O. Andrianova, Y. Golovko, G. Lomakov, Yevgeniya S. Teplukhina, Gennady M. Zherdev
The paper presents the results of a comparative analysis of criticality calculations using a Monte-Carlo code with the BNAB-93 and BNAB-RF neutron group constants, as well as with evaluated neutron data files from the Russian ROSFOND evaluated nuclear data library and other evaluated nuclear data libraries (ENDF, JEFF, JENDL) from different years. A set of integral experiments on BFS critical assemblies carried out in different years at the Institute of Physics and Power Engineering (60 different critical configurations) was analyzed. The considered integral experiments are included in the database of evaluated experimental neutronic data used to justify the neutronic performance of sodium and lead cooled fast reactors, to verify codes and nuclear data as well as to estimate uncertainties in neutronic parameters due to the nuclear data uncertainties. It has been shown that the ROSFOND evaluated nuclear data library is a library that minimizes the calculation and experimental discrepancies for the considered set of integral experiments. The paper also presents the results of criticality calculations for models of sodium and lead cooled fast reactors based on different evaluated neutron data libraries and provides estimates for the uncertainty in criticality associated with nuclear data.
{"title":"Calculation and experimental analysis of benchmark experiments with a fast neutron spectrum and models of sodium and lead cooled fast reactors using different evaluated nuclear data libraries","authors":"O. Andrianova, Y. Golovko, G. Lomakov, Yevgeniya S. Teplukhina, Gennady M. Zherdev","doi":"10.3897/NUCET.7.68951","DOIUrl":"https://doi.org/10.3897/NUCET.7.68951","url":null,"abstract":"The paper presents the results of a comparative analysis of criticality calculations using a Monte-Carlo code with the BNAB-93 and BNAB-RF neutron group constants, as well as with evaluated neutron data files from the Russian ROSFOND evaluated nuclear data library and other evaluated nuclear data libraries (ENDF, JEFF, JENDL) from different years. A set of integral experiments on BFS critical assemblies carried out in different years at the Institute of Physics and Power Engineering (60 different critical configurations) was analyzed. The considered integral experiments are included in the database of evaluated experimental neutronic data used to justify the neutronic performance of sodium and lead cooled fast reactors, to verify codes and nuclear data as well as to estimate uncertainties in neutronic parameters due to the nuclear data uncertainties. It has been shown that the ROSFOND evaluated nuclear data library is a library that minimizes the calculation and experimental discrepancies for the considered set of integral experiments. The paper also presents the results of criticality calculations for models of sodium and lead cooled fast reactors based on different evaluated neutron data libraries and provides estimates for the uncertainty in criticality associated with nuclear data.","PeriodicalId":100969,"journal":{"name":"Nuclear Energy and Technology","volume":"38 1","pages":"103-109"},"PeriodicalIF":0.0,"publicationDate":"2021-06-21","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"83593058","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 problem of selecting a turbulence model to simulate natural convection near the surface of a VVER-1000 fuel assembly unloaded from the reactor by computational fluid dynamics (CFD simulation) methods. The turbulence model is selected by comparing the calculated data obtained using the Ansys Fluent software package with the results of experimental studies on the natural convection near the surface of a heated vertical plate immersed in water, which simulates the side face of the VVER-1000 fuel assembly in a first approximation. Two-parameter semi-empirical models of turbulence, k-ε and k-ω, are considered as those most commonly used in engineering design. The calculated and experimental data were compared based on the excessive temperature of the plate surface and the water temperature profiles in the turbulent boundary layer for convection modes with a Rayleigh number of 8∙1013 to 3.28∙1014. It has been shown that the best agreement with experimental data, with an average deviation not exceeding ~ 8%, is provided by the RNG k-ε model which is recommended to be used to simulate natural convection near the surface of VVER-1000 FAs in the NPP spent fuel storage pool.
{"title":"Selection of a turbulence model to calculate the temperature profile near the surface of VVER-1000 fuel assemblies in the NPP spent fuel pool","authors":"A. Voronina, S. Pavlov","doi":"10.3897/NUCET.7.68939","DOIUrl":"https://doi.org/10.3897/NUCET.7.68939","url":null,"abstract":"The paper considers the problem of selecting a turbulence model to simulate natural convection near the surface of a VVER-1000 fuel assembly unloaded from the reactor by computational fluid dynamics (CFD simulation) methods. The turbulence model is selected by comparing the calculated data obtained using the Ansys Fluent software package with the results of experimental studies on the natural convection near the surface of a heated vertical plate immersed in water, which simulates the side face of the VVER-1000 fuel assembly in a first approximation. Two-parameter semi-empirical models of turbulence, k-ε and k-ω, are considered as those most commonly used in engineering design. The calculated and experimental data were compared based on the excessive temperature of the plate surface and the water temperature profiles in the turbulent boundary layer for convection modes with a Rayleigh number of 8∙1013 to 3.28∙1014. It has been shown that the best agreement with experimental data, with an average deviation not exceeding ~ 8%, is provided by the RNG k-ε model which is recommended to be used to simulate natural convection near the surface of VVER-1000 FAs in the NPP spent fuel storage pool.","PeriodicalId":100969,"journal":{"name":"Nuclear Energy and Technology","volume":"11 1","pages":"79-84"},"PeriodicalIF":0.0,"publicationDate":"2021-06-21","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"83061722","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 task of determining the radiation situation, including neutron and gamma-quantum flux density, radiation spectrum, specific volumetric activity of radioactive gases in the air, etc. behind the protective composition having inhomogeneities, has always been important in matters of radiation safety. One of the ways to solve the problem of determining gamma radiation fluxes was to divide the total ionizing radiation flux into four components: line-of-sight (LOS), leakage, line-of-sight albedo, and leakage albedo, and obtain an analytical solution for each component. The first three components have been studied in detail in relation to simple geometries and there are analytical solutions for them, but there is no such a solution for the last component. The authors of this work have derived an analytical representation for the leakage albedo component, which, in contrast to numerical methods (such as Monte Carlo methods), makes it possible to analyze the effect of inhomogeneities in protective compositions on the radiation environment as well as to quickly obtain estimated values of fluxes and dose rates. Performing a component-by-component comparison, it becomes possible to single out the most significant mechanisms of the dose load formation behind the nuclear reactor protection, to draw conclusions about the effectiveness of design solutions in the protection design and to improve the protection at significantly lower computational costs. Finally, the authors present calculations for the four components of the total ionizing radiation flux for various parameters of the cylindrical inhomogeneity in the reactor protection. Based on the obtained values, conclusions are made about the importance of taking into account the leakage albedo component in the formation of the radiation situation behind the core vessel.
{"title":"Analytical model for determining the leakage albedo component for a direct cylindrical channel passing through the nuclear reactor protective layer","authors":"K. S. Kupriyanov, Vladimir V. Pereverzentsev","doi":"10.3897/NUCET.7.68941","DOIUrl":"https://doi.org/10.3897/NUCET.7.68941","url":null,"abstract":"The task of determining the radiation situation, including neutron and gamma-quantum flux density, radiation spectrum, specific volumetric activity of radioactive gases in the air, etc. behind the protective composition having inhomogeneities, has always been important in matters of radiation safety. One of the ways to solve the problem of determining gamma radiation fluxes was to divide the total ionizing radiation flux into four components: line-of-sight (LOS), leakage, line-of-sight albedo, and leakage albedo, and obtain an analytical solution for each component. The first three components have been studied in detail in relation to simple geometries and there are analytical solutions for them, but there is no such a solution for the last component. The authors of this work have derived an analytical representation for the leakage albedo component, which, in contrast to numerical methods (such as Monte Carlo methods), makes it possible to analyze the effect of inhomogeneities in protective compositions on the radiation environment as well as to quickly obtain estimated values of fluxes and dose rates. Performing a component-by-component comparison, it becomes possible to single out the most significant mechanisms of the dose load formation behind the nuclear reactor protection, to draw conclusions about the effectiveness of design solutions in the protection design and to improve the protection at significantly lower computational costs.\u0000 Finally, the authors present calculations for the four components of the total ionizing radiation flux for various parameters of the cylindrical inhomogeneity in the reactor protection. Based on the obtained values, conclusions are made about the importance of taking into account the leakage albedo component in the formation of the radiation situation behind the core vessel.","PeriodicalId":100969,"journal":{"name":"Nuclear Energy and Technology","volume":"26 1","pages":""},"PeriodicalIF":0.0,"publicationDate":"2021-06-21","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"77579272","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}
At the present time, JSC Baltiskiy zavod has built and transported to the deployment site at Pevek Akademik Lomonosov, a floating nuclear power unit (FNPU), project 20870. There are also three multi-purpose nuclear icebreakers of project 22220 (Arktika, Sibir, Ural) under construction at Baltiskiy being at different readiness stages. A decision has been made to build a nuclear icebreaker, Lider, of even a higher power. Integral reactors developed by JSC OKBM Afrikantov are installed in the nuclear icebreakers using new assembly-type cores which have not been used earlier in floating facilities. A great deal of preliminary calculation is required to give these cores as advantageous characteristics as possible. The paper proposes a procedure for rapid modeling of floating cores with varied operating and design characteristics. This procedure can be used as part of preliminary modeling. The procedure is based on using a combined dimensionless parameter proposed by the author in (Korolev 2009). A chart is presented to model the key performance of cores for floating objects with a nuclear reactor NPPs. Eight assembly-type core options, which can be installed in transport reactors of a modular or integral design, are analyzed.
目前，JSC Baltiskiy zavod已经建造并运输到Pevek Akademik Lomonosov的部署地点，一个浮动核电机组(FNPU)，项目20870。还有三艘22220项目(北极、西伯利亚、乌拉尔)的多用途核破冰船正在波罗的海建造，处于不同的准备阶段。已经决定建造一艘更大功率的核动力破冰船Lider。由JSC OKBM Afrikantov开发的整体反应堆安装在核破冰船上，使用新的装配式堆芯，这种堆芯以前没有在浮式设施中使用过。要使这些核具有尽可能有利的特性，需要进行大量的初步计算。本文提出了一种具有不同工作特性和设计特性的浮式岩心快速建模方法。这个过程可以作为初步建模的一部分。该程序是基于使用作者在(Korolev 2009)中提出的组合无量纲参数。提出了一种基于核反应堆核电厂的浮物堆芯关键性能模型。分析了可安装在模块化或整体设计的运输堆中的8种组合型堆芯方案。
{"title":"Rapid preliminary modeling of transport reactor cores","authors":"V. Korolev","doi":"10.3897/NUCET.7.65310","DOIUrl":"https://doi.org/10.3897/NUCET.7.65310","url":null,"abstract":"At the present time, JSC Baltiskiy zavod has built and transported to the deployment site at Pevek Akademik Lomonosov, a floating nuclear power unit (FNPU), project 20870. There are also three multi-purpose nuclear icebreakers of project 22220 (Arktika, Sibir, Ural) under construction at Baltiskiy being at different readiness stages. A decision has been made to build a nuclear icebreaker, Lider, of even a higher power. Integral reactors developed by JSC OKBM Afrikantov are installed in the nuclear icebreakers using new assembly-type cores which have not been used earlier in floating facilities. A great deal of preliminary calculation is required to give these cores as advantageous characteristics as possible. The paper proposes a procedure for rapid modeling of floating cores with varied operating and design characteristics. This procedure can be used as part of preliminary modeling. The procedure is based on using a combined dimensionless parameter proposed by the author in (Korolev 2009). A chart is presented to model the key performance of cores for floating objects with a nuclear reactor NPPs. Eight assembly-type core options, which can be installed in transport reactors of a modular or integral design, are analyzed.","PeriodicalId":100969,"journal":{"name":"Nuclear Energy and Technology","volume":"24 1","pages":"41-47"},"PeriodicalIF":0.0,"publicationDate":"2021-03-30","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"81728510","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. V. Didenko, Dmitry Ye. Baluyev, O. Nikanorov, S. Rogozhkin, S. Shepelev, A. Aksenov, Maksim N. Zhestkov, Aleksandr Ye. Shchelyaev
A methodological approach has been developed for the computational investigation of the thermal-hydraulic processes taking place in a sodium cooled fast neutron reactor based on a Russian computational fluid dynamics code, FlowVision. The approach takes into account the integral layout of the reactor primary circuit equipment and the peculiarities of heat exchange in the liquid metal coolant, and makes it possible to model, using well-defined simplifications, the heat and mass exchange in the process of the coolant flowing through the reactor core, and the reactor heat-exchange equipment. Specifically, the methodological approach can be used for justification of safety during the reactor cooldown, as well as for other computational studies which require simulation of the integral reactor core and heat-exchange equipment. The paper presents a brief overview of the methodological approaches developed earlier to study the liquid metal cooled reactor cooldown processes. General principles of these approaches, as well as their advantages and drawbacks have been identified. A three-dimensional computational model of an advanced reactor has been developed, including one heat-exchange loop (a fourth part of the reactor). It has been demonstrated that the FlowVision gap model can be applied to model the space between the reactor core fuel assemblies (interwrapper space), and a porous skeleton model can be used to model the reactor’s heat-exchange equipment. It has been shown that the developed methodological approach is applicable to solving problems of the coolant flow in different operating modes of liquid metal cooled reactor facilities.
{"title":"Development of a methodological approach for the computational investigation of the coolant flow in the process of the sodium cooled reactor cooldown","authors":"D. V. Didenko, Dmitry Ye. Baluyev, O. Nikanorov, S. Rogozhkin, S. Shepelev, A. Aksenov, Maksim N. Zhestkov, Aleksandr Ye. Shchelyaev","doi":"10.3897/NUCET.7.65442","DOIUrl":"https://doi.org/10.3897/NUCET.7.65442","url":null,"abstract":"A methodological approach has been developed for the computational investigation of the thermal-hydraulic processes taking place in a sodium cooled fast neutron reactor based on a Russian computational fluid dynamics code, FlowVision. The approach takes into account the integral layout of the reactor primary circuit equipment and the peculiarities of heat exchange in the liquid metal coolant, and makes it possible to model, using well-defined simplifications, the heat and mass exchange in the process of the coolant flowing through the reactor core, and the reactor heat-exchange equipment. Specifically, the methodological approach can be used for justification of safety during the reactor cooldown, as well as for other computational studies which require simulation of the integral reactor core and heat-exchange equipment. The paper presents a brief overview of the methodological approaches developed earlier to study the liquid metal cooled reactor cooldown processes. General principles of these approaches, as well as their advantages and drawbacks have been identified. A three-dimensional computational model of an advanced reactor has been developed, including one heat-exchange loop (a fourth part of the reactor). It has been demonstrated that the FlowVision gap model can be applied to model the space between the reactor core fuel assemblies (interwrapper space), and a porous skeleton model can be used to model the reactor’s heat-exchange equipment. It has been shown that the developed methodological approach is applicable to solving problems of the coolant flow in different operating modes of liquid metal cooled reactor facilities.","PeriodicalId":100969,"journal":{"name":"Nuclear Energy and Technology","volume":"64 1","pages":"61-66"},"PeriodicalIF":0.0,"publicationDate":"2021-03-30","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"86777813","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 describes experimental studies into the hydrodynamics and heat exchange in a forced water flow in small-diameter channels at low pressures. The timeliness of the studies has been defined by the growing interest in small-size heat exchangers. Small-diameter channels are actively used in components of compact heat exchangers for present-day engineering development applications. The major difficulty involved in investigation of heat-transfer processes in small-diameter channels consists in the absence of common methodologies to calculate coefficients of hydraulic resistance and heat transfer in a two-phase flow. The channel size influences the heat exchange and hydrodynamics of a two-phase flow as one of the determining parameters since the existing internal scales (vapor bubble size, liquid droplet diameter, film thickness) may become commensurable with the channel diameter, this leading potentially to different flow conditions. It is evident that one cannot justifiably expect a change in the momentum and energy transfer regularities in single-phase flows as the channel size is reduced for as long as the continuum approximation remains valid. The authors have analyzed the experiments undertaken by Russian scientists to investigate the distribution of thermal-hydraulic parameters in channels with a small cross-section in the entire variation range of the flow parameters in the channel up to the critical heat flux conditions when the wall temperature increases sharply as the thermal load grows slowly. The experimental critical heat flux data obtained by Russian and foreign authors has been compared.
{"title":"Investigation of the critical heat flux in small-diameter channels","authors":"V. Belozerov, A. Gorbach","doi":"10.3897/NUCET.7.65754","DOIUrl":"https://doi.org/10.3897/NUCET.7.65754","url":null,"abstract":"The paper describes experimental studies into the hydrodynamics and heat exchange in a forced water flow in small-diameter channels at low pressures. The timeliness of the studies has been defined by the growing interest in small-size heat exchangers. Small-diameter channels are actively used in components of compact heat exchangers for present-day engineering development applications. The major difficulty involved in investigation of heat-transfer processes in small-diameter channels consists in the absence of common methodologies to calculate coefficients of hydraulic resistance and heat transfer in a two-phase flow. The channel size influences the heat exchange and hydrodynamics of a two-phase flow as one of the determining parameters since the existing internal scales (vapor bubble size, liquid droplet diameter, film thickness) may become commensurable with the channel diameter, this leading potentially to different flow conditions. It is evident that one cannot justifiably expect a change in the momentum and energy transfer regularities in single-phase flows as the channel size is reduced for as long as the continuum approximation remains valid. The authors have analyzed the experiments undertaken by Russian scientists to investigate the distribution of thermal-hydraulic parameters in channels with a small cross-section in the entire variation range of the flow parameters in the channel up to the critical heat flux conditions when the wall temperature increases sharply as the thermal load grows slowly. The experimental critical heat flux data obtained by Russian and foreign authors has been compared.","PeriodicalId":100969,"journal":{"name":"Nuclear Energy and Technology","volume":"12 1","pages":"73-78"},"PeriodicalIF":0.0,"publicationDate":"2021-03-30","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"86793975","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}
Yevgeniy L. Trykov, A. A. Kudryaev, Konstantin I. Kotsoyev, A. Ananyev
In accordance with Ref. (GOST R 58328-2018 “Pipelines of Nuclear Power Plants. Leak Before Break Concept”), NPPs with VVER-1200 reactors operate an acoustic leak monitoring system (ALMS) and a humidity leak monitoring system (HLMS), each performing the leak monitoring functions locally, independently of the other. The diagnostics results are conveyed to the upper level control system (LCS) to be further displayed for the main control room (MCR) operating personnel. There is also an integrated diagnostics system (IDS) intended to confirm the diagnosis and to update the leak rate values and coordinates based on analyzing the leak monitoring system readings and I&C signals. The system measuring channel readings are composed of background noise, the source for which are processes on the part of the reactor facility’s key components and auxiliary systems, and the leak signal in response to the leak occurrence. A major factor that affects the capability of leak monitoring systems to detect the leak is the quality of the background noise filtering. A new efficient global noise filtering method is proposed for being used as part of the integrated diagnostics system (IDS).
根据参考(GOST R 58328-2018)“核电站管道”。VVER-1200反应堆的核电站运行一个声学泄漏监测系统(ALMS)和一个湿度泄漏监测系统(HLMS)，每个系统在本地执行泄漏监测功能，彼此独立。诊断结果被传送到上层控制系统(LCS)，进一步显示给主控制室(MCR)操作人员。还有一个综合诊断系统(IDS)，用于确认诊断，并根据分析泄漏监测系统读数和I&C信号更新泄漏率值和坐标。该系统测量通道读数由背景噪声和响应泄漏发生的泄漏信号组成。背景噪声的来源是反应堆设施关键部件和辅助系统的过程。影响泄漏监测系统检测泄漏能力的一个主要因素是背景噪声滤波的质量。提出了一种新的有效的全局噪声滤波方法，用于综合诊断系统(IDS)。
{"title":"Efficient method for the global noise filtering in measuring channels of the VVER NPP leak monitoring systems","authors":"Yevgeniy L. Trykov, A. A. Kudryaev, Konstantin I. Kotsoyev, A. Ananyev","doi":"10.3897/NUCET.7.65446","DOIUrl":"https://doi.org/10.3897/NUCET.7.65446","url":null,"abstract":"In accordance with Ref. (GOST R 58328-2018 “Pipelines of Nuclear Power Plants. Leak Before Break Concept”), NPPs with VVER-1200 reactors operate an acoustic leak monitoring system (ALMS) and a humidity leak monitoring system (HLMS), each performing the leak monitoring functions locally, independently of the other. The diagnostics results are conveyed to the upper level control system (LCS) to be further displayed for the main control room (MCR) operating personnel. There is also an integrated diagnostics system (IDS) intended to confirm the diagnosis and to update the leak rate values and coordinates based on analyzing the leak monitoring system readings and I&C signals. The system measuring channel readings are composed of background noise, the source for which are processes on the part of the reactor facility’s key components and auxiliary systems, and the leak signal in response to the leak occurrence. A major factor that affects the capability of leak monitoring systems to detect the leak is the quality of the background noise filtering. A new efficient global noise filtering method is proposed for being used as part of the integrated diagnostics system (IDS).","PeriodicalId":100969,"journal":{"name":"Nuclear Energy and Technology","volume":"42 1","pages":"67-72"},"PeriodicalIF":0.0,"publicationDate":"2021-03-30","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"84724130","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
O. Vlasov, I. Zvonova, P. Krajewski, N. V. Shchukina, S. Chekin, K. Tumanov
The dynamics of 137Cs and 131I radioactivity in the crude biomass of the grass fodder and food vegetation in Mazovia, Poland, in 1986, the year of the Chernobyl accident, has been estimated. Density of 137Cs and 131I in the soil and vegetation have been measured as a function of rainfall and biomass density as of the time most of the fallout took place. A method is described to convert the instrumental data for the radionuclide activity dynamics in vegetation of one type to vegetation of other types. The results of such data conversion from lawn grass to other types of food and fodder grass vegetation are presented. A method is described for adjusting the dynamics of the radionuclide transport through the food chain components (pasture grass, green meat – milk – human body) by normalizing successively the estimated data in each next component for the average value of the instrumental data ratio to the estimated data in the preceding component. The proposed methods are intended to generate a mutually consistent base of estimated and reconstructed instrumental data: 137Cs and 131I activity in the atmosphere – rainfall – 137Cs fallout density on terrain – specific activity of 131I in vegetation. Such radioecological database will provide for a longer reliability of the estimated 131I specific activity dynamics in milk and in human body and, in the long run, when estimating the thyroid internal exposure doses.
{"title":"Radioecological modeling of the 131I activity dynamics in different types of grass vegetation in the Chernobyl accident year","authors":"O. Vlasov, I. Zvonova, P. Krajewski, N. V. Shchukina, S. Chekin, K. Tumanov","doi":"10.3897/NUCET.7.64979","DOIUrl":"https://doi.org/10.3897/NUCET.7.64979","url":null,"abstract":"The dynamics of 137Cs and 131I radioactivity in the crude biomass of the grass fodder and food vegetation in Mazovia, Poland, in 1986, the year of the Chernobyl accident, has been estimated. Density of 137Cs and 131I in the soil and vegetation have been measured as a function of rainfall and biomass density as of the time most of the fallout took place. A method is described to convert the instrumental data for the radionuclide activity dynamics in vegetation of one type to vegetation of other types. The results of such data conversion from lawn grass to other types of food and fodder grass vegetation are presented. A method is described for adjusting the dynamics of the radionuclide transport through the food chain components (pasture grass, green meat – milk – human body) by normalizing successively the estimated data in each next component for the average value of the instrumental data ratio to the estimated data in the preceding component. The proposed methods are intended to generate a mutually consistent base of estimated and reconstructed instrumental data: 137Cs and 131I activity in the atmosphere – rainfall – 137Cs fallout density on terrain – specific activity of 131I in vegetation. Such radioecological database will provide for a longer reliability of the estimated 131I specific activity dynamics in milk and in human body and, in the long run, when estimating the thyroid internal exposure doses.","PeriodicalId":100969,"journal":{"name":"Nuclear Energy and Technology","volume":"90 1","pages":"27-32"},"PeriodicalIF":0.0,"publicationDate":"2021-03-30","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"80395154","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}
Mathematical simulation of fast neutron reactors requires high-precision calculations of protection problems based on unstructured meshes. The paper considers and analyzes a parallel version of the ODETTA code (Belousov et al. 2019) with the use of the MPI (Message Passing Interface) library technology (Knyazeva et al. 2006). The code is designed for numerical simulation of neutronic processes in shielding compositions of fast neutron lead cooled reactor plants in normal operating modes, and can be used to calculate the radiation conditions of using structural components and equipment of nuclear power facilities which are assumed to be the sources of and/or exposed to ionizing radiation during their safety justification. The operation of the generated code is compared against the previous version. The MPIbased development of the ODETTA code’s algorithmic part is described. Peculiarities and specific features of the code parallelization are presented, the code modification is given, and respective algorithms are considered. The structure of the ODETTA code based on the MPI is described in brief. The results of using the ODETTA code’s serial and parallel versions in OS Linux (Kostromin 2012) for NRNU MEPhI’s HPC cluster are provided (Savchenko et al. 2020). A comparative analysis is presented for two code implementation options in terms of speed and accuracy of results when using two different clusters and different numbers of nodes for these. Peculiarities of cluster-based calculations are noted.
快中子反应堆的数学模拟需要基于非结构化网格的高精度保护问题计算。本文考虑并分析了ODETTA代码的并行版本(Belousov等人，2019)，使用MPI(消息传递接口)库技术(Knyazeva等人，2006)。本程序是为快中子铅冷反应堆在正常运行模式下屏蔽成分中的中子过程的数值模拟而设计的，并可用于计算在其安全论证过程中被假定为电离辐射源和/或暴露于电离辐射源的核电设施结构部件和设备的辐射条件。将生成的代码的操作与以前的版本进行比较。介绍了基于mpi的ODETTA代码算法部分的开发。介绍了代码并行化的特点和特点，给出了代码的修改，并考虑了相应的算法。简要介绍了基于MPI的ODETTA代码结构。为NRNU MEPhI的高性能计算集群提供了在Linux操作系统(Kostromin 2012)中使用ODETTA代码的串行和并行版本的结果(Savchenko et al. 2020)。在使用两种不同的集群和不同数量的节点时，对两种代码实现选项在结果的速度和准确性方面进行了比较分析。注意到基于聚类的计算的特点。
{"title":"Speeding up the ODETTA code for solving particle transport problems","authors":"A. Shoshina, V. Belousov","doi":"10.3897/NUCET.7.64365","DOIUrl":"https://doi.org/10.3897/NUCET.7.64365","url":null,"abstract":"Mathematical simulation of fast neutron reactors requires high-precision calculations of protection problems based on unstructured meshes. The paper considers and analyzes a parallel version of the ODETTA code (Belousov et al. 2019) with the use of the MPI (Message Passing Interface) library technology (Knyazeva et al. 2006). The code is designed for numerical simulation of neutronic processes in shielding compositions of fast neutron lead cooled reactor plants in normal operating modes, and can be used to calculate the radiation conditions of using structural components and equipment of nuclear power facilities which are assumed to be the sources of and/or exposed to ionizing radiation during their safety justification. The operation of the generated code is compared against the previous version. The MPIbased development of the ODETTA code’s algorithmic part is described. Peculiarities and specific features of the code parallelization are presented, the code modification is given, and respective algorithms are considered. The structure of the ODETTA code based on the MPI is described in brief. The results of using the ODETTA code’s serial and parallel versions in OS Linux (Kostromin 2012) for NRNU MEPhI’s HPC cluster are provided (Savchenko et al. 2020). A comparative analysis is presented for two code implementation options in terms of speed and accuracy of results when using two different clusters and different numbers of nodes for these. Peculiarities of cluster-based calculations are noted.","PeriodicalId":100969,"journal":{"name":"Nuclear Energy and Technology","volume":"34 1","pages":"15-20"},"PeriodicalIF":0.0,"publicationDate":"2021-03-24","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"82886241","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}
Analysis of corrosion processes has a major role in justifying the reliability and safety of developed nuclear reactors of a new generation with heavy liquid metal coolants. An approach has been developed which allows practical conclusions to be made with respect to the processes in the given metal-oxide system based on analyzing state diagrams for these systems in the “oxidation potential – temperature” coordinates. The proposed approach relies on a long-term experience of experimental and computational studies concerned with the interaction of various steel grades with molten lead and lead-bismuth, as well as with the transport of metal impurities within these molten metals. The oxidation potential of a metal-oxide system is measured in experimental studies using oxygen activity sensors developed and manufactured at IPPE. The applicability of the proposed approach to analyzing the processes of mass transport in iron-oxygen, lead-oxygen, sodium-oxygen, and iron-water vapor systems has been demonstrated.
{"title":"Thermodynamics of equilibrium states and approaches to analyzing the mass transport in metal-oxide systems","authors":"O. Lavrova, Aleksandr Yur’evich Legkikh","doi":"10.3897/NUCET.6.60300","DOIUrl":"https://doi.org/10.3897/NUCET.6.60300","url":null,"abstract":"Analysis of corrosion processes has a major role in justifying the reliability and safety of developed nuclear reactors of a new generation with heavy liquid metal coolants. An approach has been developed which allows practical conclusions to be made with respect to the processes in the given metal-oxide system based on analyzing state diagrams for these systems in the “oxidation potential – temperature” coordinates. The proposed approach relies on a long-term experience of experimental and computational studies concerned with the interaction of various steel grades with molten lead and lead-bismuth, as well as with the transport of metal impurities within these molten metals. The oxidation potential of a metal-oxide system is measured in experimental studies using oxygen activity sensors developed and manufactured at IPPE. The applicability of the proposed approach to analyzing the processes of mass transport in iron-oxygen, lead-oxygen, sodium-oxygen, and iron-water vapor systems has been demonstrated.","PeriodicalId":100969,"journal":{"name":"Nuclear Energy and Technology","volume":"46 1","pages":"261-268"},"PeriodicalIF":0.0,"publicationDate":"2020-11-20","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"81141067","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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