Denis A. Plotnikov, Aleksey L. Lobarev, Ivan N. Krivoshein, Pavel B. Kuznetsov, Anastasia N. Ivanyuta
The evolution of nuclear power is inseparably linked with the development of breakthrough solutions in the field of economic development of new territories. A pressing issue in this connection nowadays is generation of power for remote and hard-to-reach areas with decentralized power supply. To resolve this issue, JSC NIKIET is developing a version of the SHELF-M modular water-cooled water-moderated reactor facility as a source of power for offshore installations, including the Arctic coast areas, as well as localities with practically no power and transport infrastructure. One of the stages in justifying the safety of the reactor facility operation is to investigate the behavior of the reactor facility in dynamic transient modes at various power levels. To this end, a spatial dynamic model has been developed for the reactor facility with fuel and coolant temperature feedbacks. The dynamic model development process is a complex task that includes both preparation of constants for subsequent calculations and generation of the reactor neutronic and thermophysical models. The paper describes the development stages of the SHELF-M reactor facility spatial dynamic model and the results of coupled neutronic and thermophysical calculations for transients using the developed dynamic model of the reactor. Shim rod movement in the cold and hot states of the SHELF-M reactor facility is considered as transients.
{"title":"A spatial dynamic model of the SHELF-M reactor facility with fuel and coolant temperature feedbacks","authors":"Denis A. Plotnikov, Aleksey L. Lobarev, Ivan N. Krivoshein, Pavel B. Kuznetsov, Anastasia N. Ivanyuta","doi":"10.3897/nucet.9.102912","DOIUrl":"https://doi.org/10.3897/nucet.9.102912","url":null,"abstract":"The evolution of nuclear power is inseparably linked with the development of breakthrough solutions in the field of economic development of new territories. A pressing issue in this connection nowadays is generation of power for remote and hard-to-reach areas with decentralized power supply. To resolve this issue, JSC NIKIET is developing a version of the SHELF-M modular water-cooled water-moderated reactor facility as a source of power for offshore installations, including the Arctic coast areas, as well as localities with practically no power and transport infrastructure. One of the stages in justifying the safety of the reactor facility operation is to investigate the behavior of the reactor facility in dynamic transient modes at various power levels. To this end, a spatial dynamic model has been developed for the reactor facility with fuel and coolant temperature feedbacks. The dynamic model development process is a complex task that includes both preparation of constants for subsequent calculations and generation of the reactor neutronic and thermophysical models. The paper describes the development stages of the SHELF-M reactor facility spatial dynamic model and the results of coupled neutronic and thermophysical calculations for transients using the developed dynamic model of the reactor. Shim rod movement in the cold and hot states of the SHELF-M reactor facility is considered as transients.","PeriodicalId":100969,"journal":{"name":"Nuclear Energy and Technology","volume":"128 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2023-03-17","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"134940152","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}
Valery V. Korobeynikov, Valery V. Kolesov, Igor A. Ignatiev
The paper presents the results of studies on the burning of minor actinides (MA) extracted from SNF of thermal reactors in a BN-600 reactor, which uses the complete set of MAs instead of traditional nuclear fuel types: uranium and/or plutonium. The advantages of such approach to MA burning are that long-lived waste is recycled and energy is produced that can be used, e.g., to generate electricity. Besides, where, e.g., a reactor with uranium or MOX fuel is used for transmutation, apart from burning “foreign” minor actinides, it will additionally generate “its own” MAs. The studies have shown that such reactor can be efficient only if based on fast neutrons, which is due to the specific properties of the minor actinide neutron capture and fission cross-sections as compared with traditional fuel nuclides. The calculation results have shown rather a high rate of MA transmutation and burning in a reactor fueled with minor actinides.
{"title":"Computational simulation of minor actinide burning in a BN-600 reactor with fuel without uranium and plutonium","authors":"Valery V. Korobeynikov, Valery V. Kolesov, Igor A. Ignatiev","doi":"10.3897/nucet.9.102776","DOIUrl":"https://doi.org/10.3897/nucet.9.102776","url":null,"abstract":"The paper presents the results of studies on the burning of minor actinides (MA) extracted from SNF of thermal reactors in a BN-600 reactor, which uses the complete set of MAs instead of traditional nuclear fuel types: uranium and/or plutonium. The advantages of such approach to MA burning are that long-lived waste is recycled and energy is produced that can be used, e.g., to generate electricity. Besides, where, e.g., a reactor with uranium or MOX fuel is used for transmutation, apart from burning “foreign” minor actinides, it will additionally generate “its own” MAs. The studies have shown that such reactor can be efficient only if based on fast neutrons, which is due to the specific properties of the minor actinide neutron capture and fission cross-sections as compared with traditional fuel nuclides. The calculation results have shown rather a high rate of MA transmutation and burning in a reactor fueled with minor actinides.","PeriodicalId":100969,"journal":{"name":"Nuclear Energy and Technology","volume":"13 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2023-03-17","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"134940153","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}
Sergey N. Ivanov, Sergey I. Porollo, Sergey V. Shulepin, Yury D. Baranaev, Vladimir F. Timofeev, Yury V. Kharizomenov
Examinations of fuel elements with two different fuel compositions, U-Mo+Mg and UO 2 +Mg, irradiated in the AM reactor after their long-term storage do not reveal any visible defects on the surface of their outer claddings. However, in the fuel elements with U-Mo fuel, an increase in the diameter of the outer cladding is observed. This is most noticeable in the upper part of the fuel element. Storage of the fuel elements with UO 2 fuel for 15–22 years does not lead to a change in their diameter within the measurement accuracy. At the same time, metallographic studies have shown that on the external surface of the outer cladding and the internal surface of the inner cladding of the fuel elements with U-Mo+Mg and UO 2 +Mg fuel compositions, after long-term storage, defects are observed in the form of intergranular and irregular frontal corrosion, pits and pittings up to 20 µm deep. No interaction is found at the points of contact between the fuel claddings and the fuel composition of the layers. There is no noticeable decrease in the thickness of the outer and inner claddings of the fuel elements after long-term storage, nor does the thickness of the claddings at the locations of defects go beyond its minimum initial value, taking into account the technological tolerance for variations in thickness. It is noteworthy, however, that cracks are found in both types of fuel elements both in the fuel grains and in the magnesium matrix. As a result of long-term storage of the fuel elements with U-Mo fuel for 45–55 years, the mechanical properties of their outer claddings gradually degrade, due to which the plasticity of the cladding is significantly reduced.
{"title":"Examination of fuel elements irradiated in the reactor of the World’s First NPP after long-term storage","authors":"Sergey N. Ivanov, Sergey I. Porollo, Sergey V. Shulepin, Yury D. Baranaev, Vladimir F. Timofeev, Yury V. Kharizomenov","doi":"10.3897/nucet.9.102492","DOIUrl":"https://doi.org/10.3897/nucet.9.102492","url":null,"abstract":"Examinations of fuel elements with two different fuel compositions, U-Mo+Mg and UO 2 +Mg, irradiated in the AM reactor after their long-term storage do not reveal any visible defects on the surface of their outer claddings. However, in the fuel elements with U-Mo fuel, an increase in the diameter of the outer cladding is observed. This is most noticeable in the upper part of the fuel element. Storage of the fuel elements with UO 2 fuel for 15–22 years does not lead to a change in their diameter within the measurement accuracy. At the same time, metallographic studies have shown that on the external surface of the outer cladding and the internal surface of the inner cladding of the fuel elements with U-Mo+Mg and UO 2 +Mg fuel compositions, after long-term storage, defects are observed in the form of intergranular and irregular frontal corrosion, pits and pittings up to 20 µm deep. No interaction is found at the points of contact between the fuel claddings and the fuel composition of the layers. There is no noticeable decrease in the thickness of the outer and inner claddings of the fuel elements after long-term storage, nor does the thickness of the claddings at the locations of defects go beyond its minimum initial value, taking into account the technological tolerance for variations in thickness. It is noteworthy, however, that cracks are found in both types of fuel elements both in the fuel grains and in the magnesium matrix. As a result of long-term storage of the fuel elements with U-Mo fuel for 45–55 years, the mechanical properties of their outer claddings gradually degrade, due to which the plasticity of the cladding is significantly reduced.","PeriodicalId":100969,"journal":{"name":"Nuclear Energy and Technology","volume":"90 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2023-03-17","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"134940149","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}
Aleksandr P. Sorokin, Yuliya A. Kuzina, Radomir Sh. Ashadullin, Viktor V. Alekseev
It is shown that, as the result of developing alkali liquid metal coolants, including sodium, eutectic sodium-potassium alloy, lithium and cesium, the scientific basis has been established for their application in nuclear power. The paper presents data from investigations of thermophysical, neutronic and physicochemical properties and characteristics of various alkali liquid metal coolants, the content of solid-phase and dissolved impurities in coolants, mass transport of impurities in circulation circuits with alkali liquid metal coolants, development of systems for removal of impurities, and control of the content of impurities in alkali liquid metal coolants. Alkali liquid metal coolants are considered as a part of a system that includes a structural material in contact with the coolant, and a gas space that compensates for the thermal expansion of the coolant. The state of the system is defined by the physicochemical properties of the system’s components. And the coolant and the structural materials also represent subsystems consisting of a base material, a coolant and impurities contained both in the material and in the coolant. It has been shown that each alkali liquid metal coolant has its own set of impurities that define its technology. It depends on the physicochemical properties of the solution of the structural material impurities and components in the coolant. Objectives have been formulated for investigating further alkali liquid metal coolants, as stemming from the need to improve the efficiency, environmental friendliness, reliability and safety, and for extending the life of nuclear power plants in operation or under design. Alkali liquid metals are promising candidate materials for being used in thermonuclear power not only as the coolant but also as the tritium breeding medium. These include, first of all, lithium and its eutectic alloy with lead (17 at. % of lithium). The possibility for using lithium or a lithium-lead alloy as a coolant in the blanket of the international thermonuclear power reactor is compared.
{"title":"Study into the physical chemistry and technology of alkali liquid metal coolants for nuclear and thermonuclear power plants","authors":"Aleksandr P. Sorokin, Yuliya A. Kuzina, Radomir Sh. Ashadullin, Viktor V. Alekseev","doi":"10.3897/nucet.9.101761","DOIUrl":"https://doi.org/10.3897/nucet.9.101761","url":null,"abstract":"It is shown that, as the result of developing alkali liquid metal coolants, including sodium, eutectic sodium-potassium alloy, lithium and cesium, the scientific basis has been established for their application in nuclear power. The paper presents data from investigations of thermophysical, neutronic and physicochemical properties and characteristics of various alkali liquid metal coolants, the content of solid-phase and dissolved impurities in coolants, mass transport of impurities in circulation circuits with alkali liquid metal coolants, development of systems for removal of impurities, and control of the content of impurities in alkali liquid metal coolants. Alkali liquid metal coolants are considered as a part of a system that includes a structural material in contact with the coolant, and a gas space that compensates for the thermal expansion of the coolant. The state of the system is defined by the physicochemical properties of the system’s components. And the coolant and the structural materials also represent subsystems consisting of a base material, a coolant and impurities contained both in the material and in the coolant. It has been shown that each alkali liquid metal coolant has its own set of impurities that define its technology. It depends on the physicochemical properties of the solution of the structural material impurities and components in the coolant. Objectives have been formulated for investigating further alkali liquid metal coolants, as stemming from the need to improve the efficiency, environmental friendliness, reliability and safety, and for extending the life of nuclear power plants in operation or under design. Alkali liquid metals are promising candidate materials for being used in thermonuclear power not only as the coolant but also as the tritium breeding medium. These include, first of all, lithium and its eutectic alloy with lead (17 at. % of lithium). The possibility for using lithium or a lithium-lead alloy as a coolant in the blanket of the international thermonuclear power reactor is compared.","PeriodicalId":100969,"journal":{"name":"Nuclear Energy and Technology","volume":"59 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2023-03-17","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"135294910","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}
Sergey V. Bedenko, Igor O. Lutsik, Anton A. Matyushin, Sergey D. Polozkov, Vladimir M. Shmakov, Dmitry G. Modestov, Vadim V. Prikhodko, Andrey V. Arzhannikov
The current state of research in the field of nuclear and thermonuclear power aimed at creating power generation plants makes it possible to predict the further development of modern power industry in the direction hybrid reactor power plants. Such hybrid systems include a tokamak with reactor technologies, worked out in detail in Russia, and systems with an additional source of neutrons. Power generation plants using tokamaks and accelerators with the required level of proton energy will be of exceptionally large size and power, which will postpone their construction on an industrial scale to the distant future. The ongoing research is aimed at the development of small generation and has the prospect of entering the field of energy use in a shorter period. The hybrid reactor facility under study consists of an axisymmetric assembly of fuel blocks of a high-temperature gas-cooled reactor and a linear plasma source of additional neutrons. The paper demonstrates the results of optimization plasma-physical, thermophysical and gas-dynamic studies, the purpose of which is to level the distortions of the power density field, which are formed in the volume of the multiplicating part of the facility due to the pulsed operation of the plasma source of D-T-neutrons. The studies on increasing the “brightness” of the source and modeling its operating modes were carried out using the DOL and PRIZMA programs. The thermophysical optimization and gas-dynamic calculations were performed using the verified SERPENT and FloEFD software codes. The calculations were made on a high-performance cluster of the Tomsk Polytechnic University.
{"title":"Fusion-fission hybrid reactor facility: power profiling","authors":"Sergey V. Bedenko, Igor O. Lutsik, Anton A. Matyushin, Sergey D. Polozkov, Vladimir M. Shmakov, Dmitry G. Modestov, Vadim V. Prikhodko, Andrey V. Arzhannikov","doi":"10.3897/nucet.9.102781","DOIUrl":"https://doi.org/10.3897/nucet.9.102781","url":null,"abstract":"The current state of research in the field of nuclear and thermonuclear power aimed at creating power generation plants makes it possible to predict the further development of modern power industry in the direction hybrid reactor power plants. Such hybrid systems include a tokamak with reactor technologies, worked out in detail in Russia, and systems with an additional source of neutrons. Power generation plants using tokamaks and accelerators with the required level of proton energy will be of exceptionally large size and power, which will postpone their construction on an industrial scale to the distant future. The ongoing research is aimed at the development of small generation and has the prospect of entering the field of energy use in a shorter period. The hybrid reactor facility under study consists of an axisymmetric assembly of fuel blocks of a high-temperature gas-cooled reactor and a linear plasma source of additional neutrons. The paper demonstrates the results of optimization plasma-physical, thermophysical and gas-dynamic studies, the purpose of which is to level the distortions of the power density field, which are formed in the volume of the multiplicating part of the facility due to the pulsed operation of the plasma source of D-T-neutrons. The studies on increasing the “brightness” of the source and modeling its operating modes were carried out using the DOL and PRIZMA programs. The thermophysical optimization and gas-dynamic calculations were performed using the verified SERPENT and FloEFD software codes. The calculations were made on a high-performance cluster of the Tomsk Polytechnic University.","PeriodicalId":100969,"journal":{"name":"Nuclear Energy and Technology","volume":"20 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2023-03-17","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"135349887","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}
Fast reactors with heavy liquid metal coolant (lead or eutectic bismuth-lead alloy) are one of the most advanced technologies capable to address the accumulated world nuclear energy issues. This innovative power technology is being developed in Russia, the USA, China and the European Union. Russia is the leader since it has focused on this topic for a number of decades. First concrete started to be poured in June 2021 to form the foundation of the Russian BREST-OD-300 lead cooled reactor scheduled to be started up in 2026. Attention is also given to the development status of the Chinese CLEAR reactor series. A large scope of R&D has been undertaken, and large-scale nonnuclear experimental facilities are under construction. International Euro-US consortiums for the development of the ALFRED, PLFR and MYRRHA reactors do not expect any unsolvable technical issues either and are currently formulating requirements to the test facilities and candidate materials and technologies required for further activities.
{"title":"Heavy liquid metal cooled fast reactors: peculiarities and development status of the major projects","authors":"A. Orlov, B. Gabaraev","doi":"10.3897/nucet.9.90993","DOIUrl":"https://doi.org/10.3897/nucet.9.90993","url":null,"abstract":"Fast reactors with heavy liquid metal coolant (lead or eutectic bismuth-lead alloy) are one of the most advanced technologies capable to address the accumulated world nuclear energy issues. This innovative power technology is being developed in Russia, the USA, China and the European Union. Russia is the leader since it has focused on this topic for a number of decades. First concrete started to be poured in June 2021 to form the foundation of the Russian BREST-OD-300 lead cooled reactor scheduled to be started up in 2026. Attention is also given to the development status of the Chinese CLEAR reactor series. A large scope of R&D has been undertaken, and large-scale nonnuclear experimental facilities are under construction. International Euro-US consortiums for the development of the ALFRED, PLFR and MYRRHA reactors do not expect any unsolvable technical issues either and are currently formulating requirements to the test facilities and candidate materials and technologies required for further activities.","PeriodicalId":100969,"journal":{"name":"Nuclear Energy and Technology","volume":"28 1","pages":""},"PeriodicalIF":0.0,"publicationDate":"2023-03-08","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"84740769","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}
Modern uranium enrichment facilities can simultaneously use several raw materials as feed, including natural uranium, regenerated uranium obtained as a result of SNF reprocessing, or depleted uranium (all in the form of uranium hexafluoride). As the output of the separating cascade, several types of enriched uranium product with different levels of enrichment can be fabricated simultaneously. The paper proposes a methodology, absent in literature, for calculating the cost of each enriched uranium product in multi-stream separating cascades. The proposed methodology uses standard definitions of the isotopic value of feed and product stream and the Peierls-Dirac separation potential. Numerical calculations of the cost of enriched uranium products for three production problems are provided as examples of the methodology effectiveness: 1) involvement of depleted uranium hexafluoride (DUHF) in fabrication of enriched uranium product; 2) simultaneous fabrication of two enriched products; 3) use of depleted uranium to reduce the cost of the product with a higher enrichment level out of two (as applied, e.g., to advanced tolerant fuel). It has been shown that partial additions of DUHF as feed for a multi-product separating cascade make it possible to reduce the cost of a product with a higher level of enrichment; with the current market prices for natural uranium and separative work, there is a range of tails assays in which it is more profitable to enrich DUHF rather than natural uranium.
{"title":"Calculation of the cost of enriched uranium products in multi-stream cascades of enrichment process","authors":"E. Semenov, V. V. Kharitonov","doi":"10.3897/nucet.9.100752","DOIUrl":"https://doi.org/10.3897/nucet.9.100752","url":null,"abstract":"Modern uranium enrichment facilities can simultaneously use several raw materials as feed, including natural uranium, regenerated uranium obtained as a result of SNF reprocessing, or depleted uranium (all in the form of uranium hexafluoride). As the output of the separating cascade, several types of enriched uranium product with different levels of enrichment can be fabricated simultaneously. The paper proposes a methodology, absent in literature, for calculating the cost of each enriched uranium product in multi-stream separating cascades. The proposed methodology uses standard definitions of the isotopic value of feed and product stream and the Peierls-Dirac separation potential. Numerical calculations of the cost of enriched uranium products for three production problems are provided as examples of the methodology effectiveness: 1) involvement of depleted uranium hexafluoride (DUHF) in fabrication of enriched uranium product; 2) simultaneous fabrication of two enriched products; 3) use of depleted uranium to reduce the cost of the product with a higher enrichment level out of two (as applied, e.g., to advanced tolerant fuel). It has been shown that partial additions of DUHF as feed for a multi-product separating cascade make it possible to reduce the cost of a product with a higher level of enrichment; with the current market prices for natural uranium and separative work, there is a range of tails assays in which it is more profitable to enrich DUHF rather than natural uranium.","PeriodicalId":100969,"journal":{"name":"Nuclear Energy and Technology","volume":"74 1","pages":""},"PeriodicalIF":0.0,"publicationDate":"2023-03-08","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"80386259","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}
The authors consider phenomena that have signs of ‘cliff edge effects’ according to the definitions of the IAEA and NP-001-15: (1) degradation of the protective barrier (fuel rod claddings in surface boiling mode with the deposition of impurities and borates on their surface and heating of the claddings) and (2) departure from nucleate boiling (DNB) on the fuel rod claddings. Despite the fact that the first phenomenon was previously unknown, the safety of the power unit is ensured by the decisions adopted in the project. The DNB was studied and measures were taken in the project to prevent it under normal operating conditions and anticipated operational occurrences. The protection against the DNB is also obviously ensured by reducing the reactor power due to the control systems and reactor scram. These phenomena do not reach the state of ‘cliff edge effects’ (according to the terminology of the IAEA and federal NPs of the Russian Federation) and are prevented at the initial stages. For a small-size reactor using dispersive fuel, it is possible to provide self-protection against the DNB, namely, due to partial washout of the fuel with the insertion of negative reactivity, followed by a decrease in power and termination of the crisis.
{"title":"‘Cliff edge effects’ in safety justification and operation of NPP units","authors":"Valentin M. Makhin, Alexander K. Podshibyakin","doi":"10.3897/nucet.9.100755","DOIUrl":"https://doi.org/10.3897/nucet.9.100755","url":null,"abstract":"The authors consider phenomena that have signs of ‘cliff edge effects’ according to the definitions of the IAEA and NP-001-15: (1) degradation of the protective barrier (fuel rod claddings in surface boiling mode with the deposition of impurities and borates on their surface and heating of the claddings) and (2) departure from nucleate boiling (DNB) on the fuel rod claddings. Despite the fact that the first phenomenon was previously unknown, the safety of the power unit is ensured by the decisions adopted in the project. The DNB was studied and measures were taken in the project to prevent it under normal operating conditions and anticipated operational occurrences. The protection against the DNB is also obviously ensured by reducing the reactor power due to the control systems and reactor scram. These phenomena do not reach the state of ‘cliff edge effects’ (according to the terminology of the IAEA and federal NPs of the Russian Federation) and are prevented at the initial stages. For a small-size reactor using dispersive fuel, it is possible to provide self-protection against the DNB, namely, due to partial washout of the fuel with the insertion of negative reactivity, followed by a decrease in power and termination of the crisis.","PeriodicalId":100969,"journal":{"name":"Nuclear Energy and Technology","volume":"15 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2023-03-08","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"136180947","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}
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 constraints","authors":"Oleg V. Marchenko, Sergei V. Solomin","doi":"10.3897/nucet.9.100754","DOIUrl":"https://doi.org/10.3897/nucet.9.100754","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":100969,"journal":{"name":"Nuclear Energy and Technology","volume":"1 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2023-03-08","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"136180955","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, Ilya A. Bessonov, Artem D. Lezov, S. V. Chalpanov, Maxim S. Smykov, G. I. Skvortsov, V. A. Klimova
Formation of radioactive waste (RW) is specific to the NPP operation. Liquid radioactive waste (LRW) forms in the process of the reactor plant operation, and in decontamination of equipment, rooms and overalls. The radionuclides found mostly in vat residues are 134, 137 Cs in the form of ions and 60Co and 54Mn isotopes in the form of chelates including substances used for equipment decontamination. Among the well-known conditioning techniques, selective sorption provides for the greatest reduction of LRW amounts. The efficiency of using the amount of the filter material can be increased by supplying the treated medium simultaneously to several sorbent layers. The paper presents computer simulation results for three proposed options of improved container filter designs for ion-selective treatment differing in the ways used both to separate the treated water flows and to deliver these to the sorbent layers. The improved efficiency of the sorption processes in the proposed designs was estimated using computer simulation in SolidWorks Flow Simulation. Three sorbent grades from NPP Eksorb were used for the study. A series of experimental studies of the flow through the sorbent layer was undertaken to determine the hydraulic resistance of the studied samples. The obtained experimental data was added to the Solidworks Flow Simulation engineering database for simulation of the earlier presented designs. Representative parameters of the flow inside of container filters were obtained as a result of the simulation.
{"title":"Computational and experimental studies into the hydrodynamic operation conditions of container filters for ion-selective treatment","authors":"O. L. Tashlykov, Ilya A. Bessonov, Artem D. Lezov, S. V. Chalpanov, Maxim S. Smykov, G. I. Skvortsov, V. A. Klimova","doi":"10.3897/nucet.8.94105","DOIUrl":"https://doi.org/10.3897/nucet.8.94105","url":null,"abstract":"Formation of radioactive waste (RW) is specific to the NPP operation. Liquid radioactive waste (LRW) forms in the process of the reactor plant operation, and in decontamination of equipment, rooms and overalls. The radionuclides found mostly in vat residues are 134, 137 Cs in the form of ions and 60Co and 54Mn isotopes in the form of chelates including substances used for equipment decontamination. Among the well-known conditioning techniques, selective sorption provides for the greatest reduction of LRW amounts. The efficiency of using the amount of the filter material can be increased by supplying the treated medium simultaneously to several sorbent layers.\u0000 The paper presents computer simulation results for three proposed options of improved container filter designs for ion-selective treatment differing in the ways used both to separate the treated water flows and to deliver these to the sorbent layers. The improved efficiency of the sorption processes in the proposed designs was estimated using computer simulation in SolidWorks Flow Simulation.\u0000 Three sorbent grades from NPP Eksorb were used for the study. A series of experimental studies of the flow through the sorbent layer was undertaken to determine the hydraulic resistance of the studied samples. The obtained experimental data was added to the Solidworks Flow Simulation engineering database for simulation of the earlier presented designs. Representative parameters of the flow inside of container filters were obtained as a result of the simulation.","PeriodicalId":100969,"journal":{"name":"Nuclear Energy and Technology","volume":"10 1","pages":""},"PeriodicalIF":0.0,"publicationDate":"2022-09-27","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"88571000","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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