The operating experience of Novovoronezh NPP II-1 shows that, in the summer period, the temperature of the cooling water exceeds the design value: this indicates the insufficient performance of the service water supply system. The main factor that has a negative impact on the performance of this system is the formation of carbonate deposits on the cooling tower filler. At Novovoronezh NPP II-1, the cooling tower water distribution system was cleaned from carbonate deposits by the method of combined vibration and aerohydraulic impact. The tested method of cleaning the filler cannot be considered optimal, since the main stage that determines the entire cleaning duration is the assembly/disassembly of the cooling tower filler. It is necessary to continue research on the choice of a strategy for controlling the carbonate deposition rate, taking into account the revealed influence of the design features of the main cooling water pipelines and pipelines of the cooling tower water distribution system on the mechanism of deposit formation in the peripheral spraying area. As compensating measures to ensure the required temperature regime of the turbine plant equipment at Novovoronezh NPP II-1, it is practiced during the summer period to put the standby heat exchangers of the lubrication system and the standby pump of the nonessential services cooling water system into parallel operation. This solution is fraught with the risk of an unplanned decrease in the electrical load if this equipment is turned off in the event of a malfunction. To increase the operating stability of Novovoronezh NPP II-1 and -2 in the summer period, it is proposed to carry out a number of measures aimed at mitigating the negative consequences caused by the elevated service water temperature. Equipment upgrade options are evaluated, e.g., by installing an additional pump for the turbine building services cooling system and (or) laying an additional pipeline to supply part of the makeup water from the Don River directly to the suction pipelines of the pumps of the turbine building services cooling system.
{"title":"Operating experience and ways to improve the performance of the service water supply system at the Novovoronezh NPP II (Units 1 and 2)","authors":"V. Povarov, D. B. Statsura, Dmitry Y. Usachev","doi":"10.3897/nucet.6.60461","DOIUrl":"https://doi.org/10.3897/nucet.6.60461","url":null,"abstract":"The operating experience of Novovoronezh NPP II-1 shows that, in the summer period, the temperature of the cooling water exceeds the design value: this indicates the insufficient performance of the service water supply system. The main factor that has a negative impact on the performance of this system is the formation of carbonate deposits on the cooling tower filler. At Novovoronezh NPP II-1, the cooling tower water distribution system was cleaned from carbonate deposits by the method of combined vibration and aerohydraulic impact. The tested method of cleaning the filler cannot be considered optimal, since the main stage that determines the entire cleaning duration is the assembly/disassembly of the cooling tower filler. It is necessary to continue research on the choice of a strategy for controlling the carbonate deposition rate, taking into account the revealed influence of the design features of the main cooling water pipelines and pipelines of the cooling tower water distribution system on the mechanism of deposit formation in the peripheral spraying area. As compensating measures to ensure the required temperature regime of the turbine plant equipment at Novovoronezh NPP II-1, it is practiced during the summer period to put the standby heat exchangers of the lubrication system and the standby pump of the nonessential services cooling water system into parallel operation. This solution is fraught with the risk of an unplanned decrease in the electrical load if this equipment is turned off in the event of a malfunction. To increase the operating stability of Novovoronezh NPP II-1 and -2 in the summer period, it is proposed to carry out a number of measures aimed at mitigating the negative consequences caused by the elevated service water temperature. Equipment upgrade options are evaluated, e.g., by installing an additional pump for the turbine building services cooling system and (or) laying an additional pipeline to supply part of the makeup water from the Don River directly to the suction pipelines of the pumps of the turbine building services cooling system.","PeriodicalId":100969,"journal":{"name":"Nuclear Energy and Technology","volume":"46 1","pages":"253-260"},"PeriodicalIF":0.0,"publicationDate":"2020-11-20","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"90536037","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
A. Zrodnikov, V. Korobeynikov, A. Moseev, A. Egorov
Multi-criteria analysis is used in many areas of research where it is required to compare several alternatives according to a selected set of criteria. Of particular interest is the application of this method for a comparative assessment of the efficiency of scenarios for the development of innovative nuclear systems. The article proposes an approach to the computational substantiation of the step-by-step transfer of the Russian nuclear industry to a two-component nuclear energy system (NES) with a centralized closed nuclear fuel cycle (NFC) based on the multi-criteria analysis method. At the same time, consideration is given to options for the development of the domestic nuclear industry in view of the uncertain prospects for the future. Taking into account various trends in the nuclear energy development, the authors identify the following three groups of possible scenarios. The first group includes ‘growing’ scenarios in which the number of units and their total installed capacity grow over time. The second group assumes that after a certain time of growth of the installed capacities, the stationary level will be reached, in which there will be no time-dependent capacity changes. The third group simulates a decrease in the installed nuclear energy capacities in the country after some growth. To select the most preferable ways of technological development and assess the efficiency of a nuclear energy system, a limited set of selection criteria and performance indicators are used, covering the economy, export potential, competitiveness, efficient SNF and RW management, natural uranium consumption, and innovative development potential. An important part of this work was a detailed analysis of the uncertainties in the weights and input data used to derive the criteria.
{"title":"Multi-criteria analysis of the efficiency of scenarios for the development of the Russian nuclear industry in view of the uncertain prospects for the future","authors":"A. Zrodnikov, V. Korobeynikov, A. Moseev, A. Egorov","doi":"10.3897/nucet.6.60557","DOIUrl":"https://doi.org/10.3897/nucet.6.60557","url":null,"abstract":"Multi-criteria analysis is used in many areas of research where it is required to compare several alternatives according to a selected set of criteria. Of particular interest is the application of this method for a comparative assessment of the efficiency of scenarios for the development of innovative nuclear systems. The article proposes an approach to the computational substantiation of the step-by-step transfer of the Russian nuclear industry to a two-component nuclear energy system (NES) with a centralized closed nuclear fuel cycle (NFC) based on the multi-criteria analysis method. At the same time, consideration is given to options for the development of the domestic nuclear industry in view of the uncertain prospects for the future. Taking into account various trends in the nuclear energy development, the authors identify the following three groups of possible scenarios. The first group includes ‘growing’ scenarios in which the number of units and their total installed capacity grow over time. The second group assumes that after a certain time of growth of the installed capacities, the stationary level will be reached, in which there will be no time-dependent capacity changes. The third group simulates a decrease in the installed nuclear energy capacities in the country after some growth. To select the most preferable ways of technological development and assess the efficiency of a nuclear energy system, a limited set of selection criteria and performance indicators are used, covering the economy, export potential, competitiveness, efficient SNF and RW management, natural uranium consumption, and innovative development potential. An important part of this work was a detailed analysis of the uncertainties in the weights and input data used to derive the criteria.","PeriodicalId":100969,"journal":{"name":"Nuclear Energy and Technology","volume":"57 1","pages":"299-305"},"PeriodicalIF":0.0,"publicationDate":"2020-11-20","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"74171698","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 kinetics of nuclear reactors is determined by the average neutron lifetime. When the inserted reactivity is more than the effective delayed neutron fraction, the reactor kinetics becomes very rapid. It is possible to slow down the fast reactor kinetics by increasing the neutron lifetime. The authors consider the possibility of using the lead isotope, 208Pb, as a neutron reflector with specific properties in a lead-cooled fast reactor. To analyze the emerging effects in a reactor of this type, a point kinetics model was selected, which takes into account neutrons returning from the 208Pb reflector to the reactor core. Such specific properties of 208Pb as the high atomic weight and weak neutron absorption allow neutrons from the reactor core to penetrate deeply into the 208Pb reflector, slow down in it, and have a noticeable probability to return to the reactor core and affect the chain fission reaction. The neutrons coming back from the 208Pb reflector have a long ‘deadtime’, i.e., the sum of times when neutrons leave the reactor core, entering the 208Pb reflector, and then diffuse back into the reactor core. During the ‘dead-time’, these neutrons cannot affect the chain fission reaction. In terms of the delay time, the neutrons returning from the deep layers of the 208Pb reflector are close to the delayed neutrons. Moreover, the number of the neutrons coming back from the 208Pb reflector considerably exceeds the number of the delayed neutrons. As a result, the neutron lifetime formed by the prompt neutron lifetime and the ‘dead-time’ of the neutrons from the 208Pb reflector can be substantially increased. This will lead to a longer reactor acceleration period, which will mitigate the effects of prompt supercriticality. Thus, the use of 208Pb as a neutron reflector can significantly improve the fast reactor nuclear safety.
{"title":"On a significant slowing-down of the kinetics of fast transient processes in a fast reactor","authors":"G. Kulikov, A. Shmelev, V. Apse, E. Kulikov","doi":"10.3897/nucet.6.60379","DOIUrl":"https://doi.org/10.3897/nucet.6.60379","url":null,"abstract":"The kinetics of nuclear reactors is determined by the average neutron lifetime. When the inserted reactivity is more than the effective delayed neutron fraction, the reactor kinetics becomes very rapid. It is possible to slow down the fast reactor kinetics by increasing the neutron lifetime. The authors consider the possibility of using the lead isotope, 208Pb, as a neutron reflector with specific properties in a lead-cooled fast reactor. To analyze the emerging effects in a reactor of this type, a point kinetics model was selected, which takes into account neutrons returning from the 208Pb reflector to the reactor core. Such specific properties of 208Pb as the high atomic weight and weak neutron absorption allow neutrons from the reactor core to penetrate deeply into the 208Pb reflector, slow down in it, and have a noticeable probability to return to the reactor core and affect the chain fission reaction. The neutrons coming back from the 208Pb reflector have a long ‘deadtime’, i.e., the sum of times when neutrons leave the reactor core, entering the 208Pb reflector, and then diffuse back into the reactor core. During the ‘dead-time’, these neutrons cannot affect the chain fission reaction. In terms of the delay time, the neutrons returning from the deep layers of the 208Pb reflector are close to the delayed neutrons. Moreover, the number of the neutrons coming back from the 208Pb reflector considerably exceeds the number of the delayed neutrons. As a result, the neutron lifetime formed by the prompt neutron lifetime and the ‘dead-time’ of the neutrons from the 208Pb reflector can be substantially increased. This will lead to a longer reactor acceleration period, which will mitigate the effects of prompt supercriticality. Thus, the use of 208Pb as a neutron reflector can significantly improve the fast reactor nuclear safety.","PeriodicalId":100969,"journal":{"name":"Nuclear Energy and Technology","volume":"238 1","pages":"295-298"},"PeriodicalIF":0.0,"publicationDate":"2020-11-20","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"79620890","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 article presents a method for obtaining an analytical expression for the criterion of stability of a VVER-1000 (1200) reactor with respect to xenon oscillations of the local power in the core, containing an explicit dependence of the criterion ratio coefficients on the arbitrary axial neutron field distribution in steady states of the core. Based on the data of numerical experiments using a full-scale model of the Kalinin NPP power units, the authors present the results of checking the validity of this expression for the reactor stability criterion with respect to xenon oscillations for different NPPs with VVER-1000 (1200) reactors.
{"title":"Investigation of the impact of steady-state VVER-1000 (1200) core characteristics on the reactor stability with respect to xenon oscillations","authors":"R. Malkawi, Sergey B. Vygovsky, O. Batayneh","doi":"10.3897/nucet.6.60464","DOIUrl":"https://doi.org/10.3897/nucet.6.60464","url":null,"abstract":"The article presents a method for obtaining an analytical expression for the criterion of stability of a VVER-1000 (1200) reactor with respect to xenon oscillations of the local power in the core, containing an explicit dependence of the criterion ratio coefficients on the arbitrary axial neutron field distribution in steady states of the core. Based on the data of numerical experiments using a full-scale model of the Kalinin NPP power units, the authors present the results of checking the validity of this expression for the reactor stability criterion with respect to xenon oscillations for different NPPs with VVER-1000 (1200) reactors.","PeriodicalId":100969,"journal":{"name":"Nuclear Energy and Technology","volume":"61 1","pages":"289-294"},"PeriodicalIF":0.0,"publicationDate":"2020-11-20","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"83791983","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. Evdokimov, A. G. Khromov, Petr M. Kalinichev, V. Likhanskii, A. Kovalishin, M. N. Laletin
Fuel failures may occur during operation of nuclear power plants. One of the possible and most severe consequences of a fuel failure is that fuel may be washed out from the leaking fuel rod into the coolant. Reliable detection of fuel washout is important for handling of leaking fuel assemblies after irradiation is over. Detection of fuel washout is achievable in the framework of coolant activity evaluation during reactor operation. For this purpose, 134I activity is historically used in WWER power units. However, observed 134I activity may increase during operation even if leaking fuel in the core is absent, and fuel deposits are the only source of the fission products release. The paper describes a criterion which enables to reveal the cases when the increase in 134I activity results from the fuel washout from the leaking fuel rods during operation of the WWER-type reactor. Some examples of applications at nuclear power plants are discussed.
{"title":"Development of a criterion for assessment of fuel washout during operation of WWER power units","authors":"I. Evdokimov, A. G. Khromov, Petr M. Kalinichev, V. Likhanskii, A. Kovalishin, M. N. Laletin","doi":"10.3897/nucet.6.60559","DOIUrl":"https://doi.org/10.3897/nucet.6.60559","url":null,"abstract":"Fuel failures may occur during operation of nuclear power plants. One of the possible and most severe consequences of a fuel failure is that fuel may be washed out from the leaking fuel rod into the coolant. Reliable detection of fuel washout is important for handling of leaking fuel assemblies after irradiation is over. Detection of fuel washout is achievable in the framework of coolant activity evaluation during reactor operation. For this purpose, 134I activity is historically used in WWER power units. However, observed 134I activity may increase during operation even if leaking fuel in the core is absent, and fuel deposits are the only source of the fission products release. The paper describes a criterion which enables to reveal the cases when the increase in 134I activity results from the fuel washout from the leaking fuel rods during operation of the WWER-type reactor. Some examples of applications at nuclear power plants are discussed.","PeriodicalId":100969,"journal":{"name":"Nuclear Energy and Technology","volume":"38 1","pages":"307-312"},"PeriodicalIF":0.0,"publicationDate":"2020-11-20","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"83631398","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. I. Baranenko, O. M. Gulina, S. Mironov, N. L. Salnikov
The article presents a study carried out on carbon steel pipe components subjected to erosion-corrosion wear (ECW). Based on the repeated control data, the authors present the calculated ECW characteristics, i.e., the wall-thinning and rate. It is shown that such estimates contain great uncertainty due to corrosion products deposited on the pipeline inner surface and their migration during operation. In addition, with an increase in the operating time, for example, when the lifetime is extended, the difference between the forecast and the results of control becomes larger. This means that the error in the estimates of the residual lifetime also increases. The study is based on the data of wall thickness measurements of the feedwater pipeline (273×16 mm) and steam pipeline (465×16mm) of nuclear power plants with the VVER-440 reactors, for which a sufficient number of repeated measurements were performed over a large time interval. An analysis is made of the error in estimating the pipeline wall-thinning and ECW rate using Chexal-Horowitz Flow-Accelerated Corrosion (FAC) Model (EKI-02 and EKI-03 software tools). The estimate of the ECW rate according to the above forecast model differs from the estimate according to the current control data by no more than 12.5%, since the corrosion products deposited on the pipeline inner surface wall are leveled at a large time base. When calculating the wall-thinning, due to the obvious filtering of the control data, it is possible to achieve an acceptable accuracy of estimates, i.e., about 16% without upgrading the model.
{"title":"Repeated measurements and quality of estimates in the analysis of NPP pipeline erosion-corrosion wear","authors":"V. I. Baranenko, O. M. Gulina, S. Mironov, N. L. Salnikov","doi":"10.3897/nucet.6.60459","DOIUrl":"https://doi.org/10.3897/nucet.6.60459","url":null,"abstract":"The article presents a study carried out on carbon steel pipe components subjected to erosion-corrosion wear (ECW). Based on the repeated control data, the authors present the calculated ECW characteristics, i.e., the wall-thinning and rate. It is shown that such estimates contain great uncertainty due to corrosion products deposited on the pipeline inner surface and their migration during operation. In addition, with an increase in the operating time, for example, when the lifetime is extended, the difference between the forecast and the results of control becomes larger. This means that the error in the estimates of the residual lifetime also increases. The study is based on the data of wall thickness measurements of the feedwater pipeline (273×16 mm) and steam pipeline (465×16mm) of nuclear power plants with the VVER-440 reactors, for which a sufficient number of repeated measurements were performed over a large time interval. An analysis is made of the error in estimating the pipeline wall-thinning and ECW rate using Chexal-Horowitz Flow-Accelerated Corrosion (FAC) Model (EKI-02 and EKI-03 software tools). The estimate of the ECW rate according to the above forecast model differs from the estimate according to the current control data by no more than 12.5%, since the corrosion products deposited on the pipeline inner surface wall are leveled at a large time base. When calculating the wall-thinning, due to the obvious filtering of the control data, it is possible to achieve an acceptable accuracy of estimates, i.e., about 16% without upgrading the model.","PeriodicalId":100969,"journal":{"name":"Nuclear Energy and Technology","volume":"14 1","pages":"281-287"},"PeriodicalIF":0.0,"publicationDate":"2020-11-20","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"86142942","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. Andrianova, Evgeniya S. Teplukhina, Gennady M. Zherdev, Zhanna V. Borovskaya, A. P. Zhirnov
The paper presents the results of the efforts concerned with expanding the verification database and estimating the calculation uncertainty of the power density in the steel reflector of lead cooled fast reactor designs based on experiments performed in different years at the BFS critical assemblies by analyzing and revising earlier calculation and experimental studies on the transmission of neutrons through the steel reflector layers. The discussion includes experiments at the BFS-66 critical assembly to model neutron and photon fluxes in the reactor core shielding compositions, as well as experiments at the BFS-64 and BFS-80-2 critical assemblies to model the transmission of neutrons and gamma quanta through the reflector layers of various materials. The information provided in earlier materials with the descriptions of the above experiments has been analyzed and expanded through respective data required to prepare precision calculation models for Monte-Carlo neutronic codes. Precision neutronic models have been developed based on actualized and updated data with a detailed description of the BFS heterogeneous structure and experimental devices, and test calculations have been carried out to confirm their efficiency. The calculations of key neutronic characteristics measured at the BFS-66, -64 and -80-2 assemblies were performed using codes based on the Monte Carlo method (MCU-BR, MCNP, MMK-RF, MMK-ROKOKO) with BNAB-RF and MDBBR50 neutron data and the ROSFOND evaluated neutron data library. The developed precision calculation neutronic models of the experiments discussed can be used to justify lead cooled fast reactor designs, to verify neutronic codes and neutron data, and to evaluate the associated uncertainties.
{"title":"Precision neutronic calculations of experiments on the neutron transmission through the reflector layers at the BFS critical facilities for expanding the verification database to justify lead cooled fast reactor designs","authors":"O. Andrianova, Evgeniya S. Teplukhina, Gennady M. Zherdev, Zhanna V. Borovskaya, A. P. Zhirnov","doi":"10.3897/nucet.6.60303","DOIUrl":"https://doi.org/10.3897/nucet.6.60303","url":null,"abstract":"The paper presents the results of the efforts concerned with expanding the verification database and estimating the calculation uncertainty of the power density in the steel reflector of lead cooled fast reactor designs based on experiments performed in different years at the BFS critical assemblies by analyzing and revising earlier calculation and experimental studies on the transmission of neutrons through the steel reflector layers. The discussion includes experiments at the BFS-66 critical assembly to model neutron and photon fluxes in the reactor core shielding compositions, as well as experiments at the BFS-64 and BFS-80-2 critical assemblies to model the transmission of neutrons and gamma quanta through the reflector layers of various materials. The information provided in earlier materials with the descriptions of the above experiments has been analyzed and expanded through respective data required to prepare precision calculation models for Monte-Carlo neutronic codes. Precision neutronic models have been developed based on actualized and updated data with a detailed description of the BFS heterogeneous structure and experimental devices, and test calculations have been carried out to confirm their efficiency. The calculations of key neutronic characteristics measured at the BFS-66, -64 and -80-2 assemblies were performed using codes based on the Monte Carlo method (MCU-BR, MCNP, MMK-RF, MMK-ROKOKO) with BNAB-RF and MDBBR50 neutron data and the ROSFOND evaluated neutron data library. The developed precision calculation neutronic models of the experiments discussed can be used to justify lead cooled fast reactor designs, to verify neutronic codes and neutron data, and to evaluate the associated uncertainties.","PeriodicalId":100969,"journal":{"name":"Nuclear Energy and Technology","volume":"134 1","pages":"269-274"},"PeriodicalIF":0.0,"publicationDate":"2020-11-20","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"78174413","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 article presents the first data on EK-164ID steel swelling after operational irradiation in a fast nuclear reactor in the temperature range of 370–630 °C and maximum damaging doses of 66–77 dpa. The dose accumulation rate along the cladding tubes made of this material was 1×10-8–1.6×10-6 dpa/s. The swelling was determined by the hydrostatic weighing method with an error of no more than 0.5%. The results obtained were analyzed depending on the irradiation parameters and in comparison with the 16Cr-15Ni grade material. The objectives of the study were to estimate the characteristic values of the maximum swelling temperature and dose as well as to calculate the average material swelling rate at the working temperature of irradiation, the incubation period for the onset of swelling, and the stationary swelling rate. It was found that the tube samples, characterized with austenite grain sizes of 9–12 μm before irradiation, have an average swelling rate of 0.035–0.05 %/dpa after reaching the maximum damaging doses of 66–77 dpa (at a rate of (1–1.5)×10-6 dpa/s) and not more than 0.035%/dpa at doses less than 20 dpa (at a rate of 5×10-7 dpa/s). The characteristic maximum swelling temperature of the studied material is in the range of 430–500 °C. The characteristic maximum swelling dose is in the range of 61–72.5 dpa or 70–80% of the maximum accumulated dose. The incubation stationary swelling period for the material is 30 dpa. The stationary swelling rate is 0.1% /dpa. The radiation resistance characteristics of the studied material have an advantage over those for 16Cr-15Ni grade cladding materials under similar irradiation conditions and a similar structural state, which inherits grain sizes of 9–14 μm during the tube processing.
{"title":"16Cr-19Ni steel swelling at dose rates from 1×10-8 to 1.6×10-6 dpa/s","authors":"E. Kinev","doi":"10.3897/nucet.6.60371","DOIUrl":"https://doi.org/10.3897/nucet.6.60371","url":null,"abstract":"The article presents the first data on EK-164ID steel swelling after operational irradiation in a fast nuclear reactor in the temperature range of 370–630 °C and maximum damaging doses of 66–77 dpa. The dose accumulation rate along the cladding tubes made of this material was 1×10-8–1.6×10-6 dpa/s. The swelling was determined by the hydrostatic weighing method with an error of no more than 0.5%. The results obtained were analyzed depending on the irradiation parameters and in comparison with the 16Cr-15Ni grade material. The objectives of the study were to estimate the characteristic values of the maximum swelling temperature and dose as well as to calculate the average material swelling rate at the working temperature of irradiation, the incubation period for the onset of swelling, and the stationary swelling rate. It was found that the tube samples, characterized with austenite grain sizes of 9–12 μm before irradiation, have an average swelling rate of 0.035–0.05 %/dpa after reaching the maximum damaging doses of 66–77 dpa (at a rate of (1–1.5)×10-6 dpa/s) and not more than 0.035%/dpa at doses less than 20 dpa (at a rate of 5×10-7 dpa/s). The characteristic maximum swelling temperature of the studied material is in the range of 430–500 °C. The characteristic maximum swelling dose is in the range of 61–72.5 dpa or 70–80% of the maximum accumulated dose. The incubation stationary swelling period for the material is 30 dpa. The stationary swelling rate is 0.1% /dpa. The radiation resistance characteristics of the studied material have an advantage over those for 16Cr-15Ni grade cladding materials under similar irradiation conditions and a similar structural state, which inherits grain sizes of 9–14 μm during the tube processing.","PeriodicalId":100969,"journal":{"name":"Nuclear Energy and Technology","volume":"203 1","pages":"249-252"},"PeriodicalIF":0.0,"publicationDate":"2020-11-18","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"75082026","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
A. Avdeenkov, O. I. Achakovsky, V. Ketlerov, V. Kumaev, A. I. Orlov
The article presents the results of corrosion processes, kinetics and changes in the oxide layer modeling using MASKA-LM software complex. The complex is intended for a numerical simulation of three-dimensional non-stationary processes of mass transfer and interaction of impurity components in a heavy liquid metal coolant (HLMC: lead, lead-bismuth). The software complex is based on the numerical solution of coupled three-dimensional equations of hydrodynamics, heat transfer, formation and convective-diffusive transport of chemically interacting components of impurities. Examples of calculations of mass transfer processes and interaction of impurity components in HLMC, formation of protective oxide films on the surfaces of steels are given to justify the coolant technology.
{"title":"Basic models and approximation for the engineering description of the kinetics of the oxide layer of steel in a flow of heavy liquid metal coolant under various oxygen conditions","authors":"A. Avdeenkov, O. I. Achakovsky, V. Ketlerov, V. Kumaev, A. I. Orlov","doi":"10.3897/nucet.6.59068","DOIUrl":"https://doi.org/10.3897/nucet.6.59068","url":null,"abstract":"The article presents the results of corrosion processes, kinetics and changes in the oxide layer modeling using MASKA-LM software complex. The complex is intended for a numerical simulation of three-dimensional non-stationary processes of mass transfer and interaction of impurity components in a heavy liquid metal coolant (HLMC: lead, lead-bismuth). The software complex is based on the numerical solution of coupled three-dimensional equations of hydrodynamics, heat transfer, formation and convective-diffusive transport of chemically interacting components of impurities. Examples of calculations of mass transfer processes and interaction of impurity components in HLMC, formation of protective oxide films on the surfaces of steels are given to justify the coolant technology.","PeriodicalId":100969,"journal":{"name":"Nuclear Energy and Technology","volume":"7 1","pages":"215-234"},"PeriodicalIF":0.0,"publicationDate":"2020-11-16","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"88695348","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 article describes the status of development of codes of new generation for the “PRORYV” Project by the end of 2019: twenty-five commercial-grade software products to justify design solutions and safety of power units with fast neutron reactors and liquid metal coolant (sodium and lead) in a closed nuclear fuel cycle. The developed system of codes is multi-physical and multi-scale that allows performing both calculations of the whole installations and high precision calculations of their individual elements. The developed codes offer unique features. Twelve developed codes have already been certified by Rostechnadzor, and six more have been submitted for certification. In addition to creating the software products, a large-scale work is being carried out to conduct experimental studies for code validation that meet modern requirements imposed by the codes: unique measurement techniques have been created; experimental data on flow characteristics of heavy liquid metal coolant (HLMC) in a fuel assembly simulator have been obtained, as well as of “gas-HLMC” interphase interaction after inert gas injection in HLMC and characteristics of heat exchange between the inert gas and HLMC. The results are already used for validation of system and CFD codes used in the “PRORYV” Project.
{"title":"Codes of new generation for safety justification of power units with a closed nuclear fuel cycle developed for the “PRORYV” project","authors":"L. Bolshov, V. Strizhov, N. Mosunova","doi":"10.3897/nucet.6.54710","DOIUrl":"https://doi.org/10.3897/nucet.6.54710","url":null,"abstract":"The article describes the status of development of codes of new generation for the “PRORYV” Project by the end of 2019: twenty-five commercial-grade software products to justify design solutions and safety of power units with fast neutron reactors and liquid metal coolant (sodium and lead) in a closed nuclear fuel cycle. The developed system of codes is multi-physical and multi-scale that allows performing both calculations of the whole installations and high precision calculations of their individual elements. The developed codes offer unique features. Twelve developed codes have already been certified by Rostechnadzor, and six more have been submitted for certification. In addition to creating the software products, a large-scale work is being carried out to conduct experimental studies for code validation that meet modern requirements imposed by the codes: unique measurement techniques have been created; experimental data on flow characteristics of heavy liquid metal coolant (HLMC) in a fuel assembly simulator have been obtained, as well as of “gas-HLMC” interphase interaction after inert gas injection in HLMC and characteristics of heat exchange between the inert gas and HLMC. The results are already used for validation of system and CFD codes used in the “PRORYV” Project.","PeriodicalId":100969,"journal":{"name":"Nuclear Energy and Technology","volume":"21 1","pages":"203-214"},"PeriodicalIF":0.0,"publicationDate":"2020-11-06","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"91351140","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}