Pub Date : 2020-12-26DOI: 10.55176/2414-1038-2020-4-16-25
V. Koscheev, I. Tormischev, V. Mischin, A. Peregudov, K. Raskach, M. Semenov, A. Yakunin
The purpose of this work was to test the CROSSER high-speed constant preparation module for use in the practice of engineering calculations of the neutron-physical characteristics of fast reactors. Computational models of fast reactors BN-600, BN-800 and BN-1200 were used as test tasks. The following neutron-physical parameters were calculated: the criticality of the reactor model, the efficiency of the CPS and the sodium void coefficient of reactivity. The main reactor functionals were calculated using the MMKK engineering program, the constants for which were prepared using the CROSSER module. The initial sets of nuclear data for the MMKK program were the ABBN-93 library of constants and the constants in the ABBN-RFE formats obtained on the basis of the ROSFOND-2010 library. The results obtained using the MMKC program with detailed tracking of the particle energy were used as the reference calculation results. Detailed data in ACE format obtained from the ROSFOND 2010 library were used as the initial nuclear data for this program.
{"title":"CROSSER - SOFTWARE MODULE FOR PREPARATION GROUP CONSTANTS FOR ENGINEERING CALCULATIONS FAST REACTORS","authors":"V. Koscheev, I. Tormischev, V. Mischin, A. Peregudov, K. Raskach, M. Semenov, A. Yakunin","doi":"10.55176/2414-1038-2020-4-16-25","DOIUrl":"https://doi.org/10.55176/2414-1038-2020-4-16-25","url":null,"abstract":"The purpose of this work was to test the CROSSER high-speed constant preparation module for use in the practice of engineering calculations of the neutron-physical characteristics of fast reactors. Computational models of fast reactors BN-600, BN-800 and BN-1200 were used as test tasks. The following neutron-physical parameters were calculated: the criticality of the reactor model, the efficiency of the CPS and the sodium void coefficient of reactivity. The main reactor functionals were calculated using the MMKK engineering program, the constants for which were prepared using the CROSSER module. The initial sets of nuclear data for the MMKK program were the ABBN-93 library of constants and the constants in the ABBN-RFE formats obtained on the basis of the ROSFOND-2010 library. The results obtained using the MMKC program with detailed tracking of the particle energy were used as the reference calculation results. Detailed data in ACE format obtained from the ROSFOND 2010 library were used as the initial nuclear data for this program.","PeriodicalId":20426,"journal":{"name":"PROBLEMS OF ATOMIC SCIENCE AND TECHNOLOGY. SERIES: NUCLEAR AND REACTOR CONSTANTS","volume":null,"pages":null},"PeriodicalIF":0.0,"publicationDate":"2020-12-26","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"85847794","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2020-12-26DOI: 10.55176/2414-1038-2020-4-33-39
V. Namakonov, S. Andreyev, D. Gabbasov, A. Moseyeva, D. Sergina
The results of experiments on transmission of 14 MeV neutrons through lithium hydride layers of thickness up to 25 cm are presented in the article. The measurements were performed with time-of-flight method on a pulse channel of neutron generator NG-12I. The operating mode of the neutron generator is pulse-periodic. Neutrons passing through the layers of Li hydrides were registered by detector based on a 70×70 mm stilbene crystal scintillator. NIM standard modules were used as measuring equipment. The neutron yield from the generator target was estimated by neutron monitor with a fluorine plastic activation detector. Activity of radiation-exposed neutron activation detectors was measured using a gamma spectrometer with high purity germanium detector (HPGe). The averaged 14 MeV neutrons flux from the target was ∼2⋅108 n/s. The measurement results were used to obtain instrumental neutron spectra for samples of various thicknesses and to estimate coefficients of 14 MeV neutrons passing through the layers of Li hydrides. The obtained results can be used for validation of neutron-physical calculations and for improvement of neutron constants libraries.
{"title":"MEASUREMENT OF 14 MEV NEUTRONS TRANSMISSION THROUGH LITHIUM HYDRIDE LAYERS WITH TIME-OF-FLIGHT METHOD","authors":"V. Namakonov, S. Andreyev, D. Gabbasov, A. Moseyeva, D. Sergina","doi":"10.55176/2414-1038-2020-4-33-39","DOIUrl":"https://doi.org/10.55176/2414-1038-2020-4-33-39","url":null,"abstract":"The results of experiments on transmission of 14 MeV neutrons through lithium hydride layers of thickness up to 25 cm are presented in the article. The measurements were performed with time-of-flight method on a pulse channel of neutron generator NG-12I. The operating mode of the neutron generator is pulse-periodic. Neutrons passing through the layers of Li hydrides were registered by detector based on a 70×70 mm stilbene crystal scintillator. NIM standard modules were used as measuring equipment. The neutron yield from the generator target was estimated by neutron monitor with a fluorine plastic activation detector. Activity of radiation-exposed neutron activation detectors was measured using a gamma spectrometer with high purity germanium detector (HPGe). The averaged 14 MeV neutrons flux from the target was ∼2⋅108 n/s. The measurement results were used to obtain instrumental neutron spectra for samples of various thicknesses and to estimate coefficients of 14 MeV neutrons passing through the layers of Li hydrides. The obtained results can be used for validation of neutron-physical calculations and for improvement of neutron constants libraries.","PeriodicalId":20426,"journal":{"name":"PROBLEMS OF ATOMIC SCIENCE AND TECHNOLOGY. SERIES: NUCLEAR AND REACTOR CONSTANTS","volume":null,"pages":null},"PeriodicalIF":0.0,"publicationDate":"2020-12-26","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"88338031","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2020-12-26DOI: 10.55176/2414-1038-2020-4-163-171
A. Semchenkov, I. Kustova, O. Nikel, Y. Kabanov
The stability of the circulation in the steam-water circuit, that is, the preservation of the thermohydraulic parameters within acceptable limits for perturbations characteristic of the operation, largely determines the reliability of the steam generator itself. Under certain conditions, pulsating pressure, flow and temperature pulsations may appear in boiling apparatus. For the first time, ripples were encountered during the creation of once-through boilers, and the harmful effect of flow instability on the apparatus design led to the need for a detailed study of the problem. In this work, a computational analysis of the thermal-hydraulic stability of the reverse steam generator (OPG) of the MBIR nuclear research facility (NRF) is carried out. The need for such a study is due to the fact that the OPG consists of three parallel connected steam generating modules. With such a design, water flow rate pulsations may appear, including with overturning flow rates in separate steam-generating channels. To avoid this, it is necessary to conduct throttling at the inlet to each OPG module, therefore, in this study, the hydraulic resistance of the chokes is determined. The research methodology is based on the use of a numerical thermohydraulic model of a reverse steam generator, made by means of a certified thermohydraulic design code HYDRA-IBRAE/LM/V1. The OPG RNU MBIR at the economizer and superheating sections has twisted heat exchange intensifiers that direct water and steam both in the longitudinal and transverse directions relative to the bundle of heat exchange tubes. This circumstance required a special approach in calculating the hydraulic resistance and heat transfer coefficients along the path of the OPG modules. After the creation of the computational model, it had to be verified, and since the design of the OPG of the NRF MBIR is similar to the design of the OPG of the BOR-60 reactor, the experimental data obtained on the OPG BOR-60 were used for verification. The calculations of the normal operation modes of the OPG NRF MBIR after the throttling of the modules confirmed the absence of fluctuations in the flow rate in these modes.
{"title":"COMPUTATIONAL ANALYSIS OF THERMAL HYDRAULIC STABILITY REVERSE STEAM GENERATOR NUCLEAR RESEARCH FACILITY MBIR","authors":"A. Semchenkov, I. Kustova, O. Nikel, Y. Kabanov","doi":"10.55176/2414-1038-2020-4-163-171","DOIUrl":"https://doi.org/10.55176/2414-1038-2020-4-163-171","url":null,"abstract":"The stability of the circulation in the steam-water circuit, that is, the preservation of the thermohydraulic parameters within acceptable limits for perturbations characteristic of the operation, largely determines the reliability of the steam generator itself. Under certain conditions, pulsating pressure, flow and temperature pulsations may appear in boiling apparatus. For the first time, ripples were encountered during the creation of once-through boilers, and the harmful effect of flow instability on the apparatus design led to the need for a detailed study of the problem. In this work, a computational analysis of the thermal-hydraulic stability of the reverse steam generator (OPG) of the MBIR nuclear research facility (NRF) is carried out. The need for such a study is due to the fact that the OPG consists of three parallel connected steam generating modules. With such a design, water flow rate pulsations may appear, including with overturning flow rates in separate steam-generating channels. To avoid this, it is necessary to conduct throttling at the inlet to each OPG module, therefore, in this study, the hydraulic resistance of the chokes is determined. The research methodology is based on the use of a numerical thermohydraulic model of a reverse steam generator, made by means of a certified thermohydraulic design code HYDRA-IBRAE/LM/V1. The OPG RNU MBIR at the economizer and superheating sections has twisted heat exchange intensifiers that direct water and steam both in the longitudinal and transverse directions relative to the bundle of heat exchange tubes. This circumstance required a special approach in calculating the hydraulic resistance and heat transfer coefficients along the path of the OPG modules. After the creation of the computational model, it had to be verified, and since the design of the OPG of the NRF MBIR is similar to the design of the OPG of the BOR-60 reactor, the experimental data obtained on the OPG BOR-60 were used for verification. The calculations of the normal operation modes of the OPG NRF MBIR after the throttling of the modules confirmed the absence of fluctuations in the flow rate in these modes.","PeriodicalId":20426,"journal":{"name":"PROBLEMS OF ATOMIC SCIENCE AND TECHNOLOGY. SERIES: NUCLEAR AND REACTOR CONSTANTS","volume":null,"pages":null},"PeriodicalIF":0.0,"publicationDate":"2020-12-26","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"76327268","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2020-12-26DOI: 10.55176/2414-1038-2020-4-26-32
E. Gomin, V. Davidenko, R. Shirokov
The article presents the results of testing the KIR code designed to solve the neutron transfer equation by the Monte Carlo method, based on available experimental data from the international Bank ICSBEP on solution benchmarks. The KIR program is part of the DAREUS program complex designed for modeling dynamic processes in solution reactors. Computer models of 13 critical assemblies were created. The total number of calculated configurations was 137. The results were analyzed and the calculation errors were determined both in comparison with the experiment and with other programs. In most cases, the calculated keff multiplication factor falls within the experimental error. The obtained calculation results show a mean square deviation of the multiplication coefficient of 0.7 % for all calculated configurations of experimental assemblies. However, for a number of assemblies, there are significant differences in the assessment. Similar deviations in the calculated values of keff were observed in a variety of previously performed calculations for other codes. The article presents the main factors affecting the results of calculations. Mainly they are related to the insufficiently complete description of experiments submitted to the ICSBEP International Bank.
{"title":"A PROGRAM COMPLEX KIR VERIFICATION ON CRITICAL EXPERIMENTS WITH SOLUTION ASSEMBLIES","authors":"E. Gomin, V. Davidenko, R. Shirokov","doi":"10.55176/2414-1038-2020-4-26-32","DOIUrl":"https://doi.org/10.55176/2414-1038-2020-4-26-32","url":null,"abstract":"The article presents the results of testing the KIR code designed to solve the neutron transfer equation by the Monte Carlo method, based on available experimental data from the international Bank ICSBEP on solution benchmarks. The KIR program is part of the DAREUS program complex designed for modeling dynamic processes in solution reactors. Computer models of 13 critical assemblies were created. The total number of calculated configurations was 137. The results were analyzed and the calculation errors were determined both in comparison with the experiment and with other programs. In most cases, the calculated keff multiplication factor falls within the experimental error. The obtained calculation results show a mean square deviation of the multiplication coefficient of 0.7 % for all calculated configurations of experimental assemblies. However, for a number of assemblies, there are significant differences in the assessment. Similar deviations in the calculated values of keff were observed in a variety of previously performed calculations for other codes. The article presents the main factors affecting the results of calculations. Mainly they are related to the insufficiently complete description of experiments submitted to the ICSBEP International Bank.","PeriodicalId":20426,"journal":{"name":"PROBLEMS OF ATOMIC SCIENCE AND TECHNOLOGY. SERIES: NUCLEAR AND REACTOR CONSTANTS","volume":null,"pages":null},"PeriodicalIF":0.0,"publicationDate":"2020-12-26","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"86070053","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2020-12-26DOI: 10.55176/2414-1038-2020-4-148-157
S. Dmitriev, A. Dobrov, D. Doronkov, V. Lyskova, A. Pronin, E. Rubtsova, A. Ryazanov, D. Solntsev, A. Khrobostov
The results of experimental studies of the interassembly interaction of coolant in VVER reactor core which consists of TVSA-T and upgraded TVSA are presented. The modeling of coolant flow in the fuel assembly (FA) was carried out on an aerodynamic stand. The studies were carried out on a fragment model of VVER reactor core and consisted in measuring the velocity vector modulus in the characteristic zones of both TVSA and interassembly space of VVER reactor core. The measurements were carried out by a five-channel pneumometric probe. An analysis of the spatial distribution of projections of the absolute flow velocity allowed to detail the pattern of flow-round by the coolant flow of spacer, mixing and combined spacer grids TVSA. The results of study of interassembly interaction of coolant between neighboring TVSA-T and upgraded TVSA were adopted for practical use by JSC “Afrikantov OKBM” in assessing heat engineering reliability of VVER reactor cores and are included in the data-base for verification of computational fluid dynamics programs (CFD codes) and detailed cellular calculation of VVER reactor core.
{"title":"FLUID DYNAMICS IN THE CORE OF A WWER REACTOR WITH TVSA OF VARIOUS DESIGNS","authors":"S. Dmitriev, A. Dobrov, D. Doronkov, V. Lyskova, A. Pronin, E. Rubtsova, A. Ryazanov, D. Solntsev, A. Khrobostov","doi":"10.55176/2414-1038-2020-4-148-157","DOIUrl":"https://doi.org/10.55176/2414-1038-2020-4-148-157","url":null,"abstract":"The results of experimental studies of the interassembly interaction of coolant in VVER reactor core which consists of TVSA-T and upgraded TVSA are presented. The modeling of coolant flow in the fuel assembly (FA) was carried out on an aerodynamic stand. The studies were carried out on a fragment model of VVER reactor core and consisted in measuring the velocity vector modulus in the characteristic zones of both TVSA and interassembly space of VVER reactor core. The measurements were carried out by a five-channel pneumometric probe. An analysis of the spatial distribution of projections of the absolute flow velocity allowed to detail the pattern of flow-round by the coolant flow of spacer, mixing and combined spacer grids TVSA. The results of study of interassembly interaction of coolant between neighboring TVSA-T and upgraded TVSA were adopted for practical use by JSC “Afrikantov OKBM” in assessing heat engineering reliability of VVER reactor cores and are included in the data-base for verification of computational fluid dynamics programs (CFD codes) and detailed cellular calculation of VVER reactor core.","PeriodicalId":20426,"journal":{"name":"PROBLEMS OF ATOMIC SCIENCE AND TECHNOLOGY. SERIES: NUCLEAR AND REACTOR CONSTANTS","volume":null,"pages":null},"PeriodicalIF":0.0,"publicationDate":"2020-12-26","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"81085369","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2020-09-26DOI: 10.55176/2414-1038-2020-3-135-147
V. Grabezhnaya, A. Mikheyev, A. Kryukov
The BREST-OD-300 steam generator project being developed at NIKIET is pioneering both in terms of the heat carrier used (lead) and in design implementation (coils of helical heat transfer tubes). The advantages of the designs of steam generators made in the form of helical coiled tubes, in comparison with straight tube designs are obvious. Helical coiled tubes are used in heat exchange equipment not only to increase the heat transfer surface, to solve the problem of thermal expansion, but also to increase the coefficient of heat transfer to the fluid flowing inside the tubes. In 2011-2017 years the thermohydraulic tests of various models of lead-heated steam generator were carried out at the IPPE SPRUT facility (IPPE). The test program was aimed to study the heat transfer and the thermal-hydraulic stability of the steam generating tubes. Throughout the entire range of changes in operating parameters, no pulsating modes were detected with overturn of circulation in the water circuit. The design temperatures of superheated steam were obtained in nominal operation. The results provide extensive information on water heat transfer in different zones of the steam generating channel under various operating conditions (nominal and partial modes, starting modes). However, to verify the computer codes, experimental data on the heat transfer of lead coolant around the bundles of heat transfer tubes are necessary. Due to the small twist angle of the tubes in a full-scale steam generator, it can be said that heat transfer is close to heat transfer during transverse flow. A model with a transverse flow of lead coolant around steam-generating tubes was developed at the SSC RF - IPPE. The main goal of the research was to obtain data on the effect of oxygen concentration in lead on heat transfer in normal heat transfer modes and with lead freezing. Throughout the entire range of changes in the initial temperature values of lead and water, blocking of the annular space by frozen lead was not recorded.
{"title":"ON THE EFFECT OF OXYGEN IMPURITIES ON HEAT TRANSFER AT TRANSVERSAL FLOW OF STEAM-GENERATING TUBES IN NORMAL HEAT TRANSFER MODES AND WITH LEAD FREEZE","authors":"V. Grabezhnaya, A. Mikheyev, A. Kryukov","doi":"10.55176/2414-1038-2020-3-135-147","DOIUrl":"https://doi.org/10.55176/2414-1038-2020-3-135-147","url":null,"abstract":"The BREST-OD-300 steam generator project being developed at NIKIET is pioneering both in terms of the heat carrier used (lead) and in design implementation (coils of helical heat transfer tubes). The advantages of the designs of steam generators made in the form of helical coiled tubes, in comparison with straight tube designs are obvious. Helical coiled tubes are used in heat exchange equipment not only to increase the heat transfer surface, to solve the problem of thermal expansion, but also to increase the coefficient of heat transfer to the fluid flowing inside the tubes. In 2011-2017 years the thermohydraulic tests of various models of lead-heated steam generator were carried out at the IPPE SPRUT facility (IPPE). The test program was aimed to study the heat transfer and the thermal-hydraulic stability of the steam generating tubes. Throughout the entire range of changes in operating parameters, no pulsating modes were detected with overturn of circulation in the water circuit. The design temperatures of superheated steam were obtained in nominal operation. The results provide extensive information on water heat transfer in different zones of the steam generating channel under various operating conditions (nominal and partial modes, starting modes). However, to verify the computer codes, experimental data on the heat transfer of lead coolant around the bundles of heat transfer tubes are necessary. Due to the small twist angle of the tubes in a full-scale steam generator, it can be said that heat transfer is close to heat transfer during transverse flow. A model with a transverse flow of lead coolant around steam-generating tubes was developed at the SSC RF - IPPE. The main goal of the research was to obtain data on the effect of oxygen concentration in lead on heat transfer in normal heat transfer modes and with lead freezing. Throughout the entire range of changes in the initial temperature values of lead and water, blocking of the annular space by frozen lead was not recorded.","PeriodicalId":20426,"journal":{"name":"PROBLEMS OF ATOMIC SCIENCE AND TECHNOLOGY. SERIES: NUCLEAR AND REACTOR CONSTANTS","volume":null,"pages":null},"PeriodicalIF":0.0,"publicationDate":"2020-09-26","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"90201203","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2020-09-26DOI: 10.55176/2414-1038-2020-3-5-18
R. Sinyukov, P. Blokhin, A. Pryanichnikov, A. Simakov, M. Belikhin, I. Degtyarev, F. Novoskoltsev, E. Altukhova, Yu. V. Altukhov, A. Blokhin
The paper describes the physical basis of the mathematical model of radiation-ion damage in solid matter. The following stages of defect generation are considered: primary knocked out atoms, atomic collision cascades (dynamic stage of damage formation) based on 6 types of cascade functions. The software implementation of the described model is included in the RTS&T code designed for statistical modeling of multi-particle (200 types of particles, resonances and ions) radiation transport in heterogeneous 3D geometries in the energy range up to 20 TeV. In the region of low energies, the RTS&T transfer model is based on direct use of all the information contained in the evaluated nuclear data files submitted in the ENDF-6 format. The report compares the results of numerical modeling with experimental data.
{"title":"MATHEMATICAL MODEL OF PRIMARY RADIATION DAMAGE TO THE BOROSILICATE GLASS INTENDED FOR IMMOBILIZATION OF RADIOACTIVE WASTE","authors":"R. Sinyukov, P. Blokhin, A. Pryanichnikov, A. Simakov, M. Belikhin, I. Degtyarev, F. Novoskoltsev, E. Altukhova, Yu. V. Altukhov, A. Blokhin","doi":"10.55176/2414-1038-2020-3-5-18","DOIUrl":"https://doi.org/10.55176/2414-1038-2020-3-5-18","url":null,"abstract":"The paper describes the physical basis of the mathematical model of radiation-ion damage in solid matter. The following stages of defect generation are considered: primary knocked out atoms, atomic collision cascades (dynamic stage of damage formation) based on 6 types of cascade functions. The software implementation of the described model is included in the RTS&T code designed for statistical modeling of multi-particle (200 types of particles, resonances and ions) radiation transport in heterogeneous 3D geometries in the energy range up to 20 TeV. In the region of low energies, the RTS&T transfer model is based on direct use of all the information contained in the evaluated nuclear data files submitted in the ENDF-6 format. The report compares the results of numerical modeling with experimental data.","PeriodicalId":20426,"journal":{"name":"PROBLEMS OF ATOMIC SCIENCE AND TECHNOLOGY. SERIES: NUCLEAR AND REACTOR CONSTANTS","volume":null,"pages":null},"PeriodicalIF":0.0,"publicationDate":"2020-09-26","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"91279633","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2020-09-26DOI: 10.55176/2414-1038-2020-3-127-134
A. Osipov, K. Ivanov
In the initial period of HLMT development, it was shown that ensuring the corrosion resistance of structural steels is associated with ensuring a certain oxygen regime of the coolant. To control it, oxygen TDA sensors were created later. It was assumed that the oxygen activity in HLMC in different temperature zones of the circulation loop obeys the so-called "isoconcentration" distribution, which, when formally recalculated the measured oxygen TDA values, gives a constant concentration value at different temperatures. However, later it turned out that such a distribution of TDA is not always realized. The observed character of deviations could be explained by the influence of iron impurity. At the same time, quantitative estimates of this effect were carried out under the assumption of the formation of stoichiometric magnetite under conditions of its thermodynamic stability in the entire range of temperatures and concentrations of the initial components. The limitation of this approach lies, first of all, in the fact that it does not take into account the processes of dissociation of solid-phase iron oxides, which can occur in the hot zone when appropriate conditions are created, which can be realized with a decrease in the content of dispersed iron oxides. The importance of taking this factor into account is due to the fact that, during corrosion testing of steels in HLMC, the processes of dissociation of magnetite are actually observed in practice. Within the framework of this work, a computational method for assessing the effect of filtration processes on the thermodynamic state of HLMC has been developed. Quantitative estimates of the effect of the content of dispersed oxides in HLMC on the thermodynamic state of the coolant under non-isothermal conditions in iso-concentration and non-iso-concentration modes have been obtained and which can be used in calculation codes and comparison of experimental results with calculations.
{"title":"TO THE QUESTION ABOUT THE INFLUENCE OF DISPERSED IRON OXIDES ON THERMODYNAMIC STATE OF HLMC UNDER NON-ISOTHERMAL CONDITIONS","authors":"A. Osipov, K. Ivanov","doi":"10.55176/2414-1038-2020-3-127-134","DOIUrl":"https://doi.org/10.55176/2414-1038-2020-3-127-134","url":null,"abstract":"In the initial period of HLMT development, it was shown that ensuring the corrosion resistance of structural steels is associated with ensuring a certain oxygen regime of the coolant. To control it, oxygen TDA sensors were created later. It was assumed that the oxygen activity in HLMC in different temperature zones of the circulation loop obeys the so-called \"isoconcentration\" distribution, which, when formally recalculated the measured oxygen TDA values, gives a constant concentration value at different temperatures. However, later it turned out that such a distribution of TDA is not always realized. The observed character of deviations could be explained by the influence of iron impurity. At the same time, quantitative estimates of this effect were carried out under the assumption of the formation of stoichiometric magnetite under conditions of its thermodynamic stability in the entire range of temperatures and concentrations of the initial components. The limitation of this approach lies, first of all, in the fact that it does not take into account the processes of dissociation of solid-phase iron oxides, which can occur in the hot zone when appropriate conditions are created, which can be realized with a decrease in the content of dispersed iron oxides. The importance of taking this factor into account is due to the fact that, during corrosion testing of steels in HLMC, the processes of dissociation of magnetite are actually observed in practice. Within the framework of this work, a computational method for assessing the effect of filtration processes on the thermodynamic state of HLMC has been developed. Quantitative estimates of the effect of the content of dispersed oxides in HLMC on the thermodynamic state of the coolant under non-isothermal conditions in iso-concentration and non-iso-concentration modes have been obtained and which can be used in calculation codes and comparison of experimental results with calculations.","PeriodicalId":20426,"journal":{"name":"PROBLEMS OF ATOMIC SCIENCE AND TECHNOLOGY. SERIES: NUCLEAR AND REACTOR CONSTANTS","volume":null,"pages":null},"PeriodicalIF":0.0,"publicationDate":"2020-09-26","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"74722106","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2020-09-26DOI: 10.55176/2414-1038-2020-3-72-79
G. Zherdev, A. Suvorov
The computer program is intended for design and operational calculations that serve to substantiate the safety of nuclear facilities in documents submitted to ROSTEKHNADZOR and declared as a means of engineering calculations of radiation characteristics of spent fuel assemblies declared in the following topics: calculation of radiation protection and radiation safety of nuclear facilities; neutron-physical calculations (nuclide composition calculations). The computer program can be used independently or as part of other software systems for calculating the characteristics of objects consisting of spent fuel assemblies of uranium-graphite reactors EGP-6, AMB and AM and / or their parts. Computer program Nuclear Calculator EGP-6 2.0 is a set of WIMS-D5B and ORIGEN-S programs, as well as a number of auxiliary programs that provide both the preparation of initial data for the above programs and their interaction, and the processing of results to obtain radiation characteristics. The program has been thoroughly verified and validated. Today the program is a working tool for engineering and operational calculations, focused on the end user. Uses simple and straightforward inputs. The results are prepared in the format required by the customer. The state registration of the program was carried out.
{"title":"PROGRAM “NUCLEAR CALCULATOR EGP-6” FOR ENGINEERING CALCULATIONS OF RADIATION CHARACTERISTICS OF BILIBIN NPP","authors":"G. Zherdev, A. Suvorov","doi":"10.55176/2414-1038-2020-3-72-79","DOIUrl":"https://doi.org/10.55176/2414-1038-2020-3-72-79","url":null,"abstract":"The computer program is intended for design and operational calculations that serve to substantiate the safety of nuclear facilities in documents submitted to ROSTEKHNADZOR and declared as a means of engineering calculations of radiation characteristics of spent fuel assemblies declared in the following topics: calculation of radiation protection and radiation safety of nuclear facilities; neutron-physical calculations (nuclide composition calculations). The computer program can be used independently or as part of other software systems for calculating the characteristics of objects consisting of spent fuel assemblies of uranium-graphite reactors EGP-6, AMB and AM and / or their parts. Computer program Nuclear Calculator EGP-6 2.0 is a set of WIMS-D5B and ORIGEN-S programs, as well as a number of auxiliary programs that provide both the preparation of initial data for the above programs and their interaction, and the processing of results to obtain radiation characteristics. The program has been thoroughly verified and validated. Today the program is a working tool for engineering and operational calculations, focused on the end user. Uses simple and straightforward inputs. The results are prepared in the format required by the customer. The state registration of the program was carried out.","PeriodicalId":20426,"journal":{"name":"PROBLEMS OF ATOMIC SCIENCE AND TECHNOLOGY. SERIES: NUCLEAR AND REACTOR CONSTANTS","volume":null,"pages":null},"PeriodicalIF":0.0,"publicationDate":"2020-09-26","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"86354834","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2020-09-26DOI: 10.55176/2414-1038-2020-3-19-29
V. Los', A. Kuryndin, A. Kirkin, S. Sinegribov, S. Makovskiy
Features of calculation methods implemented in computer codes, evaluated nuclear data libraries and computer models have a significant impact on the results of burnup calculations. This article contains comparative calculations of VVER-1000 reactor nuclear fuel burnup and demonstrates an influence of different factors on the results of these calculations. The results of comparative analisys demonstrate the necessity to intensify creation of benchmarks with experimental data on nuclide concentarions in fuel including fuel for fast neutron reactors.
{"title":"IMPLEMENTATION OF COMPARATIVE BURNUP CALCULATIONS","authors":"V. Los', A. Kuryndin, A. Kirkin, S. Sinegribov, S. Makovskiy","doi":"10.55176/2414-1038-2020-3-19-29","DOIUrl":"https://doi.org/10.55176/2414-1038-2020-3-19-29","url":null,"abstract":"Features of calculation methods implemented in computer codes, evaluated nuclear data libraries and computer models have a significant impact on the results of burnup calculations. This article contains comparative calculations of VVER-1000 reactor nuclear fuel burnup and demonstrates an influence of different factors on the results of these calculations. The results of comparative analisys demonstrate the necessity to intensify creation of benchmarks with experimental data on nuclide concentarions in fuel including fuel for fast neutron reactors.","PeriodicalId":20426,"journal":{"name":"PROBLEMS OF ATOMIC SCIENCE AND TECHNOLOGY. SERIES: NUCLEAR AND REACTOR CONSTANTS","volume":null,"pages":null},"PeriodicalIF":0.0,"publicationDate":"2020-09-26","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"78386093","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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