Pub Date : 2020-09-26DOI: 10.55176/2414-1038-2020-3-63-67
Yu. I. Alimov, N. Galeyeva, V. Davydov, A. Zhirnov, P. Kuznetsov, I. Rozhdestvenskiy
The structure of the Russian nuclear power industry includes reactors with different designed fuel enrichment. It is possible to mix, in certain proportions, nuclear fuel (NF) from various reactors, thus closing the nuclear fuel cycle. Reprocessed uranium is a product of radiochemical reprocessing of spent nuclear fuel (SNF) from NF with a high initial enrichment. Use of uranium-erbium fuel based on reprocessed uranium is planned for the RBMK-1000 reactor. Along with 235U and 238U, SNF contains non-fissionable ballast isotopes of uranium (232, 234, 236 U). The 232,234U isotopes have a relatively high radioactivity and the presence of these leads to an increased dose rate of ionizing radiation but, due to their small content in fuel, does not affect the neutron balance, the neutron multiplication factor, and the reactivity margin. A large presence of 236U requires additional enrichment with 235U due a greater probability of inefficient neutron absorption by the 236U nuclei. This absorption with no fission leads to a reduced neutron multiplication factor, a reduced reactivity margin in fresh fuel, and a smaller burn-up of unloaded fuel. Analyzing the effects the increased content of even uranium isotopes (IEI) has on the reactor’s neutronic performance and fuel burn-up makes it possible to determine the amount of additional 235U fuel enrichment to make up for the negative effects of 236U on the RBMK-1000 neutronic performance.
{"title":"CONVERSION OF RBMK-1000 REACTORS TO REPROCESSED FUEL WITH AN INCREASED CONTENT OF EVEN URANIUM ISOTOPES","authors":"Yu. I. Alimov, N. Galeyeva, V. Davydov, A. Zhirnov, P. Kuznetsov, I. Rozhdestvenskiy","doi":"10.55176/2414-1038-2020-3-63-67","DOIUrl":"https://doi.org/10.55176/2414-1038-2020-3-63-67","url":null,"abstract":"The structure of the Russian nuclear power industry includes reactors with different designed fuel enrichment. It is possible to mix, in certain proportions, nuclear fuel (NF) from various reactors, thus closing the nuclear fuel cycle. Reprocessed uranium is a product of radiochemical reprocessing of spent nuclear fuel (SNF) from NF with a high initial enrichment. Use of uranium-erbium fuel based on reprocessed uranium is planned for the RBMK-1000 reactor. Along with 235U and 238U, SNF contains non-fissionable ballast isotopes of uranium (232, 234, 236 U). The 232,234U isotopes have a relatively high radioactivity and the presence of these leads to an increased dose rate of ionizing radiation but, due to their small content in fuel, does not affect the neutron balance, the neutron multiplication factor, and the reactivity margin. A large presence of 236U requires additional enrichment with 235U due a greater probability of inefficient neutron absorption by the 236U nuclei. This absorption with no fission leads to a reduced neutron multiplication factor, a reduced reactivity margin in fresh fuel, and a smaller burn-up of unloaded fuel. Analyzing the effects the increased content of even uranium isotopes (IEI) has on the reactor’s neutronic performance and fuel burn-up makes it possible to determine the amount of additional 235U fuel enrichment to make up for the negative effects of 236U on the RBMK-1000 neutronic performance.","PeriodicalId":20426,"journal":{"name":"PROBLEMS OF ATOMIC SCIENCE AND TECHNOLOGY. SERIES: NUCLEAR AND REACTOR CONSTANTS","volume":null,"pages":null},"PeriodicalIF":0.0,"publicationDate":"2020-09-26","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"77491582","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}
Pub Date : 2020-09-26DOI: 10.55176/2414-1038-2020-3-117-126
K. Ivanov, R. Askhadullin, A. Osipov, Sh. A. Niyazov
The problem of taking into account the release of metallic impurities from steels into heavy heat carriers is important from two points of view. First, the intensity of these impurities entering the coolant directly affects its quality, the formation of deposits based on oxides of metal components of structural steels in the primary circuit, as well as contamination of the gas system and the radiation environment. In addition, the process of the entry of metallic impurities into the coolant during the development of the oxidative nature of its interaction with steels determines the kinetics of this interaction and should be taken into account in oxidation models, especially with an increase in the duration of contact between steel and the coolant. At present, in world practice, despite the understanding of the importance of taking into account the losses of metal components of steels into the coolant, there is no adequate physical model for accounting for these losses. Basically, one or another empirical or semi-empirical approach is proposed. A new experimental data processing technique to verify the model of the release of iron into the lead coolant is presented in the article. It’s based on the analysis of the deoxidizing stage of the process of regulating oxygen TDA in the volume of the coolant. Series of experiments was carried out with varying temperature conditions in the range from 500 to 635 °C and the oxygen mode of CO = (1 ÷ 4) 10-6 wt%. It is shown that the diffusion models of iron yield and oxygen consumption describe well the experimental results and can be used in calculation codes for mass transfer in circuits with HLMC. The numerical values of the parameters characterizing the yield of iron in HLMC depending on the TDA of oxygen and the temperature of the liquid metal under conditions of natural convection are obtained.
{"title":"STUDY OF DIFFUSION OUTPUT OF IRON FROM EP823 STEEL INTO LEAD MELT","authors":"K. Ivanov, R. Askhadullin, A. Osipov, Sh. A. Niyazov","doi":"10.55176/2414-1038-2020-3-117-126","DOIUrl":"https://doi.org/10.55176/2414-1038-2020-3-117-126","url":null,"abstract":"The problem of taking into account the release of metallic impurities from steels into heavy heat carriers is important from two points of view. First, the intensity of these impurities entering the coolant directly affects its quality, the formation of deposits based on oxides of metal components of structural steels in the primary circuit, as well as contamination of the gas system and the radiation environment. In addition, the process of the entry of metallic impurities into the coolant during the development of the oxidative nature of its interaction with steels determines the kinetics of this interaction and should be taken into account in oxidation models, especially with an increase in the duration of contact between steel and the coolant. At present, in world practice, despite the understanding of the importance of taking into account the losses of metal components of steels into the coolant, there is no adequate physical model for accounting for these losses. Basically, one or another empirical or semi-empirical approach is proposed. A new experimental data processing technique to verify the model of the release of iron into the lead coolant is presented in the article. It’s based on the analysis of the deoxidizing stage of the process of regulating oxygen TDA in the volume of the coolant. Series of experiments was carried out with varying temperature conditions in the range from 500 to 635 °C and the oxygen mode of CO = (1 ÷ 4) 10-6 wt%. It is shown that the diffusion models of iron yield and oxygen consumption describe well the experimental results and can be used in calculation codes for mass transfer in circuits with HLMC. The numerical values of the parameters characterizing the yield of iron in HLMC depending on the TDA of oxygen and the temperature of the liquid metal under conditions of natural convection are obtained.","PeriodicalId":20426,"journal":{"name":"PROBLEMS OF ATOMIC SCIENCE AND TECHNOLOGY. SERIES: NUCLEAR AND REACTOR CONSTANTS","volume":null,"pages":null},"PeriodicalIF":0.0,"publicationDate":"2020-09-26","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"87046672","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}
Pub Date : 2020-09-26DOI: 10.55176/2414-1038-2020-3-30-38
A. Balovnev, V. Davidov, A. Zhirnov, A. Ivanyuta, A. Moiseev, E. Soldatov, V. Yufereva
One of the actual task at present is the substantiation of the project of the pilot demonstration reactor with a lead coolant BREST-OD-300. For the implementation of large-scale development of nuclear power, which meets modern requirements for new generation reactors, a competitive commercial power unit BR-1200 with an electric capacity of 1200 MW is being designed. To solve complex problems in the study for the optimal configurations of the core, it is required to develop a system of design codes that allows us to perform works on physical design and safety justification. System of codes includes the diffusion software package FACT-BR, the precision software package MCU-BR and the thermos-physical module IVIS-BR. The system is designed to conduct accurate studies of the physical characteristics during the design and computational support of a fast reactor with a lead coolant operating in a closed nuclear fuel cycle. The article provides a brief description of the system of codes. To demonstrate the capabilities of the system of codes, test calculation of the BR-1200 core was carried out.
{"title":"SYSTEM OF CODES FOR PHYSICAL DESIGN OF THE LEAD-COOLED FAST REACTOR","authors":"A. Balovnev, V. Davidov, A. Zhirnov, A. Ivanyuta, A. Moiseev, E. Soldatov, V. Yufereva","doi":"10.55176/2414-1038-2020-3-30-38","DOIUrl":"https://doi.org/10.55176/2414-1038-2020-3-30-38","url":null,"abstract":"One of the actual task at present is the substantiation of the project of the pilot demonstration reactor with a lead coolant BREST-OD-300. For the implementation of large-scale development of nuclear power, which meets modern requirements for new generation reactors, a competitive commercial power unit BR-1200 with an electric capacity of 1200 MW is being designed. To solve complex problems in the study for the optimal configurations of the core, it is required to develop a system of design codes that allows us to perform works on physical design and safety justification. System of codes includes the diffusion software package FACT-BR, the precision software package MCU-BR and the thermos-physical module IVIS-BR. The system is designed to conduct accurate studies of the physical characteristics during the design and computational support of a fast reactor with a lead coolant operating in a closed nuclear fuel cycle. The article provides a brief description of the system of codes. To demonstrate the capabilities of the system of codes, test calculation of the BR-1200 core was carried out.","PeriodicalId":20426,"journal":{"name":"PROBLEMS OF ATOMIC SCIENCE AND TECHNOLOGY. SERIES: NUCLEAR AND REACTOR CONSTANTS","volume":null,"pages":null},"PeriodicalIF":0.0,"publicationDate":"2020-09-26","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"76693224","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}
Pub Date : 2020-09-26DOI: 10.55176/2414-1038-2020-3-51-62
A. Lapin, A. Bobryashov, V. Blandinsky, E. Bobrov
Nowadays nuclear energy operates in an open fuel cycle. One of the most important directions in the development of nuclear energy is the closure of the nuclear fuel cycle. The solution to this problem is possible with the use of fast neutron reactors. To achieve this goal, the possibility of using a reactor with a fast-resonance neutron spectrum cooled by supercritical water (SCWR) was considered. The SCWR reactor can be effectively used in a closed nuclear fuel cycle, since it makes it possible to use spent fuel and dump uranium with a small amount of plutonium added. The layout options of the core with a change in the size of the core and reproduction zones are considered. The possibility of placing reproduction zones from various materials inside the active zone was evaluated. Based on the studies, an acceptable version of the core is selected in terms of system characteristics. For the considered arrangement of the reactor core, the possibility of shorting the uranium-plutonium and uranium-thorium fuel cycles has been investigated. The system characteristics of the reactor installation were studied for the following fuel load options: 1. Loading MOX fuel into the core, depleted uranium in the lateral zone of reproduction. 2. Loading of uranium-thorium fuel into the core and side screens. The results of the assessments of the system characteristics of the reactor are considered in the article.
{"title":"ASSESSMENT OF THE SYSTEM CHARACTERISTICS OF A REACTOR WITH SUPERCRITICAL COOLANT PARAMETERS FOR VARIOUS FUEL CYCLES","authors":"A. Lapin, A. Bobryashov, V. Blandinsky, E. Bobrov","doi":"10.55176/2414-1038-2020-3-51-62","DOIUrl":"https://doi.org/10.55176/2414-1038-2020-3-51-62","url":null,"abstract":"Nowadays nuclear energy operates in an open fuel cycle. One of the most important directions in the development of nuclear energy is the closure of the nuclear fuel cycle. The solution to this problem is possible with the use of fast neutron reactors. To achieve this goal, the possibility of using a reactor with a fast-resonance neutron spectrum cooled by supercritical water (SCWR) was considered. The SCWR reactor can be effectively used in a closed nuclear fuel cycle, since it makes it possible to use spent fuel and dump uranium with a small amount of plutonium added. The layout options of the core with a change in the size of the core and reproduction zones are considered. The possibility of placing reproduction zones from various materials inside the active zone was evaluated. Based on the studies, an acceptable version of the core is selected in terms of system characteristics. For the considered arrangement of the reactor core, the possibility of shorting the uranium-plutonium and uranium-thorium fuel cycles has been investigated. The system characteristics of the reactor installation were studied for the following fuel load options: 1. Loading MOX fuel into the core, depleted uranium in the lateral zone of reproduction. 2. Loading of uranium-thorium fuel into the core and side screens. The results of the assessments of the system characteristics of the reactor are considered in the article.","PeriodicalId":20426,"journal":{"name":"PROBLEMS OF ATOMIC SCIENCE AND TECHNOLOGY. SERIES: NUCLEAR AND REACTOR CONSTANTS","volume":null,"pages":null},"PeriodicalIF":0.0,"publicationDate":"2020-09-26","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"75919983","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}
Pub Date : 2020-09-26DOI: 10.55176/2414-1038-2020-3-97-109
S. Shcherbakov
The conditions and mechanisms of events in a moving fluid are analyzed, leading to the apparent disorder of unsteady flow, known as turbulence. The method of analysis is the use of different forms of equations of motion and transfer of characteristics, the selection of stable formations in the flow structure and a description of the interaction between them. The non-trivial results of previous works are used. The transfer and transformation of disturbances of a vortex distributed in the flow is analyzed, the conditions under which insulated tubes with a helical flow appear inside the shear flow. An important condition is the short duration of vortex disturbances. Equations are obtained that describe the interaction of the main shear flow and the vortex tube, the features of which lead to flow instability. The existence of two mechanisms for the development of turbulence is shown - the autogeneration of local decelerations and the instability of stretching of vortex tubes. The self-generation mechanism is the transfer of kinetic energy from the main flow to an annular vortex with the generation of a new annular vortex. This is the main mechanism that ensures the propagation of instability downstream, arises first when the Re number increases. The tensile instability leads to the splitting of the vortex tube into independent sections, the generation of many annular vortices that fill the space and drift in it. The vortex multiplication factor in each generation increases with the Re number and can reach many thousands. The role of ordered unsteady flows in the initiation of turbulence is shown.
{"title":"A COMPLETE DESCRIPTION OF THE MECHANICS OF TURBULENCE IN A MOVING FLUID","authors":"S. Shcherbakov","doi":"10.55176/2414-1038-2020-3-97-109","DOIUrl":"https://doi.org/10.55176/2414-1038-2020-3-97-109","url":null,"abstract":"The conditions and mechanisms of events in a moving fluid are analyzed, leading to the apparent disorder of unsteady flow, known as turbulence. The method of analysis is the use of different forms of equations of motion and transfer of characteristics, the selection of stable formations in the flow structure and a description of the interaction between them. The non-trivial results of previous works are used. The transfer and transformation of disturbances of a vortex distributed in the flow is analyzed, the conditions under which insulated tubes with a helical flow appear inside the shear flow. An important condition is the short duration of vortex disturbances. Equations are obtained that describe the interaction of the main shear flow and the vortex tube, the features of which lead to flow instability. The existence of two mechanisms for the development of turbulence is shown - the autogeneration of local decelerations and the instability of stretching of vortex tubes. The self-generation mechanism is the transfer of kinetic energy from the main flow to an annular vortex with the generation of a new annular vortex. This is the main mechanism that ensures the propagation of instability downstream, arises first when the Re number increases. The tensile instability leads to the splitting of the vortex tube into independent sections, the generation of many annular vortices that fill the space and drift in it. The vortex multiplication factor in each generation increases with the Re number and can reach many thousands. The role of ordered unsteady flows in the initiation of turbulence is shown.","PeriodicalId":20426,"journal":{"name":"PROBLEMS OF ATOMIC SCIENCE AND TECHNOLOGY. SERIES: NUCLEAR AND REACTOR CONSTANTS","volume":null,"pages":null},"PeriodicalIF":0.0,"publicationDate":"2020-09-26","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"84721178","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}
Pub Date : 2020-09-26DOI: 10.55176/2414-1038-2020-3-80-87
V. Kulikov, N. Zhylmaganbetov, A. Popykin, A. Smirnova, L. Kryakvin, V. Pitilimov, O. Sedov, V. Tereshonok, R. Sizov
The article represents the results of measurement and numerical simulation of measurement of scram-system efficiency of reactor plants performed during the reactor physical startup tests of Novovoronezh-2 units 1, 2 and Leningrad-2 unit 1. These power units are equipped with a reactor VVER-1200, therefore, one numerical simulation of measurement of scram-system efficiency on the basis of the reactor project was performed. The article represents comparisons of the calculated and the measured currents of the ionization chambers and the readings of reactimeter calculate for them during measurement of scram-system efficiency.
{"title":"COMPARISON OF THE RESULTS OF NUMERICAL SIMULATION OF MEASUREMENT OF SCRAM-SYSTEM EFFICIENCY AND MEASUREMENT PERFORMED DURING THE REACTOR PHYSICAL STARTUP TESTS OF UNITS 1, 2 NOVOVORONEZH-2 NPP AND UNIT 1 LENINGRAD-2 NPP","authors":"V. Kulikov, N. Zhylmaganbetov, A. Popykin, A. Smirnova, L. Kryakvin, V. Pitilimov, O. Sedov, V. Tereshonok, R. Sizov","doi":"10.55176/2414-1038-2020-3-80-87","DOIUrl":"https://doi.org/10.55176/2414-1038-2020-3-80-87","url":null,"abstract":"The article represents the results of measurement and numerical simulation of measurement of scram-system efficiency of reactor plants performed during the reactor physical startup tests of Novovoronezh-2 units 1, 2 and Leningrad-2 unit 1. These power units are equipped with a reactor VVER-1200, therefore, one numerical simulation of measurement of scram-system efficiency on the basis of the reactor project was performed. The article represents comparisons of the calculated and the measured currents of the ionization chambers and the readings of reactimeter calculate for them during measurement of scram-system efficiency.","PeriodicalId":20426,"journal":{"name":"PROBLEMS OF ATOMIC SCIENCE AND TECHNOLOGY. SERIES: NUCLEAR AND REACTOR CONSTANTS","volume":null,"pages":null},"PeriodicalIF":0.0,"publicationDate":"2020-09-26","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"79047093","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}
Pub Date : 2020-09-26DOI: 10.55176/2414-1038-2020-3-110-116
Yu. A. Kuzina, V. Alekseev, A. Sorokin, I. Voronin, M. Konovalov, R. Zykova
The article focuses on a comprehensive analysis of the new concept of a combined sodium oxygen purification module. The combined purification module includes a combination of two modules: chemisorption purification with a soluble getter and high-temperature getter purification with a high-temperature getter. A special zirconium alloy in a finely dispersed granular form can be used as a high-temperature getter. Other materials (e.g. titanium and various alloys) can be used and must be tested accordingly. At the first stage of the work, research was carried out on the module for chemisorption purification of sodium from oxygen at the Protva-1 stand, which showed the effectiveness of purification using the proposed getter materials, at the second, research on the module for high-temperature getter purification. For 20 hours of operation of the chemisorption module, the oxygen concentration in sodium was reduced by 12 ppm (3 g oxygen), while 4.5 g of magnesium reacted (the initial mass of magnesium was 25 g). According to the results of experiments in the sodium loop, the decrease in the concentration of dissolved oxygen in sodium during purification with a zirconium getter for 20 h was about 50 ppm. Two promising options for the development of a combined module for the purification of sodium from oxygen are proposed. Experimental justifications of selected chemisorbent, getter and filtering material are carried out. The main part of the study is analysis of chemisorption efficiency, purification process dynamics and influence on it of temperature factor for selected chemisorbent.
{"title":"INVESTIGATION OF GETTER PURIFICATION OF SODIUM FROM OXYGEN","authors":"Yu. A. Kuzina, V. Alekseev, A. Sorokin, I. Voronin, M. Konovalov, R. Zykova","doi":"10.55176/2414-1038-2020-3-110-116","DOIUrl":"https://doi.org/10.55176/2414-1038-2020-3-110-116","url":null,"abstract":"The article focuses on a comprehensive analysis of the new concept of a combined sodium oxygen purification module. The combined purification module includes a combination of two modules: chemisorption purification with a soluble getter and high-temperature getter purification with a high-temperature getter. A special zirconium alloy in a finely dispersed granular form can be used as a high-temperature getter. Other materials (e.g. titanium and various alloys) can be used and must be tested accordingly. At the first stage of the work, research was carried out on the module for chemisorption purification of sodium from oxygen at the Protva-1 stand, which showed the effectiveness of purification using the proposed getter materials, at the second, research on the module for high-temperature getter purification. For 20 hours of operation of the chemisorption module, the oxygen concentration in sodium was reduced by 12 ppm (3 g oxygen), while 4.5 g of magnesium reacted (the initial mass of magnesium was 25 g). According to the results of experiments in the sodium loop, the decrease in the concentration of dissolved oxygen in sodium during purification with a zirconium getter for 20 h was about 50 ppm. Two promising options for the development of a combined module for the purification of sodium from oxygen are proposed. Experimental justifications of selected chemisorbent, getter and filtering material are carried out. The main part of the study is analysis of chemisorption efficiency, purification process dynamics and influence on it of temperature factor for selected chemisorbent.","PeriodicalId":20426,"journal":{"name":"PROBLEMS OF ATOMIC SCIENCE AND TECHNOLOGY. SERIES: NUCLEAR AND REACTOR CONSTANTS","volume":null,"pages":null},"PeriodicalIF":0.0,"publicationDate":"2020-09-26","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"73615397","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}
Pub Date : 2020-09-26DOI: 10.55176/2414-1038-2020-3-39-50
I. Dyachkov, M. Ioannisian
Describes the software used in the calculation, according to the KIR2 program and the algorithms developed by the MAGMA group module (Group modulus taking into account anisotropic scattering). The results of the verification stage of the KIR2 program with a group physical module are presented. For verification, we used the C5G7 mathematical benchmark test, in which the energy field is divided into 7 groups. The results of comparison with the MCNP benchmark program on the effective breeding coefficient, confidence intervals and deviations from the benchmark program are presented. A comparison with other programs based on the Monte Carlo method (VIM and UNKMK and other independent calculations using the MCNP program) is also given. In addition, the use of the calculated data of the KIR2 program with high statistics as reference data for deterministic class codes (ATTILA, MCCG3D, CRX, PARTISN, UNKGRO) and programs based on the Monte Carlo method is presented. The analysis of the effectiveness of using comparison methods by deviations and confidence intervals is presented. The results of the verification phase of the KIR2 program with the MAGMA group module are summarized.
{"title":"VALIDATION OF THE GROUP PHYSICAL MODULE IN THE NEUTRON ENERGY CALCULATION PROGRAM KINETICS OF KIR2 BASED ON BENCHMARK TEST C5G7","authors":"I. Dyachkov, M. Ioannisian","doi":"10.55176/2414-1038-2020-3-39-50","DOIUrl":"https://doi.org/10.55176/2414-1038-2020-3-39-50","url":null,"abstract":"Describes the software used in the calculation, according to the KIR2 program and the algorithms developed by the MAGMA group module (Group modulus taking into account anisotropic scattering). The results of the verification stage of the KIR2 program with a group physical module are presented. For verification, we used the C5G7 mathematical benchmark test, in which the energy field is divided into 7 groups. The results of comparison with the MCNP benchmark program on the effective breeding coefficient, confidence intervals and deviations from the benchmark program are presented. A comparison with other programs based on the Monte Carlo method (VIM and UNKMK and other independent calculations using the MCNP program) is also given. In addition, the use of the calculated data of the KIR2 program with high statistics as reference data for deterministic class codes (ATTILA, MCCG3D, CRX, PARTISN, UNKGRO) and programs based on the Monte Carlo method is presented. The analysis of the effectiveness of using comparison methods by deviations and confidence intervals is presented. The results of the verification phase of the KIR2 program with the MAGMA group module are summarized.","PeriodicalId":20426,"journal":{"name":"PROBLEMS OF ATOMIC SCIENCE AND TECHNOLOGY. SERIES: NUCLEAR AND REACTOR CONSTANTS","volume":null,"pages":null},"PeriodicalIF":0.0,"publicationDate":"2020-09-26","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"73883164","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}
Pub Date : 2020-09-26DOI: 10.55176/2414-1038-2020-3-88-96
M. Ioannisian, V. Davidenko, I. Dyachkov
The work is devoted to the problem of initial neutron source distribution for modeling kinetic transient processes by the direct analogue Monte Carlo method. An algorithm is proposed for modeling such a source. The source algorithm implemented in the KIR2 program and designed to solve the problem of taking the system out of the critical state by introducing a perturbation (reactivity). Only prompt neutrons are considered. For the neutron source distribution, a phase coordinates of the neutron are used. The phase coordinates are determined in the preliminary calculation of the critical state. The results of testing the algorithm on non-stationary test problems (an infinite environment and RP1GS) are presented. In each task, the geometric area is filled with material presented in the form of single-group macroparameters. The results of calculating the integral neutron flux density were compared with the solution of the point kinetics equations. Good agreement was obtained - the deviation during all the calculated processes does not exceed 0.4 % in the integrated flux density, but it increases at the end of the processes with the introduction of negative reactivity up to 2.5 % with a small number of neutrons in the system.
{"title":"ON MODELLING OF THE INITIAL SOURCE OF PROMPT NEUTRONS FOR SOLV-ING NONSTATIONARY PROBLEMS BY THE MONTE CARLO METHOD","authors":"M. Ioannisian, V. Davidenko, I. Dyachkov","doi":"10.55176/2414-1038-2020-3-88-96","DOIUrl":"https://doi.org/10.55176/2414-1038-2020-3-88-96","url":null,"abstract":"The work is devoted to the problem of initial neutron source distribution for modeling kinetic transient processes by the direct analogue Monte Carlo method. An algorithm is proposed for modeling such a source. The source algorithm implemented in the KIR2 program and designed to solve the problem of taking the system out of the critical state by introducing a perturbation (reactivity). Only prompt neutrons are considered. For the neutron source distribution, a phase coordinates of the neutron are used. The phase coordinates are determined in the preliminary calculation of the critical state. The results of testing the algorithm on non-stationary test problems (an infinite environment and RP1GS) are presented. In each task, the geometric area is filled with material presented in the form of single-group macroparameters. The results of calculating the integral neutron flux density were compared with the solution of the point kinetics equations. Good agreement was obtained - the deviation during all the calculated processes does not exceed 0.4 % in the integrated flux density, but it increases at the end of the processes with the introduction of negative reactivity up to 2.5 % with a small number of neutrons in the system.","PeriodicalId":20426,"journal":{"name":"PROBLEMS OF ATOMIC SCIENCE AND TECHNOLOGY. SERIES: NUCLEAR AND REACTOR CONSTANTS","volume":null,"pages":null},"PeriodicalIF":0.0,"publicationDate":"2020-09-26","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"83958821","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}
Pub Date : 2020-09-26DOI: 10.55176/2414-1038-2020-3-68-71
Yu. A. Alimov, N. Galeyeva, A. Zhirnov, P. Kuznetsov, I. Rozhdestvenskiy
The RBMK-1000 reactor operates in a continuous refueling mode. The reactor core comprises FAs across the burn-up spectrum, from fresh fuel assemblies to the most burnt-up ones. Following the reactor shutdown for decommissioning, the majority of irradiated fuel assemblies (IFA) have the potential for further use (the so-called afterburning) at operating NPP units. Most of the irradiated fuel assemblies have a burn-up fraction, which is far from the specified threshold value. The neutron multiplication factor in the cell exceeds the core average value and, therefore, the IFA loading makes it possible to increase the reactivity margin to the desired value. Reuse of spent fuel assemblies in reactors at other power units offers a number of advantages. In economic terms, afterburning is able to reduce the use of fresh fuel assemblies (FFA). In terms of radioactive waste handling, afterburning is able to reduce the number of irradiated fuel assemblies, which require long-time storage in a spent nuclear fuel repository thus reducing the IFA and radioactive waste disposal loads. JSC NIKIET has proposed the use of irradiated fuel assemblies from units 1 and 2 of Leningrad NPP shut down after decommissioning in reactors at units 3 and 4 of the same NPP.
{"title":"COMPUTATIONAL SIMULATION OF POST-DECOMMISSIONING AFTERBURNING OF FAS FROM UNITS 1 AND 2 OF LENINGRAD NPP","authors":"Yu. A. Alimov, N. Galeyeva, A. Zhirnov, P. Kuznetsov, I. Rozhdestvenskiy","doi":"10.55176/2414-1038-2020-3-68-71","DOIUrl":"https://doi.org/10.55176/2414-1038-2020-3-68-71","url":null,"abstract":"The RBMK-1000 reactor operates in a continuous refueling mode. The reactor core comprises FAs across the burn-up spectrum, from fresh fuel assemblies to the most burnt-up ones. Following the reactor shutdown for decommissioning, the majority of irradiated fuel assemblies (IFA) have the potential for further use (the so-called afterburning) at operating NPP units. Most of the irradiated fuel assemblies have a burn-up fraction, which is far from the specified threshold value. The neutron multiplication factor in the cell exceeds the core average value and, therefore, the IFA loading makes it possible to increase the reactivity margin to the desired value. Reuse of spent fuel assemblies in reactors at other power units offers a number of advantages. In economic terms, afterburning is able to reduce the use of fresh fuel assemblies (FFA). In terms of radioactive waste handling, afterburning is able to reduce the number of irradiated fuel assemblies, which require long-time storage in a spent nuclear fuel repository thus reducing the IFA and radioactive waste disposal loads. JSC NIKIET has proposed the use of irradiated fuel assemblies from units 1 and 2 of Leningrad NPP shut down after decommissioning in reactors at units 3 and 4 of the same NPP.","PeriodicalId":20426,"journal":{"name":"PROBLEMS OF ATOMIC SCIENCE AND TECHNOLOGY. SERIES: NUCLEAR AND REACTOR CONSTANTS","volume":null,"pages":null},"PeriodicalIF":0.0,"publicationDate":"2020-09-26","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"87734098","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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