Konstantin I. Kotsoyev, Yevgeny L. Trykov, I. Trykova
Motor operated valves (MOV) are one of the most numerous classes of the nuclear power plant components. An important issue concerned with the MOV diagnostics is the lack of in-process (online) automated control for the MOV technical condition during full power operation of the NPP unit. In this regard, a vital task is that of the MOV diagnostics based on the signals of the current and voltage consumed during MOV ‘opening’ and ‘closing’ operations. The current and voltage signals represent time series measured at regular intervals. The current (and voltage) signals can be received online and contain all necessary information for the online diagnostics of the MOV status. Essentially, the approach allows active power signals to be calculated from the current and voltage signals, and characteristics (‘diagnostic signs’) to be extracted from particular portions (segments) of the active power signals using the values of which MOVs can be diagnosed. The paper deals with the problem of automating the segmentation of active power signals. To accomplish this, an algorithm has been developed based on using a convolutional neural network.
{"title":"Use of a convolutional neural network to segment signals of motor operated valves","authors":"Konstantin I. Kotsoyev, Yevgeny L. Trykov, I. Trykova","doi":"10.3897/nucet.7.73489","DOIUrl":"https://doi.org/10.3897/nucet.7.73489","url":null,"abstract":"Motor operated valves (MOV) are one of the most numerous classes of the nuclear power plant components. An important issue concerned with the MOV diagnostics is the lack of in-process (online) automated control for the MOV technical condition during full power operation of the NPP unit.\u0000 In this regard, a vital task is that of the MOV diagnostics based on the signals of the current and voltage consumed during MOV ‘opening’ and ‘closing’ operations. The current and voltage signals represent time series measured at regular intervals. The current (and voltage) signals can be received online and contain all necessary information for the online diagnostics of the MOV status.\u0000 Essentially, the approach allows active power signals to be calculated from the current and voltage signals, and characteristics (‘diagnostic signs’) to be extracted from particular portions (segments) of the active power signals using the values of which MOVs can be diagnosed.\u0000 The paper deals with the problem of automating the segmentation of active power signals. To accomplish this, an algorithm has been developed based on using a convolutional neural network.","PeriodicalId":100969,"journal":{"name":"Nuclear Energy and Technology","volume":"7 1","pages":""},"PeriodicalIF":0.0,"publicationDate":"2021-09-23","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"74661311","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}
R. A. Vnukov, V. Kolesov, I. Zhavoronkova, Ya. A. Kotov, Md Masum Rana Pramanik
Optimizing the use of fuel in a power reactor is a task of current concern. However, little attention has been given to investigating the dependences among the enrichment used, the content of gadolinium oxide in fuel elements, and the life time in combination with assessing the efficiency of using Gd fuel elements with different Gd2O3 contents. The paper considers fuel assembly (FA) versions for VVER-1200 reactors having different enrichments for fuel elements, including those with Gd, and different contents of gadolinium oxide in fuel. A comparative analysis is presented for assemblies with homogeneous Gd2O3 arrangements in each fuel element and with profiled Gd2O3 arrangements. In the latter case, profiling depends on the neutron flux density in the layer which includes Gd fuel elements. This suggests that the arrangement of gadolinium oxide proportionally to the neutron flux density will improve the FA neutronic performance. The results were obtained using SERPENT (a continuous-energy multi-purpose three-dimensional Monte Carlo particle transport code). The assemblies with the used parameters for a 12-month fuel cycle have shown the method under consideration to be inefficient for a period of over 300 eff. days. With increased enrichment and content of gadolinium oxide, the use of profiled versions has turned out to be more rational for longer periods (up to 900 eff. days). Therefore, this phenomenon is relevant for the reactor life, whereas it proves to be insignificant for the fuel life. A complex relationship is noted between the gadolinium and uranium content in an assembly and the effective multiplication factor for the profiled and standard assemblies. This relationship requires further detailed consideration.
{"title":"Effect of the burnable absorber arrangement on the VVER-1200 fuel assembly neutronic performance","authors":"R. A. Vnukov, V. Kolesov, I. Zhavoronkova, Ya. A. Kotov, Md Masum Rana Pramanik","doi":"10.3897/nucet.7.73490","DOIUrl":"https://doi.org/10.3897/nucet.7.73490","url":null,"abstract":"Optimizing the use of fuel in a power reactor is a task of current concern. However, little attention has been given to investigating the dependences among the enrichment used, the content of gadolinium oxide in fuel elements, and the life time in combination with assessing the efficiency of using Gd fuel elements with different Gd2O3 contents.\u0000 The paper considers fuel assembly (FA) versions for VVER-1200 reactors having different enrichments for fuel elements, including those with Gd, and different contents of gadolinium oxide in fuel. A comparative analysis is presented for assemblies with homogeneous Gd2O3 arrangements in each fuel element and with profiled Gd2O3 arrangements. In the latter case, profiling depends on the neutron flux density in the layer which includes Gd fuel elements. This suggests that the arrangement of gadolinium oxide proportionally to the neutron flux density will improve the FA neutronic performance.\u0000 The results were obtained using SERPENT (a continuous-energy multi-purpose three-dimensional Monte Carlo particle transport code). The assemblies with the used parameters for a 12-month fuel cycle have shown the method under consideration to be inefficient for a period of over 300 eff. days. With increased enrichment and content of gadolinium oxide, the use of profiled versions has turned out to be more rational for longer periods (up to 900 eff. days). Therefore, this phenomenon is relevant for the reactor life, whereas it proves to be insignificant for the fuel life. A complex relationship is noted between the gadolinium and uranium content in an assembly and the effective multiplication factor for the profiled and standard assemblies. This relationship requires further detailed consideration.","PeriodicalId":100969,"journal":{"name":"Nuclear Energy and Technology","volume":"153 2 1","pages":""},"PeriodicalIF":0.0,"publicationDate":"2021-09-23","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"77341880","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}
V. Potetnya, E. Koryakina, M. Troshina, S. Koryakin
The paper investigates the characteristics of the chemical Fricke dosimeter (with the standard composition (D1), without NaCl addition to the solution (D2), without NaCl but with a tenfold increased concentration of Fe2+ (D3)) under continuous and pulsed irradiation with an ultra-high dose rate of the BARS-6 reactor with unshielded metallic cores. The dosimeter radiosensitivity had a linear dependence on the gamma neutron radiation dose in a range of 25 to 750 Gy and was respectively 1.96 ± 0.05 μGy–1 (D1), 2.04 ± 0.05 μGy–1 (D2), and 2.08 ± 0.5 μGy–1 (D3) in the continuous irradiation mode, and 1.24 ± 0.05 μGy–1, 2.00 ± 0.05 μGy–1, and 1.94 ± 0.05 μGy–1 in the pulsed irradiation mode. This makes ≈ 60% of their sensitivity to the 60Со gamma radiation (3.40 ± 0.02 μGy–1), and 36%, 1.6 times as less, for a standard Fricke dosimeter irradiated in the pulsed mode. The experimental value of the radiation chemical yield, Gn(Fe3+), for all solution modifications and both irradiation modes varied slightly and was 0.84 ± 0.11 μM/J on the average, except for the standard solution in the pulsed mode (0.66 ± 0.07 μM/J). The neutron doses determined by chemical and activation dosimeters coincided within the error limits, but the chemical dosimeter readings were systematically higher, by about 20%. Therefore, in the fission spectrum neutron dose rate range of 0.4 to 7×108 Gy/min, there is no dose rate effect both in the standard Fricke dosimeter version (without NaCl) and in the modified version, which makes it possible to use modified Fricke dosimeters to assess the physical and dosimetry characteristics of mixed gamma neutron radiation beams.
{"title":"Use of the chemical Fricke dosimeter and its modifications for dosimetry of gamma neutron radiation of a pulsed reactor","authors":"V. Potetnya, E. Koryakina, M. Troshina, S. Koryakin","doi":"10.3897/nucet.7.74149","DOIUrl":"https://doi.org/10.3897/nucet.7.74149","url":null,"abstract":"The paper investigates the characteristics of the chemical Fricke dosimeter (with the standard composition (D1), without NaCl addition to the solution (D2), without NaCl but with a tenfold increased concentration of Fe2+ (D3)) under continuous and pulsed irradiation with an ultra-high dose rate of the BARS-6 reactor with unshielded metallic cores.\u0000 The dosimeter radiosensitivity had a linear dependence on the gamma neutron radiation dose in a range of 25 to 750 Gy and was respectively 1.96 ± 0.05 μGy–1 (D1), 2.04 ± 0.05 μGy–1 (D2), and 2.08 ± 0.5 μGy–1 (D3) in the continuous irradiation mode, and 1.24 ± 0.05 μGy–1, 2.00 ± 0.05 μGy–1, and 1.94 ± 0.05 μGy–1 in the pulsed irradiation mode. This makes ≈ 60% of their sensitivity to the 60Со gamma radiation (3.40 ± 0.02 μGy–1), and 36%, 1.6 times as less, for a standard Fricke dosimeter irradiated in the pulsed mode. The experimental value of the radiation chemical yield, Gn(Fe3+), for all solution modifications and both irradiation modes varied slightly and was 0.84 ± 0.11 μM/J on the average, except for the standard solution in the pulsed mode (0.66 ± 0.07 μM/J). The neutron doses determined by chemical and activation dosimeters coincided within the error limits, but the chemical dosimeter readings were systematically higher, by about 20%.\u0000 Therefore, in the fission spectrum neutron dose rate range of 0.4 to 7×108 Gy/min, there is no dose rate effect both in the standard Fricke dosimeter version (without NaCl) and in the modified version, which makes it possible to use modified Fricke dosimeters to assess the physical and dosimetry characteristics of mixed gamma neutron radiation beams.","PeriodicalId":100969,"journal":{"name":"Nuclear Energy and Technology","volume":"63 1","pages":""},"PeriodicalIF":0.0,"publicationDate":"2021-09-23","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"88666318","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}
I. Gusev, Aleksandr P. Vorobyev, Mikhail N. Kozlovsky, S. P. Padun
Introduction. The article analyzes the operation of Unit 1 and 2 of Novovoronezh Nuclear Power Plant II (equipped with VVER-1200 reactors) with two electrical feed pumps disabled and the backup pump not enabled. These operating conditions are subsequently simulated using the power unit model software-hardware package (PUM SHC) developed by LLC IF SNIIP ATOM. Research objectives. The objective of this work was to check the reliability of the forecasts of changes in the power unit parameters obtained using the PUM SHC, based on operational data. Methods. The simulated power unit parameter changes in transient conditions were in good agreement with the data collected in real tests. During the simulation, the power unit dynamic stability was preserved, i.e., the operational parameters were within the design limits and did not exceed the protection operation set points. Results. The results of the work suggest the possibility of using current NPP power unit simulations: for developing proposals for adjusting the operation control algorithms in case of malfunctions and emergency modes with the main equipment shutdown and power unit protection actuation; and for verifying design solutions for updating the power unit systems, which are associated with the use of new equipment or changes in flow diagrams. Conclusion. Current power unit models can be applied both for existing power units and for new ones that are being commissioned.
介绍。本文分析了新沃罗涅日核电站1号机组和2号机组(配备VVER-1200反应堆)在两台给水泵停用、备用泵未启用的情况下的运行情况。随后，使用LLC IF SNIIP ATOM开发的动力单元模型软硬件包(PUM SHC)对这些运行条件进行了模拟。研究的目标。这项工作的目的是检查使用PUM SHC根据运行数据获得的动力单元参数变化预测的可靠性。方法。模拟的机组暂态参数变化与实际试验数据吻合较好。在仿真过程中，保持了机组的动态稳定性，即运行参数在设计范围内，不超过保护运行设定点。结果。研究结果表明，有可能采用目前的核电站动力单元模拟:在主设备关闭和动力单元保护启动的故障和紧急模式下，为调整运行控制算法提出建议;并验证更新动力单元系统的设计解决方案，这些解决方案与使用新设备或流程图的变化有关。结论。目前的动力单元模型既适用于现有的动力单元，也适用于正在服役的新动力单元。
{"title":"Simulating operation of power units 1 and 2 at Novovoronezh NPP II with two electrical feed pumps disabled and the backup pump not enabled","authors":"I. Gusev, Aleksandr P. Vorobyev, Mikhail N. Kozlovsky, S. P. Padun","doi":"10.3897/nucet.7.72394","DOIUrl":"https://doi.org/10.3897/nucet.7.72394","url":null,"abstract":"Introduction. The article analyzes the operation of Unit 1 and 2 of Novovoronezh Nuclear Power Plant II (equipped with VVER-1200 reactors) with two electrical feed pumps disabled and the backup pump not enabled. These operating conditions are subsequently simulated using the power unit model software-hardware package (PUM SHC) developed by LLC IF SNIIP ATOM.\u0000 Research objectives. The objective of this work was to check the reliability of the forecasts of changes in the power unit parameters obtained using the PUM SHC, based on operational data.\u0000 Methods. The simulated power unit parameter changes in transient conditions were in good agreement with the data collected in real tests. During the simulation, the power unit dynamic stability was preserved, i.e., the operational parameters were within the design limits and did not exceed the protection operation set points.\u0000 Results. The results of the work suggest the possibility of using current NPP power unit simulations:\u0000 for developing proposals for adjusting the operation control algorithms in case of malfunctions and emergency modes with the main equipment shutdown and power unit protection actuation; and\u0000 for verifying design solutions for updating the power unit systems, which are associated with the use of new equipment or changes in flow diagrams.\u0000 Conclusion. Current power unit models can be applied both for existing power units and for new ones that are being commissioned.","PeriodicalId":100969,"journal":{"name":"Nuclear Energy and Technology","volume":"72 1","pages":""},"PeriodicalIF":0.0,"publicationDate":"2021-09-23","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"87914233","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}
Expensive permanent storage facilities with massive engineered structures are used traditionally to ensure safe temporary storage of solidified radioactive waste at the NPP sites. Such approach is dictated by the need to comply with the regulatory requirements for limiting the gamma background in the area adjacent to the storage facility. The costs involved in temporary storage of solidified RW can be optimized by using light hangar-type storage facilities. At the same time, the safety of storage, including radiation protection of the personnel, the public and the environment, is undoubtedly ensured through the use of special organizational and engineering solutions. The Novovoronezh NPP, a branch of JSC Concern Rosenergoatom, operates successfully light hangar-type facilities for temporary storage of solidified RW classified as medium-level waste in accordance with OSPORB-99/2009. In the process of operation, a methodology and a method for conditioning and temporary storage of solidified RW were developed to ensure the RW removal for final disposal with no extra process operations and unreasonable costs. A methodology has been developed to assess the radiation situation around storage facilities during temporary storage of RW, as well as a software package for predicting the radiation situation when deciding on the arrangement of the storage facility’s peripheral rows.
{"title":"Ensuring radiation safety during temporary storage of solidified radioactive waste in light hangar-type facilities","authors":"Sergey V. Rosnovsky, V. Povarov","doi":"10.3897/nucet.7.73487","DOIUrl":"https://doi.org/10.3897/nucet.7.73487","url":null,"abstract":"Expensive permanent storage facilities with massive engineered structures are used traditionally to ensure safe temporary storage of solidified radioactive waste at the NPP sites. Such approach is dictated by the need to comply with the regulatory requirements for limiting the gamma background in the area adjacent to the storage facility.\u0000 The costs involved in temporary storage of solidified RW can be optimized by using light hangar-type storage facilities. At the same time, the safety of storage, including radiation protection of the personnel, the public and the environment, is undoubtedly ensured through the use of special organizational and engineering solutions.\u0000 The Novovoronezh NPP, a branch of JSC Concern Rosenergoatom, operates successfully light hangar-type facilities for temporary storage of solidified RW classified as medium-level waste in accordance with OSPORB-99/2009. In the process of operation, a methodology and a method for conditioning and temporary storage of solidified RW were developed to ensure the RW removal for final disposal with no extra process operations and unreasonable costs.\u0000 A methodology has been developed to assess the radiation situation around storage facilities during temporary storage of RW, as well as a software package for predicting the radiation situation when deciding on the arrangement of the storage facility’s peripheral rows.","PeriodicalId":100969,"journal":{"name":"Nuclear Energy and Technology","volume":"11 1","pages":""},"PeriodicalIF":0.0,"publicationDate":"2021-09-23","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"77762819","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}
V. Ulyanov, M. Koshelev, V. S. Kremlyova, S. Kharchuk
The paper presents a computational analysis of regularities in the accumulation of slags during the interaction of lead and lead-bismuth coolants with oxygen gas. Oxidation of lead-containing coolants will cause the formation of lead oxide, while the formation of bismuth oxide is unlikely. Dosed supply of oxidizing gas to lead-containing coolants makes it possible to oxidize, selectively, chromium and nickel to their oxides without the slag formation from solid lead oxide. Regularities were studied which are involved in the lead oxide formation during the interaction of lead-containing coolants with oxygen gas. It has been found that, in the process of interacting with oxygen gas, a lead-bismuth alloy is oxidized 1.7 times as intensively as lead, this being explained by the presence of bismuth in the alloy. Bismuth is oxidized more intensively than both lead and the lead-bismuth alloy. The inert gas overpressure during depressurization does not prevent air oxygen from entering the circuit, and the dependence of the nitrogen and oxygen flow into the circuit on the argon flow out of the loop is close to linear regardless of the circuit state (cold, without coolant; heated, without coolant; heated, with circulating coolant). Oxygen is a chemically active impurity and is absorbed by the circuit; it is therefore important to control nitrogen in the gas spaces of the reactor and research plant circuits with lead-containing coolants. This will make it possible to signal, in a timely manner, the ingress of oxygen into the circuit and to take measures required to avoid or reduce the scale of the slag formation from lead oxides.
{"title":"Investigations of regularities in the accumulation of hydrogen-reduced slags in circulation circuits with lead-containing coolants","authors":"V. Ulyanov, M. Koshelev, V. S. Kremlyova, S. Kharchuk","doi":"10.3897/nucet.7.74154","DOIUrl":"https://doi.org/10.3897/nucet.7.74154","url":null,"abstract":"The paper presents a computational analysis of regularities in the accumulation of slags during the interaction of lead and lead-bismuth coolants with oxygen gas. Oxidation of lead-containing coolants will cause the formation of lead oxide, while the formation of bismuth oxide is unlikely. Dosed supply of oxidizing gas to lead-containing coolants makes it possible to oxidize, selectively, chromium and nickel to their oxides without the slag formation from solid lead oxide. Regularities were studied which are involved in the lead oxide formation during the interaction of lead-containing coolants with oxygen gas. It has been found that, in the process of interacting with oxygen gas, a lead-bismuth alloy is oxidized 1.7 times as intensively as lead, this being explained by the presence of bismuth in the alloy. Bismuth is oxidized more intensively than both lead and the lead-bismuth alloy. The inert gas overpressure during depressurization does not prevent air oxygen from entering the circuit, and the dependence of the nitrogen and oxygen flow into the circuit on the argon flow out of the loop is close to linear regardless of the circuit state (cold, without coolant; heated, without coolant; heated, with circulating coolant). Oxygen is a chemically active impurity and is absorbed by the circuit; it is therefore important to control nitrogen in the gas spaces of the reactor and research plant circuits with lead-containing coolants. This will make it possible to signal, in a timely manner, the ingress of oxygen into the circuit and to take measures required to avoid or reduce the scale of the slag formation from lead oxides.","PeriodicalId":100969,"journal":{"name":"Nuclear Energy and Technology","volume":"159 1","pages":""},"PeriodicalIF":0.0,"publicationDate":"2021-09-23","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"86243269","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}
V. Usanov, Stepan A. Kvyatkovskiy, A. Andrianov, I. Kuptsov
The paper presents the results from a multi-criteria comparative evaluation of potential deployment scenarios for Russian nuclear power with thermal and sodium-cooled fast reactors in a closed nuclear fuel cycle (the so-called two-component nuclear energy system). The comparison and the ranking were performed taking into account the recommendations and using the IAEA/INPRO software tools for comparative evaluation of nuclear energy systems, including tools for sensitivity/uncertainty analysis with respect to weighting factors. Ten potential Russian nuclear power deployment scenarios with different shares of thermal and sodium-cooled fast reactors were considered, including options involving the use of MOX fuel in VVER reactors. Eight key indicators were used, estimated as of 2100 and structured into a three-level objectives tree. The comparative evaluation and the ranking were carried out based on the multi-attribute value theory. The model for assessing the key indicators was developed using the IAEA/INPRO MESSAGE-NES energy system planning software tool. The information base for the study was formed by publications of experts from JSC SSC RF-IPPE, NRC Kurchatov Institute and NRNU MEPhI. The presented results show that it is possible to enhance significantly the sustainability of the Russian nuclear energy system, when considering multiple performance indicators, through the intensive deployment of sodium-cooled fast reactors and the transition to a closed nuclear fuel cycle. Tasks have been outlined for the follow-up studies to make it possible to obtain more rigorous conclusions regarding the preferred options for the evolution of a two-component nuclear energy system.
{"title":"Multi-criteria evaluation and ranking of potential scenarios for the development of Russian two-component nuclear energy system with thermal and sodium-cooled fast reactors","authors":"V. Usanov, Stepan A. Kvyatkovskiy, A. Andrianov, I. Kuptsov","doi":"10.3897/nucet.7.72391","DOIUrl":"https://doi.org/10.3897/nucet.7.72391","url":null,"abstract":"The paper presents the results from a multi-criteria comparative evaluation of potential deployment scenarios for Russian nuclear power with thermal and sodium-cooled fast reactors in a closed nuclear fuel cycle (the so-called two-component nuclear energy system). The comparison and the ranking were performed taking into account the recommendations and using the IAEA/INPRO software tools for comparative evaluation of nuclear energy systems, including tools for sensitivity/uncertainty analysis with respect to weighting factors. Ten potential Russian nuclear power deployment scenarios with different shares of thermal and sodium-cooled fast reactors were considered, including options involving the use of MOX fuel in VVER reactors. Eight key indicators were used, estimated as of 2100 and structured into a three-level objectives tree. The comparative evaluation and the ranking were carried out based on the multi-attribute value theory. The model for assessing the key indicators was developed using the IAEA/INPRO MESSAGE-NES energy system planning software tool. The information base for the study was formed by publications of experts from JSC SSC RF-IPPE, NRC Kurchatov Institute and NRNU MEPhI. The presented results show that it is possible to enhance significantly the sustainability of the Russian nuclear energy system, when considering multiple performance indicators, through the intensive deployment of sodium-cooled fast reactors and the transition to a closed nuclear fuel cycle. Tasks have been outlined for the follow-up studies to make it possible to obtain more rigorous conclusions regarding the preferred options for the evolution of a two-component nuclear energy system.","PeriodicalId":100969,"journal":{"name":"Nuclear Energy and Technology","volume":"105 1","pages":""},"PeriodicalIF":0.0,"publicationDate":"2021-09-14","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"80596110","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}
I. Volkov, Viktoriya O. Zharkova, Y. Y. Karaseva, Еlena I. Lysakova, Е.V. Zakharova
The purpose of the work was to investigate the sorptive capacity of natural clay samples with respect to 90Sr and 137Cs to assess the possibility of using these as components of protective barriers at radioactive waste isolation facilities. Bentonite clays of the Zyryanskoye and Desyaty Khutor deposits and high-melting clay of the Kampanovskoye deposit were selected for the investigation. The capacity of clays for sorption through ionic exchange is characterized by the value of the cation exchange capacity (CEC). In the process of sorption experiments, all of the test clays displayed a high rate of extracting strontium and cesium radionuclides from aqueous solutions. It was shown that the sorption of 90Sr is affected by the content of montmorillonite in the samples: bentonite clays absorb up to 98–99% of the initial radionuclide content in the solution, while about 80% of strontium is sorbed by high-melting clay. Cesium is practically fully sorbed by the tested samples and the degree of sorption amounts to over 99%, the highest value of the distribution coefficient having been recorded for the Kampanovskoye sample (Kd = 5.0×103 cm3/g). The method of sorbed radionuclides fixation on the clay samples were identified by selective desorption using the modified Tessier methodology. It was shown that strontium ions are more mobile than ions of cesium up to 97% of which is retained by clays.
{"title":"Sorption of 90Sr and 137Cs on clays used to build safety barriers in radioactive waste storage facilities","authors":"I. Volkov, Viktoriya O. Zharkova, Y. Y. Karaseva, Еlena I. Lysakova, Е.V. Zakharova","doi":"10.3897/nucet.7.69930","DOIUrl":"https://doi.org/10.3897/nucet.7.69930","url":null,"abstract":"The purpose of the work was to investigate the sorptive capacity of natural clay samples with respect to 90Sr and 137Cs to assess the possibility of using these as components of protective barriers at radioactive waste isolation facilities. Bentonite clays of the Zyryanskoye and Desyaty Khutor deposits and high-melting clay of the Kampanovskoye deposit were selected for the investigation. The capacity of clays for sorption through ionic exchange is characterized by the value of the cation exchange capacity (CEC). In the process of sorption experiments, all of the test clays displayed a high rate of extracting strontium and cesium radionuclides from aqueous solutions. It was shown that the sorption of 90Sr is affected by the content of montmorillonite in the samples: bentonite clays absorb up to 98–99% of the initial radionuclide content in the solution, while about 80% of strontium is sorbed by high-melting clay. Cesium is practically fully sorbed by the tested samples and the degree of sorption amounts to over 99%, the highest value of the distribution coefficient having been recorded for the Kampanovskoye sample (Kd = 5.0×103 cm3/g). The method of sorbed radionuclides fixation on the clay samples were identified by selective desorption using the modified Tessier methodology. It was shown that strontium ions are more mobile than ions of cesium up to 97% of which is retained by clays.","PeriodicalId":100969,"journal":{"name":"Nuclear Energy and Technology","volume":"1 1","pages":""},"PeriodicalIF":0.0,"publicationDate":"2021-06-24","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"87050298","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}
Valeriy Ivanovich Baranenko, O. M. Gulina, N. L. Salnikov
Modern foreign computer codes predict a linear growth in the pipeline wall thinning with time due to the process of flow-accelerated corrosion (FAC), i.e. erosion-corrosion wear (ECW). Linear time-thinning dependence and corrosion rate constancy are not however typical of the NPP piping operating conditions. And the associated excessive conservatism of the residual life estimates leads to increased economic costs of repeated inspections. In domestic software tools, EKI-02 and EKI-03, the influence of operating time are taken into account by introducing the respective coefficient into the Chexal-Horowitz model based on the yield of corrosion products into the coolant. The ECW intensity can be however reduced through improvements in operating conditions, preventive measures, improvements in water chemistry, etc., and the use of the dependences once obtained may turn out to be too conservative. Based on a large number of repeated measurements as well as on data from corrosion testers, it has been shown that the influence of time can be described by the function of a particular form, the coefficients of which differ for different units and component and subsystem types. This makes it possible to determine the ‘aging function’ based on inspection data, and then use it in a targeted way for particular components. It has been shown that such estimates are much less conservative.
{"title":"Influence of operating time on the corrosion in single-phase and two-phase media","authors":"Valeriy Ivanovich Baranenko, O. M. Gulina, N. L. Salnikov","doi":"10.3897/NUCET.7.69175","DOIUrl":"https://doi.org/10.3897/NUCET.7.69175","url":null,"abstract":"Modern foreign computer codes predict a linear growth in the pipeline wall thinning with time due to the process of flow-accelerated corrosion (FAC), i.e. erosion-corrosion wear (ECW). Linear time-thinning dependence and corrosion rate constancy are not however typical of the NPP piping operating conditions. And the associated excessive conservatism of the residual life estimates leads to increased economic costs of repeated inspections. In domestic software tools, EKI-02 and EKI-03, the influence of operating time are taken into account by introducing the respective coefficient into the Chexal-Horowitz model based on the yield of corrosion products into the coolant. The ECW intensity can be however reduced through improvements in operating conditions, preventive measures, improvements in water chemistry, etc., and the use of the dependences once obtained may turn out to be too conservative. Based on a large number of repeated measurements as well as on data from corrosion testers, it has been shown that the influence of time can be described by the function of a particular form, the coefficients of which differ for different units and component and subsystem types. This makes it possible to determine the ‘aging function’ based on inspection data, and then use it in a targeted way for particular components. It has been shown that such estimates are much less conservative.","PeriodicalId":100969,"journal":{"name":"Nuclear Energy and Technology","volume":"14 1","pages":"127-132"},"PeriodicalIF":0.0,"publicationDate":"2021-06-23","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"88976387","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
S. R. Askhadullin, Victor Konstantinovich Milinchuk
The process of hydrogen formation and the associated risk of combustion and explosion is a complex problem concerned with the hydrogen and radiation safety of nuclear reactors. Lithium, potassium and sodium hydroxides are used in VVER reactors as corrective additives for keeping the hydrogen potential of the water coolant with boric acid at a controlled level of 5.8 to 10.3. In the process of investigating the interaction of aqueous solutions of the above hydroxides with aluminum, the most chemically active of these is lithium hydroxide; this reaction proceeds with hydrogen formed at a high rate at room temperature (in an exothermic mode). The processes of hydrogen generation in hydroheterogeneous compositions with potassium and sodium hydroxides proceed at an acceptable rate with heating to ~ 60 °C. The kinetics of hydrogen generation depends in a complex way on the content of boric acid, namely, the hydrogen yield is at a level of ~ 1000 ml at a low concentration of 0.01 to 0.05 g/l, and there is no hydrogen formation at a concentration of 0.6 g/l. According to the coolant quality standards, in the hot state of a VVER-1000 unit or in the reactor state at the minimum controlled power level, the total concentration of alkali metals is about 1 mg/dm3, i.e. two to three orders of magnitude as less as in the investigated compositions. The discovery of the influence of alkali metal hydroxides on the formation of hydrogen with the participation of structural materials based on the example of aluminum makes it possible to suggest that the hydroxides of these metals contained in the coolant in a small amount can also take part in the hydroheterogeneous process of formation of minor hydrogen amounts. The potential for hydrogen formation in such a way needs to be taken into account during long-term operation of nuclear reactors, and during accidents and incidents at NPPs
{"title":"Generation of hydrogen by hydroheterogeneous compositions based on aluminum and alkali metals","authors":"S. R. Askhadullin, Victor Konstantinovich Milinchuk","doi":"10.3897/NUCET.7.69178","DOIUrl":"https://doi.org/10.3897/NUCET.7.69178","url":null,"abstract":"The process of hydrogen formation and the associated risk of combustion and explosion is a complex problem concerned with the hydrogen and radiation safety of nuclear reactors. Lithium, potassium and sodium hydroxides are used in VVER reactors as corrective additives for keeping the hydrogen potential of the water coolant with boric acid at a controlled level of 5.8 to 10.3. In the process of investigating the interaction of aqueous solutions of the above hydroxides with aluminum, the most chemically active of these is lithium hydroxide; this reaction proceeds with hydrogen formed at a high rate at room temperature (in an exothermic mode). The processes of hydrogen generation in hydroheterogeneous compositions with potassium and sodium hydroxides proceed at an acceptable rate with heating to ~ 60 °C. The kinetics of hydrogen generation depends in a complex way on the content of boric acid, namely, the hydrogen yield is at a level of ~ 1000 ml at a low concentration of 0.01 to 0.05 g/l, and there is no hydrogen formation at a concentration of 0.6 g/l. According to the coolant quality standards, in the hot state of a VVER-1000 unit or in the reactor state at the minimum controlled power level, the total concentration of alkali metals is about 1 mg/dm3, i.e. two to three orders of magnitude as less as in the investigated compositions. The discovery of the influence of alkali metal hydroxides on the formation of hydrogen with the participation of structural materials based on the example of aluminum makes it possible to suggest that the hydroxides of these metals contained in the coolant in a small amount can also take part in the hydroheterogeneous process of formation of minor hydrogen amounts. The potential for hydrogen formation in such a way needs to be taken into account during long-term operation of nuclear reactors, and during accidents and incidents at NPPs","PeriodicalId":100969,"journal":{"name":"Nuclear Energy and Technology","volume":"18 1","pages":"133-138"},"PeriodicalIF":0.0,"publicationDate":"2021-06-23","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"81134629","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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