D. Plotnikov, A. L. Lobarev, Ivan Nikolayevich Krivoshein, P. B. Kuznetsov, A. 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":"D. Plotnikov, A. L. Lobarev, Ivan Nikolayevich Krivoshein, P. B. Kuznetsov, A. Ivanyuta","doi":"10.26583/npe.2022.3.03","DOIUrl":"https://doi.org/10.26583/npe.2022.3.03","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":37826,"journal":{"name":"Izvestiya Wysshikh Uchebnykh Zawedeniy, Yadernaya Energetika","volume":"99 1","pages":""},"PeriodicalIF":0.0,"publicationDate":"2022-09-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"86148416","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. V. Bedenko, I. Lutsik, A. Matyushin, S. D. Polozkov, V. Shmakov, D. Modestov, V. Prikhodko, A. Arzhannikov
{"title":"Hybrid Fusion-Fission Reactor Facility: Power Profiling","authors":"S. V. Bedenko, I. Lutsik, A. Matyushin, S. D. Polozkov, V. Shmakov, D. Modestov, V. Prikhodko, A. Arzhannikov","doi":"10.26583/npe.2022.3.04","DOIUrl":"https://doi.org/10.26583/npe.2022.3.04","url":null,"abstract":"","PeriodicalId":37826,"journal":{"name":"Izvestiya Wysshikh Uchebnykh Zawedeniy, Yadernaya Energetika","volume":"15 1","pages":""},"PeriodicalIF":0.0,"publicationDate":"2022-09-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"85621027","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 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":"V. Korobeynikov, V. V. Kolesov, Igor A. Ignatiev","doi":"10.26583/npe.2022.3.12","DOIUrl":"https://doi.org/10.26583/npe.2022.3.12","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":37826,"journal":{"name":"Izvestiya Wysshikh Uchebnykh Zawedeniy, Yadernaya Energetika","volume":"2 1","pages":""},"PeriodicalIF":0.0,"publicationDate":"2022-09-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"90358526","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}
E. Kuleshova, I. Fedotov, N. Stepanov, A. Frolov, D. A. Maltsev, D. Safonov
Nickel is an essential alloying element in steels used as structural materials in the most common nuclear power reactors of the VVER type. The paper considers reviews the results of structural studies of traditional and advanced materials of the vessels and internals of VVER-type reactors with high nickel contents in their compositions. It is shown that an increased nickel content (up to 5 wt.%) in the steels of VVER pressure vessels contributes to the formation of a more dispersed structure with a smaller size of substructural elements and an increased density of dislocations, as well as a higher volume density of carbide phases. The revealed features of the structure of the reactor pressure vessel steel with high nickel content have the prerequisites for improving the strength and viscoplastic properties due to the increased number of barriers both for the dislocation motion and brittle crack propagation. Using the example of materials for VVER internals, it is shown that the nickel content increased in them up to 25 wt.% contributes to an increase in the volume density of radiation defects (dislocation loops of various types) and radiation-induced phase precipitates (G-phase). As nickel increases from 10 to 25 wt.%, there is a tendency to reduce swelling, which contributes to less shape change of the components of the reactor vessel internals. At the same time, in the steel with the highest nickel content, the highest nickel content was found in the near-boundary regions of the matrix, which contributes to greater austenite stability and a lower probability of the formation of an embrittling α-phase. The data obtained in the work on the effect of nickel alloying on the steel structural phase state and service characteristics were used in the development of new materials for the vessels and internals of advanced reactors.
{"title":"The Role of Nickel in Forming a Structure Providing Increased Service Properties of Reactor Structural Materials","authors":"E. Kuleshova, I. Fedotov, N. Stepanov, A. Frolov, D. A. Maltsev, D. Safonov","doi":"10.26583/npe.2022.3.11","DOIUrl":"https://doi.org/10.26583/npe.2022.3.11","url":null,"abstract":"Nickel is an essential alloying element in steels used as structural materials in the most common nuclear power reactors of the VVER type. The paper considers reviews the results of structural studies of traditional and advanced materials of the vessels and internals of VVER-type reactors with high nickel contents in their compositions. It is shown that an increased nickel content (up to 5 wt.%) in the steels of VVER pressure vessels contributes to the formation of a more dispersed structure with a smaller size of substructural elements and an increased density of dislocations, as well as a higher volume density of carbide phases. The revealed features of the structure of the reactor pressure vessel steel with high nickel content have the prerequisites for improving the strength and viscoplastic properties due to the increased number of barriers both for the dislocation motion and brittle crack propagation. Using the example of materials for VVER internals, it is shown that the nickel content increased in them up to 25 wt.% contributes to an increase in the volume density of radiation defects (dislocation loops of various types) and radiation-induced phase precipitates (G-phase). As nickel increases from 10 to 25 wt.%, there is a tendency to reduce swelling, which contributes to less shape change of the components of the reactor vessel internals. At the same time, in the steel with the highest nickel content, the highest nickel content was found in the near-boundary regions of the matrix, which contributes to greater austenite stability and a lower probability of the formation of an embrittling α-phase. The data obtained in the work on the effect of nickel alloying on the steel structural phase state and service characteristics were used in the development of new materials for the vessels and internals of advanced reactors.","PeriodicalId":37826,"journal":{"name":"Izvestiya Wysshikh Uchebnykh Zawedeniy, Yadernaya Energetika","volume":"63 1","pages":""},"PeriodicalIF":0.0,"publicationDate":"2022-09-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"73989724","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}
Y. Kazansky, Nikita O. Kushnir, Ekaterina Sergeevna Khnykina
This paper considers the use of unconventional fuel in nuclear power reactors, using the example of a VVER-type unit, in order to find out the possibility of saving natural fissile uranium nuclei. Saving fissile uranium is one of the important tasks, the solution of which will give time for the development of a two-component nuclear power industry that will have no problems with fuel resources. However, at present, the reserves of cheap uranium can provide the existing level of global nuclear energy for only 80–100 years. The main components of this proposed fuel are 232Th and fissile isotopes of uranium: 235U (loaded) and 233U (produced from thorium). All the uranium isotopes and added 235U nuclei at the beginning of the campaign account for about 6% of the number of thorium nuclei and uranium isotopes. The abbreviated name of this fuel is TORUR-5. To keep fissionable nuclei in the fuel cycle after the spent fuel is unloaded, it is envisaged that all the heavy nuclei will be returned back to the reactor after they have been cleaned from fission fragments, i.e., the fuel cycle will be closed. At the same time, the principle of annual movement of fuel assemblies (as they burn up) is the same as in the existing VVER-1000 reactors. Using the Serpent software, a reactor model was built, the composition and dimensions of which were close to the parameters of the VVER-1000 serial unit. The main results of calculations were the quantitative compositions of isotopes annually loaded into the reactor as well as the amounts of 235U and thorium added also annually. The analysis of the obtained results allowed us to make the following conclusions. The annual reloading of 235U during the computation period is required almost at a constant level and, in comparison with uranium fuel, is about half as much. This is feasible for the following reasons. Part of the fissions of 235U is replaced by the fission of 233U produced from 232Th. In addition, fissionable nuclei are kept in the closed Th-U fuel cycle. This is the first “advantage” of the proposed fuel. TORUR-5 requires uranium enriched to at least 90%, the cost of which is several times higher than that of 3–5% enriched uranium. But since much less highly enriched uranium is required, the cost of fuel for a TORUR-5-fueled VVER-1000 reactor is significantly lower. This is the second “advantage” of the proposed fuel. The negative characteristic of TORUR-5, which requires further investigation, is that, after the initial loading, several uranium isotopes appear in the returned fuel, the total radioactivity of which, according to estimates, exceeds the radioactivity of traditional 3–5% enriched uranium fuel by several thousand times. At the same time, the radioactivity of discharged spent conventional fuel exceeds the radioactivity of fresh fuel by millions of times, and this problem has been solved at NPPs both organizationally and technically. Therefore, it will be necessary to develop a technology for loa
{"title":"Multiple Usage of Thorium-Based Fuel in a VVER-1000 Reactor","authors":"Y. Kazansky, Nikita O. Kushnir, Ekaterina Sergeevna Khnykina","doi":"10.26583/npe.2022.3.05","DOIUrl":"https://doi.org/10.26583/npe.2022.3.05","url":null,"abstract":"This paper considers the use of unconventional fuel in nuclear power reactors, using the example of a VVER-type unit, in order to find out the possibility of saving natural fissile uranium nuclei. Saving fissile uranium is one of the important tasks, the solution of which will give time for the development of a two-component nuclear power industry that will have no problems with fuel resources. However, at present, the reserves of cheap uranium can provide the existing level of global nuclear energy for only 80–100 years.\u0000 The main components of this proposed fuel are 232Th and fissile isotopes of uranium: 235U (loaded) and 233U (produced from thorium). All the uranium isotopes and added 235U nuclei at the beginning of the campaign account for about 6% of the number of thorium nuclei and uranium isotopes. The abbreviated name of this fuel is TORUR-5.\u0000 To keep fissionable nuclei in the fuel cycle after the spent fuel is unloaded, it is envisaged that all the heavy nuclei will be returned back to the reactor after they have been cleaned from fission fragments, i.e., the fuel cycle will be closed. At the same time, the principle of annual movement of fuel assemblies (as they burn up) is the same as in the existing VVER-1000 reactors.\u0000 Using the Serpent software, a reactor model was built, the composition and dimensions of which were close to the parameters of the VVER-1000 serial unit. The main results of calculations were the quantitative compositions of isotopes annually loaded into the reactor as well as the amounts of 235U and thorium added also annually. The analysis of the obtained results allowed us to make the following conclusions.\u0000 The annual reloading of 235U during the computation period is required almost at a constant level and, in comparison with uranium fuel, is about half as much. This is feasible for the following reasons. Part of the fissions of 235U is replaced by the fission of 233U produced from 232Th. In addition, fissionable nuclei are kept in the closed Th-U fuel cycle. This is the first “advantage” of the proposed fuel. TORUR-5 requires uranium enriched to at least 90%, the cost of which is several times higher than that of 3–5% enriched uranium.\u0000 But since much less highly enriched uranium is required, the cost of fuel for a TORUR-5-fueled VVER-1000 reactor is significantly lower. This is the second “advantage” of the proposed fuel.\u0000 The negative characteristic of TORUR-5, which requires further investigation, is that, after the initial loading, several uranium isotopes appear in the returned fuel, the total radioactivity of which, according to estimates, exceeds the radioactivity of traditional 3–5% enriched uranium fuel by several thousand times. At the same time, the radioactivity of discharged spent conventional fuel exceeds the radioactivity of fresh fuel by millions of times, and this problem has been solved at NPPs both organizationally and technically. Therefore, it will be necessary to develop a technology for loa","PeriodicalId":37826,"journal":{"name":"Izvestiya Wysshikh Uchebnykh Zawedeniy, Yadernaya Energetika","volume":"9 1","pages":""},"PeriodicalIF":0.0,"publicationDate":"2022-09-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"73428719","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. M. Gulina, Anastasiya Valerievna Merkun, V. P. Semishkin
{"title":"Probabilistic Estimation of the Residual Lifetime of the VVER Coolant Loop Elements in Aging Control","authors":"O. M. Gulina, Anastasiya Valerievna Merkun, V. P. Semishkin","doi":"10.26583/npe.2022.3.09","DOIUrl":"https://doi.org/10.26583/npe.2022.3.09","url":null,"abstract":"","PeriodicalId":37826,"journal":{"name":"Izvestiya Wysshikh Uchebnykh Zawedeniy, Yadernaya Energetika","volume":"292 2 1","pages":""},"PeriodicalIF":0.0,"publicationDate":"2022-09-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"73170173","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. Sorokin, J. Kuzina, R. Askhadullin, V. 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 Coolantsn for Nuclear and Thermonuclear Power Plants","authors":"A. Sorokin, J. Kuzina, R. Askhadullin, V. V. Alekseev","doi":"10.26583/npe.2022.3.01","DOIUrl":"https://doi.org/10.26583/npe.2022.3.01","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":37826,"journal":{"name":"Izvestiya Wysshikh Uchebnykh Zawedeniy, Yadernaya Energetika","volume":"87 1","pages":""},"PeriodicalIF":0.0,"publicationDate":"2022-09-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"83454861","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":"Ranking of Information Flows in the Technical Diagnostics Systems of the VVER-1200 Power Unit","authors":"N. Bocharova, A. Voronov, M. T. Slepov","doi":"10.26583/npe.2022.3.06","DOIUrl":"https://doi.org/10.26583/npe.2022.3.06","url":null,"abstract":"","PeriodicalId":37826,"journal":{"name":"Izvestiya Wysshikh Uchebnykh Zawedeniy, Yadernaya Energetika","volume":"32 1","pages":""},"PeriodicalIF":0.0,"publicationDate":"2022-09-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"87428466","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}
Sergei Terent’evich Leskin, V. Slobodchuk, Aleksandr Sergeevich Shelegov
{"title":"Computational Analysis of the Power Conversion Loop of a Nuclear Power Plant Unit with the Closed S-CO2 Brayton Cycle","authors":"Sergei Terent’evich Leskin, V. Slobodchuk, Aleksandr Sergeevich Shelegov","doi":"10.26583/npe.2022.3.02","DOIUrl":"https://doi.org/10.26583/npe.2022.3.02","url":null,"abstract":"","PeriodicalId":37826,"journal":{"name":"Izvestiya Wysshikh Uchebnykh Zawedeniy, Yadernaya Energetika","volume":"28 1","pages":""},"PeriodicalIF":0.0,"publicationDate":"2022-09-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"78206062","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}
Anatolij Aleksandrovich Kazantsev, O. Supotnitskaya, E. A. Ivanova, I. V. Moskovchenko, V. Timofeev, Nataliya Eduardovna Astakhova
{"title":"Radiolysis of the Coolant in the EGP-6 Reactor of the Bilibino NPP","authors":"Anatolij Aleksandrovich Kazantsev, O. Supotnitskaya, E. A. Ivanova, I. V. Moskovchenko, V. Timofeev, Nataliya Eduardovna Astakhova","doi":"10.26583/npe.2022.3.07","DOIUrl":"https://doi.org/10.26583/npe.2022.3.07","url":null,"abstract":"","PeriodicalId":37826,"journal":{"name":"Izvestiya Wysshikh Uchebnykh Zawedeniy, Yadernaya Energetika","volume":"81 1","pages":""},"PeriodicalIF":0.0,"publicationDate":"2022-09-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"90389667","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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