Pub Date : 2021-09-26DOI: 10.55176/2414-1038-2021-3-191-205
M. Koshelev, V. Ulyanov, S. Kharchuk
The results of computational and experimental studies of the properties of lead and lead-bismuth coolants and the mechanism of physicochemical processes occurring in their circulation circuits explain the reasons for the formation of slag deposits based on lead oxide. First of all, this is the uncontrolled interaction of lead and lead-bismuth coolants with gaseous oxygen, which, along with nitrogen, is one of the main components of air. Slag deposits are based on lead oxides, but they also contain a significant amount of unoxidized lead (and bismuth in the case of using a lead-bismuth coolant) bound by a lead oxide framework into a single structure, as well as a small amount of components of construction materials (iron , chrome, nickel). The formation of slags in the circulation loop with a lead-containing coolant has a negative effect on its operation: the flow cross-sections are narrowed; heat transfer surfaces are blocked; the normal operation of pipeline fittings is disrupted. To minimize slag accumulation, it is necessary to limit the contact of the coolant with air oxygen, avoid draining the coolant from the circulation circuit, control whether the circulation circuit is in a depressurized state both during research and during non-working periods, when heating and filling the circuits with a coolant, create and maintain a vacuum. The best way to prevent slag accumulation is periodic hydrogen cleaning of the circuit with gas mixtures "hydrogen - water vapor - inert gas". The greatest cleaning efficiency is achieved when gas mixtures are introduced directly into the coolant flow using gas injection devices. Mechanical devices with moving parts, ejectors, nozzle nozzles can act as gas injection devices.
{"title":"INVESTIGATION OF THE SLAGS ACCUMULATION MECHANISM IN LEAD-CONTAINING COOLANTS LOOPS AND ITS PREVENTION BY HIGH-TEMPERATURE TREATMENT WITH HYDROGEN AND WATER STEAM MIXTURES","authors":"M. Koshelev, V. Ulyanov, S. Kharchuk","doi":"10.55176/2414-1038-2021-3-191-205","DOIUrl":"https://doi.org/10.55176/2414-1038-2021-3-191-205","url":null,"abstract":"The results of computational and experimental studies of the properties of lead and lead-bismuth coolants and the mechanism of physicochemical processes occurring in their circulation circuits explain the reasons for the formation of slag deposits based on lead oxide. First of all, this is the uncontrolled interaction of lead and lead-bismuth coolants with gaseous oxygen, which, along with nitrogen, is one of the main components of air. Slag deposits are based on lead oxides, but they also contain a significant amount of unoxidized lead (and bismuth in the case of using a lead-bismuth coolant) bound by a lead oxide framework into a single structure, as well as a small amount of components of construction materials (iron , chrome, nickel). The formation of slags in the circulation loop with a lead-containing coolant has a negative effect on its operation: the flow cross-sections are narrowed; heat transfer surfaces are blocked; the normal operation of pipeline fittings is disrupted. To minimize slag accumulation, it is necessary to limit the contact of the coolant with air oxygen, avoid draining the coolant from the circulation circuit, control whether the circulation circuit is in a depressurized state both during research and during non-working periods, when heating and filling the circuits with a coolant, create and maintain a vacuum. The best way to prevent slag accumulation is periodic hydrogen cleaning of the circuit with gas mixtures \"hydrogen - water vapor - inert gas\". The greatest cleaning efficiency is achieved when gas mixtures are introduced directly into the coolant flow using gas injection devices. Mechanical devices with moving parts, ejectors, nozzle nozzles can act as gas injection devices.","PeriodicalId":20426,"journal":{"name":"PROBLEMS OF ATOMIC SCIENCE AND TECHNOLOGY. SERIES: NUCLEAR AND REACTOR CONSTANTS","volume":null,"pages":null},"PeriodicalIF":0.0,"publicationDate":"2021-09-26","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"76307664","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2021-09-26DOI: 10.55176/2414-1038-2021-3-68-76
I. Kurina, M. Frolova, E. Chesnokov, V. Ryaby, A. Dvoryashin, M. Kanunnikov
Density, phase composition, microstructure and thermal conductivity of the U-10 wt. % Zr alloy manufactured by induction melting with subsequent casting into quartz molds and turning to size have been investigated at JSC “SSC RF - IPPE”. For comparison, the density and thermal conductivity of the U-10 wt. % Zr alloy produced by melting followed by extrusion and turning to size were investigated. To determine the density, a hydrostatic weighing method was used. The average density values of the cast and extruded alloy were respectively 98.8 and 97.5 % of the theoretical density, which was calculated according to the rule of mixtures. The results of studying the microstructure using a scanning electron microscope are presented. It is shown that the cast alloy U-10 % Zr is a metal matrix in which zirconium-enriched particles of arbitrary shape are distributed. In a metal matrix, the bulk of the volume is occupied by the α-U, and there are also precipitates of the δ-phase in the form of thin plates. A lower value of the microhardness of the alloy is noted in comparison with the data published in the known literature. The results of measuring the thermal conductivity at temperatures from 100 to 750 °C for the U-10 wt. % Zr alloy obtained by casting and extrusion are presented. The stationary axial heat flux method (or method of plate) was used to measure the thermal conductivity. Alloy samples made in different ways have almost the same thermal conductivity at 200 °C. With an increase in temperature, the discrepancy in thermal conductivity between the samples of the cast and extruded alloy gradually increases, and the thermal conductivity of the extruded alloy turns out to be lower, which is especially noticeable in the temperature range of 600-750 °C. The data obtained are compared with the results of published works. The measured values of the thermal conductivity of the cast alloy U-10 wt. % Zr up to a temperature of 750 °C do not disagree with the literature data. It was found that at a higher temperature, the alloy softens, which, in turn, leads to deformation of the test specimen and an increase in the measurement error when using the axial heat flux method.
{"title":"STUDY OF U-10 % Zr ALLOY PROPERTIES","authors":"I. Kurina, M. Frolova, E. Chesnokov, V. Ryaby, A. Dvoryashin, M. Kanunnikov","doi":"10.55176/2414-1038-2021-3-68-76","DOIUrl":"https://doi.org/10.55176/2414-1038-2021-3-68-76","url":null,"abstract":"Density, phase composition, microstructure and thermal conductivity of the U-10 wt. % Zr alloy manufactured by induction melting with subsequent casting into quartz molds and turning to size have been investigated at JSC “SSC RF - IPPE”. For comparison, the density and thermal conductivity of the U-10 wt. % Zr alloy produced by melting followed by extrusion and turning to size were investigated. To determine the density, a hydrostatic weighing method was used. The average density values of the cast and extruded alloy were respectively 98.8 and 97.5 % of the theoretical density, which was calculated according to the rule of mixtures. The results of studying the microstructure using a scanning electron microscope are presented. It is shown that the cast alloy U-10 % Zr is a metal matrix in which zirconium-enriched particles of arbitrary shape are distributed. In a metal matrix, the bulk of the volume is occupied by the α-U, and there are also precipitates of the δ-phase in the form of thin plates. A lower value of the microhardness of the alloy is noted in comparison with the data published in the known literature. The results of measuring the thermal conductivity at temperatures from 100 to 750 °C for the U-10 wt. % Zr alloy obtained by casting and extrusion are presented. The stationary axial heat flux method (or method of plate) was used to measure the thermal conductivity. Alloy samples made in different ways have almost the same thermal conductivity at 200 °C. With an increase in temperature, the discrepancy in thermal conductivity between the samples of the cast and extruded alloy gradually increases, and the thermal conductivity of the extruded alloy turns out to be lower, which is especially noticeable in the temperature range of 600-750 °C. The data obtained are compared with the results of published works. The measured values of the thermal conductivity of the cast alloy U-10 wt. % Zr up to a temperature of 750 °C do not disagree with the literature data. It was found that at a higher temperature, the alloy softens, which, in turn, leads to deformation of the test specimen and an increase in the measurement error when using the axial heat flux method.","PeriodicalId":20426,"journal":{"name":"PROBLEMS OF ATOMIC SCIENCE AND TECHNOLOGY. SERIES: NUCLEAR AND REACTOR CONSTANTS","volume":null,"pages":null},"PeriodicalIF":0.0,"publicationDate":"2021-09-26","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"83989016","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2021-09-26DOI: 10.55176/2414-1038-2021-3-167-173
A. Sakhipgareev, A. Shlepkin, A. Morozov
The results of experimental studies of the solubility of boric acid in steam in the concentration range of 16-440 g/kg H2O and the pressure range of 0.1-0.3 MPa is considered in the article. An analysis of the literature data on the solubility of boric acid in steam showed that the results do not cover the entire range of parameters (temperature, pressure, concentration of boric acid) typical of a possible emergency situation at NNPs with the WWER. In paper the test facility is described. The obtained experimental results confirm the data available in the literature that the change in the concentration of boric acid in steam is described by a linear law. The range of application of the dependence of the distribution of H3BO3 between the liquid and steam phases of the boiling coolant has been expanded to a concentration of 440 g/kg H2O, which is close to the solubility limit of boric acid in water at a pressure of 0.3 MPa. The obtained experimental data can be used in the calculating of the emergency at nuclear power plants with WWER, taking into account the operating of passive safety systems.
{"title":"EXPERIMENTAL STUDY OF THE FEATURES OF PROCESS OF SOLUBILITY OF BORIC ACID IN STEAM DURING OF BOILING OF THE COOLANT OF THE WWER IN CASE OF ACCIDENT","authors":"A. Sakhipgareev, A. Shlepkin, A. Morozov","doi":"10.55176/2414-1038-2021-3-167-173","DOIUrl":"https://doi.org/10.55176/2414-1038-2021-3-167-173","url":null,"abstract":"The results of experimental studies of the solubility of boric acid in steam in the concentration range of 16-440 g/kg H2O and the pressure range of 0.1-0.3 MPa is considered in the article. An analysis of the literature data on the solubility of boric acid in steam showed that the results do not cover the entire range of parameters (temperature, pressure, concentration of boric acid) typical of a possible emergency situation at NNPs with the WWER. In paper the test facility is described. The obtained experimental results confirm the data available in the literature that the change in the concentration of boric acid in steam is described by a linear law. The range of application of the dependence of the distribution of H3BO3 between the liquid and steam phases of the boiling coolant has been expanded to a concentration of 440 g/kg H2O, which is close to the solubility limit of boric acid in water at a pressure of 0.3 MPa. The obtained experimental data can be used in the calculating of the emergency at nuclear power plants with WWER, taking into account the operating of passive safety systems.","PeriodicalId":20426,"journal":{"name":"PROBLEMS OF ATOMIC SCIENCE AND TECHNOLOGY. SERIES: NUCLEAR AND REACTOR CONSTANTS","volume":null,"pages":null},"PeriodicalIF":0.0,"publicationDate":"2021-09-26","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"91499326","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2021-09-26DOI: 10.55176/2414-1038-2021-3-136-142
V. Perevezentsev
Experimental studies of hydrodynamic induced of fuel rods vibrations WWER-440 tweers using a full-scale mock-up with lead imitators of fuel tablets have been carried out. It is shown that the speed of the turbulent flow in the fuel rods beam cannot be the only hydrodynamic characteristic that determines the vibrational characteristics of the fuel rods beams. The presence of indignant flow of various elements of the tract of the coolant to the fuel rods beams inlet significantly affects the intensity of their vibrations. An energy model of the balance of pulsating energy of turbulent flow and expendable mechanical energy to the fuel rods beams has been developed. On the basis of the proposed model, a functional connection of the intensity of the vibrations of the fuel rods beams with the levels of pressure pulsations in the turbulent flow of the coolant at the entrance to the fuel rods beams is established. The modelling of Fuel Assembly (FA) as a mechanical vibrational system has received considerable attention. At the same time, the end-element methods, as the most universal for the sampling of spatial continual systems, have become widespread. However, the analysis of hydrodynamically excited vibrations is impossible without describing the processes of interaction of the flow with the streamlined surfaces of the mechanical vibration system. In most cases, such information can be obtained only through experimental research.
{"title":"FEATURES OF THE FORMATION OF HYDRODYNAMIC LOADS ON THE FUEL RODS DEPENDING ON THE STRUCTURE OF THE COOLANT FLOW AT THE ENTRANCE TO THE FUEL ASSEMBLY WWER","authors":"V. Perevezentsev","doi":"10.55176/2414-1038-2021-3-136-142","DOIUrl":"https://doi.org/10.55176/2414-1038-2021-3-136-142","url":null,"abstract":"Experimental studies of hydrodynamic induced of fuel rods vibrations WWER-440 tweers using a full-scale mock-up with lead imitators of fuel tablets have been carried out. It is shown that the speed of the turbulent flow in the fuel rods beam cannot be the only hydrodynamic characteristic that determines the vibrational characteristics of the fuel rods beams. The presence of indignant flow of various elements of the tract of the coolant to the fuel rods beams inlet significantly affects the intensity of their vibrations. An energy model of the balance of pulsating energy of turbulent flow and expendable mechanical energy to the fuel rods beams has been developed. On the basis of the proposed model, a functional connection of the intensity of the vibrations of the fuel rods beams with the levels of pressure pulsations in the turbulent flow of the coolant at the entrance to the fuel rods beams is established. The modelling of Fuel Assembly (FA) as a mechanical vibrational system has received considerable attention. At the same time, the end-element methods, as the most universal for the sampling of spatial continual systems, have become widespread. However, the analysis of hydrodynamically excited vibrations is impossible without describing the processes of interaction of the flow with the streamlined surfaces of the mechanical vibration system. In most cases, such information can be obtained only through experimental research.","PeriodicalId":20426,"journal":{"name":"PROBLEMS OF ATOMIC SCIENCE AND TECHNOLOGY. SERIES: NUCLEAR AND REACTOR CONSTANTS","volume":null,"pages":null},"PeriodicalIF":0.0,"publicationDate":"2021-09-26","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"83035873","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2021-09-26DOI: 10.55176/2414-1038-2021-3-106-122
O. Vilenskii, S. Dushev, D. Lapshin, E. Novinskii, A. Tatarskii
The aim of the paper is to substantiate the developed shipping package integrity on the basis of mathematical analysis of postulated scenarios for the heaviest dynamic effects using verified behavior models for applied structural materials and modern certified finite element software package. The shipping package is permitted to transfer to the limiting condition in the result of impact, when its further operation is not acceptable, and recovery of its functional condition is not expedient, excluding the possibility of falling out of the nuclear fuel (NF) or distortion of the relative position of the NF in the shipping package. The paper presents main analysis results of the process of an aircraft turbine collision with a shipping package. The calculation analysis was performed using the dynamic calculation module LS-DYNA of the certified software package ANSYS. The LS-DYNA module is meant for computational analysis of high linear dynamic processes under explicit scheme of integration of the dynamics equations. The completed full-scale mathematical 3D modeling permitted to carry out sufficiently deep and detailed analysis of dynamic processes, reducing the design duration, and it permitted to reduce the self-cost of the developed design. The obtained results of mathematical analysis of shipping package behavior during interaction with aircraft turbine enabled to form the approach to substantiation of equipment safety in incidents. The detailed design study performed at the design stage formed basis for the final appearance of the developed structure.
{"title":"MATHEMATICAL ANALYSIS OF AIRCRAFT TURBINE COLLISION WITH SHIPPING PACKAGE","authors":"O. Vilenskii, S. Dushev, D. Lapshin, E. Novinskii, A. Tatarskii","doi":"10.55176/2414-1038-2021-3-106-122","DOIUrl":"https://doi.org/10.55176/2414-1038-2021-3-106-122","url":null,"abstract":"The aim of the paper is to substantiate the developed shipping package integrity on the basis of mathematical analysis of postulated scenarios for the heaviest dynamic effects using verified behavior models for applied structural materials and modern certified finite element software package. The shipping package is permitted to transfer to the limiting condition in the result of impact, when its further operation is not acceptable, and recovery of its functional condition is not expedient, excluding the possibility of falling out of the nuclear fuel (NF) or distortion of the relative position of the NF in the shipping package. The paper presents main analysis results of the process of an aircraft turbine collision with a shipping package. The calculation analysis was performed using the dynamic calculation module LS-DYNA of the certified software package ANSYS. The LS-DYNA module is meant for computational analysis of high linear dynamic processes under explicit scheme of integration of the dynamics equations. The completed full-scale mathematical 3D modeling permitted to carry out sufficiently deep and detailed analysis of dynamic processes, reducing the design duration, and it permitted to reduce the self-cost of the developed design. The obtained results of mathematical analysis of shipping package behavior during interaction with aircraft turbine enabled to form the approach to substantiation of equipment safety in incidents. The detailed design study performed at the design stage formed basis for the final appearance of the developed structure.","PeriodicalId":20426,"journal":{"name":"PROBLEMS OF ATOMIC SCIENCE AND TECHNOLOGY. SERIES: NUCLEAR AND REACTOR CONSTANTS","volume":null,"pages":null},"PeriodicalIF":0.0,"publicationDate":"2021-09-26","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"76881084","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2021-09-26DOI: 10.55176/2414-1038-2021-3-158-166
N. Loginov, A. Mikheyev, T. Vereshchagina
The article discusses the designs of two nuclear reactors cooled by heat pipes with liquid metal coolants. Both designs were developed at IPPE JSC. The first design is a research reactor with a capacity of 1.2 MW. Evaporating sodium is used as a coolant in the core. The evaporating eutectic sodium-potassium alloy is used as a coolant in the secondary loop. The third loop contains gas as working fluid of the Stirling or Brighton cycle. The report presents the results of thermo-hydraulic experiments that confirmed the main design parameters. The second design called RIFMA is an innovative super small NPP with direct conversion of thermal energy into electricity and is supposed, in particular, to be used in the Arctic region. Thermal power is 100 kW, efficiency is not less than 10 %. Molybdenum heat pipes filled with lithium are used to transfer heat from the core. To convert energy, thermophotovoltaic converters are proposed. They are cooled by low-temperature heat pipes that remove residual heat and transfer it to air radiators. A nuclear power plant concept and three versions of the core are presented.
{"title":"ON DEVELOPMENT OF HEAT PIPES FOR NUCLEAR POWER PLANTS","authors":"N. Loginov, A. Mikheyev, T. Vereshchagina","doi":"10.55176/2414-1038-2021-3-158-166","DOIUrl":"https://doi.org/10.55176/2414-1038-2021-3-158-166","url":null,"abstract":"The article discusses the designs of two nuclear reactors cooled by heat pipes with liquid metal coolants. Both designs were developed at IPPE JSC. The first design is a research reactor with a capacity of 1.2 MW. Evaporating sodium is used as a coolant in the core. The evaporating eutectic sodium-potassium alloy is used as a coolant in the secondary loop. The third loop contains gas as working fluid of the Stirling or Brighton cycle. The report presents the results of thermo-hydraulic experiments that confirmed the main design parameters. The second design called RIFMA is an innovative super small NPP with direct conversion of thermal energy into electricity and is supposed, in particular, to be used in the Arctic region. Thermal power is 100 kW, efficiency is not less than 10 %. Molybdenum heat pipes filled with lithium are used to transfer heat from the core. To convert energy, thermophotovoltaic converters are proposed. They are cooled by low-temperature heat pipes that remove residual heat and transfer it to air radiators. A nuclear power plant concept and three versions of the core are presented.","PeriodicalId":20426,"journal":{"name":"PROBLEMS OF ATOMIC SCIENCE AND TECHNOLOGY. SERIES: NUCLEAR AND REACTOR CONSTANTS","volume":null,"pages":null},"PeriodicalIF":0.0,"publicationDate":"2021-09-26","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"81765282","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2021-09-26DOI: 10.55176/2414-1038-2021-3-77-87
M. Kolesnik, T. Aliev, V. Likhanskii
Computation study of the average zirconium hydride length on the cooling rate was performed using the precipitate nucleation and growth model. The cooling rate was varied in the range equal to six orders between typical values for the spent nuclear fuel dry storage conditions to values typical for laboratory tests modeling the dry storage. The calculations showed that as the cooling rate decreases, the hydrides concentration decreases, and their average length increases linearly on a double logarithmic scale. These dependencies have no limit if hydrides were abscended in the sample before the cooling began. If there were hydrides in the sample before the start of cooling, then they will grow and new hydrides will not nucleate in the limit of low cooling rates. For spent nuclear fuel dry storage, these results mean that if hydrides remain in the fuel claddings at the initial storage period, then hydrides morphology and hydrogen embrittlement at the end of the storage period are similar values gained under laboratory conditions with sufficiently slow cooling. If hydrides in fuel claddings are completely dissolved at the beginning of dry storage, then their length will be significantly greater than in laboratory tests at the end of the storage. Therefore, if the threshold values for the circumferential stresses are exceeded in fuel claddings, the hydrogen embrittlement can be expected to be higher than after faster cooling in typical laboratory studies. In this case, the hydrogen embrittlement assessment should be performed in a conservative approach assuming that radial hydrides have an average length equal to the thickness of the fuel cladding.
{"title":"COMPUTATIONAL STUDY OF ZIRCONIUM HYDRIDES MORPHOLOGY AT WIDELY VARIED COOLING RATES","authors":"M. Kolesnik, T. Aliev, V. Likhanskii","doi":"10.55176/2414-1038-2021-3-77-87","DOIUrl":"https://doi.org/10.55176/2414-1038-2021-3-77-87","url":null,"abstract":"Computation study of the average zirconium hydride length on the cooling rate was performed using the precipitate nucleation and growth model. The cooling rate was varied in the range equal to six orders between typical values for the spent nuclear fuel dry storage conditions to values typical for laboratory tests modeling the dry storage. The calculations showed that as the cooling rate decreases, the hydrides concentration decreases, and their average length increases linearly on a double logarithmic scale. These dependencies have no limit if hydrides were abscended in the sample before the cooling began. If there were hydrides in the sample before the start of cooling, then they will grow and new hydrides will not nucleate in the limit of low cooling rates. For spent nuclear fuel dry storage, these results mean that if hydrides remain in the fuel claddings at the initial storage period, then hydrides morphology and hydrogen embrittlement at the end of the storage period are similar values gained under laboratory conditions with sufficiently slow cooling. If hydrides in fuel claddings are completely dissolved at the beginning of dry storage, then their length will be significantly greater than in laboratory tests at the end of the storage. Therefore, if the threshold values for the circumferential stresses are exceeded in fuel claddings, the hydrogen embrittlement can be expected to be higher than after faster cooling in typical laboratory studies. In this case, the hydrogen embrittlement assessment should be performed in a conservative approach assuming that radial hydrides have an average length equal to the thickness of the fuel cladding.","PeriodicalId":20426,"journal":{"name":"PROBLEMS OF ATOMIC SCIENCE AND TECHNOLOGY. SERIES: NUCLEAR AND REACTOR CONSTANTS","volume":null,"pages":null},"PeriodicalIF":0.0,"publicationDate":"2021-09-26","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"78853042","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2021-09-26DOI: 10.55176/2414-1038-2021-3-123-135
B. Alchagirov, A. Khibiev
In the 1950s liquid lead and the lead-bismuth eutectic alloy (Pb45Bi55) were considered candidates for their use as coolants for nuclear power systems in the USSR and the USA. At the same time, the alloy (Pb45Bi55), first proposed by A.I. Leypunsky, was chosen as a coolant for the nuclear submarine “Alpha”, which was ahead of its time in terms of its tactical and technical data. However, in 1968, one of them suffered a severe radiation accident caused by the melting of fuel elements in the core of an onboard nuclear reactor due to the precipitation of oxides from the coolant and their accumulation, which blocked the pipeline cross-section and sharply worsened the cooling of the reactor, which led to its failure. In fact, the primary cause of the accident was a lack of knowledge about the physicochemical and technological properties of the lead-bismuth coolant. Thus, the main disadvantage of the Pb45Bi55 coolant is its corrosiveness to structural materials used in the nuclear power plant. But it has been found that corrosion by liquid lead alloys can be reduced by adjusting the oxygen level in the coolant. For example, the corrosion rate of martensitic steel at 770 K in a Pb45Bi55 flowing coolant without oxygen is about 1 mm per year, but it can be reduced to 0.01 mm per year, i.e. 100 times, if oxygen is dissolved in Pb45Bi55 coolant and its mass concentration is maintained at the level of 0.01 ppm. The observed effect is explained by the protection provided by the oxide layer formed on the steel surface of the pipeline. Thus, for a deeper understanding of the phenomena occurring at the boundaries of the liquid metal “coolants - gases” section, it remains relevant to study the processes of formation and destruction of the protective oxide layer and its behavior in coolant fluids, especially from the point of view of long-term operation of nuclear power plants. In this regard, data on the surface tension of the “coolant - gas” interphase boundaries are of great scientific and practical importance. In this connection, the present work sets the task of experimentally studying the influence of atmospheric air on the surface tension of Pb45Bi55 eutectic melt. Measurements of the surface tension of the eutectic melt Pb44,6Bi55,4 prepared by the authors were carried out in a non-stop mode sequentially, under static vacuum and atmospheric air, on the same surface. About three hundred experimental points obtained in this work made it possible to describe the dynamics of the surface tension changing process depending on the time of exposure of the coolant surface in vacuum and atmospheric air. It is shown that in comparison with the results obtained by the authors by the large lying drop method in a static vacuum, in the first 10 minutes from the beginning of the exposure of the eutectic melt in atmospheric air at a pressure of about 300 mmHg), the surface tension of the eutectic Pb44.6Bi55.4 decreases by 55 mN/m and more, which is an order of magnitude hi
在20世纪50年代，液态铅和铅铋共晶合金(Pb45Bi55)被认为是苏联和美国核电系统冷却剂的候选材料。与此同时，由A.I.雷普斯基首先提出的合金(Pb45Bi55)被选为“阿尔法”号核潜艇的冷却剂，在战术和技术数据上都领先于当时。然而，1968年，其中一艘遭遇了严重的辐射事故，由于冷却剂的氧化物沉淀和积聚，导致船载核反应堆堆芯燃料元件熔化，堵塞了管道截面，急剧恶化了反应堆的冷却，导致其失效。事实上，事故的主要原因是缺乏对铅铋冷却剂的物理化学和技术特性的了解。因此，Pb45Bi55冷却剂的主要缺点是对核电站结构材料的腐蚀性。但研究发现，通过调节冷却剂中的含氧量，可以减少液铅合金的腐蚀。例如，在770 K时，马氏体钢在Pb45Bi55无氧流动冷却液中的腐蚀速率约为1 mm /年，而在Pb45Bi55冷却液中溶解氧气并保持其质量浓度为0.01 ppm的情况下，腐蚀速率可降至0.01 mm /年，即100次。观察到的效果是由管道钢表面形成的氧化层提供的保护来解释的。因此，为了更深入地了解液态金属“冷却剂-气体”部分边界上发生的现象，研究保护氧化层的形成和破坏过程及其在冷却剂流体中的行为，特别是从核电厂长期运行的角度来看，仍然具有重要意义。在这方面，关于“冷却剂-气体”相界面表面张力的数据具有重要的科学和实际意义。为此，本工作的任务是实验研究大气对Pb45Bi55共晶熔体表面张力的影响。在静态真空和常压条件下，对制备的共晶熔体Pb44, 6bi55,4的表面张力进行了连续测量。在这项工作中获得的大约300个实验点使得描述表面张力变化过程的动力学成为可能，这取决于冷却剂表面在真空和大气中暴露的时间。结果表明，与作者在静态真空中采用大垂滴法得到的结果相比，共晶熔体在大气压力约300 mmHg下暴露后的前10分钟内，共晶Pb44.6Bi55.4的表面张力下降了55 mN/m以上，比我们测量的总误差(2%)高一个数量级。
{"title":"SURFACE TENSION OF Pb44,6Bi55,4 EUTECTIC MELT IN AIR ATMOSPHERE","authors":"B. Alchagirov, A. Khibiev","doi":"10.55176/2414-1038-2021-3-123-135","DOIUrl":"https://doi.org/10.55176/2414-1038-2021-3-123-135","url":null,"abstract":"In the 1950s liquid lead and the lead-bismuth eutectic alloy (Pb45Bi55) were considered candidates for their use as coolants for nuclear power systems in the USSR and the USA. At the same time, the alloy (Pb45Bi55), first proposed by A.I. Leypunsky, was chosen as a coolant for the nuclear submarine “Alpha”, which was ahead of its time in terms of its tactical and technical data. However, in 1968, one of them suffered a severe radiation accident caused by the melting of fuel elements in the core of an onboard nuclear reactor due to the precipitation of oxides from the coolant and their accumulation, which blocked the pipeline cross-section and sharply worsened the cooling of the reactor, which led to its failure. In fact, the primary cause of the accident was a lack of knowledge about the physicochemical and technological properties of the lead-bismuth coolant. Thus, the main disadvantage of the Pb45Bi55 coolant is its corrosiveness to structural materials used in the nuclear power plant. But it has been found that corrosion by liquid lead alloys can be reduced by adjusting the oxygen level in the coolant. For example, the corrosion rate of martensitic steel at 770 K in a Pb45Bi55 flowing coolant without oxygen is about 1 mm per year, but it can be reduced to 0.01 mm per year, i.e. 100 times, if oxygen is dissolved in Pb45Bi55 coolant and its mass concentration is maintained at the level of 0.01 ppm. The observed effect is explained by the protection provided by the oxide layer formed on the steel surface of the pipeline. Thus, for a deeper understanding of the phenomena occurring at the boundaries of the liquid metal “coolants - gases” section, it remains relevant to study the processes of formation and destruction of the protective oxide layer and its behavior in coolant fluids, especially from the point of view of long-term operation of nuclear power plants. In this regard, data on the surface tension of the “coolant - gas” interphase boundaries are of great scientific and practical importance. In this connection, the present work sets the task of experimentally studying the influence of atmospheric air on the surface tension of Pb45Bi55 eutectic melt. Measurements of the surface tension of the eutectic melt Pb44,6Bi55,4 prepared by the authors were carried out in a non-stop mode sequentially, under static vacuum and atmospheric air, on the same surface. About three hundred experimental points obtained in this work made it possible to describe the dynamics of the surface tension changing process depending on the time of exposure of the coolant surface in vacuum and atmospheric air. It is shown that in comparison with the results obtained by the authors by the large lying drop method in a static vacuum, in the first 10 minutes from the beginning of the exposure of the eutectic melt in atmospheric air at a pressure of about 300 mmHg), the surface tension of the eutectic Pb44.6Bi55.4 decreases by 55 mN/m and more, which is an order of magnitude hi","PeriodicalId":20426,"journal":{"name":"PROBLEMS OF ATOMIC SCIENCE AND TECHNOLOGY. SERIES: NUCLEAR AND REACTOR CONSTANTS","volume":null,"pages":null},"PeriodicalIF":0.0,"publicationDate":"2021-09-26","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"79963658","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2021-09-26DOI: 10.55176/2414-1038-2021-3-49-67
G. Bezhunov, N. Rykov, B. Ryazanov
A system for measuring the mass of fissile nuclides (FN) in solid radioactive waste (SRW) in containers up to 200 liters has been developed, manufactured and tested. The measuring system includes a pulsed neutron generator (PNG) with a neutron output of 2·108 neutrons/s, a neutron moderating unit with counters of supra-cadmium neutrons and a chamber for containers up to 200 liters, a time pulse analyzer, a personal computer with a software for accumulation and processing of time spectra of pulses from neutron registration. Experimental and computational studies have been carried out for models of containers with solid radioactive waste of 120 and 211 liters with matrices of quartz sand, graphite and paper of various densities with FN content in an amount from 0.01 to 100 g using reference or well-characterized samples with different enrichment in U-235 (from 5 to 90 %) and PuO2 reference sample, with different chemical composition (dioxide, uranyl nitrate solution, oxide-nitrous oxide), different geometrical sizes, placed homogeneously or heterogeneously over the volume of the container. The parameters of the measuring system were determined, including the response values in units of counts per second per gram of FN in the container and per the PNG neutron. The measurement time at an PNG frequency of 20 Hz with a neutron yield of ~107 neutrons/pulse is from 100 to 300 s. The lower limit for U-235 mass measuring for typical TPO matrices is 0.01 g per container. The influence of various factors on the measurement results was estimated: the moisture content of the matrix, the chemical composition and density of the matrices, the mass fraction of U-238 in uranium, the heterogeneity of the FN arrangement in the container, the presence of internal neutron sources. Measurements of the mass of fissile nuclides in containers using the developed system are possible for the case of the presence of internal neutron sources in the container with an intensity of up to 5·107 neutrons/s per container.
{"title":"DEVELOPMENT OF A COMPACT PULSED NEUTRON SYSTEM FOR MEASURING THE MASS OF FISSILE NUCLIDES IN SOLID RADIOACTIVE WASTE","authors":"G. Bezhunov, N. Rykov, B. Ryazanov","doi":"10.55176/2414-1038-2021-3-49-67","DOIUrl":"https://doi.org/10.55176/2414-1038-2021-3-49-67","url":null,"abstract":"A system for measuring the mass of fissile nuclides (FN) in solid radioactive waste (SRW) in containers up to 200 liters has been developed, manufactured and tested. The measuring system includes a pulsed neutron generator (PNG) with a neutron output of 2·108 neutrons/s, a neutron moderating unit with counters of supra-cadmium neutrons and a chamber for containers up to 200 liters, a time pulse analyzer, a personal computer with a software for accumulation and processing of time spectra of pulses from neutron registration. Experimental and computational studies have been carried out for models of containers with solid radioactive waste of 120 and 211 liters with matrices of quartz sand, graphite and paper of various densities with FN content in an amount from 0.01 to 100 g using reference or well-characterized samples with different enrichment in U-235 (from 5 to 90 %) and PuO2 reference sample, with different chemical composition (dioxide, uranyl nitrate solution, oxide-nitrous oxide), different geometrical sizes, placed homogeneously or heterogeneously over the volume of the container. The parameters of the measuring system were determined, including the response values in units of counts per second per gram of FN in the container and per the PNG neutron. The measurement time at an PNG frequency of 20 Hz with a neutron yield of ~107 neutrons/pulse is from 100 to 300 s. The lower limit for U-235 mass measuring for typical TPO matrices is 0.01 g per container. The influence of various factors on the measurement results was estimated: the moisture content of the matrix, the chemical composition and density of the matrices, the mass fraction of U-238 in uranium, the heterogeneity of the FN arrangement in the container, the presence of internal neutron sources. Measurements of the mass of fissile nuclides in containers using the developed system are possible for the case of the presence of internal neutron sources in the container with an intensity of up to 5·107 neutrons/s per container.","PeriodicalId":20426,"journal":{"name":"PROBLEMS OF ATOMIC SCIENCE AND TECHNOLOGY. SERIES: NUCLEAR AND REACTOR CONSTANTS","volume":null,"pages":null},"PeriodicalIF":0.0,"publicationDate":"2021-09-26","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"80714651","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2021-09-26DOI: 10.55176/2414-1038-2021-3-184-190
D. Skobeev, A. Legkikh
In order to ensure the safe operation of reactor installations under development with heavy liquid metal coolants (HLMC), such as lead and lead-bismuth, it is necessary to address issues related to the control and regulation of the oxygen potential of the coolant. This is necessary to maintain the normalized range of dissolved oxygen concentrations to ensure the conditions for the formation and maintenance of the integrity of protective oxide coatings on structural materials, as well as to prevent the formation of oxide slags from the coolant during the operation of the reactor plant. Specialists of IPPE have developed a method and a means of solid-phase regulation of the thermo-dynamic activity of oxygen in HLMC. In mass transfer devices developed for the implementation of the solid-phase method of controlled feeding of a heavy liquid metal coolant with dissolved oxygen, lead oxide granules are used as a filler. One of the important issues in substantiating the reliability of mass transfer devices is the question of the constancy of the mechanical properties and chemical composition of lead oxide granules after their prolonged stay under the level of a heavy liquid metal coolant at operating temperature. The lead oxide granules were aged in a sealed container filled with lead. The tank was equipped with the necessary means to control the temperature of the lead and the pressure of the protective gas (argon). The article presents the results of experimental studies of lead oxide granules after their exposure in a lead melt at a temperature of (420±10) °C in a non-carbon mode for 6000 hours. To investigate the characteristics of the lead oxide pellets, a batch of pellets was excavated after 500, 750, 1000, 1750, 2500, 3000, 4500 and 6000 hours from the start of the tests. For the pellets from each recess, the following studies were performed: - pellet density measurements; - measurement of breaking forces of lead oxide granules; - determination of pellet impact strength; - chemical analysis of granule composition.
{"title":"INVESTIGATION OF THE CHARACTERISTICS OF LEAD OXIDE GRANULES AFTER PROLONGED EXPOSURE IN LIQUID LEAD","authors":"D. Skobeev, A. Legkikh","doi":"10.55176/2414-1038-2021-3-184-190","DOIUrl":"https://doi.org/10.55176/2414-1038-2021-3-184-190","url":null,"abstract":"In order to ensure the safe operation of reactor installations under development with heavy liquid metal coolants (HLMC), such as lead and lead-bismuth, it is necessary to address issues related to the control and regulation of the oxygen potential of the coolant. This is necessary to maintain the normalized range of dissolved oxygen concentrations to ensure the conditions for the formation and maintenance of the integrity of protective oxide coatings on structural materials, as well as to prevent the formation of oxide slags from the coolant during the operation of the reactor plant. Specialists of IPPE have developed a method and a means of solid-phase regulation of the thermo-dynamic activity of oxygen in HLMC. In mass transfer devices developed for the implementation of the solid-phase method of controlled feeding of a heavy liquid metal coolant with dissolved oxygen, lead oxide granules are used as a filler. One of the important issues in substantiating the reliability of mass transfer devices is the question of the constancy of the mechanical properties and chemical composition of lead oxide granules after their prolonged stay under the level of a heavy liquid metal coolant at operating temperature. The lead oxide granules were aged in a sealed container filled with lead. The tank was equipped with the necessary means to control the temperature of the lead and the pressure of the protective gas (argon). The article presents the results of experimental studies of lead oxide granules after their exposure in a lead melt at a temperature of (420±10) °C in a non-carbon mode for 6000 hours. To investigate the characteristics of the lead oxide pellets, a batch of pellets was excavated after 500, 750, 1000, 1750, 2500, 3000, 4500 and 6000 hours from the start of the tests. For the pellets from each recess, the following studies were performed: - pellet density measurements; - measurement of breaking forces of lead oxide granules; - determination of pellet impact strength; - chemical analysis of granule composition.","PeriodicalId":20426,"journal":{"name":"PROBLEMS OF ATOMIC SCIENCE AND TECHNOLOGY. SERIES: NUCLEAR AND REACTOR CONSTANTS","volume":null,"pages":null},"PeriodicalIF":0.0,"publicationDate":"2021-09-26","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"82239001","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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