The Experts below are selected from a list of 8535 Experts worldwide ranked by ideXlab platform
R. R. Khayrutdinov - One of the best experts on this subject based on the ideXlab platform.
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Concept of a demonstrational hybrid reactor—a tokamak with molten-salt blanket for ^233U fuel production: 1. Concept of a stationary Tokamak as a neutron source
Physics of Atomic Nuclei, 2015Co-Authors: E. A. Azizov, G. G. Gladush, V. N. Dokuka, R. R. KhayrutdinovAbstract:On the basis of current understanding of physical processes in tokamaks and taking into account engineering constraints, it is shown that a low-cost facility of a moderate size can be designed within the adopted concept. This facility makes it possible to achieve the power density of neutron flux which is of interest, in particular, for solving the problem of ^233U fuel production from thorium. By using a molten-salt blanket, the important task of ensuring the safe operation of such a reactor in the case of possible Coolant Loss is accomplished. Moreover, in a hybrid reactor with the blanket based on liquid salts, the problem of periodic refueling that is difficult to perform in solid blankets can be solved.
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Concept of a demonstrational hybrid reactor—a tokamak with molten-salt blanket for 233 U fuel production: 1. Concept of a stationary Tokamak as a neutron source
Physics of Atomic Nuclei, 2015Co-Authors: E. A. Azizov, G. G. Gladush, V. N. Dokuka, R. R. KhayrutdinovAbstract:On the basis of current understanding of physical processes in tokamaks and taking into account engineering constraints, it is shown that a low-cost facility of a moderate size can be designed within the adopted concept. This facility makes it possible to achieve the power density of neutron flux which is of interest, in particular, for solving the problem of 233U fuel production from thorium. By using a molten-salt blanket, the important task of ensuring the safe operation of such a reactor in the case of possible Coolant Loss is accomplished. Moreover, in a hybrid reactor with the blanket based on liquid salts, the problem of periodic refueling that is difficult to perform in solid blankets can be solved.
Heng Xie - One of the best experts on this subject based on the ideXlab platform.
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A Simulation of Small Break Loss of Coolant Accident in Nuclear Heating Reactor Based on RELAP5
Volume 6A: Thermal-Hydraulics and Safety Analyses, 2018Co-Authors: Heng XieAbstract:Nuclear heating reactor is integrated designed without main pump and safety injection system. The Loss of Coolant accidents are mainly in the form of small break LOCA. As no safety injection system is designed for Coolant makeup, the water volume in the reactor vessel is critical since it determines whether the reactor will be submerged during the whole scenario. Therefore, the study on Coolant Loss in this pool system is indispensable. The RELAP5 code has been developed for best-estimate transient simulation of light water reactor Coolant systems during postulated accidents. The long term effect in nuclear heating reactor is important. In this paper we investigated the influential factors on SBLOCA scenario and found the long term residual heat removal capacity is decisive in determining the Loss of Coolant. The residual heat removal capacity should be greater than 2% of reactor thermal power if ensuring the core submerged in the long run.
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A Numerical Study of Liquid Carryover Based on the VOF Model
Volume 8: Computational Fluid Dynamics (CFD) and Coupled Codes; Nuclear Education Public Acceptance and Related Issues, 2017Co-Authors: Heng XieAbstract:Liquid carryover is an important phenomenon during the small break Loss of Coolant accidents (LOCA). The Coolant Loss rate due to liquid entrained decides the decreasing height of water level and the ability that the core can be cooled. It is especially important in nuclear heating reactor as it is integral arranged without safety injection system. Therefore, the initial water volume in the reactor vessel and Coolant Loss rate determines whether the core could be submerged. Numerous experiments have been conducted to investigate the phenomenon of liquid carryover in the pool system. Fruitful outcomes have been proposed involving mechanism of entrained liquid produced as well as semi-empirical formula for engineering purposes. However, the semi-empirical correlations on entrainment are highly data-depended and are less applicable under different experiment settings. This paper presents a numerical method based on the VOF to study the liquid carryover in the pool system. The numerical results are analyzed with mesh independence analysis and compared with previous experiments. Several types of liquid carryover are found in the simulation. The exit effect on liquid Loss rate is specifically studied including the setting of exit as well as the range that exit effect dominates.
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An investigation of pool entrainment based on the method of Volume of Fluid
Nuclear Engineering and Design, 2017Co-Authors: Heng XieAbstract:Abstract Liquid entrainment in pool system is a universal phenomenon that has been studied for many years. Many experiments have been conducted to investigate the Coolant Loss rate during the scenario of small break Loss of Coolant accident (SBLOCA). However, these studies are limited by specific experiment settings whose results cannot be directly widely used. In this paper, a numerical simulation based on the Volume of Fluid (VOF) method is performed and the transient calculation is based on the Pressure Implicit with Splitting of Operator (PISO) algorithm. The averaged entrainment rate is calculated by monitoring the liquid mass Loss during a period of time. Meanwhile, the two-phase mixture level is determined by monitoring the averaged density variation along the height within a span of time. The numerical result shows a good tendency with the experiment data. Several types of entrainment are demonstrated to show different ways of entrained liquid produced under different superficial air velocities. Moreover, exit effect is non-negligible factor that dominates whether the entrained liquid could be eventually carried out or not. A theoretical formula is proposed to predict the region where exit effect dominates, which agrees well with the numerical results.
E. A. Azizov - One of the best experts on this subject based on the ideXlab platform.
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Concept of a demonstrational hybrid reactor—a tokamak with molten-salt blanket for ^233U fuel production: 1. Concept of a stationary Tokamak as a neutron source
Physics of Atomic Nuclei, 2015Co-Authors: E. A. Azizov, G. G. Gladush, V. N. Dokuka, R. R. KhayrutdinovAbstract:On the basis of current understanding of physical processes in tokamaks and taking into account engineering constraints, it is shown that a low-cost facility of a moderate size can be designed within the adopted concept. This facility makes it possible to achieve the power density of neutron flux which is of interest, in particular, for solving the problem of ^233U fuel production from thorium. By using a molten-salt blanket, the important task of ensuring the safe operation of such a reactor in the case of possible Coolant Loss is accomplished. Moreover, in a hybrid reactor with the blanket based on liquid salts, the problem of periodic refueling that is difficult to perform in solid blankets can be solved.
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Concept of a demonstrational hybrid reactor—a tokamak with molten-salt blanket for 233 U fuel production: 1. Concept of a stationary Tokamak as a neutron source
Physics of Atomic Nuclei, 2015Co-Authors: E. A. Azizov, G. G. Gladush, V. N. Dokuka, R. R. KhayrutdinovAbstract:On the basis of current understanding of physical processes in tokamaks and taking into account engineering constraints, it is shown that a low-cost facility of a moderate size can be designed within the adopted concept. This facility makes it possible to achieve the power density of neutron flux which is of interest, in particular, for solving the problem of 233U fuel production from thorium. By using a molten-salt blanket, the important task of ensuring the safe operation of such a reactor in the case of possible Coolant Loss is accomplished. Moreover, in a hybrid reactor with the blanket based on liquid salts, the problem of periodic refueling that is difficult to perform in solid blankets can be solved.
Iooss Ertrand - One of the best experts on this subject based on the ideXlab platform.
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Statistical identification of penalizing configurations in high-dimensional thermalhydraulic numerical experiments: The ICSCREAM methodology
HAL CCSD, 2020Co-Authors: Marrel A., Iooss ErtrandAbstract:In the framework of risk assessment in nuclear accident analysis, best-estimate computer codes are used to estimate safety margins. Several inputs of the code can be uncertain, due to a lack of knowledge but also to the particular choice of accidental scenario being considered. The objective of this work is to identify the most penalizing (or critical) configurations (corresponding to extreme values of the code output) of several input parameters (called ``scenario inputs''), independently of the uncertainty of the other input parameters. However, complex computer codes, as the ones used in thermal-hydraulic accident scenario simulations, are often too CPU-time expensive to be directly used to perform these studies. A solution consists in fitting the code output by a metamodel, built from a reduced number of code simulations. When the number of input parameters is very large (e.g., around a hundred here), the metamodel building remains a challenge. To overcome this, we propose a methodology, called ICSCREAM (Identification of penalizing Configurations using SCREening And Metamodel), based on screening techniques and Gaussian process (Gp) metamodeling. The efficiency of this methodology is illustrated on a thermal-hydraulic industrial case simulating an accident of primary Coolant Loss in a Pressurized Water Reactor. This use-case includes 97 uncertain inputs, two scenario inputs to be penalized and 500 code simulations for the learning database. The study focuses on the peak cladding temperature (PCT) and critical configurations are defined by exceeding the 90%-quantile of PCT.For the screening step, statistical tests of independence based on the Hilbert-Schmidt independence criterion are used for global and target sensitivity analyses. They allow a significant reduction of inputs (from 97 to 20) and a ranking of these influential inputs by order of influence. Then, a Gp metamodel is sequentially built to reach a satisfactory predictivity of 82% of explained PTC variance, and a high capacity of identifying PTC critical areas (94% of good ranking rate above the threshold). Finally, the Gp is used to estimate, within a Bayesian framework, the conditional probabilities of exceeding the threshold, according to the two scenario inputs. The analysis reveals the strong interaction of the two scenario inputs in the occurrence of critical configurations, worst cases corresponding to medium values of both inputs
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STATISTICAL IDENTIFICATION OF PENALIZING CONFIGURATIONS IN HIGH-DIMENSIONAL THERMAL-HYDRAULIC NUMERICAL EXPERIMENTS: THE ICSCREAM METHODOLOGY
HAL CCSD, 2020Co-Authors: Marrel A., Iooss ErtrandAbstract:In the framework of risk assessment in nuclear accident analysis, best-estimate computer codes are used to estimate safety margins. Several inputs of the code can be uncertain, due to a lack of knowledge but also to the particular choice of accidental scenario being considered. The objective of this work is to identify the most penalizing (or critical) configurations (corresponding to extreme values of the code output) of several input parameters (called "scenario inputs"), independently of the uncertainty of the other input parameters. However, complex computer codes, as the ones used in thermal-hydraulic accident scenario simulations, are often too CPU-time expensive to be directly used to perform these studies. A solution consists in fitting the code output by a metamodel, built from a reduced number of code simulations. When the number of input parameters is very large (around a hundred here), the metamodel building remains a challenge. To overcome this, we propose a methodology, called ICSCREAM (Identification of penalizing Configurations using SCREening And Metamodel), based on screening techniques and Gaussian process (Gp) metamodeling. The Gp metamodel is used to estimate, within a Bayesian framework, the conditional probabilities of exceeding a critical value, according to the scenario inputs. Critical configurations of these inputs are then identified. The efficiency of this methodology is illustrated on a thermal-hydraulic industrial case simulating an accident of primary Coolant Loss in a Pressurized Water Reactor with 97 uncertain inputs and two scenario inputs to be penalized
Li Weitong - One of the best experts on this subject based on the ideXlab platform.
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Experimental and numerical investigations on effect of reverse flow on transient from forced circulation to natural circulation
Nuclear Engineering and Technology, 2020Co-Authors: Wenzhen Chen, Jianli Hao, Li WeitongAbstract:Abstract In a sudden shutdown of primary pump or Coolant Loss accident in a marine nuclear power plant, the primary flow decreases rapidly in a transition process from forced circulation (FC) to natural circulation (NC), and the lower flow enters the steam generator (SG) causing reverse flow in the U-tube. This can significantly compromise the safety of nuclear power plants. Based on the marine natural circulation steam generator (NCSG), an experimental loop is constructed to study the characteristics of reverse flow under middle-temperature and middle-pressure conditions. The transition from FC to NC is simulated experimentally, and the characteristics of SG reverse flow are studied. On this basis, the experimental loop is numerically modeled using RELAP5/MOD3.3 code for system analysis, and the accuracy of the model is verified according to the experimental data. The influence of the flow variation rate on the reverse flow phenomenon and flow distribution is investigated. The experimental and numerical results show that in comparison with the case of adjusting the mass flow discontinuously, the number of reverse flow tubes increases significantly during the transition from FC to NC, and the reverse flow has a more severe impact on the operating characteristics of the SG. With the increase of flow variation rate, the reverse flow is less likely to occur. The mass flow in the reverse flow U-tubes increases at first and then decreases. When the system is approximately stable, the reverse flow is slightly lower than obverse flow in the same U-tube, while the flow in the obverse flow U-tube increases.