT (Pb, Pb-Bi)","authors":"R. Askhadullin, A. Legkikh, V. Ulyanov, I. Voronin","doi":"10.55176/2414-1038-2021-2-105-115","DOIUrl":"https://doi.org/10.55176/2414-1038-2021-2-105-115","url":null,"abstract":"The technology of heavy liquid metal coolant (HLMC) is an important part of the safety system for the operation of reactor facilities with HLMC at all stages of their life cycle: from the preparation of a coolant and loading into the reactor to the decommissioning of the reactor facility (RF). The technology of heavy liquid metal coolant is a set of measures that allow: - to prepare the coolant for filling in the primary circuit of the RF; - to maintain the conditions in the coolant to ensure the corrosion resistance of structural steels; - to perform the coolant cleaning from solid-phase slags both based on lead oxides and based on oxides of structural steel components; - to clean the protective gas from coolant aerosols and corrosion products; - to perform the diagnostics of the contour state. HLMC technology includes both methods of performing technological operations, as well as special equipment, controls, and analysis methods. To date, a significant amount of research and development work has been carried out for the implementation of HLMC in the “BREST-OD-300” and special-purpose RF. The report paper presents the achievements in the field of HLMC technology in recent years of the work of the staff of IPPE JSC.","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":"80607890","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: 10.55176/2414-1038-2021-2-52-63
A. Zrodnikov, V. Korobeinikov, A. Moseev, A. Egorov, V. Decusar
In accordance with the approved order of the State Atomic Energy Corporation Rosatom dated May 24, 2018 No. 1-1/366-r “Strategy for the development of nuclear power until 2050 and the outlook for the period up to 2100”, the transition to a two-component structure of nuclear power in Russia is a key area of its technological development. The structure of a two-component nuclear power system (hereinafter NES) can include various elements, the search for the optimal combination of which in development is a multi-parameter problem. The relevance of determining the appearance and structure of a two-component NES is that it is designed to ensure sustainable, cost-effective electricity production in the face of growing global restrictions on natural resources of fossil fuel and environmentally harmful emissions, while meeting the requirements of nuclear, radiation safety and nuclear non-proliferation, as well as to realize the potential increasing the commercial attractiveness of the nuclear industry through the provision of services in the field of nuclear fuel cycle. The article presents the results of computational and analytical studies to substantiate the creation of a two-component NES and recommendations for choosing the structure of a two-component NES with optimal technical and economic indicators, taking into account the export potential. To select the optimal scenario for the development of a nuclear power plant, a multi-criteria analysis of the simulation results with a given set of system criteria was carried out.
{"title":"CHOICE OF THE STRUCTURE OF A TWO-COMPONENT NES WITH OPTIMAL TAKING INTO ACCOUNT AN EXPORT POTENTIAL AND TECHNICAL AND ECONOMIC INDICATORS","authors":"A. Zrodnikov, V. Korobeinikov, A. Moseev, A. Egorov, V. Decusar","doi":"10.55176/2414-1038-2021-2-52-63","DOIUrl":"https://doi.org/10.55176/2414-1038-2021-2-52-63","url":null,"abstract":"In accordance with the approved order of the State Atomic Energy Corporation Rosatom dated May 24, 2018 No. 1-1/366-r “Strategy for the development of nuclear power until 2050 and the outlook for the period up to 2100”, the transition to a two-component structure of nuclear power in Russia is a key area of its technological development. The structure of a two-component nuclear power system (hereinafter NES) can include various elements, the search for the optimal combination of which in development is a multi-parameter problem. The relevance of determining the appearance and structure of a two-component NES is that it is designed to ensure sustainable, cost-effective electricity production in the face of growing global restrictions on natural resources of fossil fuel and environmentally harmful emissions, while meeting the requirements of nuclear, radiation safety and nuclear non-proliferation, as well as to realize the potential increasing the commercial attractiveness of the nuclear industry through the provision of services in the field of nuclear fuel cycle. The article presents the results of computational and analytical studies to substantiate the creation of a two-component NES and recommendations for choosing the structure of a two-component NES with optimal technical and economic indicators, taking into account the export potential. To select the optimal scenario for the development of a nuclear power plant, a multi-criteria analysis of the simulation results with a given set of system criteria was carried out.","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":"81065070","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: 10.55176/2414-1038-2021-2-181-199
V. Ulyanov, M. Koshelev, S. Kharchuk
The unique properties of heavy liquid metal coolants (low chemical activity when interacting with air, organic compounds, water and water vapor, high boiling point, low vapor pressure) allow them to be used both in the first circuits of fast neutron reactor plants and in heat exchangers used for processing various organic and inorganic raw materials. Such devices can be either recuperative or mixing type. The latter additionally have the following advantages: lower cost and simplicity of construction; no heat transfer surfaces that can be subject to corrosion and contamination; reduced dimensions due to a larger specific heat transfer surface. The greatest degree of study relates to liquid metal pyrolysis of solid substances, direct contact distillation of aqueous solutions without preliminary preparation, improving the efficiency of continuous casting machines due to the replacement of water with lead-bismuth eutectic. The results of research in support of the above processes are presented. The methods of maintaining the quality of lead-containing heat carriers do not have fundamental differences for the conditions of reactor installations and in relation to promising technologies for processing organic and inorganic raw materials. At the same time, the requirements for devices for implementing these methods in promising technologies for processing various raw materials are significantly lower: in general, a sensor for measuring the thermodynamic activity of oxygen and a device for entering gas (hydrogen-containing or oxygen-containing) mixtures in the coolants.
{"title":"MAINTAINING THE QUALITY OF LEAD CONTAINING COOLANT IN ADVANCED TECHNOLOGIES RECYCLING ORGANIC AND INORGANIC WASTE","authors":"V. Ulyanov, M. Koshelev, S. Kharchuk","doi":"10.55176/2414-1038-2021-2-181-199","DOIUrl":"https://doi.org/10.55176/2414-1038-2021-2-181-199","url":null,"abstract":"The unique properties of heavy liquid metal coolants (low chemical activity when interacting with air, organic compounds, water and water vapor, high boiling point, low vapor pressure) allow them to be used both in the first circuits of fast neutron reactor plants and in heat exchangers used for processing various organic and inorganic raw materials. Such devices can be either recuperative or mixing type. The latter additionally have the following advantages: lower cost and simplicity of construction; no heat transfer surfaces that can be subject to corrosion and contamination; reduced dimensions due to a larger specific heat transfer surface. The greatest degree of study relates to liquid metal pyrolysis of solid substances, direct contact distillation of aqueous solutions without preliminary preparation, improving the efficiency of continuous casting machines due to the replacement of water with lead-bismuth eutectic. The results of research in support of the above processes are presented. The methods of maintaining the quality of lead-containing heat carriers do not have fundamental differences for the conditions of reactor installations and in relation to promising technologies for processing organic and inorganic raw materials. At the same time, the requirements for devices for implementing these methods in promising technologies for processing various raw materials are significantly lower: in general, a sensor for measuring the thermodynamic activity of oxygen and a device for entering gas (hydrogen-containing or oxygen-containing) mixtures in the coolants.","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":"85027660","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: 10.55176/2414-1038-2021-2-82-104
I. Kurina, M. Frolova, E. Chesnokov
The article provides a review of well-known foreign scientific publications devoted to the study of the properties of metallic nuclear fuel based on U-Zr, in composition close to U 10 wt. % Zr, which is widely used in reactors. Differences in the microstructure of fuel made by different methods: extrusion and casting - are considered. The effect of thermal annealing on the change in the microstructure of the alloy is shown. The photographs obtained using optical and electron microscopes are presented, as well as crystallographic data for two phases: α-U and δ-UZr2. The known literature data indicate that the density of uranium-rich U-Zr alloys corresponds to the rule of mixtures. The theoretical density of the alloy U-10 wt. % Zr (U-22.5 at. % Zr) should be taken as 16.2 g/cm3. The results of thermophysical studies of 10 wt. % Zr fuel using the method of differential scanning calorimetry (DSC) are presented. Data on measurements of thermal expansion of U-Zr alloys, as well as thermal conductivity are presented. Most of the thermal conductivity data are either calculated from the measured density, specific heat and thermal diffusivity, or obtained from simulations.
{"title":"REVIEW OF METAL FUEL U-10 wt. % Zr STUDIES","authors":"I. Kurina, M. Frolova, E. Chesnokov","doi":"10.55176/2414-1038-2021-2-82-104","DOIUrl":"https://doi.org/10.55176/2414-1038-2021-2-82-104","url":null,"abstract":"The article provides a review of well-known foreign scientific publications devoted to the study of the properties of metallic nuclear fuel based on U-Zr, in composition close to U 10 wt. % Zr, which is widely used in reactors. Differences in the microstructure of fuel made by different methods: extrusion and casting - are considered. The effect of thermal annealing on the change in the microstructure of the alloy is shown. The photographs obtained using optical and electron microscopes are presented, as well as crystallographic data for two phases: α-U and δ-UZr2. The known literature data indicate that the density of uranium-rich U-Zr alloys corresponds to the rule of mixtures. The theoretical density of the alloy U-10 wt. % Zr (U-22.5 at. % Zr) should be taken as 16.2 g/cm3. The results of thermophysical studies of 10 wt. % Zr fuel using the method of differential scanning calorimetry (DSC) are presented. Data on measurements of thermal expansion of U-Zr alloys, as well as thermal conductivity are presented. Most of the thermal conductivity data are either calculated from the measured density, specific heat and thermal diffusivity, or obtained from simulations.","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":"76606433","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: 10.55176/2414-1038-2021-2-268-280
A. Shlepkin, A. Morozov, A. Sakhipgareev, D. Kalyakin
The article presents the results of experimental studies of the thermophysical (density, viscosity) and physicochemical (degree of acidity - pH) properties of water solutions of boric acid. A review of the available literature data on the effect of the properties of boric acid solutions on heat removal from the reactor core is presented. It has been established that the available information is very general and does not cover the entire range of parameters (temperature, pressure, acid concentration) characteristic of a possible emergency at a nuclear power plant with a VVER reactor. Methods and facilities for conducting experimental studies are described. The results of experimental studies are presented. The density of aqueous solutions of boric acid with a concentration of 2.5-450 g/kg H2O at a temperature of 25-130 °C was determined. The dependence of the investigated characteristic on temperature and concentration was also obtained. The results of an experimental study of the kinematic viscosity of water solutions of boric acid with a concentration of 2.5-200 g/kg H2O at a temperature of 25-90 °C were obtained. The total error in measuring the viscosity of aqueous solutions of boric acid did not exceed 2 %. The pH values of water solutions of boric acid in the temperature range 25-50 and concentrations of 2.5-450 g/kg H2O were determined. The dependence for calculating the degree of acidity of boric acid is obtained. Experimental data on the thermophysical and physicochemical properties of water solutions of boric acid can be used to refine the results of calculations of emergency heat removal processes in a reactor facility, carried out using both one-dimensional calculation programs and three-dimensional CFD codes.
{"title":"DETERMINATION OF THE COMPLEX OF THERMOPHYSICAL PROPERTIES OF BORIC ACID WATER SOLUTIONS AT PARAMETERS CHARACTERISTIC FOR VVER EMERGENCY MODE","authors":"A. Shlepkin, A. Morozov, A. Sakhipgareev, D. Kalyakin","doi":"10.55176/2414-1038-2021-2-268-280","DOIUrl":"https://doi.org/10.55176/2414-1038-2021-2-268-280","url":null,"abstract":"The article presents the results of experimental studies of the thermophysical (density, viscosity) and physicochemical (degree of acidity - pH) properties of water solutions of boric acid. A review of the available literature data on the effect of the properties of boric acid solutions on heat removal from the reactor core is presented. It has been established that the available information is very general and does not cover the entire range of parameters (temperature, pressure, acid concentration) characteristic of a possible emergency at a nuclear power plant with a VVER reactor. Methods and facilities for conducting experimental studies are described. The results of experimental studies are presented. The density of aqueous solutions of boric acid with a concentration of 2.5-450 g/kg H2O at a temperature of 25-130 °C was determined. The dependence of the investigated characteristic on temperature and concentration was also obtained. The results of an experimental study of the kinematic viscosity of water solutions of boric acid with a concentration of 2.5-200 g/kg H2O at a temperature of 25-90 °C were obtained. The total error in measuring the viscosity of aqueous solutions of boric acid did not exceed 2 %. The pH values of water solutions of boric acid in the temperature range 25-50 and concentrations of 2.5-450 g/kg H2O were determined. The dependence for calculating the degree of acidity of boric acid is obtained. Experimental data on the thermophysical and physicochemical properties of water solutions of boric acid can be used to refine the results of calculations of emergency heat removal processes in a reactor facility, carried out using both one-dimensional calculation programs and three-dimensional CFD codes.","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":"74078702","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: 10.55176/2414-1038-2021-2-127-138
T. Vereshchagina, A. Mikheev, Y. Kudryaeva
This paper presents a brief review of research and developments history in the field of thermoacoustics from the 18th to the 20th century. The modern state of the art in the field of thermoacoustics is presented too. The basic equations of the linear thermoacoustics (generalized Rott - Swift equations) theory are presented. It is shown that now there are wide opportunities to use various devices based on the direct and reverse thermoacoustic effects. The features and advantages of thermoacoustic devices are listed. Particular attention is paid to cooling systems. It is shown that the development of refrigeration units based on the thermoacoustic effect is an actual and promising task. The conclusion is made about the relevance of theoretical and experimental studies in the field of thermoacoustics. The article also presents the results of publication activity analysis in the field of thermoacoustics over the past 20 years. It is concluded that there has been a dramatic increasing of publications number deal with the results of scientific and practical developments in the field of thermoacoustics over the past 20 years, both in Russia and abroad. Based on the analysis of inventive activity, conclusions are drawn about the wide scope of application of thermoacoustic devices and the directions of their improvement.
{"title":"THERMOACOUSTIC EFFECT AND ITS APPLICATION","authors":"T. Vereshchagina, A. Mikheev, Y. Kudryaeva","doi":"10.55176/2414-1038-2021-2-127-138","DOIUrl":"https://doi.org/10.55176/2414-1038-2021-2-127-138","url":null,"abstract":"This paper presents a brief review of research and developments history in the field of thermoacoustics from the 18th to the 20th century. The modern state of the art in the field of thermoacoustics is presented too. The basic equations of the linear thermoacoustics (generalized Rott - Swift equations) theory are presented. It is shown that now there are wide opportunities to use various devices based on the direct and reverse thermoacoustic effects. The features and advantages of thermoacoustic devices are listed. Particular attention is paid to cooling systems. It is shown that the development of refrigeration units based on the thermoacoustic effect is an actual and promising task. The conclusion is made about the relevance of theoretical and experimental studies in the field of thermoacoustics. The article also presents the results of publication activity analysis in the field of thermoacoustics over the past 20 years. It is concluded that there has been a dramatic increasing of publications number deal with the results of scientific and practical developments in the field of thermoacoustics over the past 20 years, both in Russia and abroad. Based on the analysis of inventive activity, conclusions are drawn about the wide scope of application of thermoacoustic devices and the directions of their improvement.","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":"80083842","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: 10.55176/2414-1038-2021-2-34-51
P. Alekseev, A. Andrianov, M. Bakanov, A. Balanin, A. Gulevich, V. Dekusar, A. Egorov, V. Korobeinikov, E. Marova, A. Maslov, A. Moseev, V. Nevinitsa, P. Teplov, M. Farakshin, P. Fomichenko, S. Shepelev
The development of an energy system, including nuclear, is a long and multi-stage process. The complexity of the process is determined by the deep degree of workforce decomposition at nuclear power facilities, high capital and science intensity of the industry. When assessing the prospects, it is necessary to take into account many significant external factors and uncertainty of the future macroeconomics development trends. In the process of analysis and decision-making on the development of the country's energy system, there are many participants of technological concepts are involved, often with divergent interests. The main task of justifying decisions on the development of the energy industry and, in particular, nuclear power are coordination of interests of all subjects of relations and methodology the formation of mechanisms to ensure the development process. In such an assessment, the ultimate goal is sustainable development. To determine the technological development of the energy system with nuclear and non-nuclear power technologies for the next 5-10 years, it is necessary to assess the long-term prospects of such scenarios. This task requires a multi-criteria approach to analyze the competitiveness of technology using a list of criteria of competitive advantages, combining economic indicators and indicators that characterize safety, environmental impact, risks of implementation, development prospects. This paper discusses the results of multi-criteria system analysis using international assessment tools. The transfer of Russian nuclear power to the two-component nuclear energy system with VVER and BN reactors and the expansion of the international fuel business by providing a full range of NFC services was evaluated using a set of key criteria.
{"title":"ANALYSIS OF THE COMPETITIVENESS OF A NEW GENERATION BN POWER UNIT PROJECT, TAKING INTO ACCOUNT SYSTEM REQUIREMENTS","authors":"P. Alekseev, A. Andrianov, M. Bakanov, A. Balanin, A. Gulevich, V. Dekusar, A. Egorov, V. Korobeinikov, E. Marova, A. Maslov, A. Moseev, V. Nevinitsa, P. Teplov, M. Farakshin, P. Fomichenko, S. Shepelev","doi":"10.55176/2414-1038-2021-2-34-51","DOIUrl":"https://doi.org/10.55176/2414-1038-2021-2-34-51","url":null,"abstract":"The development of an energy system, including nuclear, is a long and multi-stage process. The complexity of the process is determined by the deep degree of workforce decomposition at nuclear power facilities, high capital and science intensity of the industry. When assessing the prospects, it is necessary to take into account many significant external factors and uncertainty of the future macroeconomics development trends. In the process of analysis and decision-making on the development of the country's energy system, there are many participants of technological concepts are involved, often with divergent interests. The main task of justifying decisions on the development of the energy industry and, in particular, nuclear power are coordination of interests of all subjects of relations and methodology the formation of mechanisms to ensure the development process. In such an assessment, the ultimate goal is sustainable development. To determine the technological development of the energy system with nuclear and non-nuclear power technologies for the next 5-10 years, it is necessary to assess the long-term prospects of such scenarios. This task requires a multi-criteria approach to analyze the competitiveness of technology using a list of criteria of competitive advantages, combining economic indicators and indicators that characterize safety, environmental impact, risks of implementation, development prospects. This paper discusses the results of multi-criteria system analysis using international assessment tools. The transfer of Russian nuclear power to the two-component nuclear energy system with VVER and BN reactors and the expansion of the international fuel business by providing a full range of NFC services was evaluated using a set of key criteria.","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":"83871061","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: 10.55176/2414-1038-2021-2-218-235
V. Grabezhnaya, A. Mikheyev, A. Alekhin, A. Kryukov, A. Tikhomirov
The project BREST-OD-300 reactor plant (RP) with a fast neutron reactor and a lead coolant in the primary circuit is being developed in NIKIET JSC. As a steam generator (SG), a helical-type steam generator with coiled tubes with subcritical pressure water in the second circuit is considered. To substantiate the design characteristics of the secondary coolant at the State Research Center of the Russian Federation - IPPE, thermohydraulic tests of various SG models were carried out at the SPRUT stand Initially, tests were carried out on a model of a coiled steam generator consisting of two three-tube modules with a longitudinal lead flow around a three-tube bundle of coiled tubes. The influence of operating parameters on thermohydraulic characteristics and hydrodynamic stability is shown in the case of operation of one module, as well as in the joint operation of two models in the investigated range of operating parameters. At the second stage, tests of a standard steam generator model were carried out with lead flowing around 18 heat exchange tubes. In the multitube model, the downward movement of the heating coolant took place with the flow around the bundle of heat transfer tubes close to the transverse flow. Data were obtained on the hydrodynamic stability of steam generating tubes and the entire model as a whole when operating in the entire range of changes in operating parameters, which are necessary for creating a databank and further verification of calculation codes describing the ongoing thermohydraulic processes. During the tests in both models of the steam generator, there were no noises inherent in unstable operating modes of the circuit. No pulsations of water and steam temperature were found, respectively, in the inlet and outlet collectors. At high lead temperatures, the temperature of the superheated steam was always close to the lead inlet temperature. A series of works devoted to the study of heat transfer from the side of a lead coolant with a transverse flow around a package of heat exchange tubes in normal heat transfer modes and with freezing of lead has been completed. Studies have been carried out on the effect of oxygen concentration in lead on heat transfer.
{"title":"EXPERIMENTAL JUSTIFICATION OF DESIGN CHARACTERISTICS OF STEAM GENERATOR RP BREST-OD-300","authors":"V. Grabezhnaya, A. Mikheyev, A. Alekhin, A. Kryukov, A. Tikhomirov","doi":"10.55176/2414-1038-2021-2-218-235","DOIUrl":"https://doi.org/10.55176/2414-1038-2021-2-218-235","url":null,"abstract":"The project BREST-OD-300 reactor plant (RP) with a fast neutron reactor and a lead coolant in the primary circuit is being developed in NIKIET JSC. As a steam generator (SG), a helical-type steam generator with coiled tubes with subcritical pressure water in the second circuit is considered. To substantiate the design characteristics of the secondary coolant at the State Research Center of the Russian Federation - IPPE, thermohydraulic tests of various SG models were carried out at the SPRUT stand Initially, tests were carried out on a model of a coiled steam generator consisting of two three-tube modules with a longitudinal lead flow around a three-tube bundle of coiled tubes. The influence of operating parameters on thermohydraulic characteristics and hydrodynamic stability is shown in the case of operation of one module, as well as in the joint operation of two models in the investigated range of operating parameters. At the second stage, tests of a standard steam generator model were carried out with lead flowing around 18 heat exchange tubes. In the multitube model, the downward movement of the heating coolant took place with the flow around the bundle of heat transfer tubes close to the transverse flow. Data were obtained on the hydrodynamic stability of steam generating tubes and the entire model as a whole when operating in the entire range of changes in operating parameters, which are necessary for creating a databank and further verification of calculation codes describing the ongoing thermohydraulic processes. During the tests in both models of the steam generator, there were no noises inherent in unstable operating modes of the circuit. No pulsations of water and steam temperature were found, respectively, in the inlet and outlet collectors. At high lead temperatures, the temperature of the superheated steam was always close to the lead inlet temperature. A series of works devoted to the study of heat transfer from the side of a lead coolant with a transverse flow around a package of heat exchange tubes in normal heat transfer modes and with freezing of lead has been completed. Studies have been carried out on the effect of oxygen concentration in lead on heat transfer.","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":"79261523","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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