For traditional Direct Power Control (DPC), the switching frequency is not fixed and the characteristics in steady state are not well. In order to solve these problems, a proposed DPC strategy based on virtual flux is...
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Aimed to problems that exist in the regeneration energy feedback technology of the mine hoist, a novel dual-PWM variable frequency control system was proposed for three-phase AC motor in this paper. Compared to the un...
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As the important parts of the fuel assemblies of Pressurized Water reactors (PWR), the presence of spacer grids promotes the local heat transfer adjacent to the wall downstream the grids by causing cross and swirl flo...
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ISBN:
(纸本)9781629937212
As the important parts of the fuel assemblies of Pressurized Water reactors (PWR), the presence of spacer grids promotes the local heat transfer adjacent to the wall downstream the grids by causing cross and swirl flow between and within sub-channels, which greatly widen the thermal hydraulic safety margin. Recent years, with tremendous development of computer technologies and numerical simulations, Computational Fluid Dynamics codes (CFD) are widely adapted to the designs and selections of spacer grid. This paper presents a CFD methodology to evaluate two-phase mixing characteristics on PWR fuel rod bundle with spacer grids. Numerical simulations on air and water flow were carried out with CFX 12.0 for two different patterns of spacer grids. In addition, they were selected for the critical heat flux (CHF) tests with FREON in Nuclear Power Institute of China (NPIC). Comparisons of numerical researches and experimental results show that the CFD methodology to predict bubble congregating tendency adjacent to the surface of fuel rod was validated.
The paper presents an innovative design for a wheeled hopping robot. Our main objective was to obtain a stable and efficient robot, which can jump over obstacles and has the ability to reach a certain ledge. The innov...
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The paper presents an innovative design for a wheeled hopping robot. Our main objective was to obtain a stable and efficient robot, which can jump over obstacles and has the ability to reach a certain ledge. The innovative design of our robot has the capability to jump to a certain distance even with a zero initial velocity provided by the wheels movement. Also, we took into consideration the wind speed which can prevent our robot to reach the desired destination, and compensated with an initial velocity on the direction of movement. For testing we designed a simulation which had as inputs different initial conditions to test and present the jumping capability of our robot in different jumping conditions. In the end we provide a jump area which the robot can reach for certain initial conditions, so we can later chose the optimal one for reaching the target position.
On the premise of safety, economy is one of the important objectives of nuclear power plants. For VVER-1000, there are several cases of operation modes when the plant is off the grid, such as the critical operation af...
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On the premise of safety, economy is one of the important objectives of nuclear power plants. For VVER-1000, there are several cases of operation modes when the plant is off the grid, such as the critical operation after refueling, the hot shutdown and critical operation, and the critical operation after test. For the purpose to reduce the time of the operations, the consumption of materials, and the amount of waste water produced, the optimization researches are carried out on these cases. The qualitative and quantitative analysis is carried out, on the configurations of the control rod and boric acid, which are the important factors to influence the shutdown and critical operations. The general steps and the basic principles of the optimizations are described. Three cases are optimized.
As a key equipment,reactor Vessel Internals influence the safety of nuclear power *** is a rigid requirement to stock of reactor Vessel Internals because of the high temperature,high pressure and high dose *** the sta...
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As a key equipment,reactor Vessel Internals influence the safety of nuclear power *** is a rigid requirement to stock of reactor Vessel Internals because of the high temperature,high pressure and high dose *** the stainless steel usually used in different nations,the paper analyze the corrosion resistance property,weldability and irradiation property and recommend the optimum stock option.
In recent years many researchers are trying to combine wheeled movement with legged movement or hopping movement to strengthen the robots adaptability to environment. Hybrid robots are playing a more and more importan...
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In recent years many researchers are trying to combine wheeled movement with legged movement or hopping movement to strengthen the robots adaptability to environment. Hybrid robots are playing a more and more important role in our modern life. However, their controls are mostly complicated. So in this paper we present a new solution for the wheeled hopping robot. Based on the mechanical design, this robot has two movement modes: wheeled and hopping movement mode. While the center of gravity of the mechanism is under the wheels center, the wheeled hopping robot which is like a tumbler can keep its stability even when the wheels' velocity is zero. In this paper, we firstly introduce the description and design of the wheeled hopping robot. Then we explain how to choose the robot parameters and lastly we describe, how the robot through hopping, successfully traverses obstacles and ditches.
For nuclear power plant, station black- out (SBO) is the events that contribute significantly to the level-I core damage risk. For an SBO, it is assumed that both the off-site power and on-site diesel generators fail ...
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CSR1000(China Supercritical water-cooled reactor with the rated electric power of 1000 MWe) put forward by NPIC (Nuclear Power Institute of China) proposed a new design of fuel assembly cluster with two-pass core arra...
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ISBN:
(纸本)9781629937212
CSR1000(China Supercritical water-cooled reactor with the rated electric power of 1000 MWe) put forward by NPIC (Nuclear Power Institute of China) proposed a new design of fuel assembly cluster with two-pass core arrangement. In this paper, subchannel analysis code ATHAS which has added new heat transfer and friction coefficient correlations obtained from experiment on T/H characteristics of SCW carried out by NPIC has been used to analyze the thermal-hydraulics characteristics of CSR1000. The subchannel analysis results have shown that the design basically meets the requirements of thermal-hydraulics design and remains some margins for safety analysis.
The failure issues of power insulated gate bipolar transistor (IGBT) modules are mainly related to thermal and ther mo-mechanical aging mechanism. This aging causes degradation of the device performance and faults whi...
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