The aim of the work is forecasting the development of nuclear power in Russia and the world for the period up to 2050 under various scenarios of constraints on carbon dioxide emissions. A brief comparative analysis of the main characteristics of the forecasts of the International Energy Agency (IEA) and the International Renewable Energy Agency (IRENA) has been carried out. Additionally, calculations were performed using the mathematical models of the world energy system GEM and GEM-Dyn developed at the ISEM SB RAS. The optimal ratio of nuclear and non-nuclear energy sources has been determined. It is shown that nuclear power, including nuclear power plants operating on a closed fuel cycle, along with renewable energy sources, is an effective technology that can solve the problem of reducing carbon dioxide emissions. Calculations have shown that in the sustainable development scenario, the capacity of nuclear power plants in Russia in the period from 2020 to 2050 can increase by 2.7 times, and their share in electricity generation can reach 21–25% in 2030 and 26–35% in 2050. The average annual growth rate (for 30 years) of the installed capacity of nuclear power plants in Russia in the sustainable development scenario is 3.1% compared to 2.7% for the world as a whole. In the GEM and GEM-Dyn calculations performed by the authors, the scale of nuclear energy use turned out to be about 30% higher than in the scenarios of the International Energy Agency due to more conservative estimates of the opportunities for improving the performance of renewable energy sources and taking into account the need to back-up their capacity.
{"title":"The Development Options of Nuclear Power Under Carbon Dioxide Emissions Constrains","authors":"O. Marchenko, S. Solomin","doi":"10.26583/npe.2022.2.01","DOIUrl":"https://doi.org/10.26583/npe.2022.2.01","url":null,"abstract":"The aim of the work is forecasting the development of nuclear power in Russia and the world for the period up to 2050 under various scenarios of constraints on carbon dioxide emissions. A brief comparative analysis of the main characteristics of the forecasts of the International Energy Agency (IEA) and the International Renewable Energy Agency (IRENA) has been carried out. Additionally, calculations were performed using the mathematical models of the world energy system GEM and GEM-Dyn developed at the ISEM SB RAS. The optimal ratio of nuclear and non-nuclear energy sources has been determined. It is shown that nuclear power, including nuclear power plants operating on a closed fuel cycle, along with renewable energy sources, is an effective technology that can solve the problem of reducing carbon dioxide emissions. Calculations have shown that in the sustainable development scenario, the capacity of nuclear power plants in Russia in the period from 2020 to 2050 can increase by 2.7 times, and their share in electricity generation can reach 21–25% in 2030 and 26–35% in 2050. The average annual growth rate (for 30 years) of the installed capacity of nuclear power plants in Russia in the sustainable development scenario is 3.1% compared to 2.7% for the world as a whole. In the GEM and GEM-Dyn calculations performed by the authors, the scale of nuclear energy use turned out to be about 30% higher than in the scenarios of the International Energy Agency due to more conservative estimates of the opportunities for improving the performance of renewable energy sources and taking into account the need to back-up their capacity.","PeriodicalId":37826,"journal":{"name":"Izvestiya Wysshikh Uchebnykh Zawedeniy, Yadernaya Energetika","volume":"11 1","pages":""},"PeriodicalIF":0.0,"publicationDate":"2022-06-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"89508407","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}
O. L. Tashlykov, I. A. Bessonov, Artyom Dmitrievich Lezov, S. V. Chalpanov, Maksim Sergeevich Smykov, G. I. Skvortsov, V. A. Klimova
{"title":"Computational and Experimental Studies of Hydrodynamic Operating Conditions of Filter Containers for Ion-Selective Purification","authors":"O. L. Tashlykov, I. A. Bessonov, Artyom Dmitrievich Lezov, S. V. Chalpanov, Maksim Sergeevich Smykov, G. I. Skvortsov, V. A. Klimova","doi":"10.26583/npe.2022.2.06","DOIUrl":"https://doi.org/10.26583/npe.2022.2.06","url":null,"abstract":"","PeriodicalId":37826,"journal":{"name":"Izvestiya Wysshikh Uchebnykh Zawedeniy, Yadernaya Energetika","volume":"320 1","pages":""},"PeriodicalIF":0.0,"publicationDate":"2022-06-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"75228929","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}
A. Belov, V. Bereznev, Galina Sergeevna Blokhina, D. P. Veprev, D. Koltashev, V. S. Potapov, O. Chertovskikh, Aleksey Vladimirovich Shershov
The paper presents the results of the development of the EUCLID/V1 integrated dynamic code designed to analyze and justify the safety of fast neutron reactor facilities with a liquid-metal coolant, in terms of simulating the reactor campaign. The relevance of this study lies in the need to simulate the behavior of the core at any time during the campaign. It lets us to obtain a full dataset for subsequent simulations of the reactor dynamic conditions (including transient states or accidents). The authors have developed a fuel archive to store calculated data in HDF5 format, created a computational model editor to generate input data in the fuel archive format, and also provided an example of computing the campaign of a lead-cooled fast reactor for three core design models shown in this paper. The main array of fuel assemblies was simulated as a single unit in the first model, as three units in the second model, and in the third every single assembly was unique. In addition, the authors have shown changes in the total masses of actinides in the core, revealed that the different core models have an insignificant effect on the evolution of the total masses of actinides, and given the fuel assembly burnup values for the three core models. For the third model, the largest difference between the minimum and maximum burnup values was obtained with an almost identical average over the fuel assemblies. The reactivity margin over time for the three core models was presented. It was shown that the values and behavior of the reactivity margin during the three micro-campaigns are almost equal. From the fourth to the sixth cycle, the reactivity margin value for the third core model was lower than for the first and the second ones. Finally, the authors conclude that it is desirable to evaluate the behavior of the reactivity margin for lead-cooled fast reactor campaigns based on the detailed model of the core.
{"title":"Simulating a Lead&Cooled Reactor Campaign Using the EUCLID/V1 Code","authors":"A. Belov, V. Bereznev, Galina Sergeevna Blokhina, D. P. Veprev, D. Koltashev, V. S. Potapov, O. Chertovskikh, Aleksey Vladimirovich Shershov","doi":"10.26583/npe.2022.2.13","DOIUrl":"https://doi.org/10.26583/npe.2022.2.13","url":null,"abstract":"The paper presents the results of the development of the EUCLID/V1 integrated dynamic code designed to analyze and justify the safety of fast neutron reactor facilities with a liquid-metal coolant, in terms of simulating the reactor campaign. The relevance of this study lies in the need to simulate the behavior of the core at any time during the campaign. It lets us to obtain a full dataset for subsequent simulations of the reactor dynamic conditions (including transient states or accidents). The authors have developed a fuel archive to store calculated data in HDF5 format, created a computational model editor to generate input data in the fuel archive format, and also provided an example of computing the campaign of a lead-cooled fast reactor for three core design models shown in this paper. The main array of fuel assemblies was simulated as a single unit in the first model, as three units in the second model, and in the third every single assembly was unique. In addition, the authors have shown changes in the total masses of actinides in the core, revealed that the different core models have an insignificant effect on the evolution of the total masses of actinides, and given the fuel assembly burnup values for the three core models. For the third model, the largest difference between the minimum and maximum burnup values was obtained with an almost identical average over the fuel assemblies. The reactivity margin over time for the three core models was presented. It was shown that the values and behavior of the reactivity margin during the three micro-campaigns are almost equal. From the fourth to the sixth cycle, the reactivity margin value for the third core model was lower than for the first and the second ones. Finally, the authors conclude that it is desirable to evaluate the behavior of the reactivity margin for lead-cooled fast reactor campaigns based on the detailed model of the core.","PeriodicalId":37826,"journal":{"name":"Izvestiya Wysshikh Uchebnykh Zawedeniy, Yadernaya Energetika","volume":"74 1","pages":""},"PeriodicalIF":0.0,"publicationDate":"2022-06-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"86926192","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}
P. A. Pisarev, V. V. Kolesov, Dmitry Valerievich Kolesov
Computational studies have been carried out showing the complex time dependence of uncertainties in nuclear concentrations of various nuclides arising from the propagation of the neutron flux density errors in the burnup calculation process in cells with different neutron spectra on the above errors. It is found that these uncertainties not only depend on the burnup time in a complex way, but also depend on the spectrum of the cell. The variants of the cell with thermal and fast neutron spectra were considered. The calculations were performed using the VisualBurnOut program (Kolesov et al. 2009), which makes it possible to estimate these uncertainties arising due to errors in the input parameters of the burnup problem (reaction rates, neutron flux density, etc.). The influence of the number of calculated burnup points on the results of burnup calculations by the Monte Carlo method was investigated. Uncertainties arising in nuclear concentrations at intermediate calculation steps due to errors in nuclear concentrations appearing at the previous step were taken into account in the calculations.
计算研究表明，在不同中子谱的细胞中，燃耗计算过程中中子通量密度误差的传播所引起的各种核素核浓度的不确定性对上述误差具有复杂的时间依赖性。研究发现，这些不确定性不仅与燃耗时间有关，而且与电池的光谱有关。考虑了热中子能谱和快中子能谱的变化。使用VisualBurnOut程序(Kolesov et al. 2009)进行计算，这使得可以估计由于燃耗问题(反应速率，中子通量密度等)的输入参数错误而产生的这些不确定性。研究了燃耗点计算个数对用蒙特卡罗方法计算燃耗结果的影响。在计算中考虑了中间计算步骤中由于前一步出现的核浓度误差而引起的核浓度不确定性。
{"title":"The Effect of Errors in the Neutron Flux Density on the Uncertainties of Nuclear Concentrations of Nuclides Arising During the Calculation of Fuel Burnup in Cells with Different Neutron Spectra","authors":"P. A. Pisarev, V. V. Kolesov, Dmitry Valerievich Kolesov","doi":"10.26583/npe.2022.2.12","DOIUrl":"https://doi.org/10.26583/npe.2022.2.12","url":null,"abstract":"Computational studies have been carried out showing the complex time dependence of uncertainties in nuclear concentrations of various nuclides arising from the propagation of the neutron flux density errors in the burnup calculation process in cells with different neutron spectra on the above errors.\u0000 It is found that these uncertainties not only depend on the burnup time in a complex way, but also depend on the spectrum of the cell. The variants of the cell with thermal and fast neutron spectra were considered.\u0000 The calculations were performed using the VisualBurnOut program (Kolesov et al. 2009), which makes it possible to estimate these uncertainties arising due to errors in the input parameters of the burnup problem (reaction rates, neutron flux density, etc.).\u0000 The influence of the number of calculated burnup points on the results of burnup calculations by the Monte Carlo method was investigated. Uncertainties arising in nuclear concentrations at intermediate calculation steps due to errors in nuclear concentrations appearing at the previous step were taken into account in the calculations.","PeriodicalId":37826,"journal":{"name":"Izvestiya Wysshikh Uchebnykh Zawedeniy, Yadernaya Energetika","volume":"115 1","pages":""},"PeriodicalIF":0.0,"publicationDate":"2022-06-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"80284715","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}
{"title":"Problems of Decommissioning Fast Reactors and Ways of their Solution on the Basis of the BR-10 Research Reactor","authors":"V. Smykov","doi":"10.26583/npe.2022.2.09","DOIUrl":"https://doi.org/10.26583/npe.2022.2.09","url":null,"abstract":"","PeriodicalId":37826,"journal":{"name":"Izvestiya Wysshikh Uchebnykh Zawedeniy, Yadernaya Energetika","volume":"14 1","pages":""},"PeriodicalIF":0.0,"publicationDate":"2022-06-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"86097483","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}
Denis Alekseevich Soloviev, A. Khachatryan, E. V. Chernov, Rashdan Talal Al
This paper presents the results of numerical studies of various algorithms for suppression of xenon offset and power distribution oscillations in the core of a VVER-1200 reactor. The purpose of the research is to select an algorithm that minimizes the amount of liquid radioactive wastes during water exchange in the primary circuit of a nuclear power plant. For this, several algorithms for xenon oscillations suppression were considered. The first algorithm considered was an algorithm for suppression of xenon oscillations, which uses regulation due to AWP only, without utilization of any additional regulation. The second algorithm considered was an algorithm based on the use both AWP and boron regulation. In this algorithm suppression of xenon oscillations was carried out with the help of accelerated initiation of the work of the AWP by changing the boric acid concentration with constant second circuit pressure of the NPP and by utilization of the second control rods group. Last algorithm considered was algorithm based on the use of temperature control for accelerated initiation of the work of the AWP. In this algorithm, xenon oscillations suppression was carried out by changing coolant temperature at the reactor inlet caused by pressure change in the secondary circuit in the normal operation margins, and by involving the second group of control rods. It was shown that the best way to suppress xenon offset and power distribution oscillations in terms of minimization of radioactive liquid wastes amount is the algorithm with accelerated initiation of the AWP due to temperature regulation, with elimination of temperature regulation after minimizing of current axial offset value deviation from the nominal one.
{"title":"Investigation of Algorithms for Suppressing Xenon Oscillations in a VVER-1200 Reactor","authors":"Denis Alekseevich Soloviev, A. Khachatryan, E. V. Chernov, Rashdan Talal Al","doi":"10.26583/npe.2022.2.04","DOIUrl":"https://doi.org/10.26583/npe.2022.2.04","url":null,"abstract":"This paper presents the results of numerical studies of various algorithms for suppression of xenon offset and power distribution oscillations in the core of a VVER-1200 reactor. The purpose of the research is to select an algorithm that minimizes the amount of liquid radioactive wastes during water exchange in the primary circuit of a nuclear power plant. For this, several algorithms for xenon oscillations suppression were considered. The first algorithm considered was an algorithm for suppression of xenon oscillations, which uses regulation due to AWP only, without utilization of any additional regulation.\u0000 The second algorithm considered was an algorithm based on the use both AWP and boron regulation. In this algorithm suppression of xenon oscillations was carried out with the help of accelerated initiation of the work of the AWP by changing the boric acid concentration with constant second circuit pressure of the NPP and by utilization of the second control rods group.\u0000 Last algorithm considered was algorithm based on the use of temperature control for accelerated initiation of the work of the AWP. In this algorithm, xenon oscillations suppression was carried out by changing coolant temperature at the reactor inlet caused by pressure change in the secondary circuit in the normal operation margins, and by involving the second group of control rods.\u0000 It was shown that the best way to suppress xenon offset and power distribution oscillations in terms of minimization of radioactive liquid wastes amount is the algorithm with accelerated initiation of the AWP due to temperature regulation, with elimination of temperature regulation after minimizing of current axial offset value deviation from the nominal one.","PeriodicalId":37826,"journal":{"name":"Izvestiya Wysshikh Uchebnykh Zawedeniy, Yadernaya Energetika","volume":"56 1 1","pages":""},"PeriodicalIF":0.0,"publicationDate":"2022-06-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"77866520","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}
P. Alekseev, G. E. Lazarenko, V. A. Linnik, A. P. Pyshko
As a result of the analytical studies of the designs of thermionic reactor-converters, four groups of technical solutions have been identified that differ in the method of heat transfer from the fuel to the emitters of the thermionic converter: one option with direct in-core transfer (combining the fuel cladding with the emitter) and three options with thermionic converters taken away from the reactor core, in which case the heat is removed either by heat pipes (common or individual for each fuel element) or is arranged based on the principle of a steam chamber. The article describes the advantages and disadvantages for each of these methods. It is shown that at present the most developed design remains the version with in-core power conversion and, in the future it will be based on the steam chamber since the ingress of gaseous fission products into the inter-electrode gap as well as the influence of fuel swelling on the inter-electrode gap size are excluded and it ensures constant temperature and heat flux density on the surface of all emitters of the thermionic converters, which makes it possible to select the optimal operating parameters for them. A model of a thermionic reactor-converter with a steam chamber containing a reactor core and a zone of thermionic converters has been developed in which the fuel element of the reactor core and the power generating channels of the thermionic converter are separated in space, covered with a capillary porous structure and interconnected by a honeycomb capillary porous spacer plate to provide for circulation of the liquid metal coolant and to let its steam pass through. Neutronic calculations have demonstrated the possibility of a duration for the reactor campaign in excess of ten years following the nuclear safety regulations when a gadolinium oxide coating is applied to the surface of the fuel rods and the reactor vessel in the area of the reactor core. The assessment of thermal and electrical parameters shows that, due to the constant temperature and heat flux density on the surface of all emitters and optimization of the power conversion process for all the thermionic converters, one can expect to reach the maximum efficiency of 20%.
{"title":"The Concept of a Thermionic Reactor-Converter with Evaporative Heat Transfer","authors":"P. Alekseev, G. E. Lazarenko, V. A. Linnik, A. P. Pyshko","doi":"10.26583/npe.2022.1.10","DOIUrl":"https://doi.org/10.26583/npe.2022.1.10","url":null,"abstract":"As a result of the analytical studies of the designs of thermionic reactor-converters, four groups of technical solutions have been identified that differ in the method of heat transfer from the fuel to the emitters of the thermionic converter: one option with direct in-core transfer (combining the fuel cladding with the emitter) and three options with thermionic converters taken away from the reactor core, in which case the heat is removed either by heat pipes (common or individual for each fuel element) or is arranged based on the principle of a steam chamber.\u0000 The article describes the advantages and disadvantages for each of these methods. It is shown that at present the most developed design remains the version with in-core power conversion and, in the future it will be based on the steam chamber since the ingress of gaseous fission products into the inter-electrode gap as well as the influence of fuel swelling on the inter-electrode gap size are excluded and it ensures constant temperature and heat flux density on the surface of all emitters of the thermionic converters, which makes it possible to select the optimal operating parameters for them.\u0000 A model of a thermionic reactor-converter with a steam chamber containing a reactor core and a zone of thermionic converters has been developed in which the fuel element of the reactor core and the power generating channels of the thermionic converter are separated in space, covered with a capillary porous structure and interconnected by a honeycomb capillary porous spacer plate to provide for circulation of the liquid metal coolant and to let its steam pass through.\u0000 Neutronic calculations have demonstrated the possibility of a duration for the reactor campaign in excess of ten years following the nuclear safety regulations when a gadolinium oxide coating is applied to the surface of the fuel rods and the reactor vessel in the area of the reactor core.\u0000 The assessment of thermal and electrical parameters shows that, due to the constant temperature and heat flux density on the surface of all emitters and optimization of the power conversion process for all the thermionic converters, one can expect to reach the maximum efficiency of 20%.","PeriodicalId":37826,"journal":{"name":"Izvestiya Wysshikh Uchebnykh Zawedeniy, Yadernaya Energetika","volume":"115 1","pages":""},"PeriodicalIF":0.0,"publicationDate":"2022-01-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"85257978","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}
V. Melnikov, T. Bokova, V. V. Ivanov, A. Marov, Nataliya Aleksandrovna Lobaeva, A. S. Kvashennikov, P. Bokov, N. Volkov
{"title":"Experimental Study of a Microwave Reflex-Radar Level Gauge of a Liquid-Metal Coolant","authors":"V. Melnikov, T. Bokova, V. V. Ivanov, A. Marov, Nataliya Aleksandrovna Lobaeva, A. S. Kvashennikov, P. Bokov, N. Volkov","doi":"10.26583/npe.2022.1.07","DOIUrl":"https://doi.org/10.26583/npe.2022.1.07","url":null,"abstract":"","PeriodicalId":37826,"journal":{"name":"Izvestiya Wysshikh Uchebnykh Zawedeniy, Yadernaya Energetika","volume":"52 1","pages":""},"PeriodicalIF":0.0,"publicationDate":"2022-01-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"90576550","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}
R. Irkimbekov, A. Vurim, S. V. Bedenko, A. Surayev, G. Vityuk
IVG.1M is a research pressurized water reactor designed to use high-enriched fuel. As part of the core conversion program, the reactor will be switched to a new low-enriched composite uranium fuel. Further operation of the reactor is determined by the availability of fresh fuel to replace the core after the next campaign and the possibility of ensuring safe storage of irradiated spent nuclear fuel (SNF) unloaded from the core. The SNF storage conditions are assessed in terms of ensuring nuclear and radiation safety. Radiation safety of the research reactor fuel storage is achieved, first of all, by solving problems of protection against γ-radiation, while neutron radiation, as a rule, is not considered due to its significantly lower intensity compared to γ-radiation. As for the new low-enriched fuel of the IVG.1M reactor, which is characterized by a set of elements with low and medium atomic masses, on which the (α, n) reaction is possible, the assessment of the neutron component is a necessary procedure to ensure safe fuel storage. The authors of the article propose a procedure for calculating the neutron component of the radiation characteristics of fresh and irradiated composite fuel of the IVG.1M reactor, and also estimate the (α, n)-component. The results of the research will be useful in selecting SNF storage and transportation technologies as well as in providing scientific justification for the possibility of using neutron radiation to control burnup. The research was carried out using verified computational codes MCNP5 and Sources-4C, high-precision experimental EXFOR and computational ENDSF data, as well as evaluated nuclear data libraries.
{"title":"Neutron Background of Composite Low-Enriched Uranium Fuel of the IVG.1M Research Reactor","authors":"R. Irkimbekov, A. Vurim, S. V. Bedenko, A. Surayev, G. Vityuk","doi":"10.26583/npe.2022.1.11","DOIUrl":"https://doi.org/10.26583/npe.2022.1.11","url":null,"abstract":"IVG.1M is a research pressurized water reactor designed to use high-enriched fuel. As part of the core conversion program, the reactor will be switched to a new low-enriched composite uranium fuel. Further operation of the reactor is determined by the availability of fresh fuel to replace the core after the next campaign and the possibility of ensuring safe storage of irradiated spent nuclear fuel (SNF) unloaded from the core. The SNF storage conditions are assessed in terms of ensuring nuclear and radiation safety.\u0000 Radiation safety of the research reactor fuel storage is achieved, first of all, by solving problems of protection against γ-radiation, while neutron radiation, as a rule, is not considered due to its significantly lower intensity compared to γ-radiation. As for the new low-enriched fuel of the IVG.1M reactor, which is characterized by a set of elements with low and medium atomic masses, on which the (α, n) reaction is possible, the assessment of the neutron component is a necessary procedure to ensure safe fuel storage.\u0000 The authors of the article propose a procedure for calculating the neutron component of the radiation characteristics of fresh and irradiated composite fuel of the IVG.1M reactor, and also estimate the (α, n)-component. The results of the research will be useful in selecting SNF storage and transportation technologies as well as in providing scientific justification for the possibility of using neutron radiation to control burnup.\u0000 The research was carried out using verified computational codes MCNP5 and Sources-4C, high-precision experimental EXFOR and computational ENDSF data, as well as evaluated nuclear data libraries.","PeriodicalId":37826,"journal":{"name":"Izvestiya Wysshikh Uchebnykh Zawedeniy, Yadernaya Energetika","volume":"1 1","pages":""},"PeriodicalIF":0.0,"publicationDate":"2022-01-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"86867851","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}
S. Soloviev, D. G. Zaryugin, S. G. Kalyakin, Sergei Terent’evich Leskin
In paper considered Small Modular Reactors (SMR) main advantages of design such as: In paper considered possible areas of SMR application, including consumer demands, which are as follows: power supply of remote (Arctic) territories, replacement (renovation) of old coal generation, production of high"potential heat and hydrogen for industrial consumers and other applications. The necessity of development and implementation of a new technological platform for nuclear energy based on SMRs is shown in order to implement the global decarbonization of the world economy by means of significant expansion of nuclear energy technologies application. This technological platform should be developed in addition to the currently developing one based on the closed nuclear fuel cycle with fast reactors (solving the problem of fuel supply and waste disposal) and also developing technological platform of controlled thermonuclear fusion(solving the problem of global energy supply in the long term). The new technological platform should be created on the bases of broad international cooperation with creation of international consortiums. An experimental testing facility (research reactor)is proposed to be created for the development of captive hydrogen (heat)production technologies for industrial consumers as well as other technologies for the application of small modular reactors.
{"title":"Identifying the Key Development Areas for Small Modular Reactors","authors":"S. Soloviev, D. G. Zaryugin, S. G. Kalyakin, Sergei Terent’evich Leskin","doi":"10.26583/npe.2022.1.02","DOIUrl":"https://doi.org/10.26583/npe.2022.1.02","url":null,"abstract":"In paper considered Small Modular Reactors (SMR) main advantages of design such as: In paper considered possible areas of SMR application, including consumer demands, which are as follows: power supply of remote (Arctic) territories, replacement (renovation) of old coal generation, production of high\"potential heat and hydrogen for industrial consumers and other applications. The necessity of development and implementation of a new technological platform for nuclear energy based on SMRs is shown in order to implement the global decarbonization of the world economy by means of significant expansion of nuclear energy technologies application. This technological platform should be developed in addition to the currently developing one based on the closed nuclear fuel cycle with fast reactors (solving the problem of fuel supply and waste disposal) and also developing technological platform of controlled thermonuclear fusion(solving the problem of global energy supply in the long term). The new technological platform should be created on the bases of broad international cooperation with creation of international consortiums. An experimental testing facility (research reactor)is proposed to be created for the development of captive hydrogen (heat)production technologies for industrial consumers as well as other technologies for the application of small modular reactors.","PeriodicalId":37826,"journal":{"name":"Izvestiya Wysshikh Uchebnykh Zawedeniy, Yadernaya Energetika","volume":"76 1","pages":""},"PeriodicalIF":0.0,"publicationDate":"2022-01-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"80823728","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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