In VVER reactor plants, it is impossible to completely exclude the appearance of loose, loosely fixed and foreign objects in the main circulation circuit. Operational experience shows that early detection and estimation of the parameters of such incidents can provide the time required to eliminate or minimize damage to the main equipment of the reactor plant. For this reason, most modern power units with pressurized water reactors (PWR, VVER) are equipped with a loose parts monitoring system (LPMS). At the units under construction, these systems are laid down as standard ones; the power units put into commercial operation in the Soviet period were also equipped with them. The requirements for them are established by international standards. Ongoing research work in this area is aimed at determining the root cause of the acoustic anomaly and the localization of its epicenter. Also, no less significant are the works aimed at determining the mass of a loose object (LO). The most precise definition of this parameter will make it possible to have an idea of the nature of the LO before its withdrawal from the primary circuit and to conclude about whether this object is accidentally found or it is a detached part of the steam generators, main circulation pumps, internal devices or shut-off and control valves.
{"title":"Experience of using loose parts monitoring systems at Novovoronezh NPP","authors":"A. Voronov, M. T. Slepov","doi":"10.3897/nucet.8.94106","DOIUrl":"https://doi.org/10.3897/nucet.8.94106","url":null,"abstract":"In VVER reactor plants, it is impossible to completely exclude the appearance of loose, loosely fixed and foreign objects in the main circulation circuit. Operational experience shows that early detection and estimation of the parameters of such incidents can provide the time required to eliminate or minimize damage to the main equipment of the reactor plant. For this reason, most modern power units with pressurized water reactors (PWR, VVER) are equipped with a loose parts monitoring system (LPMS). At the units under construction, these systems are laid down as standard ones; the power units put into commercial operation in the Soviet period were also equipped with them. The requirements for them are established by international standards. Ongoing research work in this area is aimed at determining the root cause of the acoustic anomaly and the localization of its epicenter. Also, no less significant are the works aimed at determining the mass of a loose object (LO). The most precise definition of this parameter will make it possible to have an idea of the nature of the LO before its withdrawal from the primary circuit and to conclude about whether this object is accidentally found or it is a detached part of the steam generators, main circulation pumps, internal devices or shut-off and control valves.","PeriodicalId":100969,"journal":{"name":"Nuclear Energy and Technology","volume":"23 1","pages":""},"PeriodicalIF":0.0,"publicationDate":"2022-09-27","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"80571367","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, Natalia A. Lobaeva, A. S. Kvashennikov, P. Bokov, N. Volkov
The article presents the results of work aimed at solving the problem of measuring the coolant level in miscellaneous tanks of liquid-metal-cooled reactor plants, mainly of an integral layout with a free level of the primary coolant. The choice of relevant measuring means and methods is limited by the extreme parameters of the liquid metal coolant (LMC) and operating conditions. Traditional measuring means are practically unsuitable; therefore, measuring the HLMC level is a complex technical task. Based on this review, they propose and describe a method of pulsed microwave reflectometry as the most promising in terms of combining the characteristics of reliability, accuracy and ease of use. The results of the experimental study demonstrated the efficiency of the level gauge, which worked according to this method, for measuring the level of lead-bismuth coolant in the control tank under conditions close to natural ones. An analysis of the results confirmed the possibility of using this method to control the level of melts of various metals as applied to HLMC reactor plants. Using the device for measuring the level, which works according to the proposed method, it is possible to control the level of melt of various metals in tanks in real time without the need to move various parts of the sensitive element of the level gauge while maintaining the tightness of the circuit. This device is applicable for various nuclear power plants, accelerator-controlled systems, research reactors and experimental facilities with liquid metal coolants.
{"title":"Experimental study of using microwave reflex-radar level gauges for liquid metal coolants","authors":"V. Melnikov, T. Bokova, V. V. Ivanov, A. Marov, Natalia A. Lobaeva, A. S. Kvashennikov, P. Bokov, N. Volkov","doi":"10.3897/nucet.8.94540","DOIUrl":"https://doi.org/10.3897/nucet.8.94540","url":null,"abstract":"The article presents the results of work aimed at solving the problem of measuring the coolant level in miscellaneous tanks of liquid-metal-cooled reactor plants, mainly of an integral layout with a free level of the primary coolant. The choice of relevant measuring means and methods is limited by the extreme parameters of the liquid metal coolant (LMC) and operating conditions. Traditional measuring means are practically unsuitable; therefore, measuring the HLMC level is a complex technical task. Based on this review, they propose and describe a method of pulsed microwave reflectometry as the most promising in terms of combining the characteristics of reliability, accuracy and ease of use. The results of the experimental study demonstrated the efficiency of the level gauge, which worked according to this method, for measuring the level of lead-bismuth coolant in the control tank under conditions close to natural ones. An analysis of the results confirmed the possibility of using this method to control the level of melts of various metals as applied to HLMC reactor plants.\u0000 Using the device for measuring the level, which works according to the proposed method, it is possible to control the level of melt of various metals in tanks in real time without the need to move various parts of the sensitive element of the level gauge while maintaining the tightness of the circuit. This device is applicable for various nuclear power plants, accelerator-controlled systems, research reactors and experimental facilities with liquid metal coolants.","PeriodicalId":100969,"journal":{"name":"Nuclear Energy and Technology","volume":"125 1","pages":""},"PeriodicalIF":0.0,"publicationDate":"2022-09-27","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"86773010","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}
N. V. Gorin, V. Kuchinov, Andrey V. Krivtsov, A. Orlov, V. V. Shidlovskiy, Daria Borisovna Matveeva
The inevitability of switching to carbon-free energy to withstand the climate change is no longer disputed by anyone today. There is no alternative to this, and the scientific community is forming an appropriate understanding of the need for the development of nuclear energy as carbon free energy source. Solutions are already being discussed at the level of the President and the Government of Russia. In this regard, the article shows that such a solution is possible only based on a new technological platform – two-component nuclear power with the development of technologies of fast reactors with a closed fuel cycle. At the same time the prevailing view in the public opinion of Russia, and not only in it, is that climate change problem can be solved only at the expense of solar and wind energy. This attitude needs to be changed, because without the understanding and support of society, it is impossible to achieve a wide spread of fast reactors with closed fuel cycle technologies. It is concluded that in order to promote a new technological platform in commercial energy and ensure the export prospects of fast reactors of Russian design with closed nuclear fuel cycle facilities, it is necessary to attract representatives of business circles and large energy businesses to the number of supporters of such development by demonstrating the profitability of solutions in the medium and long term, implemented in the case of the use of Russian technologies of fast reactors with the closure of the nuclear fuel cycle.
{"title":"Export prospects of fast reactors desined in Russia with closed nuclear fuel cycle facilities","authors":"N. V. Gorin, V. Kuchinov, Andrey V. Krivtsov, A. Orlov, V. V. Shidlovskiy, Daria Borisovna Matveeva","doi":"10.3897/nucet.8.80757","DOIUrl":"https://doi.org/10.3897/nucet.8.80757","url":null,"abstract":"The inevitability of switching to carbon-free energy to withstand the climate change is no longer disputed by anyone today. There is no alternative to this, and the scientific community is forming an appropriate understanding of the need for the development of nuclear energy as carbon free energy source. Solutions are already being discussed at the level of the President and the Government of Russia. In this regard, the article shows that such a solution is possible only based on a new technological platform – two-component nuclear power with the development of technologies of fast reactors with a closed fuel cycle. At the same time the prevailing view in the public opinion of Russia, and not only in it, is that climate change problem can be solved only at the expense of solar and wind energy. This attitude needs to be changed, because without the understanding and support of society, it is impossible to achieve a wide spread of fast reactors with closed fuel cycle technologies. It is concluded that in order to promote a new technological platform in commercial energy and ensure the export prospects of fast reactors of Russian design with closed nuclear fuel cycle facilities, it is necessary to attract representatives of business circles and large energy businesses to the number of supporters of such development by demonstrating the profitability of solutions in the medium and long term, implemented in the case of the use of Russian technologies of fast reactors with the closure of the nuclear fuel cycle.","PeriodicalId":100969,"journal":{"name":"Nuclear Energy and Technology","volume":"24 1","pages":""},"PeriodicalIF":0.0,"publicationDate":"2022-09-20","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"82364481","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}
Until the mid-1950s, scientists and engineers at the Obninsk Institute of Physics and Power Engineering (IPPE) had worked out a number of unique projects that served as the foundation for the development of domestic and world nuclear power engineering. The list of these projects includes, in particular, TES-3, the first mobile nuclear power plant, which has become a symbol of small-scale nuclear power engineering, a historical achievement of Russian scientists, and part of the heritage of the City of Peaceful Atom. TES-3, a demonstration and experimental plant, being one of the possible nuclear power sources for remote areas, was a mobile power-generating unit consisting of four tracked platforms with a reactor unit equipped with a water-cooled and water-moderated reactor with a 1.5 MW turbogenerator. The “self-propelled uranium-fueled machine” was created in record-breaking time due to the scale and cooperation of the project participants under the scientific guidance of the Laboratory V staff. The plant showed reliability in operation, good controllability, safety and maintainability. Over the entire operating period in the power generation mode, TES-3 worked for about 1300 hours without any radiation accidents. After the completion of the first fuel campaign in 1965, the reactor was shut down, but the idea of mobile low-capacity large-component nuclear power plants was further developed in the form of mobile nuclear power plants of the next generation.
{"title":"TES-3 – transportable nuclear power plant mounted on self-propelled tracked vehicles","authors":"N. Naumenko, I. M. Mokhireva","doi":"10.3897/nucet.8.89356","DOIUrl":"https://doi.org/10.3897/nucet.8.89356","url":null,"abstract":"Until the mid-1950s, scientists and engineers at the Obninsk Institute of Physics and Power Engineering (IPPE) had worked out a number of unique projects that served as the foundation for the development of domestic and world nuclear power engineering. The list of these projects includes, in particular, TES-3, the first mobile nuclear power plant, which has become a symbol of small-scale nuclear power engineering, a historical achievement of Russian scientists, and part of the heritage of the City of Peaceful Atom.\u0000 TES-3, a demonstration and experimental plant, being one of the possible nuclear power sources for remote areas, was a mobile power-generating unit consisting of four tracked platforms with a reactor unit equipped with a water-cooled and water-moderated reactor with a 1.5 MW turbogenerator. The “self-propelled uranium-fueled machine” was created in record-breaking time due to the scale and cooperation of the project participants under the scientific guidance of the Laboratory V staff. The plant showed reliability in operation, good controllability, safety and maintainability. Over the entire operating period in the power generation mode, TES-3 worked for about 1300 hours without any radiation accidents. After the completion of the first fuel campaign in 1965, the reactor was shut down, but the idea of mobile low-capacity large-component nuclear power plants was further developed in the form of mobile nuclear power plants of the next generation.","PeriodicalId":100969,"journal":{"name":"Nuclear Energy and Technology","volume":"32 1","pages":""},"PeriodicalIF":0.0,"publicationDate":"2022-06-28","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"74232790","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}
After three quarters of a century using nuclear fission to produce energy, Nuclear Reactor Safety and Risk constitutes an established technological sector. A key feature is continuous updating following new discoveries and progress in knowledge, resulting in extensive and elaborate safety methodologies, which are still not internationally accepted, generally applicable or technically consistent. Each country developed its own methods, guides, traditions and requirements to deal with evolving design, safety, siting and licensing issues. There is a clear parallel in societal risk perception between nuclear radiation exposure in accidents and viral infection in pandemics and the fear of the “unknown”. Unfortunately, over the last 20–30 years the declining introduction of electricity by nuclear fission in the countries that contributed most to its earliest development also has broken the bond between new scientific advancements and improvements of existing safety methodologies. By looking at the origins and fundaments of nuclear technology, we consider the following topics of both deterministic and probabilistic interest: a) Loss of Coolant analysis; b) nuclear fuel accident performance weaknesses; c) role of containment and ultimate heat sinks; d) residual risk and emergency system deployment, and e) independent and risk informed decision making assessment. As a key outcome, we propose modifying the traditional licensing methodology, and the use of active and/or passive systems by being subsumed into a broader Engineered Safety Features Management process. Furthermore, we emphasize the need of connecting the As Low As Reasonably Achievable principle with the analyses to demonstrate the safety of nuclear installations minimizing the need for excessive “paper” safety analyses and licensing efforts.
{"title":"Innovation needs in nuclear reactor safety and risk","authors":"F. D’Auria, R. Duffey","doi":"10.3897/nucet.8.82296","DOIUrl":"https://doi.org/10.3897/nucet.8.82296","url":null,"abstract":"After three quarters of a century using nuclear fission to produce energy, Nuclear Reactor Safety and Risk constitutes an established technological sector. A key feature is continuous updating following new discoveries and progress in knowledge, resulting in extensive and elaborate safety methodologies, which are still not internationally accepted, generally applicable or technically consistent. Each country developed its own methods, guides, traditions and requirements to deal with evolving design, safety, siting and licensing issues. There is a clear parallel in societal risk perception between nuclear radiation exposure in accidents and viral infection in pandemics and the fear of the “unknown”. Unfortunately, over the last 20–30 years the declining introduction of electricity by nuclear fission in the countries that contributed most to its earliest development also has broken the bond between new scientific advancements and improvements of existing safety methodologies. By looking at the origins and fundaments of nuclear technology, we consider the following topics of both deterministic and probabilistic interest: a) Loss of Coolant analysis; b) nuclear fuel accident performance weaknesses; c) role of containment and ultimate heat sinks; d) residual risk and emergency system deployment, and e) independent and risk informed decision making assessment. As a key outcome, we propose modifying the traditional licensing methodology, and the use of active and/or passive systems by being subsumed into a broader Engineered Safety Features Management process. Furthermore, we emphasize the need of connecting the As Low As Reasonably Achievable principle with the analyses to demonstrate the safety of nuclear installations minimizing the need for excessive “paper” safety analyses and licensing efforts.","PeriodicalId":100969,"journal":{"name":"Nuclear Energy and Technology","volume":"1 1","pages":""},"PeriodicalIF":0.0,"publicationDate":"2022-06-27","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"73254171","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 transmutation process of three long-lived fission products (79Se, 99Tc and 107Pd) in the experimental fast reactor JOYO is postulated. The possibility of increasing the transmutation rate utilizing the high neutron flux present in the JOYO reactor by loading neutron-moderating subassemblies in the reflector zone has been investigated. A cluster of reflector subassemblies was replaced with beryllium or zirconium hydride (ZrH1.65) moderated subassemblies. These moderated subassemblies surrounded one central test subassembly that would contain the three long-lived fission products (LLFP) simultaneous and without isotopic separation. ChainSolver 2.34 code is used to calculate the transmutation rates. In this study, the new characteristics of LLFP transmutation in a fast reactor using moderator materials were shown for future applications.
{"title":"Evaluation of transmutation rate of some LLFP in experimental fast reactor JOYO","authors":"N. Amrani, A. Boucenna, A. Galahom","doi":"10.3897/nucet.8.78428","DOIUrl":"https://doi.org/10.3897/nucet.8.78428","url":null,"abstract":"A transmutation process of three long-lived fission products (79Se, 99Tc and 107Pd) in the experimental fast reactor JOYO is postulated. The possibility of increasing the transmutation rate utilizing the high neutron flux present in the JOYO reactor by loading neutron-moderating subassemblies in the reflector zone has been investigated. A cluster of reflector subassemblies was replaced with beryllium or zirconium hydride (ZrH1.65) moderated subassemblies. These moderated subassemblies surrounded one central test subassembly that would contain the three long-lived fission products (LLFP) simultaneous and without isotopic separation. ChainSolver 2.34 code is used to calculate the transmutation rates. In this study, the new characteristics of LLFP transmutation in a fast reactor using moderator materials were shown for future applications.","PeriodicalId":100969,"journal":{"name":"Nuclear Energy and Technology","volume":"91 1","pages":""},"PeriodicalIF":0.0,"publicationDate":"2022-06-27","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"80855185","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}
In terms of nuclear raw materials, the issue of involving thorium in the fuel cycle is hardly very relevant. However, in view of the large-scale nuclear power development, the use of thorium seems to be quite natural and reasonable. The substitution of traditional uranium-plutonium fuel for uranium-thorium fuel in fast neutron reactors will significantly reduce the production of minor actinides, which will make it attractive for the transmutation of long-lived radioactive isotopes of americium, curium and neptunium that have already been and are still being accumulated. Due to the absence of uranium-233 in nature, the use of thorium in the nuclear power industry requires a closed fuel cycle. At the initial stage of developing the uranium-thorium cycle, it is proposed to use uranium-235 instead of uranium-233 as nuclear fuel. Studies have been carried out on the transmutation of minor actinides in a fast neutron reactor in which the uranium-thorium cycle is implemented. Several options for the structure of the core of such a reactor have been considered. It has been shown that heterogeneous placement of americium leads to higher rates of its transmutation than homogeneous placement does.
{"title":"Comparison of the minor actinide transmutation efficiency in models of a fast neutron uranium-thorium fueled reactor","authors":"V. Korobeinikov, V. Kolesov, A. Mikhalev","doi":"10.3897/nucet.8.82757","DOIUrl":"https://doi.org/10.3897/nucet.8.82757","url":null,"abstract":"In terms of nuclear raw materials, the issue of involving thorium in the fuel cycle is hardly very relevant. However, in view of the large-scale nuclear power development, the use of thorium seems to be quite natural and reasonable. The substitution of traditional uranium-plutonium fuel for uranium-thorium fuel in fast neutron reactors will significantly reduce the production of minor actinides, which will make it attractive for the transmutation of long-lived radioactive isotopes of americium, curium and neptunium that have already been and are still being accumulated.\u0000 Due to the absence of uranium-233 in nature, the use of thorium in the nuclear power industry requires a closed fuel cycle. At the initial stage of developing the uranium-thorium cycle, it is proposed to use uranium-235 instead of uranium-233 as nuclear fuel.\u0000 Studies have been carried out on the transmutation of minor actinides in a fast neutron reactor in which the uranium-thorium cycle is implemented. Several options for the structure of the core of such a reactor have been considered. It has been shown that heterogeneous placement of americium leads to higher rates of its transmutation than homogeneous placement does.","PeriodicalId":100969,"journal":{"name":"Nuclear Energy and Technology","volume":"78 1","pages":""},"PeriodicalIF":0.0,"publicationDate":"2022-03-18","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"86566094","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}
Aleksey V. Balovnev, V. K. Davydov, A. P. Zhirnov, Andrey V. Moiseev, E. Soldatov
The development of nuclear power with fast reactors is associated with the implementation of a closed nuclear fuel cycle (CNFC). In this regard, one actual task is to simulate the stages of the fuel cycle with study of the neutron-physical characteristics of the core. The design of a reactor for operation in the closed nuclear fuel cycle mode is impossible without the using of verified and certified software packages for calculating fast reactors, capable of simulating all stages of the operation of the reactor facility and the fuel cycle. For the calculations, the FACT-BR software package was used, which has all the necessary capabilities to simulate the operation of the reactor in the closed nuclear fuel cycle mode, taking into account the stages of fuel storage and refabrication. The article presents a technique for modeling the fuel cycle, implemented for the operation of fast reactors with a lead coolant. To demonstrate methodology, a closed nuclear fuel cycle was simulated for the BREST-OD-300 and BR-1200 reactors for the design life. The article describes the scenarios in which the calculation of the burnup of reactor was carried out. In the considered scenarios, it is assumed that the unloading of fuel at the end of the micro campaign is conducted according to the maximum burnup. During the computational modeling the ranges of changes in fuel density and enrichment, reactivity margin, breeding ratio and isotopic composition of plutonium were determined.
{"title":"Simulating the fuel cycle of a lead-cooled fast reactor","authors":"Aleksey V. Balovnev, V. K. Davydov, A. P. Zhirnov, Andrey V. Moiseev, E. Soldatov","doi":"10.3897/nucet.8.83062","DOIUrl":"https://doi.org/10.3897/nucet.8.83062","url":null,"abstract":"The development of nuclear power with fast reactors is associated with the implementation of a closed nuclear fuel cycle (CNFC). In this regard, one actual task is to simulate the stages of the fuel cycle with study of the neutron-physical characteristics of the core. The design of a reactor for operation in the closed nuclear fuel cycle mode is impossible without the using of verified and certified software packages for calculating fast reactors, capable of simulating all stages of the operation of the reactor facility and the fuel cycle. For the calculations, the FACT-BR software package was used, which has all the necessary capabilities to simulate the operation of the reactor in the closed nuclear fuel cycle mode, taking into account the stages of fuel storage and refabrication. The article presents a technique for modeling the fuel cycle, implemented for the operation of fast reactors with a lead coolant. To demonstrate methodology, a closed nuclear fuel cycle was simulated for the BREST-OD-300 and BR-1200 reactors for the design life. The article describes the scenarios in which the calculation of the burnup of reactor was carried out. In the considered scenarios, it is assumed that the unloading of fuel at the end of the micro campaign is conducted according to the maximum burnup. During the computational modeling the ranges of changes in fuel density and enrichment, reactivity margin, breeding ratio and isotopic composition of plutonium were determined.","PeriodicalId":100969,"journal":{"name":"Nuclear Energy and Technology","volume":"23 1","pages":""},"PeriodicalIF":0.0,"publicationDate":"2022-03-18","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"82433876","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}
M. Trofimov, Yevgeny G. Murachev, A. Rogoza, Nikolay D. Yegupov
The paper considers the construction of a mathematical model for an electrohydraulic system to control automatically the Т-63-13,0/0,25 product manufactured by JSC Kaluga Turbine Plant. Mathematical simulation of control systems makes it possible to improve considerably the quality of control, that is, the accuracy and reliability of such systems, as well as to accelerate greatly the development and calculation of the control system and the parameters of its individual components. The T-63-13,0/0,25 mathematical model of the ASTCS allows estimating the effects of design parameters during any load dropping (in a range of 0 to 100%) and the quality of control for the monitored parameters both in the process of operation as part of an isolated power system (generator output, frequency) and an integrated power system (generator output). A mathematical representation has been developed in the model for the control units, the T-63-13,0/0,25 product model, and the electronic controlling part of each of the control units. It has been proposed that pulse-width modulation be used to control the synchronous motors which makes it possible to control the synchronous machine shaft speed by changing the supply voltage frequency. To this end, the control system’s model uses a frequency converter which is proposed to be used in the real control system. The developed control system with one adjustable steam extraction in the T-63-13,0/0,25 steam turbine is coupled and autonomous, that is, each of the two meters for the turbine’s controlled parameters has effect on both steam distribution systems such that a deviation for one of the controlled parameters does not lead to excitations in the other.
{"title":"Mathematical simulation of an automatic steam turbine control system","authors":"M. Trofimov, Yevgeny G. Murachev, A. Rogoza, Nikolay D. Yegupov","doi":"10.3897/nucet.8.83146","DOIUrl":"https://doi.org/10.3897/nucet.8.83146","url":null,"abstract":"The paper considers the construction of a mathematical model for an electrohydraulic system to control automatically the Т-63-13,0/0,25 product manufactured by JSC Kaluga Turbine Plant. Mathematical simulation of control systems makes it possible to improve considerably the quality of control, that is, the accuracy and reliability of such systems, as well as to accelerate greatly the development and calculation of the control system and the parameters of its individual components. The T-63-13,0/0,25 mathematical model of the ASTCS allows estimating the effects of design parameters during any load dropping (in a range of 0 to 100%) and the quality of control for the monitored parameters both in the process of operation as part of an isolated power system (generator output, frequency) and an integrated power system (generator output). A mathematical representation has been developed in the model for the control units, the T-63-13,0/0,25 product model, and the electronic controlling part of each of the control units. It has been proposed that pulse-width modulation be used to control the synchronous motors which makes it possible to control the synchronous machine shaft speed by changing the supply voltage frequency. To this end, the control system’s model uses a frequency converter which is proposed to be used in the real control system. The developed control system with one adjustable steam extraction in the T-63-13,0/0,25 steam turbine is coupled and autonomous, that is, each of the two meters for the turbine’s controlled parameters has effect on both steam distribution systems such that a deviation for one of the controlled parameters does not lead to excitations in the other.","PeriodicalId":100969,"journal":{"name":"Nuclear Energy and Technology","volume":"40 1","pages":""},"PeriodicalIF":0.0,"publicationDate":"2022-03-18","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"75714016","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. S. Shchepin, A. M. Koshcheev, Ivan V. Kuznetsov, M. Kalenova, I. M. Melnikova
The paper investigates the process of regeneration of a liquid metal medium used in the pyroelectrochemical reprocessing of spent mixed uranium-plutonium nitride fuel produced by a fast neutron reactor. The investigation concerns the interaction of liquid cadmium with sludge formed during the anodic dissolution of ceramic nitride pellets in a 3LiCl-2KCl melt medium as well as the possibility of its purification by filtration from individual metal fission products. Anode sludge is represented by fission products of the platinum group, zirconium, molybdenum and technetium. It was determined by scanning electron microscopy that the metal product is composed of several intergrowth phases. It was found that upon contact of a polymetallic alloy simulating anode sludge with a melt, the liquid metal phase is saturated to 0.025 wt% of Pd, 0.01 wt% of Rh for 50 hours at 500 °C, while zirconium forms an insoluble dispersed intermetallic compound ZrCd3. Powders of molybdenum and technetium, which are not wetted with cadmium, can be completely removed using a filter mesh of plain weaving of the P-200 type. It is also possible to remove zirconium from anodic cadmium by filtration. The filtration efficiency of ruthenium and palladium powders did not exceed 54.3 and 13.1 wt%, respectively, due to partial dissolution and thinning of particles, which will lead to saturation of the liquid metal phase and the need to purify it by alternative methods.
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