Valerij Pavlovich Bereznev, D. Koltashev, Roman Evgenyevich Shurygin
{"title":"Cross-Verification of the Neutron-Physical Codes CORNER and MCU-FR on Models of Advanced Fast Neutron Reactors","authors":"Valerij Pavlovich Bereznev, D. Koltashev, Roman Evgenyevich Shurygin","doi":"10.26583/npe.2023.1.11","DOIUrl":"https://doi.org/10.26583/npe.2023.1.11","url":null,"abstract":"","PeriodicalId":37826,"journal":{"name":"Izvestiya Wysshikh Uchebnykh Zawedeniy, Yadernaya Energetika","volume":null,"pages":null},"PeriodicalIF":0.0,"publicationDate":"2023-03-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"84482695","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}
D. G. Chereshkov, Ternovykh Mikhail Ternovykh, G. Tikhomirov, Aleksandr Aleksandrovich Ryzhkov
The new calculation code capabilities are applied in the current work as well as important fast reactor criticality parameters uncertainty assessment articles’ results based on different nuclear data libraries and covariance matrices. A comparative analysis of uncertainty estimations related to neutron reactions is presented for lead-cooled reactor models and sodium-cooled reactor models. For the models of advanced BN and BR fast reactors with three fuel types (UO 2 , MOX, MNUP), the multiplication factor uncertainty calculations are performed using 252-group covariance matrices based on ENDF/B-VII.1 library via the SCALE 6.2.4 code system. The main nuclear data uncertainty contributors in the multiplication factor are determined. Recommendations are formulated for improving the cross sections accuracy for several nuclides in order to provide more reliable results of fast reactor criticality calculations. Lead-cooled reactors have no operational history compared to light-water and sodium-cooled reactors. The experimental data insufficiency calls in the question about reliability of the simulation results and requires a comprehensive initial data uncertainty analysis for the neutron transport simulation. The obtained results support the idea that lead-and sodium-cooled reactors have close nuclear data sensitivity using one and the same computation tools, nuclear data libraries and fuel compositions. This makes it possible to use the accumulated data of benchmarks for sodium-cooled reactors in the safety determination of lead-cooled reactors.
{"title":"Nuclear Data Uncertainty on Generation IV Fast Reactors Criticality Calculations Analysis Comparison","authors":"D. G. Chereshkov, Ternovykh Mikhail Ternovykh, G. Tikhomirov, Aleksandr Aleksandrovich Ryzhkov","doi":"10.26583/npe.2023.1.14","DOIUrl":"https://doi.org/10.26583/npe.2023.1.14","url":null,"abstract":"The new calculation code capabilities are applied in the current work as well as important fast reactor criticality parameters uncertainty assessment articles’ results based on different nuclear data libraries and covariance matrices. A comparative analysis of uncertainty estimations related to neutron reactions is presented for lead-cooled reactor models and sodium-cooled reactor models. For the models of advanced BN and BR fast reactors with three fuel types (UO 2 , MOX, MNUP), the multiplication factor uncertainty calculations are performed using 252-group covariance matrices based on ENDF/B-VII.1 library via the SCALE 6.2.4 code system. The main nuclear data uncertainty contributors in the multiplication factor are determined. Recommendations are formulated for improving the cross sections accuracy for several nuclides in order to provide more reliable results of fast reactor criticality calculations. Lead-cooled reactors have no operational history compared to light-water and sodium-cooled reactors. The experimental data insufficiency calls in the question about reliability of the simulation results and requires a comprehensive initial data uncertainty analysis for the neutron transport simulation. The obtained results support the idea that lead-and sodium-cooled reactors have close nuclear data sensitivity using one and the same computation tools, nuclear data libraries and fuel compositions. This makes it possible to use the accumulated data of benchmarks for sodium-cooled reactors in the safety determination of lead-cooled reactors.","PeriodicalId":37826,"journal":{"name":"Izvestiya Wysshikh Uchebnykh Zawedeniy, Yadernaya Energetika","volume":null,"pages":null},"PeriodicalIF":0.0,"publicationDate":"2023-03-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"73339001","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 considers the applicability of small perturbation theory to assessing the variations of the prompt neutron lifetime caused by variations in the isotope composition of a lead-cooled fast reactor. The generalized small perturbation theory formulas have been developed to calculate derivatives of the prompt neutron lifetime regarded as a bilinear neutron flux and neutron worth ratio. A numerical algorithm has been proposed for the step-by-step application of the small perturbation theory formulas to assess the prompt neutron lifetime variations caused by a major perturbation in the reactor isotope composition, e.g. by the complete change of the material used earlier as the neutron reflector. The advantage of the proposed approach has been shown which consists in that it is basically possible to determine the role of different neutron reactions, isotopes and energy groups in and their contributions to the total prompt neutron lifetime variation caused by major changes in the reactor isotope composition.
{"title":"Application of Small Perturbation Theory for Assessing Variations of Prompt Neutron Lifetime in a Lead-Cooled Fast Reactor","authors":"V. Apse, A. Shmelev, G. Kulikov, E. Kulikov","doi":"10.26583/npe.2023.1.03","DOIUrl":"https://doi.org/10.26583/npe.2023.1.03","url":null,"abstract":"The paper considers the applicability of small perturbation theory to assessing the variations of the prompt neutron lifetime caused by variations in the isotope composition of a lead-cooled fast reactor. The generalized small perturbation theory formulas have been developed to calculate derivatives of the prompt neutron lifetime regarded as a bilinear neutron flux and neutron worth ratio. A numerical algorithm has been proposed for the step-by-step application of the small perturbation theory formulas to assess the prompt neutron lifetime variations caused by a major perturbation in the reactor isotope composition, e.g. by the complete change of the material used earlier as the neutron reflector. The advantage of the proposed approach has been shown which consists in that it is basically possible to determine the role of different neutron reactions, isotopes and energy groups in and their contributions to the total prompt neutron lifetime variation caused by major changes in the reactor isotope composition.","PeriodicalId":37826,"journal":{"name":"Izvestiya Wysshikh Uchebnykh Zawedeniy, Yadernaya Energetika","volume":null,"pages":null},"PeriodicalIF":0.0,"publicationDate":"2023-03-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"80539554","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}
Vladimir A. Gorbunov, S.S. Teplyakova, N. A. Lonshakov, Sergey G. Andrianov, P.A. Mineev, Yury Kazansky
The paper presents the results of a numerical study into the efficiency of the fuel element operation in the pressurized water reactor (VVER) core filled with uranium dioxide (UO 2 ) pellets. The investigation results were obtained from a three-dimensional simulation of the fuel element power density. The dependencies of the fuel and fuel cladding temperatures on specific power per cubic meter of fuel are compared. Uranium metal and uranium dioxide have been studied as fuel. Engineering constraints on the safe operation of fuel assemblies have been selected as the determining parameters. The paper analyzes the extent of the radiation heat transfer effects on the fuel element specific power. Equations have been obtained that reflect the dependencies of specific power per cubic meter of fuel on the size of the fuel pellet hole diameter in the maximum heat flux conditions. The COMSOL Multiphysics code, a numerical thermo - physical simulation package, was used for the study. Calculations show that an additional uranium-235 enrichment with an increase in the fuel pellet hole diameter at a fixed fuel thermal power leads to a reduced reactor axial temperature field peaking factor.
{"title":"Investigation of the Influence of the Fuel Element Design Parameter on the VVER-1000 Reactor Axial Power Peaking Factor","authors":"Vladimir A. Gorbunov, S.S. Teplyakova, N. A. Lonshakov, Sergey G. Andrianov, P.A. Mineev, Yury Kazansky","doi":"10.26583/npe.2023.2.03","DOIUrl":"https://doi.org/10.26583/npe.2023.2.03","url":null,"abstract":"The paper presents the results of a numerical study into the efficiency of the fuel element operation in the pressurized water reactor (VVER) core filled with uranium dioxide (UO 2 ) pellets. The investigation results were obtained from a three-dimensional simulation of the fuel element power density. The dependencies of the fuel and fuel cladding temperatures on specific power per cubic meter of fuel are compared. Uranium metal and uranium dioxide have been studied as fuel. Engineering constraints on the safe operation of fuel assemblies have been selected as the determining parameters. The paper analyzes the extent of the radiation heat transfer effects on the fuel element specific power. Equations have been obtained that reflect the dependencies of specific power per cubic meter of fuel on the size of the fuel pellet hole diameter in the maximum heat flux conditions. The COMSOL Multiphysics code, a numerical thermo - physical simulation package, was used for the study. Calculations show that an additional uranium-235 enrichment with an increase in the fuel pellet hole diameter at a fixed fuel thermal power leads to a reduced reactor axial temperature field peaking factor.","PeriodicalId":37826,"journal":{"name":"Izvestiya Wysshikh Uchebnykh Zawedeniy, Yadernaya Energetika","volume":null,"pages":null},"PeriodicalIF":0.0,"publicationDate":"2023-03-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"83949264","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. Kovalyov, A. M. Prokoshin, A. S. Kudinov, V. Nevinitsa
The VVER-1000 thermal neutron reactor can operate on mixed uranium-plutonium fuel with a content of reactor-grade plutonium up to 5% with a 100% loaded core. In this case, plutonium burns up to 56% of odd isotopes. The energy potential of such plutonium is very low, and its further use in thermal reactors is impractical. However, such plutonium can be used in fast neutron reactors. The paper presents the results of investigating the possibility for such isotopic plutonium composition to be used in the BN-1200 thermal neutron reactor and its value be increased for the plutonium recycle in the reactor. For this purpose, a precision model of the BN-1200 reactor has been developed using the Serpent Monte Carlo code. The model has been verified against the reference values of the nuclear fuel burnup and breeding ratios. The study has shown that such plutonium can be used in the BN-1200 reactor MOX fuel. Maintaining the operating cycle length requires the plutonium fraction in the MOX fuel to be increased up to 2%. In the BN-1200 reactor, the isotopic composition has been found to improve for the further recycle of plutonium in the thermal reactor, i.e. odd plutonium isotopes increase. The fewer odd plutonium isotopes at the beginning of the BN-1200 operating cycle, the greater their increase. It can be seen as the result of the calculation that plutonium from VVER-1000 spent mixed fuel must be loaded into the BN-1200 reactor at least twice to increase the fraction of odd isotopes to the level of reactor-grade plutonium.
{"title":"Use of REMIX Spent Mixed Fuel Plutonium in the BN-1200 Reactor","authors":"N. Kovalyov, A. M. Prokoshin, A. S. Kudinov, V. Nevinitsa","doi":"10.26583/npe.2023.1.06","DOIUrl":"https://doi.org/10.26583/npe.2023.1.06","url":null,"abstract":"The VVER-1000 thermal neutron reactor can operate on mixed uranium-plutonium fuel with a content of reactor-grade plutonium up to 5% with a 100% loaded core. In this case, plutonium burns up to 56% of odd isotopes. The energy potential of such plutonium is very low, and its further use in thermal reactors is impractical. However, such plutonium can be used in fast neutron reactors. The paper presents the results of investigating the possibility for such isotopic plutonium composition to be used in the BN-1200 thermal neutron reactor and its value be increased for the plutonium recycle in the reactor. For this purpose, a precision model of the BN-1200 reactor has been developed using the Serpent Monte Carlo code. The model has been verified against the reference values of the nuclear fuel burnup and breeding ratios. The study has shown that such plutonium can be used in the BN-1200 reactor MOX fuel. Maintaining the operating cycle length requires the plutonium fraction in the MOX fuel to be increased up to 2%. In the BN-1200 reactor, the isotopic composition has been found to improve for the further recycle of plutonium in the thermal reactor, i.e. odd plutonium isotopes increase. The fewer odd plutonium isotopes at the beginning of the BN-1200 operating cycle, the greater their increase. It can be seen as the result of the calculation that plutonium from VVER-1000 spent mixed fuel must be loaded into the BN-1200 reactor at least twice to increase the fraction of odd isotopes to the level of reactor-grade plutonium.","PeriodicalId":37826,"journal":{"name":"Izvestiya Wysshikh Uchebnykh Zawedeniy, Yadernaya Energetika","volume":null,"pages":null},"PeriodicalIF":0.0,"publicationDate":"2023-03-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"79650679","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":"A Model of the Coolant Flow in Supercritical Nuclear Reactors Based on the Highest Approximations of the Chapman-Enskog Method","authors":"I. A. Chusov","doi":"10.26583/npe.2023.2.04","DOIUrl":"https://doi.org/10.26583/npe.2023.2.04","url":null,"abstract":"","PeriodicalId":37826,"journal":{"name":"Izvestiya Wysshikh Uchebnykh Zawedeniy, Yadernaya Energetika","volume":null,"pages":null},"PeriodicalIF":0.0,"publicationDate":"2023-03-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"76423766","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}
G. Kulikov, A. Shmelev, B. G. Vasily, V. Apse, E. Kulikov
{"title":"Neutronic Foundations for Large-Scale Production of 238Pu for Independent Energy Sources","authors":"G. Kulikov, A. Shmelev, B. G. Vasily, V. Apse, E. Kulikov","doi":"10.26583/npe.2023.2.13","DOIUrl":"https://doi.org/10.26583/npe.2023.2.13","url":null,"abstract":"","PeriodicalId":37826,"journal":{"name":"Izvestiya Wysshikh Uchebnykh Zawedeniy, Yadernaya Energetika","volume":null,"pages":null},"PeriodicalIF":0.0,"publicationDate":"2023-03-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"85053896","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 publication substantiates the economic advantages of using in the starting loads of inherently safe fast reactors with a closed fuel cycle of enriched uranium instead of uranium-plutonium regenerate obtained by reprocessing of thermal reactors spent nuclear fuel (SNF). The justifications are given taking into account both the preliminary technical and economic assessments carried out by the basic enterprises of TVEL JSC and SHK JSC, and the neutron-physical and system-economic studies performed at the Private Institution of the ITCP Proryv (Breakthrough). It is shown that the starting-up of a fast reactor on enriched uranium instead of uranium-plutonium fuel, taking into account the costs of preliminary reprocessing of thermal reactors spent fuel, allows achieving a significant economic gain at the stage of construction and commissioning of nuclear power plants. It is also shown that even at moderately high values of the discount coefficient, the uranium start of a fast reactor with a closed fuel cycle is economically preferable in comparison with the option of starting on uranium-plutonium fuel from the positions of the break-even tariff.
{"title":"Economic Advantages of Starting-up of Inherently Safe Fast Reactors with a Closed Fuel Cycle on Fortificated Uranium","authors":"M. A. Orlov","doi":"10.26583/npe.2023.1.01","DOIUrl":"https://doi.org/10.26583/npe.2023.1.01","url":null,"abstract":"The publication substantiates the economic advantages of using in the starting loads of inherently safe fast reactors with a closed fuel cycle of enriched uranium instead of uranium-plutonium regenerate obtained by reprocessing of thermal reactors spent nuclear fuel (SNF). The justifications are given taking into account both the preliminary technical and economic assessments carried out by the basic enterprises of TVEL JSC and SHK JSC, and the neutron-physical and system-economic studies performed at the Private Institution of the ITCP Proryv (Breakthrough). It is shown that the starting-up of a fast reactor on enriched uranium instead of uranium-plutonium fuel, taking into account the costs of preliminary reprocessing of thermal reactors spent fuel, allows achieving a significant economic gain at the stage of construction and commissioning of nuclear power plants. It is also shown that even at moderately high values of the discount coefficient, the uranium start of a fast reactor with a closed fuel cycle is economically preferable in comparison with the option of starting on uranium-plutonium fuel from the positions of the break-even tariff.","PeriodicalId":37826,"journal":{"name":"Izvestiya Wysshikh Uchebnykh Zawedeniy, Yadernaya Energetika","volume":null,"pages":null},"PeriodicalIF":0.0,"publicationDate":"2023-03-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"82990164","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, Y. Korovin, I. Kuptsov, Anastasia Alekseevna Spiridonova
{"title":"A Computer Code for Optimizing the Neutronic Model Parameters Based on Results of Integral Experiments","authors":"A. Andrianov, O. Andrianova, Y. Korovin, I. Kuptsov, Anastasia Alekseevna Spiridonova","doi":"10.26583/npe.2023.2.12","DOIUrl":"https://doi.org/10.26583/npe.2023.2.12","url":null,"abstract":"","PeriodicalId":37826,"journal":{"name":"Izvestiya Wysshikh Uchebnykh Zawedeniy, Yadernaya Energetika","volume":null,"pages":null},"PeriodicalIF":0.0,"publicationDate":"2023-03-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"81075392","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, D. A. Maltsev, Margarita G. Isaenkova, Olga A. Krymskaya, R. A. Minushkin
with
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{"title":"Thermal Resistance of Steels with Increased Strength Properties for Pressure Vessels of Advanced VVER Reactors of Various Designs","authors":"E. Kuleshova, I. Fedotov, D. A. Maltsev, Margarita G. Isaenkova, Olga A. Krymskaya, R. A. Minushkin","doi":"10.26583/npe.2023.2.08","DOIUrl":"https://doi.org/10.26583/npe.2023.2.08","url":null,"abstract":"with","PeriodicalId":37826,"journal":{"name":"Izvestiya Wysshikh Uchebnykh Zawedeniy, Yadernaya Energetika","volume":null,"pages":null},"PeriodicalIF":0.0,"publicationDate":"2023-03-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"74070398","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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