Andrey V. Kaplienko, B. Gabaraev, Yuriy S. Cherepnin
Thermal Power Plants (ThPP), along with transport facilities, are the major sources for industrial emissions of carbon dioxide (CO2) believed to be responsible for the greenhouse effect leading to overheating of the lower atmosphere. In the opinion of many scientists, there is a threshold value of the average atmospheric temperature exceeding, which entails the potential for the development of irreversible processes threatening the existence of humankind. To avoid this danger, governments in nearly 200 countries have chosen voluntarily to reach net-zero CO2 emissions by 2050. Renewable Energy Sources (RES), including wind and solar power plants, have been selected as substitutes for ThPPs. However, energy systems based on RES only need to be multiply redundant in terms of installed capacity due to their efficiency being heavily dependent on daily, seasonal and weather factors, leave alone the scale of the required material resources (metals, polymers, concrete, glass, etc.). The major drawback of such energy systems is, however, the RES common-cause failure, e.g., in the event of a global volcanic eruption, when no energy-security requirement can be met to provide energy for satisfying the most vital needs. A need is fully evident for furnishing such energy system with another power source to be not dependent on the event that has caused the mass RES failure. With the net-zero-carbon requirement taken into account, Nuclear Power (NP) appears to be the best option in this respect. Modern NP does not however fully suits this role due to its inherent drawbacks (limited fuel resources, pending Spent Nuclear Fuel (SNF) and RadioActive Wastes (RAW) handling and nuclear-material nonproliferation issues). A potential solution to these drawbacks is a two-component NP technology in a closed nuclear-fuel cycle currently in the process of development. In Russia, where the greatest progress has been achieved in this field of development, under construction is a pilot and demonstration energy complex with the BREST-OD-300 nuclear unit expected to be started up in 2026–2027. Another promising designs to be developed are Small Modular Reactors (SMRs) / Small Nuclear Power Plants (SNPPs).
{"title":"Potential role of nuclear power in a carbon-free world","authors":"Andrey V. Kaplienko, B. Gabaraev, Yuriy S. Cherepnin","doi":"10.3897/nucet.10.121449","DOIUrl":"https://doi.org/10.3897/nucet.10.121449","url":null,"abstract":"Thermal Power Plants (ThPP), along with transport facilities, are the major sources for industrial emissions of carbon dioxide (CO2) believed to be responsible for the greenhouse effect leading to overheating of the lower atmosphere. In the opinion of many scientists, there is a threshold value of the average atmospheric temperature exceeding, which entails the potential for the development of irreversible processes threatening the existence of humankind. To avoid this danger, governments in nearly 200 countries have chosen voluntarily to reach net-zero CO2 emissions by 2050. Renewable Energy Sources (RES), including wind and solar power plants, have been selected as substitutes for ThPPs. However, energy systems based on RES only need to be multiply redundant in terms of installed capacity due to their efficiency being heavily dependent on daily, seasonal and weather factors, leave alone the scale of the required material resources (metals, polymers, concrete, glass, etc.). The major drawback of such energy systems is, however, the RES common-cause failure, e.g., in the event of a global volcanic eruption, when no energy-security requirement can be met to provide energy for satisfying the most vital needs. A need is fully evident for furnishing such energy system with another power source to be not dependent on the event that has caused the mass RES failure. With the net-zero-carbon requirement taken into account, Nuclear Power (NP) appears to be the best option in this respect. Modern NP does not however fully suits this role due to its inherent drawbacks (limited fuel resources, pending Spent Nuclear Fuel (SNF) and RadioActive Wastes (RAW) handling and nuclear-material nonproliferation issues). A potential solution to these drawbacks is a two-component NP technology in a closed nuclear-fuel cycle currently in the process of development. In Russia, where the greatest progress has been achieved in this field of development, under construction is a pilot and demonstration energy complex with the BREST-OD-300 nuclear unit expected to be started up in 2026–2027. Another promising designs to be developed are Small Modular Reactors (SMRs) / Small Nuclear Power Plants (SNPPs).","PeriodicalId":100969,"journal":{"name":"Nuclear Energy and Technology","volume":"11 2","pages":""},"PeriodicalIF":0.0,"publicationDate":"2024-06-04","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"141266972","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}
Rafael Sh. Isayev, Pavel Dzhumaev, Irina A. Naumenko, Maria V. Leontieva-Smirnova
The processes of corrosion damage of the inner surface of the cladding are determined by corrosive reagents aggressive with respect to the cladding and the type of fuel used. Reactor irradiation of cladding made of EP823-Sh steel with mixed nitride fuel planned for use in the BREST-OD-300 reactor revealed non-uniform corrosion of the inner surface of the cladding. In this paper, the use of the chromium coating is proposed to prevent the corrosion of the inner surface of the steel fuel cladding. The results of corrosion tests of chromium coating applied to the inner surface of cladding made of EP823-Sh steel by electrolytic deposition are presented. Electron-microscopic studies of the chromium coating on EP823-Sh steel showed no significant signs of corrosion damage when tested in the environment of simulant fission products (CsI+Te) and in liquid lead at 650 °C.
{"title":"Corrosion resistance of chromium coating on the inner surface of EP823-Sh steel cladding","authors":"Rafael Sh. Isayev, Pavel Dzhumaev, Irina A. Naumenko, Maria V. Leontieva-Smirnova","doi":"10.3897/nucet.10.119642","DOIUrl":"https://doi.org/10.3897/nucet.10.119642","url":null,"abstract":"The processes of corrosion damage of the inner surface of the cladding are determined by corrosive reagents aggressive with respect to the cladding and the type of fuel used. Reactor irradiation of cladding made of EP823-Sh steel with mixed nitride fuel planned for use in the BREST-OD-300 reactor revealed non-uniform corrosion of the inner surface of the cladding.\u0000 In this paper, the use of the chromium coating is proposed to prevent the corrosion of the inner surface of the steel fuel cladding. The results of corrosion tests of chromium coating applied to the inner surface of cladding made of EP823-Sh steel by electrolytic deposition are presented. Electron-microscopic studies of the chromium coating on EP823-Sh steel showed no significant signs of corrosion damage when tested in the environment of simulant fission products (CsI+Te) and in liquid lead at 650 °C.","PeriodicalId":100969,"journal":{"name":"Nuclear Energy and Technology","volume":"143 2","pages":""},"PeriodicalIF":0.0,"publicationDate":"2024-05-07","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"141002118","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 calculation and visualization of the dose field in the decommissioning of nuclear facilities is one of the important functions of the decommissioning virtual simulation system. The dose field simulation tools can provide radiation field distribution and play an important role in determining the decommissioning plan and protecting personnel during the engineering implementation process. This article investigates the development of dose field calculation and visualization in the reactor decommissioning virtual simulation systems. A preliminary technology plan suitable for the development of the decommissioning dose field calculation and visualization display programs of the first Heavy Water Research Reactor (HWRR) in China is proposed. The applicability of the selected scheme is analyzed. The functional requirement and development direction of the HWRR reactor decommissioning dose field tool are preliminarily determined. Furthermore, the reactor vessel of HWRR reactor is modeled, the dose field distribution is calculated and visualized based on the preliminary decommissioning code. This research can provide technical support for the development of the decommissioning simulation system for the first HWRR reactor in China.
{"title":"A brief investigation of the dose field virtual simulation tools for reactor decommissioning and preliminary design for the HWRR reactor in China","authors":"Yaping Guo, Peng Nie, Ruizhi Li, Lijun Zhang, Xingwang Zhang, Ren Ren, Zelong Zhao","doi":"10.3897/nucet.10.114088","DOIUrl":"https://doi.org/10.3897/nucet.10.114088","url":null,"abstract":"The calculation and visualization of the dose field in the decommissioning of nuclear facilities is one of the important functions of the decommissioning virtual simulation system. The dose field simulation tools can provide radiation field distribution and play an important role in determining the decommissioning plan and protecting personnel during the engineering implementation process. This article investigates the development of dose field calculation and visualization in the reactor decommissioning virtual simulation systems. A preliminary technology plan suitable for the development of the decommissioning dose field calculation and visualization display programs of the first Heavy Water Research Reactor (HWRR) in China is proposed. The applicability of the selected scheme is analyzed. The functional requirement and development direction of the HWRR reactor decommissioning dose field tool are preliminarily determined. Furthermore, the reactor vessel of HWRR reactor is modeled, the dose field distribution is calculated and visualized based on the preliminary decommissioning code. This research can provide technical support for the development of the decommissioning simulation system for the first HWRR reactor in China.","PeriodicalId":100969,"journal":{"name":"Nuclear Energy and Technology","volume":"101 5","pages":""},"PeriodicalIF":0.0,"publicationDate":"2024-01-22","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"139605923","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. Shmelev, N. Geraskin, V. Apse, G. G. Kulikov, E. G. Kulikov, Vasily B. Glebov
The paper presents the estimates for the possibility for large-scale production of 238Pu in the core of a VVER-1000 power reactor. The Np-fraction of minor actinides extracted from transuranic radioactive waste is proposed to be used as the starting material. The irradiation device with NpO2 fuel elements is installed at the reactor core center. The NpO2 fuel lattice pitch is varied and the irradiation device is surrounded by a heavy moderator layer to create the best possible spectral conditions for large-scale production (~ 3 kg/year) of conditioned plutonium with the required isotopic composition (not less than 85% of 238Pu and not more than 2 ppm of 236Pu). Plutonium with such isotopic composition can be used as the thermal source in thermoelectric radioisotope generators and in cardiac pacemakers. It has been demonstrated that the estimated scale of the 238Pu production in a VVER-type power reactor exceeds considerably the existing scale of its production in research reactors.
{"title":"Assessment of the possibility for large-scale 238Pu production in a VVER-1000 power reactor","authors":"A. Shmelev, N. Geraskin, V. Apse, G. G. Kulikov, E. G. Kulikov, Vasily B. Glebov","doi":"10.3897/nucet.9.117199","DOIUrl":"https://doi.org/10.3897/nucet.9.117199","url":null,"abstract":"The paper presents the estimates for the possibility for large-scale production of 238Pu in the core of a VVER-1000 power reactor. The Np-fraction of minor actinides extracted from transuranic radioactive waste is proposed to be used as the starting material. The irradiation device with NpO2 fuel elements is installed at the reactor core center. The NpO2 fuel lattice pitch is varied and the irradiation device is surrounded by a heavy moderator layer to create the best possible spectral conditions for large-scale production (~ 3 kg/year) of conditioned plutonium with the required isotopic composition (not less than 85% of 238Pu and not more than 2 ppm of 236Pu). Plutonium with such isotopic composition can be used as the thermal source in thermoelectric radioisotope generators and in cardiac pacemakers. It has been demonstrated that the estimated scale of the 238Pu production in a VVER-type power reactor exceeds considerably the existing scale of its production in research reactors.","PeriodicalId":100969,"journal":{"name":"Nuclear Energy and Technology","volume":"32 12","pages":""},"PeriodicalIF":0.0,"publicationDate":"2023-12-21","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"138949641","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. Gulevich, V. Usanov, V. Dekusar, V. A. Eliseev, A. Moseev, E. S. Khnykina
The article analyzes the economic aspects of reducing the production of americium during the transition from a single-component nuclear energy system (NES) based on thermal reactors in an open fuel cycle to a two-component system with thermal and fast reactors in a closed nuclear fuel cycle. Scenarios for the development of these systems in Russia up to the end of the century are modeled. Two methods are considered for reducing the production of americium in a two-component NES with fast sodium reactors. The first method, closing the fuel cycle for plutonium in BN reactors of SFR type, is based on the use of plutonium separated from spent nuclear fuel of thermal reactors with the shortest possible (according to technical specifications) time for MOX fuel preparation and use thus preventing the main part of plutonium-241 from decay into americium. The second way is transmutation of americium. The study was carried out by using the mathematical code CYCLE designed for modeling of the NES with closed nuclear fuel cycle (NFC). The technical and economic data used in the paper was taken from published studies of Russian specialists and materials of European Union specialists presented in the IAEA/INPRO SYNERGIES project. The results of the research show that the efficiency of closing the NFC by using plutonium from thermal reactors in MOX fuel of fast sodium reactors is comparable to the efficiency of the homogeneous transmutation considered in the paper. The combination of the americium accumulation prevention method and transmutation method might significantly reduce the rate of the americium accumulation in a nuclear energy system, but the estimated costs of the considered homogeneous transmutation can significantly worsen the economic performance of sodium fast reactors.
{"title":"Some economic aspects of reducing americium production in a two-component system of thermal and fast reactors","authors":"A. Gulevich, V. Usanov, V. Dekusar, V. A. Eliseev, A. Moseev, E. S. Khnykina","doi":"10.3897/nucet.9.116655","DOIUrl":"https://doi.org/10.3897/nucet.9.116655","url":null,"abstract":"The article analyzes the economic aspects of reducing the production of americium during the transition from a single-component nuclear energy system (NES) based on thermal reactors in an open fuel cycle to a two-component system with thermal and fast reactors in a closed nuclear fuel cycle. Scenarios for the development of these systems in Russia up to the end of the century are modeled. Two methods are considered for reducing the production of americium in a two-component NES with fast sodium reactors. The first method, closing the fuel cycle for plutonium in BN reactors of SFR type, is based on the use of plutonium separated from spent nuclear fuel of thermal reactors with the shortest possible (according to technical specifications) time for MOX fuel preparation and use thus preventing the main part of plutonium-241 from decay into americium. The second way is transmutation of americium. The study was carried out by using the mathematical code CYCLE designed for modeling of the NES with closed nuclear fuel cycle (NFC). The technical and economic data used in the paper was taken from published studies of Russian specialists and materials of European Union specialists presented in the IAEA/INPRO SYNERGIES project. The results of the research show that the efficiency of closing the NFC by using plutonium from thermal reactors in MOX fuel of fast sodium reactors is comparable to the efficiency of the homogeneous transmutation considered in the paper. The combination of the americium accumulation prevention method and transmutation method might significantly reduce the rate of the americium accumulation in a nuclear energy system, but the estimated costs of the considered homogeneous transmutation can significantly worsen the economic performance of sodium fast reactors.","PeriodicalId":100969,"journal":{"name":"Nuclear Energy and Technology","volume":" 87","pages":""},"PeriodicalIF":0.0,"publicationDate":"2023-12-19","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"138960901","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}
Ruben A. Shaginyan, Valery V. Korobeinikov, Viktor Yu. Stogov
At present, the existing approaches to production of artificial isotopes are mostly based on the development experience from previous years. This work aims to develop an algorithm for selecting the most effective irradiation modes for target materials. The study is based on sequential modeling of irradiation of target isotopes by neutrons of different ‘single-group’ fluxes at the same neutron flux density within each energy group (BNAB-93). In this study, a flux density equal to 2×1015 n/(cm2×s) was used for each energy group. This approach will help ‘designing’ and selecting the actual neutron spectrum that has the highest efficiency compared to alternatives. The study modelled Co-60 and Lu-177g production for each energy group. The kinetics was analyzed in the most efficient groups in terms of specific activity. The maximum specific activity for Co-60 is reached in group 17 and is equal to 1 kCi/g. For the scheme of Lu-177g production through Lu-176 the maximum specific activity is reached in group 26 and is equal to 58.5 kCi/g. For the scheme of Lu-177g production through Yb-176, the maximum specific activity is reached in group 17 and is equal to 260 Ci/g, advantageous for production are groups 15–17 and 26.
{"title":"Study into the dependence of the Co-60 and Lu-177g production efficiency on the energy structure of neutron flux density","authors":"Ruben A. Shaginyan, Valery V. Korobeinikov, Viktor Yu. Stogov","doi":"10.3897/nucet.9.116662","DOIUrl":"https://doi.org/10.3897/nucet.9.116662","url":null,"abstract":"At present, the existing approaches to production of artificial isotopes are mostly based on the development experience from previous years. This work aims to develop an algorithm for selecting the most effective irradiation modes for target materials. The study is based on sequential modeling of irradiation of target isotopes by neutrons of different ‘single-group’ fluxes at the same neutron flux density within each energy group (BNAB-93). In this study, a flux density equal to 2×1015 n/(cm2×s) was used for each energy group. This approach will help ‘designing’ and selecting the actual neutron spectrum that has the highest efficiency compared to alternatives. The study modelled Co-60 and Lu-177g production for each energy group. The kinetics was analyzed in the most efficient groups in terms of specific activity. The maximum specific activity for Co-60 is reached in group 17 and is equal to 1 kCi/g. For the scheme of Lu-177g production through Lu-176 the maximum specific activity is reached in group 26 and is equal to 58.5 kCi/g. For the scheme of Lu-177g production through Yb-176, the maximum specific activity is reached in group 17 and is equal to 260 Ci/g, advantageous for production are groups 15–17 and 26.","PeriodicalId":100969,"journal":{"name":"Nuclear Energy and Technology","volume":" 8","pages":""},"PeriodicalIF":0.0,"publicationDate":"2023-12-19","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"138963034","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. Andrianov, O. Andrianova, Yury A. Korovin, Iliya S. Kuptsov, Anastasiya A. Spiridonova
The paper describes in brief the functional capabilities of a computer code for optimizing the neutronics model parameters (neutron data, technological parameters, and their covariance matrices) based on results of reactor physics experiments using conditional nonlinear multi-parameter optimization algorithms. The code’s application scope includes adjustment of neutron constants, technological parameters and their covariance matrices based on integral measurement results, formulation of requiremen117198ts with respect to the neutron data uncertainties for achieving the target accuracies in calculation of the reactor functionals, and estimation of the reactor performance prediction accuracy, as well as the informativity and similarity metrics of reactor physics experiments with respect to each other and in relation to the target reactor system. The paper also considers some examples of using the code to refine the neutronics models of nuclear reactor and fuel cycle systems based on results of reactor physics experiments.
{"title":"A computer code for optimizing the neutronics model parameters based on results of reactor physics experiments","authors":"A. Andrianov, O. Andrianova, Yury A. Korovin, Iliya S. Kuptsov, Anastasiya A. Spiridonova","doi":"10.3897/nucet.9.117198","DOIUrl":"https://doi.org/10.3897/nucet.9.117198","url":null,"abstract":"The paper describes in brief the functional capabilities of a computer code for optimizing the neutronics model parameters (neutron data, technological parameters, and their covariance matrices) based on results of reactor physics experiments using conditional nonlinear multi-parameter optimization algorithms. The code’s application scope includes adjustment of neutron constants, technological parameters and their covariance matrices based on integral measurement results, formulation of requiremen117198ts with respect to the neutron data uncertainties for achieving the target accuracies in calculation of the reactor functionals, and estimation of the reactor performance prediction accuracy, as well as the informativity and similarity metrics of reactor physics experiments with respect to each other and in relation to the target reactor system. The paper also considers some examples of using the code to refine the neutronics models of nuclear reactor and fuel cycle systems based on results of reactor physics experiments.","PeriodicalId":100969,"journal":{"name":"Nuclear Energy and Technology","volume":" 7","pages":""},"PeriodicalIF":0.0,"publicationDate":"2023-12-19","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"138959846","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}
Vyacheslav V. Borisov, Aleksey A. Kamaev, Olga I. Myazdrikova, Stepan A. Mikhin, Ilia А. Pakhomov, Sergey V. Perevoznikov
On the example of domestic fast reactors BN-350, BN-600, BN-800, the evolution of technical solutions for the automatic protection system of sodium-water steam generators (SG APS) is analyzed. The structural diagrams and main characteristics of the SG APS equipment of listed reactors are presented. The effectiveness of the SG APS in conditions of real leaks in steam generators of BN reactors is analyzed. The issues of development and creation of the SG APS of designed BN-1200 and INU MBIR are considered.
{"title":"Principles of construction and development of an automatic protection system for steam generators of fast reactors","authors":"Vyacheslav V. Borisov, Aleksey A. Kamaev, Olga I. Myazdrikova, Stepan A. Mikhin, Ilia А. Pakhomov, Sergey V. Perevoznikov","doi":"10.3897/nucet.9.116657","DOIUrl":"https://doi.org/10.3897/nucet.9.116657","url":null,"abstract":"On the example of domestic fast reactors BN-350, BN-600, BN-800, the evolution of technical solutions for the automatic protection system of sodium-water steam generators (SG APS) is analyzed. The structural diagrams and main characteristics of the SG APS equipment of listed reactors are presented. The effectiveness of the SG APS in conditions of real leaks in steam generators of BN reactors is analyzed. The issues of development and creation of the SG APS of designed BN-1200 and INU MBIR are considered.","PeriodicalId":100969,"journal":{"name":"Nuclear Energy and Technology","volume":" 36","pages":""},"PeriodicalIF":0.0,"publicationDate":"2023-12-19","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"138961379","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 is devoted to the definition of an analytical expression for estimating the burnup depth of nuclear fuel depending on its enrichment level, the periodicity of refueling, thermal stressthermal stress and the duration of the time period between refueling (reactor campaign) in a wide range of changes in key parameters for different types of thermal neutron reactors. The analytical expressions obtained in the work for the burnup depth are compared with numerous neutron physics calculations and experimental data from different authors for uranium fuel enrichment up to 9%. Calculations of the fuel share of the cost of electricity of nuclear power plants with PWR type reactors were performed and its sensitivity to changes in burnup depth and enrichment of fuel, the refueling periodicity, as well as to market prices for natural uranium, conversion, enrichment, fabrication of fuel assemblies and SNF handling were determined.
{"title":"Analytical dependence of burnup on enrichment of prospective fuel and parameters of reactors fuel campaign","authors":"Evgeny V. Semenov, V. V. Kharitonov","doi":"10.3897/nucet.9.116653","DOIUrl":"https://doi.org/10.3897/nucet.9.116653","url":null,"abstract":"The paper is devoted to the definition of an analytical expression for estimating the burnup depth of nuclear fuel depending on its enrichment level, the periodicity of refueling, thermal stressthermal stress and the duration of the time period between refueling (reactor campaign) in a wide range of changes in key parameters for different types of thermal neutron reactors. The analytical expressions obtained in the work for the burnup depth are compared with numerous neutron physics calculations and experimental data from different authors for uranium fuel enrichment up to 9%. Calculations of the fuel share of the cost of electricity of nuclear power plants with PWR type reactors were performed and its sensitivity to changes in burnup depth and enrichment of fuel, the refueling periodicity, as well as to market prices for natural uranium, conversion, enrichment, fabrication of fuel assemblies and SNF handling were determined.","PeriodicalId":100969,"journal":{"name":"Nuclear Energy and Technology","volume":"51 3","pages":""},"PeriodicalIF":0.0,"publicationDate":"2023-12-12","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"139007105","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}
Oleg K. Vlasov, I. Zvonova, N. V. Shchukina, S. Chekin
A radioecological model, which is a system of linear differential equations describing the dynamics of the transport of 137Cs and 131I radionuclides along the food chain after their release into the atmosphere after the Chernobyl accident, was used to reconstruct “instrumental” data of the 131I activities in the grass pastures in the central part of Mazovia. Four atmospheric models were used for the reconstruction: direct calculation, homogeneous cloud – inhomogeneous rainfall, inhomogeneous cloud – homogeneous rainfall, and a model with recalculation of the 137Cs and 131I activities in the atmosphere. The “instrumental” data were reconstructed based on data from direct measurements of the 131I activity in lawn grass. It has been shown that the direct calculation and homogeneous cloud models lead to a better agreement of the calculated and reconstructed “instrumental” data than the inhomogeneous cloud model. The arithmetic mean ratio of the calculated and reconstructed “instrumental” data lie in a range of 0.84 to 0.95 for the direct calculation and homogeneous cloud models, and in a range of 1.7 to 3.0 for the inhomogeneous cloud model. The mean geometric deviation for all models is constant and equal to 1.7. Instrumental and reconstructed “instrumental” data show a significant decrease in the specific activity of 131I in grass due to its wash-off by continuous rainfall, both during rainfall and after most of the deposition takes place. Due to this effect, the coefficient of the 131I retention on grass in the form of the maximum activity ratio to the 137Cs deposition density decreases from 34 to 1.4 m2/kg while it increases from 1 to 29 kBq/m2 as the result of the rainfall growth from 0 to 40 mm.
{"title":"Radioecological modeling of the 131I activity dynamics in the pasture vegetation of Mazovia in the year of the Chernobyl accident: Reconstruction, verification, reliability assessments","authors":"Oleg K. Vlasov, I. Zvonova, N. V. Shchukina, S. Chekin","doi":"10.3897/nucet.9.116654","DOIUrl":"https://doi.org/10.3897/nucet.9.116654","url":null,"abstract":"A radioecological model, which is a system of linear differential equations describing the dynamics of the transport of 137Cs and 131I radionuclides along the food chain after their release into the atmosphere after the Chernobyl accident, was used to reconstruct “instrumental” data of the 131I activities in the grass pastures in the central part of Mazovia. Four atmospheric models were used for the reconstruction: direct calculation, homogeneous cloud – inhomogeneous rainfall, inhomogeneous cloud – homogeneous rainfall, and a model with recalculation of the 137Cs and 131I activities in the atmosphere. The “instrumental” data were reconstructed based on data from direct measurements of the 131I activity in lawn grass. It has been shown that the direct calculation and homogeneous cloud models lead to a better agreement of the calculated and reconstructed “instrumental” data than the inhomogeneous cloud model. The arithmetic mean ratio of the calculated and reconstructed “instrumental” data lie in a range of 0.84 to 0.95 for the direct calculation and homogeneous cloud models, and in a range of 1.7 to 3.0 for the inhomogeneous cloud model. The mean geometric deviation for all models is constant and equal to 1.7. Instrumental and reconstructed “instrumental” data show a significant decrease in the specific activity of 131I in grass due to its wash-off by continuous rainfall, both during rainfall and after most of the deposition takes place. Due to this effect, the coefficient of the 131I retention on grass in the form of the maximum activity ratio to the 137Cs deposition density decreases from 34 to 1.4 m2/kg while it increases from 1 to 29 kBq/m2 as the result of the rainfall growth from 0 to 40 mm.","PeriodicalId":100969,"journal":{"name":"Nuclear Energy and Technology","volume":"13 5‐6","pages":""},"PeriodicalIF":0.0,"publicationDate":"2023-12-12","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"139009248","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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