Pub Date : 2021-03-26DOI: 10.55176/2414-1038-2021-1-36-47
A. Anfimov, I. Kirilov, D. Kuznetsov, O. Nikanorov, A. Salyaev
The paper presents the results of a computational analysis of simultaneously occurring beyond design basis accidents at power generating units No. 3 (BN-600) and No. 4 (BN-800) of the BNPP. Safety analysis for operating power generating units at the same time occurring beyond design basis accidents is carried out in accordance with the requirements of NP-001-15. Computational studies were performed with the SOCRAT-BN, ANSYS, ORIGEN2, and VIBROS 2.2 codes. There are considered beyond design basis accidents caused by seismic impact with intensity of 7 points on MSK-64 scale (exceeding the maximum allowed earthquake (MAE) for the site of power unit No. 3 and the corresponding MAE for the site of power unit No. 4). As a result of seismic impact at all the power generating units of the BNPP, a loss of system power supply occurs, which, together with additional failures of the main equipment, leads to beyond design basis accidents. The paper presents the results of computational studies of beyond design basis accidents in terms of reactor plant, assembly cooling pond and holding pond and radiation effects assessment. The results of a computation analysis of the beyond design basis accident showed that the release of radioactivity occurs only for the BN-600 reactor facility (where are breaking of the fuel elements cladding). At the same time, the predicted radiation doses for the first years after the accident are lower than the NRB-99/2009 criteria for making a decision on the evacuation and resettlement of inhabitants.
{"title":"CALCULATION ANALYSIS OF SIMULTANEOUSLY OCCURRING BEYOND DESIGN BASIS ACCIDENTS AT POWER GENERATING UNITS № 3 AND 4 OF THE BNPP","authors":"A. Anfimov, I. Kirilov, D. Kuznetsov, O. Nikanorov, A. Salyaev","doi":"10.55176/2414-1038-2021-1-36-47","DOIUrl":"https://doi.org/10.55176/2414-1038-2021-1-36-47","url":null,"abstract":"The paper presents the results of a computational analysis of simultaneously occurring beyond design basis accidents at power generating units No. 3 (BN-600) and No. 4 (BN-800) of the BNPP. Safety analysis for operating power generating units at the same time occurring beyond design basis accidents is carried out in accordance with the requirements of NP-001-15. Computational studies were performed with the SOCRAT-BN, ANSYS, ORIGEN2, and VIBROS 2.2 codes. There are considered beyond design basis accidents caused by seismic impact with intensity of 7 points on MSK-64 scale (exceeding the maximum allowed earthquake (MAE) for the site of power unit No. 3 and the corresponding MAE for the site of power unit No. 4). As a result of seismic impact at all the power generating units of the BNPP, a loss of system power supply occurs, which, together with additional failures of the main equipment, leads to beyond design basis accidents. The paper presents the results of computational studies of beyond design basis accidents in terms of reactor plant, assembly cooling pond and holding pond and radiation effects assessment. The results of a computation analysis of the beyond design basis accident showed that the release of radioactivity occurs only for the BN-600 reactor facility (where are breaking of the fuel elements cladding). At the same time, the predicted radiation doses for the first years after the accident are lower than the NRB-99/2009 criteria for making a decision on the evacuation and resettlement of inhabitants.","PeriodicalId":20426,"journal":{"name":"PROBLEMS OF ATOMIC SCIENCE AND TECHNOLOGY. SERIES: NUCLEAR AND REACTOR CONSTANTS","volume":null,"pages":null},"PeriodicalIF":0.0,"publicationDate":"2021-03-26","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"85552159","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 : 2021-03-26DOI: 10.55176/2414-1038-2021-1-15-26
V. Bereznev, A. Belov, D. Koltashev
The research is devoted to the features of radiation shieldind calculations by the deterministic program ODETTA, which is intended for numerical simulation of the neutron and photon transport in shielding compositions of the nuclear facilities and based on the discrete ordinates method and finite element method on unstructured tetrahedral meshes. The article describes the methods of the uncollided radiation component calculations implemented in the ODETTA program for “ray” effect elimination which is typical for discrete ordinates method in weakly scattering media with localized radiation sources. In addition, the first collision method allows to correctly simulating point sources, and the last collision method allows calculating the required functionals at the detection points located outside the computational domain. The implemented methods have been tested on computational benchmarks and experiments, a brief description of which is given in the article. The results obtained were compared with analytical and experimental data, as well as with the results of calculations by the Monte Carlo method within the Scale 6.2.3 software package. The analysis of the influence of the calculated parameters is carried out and conclusions are drawn about the effectiveness of the implemented methods.
{"title":"APPLICATION OF THE FIRST AND LAST COLLISION METHODS IN ODETTA CODE FOR RADIATION SHIELDING CALCULATIONS","authors":"V. Bereznev, A. Belov, D. Koltashev","doi":"10.55176/2414-1038-2021-1-15-26","DOIUrl":"https://doi.org/10.55176/2414-1038-2021-1-15-26","url":null,"abstract":"The research is devoted to the features of radiation shieldind calculations by the deterministic program ODETTA, which is intended for numerical simulation of the neutron and photon transport in shielding compositions of the nuclear facilities and based on the discrete ordinates method and finite element method on unstructured tetrahedral meshes. The article describes the methods of the uncollided radiation component calculations implemented in the ODETTA program for “ray” effect elimination which is typical for discrete ordinates method in weakly scattering media with localized radiation sources. In addition, the first collision method allows to correctly simulating point sources, and the last collision method allows calculating the required functionals at the detection points located outside the computational domain. The implemented methods have been tested on computational benchmarks and experiments, a brief description of which is given in the article. The results obtained were compared with analytical and experimental data, as well as with the results of calculations by the Monte Carlo method within the Scale 6.2.3 software package. The analysis of the influence of the calculated parameters is carried out and conclusions are drawn about the effectiveness of the implemented methods.","PeriodicalId":20426,"journal":{"name":"PROBLEMS OF ATOMIC SCIENCE AND TECHNOLOGY. SERIES: NUCLEAR AND REACTOR CONSTANTS","volume":null,"pages":null},"PeriodicalIF":0.0,"publicationDate":"2021-03-26","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"87653131","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 : 2021-03-26DOI: 10.55176/2414-1038-2021-1-66-73
A. Vaivod, A. Yudov, S. Besov, S. Andreev
The paper provides the precision experimental results on fission spectrum neutrons transmission through lithium hydride layers-7 with neutron activation analysis. The experiments were performed in Zababakhin All-Russia Research Institute of Technical Physics” on a critical assembly stand FKBN 2. A cylindrical multiplying system (MS) made of high enriched uranium was used as a source of fission spectrum neutrons. Activation integrals of neutron activation detectors (NADs) located at different points of the composite lithium-7 hydride reflector face-mounted on the cylindrical uranium MS were determined in the experiments. NADs based on nickel, indium, titanium, aluminum and copper were used. NADs of different types were exposed to radiation at power ~25-30 W during 1-1,5 hours. Nickel NAD was used as a monitor. Activation integrals absolute measurement error was determined by a certified technique and ranged from ~4 to ~8 % (2σ). Relative measurement error was determined by statistical accuracy, instrument peaks software processing correctness and was found not to exceed 5,4 % (2σ). Using the data obtained, spectral indices (the ratio of the normalized values of the activation integrals of various detectors types to the one of nickel detector), which provide information on the neutron spectrum were determined.
{"title":"EXPERIMENTS ON NEUTRONS TRANSMISSION THROUGH LITHIUM HYDRIDE LAYERS-7 IN URANIUM MULTIPLYING SYSTEM WITH NEUTRON ACTIVATION ANALYSIS","authors":"A. Vaivod, A. Yudov, S. Besov, S. Andreev","doi":"10.55176/2414-1038-2021-1-66-73","DOIUrl":"https://doi.org/10.55176/2414-1038-2021-1-66-73","url":null,"abstract":"The paper provides the precision experimental results on fission spectrum neutrons transmission through lithium hydride layers-7 with neutron activation analysis. The experiments were performed in Zababakhin All-Russia Research Institute of Technical Physics” on a critical assembly stand FKBN 2. A cylindrical multiplying system (MS) made of high enriched uranium was used as a source of fission spectrum neutrons. Activation integrals of neutron activation detectors (NADs) located at different points of the composite lithium-7 hydride reflector face-mounted on the cylindrical uranium MS were determined in the experiments. NADs based on nickel, indium, titanium, aluminum and copper were used. NADs of different types were exposed to radiation at power ~25-30 W during 1-1,5 hours. Nickel NAD was used as a monitor. Activation integrals absolute measurement error was determined by a certified technique and ranged from ~4 to ~8 % (2σ). Relative measurement error was determined by statistical accuracy, instrument peaks software processing correctness and was found not to exceed 5,4 % (2σ). Using the data obtained, spectral indices (the ratio of the normalized values of the activation integrals of various detectors types to the one of nickel detector), which provide information on the neutron spectrum were determined.","PeriodicalId":20426,"journal":{"name":"PROBLEMS OF ATOMIC SCIENCE AND TECHNOLOGY. SERIES: NUCLEAR AND REACTOR CONSTANTS","volume":null,"pages":null},"PeriodicalIF":0.0,"publicationDate":"2021-03-26","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"82798959","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 : 2021-03-26DOI: 10.55176/2414-1038-2021-1-55-65
M. Uvakin, A. Nikolaev, I. Makhin, E. Sotskov
Maneuvering operation test is actual problem of VVER reactors exploitation. Such tests are classified as nuclear hazardous procedures which are accompanied by power variations, regulators actions and continuous space power field fluctuations. As a result a lot of possible initial reactor plant conditions are occurred. This fact should be taken into account during experiment safety assessment. Current work presents calculation analysis results for high powered VVER unit safety assessment which was covered daily maneuvering experiment. All calculations are implemented by KORSAR/GP programming code with 3D-neutrion kinetic model. Quantitative and qualitative criteria show the successful solving of following problems: - Confirmation for normal operation limits adherence, especially for local power distribution parameters. - Reactor plant safety assessment for RIA accidents. Work contains analysis of developed method possibilities for VVER power unit calculation safety assessment during daily power maneuvering. Following criteria has been confirmed: - All conservative approach principles completely comply which has composed usual VVER safety assessment method for 3D neutron kinetic model. - All important features for reactor physics and dynamics during maneuvering have been taken into account. Maneuvering regimes calculations practical and methodical experience will be applied for the same tasks of VVER safety assessment.
{"title":"DYNAMICAL METHOD APPLICATION ANALYSIS FOR SAFETY ASSESSMENT CALCULATION OF VVER UNIT MANEUVERING REGIMES EXPERIMENT","authors":"M. Uvakin, A. Nikolaev, I. Makhin, E. Sotskov","doi":"10.55176/2414-1038-2021-1-55-65","DOIUrl":"https://doi.org/10.55176/2414-1038-2021-1-55-65","url":null,"abstract":"Maneuvering operation test is actual problem of VVER reactors exploitation. Such tests are classified as nuclear hazardous procedures which are accompanied by power variations, regulators actions and continuous space power field fluctuations. As a result a lot of possible initial reactor plant conditions are occurred. This fact should be taken into account during experiment safety assessment. Current work presents calculation analysis results for high powered VVER unit safety assessment which was covered daily maneuvering experiment. All calculations are implemented by KORSAR/GP programming code with 3D-neutrion kinetic model. Quantitative and qualitative criteria show the successful solving of following problems: - Confirmation for normal operation limits adherence, especially for local power distribution parameters. - Reactor plant safety assessment for RIA accidents. Work contains analysis of developed method possibilities for VVER power unit calculation safety assessment during daily power maneuvering. Following criteria has been confirmed: - All conservative approach principles completely comply which has composed usual VVER safety assessment method for 3D neutron kinetic model. - All important features for reactor physics and dynamics during maneuvering have been taken into account. Maneuvering regimes calculations practical and methodical experience will be applied for the same tasks of VVER safety assessment.","PeriodicalId":20426,"journal":{"name":"PROBLEMS OF ATOMIC SCIENCE AND TECHNOLOGY. SERIES: NUCLEAR AND REACTOR CONSTANTS","volume":null,"pages":null},"PeriodicalIF":0.0,"publicationDate":"2021-03-26","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"80515602","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 : 2021-03-26DOI: 10.55176/2414-1038-2021-1-74-85
A. Voronina, S. Pavlov
Engineering formulas are presented for calculating the value of methodological error in measuring the distances from the ultrasonic sensor to the surface of the irradiated fuel assembly of WWER-1000 caused by decay heat from the fuel assembly. The methodical error arises due to the water temperature gradient along ultrasonic wave path. This is due to the presence of natural convection at the surface of the fuel assembly discharged from the reactor into cooling pool at nuclear power plant. The paper presents a methodology for calculating the methodical error. It is assumed that the water temperature between the sensor and the surface of the fuel assembly is determined by convective heat transfer between the fuel assembly and the water in cooling pool at nuclear power plant. The surface of VVER-1000 fuel assemblies is modeled by a flat vertical plate with a uniform surface heat flux. The propagation of an ultrasonic wave in a medium between the surface of a fuel assembly and an ultrasonic sensor is described in the approximation of geometric acoustics. The results of numerical calculations according to the presented method are executed in the form of nomograms. Engineering formulas for calculating the value of methodological error are obtained by processing an array of calculated data. The obtained formulas are used in the development of measuring systems for monitoring the deformation of fuel assemblies of WWER-1000 in cooling pond of nuclear power plant.
{"title":"ENGINEERING FORMULAS FOR ESTIMATING THE INFLUENCE OF NATURAL CONVECTION ALONG THE SURFACE OF FUEL ASSEMBLY OF WWER-1000 ON THE RESULTS OF MEASURING ITS SIZES BY THE ULTRASONIC METHOD IN COOLING POND OF NUCLEAR POWER PLANT","authors":"A. Voronina, S. Pavlov","doi":"10.55176/2414-1038-2021-1-74-85","DOIUrl":"https://doi.org/10.55176/2414-1038-2021-1-74-85","url":null,"abstract":"Engineering formulas are presented for calculating the value of methodological error in measuring the distances from the ultrasonic sensor to the surface of the irradiated fuel assembly of WWER-1000 caused by decay heat from the fuel assembly. The methodical error arises due to the water temperature gradient along ultrasonic wave path. This is due to the presence of natural convection at the surface of the fuel assembly discharged from the reactor into cooling pool at nuclear power plant. The paper presents a methodology for calculating the methodical error. It is assumed that the water temperature between the sensor and the surface of the fuel assembly is determined by convective heat transfer between the fuel assembly and the water in cooling pool at nuclear power plant. The surface of VVER-1000 fuel assemblies is modeled by a flat vertical plate with a uniform surface heat flux. The propagation of an ultrasonic wave in a medium between the surface of a fuel assembly and an ultrasonic sensor is described in the approximation of geometric acoustics. The results of numerical calculations according to the presented method are executed in the form of nomograms. Engineering formulas for calculating the value of methodological error are obtained by processing an array of calculated data. The obtained formulas are used in the development of measuring systems for monitoring the deformation of fuel assemblies of WWER-1000 in cooling pond of nuclear power plant.","PeriodicalId":20426,"journal":{"name":"PROBLEMS OF ATOMIC SCIENCE AND TECHNOLOGY. SERIES: NUCLEAR AND REACTOR CONSTANTS","volume":null,"pages":null},"PeriodicalIF":0.0,"publicationDate":"2021-03-26","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"75725301","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 : 2021-03-26DOI: 10.55176/2414-1038-2021-1-152-161
A. Blokhina, S. Lyakishev, O. Korotkova
The article investigates the influence of coolant flow profile nonlinearity through straight-pipe steam generators tube assembly for fast neutron reactor with sodium coolant on heat exchanger power. When designing steam generators, a very important task is to correctly calculate the output parameters of the steam generator, especially the power. For the plants without reference solutions it is necessary to perform a deep analysis of the factors affecting on calculated parameters and to incorporate these parameters in the codes. An example of a new plant that do not have analogues is a shell-type steam generator for perspective fast neutron plants with liquid metal sodium coolant. The application of new solutions in steam generators design requires experimental and calculational justification of thermal hydraulic with the use of modern calculation codes. Power calculation of steam generator is carried out by thermohydraulic code “KORSAR/GP”, “PGN-2K”. One of the assumptions in coolant path parameters calculation model (tube space) is a uniform velocity profile by cross-section of tube assembly. It’s also accepted, that each heat exchange tube has the same expense of feed water. On the other hand calculational CFD and experimental studies at aerodynamic model of steam generator showed the presence of significant unevenness of coolant expense by tube space cross-section which is not taken into account in thermohydraulic calculations. The article contains the methodic of accounting for the known uneven coolant flow profile by tube space cross-section in liquid metal steam generator calculation. Based on the results obtained, measures to improve power output and reliability are proposed.
{"title":"POWER CALCULATION OF STRAIGHT-PIPE STEAM GENERATOR WITH SODIUM COOLANT","authors":"A. Blokhina, S. Lyakishev, O. Korotkova","doi":"10.55176/2414-1038-2021-1-152-161","DOIUrl":"https://doi.org/10.55176/2414-1038-2021-1-152-161","url":null,"abstract":"The article investigates the influence of coolant flow profile nonlinearity through straight-pipe steam generators tube assembly for fast neutron reactor with sodium coolant on heat exchanger power. When designing steam generators, a very important task is to correctly calculate the output parameters of the steam generator, especially the power. For the plants without reference solutions it is necessary to perform a deep analysis of the factors affecting on calculated parameters and to incorporate these parameters in the codes. An example of a new plant that do not have analogues is a shell-type steam generator for perspective fast neutron plants with liquid metal sodium coolant. The application of new solutions in steam generators design requires experimental and calculational justification of thermal hydraulic with the use of modern calculation codes. Power calculation of steam generator is carried out by thermohydraulic code “KORSAR/GP”, “PGN-2K”. One of the assumptions in coolant path parameters calculation model (tube space) is a uniform velocity profile by cross-section of tube assembly. It’s also accepted, that each heat exchange tube has the same expense of feed water. On the other hand calculational CFD and experimental studies at aerodynamic model of steam generator showed the presence of significant unevenness of coolant expense by tube space cross-section which is not taken into account in thermohydraulic calculations. The article contains the methodic of accounting for the known uneven coolant flow profile by tube space cross-section in liquid metal steam generator calculation. Based on the results obtained, measures to improve power output and reliability are proposed.","PeriodicalId":20426,"journal":{"name":"PROBLEMS OF ATOMIC SCIENCE AND TECHNOLOGY. SERIES: NUCLEAR AND REACTOR CONSTANTS","volume":null,"pages":null},"PeriodicalIF":0.0,"publicationDate":"2021-03-26","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"78810261","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 : 2021-03-26DOI: 10.55176/2414-1038-2021-1-135-144
N. Gonchar, M. Morkin
Gas-lift probe is an element of cladding failure detection system of perspective lead cooled reactor. Its function is local measurement of gaseous fission product activity in the coolant and the most defected fuel assembly localization. In the coolant leaving the defected fuel assembly the specific activity of gaseous fission products is higher than the average one in the primary circuit. In the barbotage channel of gas-lift probe gaseous fission products diffuse through the bubble interface surface into the volume of the bubbles. The bubbles deliver gaseous fission product to interface surface in the separation volume. The gas enriched with radioactive gaseous fission product goes to measurement volume of the probe. The more significant the damage and the closer the defective fuel assembly is located to the probe input, the more gaseous fission product activity will be registered. The paper presents a model of gaseous activity transfer from cladding defect to probe measuring volume. The gaseous activity transfer is described on the basis of the inert gases dissolution/degassing processes in lead. The gas-lift probe sensitivity was estimated as the ratio of the entry velocity of gaseous activity into the measurement volume to the exit one into the coolant through fuel assemblies cladding defects. A gas-lift probe sensitivity for exposed fuel surface calculated as an example. Gaseous fission products with significant gamma radiation are considered. The calculation results are presented in the article.
{"title":"A SENSITIVITY EVALUATION OF GAS-LIFT PROBE INCLUDED INTO CLADDING FAILURE DETECTION SYSTEM BY THE MODEL OF GASEOUS FISSION PRODUCT SOLUTION/DEGASSING INTO LEAD COOLANT","authors":"N. Gonchar, M. Morkin","doi":"10.55176/2414-1038-2021-1-135-144","DOIUrl":"https://doi.org/10.55176/2414-1038-2021-1-135-144","url":null,"abstract":"Gas-lift probe is an element of cladding failure detection system of perspective lead cooled reactor. Its function is local measurement of gaseous fission product activity in the coolant and the most defected fuel assembly localization. In the coolant leaving the defected fuel assembly the specific activity of gaseous fission products is higher than the average one in the primary circuit. In the barbotage channel of gas-lift probe gaseous fission products diffuse through the bubble interface surface into the volume of the bubbles. The bubbles deliver gaseous fission product to interface surface in the separation volume. The gas enriched with radioactive gaseous fission product goes to measurement volume of the probe. The more significant the damage and the closer the defective fuel assembly is located to the probe input, the more gaseous fission product activity will be registered. The paper presents a model of gaseous activity transfer from cladding defect to probe measuring volume. The gaseous activity transfer is described on the basis of the inert gases dissolution/degassing processes in lead. The gas-lift probe sensitivity was estimated as the ratio of the entry velocity of gaseous activity into the measurement volume to the exit one into the coolant through fuel assemblies cladding defects. A gas-lift probe sensitivity for exposed fuel surface calculated as an example. Gaseous fission products with significant gamma radiation are considered. The calculation results are presented in the article.","PeriodicalId":20426,"journal":{"name":"PROBLEMS OF ATOMIC SCIENCE AND TECHNOLOGY. SERIES: NUCLEAR AND REACTOR CONSTANTS","volume":null,"pages":null},"PeriodicalIF":0.0,"publicationDate":"2021-03-26","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"82055316","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 : 2021-03-26DOI: 10.55176/2414-1038-2021-1-27-35
E. Bogdanova, G. Tikhomirov, I. Suslov, Y. Homyakov
In the design and operation of nuclear power plants, one of the most important tasks is to assess the radiation protection of the reactor. Currently, the most widespread are deterministic (method of discrete ordinates) and stochastic computational methods for evaluating functionals. At large attenuations of the neutron flux (by 5-15 orders of magnitude) the deep penetration problems require large computational costs. The most accurate simulation of radiation transfer is achieved by using precision programs that implement the Monte Carlo method with a continuous energy dependence of the cross sections. A detailed description of the geometry and the use of continuous cross sections for particle interactions in calculations lead to high computational costs. To improve computational efficiency, there are variance reduction techniques (non-analog modeling). In this paper the possibility of using non-analog modeling in MCU-FR program by calculating the protection of the fast reactor full-scale model with a lead coolant is considered. The volume-integral neutron fluxes were estimated at points located in a long distance from the center of the reactor core. Analysis results were shown the significant reduction of the variance in the reactor shielding by using the non-analog Monte Carlo method.
{"title":"APPLICATION OF NON-ANALOG MONTE CARLO MODELING IN SHIELDING CALCULATIONS OF FAST REACTORS","authors":"E. Bogdanova, G. Tikhomirov, I. Suslov, Y. Homyakov","doi":"10.55176/2414-1038-2021-1-27-35","DOIUrl":"https://doi.org/10.55176/2414-1038-2021-1-27-35","url":null,"abstract":"In the design and operation of nuclear power plants, one of the most important tasks is to assess the radiation protection of the reactor. Currently, the most widespread are deterministic (method of discrete ordinates) and stochastic computational methods for evaluating functionals. At large attenuations of the neutron flux (by 5-15 orders of magnitude) the deep penetration problems require large computational costs. The most accurate simulation of radiation transfer is achieved by using precision programs that implement the Monte Carlo method with a continuous energy dependence of the cross sections. A detailed description of the geometry and the use of continuous cross sections for particle interactions in calculations lead to high computational costs. To improve computational efficiency, there are variance reduction techniques (non-analog modeling). In this paper the possibility of using non-analog modeling in MCU-FR program by calculating the protection of the fast reactor full-scale model with a lead coolant is considered. The volume-integral neutron fluxes were estimated at points located in a long distance from the center of the reactor core. Analysis results were shown the significant reduction of the variance in the reactor shielding by using the non-analog Monte Carlo method.","PeriodicalId":20426,"journal":{"name":"PROBLEMS OF ATOMIC SCIENCE AND TECHNOLOGY. SERIES: NUCLEAR AND REACTOR CONSTANTS","volume":null,"pages":null},"PeriodicalIF":0.0,"publicationDate":"2021-03-26","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"76113192","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 : 2021-03-26DOI: 10.55176/2414-1038-2021-1-5-14
A. Vaivod, S. Besov, A. Yudov, S. Andreyev
The article presents the investigation into reactivity and critical parameters affected by temperature changes in cylindrical and spherical space-effective multiplying systems (MS) of plutonium in α-phase without reflector during integral critical experiments. The experiments were performed in “Zababakhin All-Russia Research Institute of Technical Physics” on a critical assembly stand FKBN-2. Brief description of the experimental set up is provided. During the experiments, the multiplication factors and decay constant of the prompt neutrons were determined at a fixed gap between MS parts and different MS temperatures. Values of MS critical gap at different MS temperatures were established experimentally. The change in MS reactivity is initiated by energy release under the plutonium isotopes alpha-decay. As a result, the change in MS temperature during the experiments without its forced regulation may reach ~40 °C. It is proposed to consider this temperature effect when specifying experimental results including error estimation. The obtained experimental data were used to estimate the temperature coefficients of reactivity. The work results are applied to create benchmark models of the multiplying systems.
{"title":"EXPERIMENTS ON INVESTIGATION INTO PLUTONIUM TEMPERATURE EFFECT IN PLUTONIUM MULTIPLYING SYSTEMS WITHOUT REFLECTOR","authors":"A. Vaivod, S. Besov, A. Yudov, S. Andreyev","doi":"10.55176/2414-1038-2021-1-5-14","DOIUrl":"https://doi.org/10.55176/2414-1038-2021-1-5-14","url":null,"abstract":"The article presents the investigation into reactivity and critical parameters affected by temperature changes in cylindrical and spherical space-effective multiplying systems (MS) of plutonium in α-phase without reflector during integral critical experiments. The experiments were performed in “Zababakhin All-Russia Research Institute of Technical Physics” on a critical assembly stand FKBN-2. Brief description of the experimental set up is provided. During the experiments, the multiplication factors and decay constant of the prompt neutrons were determined at a fixed gap between MS parts and different MS temperatures. Values of MS critical gap at different MS temperatures were established experimentally. The change in MS reactivity is initiated by energy release under the plutonium isotopes alpha-decay. As a result, the change in MS temperature during the experiments without its forced regulation may reach ~40 °C. It is proposed to consider this temperature effect when specifying experimental results including error estimation. The obtained experimental data were used to estimate the temperature coefficients of reactivity. The work results are applied to create benchmark models of the multiplying systems.","PeriodicalId":20426,"journal":{"name":"PROBLEMS OF ATOMIC SCIENCE AND TECHNOLOGY. SERIES: NUCLEAR AND REACTOR CONSTANTS","volume":null,"pages":null},"PeriodicalIF":0.0,"publicationDate":"2021-03-26","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"87179915","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 : 2021-03-26DOI: 10.55176/2414-1038-2021-1-108-116
A. Kazantsev, O. Supotnitskaya, Vladimir N. Sergeev
The article presents the results of simulating the jet-vortex condenser operation. The jet-vortex condenser used as a part of the confinement system is designed to ensure the confinement integrity of an NPP with VVER-440 unit during LOCA accidents. To simulate the jet-vortex condenser operation it is important to keep a sufficient amount of water after a pressure peak was reached in the jet-vortex condenser hydraulic lock. The loss of water due to drop entrainment with steam-air mixture flow into the atmosphere stops when the velocity of drop sedimentation becomes higher than the velocity of drop entrainment with the mixture flow. The jet-vortex condenser model integrated into the KUPOL-M code was validated against the experimental data obtained on the VNIIAES test facility. To take into account drop entrainment with steam-air mixture flow the procedure of moisture separation and drop entrainment was used. A good agreement between calculated and experimental results was obtained when comparing the initial and final water levels in the hydraulic lock. The research results confirmed validity of the model of drop entrainment with steam-air mixture flow during the operation of the jet-vortex condenser and the preservation of water into the hydraulic lock during the accident.
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