Pub Date : 2020-12-26DOI: 10.55176/2414-1038-2020-4-138-147
V. Levchenko, M. Kascheev, S. Dorokhovich, A. Zaytsev
The problem of determining a non-stationary three-dimensional temperature field in a k-layer cylinder of length is solved. There is a symmetrically located cylindrical cavity in the center of this body. The absence of a cavity is a special case of the problem. In each layer, there are heat sources, depending on the coordinates and time. The initial temperature of the layers is a function of the coordinates. In the center of the body the symmetry condition is fulfilled. At the boundary of contact of the layers - ideal thermal contact: continuity of temperatures and heat flows. On the inner and outer side surfaces and ends, heat exchange occurs according to Newton's law with environments whose temperatures change over time according to an arbitrary law. The periodicity condition is set for the angle φ. The problem in this statement is solved for the first time. For the solution of the problem the following approach is used: by means of the method of finite integral transformations differential operations on longitudinal coordinate, angle and transverse coordinate are sequentially excluded, and the determination of time dependence of temperature is reduced to the solution of the ordinary differential equation of the first order.
{"title":"NON-STATIONARY THREE-DIMENSIONAL TEMPERATURE FIELD IN A MULTILAYER CYLINDER","authors":"V. Levchenko, M. Kascheev, S. Dorokhovich, A. Zaytsev","doi":"10.55176/2414-1038-2020-4-138-147","DOIUrl":"https://doi.org/10.55176/2414-1038-2020-4-138-147","url":null,"abstract":"The problem of determining a non-stationary three-dimensional temperature field in a k-layer cylinder of length is solved. There is a symmetrically located cylindrical cavity in the center of this body. The absence of a cavity is a special case of the problem. In each layer, there are heat sources, depending on the coordinates and time. The initial temperature of the layers is a function of the coordinates. In the center of the body the symmetry condition is fulfilled. At the boundary of contact of the layers - ideal thermal contact: continuity of temperatures and heat flows. On the inner and outer side surfaces and ends, heat exchange occurs according to Newton's law with environments whose temperatures change over time according to an arbitrary law. The periodicity condition is set for the angle φ. The problem in this statement is solved for the first time. For the solution of the problem the following approach is used: by means of the method of finite integral transformations differential operations on longitudinal coordinate, angle and transverse coordinate are sequentially excluded, and the determination of time dependence of temperature is reduced to the solution of the ordinary differential equation of the first order.","PeriodicalId":20426,"journal":{"name":"PROBLEMS OF ATOMIC SCIENCE AND TECHNOLOGY. SERIES: NUCLEAR AND REACTOR CONSTANTS","volume":null,"pages":null},"PeriodicalIF":0.0,"publicationDate":"2020-12-26","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"81400911","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 : 2020-12-26DOI: 10.55176/2414-1038-2020-4-116-128
V. Delnov
The paper considers the issues of hydrodynamics of typical distribution header systems (DHS) in heat exchangers and reactors of nuclear power plants (NPP). The analysis of hydrodynamic properties of coolant flow in DHS has been carried out. The paper presents typical flow patterns of a coolant in the flow paths of the specified DHS. The classification of different types of DHS is presented. The coolant flow is shown to have a jet nature in a relatively confined and unconfined DHS. Consideration is given to several types of jets and processes of transformation of some types of jets into other types. The paper presents a brief description and formula of the fluid distribution regularity (registered as a scientific discovery) at the outlet from the flat and cylindrical DHS with central and side fluid supply. Consideration is given to the similarity property of hydrodynamics in the flat and cylindrical RHS with different points of fluid supply to the header. Advanced areas of research into revealing hydrodynamic effects in DHS and obtaining additional information on the influence of various factors on the fluid flow in them are proposed.
{"title":"HYDRODYNAMICS OF TYPICAL DISTRIBUTION HEADER SYSTEMS OF NUCLEAR POWER PLANTS: MODERN VIEWS AND RESEARCH PERSPECTIVES","authors":"V. Delnov","doi":"10.55176/2414-1038-2020-4-116-128","DOIUrl":"https://doi.org/10.55176/2414-1038-2020-4-116-128","url":null,"abstract":"The paper considers the issues of hydrodynamics of typical distribution header systems (DHS) in heat exchangers and reactors of nuclear power plants (NPP). The analysis of hydrodynamic properties of coolant flow in DHS has been carried out. The paper presents typical flow patterns of a coolant in the flow paths of the specified DHS. The classification of different types of DHS is presented. The coolant flow is shown to have a jet nature in a relatively confined and unconfined DHS. Consideration is given to several types of jets and processes of transformation of some types of jets into other types. The paper presents a brief description and formula of the fluid distribution regularity (registered as a scientific discovery) at the outlet from the flat and cylindrical DHS with central and side fluid supply. Consideration is given to the similarity property of hydrodynamics in the flat and cylindrical RHS with different points of fluid supply to the header. Advanced areas of research into revealing hydrodynamic effects in DHS and obtaining additional information on the influence of various factors on the fluid flow in them are proposed.","PeriodicalId":20426,"journal":{"name":"PROBLEMS OF ATOMIC SCIENCE AND TECHNOLOGY. SERIES: NUCLEAR AND REACTOR CONSTANTS","volume":null,"pages":null},"PeriodicalIF":0.0,"publicationDate":"2020-12-26","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"89459627","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 : 2020-12-26DOI: 10.55176/2414-1038-2020-4-78-85
O. Gurskaya, E. Dzugkoeva, L. Korobeynikova, V. Mishin, V. Stogov
The current program in Russia to increase the fuel consumption of fast reactors and increase its burn-out causes the transition to new structural materials, which, in turn, leads to changes in the neutron-physical characteristics of reactors. In particular, the drop in the reactivity reserve noted in the BN-600 reactor of the Beloyarsk NPP at the end of 76 operational cycles, as will be shown below, is due to the transition to a new type of shell steel with an increased content of nickel, which strongly affects the reactivity. Design support for the operation of the BN-600 and BN-800 fast reactors, as well as the experiments carried out on them, is performed by IPPE. This article presents the results of a calculated analysis of the expected changes in the reactivity reserve at the end of 76 operational cycles when replacing the shell steel in BN-600. In addition, the influence of experimental assemblies located in the core on the reactivity reserve of the BN-600 is analyzed. Analysis of calculations of the actual loading of the BN-600 reactor at 76 operational cycle using the methods of the 1st-order perturbation theory, strict perturbation theory, and the Monte Carlo method showed that a partial transition at 76 operational cycle to EK-164 shell steel leads to a decrease in the reactivity margin by 0.030±0.004 %Δk/k. Replacement of steel for the entire core will reduce the reactivity margin by ~0.12 %Δk/k, which is confirmed by Monte Carlo calculations. The calculated reactivity margin obtained at the end of 76 operational cycles for the hot state of the BN-600 reactor is in good agreement with the measured reactivity margin.
{"title":"EFFECT OF REPLACING FA STRUCTURAL STEEL ON THE REACTIVITY RESERVE IN THE BN-600 REACTOR","authors":"O. Gurskaya, E. Dzugkoeva, L. Korobeynikova, V. Mishin, V. Stogov","doi":"10.55176/2414-1038-2020-4-78-85","DOIUrl":"https://doi.org/10.55176/2414-1038-2020-4-78-85","url":null,"abstract":"The current program in Russia to increase the fuel consumption of fast reactors and increase its burn-out causes the transition to new structural materials, which, in turn, leads to changes in the neutron-physical characteristics of reactors. In particular, the drop in the reactivity reserve noted in the BN-600 reactor of the Beloyarsk NPP at the end of 76 operational cycles, as will be shown below, is due to the transition to a new type of shell steel with an increased content of nickel, which strongly affects the reactivity. Design support for the operation of the BN-600 and BN-800 fast reactors, as well as the experiments carried out on them, is performed by IPPE. This article presents the results of a calculated analysis of the expected changes in the reactivity reserve at the end of 76 operational cycles when replacing the shell steel in BN-600. In addition, the influence of experimental assemblies located in the core on the reactivity reserve of the BN-600 is analyzed. Analysis of calculations of the actual loading of the BN-600 reactor at 76 operational cycle using the methods of the 1st-order perturbation theory, strict perturbation theory, and the Monte Carlo method showed that a partial transition at 76 operational cycle to EK-164 shell steel leads to a decrease in the reactivity margin by 0.030±0.004 %Δk/k. Replacement of steel for the entire core will reduce the reactivity margin by ~0.12 %Δk/k, which is confirmed by Monte Carlo calculations. The calculated reactivity margin obtained at the end of 76 operational cycles for the hot state of the BN-600 reactor is in good agreement with the measured reactivity margin.","PeriodicalId":20426,"journal":{"name":"PROBLEMS OF ATOMIC SCIENCE AND TECHNOLOGY. SERIES: NUCLEAR AND REACTOR CONSTANTS","volume":null,"pages":null},"PeriodicalIF":0.0,"publicationDate":"2020-12-26","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"86980331","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 : 2020-12-26DOI: 10.55176/2414-1038-2020-4-40-45
A. Kovalev, R. Rodionov, D. Portnov, Yu. A. Kashchuk
The work is devoted to the analysis of uranium radiation heating results under the action of neutron radiation. The revealed results discrepancies in the radiation heating modeling of the uranium 235 and 238 using FENDL 3.1d and ENDF nuclear data libraries motivate the study. At the same time, all initial data in the computational model, namely, the neutrons energy spectrum, mass of fissile material, geometry and surrounding structures composition were identical. In the course of the work, the results of energy release calculations in uranium cells were compared using various libraries. Analytical evaluation of the radiation heating results in test cells is given. The obtained data indicate erroneous values of neutron kerma factors in FENDL 3.1d nuclear data library. Fusion Evaluated Nuclear Data Library was created by the working group of the Nuclear Data Section of the IAEA in the mid-1990s. FENDL is focused on application in the design of fusion plants as well as plants for materials testing such as IFMIF. The expanded application required the addition of some materials to the library, the expansion of the energy range to higher energies and the addition of data for reactions caused by charged particles. Version 3.1d was released in January 2018 and is widely used in nuclear analysis of structural materials for the experimental fusion reactor ITER and the demonstration fusion reactor DEMO. The results of this work have been submitted to the ITER International Organization contact group for presentation to the developers of the FENDL library.
{"title":"INACCURACIES IN THE RADIATION HEATING CALCULATIONS OF URANIUM USING FENDL 3.1D LIBRARY","authors":"A. Kovalev, R. Rodionov, D. Portnov, Yu. A. Kashchuk","doi":"10.55176/2414-1038-2020-4-40-45","DOIUrl":"https://doi.org/10.55176/2414-1038-2020-4-40-45","url":null,"abstract":"The work is devoted to the analysis of uranium radiation heating results under the action of neutron radiation. The revealed results discrepancies in the radiation heating modeling of the uranium 235 and 238 using FENDL 3.1d and ENDF nuclear data libraries motivate the study. At the same time, all initial data in the computational model, namely, the neutrons energy spectrum, mass of fissile material, geometry and surrounding structures composition were identical. In the course of the work, the results of energy release calculations in uranium cells were compared using various libraries. Analytical evaluation of the radiation heating results in test cells is given. The obtained data indicate erroneous values of neutron kerma factors in FENDL 3.1d nuclear data library. Fusion Evaluated Nuclear Data Library was created by the working group of the Nuclear Data Section of the IAEA in the mid-1990s. FENDL is focused on application in the design of fusion plants as well as plants for materials testing such as IFMIF. The expanded application required the addition of some materials to the library, the expansion of the energy range to higher energies and the addition of data for reactions caused by charged particles. Version 3.1d was released in January 2018 and is widely used in nuclear analysis of structural materials for the experimental fusion reactor ITER and the demonstration fusion reactor DEMO. The results of this work have been submitted to the ITER International Organization contact group for presentation to the developers of the FENDL library.","PeriodicalId":20426,"journal":{"name":"PROBLEMS OF ATOMIC SCIENCE AND TECHNOLOGY. SERIES: NUCLEAR AND REACTOR CONSTANTS","volume":null,"pages":null},"PeriodicalIF":0.0,"publicationDate":"2020-12-26","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"79104015","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 : 2020-12-26DOI: 10.55176/2414-1038-2020-4-129-137
S. Lunina, V. Delnov
The paper presents the results of test calculations of hydrodynamics for distribution header systems (DHS) in heat exchangers and reactors of nuclear power plants (NPP). The header system is a typical element of the flow path in the specified NPP equipment, their efficient and safe operation is a priority area for Rosatom State Corporation. The calculations have been performed with the use of computational fluid dynamics software systems known as CFD codes. The results of calculations are presented for the coolant turbulent flow in the flow paths of the flat and cylindrical DHS with different conditions for the coolant supply and removal. A comparative analysis of the calculation results and experimental data on hydrodynamics of DHS typical of the nuclear power plants has been carried out, their satisfactory agreement has been shown. An assumption is made of the legitimacy of using the CFD code for the calculation of hydrodynamics of various DHS types and their specific typical elements.
{"title":"TEST CALCULATIONS OF HYDRODYNAMICS OF DISTRIBUTION HEADER SYSTEMS IN NPP HEAT EXCHANGERS AND REACTORS","authors":"S. Lunina, V. Delnov","doi":"10.55176/2414-1038-2020-4-129-137","DOIUrl":"https://doi.org/10.55176/2414-1038-2020-4-129-137","url":null,"abstract":"The paper presents the results of test calculations of hydrodynamics for distribution header systems (DHS) in heat exchangers and reactors of nuclear power plants (NPP). The header system is a typical element of the flow path in the specified NPP equipment, their efficient and safe operation is a priority area for Rosatom State Corporation. The calculations have been performed with the use of computational fluid dynamics software systems known as CFD codes. The results of calculations are presented for the coolant turbulent flow in the flow paths of the flat and cylindrical DHS with different conditions for the coolant supply and removal. A comparative analysis of the calculation results and experimental data on hydrodynamics of DHS typical of the nuclear power plants has been carried out, their satisfactory agreement has been shown. An assumption is made of the legitimacy of using the CFD code for the calculation of hydrodynamics of various DHS types and their specific typical elements.","PeriodicalId":20426,"journal":{"name":"PROBLEMS OF ATOMIC SCIENCE AND TECHNOLOGY. SERIES: NUCLEAR AND REACTOR CONSTANTS","volume":null,"pages":null},"PeriodicalIF":0.0,"publicationDate":"2020-12-26","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"79068272","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 : 2020-12-26DOI: 10.55176/2414-1038-2020-4-5-15
E. Seleznev
To analyze the features of calculating the transient behavior of fast breeder reactors, both using the classical system of kinetic equations and using the system of partial neutron transfer equations, computational studies of the test model of the MET1000 reactor developed under the Generation-IV project were carried out. As transient processes, displacements (dumping) of control rods were simulated with the assessment of reactivity effects in the specified reactor through the solution of stationary problems, i. e. through the use of an asymptotic estimate of the reactivity obtained from solving stationary homogeneous neutron transfer equations and from processing by the Inverse Solution of the Kinetics Equation method of solving a transient problem. Test calculations were performed in the 26th and 28th group approximations using the ABBN-93 and ABBN-RF libraries and eight groups of delayed neutrons. It is shown that the features of the determination of both calculation and measurement of the effects of reactivity and efficiency of control rods in Fast Breeder Reactors are associated with the presence of both instantaneous and delayed neutrons in the reactor, with the working region of the neutron spectrum and with a change in the shape of the neutron flux in the process of introducing perturbation.
{"title":"SOME CALCULATION FEATURES OF TRANSIENT BEHAVIOR OF THE REACTOR","authors":"E. Seleznev","doi":"10.55176/2414-1038-2020-4-5-15","DOIUrl":"https://doi.org/10.55176/2414-1038-2020-4-5-15","url":null,"abstract":"To analyze the features of calculating the transient behavior of fast breeder reactors, both using the classical system of kinetic equations and using the system of partial neutron transfer equations, computational studies of the test model of the MET1000 reactor developed under the Generation-IV project were carried out. As transient processes, displacements (dumping) of control rods were simulated with the assessment of reactivity effects in the specified reactor through the solution of stationary problems, i. e. through the use of an asymptotic estimate of the reactivity obtained from solving stationary homogeneous neutron transfer equations and from processing by the Inverse Solution of the Kinetics Equation method of solving a transient problem. Test calculations were performed in the 26th and 28th group approximations using the ABBN-93 and ABBN-RF libraries and eight groups of delayed neutrons. It is shown that the features of the determination of both calculation and measurement of the effects of reactivity and efficiency of control rods in Fast Breeder Reactors are associated with the presence of both instantaneous and delayed neutrons in the reactor, with the working region of the neutron spectrum and with a change in the shape of the neutron flux in the process of introducing perturbation.","PeriodicalId":20426,"journal":{"name":"PROBLEMS OF ATOMIC SCIENCE AND TECHNOLOGY. SERIES: NUCLEAR AND REACTOR CONSTANTS","volume":null,"pages":null},"PeriodicalIF":0.0,"publicationDate":"2020-12-26","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"84186988","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 : 2020-12-26DOI: 10.55176/2414-1038-2020-4-69-77
L. Bass, O. Nikolaeva, V. Davidenko, S. Gaifulin, A. Danilov, A. Khmylev
The article discusses the issue of the concept of “new generation code”, which has been actively used recently to characterize computer programs designed to solve problems of the transfer of neutrons and gamma quanta in nuclear facilities. As an example of a new generation code developed for solving the multigroup transport equation by the grid (deterministic) method, the first version of the new software package RADUGA-TV is considered, including, in particular, the UNK complex for calculating burnup. The article lists the main features of the RADUGA-TV code: the problems to be solved, the types of constants used, the methods for specifying the geometry of the calculation area, the methods for constructing an unstructured spatial mesh. The possibilities of the postprocessor for processing the obtained solution are presented. The article presents progressive algorithms included in the RADUGA-TV code, including grid schemes and methods for parallelizing computations. The advantages of using unstructured grids, including those consisting of cells of various types, are discussed. Methods for parallelizing computations on hybrid computing systems are considered. The question of the spatial grid decomposition when parallelizing computations on distributed memory systems is considered, as well as the question of organizing parallel computation on such systems. Comparison of the characteristics and capabilities of the RADUGA-TV code and other similar in purpose codes, foreign (ATTILA, AETIUS, ARES, THOR) and domestic ODETTA is performed. It is shown that the RADUGA-TV code is significantly advanced methodically and practically has no analogues. The article was written based on the materials of the report at the conference “Neutronika-19” and contains more detailed information on the issues discussed in the report.
{"title":"RADUGA-TV NEW GENERATION CODE FOR SOLUTION TRANSFER EQUATIONS","authors":"L. Bass, O. Nikolaeva, V. Davidenko, S. Gaifulin, A. Danilov, A. Khmylev","doi":"10.55176/2414-1038-2020-4-69-77","DOIUrl":"https://doi.org/10.55176/2414-1038-2020-4-69-77","url":null,"abstract":"The article discusses the issue of the concept of “new generation code”, which has been actively used recently to characterize computer programs designed to solve problems of the transfer of neutrons and gamma quanta in nuclear facilities. As an example of a new generation code developed for solving the multigroup transport equation by the grid (deterministic) method, the first version of the new software package RADUGA-TV is considered, including, in particular, the UNK complex for calculating burnup. The article lists the main features of the RADUGA-TV code: the problems to be solved, the types of constants used, the methods for specifying the geometry of the calculation area, the methods for constructing an unstructured spatial mesh. The possibilities of the postprocessor for processing the obtained solution are presented. The article presents progressive algorithms included in the RADUGA-TV code, including grid schemes and methods for parallelizing computations. The advantages of using unstructured grids, including those consisting of cells of various types, are discussed. Methods for parallelizing computations on hybrid computing systems are considered. The question of the spatial grid decomposition when parallelizing computations on distributed memory systems is considered, as well as the question of organizing parallel computation on such systems. Comparison of the characteristics and capabilities of the RADUGA-TV code and other similar in purpose codes, foreign (ATTILA, AETIUS, ARES, THOR) and domestic ODETTA is performed. It is shown that the RADUGA-TV code is significantly advanced methodically and practically has no analogues. The article was written based on the materials of the report at the conference “Neutronika-19” and contains more detailed information on the issues discussed in the report.","PeriodicalId":20426,"journal":{"name":"PROBLEMS OF ATOMIC SCIENCE AND TECHNOLOGY. SERIES: NUCLEAR AND REACTOR CONSTANTS","volume":null,"pages":null},"PeriodicalIF":0.0,"publicationDate":"2020-12-26","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"79542442","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 : 2020-12-26DOI: 10.55176/2414-1038-2020-4-86-115
A. Sorokin, A. Ivanov, Yu. A. Kuzina, A. Morozov, N. Denisova
An important problem determining the development of clean energy is the involvement of hydrogen in the fuel cycle. At present, the main method of hydrogen production is steam methane conversion. In the long term, large-scale hydrogen production, this method is not viable due to the consumption of non-renewable resources and the emission of greenhouse gases. Alternative methods of hydrogen production by water splitting methods using thermochemical or electrolysis processes require a high-temperature heat source. Nuclear reactors can serve as the most widely used high-temperature heat sources. The performed neutron-physical and thermophysical studies have shown that there is a fundamental possibility to provide the required parameters of a high-temperature (900-950 °C) with a 600 MW (thermal) fast neutron reactor with a sodium coolant for hydrogen production. It’s possible on the basis of one of the thermochemical cycles or high-temperature electrolysis with a high coefficient of thermal utilization of energy. It is shown that the temperature regime of core fuel elements is determined by a large number of parameters that have a regular and statistical nature. The developed methodology and numerical program allows to take into account, in the fuel assemblies shaped during the campaign, the effect on the temperature distribution of the fuel element cladding and temperature irregularities along the fuel element perimeter in the interchannel exchange fuel assembly, the random distribution of channel cross-sections and the heat generation of fuel elements using the Monte Carlo method, also other factors. For various reactor operating regimes, zones with stable temperature stratification with large gradients and temperature fluctuations have been identified. The results obtained make it possible to judge the amplitude and frequency characteristics of temperature pulsations in these potentially dangerous areas. The relative small size, the type of coolant, the choice of fissile material and structural materials make it possible to create a reactor with inherent properties that ensure increased nuclear and radiation safety.
{"title":"THERMAL AND HYDRAULIC STUDIES OF A HIGH-TEMPERATURE NPP FOR HYDROGEN PRODUCTION AND OTHER TECHNOLOGICAL APPLICATIONS","authors":"A. Sorokin, A. Ivanov, Yu. A. Kuzina, A. Morozov, N. Denisova","doi":"10.55176/2414-1038-2020-4-86-115","DOIUrl":"https://doi.org/10.55176/2414-1038-2020-4-86-115","url":null,"abstract":"An important problem determining the development of clean energy is the involvement of hydrogen in the fuel cycle. At present, the main method of hydrogen production is steam methane conversion. In the long term, large-scale hydrogen production, this method is not viable due to the consumption of non-renewable resources and the emission of greenhouse gases. Alternative methods of hydrogen production by water splitting methods using thermochemical or electrolysis processes require a high-temperature heat source. Nuclear reactors can serve as the most widely used high-temperature heat sources. The performed neutron-physical and thermophysical studies have shown that there is a fundamental possibility to provide the required parameters of a high-temperature (900-950 °C) with a 600 MW (thermal) fast neutron reactor with a sodium coolant for hydrogen production. It’s possible on the basis of one of the thermochemical cycles or high-temperature electrolysis with a high coefficient of thermal utilization of energy. It is shown that the temperature regime of core fuel elements is determined by a large number of parameters that have a regular and statistical nature. The developed methodology and numerical program allows to take into account, in the fuel assemblies shaped during the campaign, the effect on the temperature distribution of the fuel element cladding and temperature irregularities along the fuel element perimeter in the interchannel exchange fuel assembly, the random distribution of channel cross-sections and the heat generation of fuel elements using the Monte Carlo method, also other factors. For various reactor operating regimes, zones with stable temperature stratification with large gradients and temperature fluctuations have been identified. The results obtained make it possible to judge the amplitude and frequency characteristics of temperature pulsations in these potentially dangerous areas. The relative small size, the type of coolant, the choice of fissile material and structural materials make it possible to create a reactor with inherent properties that ensure increased nuclear and radiation safety.","PeriodicalId":20426,"journal":{"name":"PROBLEMS OF ATOMIC SCIENCE AND TECHNOLOGY. SERIES: NUCLEAR AND REACTOR CONSTANTS","volume":null,"pages":null},"PeriodicalIF":0.0,"publicationDate":"2020-12-26","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"81922854","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 : 2020-12-26DOI: 10.55176/2414-1038-2020-4-158-162
A. Osipov, R. Askhadullin, O. Lavrova, K. Ivanov, S. Niyazov, R. Cheporov
At present, installations with heavy liquid metal coolants (HLMC) based on lead melts are considered as promising nuclear power plants, since these coolants have a number of advantages over alkali metals and other coolants. A feature of heavy liquid metals is their rather high corrosivity towards structural materials; however, at present this problem has been partially solved due to the formation of protective oxide coatings on the surfaces of steels based on oxides of metal components of structural steels. This determines the main tasks of modern HLMC technology, which are to provide conditions for the formation and maintenance of oxide layers with optimal protective properties on the surfaces of structural steels. The success of solving these problems in the general case depends on an adequate representation and quantitative description of the properties of the system “liquid metal melt - oxide phase”. In this work, we consider the questions of the possibility of the formation of plumboferrite phases in liquid lead, their composition and thermodynamic properties.
{"title":"ESTIMATION OF THE LOW-OXYGEN BOUNDARY OF THE EXISTENCE OF PLUMBOFERRITE PHASES IN THE LEAD MELT","authors":"A. Osipov, R. Askhadullin, O. Lavrova, K. Ivanov, S. Niyazov, R. Cheporov","doi":"10.55176/2414-1038-2020-4-158-162","DOIUrl":"https://doi.org/10.55176/2414-1038-2020-4-158-162","url":null,"abstract":"At present, installations with heavy liquid metal coolants (HLMC) based on lead melts are considered as promising nuclear power plants, since these coolants have a number of advantages over alkali metals and other coolants. A feature of heavy liquid metals is their rather high corrosivity towards structural materials; however, at present this problem has been partially solved due to the formation of protective oxide coatings on the surfaces of steels based on oxides of metal components of structural steels. This determines the main tasks of modern HLMC technology, which are to provide conditions for the formation and maintenance of oxide layers with optimal protective properties on the surfaces of structural steels. The success of solving these problems in the general case depends on an adequate representation and quantitative description of the properties of the system “liquid metal melt - oxide phase”. In this work, we consider the questions of the possibility of the formation of plumboferrite phases in liquid lead, their composition and thermodynamic properties.","PeriodicalId":20426,"journal":{"name":"PROBLEMS OF ATOMIC SCIENCE AND TECHNOLOGY. SERIES: NUCLEAR AND REACTOR CONSTANTS","volume":null,"pages":null},"PeriodicalIF":0.0,"publicationDate":"2020-12-26","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"87171496","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 : 2020-12-26DOI: 10.55176/2414-1038-2020-4-60-68
E. Kinev, A. Ustinov
The radiation swelling of hot-pressed B4C blocks of industrial and re-fabricated quality with the participation of absorber elements providing emergency protection, up to fast neutron fluence 7⋅1022 cm-2, was investigated. The reason of metal cladding machining deformation of neutron absorber during a two-year exposure period was swelling. The volumetric swelling and the porous parameters of the absorber are measured by hydrostatic weighing and digital analysis of electron microscopic fractograms of fresh splits. The pore allocation and destruction manner of B4C was studied by scanning electron microscopy. It was found that swelling for industrial and re-fabricated B4C, with maximum neutron fluence in the lower part of absorbing elements, culminates 12 and 18 % respectively. The drastic decrease in its swelling was caused by fluence reduction along of absorber rod. Volumetric changes are explained by growth of closed porosity. The maximum swelling of irradiated briquettes for industrial and re-manufactured B4C is 25 and 30 %, respectively, with the same initial values of about 19 and 23 %. The maximum value of closed porosity from industrial and re-manufactured B4C after irradiation is 14 and 21 %, respectively, against the initial values of 4 and 9 %. All detected pores settle at the grain border. The primary bulk of the pores is of technological origin. The size and concentration of pores correlate with an increase of neutron fluence and burnout during irradiation, as well as provoke to formation of microcracks between the crystals and grains of the boron carbide matrix. Nanoscale helium pores with a diameter of more than 100 nm were not detected in the grain body. The open porosity inside maximum swelling zone and mechanical interaction absorber with the cladding is reduced. The fracture nature of the swelling B4C changes from a mixture with a predominance of intragranular splitting to a purely intergranular type under increasing fluence.
{"title":"SWELLING OF NEUTRON ABSORBER ON THE BASIS OF INDUSTRIAL AND RECONSTRUCTED QUALITIES B4C AFTER TWO-YEAR IRRADIATION IN THE EMERGENCY PROTECTION RODS","authors":"E. Kinev, A. Ustinov","doi":"10.55176/2414-1038-2020-4-60-68","DOIUrl":"https://doi.org/10.55176/2414-1038-2020-4-60-68","url":null,"abstract":"The radiation swelling of hot-pressed B4C blocks of industrial and re-fabricated quality with the participation of absorber elements providing emergency protection, up to fast neutron fluence 7⋅1022 cm-2, was investigated. The reason of metal cladding machining deformation of neutron absorber during a two-year exposure period was swelling. The volumetric swelling and the porous parameters of the absorber are measured by hydrostatic weighing and digital analysis of electron microscopic fractograms of fresh splits. The pore allocation and destruction manner of B4C was studied by scanning electron microscopy. It was found that swelling for industrial and re-fabricated B4C, with maximum neutron fluence in the lower part of absorbing elements, culminates 12 and 18 % respectively. The drastic decrease in its swelling was caused by fluence reduction along of absorber rod. Volumetric changes are explained by growth of closed porosity. The maximum swelling of irradiated briquettes for industrial and re-manufactured B4C is 25 and 30 %, respectively, with the same initial values of about 19 and 23 %. The maximum value of closed porosity from industrial and re-manufactured B4C after irradiation is 14 and 21 %, respectively, against the initial values of 4 and 9 %. All detected pores settle at the grain border. The primary bulk of the pores is of technological origin. The size and concentration of pores correlate with an increase of neutron fluence and burnout during irradiation, as well as provoke to formation of microcracks between the crystals and grains of the boron carbide matrix. Nanoscale helium pores with a diameter of more than 100 nm were not detected in the grain body. The open porosity inside maximum swelling zone and mechanical interaction absorber with the cladding is reduced. The fracture nature of the swelling B4C changes from a mixture with a predominance of intragranular splitting to a purely intergranular type under increasing fluence.","PeriodicalId":20426,"journal":{"name":"PROBLEMS OF ATOMIC SCIENCE AND TECHNOLOGY. SERIES: NUCLEAR AND REACTOR CONSTANTS","volume":null,"pages":null},"PeriodicalIF":0.0,"publicationDate":"2020-12-26","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"84924030","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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