Influence of voltage fluctuations on core vibration of a UHV shunt reactor

2021 ◽  
pp. 1-21
Author(s):  
Jiayi Guo ◽  
Jianghai Geng ◽  
Fangcheng Lü
Keyword(s):  
Power Grid ◽  
Reactor Core ◽  
Operating Conditions ◽  
Operating Voltage ◽  
Voltage Fluctuations ◽  
Shunt Reactor ◽  
The Core ◽  
Nodal Voltage ◽  
Vibration Parameters ◽  
Shunt Reactors

A new core vibration calculation method of UHV shunt reactors is proposed to deal with the problem that vibration parameters of the core of an ultra-high voltage (UHV) shunt reactor cannot be measured on site. A series of tests have validated that the method can calculate the vibration parameters of a UHV shunt reactor core with different voltage ratios. The method is adopted to study the influence of nodal voltage fluctuations of the UHV AC tie line on the core vibration characteristics of a UHV shunt reactor under normal and abnormal operating conditions. The conclusion is drawn as follows: in the three operating states assessed herein, no matter whether the power grid is in a normal state or not, the core of the UHV shunt reactor will not reach magnetic saturation, and the vibration parameters of the reactor always maintain a linear relationship with the operating voltage. The changes of power grid operation parameters are introduced to the research into reactor vibration, which is conducive to a more comprehensive understanding of the actual operating state of a UHV shunt reactor. It can also provide help for the design and operation of UHV shunt reactors.

Shunt Reactors Compensation Research of UHV Accessed to Jiangxi Power Grid

2014 ◽  
Vol 1070-1072 ◽  
pp. 1029-1034
Author(s):  
Yong Gao Zhang ◽  
Kai Xuan Chang ◽  
Yong Chun Su
Keyword(s):  
Steady State ◽  
High Voltage ◽  
Transfer Equation ◽  
Power Grid ◽  
Distributed Parameter ◽  
Shunt Reactor ◽  
Power Frequency ◽  
Shunt Reactors ◽  
Relationship Of ◽  
The Relationship

Jiangxi ultra high voltage (UHV) accessed program is introduced. Based on uniform transfer equation, UHV π-type equivalent circuit is derived by considering the distributed parameter characteristics of UHV transmission line without conductance. Based on advanced digital power system simulator, the relationship of shunt reactor and steady-state voltage, as well as the relationship of shunt reactor and transient voltage, is achieved by using electromechanical and electromagnetic hybrid simulation, where the UHV between Hubei and Jiangxi is considered. Influences of shunt reactors’ location and the degree of compensation on UHV’s steady-state voltage and transient power frequency overvoltage are analyzed. The Logical results provide a meaningful guidance for the introduction of UHV power grid into Jiangxi power grid.

3D FEA Elastoplastic Structural Analysis of Hypothetically Cracked Monju FBR Core Support Structures

2014 ◽  
Author(s):  
Daniel Garcia-Rodriguez ◽  
Shinichiro Matsubara
Keyword(s):  
Structural Reliability ◽  
Reactor Core ◽  
Operating Conditions ◽  
Severe Accident ◽  
Support Structures ◽  
Element Analysis ◽  
The Core ◽  
Fea Model ◽  
3D Fea ◽  
Accident Conditions

In this work the structural reliability of the circumferentially cracked core support mount of Monju Fast Breeder Reactor (FBR) is analyzed using Finite Element Analysis (FEA). The 3D shell model employed was derived after detailed evaluation of the core support mount behavior with a specific 3D solid model. First, elastoplastic static analysis results show that, under nominal operating conditions, the overall structure would be able to survive a total loss of the core support mount. Second, using the double elastic slope method it was inferred that earthquake loading integrity could be warranted up to a crack representing more than 50% of the total circumference. Both results highlight the ample primary loading margins taken in the design of Monju’s reactor core support structures. Furthermore, the developed 3D shell FEA model will be applied to study other extreme cases such as those under severe accident conditions.

scholarly journals Fabrication and Testing of a Modular Micro-Pocket Fission Detector Instrumentation System for Test Nuclear Reactors

2018 ◽  
Vol 170 ◽  
pp. 04018
Author(s):  
Michael A. Reichenberger ◽  
Daniel M. Nichols ◽  
Sarah R. Stevenson ◽  
Tanner M. Swope ◽  
Caden W. Hilger ◽  
...  
Keyword(s):  
Neutron Flux ◽  
Nuclear Reactor ◽  
Modular Design ◽  
Reactor Core ◽  
Operating Conditions ◽  
Sensor Nodes ◽  
Detector Array ◽  
State University ◽  
The Core ◽  
Kansas State University

Advancements in nuclear reactor core modeling and computational capability have encouraged further development of in-core neutron sensors. Measurement of the neutron-flux distribution within the reactor core provides a more complete understanding of the operating conditions in the reactor than typical ex-core sensors. Micro-Pocket Fission Detectors have been developed and tested previously but have been limited to single-node operation and have utilized highly specialized designs. The development of a widely deployable, multi-node Micro-Pocket Fission Detector assembly will enhance nuclear research capabilities. A modular, four-node Micro-Pocket Fission Detector array was designed, fabricated, and tested at Kansas State University. The array was constructed from materials that do not significantly perturb the neutron flux in the reactor core. All four sensor nodes were equally spaced axially in the array to span the fuel-region of the reactor core. The array was filled with neon gas, serving as an ionization medium in the small cavities of the Micro-Pocket Fission Detectors. The modular design of the instrument facilitates the testing and deployment of numerous sensor arrays. The unified design drastically improved device ruggedness and simplified construction from previous designs. Five 8-mm penetrations in the upper grid plate of the Kansas State University TRIGA Mk. II research nuclear reactor were utilized to deploy the array between fuel elements in the core. The Micro-Pocket Fission Detector array was coupled to an electronic support system which has been specially developed to support pulse-mode operation. The Micro-Pocket Fission Detector array composed of four sensors was used to monitor local neutron flux at a constant reactor power of 100 kWth at different axial locations simultaneously. The array was positioned at five different radial locations within the core to emulate the deployment of multiple arrays and develop a 2-dimensional measurement of neutron flux in the reactor core.

Design of Reverse Flow Blockage Device for Reactor Coolant Pumps

2016 ◽  
Vol 852 ◽  
pp. 489-497
Author(s):  
S. Aravindan ◽  
Bhagwana Ram Manda ◽  
K.V. Sreedharan ◽  
S. Athmalingam ◽  
V. Balasubramaniyan
Keyword(s):  
Cylindrical Shell ◽  
Reverse Flow ◽  
Reactor Core ◽  
Operating Conditions ◽  
Coolant Fluid ◽  
Reactor Coolant ◽  
The Core ◽  
Time Operation ◽  
Reactor Operating ◽  
Tie Rods

Future Sodium cooled Fast Reactors (SFR) in India, are being designed with 3 Reactor Coolant Pumps (RCP). Reactor Coolant Pumps (RCP) are used in pumping the coolant fluid through the reactor core. As the RCPs are operating in parallel, failure of one pump will result in a significant portion of the pumped coolant to bypass the core via tripped RCP. Thus, the reactor has to be shut down and the tripped RCP has to be replaced, which results in down time. Operation of reactor with only two RCP (2/3 mode) will supplement the power generation. Henceforth, if the flow path through the tripped RCP is made highly resistant as compared to the core, then eventually, flow from the two operating RCPs will go through reactor core. Such a flow blocking/resisting arrangement can also provide the flexibility during reactor startup. A cylindrical shell is designed for closing the suction passage of the pump and the shell is raised or lowered from above the roof slab with the help of tie-rods. The cylindrical shell (sleeve shell) is designed to withstand water hammer shock due to inadvertent sudden closure of the suction passage. To determine the shell thickness, pump performance curves are developed based on reactor operating conditions. The sleeve shell will not be able to perfectly close the suction passage as space is required for movement of sleeve shell over the pump shell, thus a study is performed on incorporating a labyrinth to minimize the leakage.

scholarly journals Core Stress Analysis of Amorphous Alloy Transformer for Rail Transit under Different Working Conditions

Energies ◽  
2020 ◽  
Vol 14 (1) ◽  
pp. 164
Author(s):  
Jianwei Shao ◽  
Cuidong Xu ◽  
Ka Wai Eric Cheng
Keyword(s):  
Amorphous Alloy ◽  
Working Conditions ◽  
Short Circuit ◽  
Element Model ◽  
Operating Conditions ◽  
Iron Core ◽  
Rail Transit ◽  
Transit System ◽  
The Core ◽  
Distribution Transformers

The rail transit system is a large electric vehicle system that is strongly dependent on the energy technologies of the power system. The use of new energy-saving amorphous alloy transformers can not only reduce the loss of rail transit power, but also help alleviate the power shortage situation and electromagnetic emissions. The application of the transformer in the field of rail transit is limited by the problem that amorphous alloy is prone to debris. this paper studied the stress conditions of amorphous alloy transformer cores under different working conditions and determined that the location where the core is prone to fragmentation, which is the key problem of smoothly integrating amorphous alloy distribution transformers on rail transit power supply systems. In this study, we investigate the changes in the electromagnetic field and stress of the amorphous alloy transformer core under different operating conditions. The finite element model of an amorphous alloy transformer is established and verified. The simulation results of the magnetic field and stress of the core under different working conditions are given. The no-load current and no-load loss are simulated and compared with the actual experimental data to verify practicability of amorphous alloy transformers. The biggest influence on the iron core is the overload state and the maximum value is higher than the core stress during short circuit. The core strain caused by the side-phase short circuit is larger than the middle-phase short circuit.

Calculation studies of coated particles performance in sodium-cooled fast reactor

Kerntechnik ◽  
2021 ◽  
Vol 86 (1) ◽  
pp. 45-49
Author(s):  
N. V. Maslov ◽  
E. I. Grishanin ◽  
P. N. Alekseev
Keyword(s):  
Fast Reactor ◽  
Reactor Core ◽  
Heat Removal ◽  
Design Solution ◽  
Fast Neutrons ◽  
Steel Shell ◽  
Primary Circuit ◽  
Coated Particles ◽  
The Core ◽  
Fuel Assemblies

Abstract This paper presents results of calculation studies of the viability of coated particles in the conditions of the reactor core on fast neutrons with sodium cooling, justifying the development of the concept of the reactor BN with microspherical fuel. Traditional rod fuel assemblies with pellet MOX fuel in the core of a fast sodium reactor are directly replaced by fuel assemblies with micro-spherical mixed (U,Pu)C-fuel. Due to the fact that the micro-spherical (U, Pu)C fuel has a developed heat removal surface and that the design solution for the fuel assembly with coated particles is horizontal cooling of the microspherical fuel, the core has additional possibilities of increasing inherent (passive) safety and improve the competitiveness of BN type of reactors. It is obvious from obtained results that the microspherical (U, Pu)C fuel is limited with the maximal burn-up depth of ∼11% of heavy atoms in conditions of the sodium-cooled fast reactor core at the conservative approach; it gives the possibility of reaching stated thermal-hydraulic and neutron-physical characteristics. Such a tolerant fuel makes it less likely that fission products will enter the primary circuit in case of accidents with loss of coolant and the introduction of positive reactivity, since the coating of microspherical fuel withstands higher temperatures than the steel shell of traditional rod-type fuel elements.

scholarly journals Financial Credit Risk Evaluation Based on Core Enterprise Supply Chains

2018 ◽  
Vol 10 (10) ◽  
pp. 3699 ◽  
Author(s):  
WeiMing Mou ◽  
Wing-Keung Wong ◽  
Michael McAleer
Keyword(s):  
Supply Chain ◽  
Credit Risk ◽  
Supply Chains ◽  
Risk Evaluation ◽  
Operating Conditions ◽  
The Core ◽  
Supply Chain Finance ◽  
Enterprise Credit ◽  
Financial Credit ◽  
Core Enterprise

Supply chain finance has broken through traditional credit modes and advanced rapidly as a creative financial business discipline. Core enterprises have played a critical role in the credit enhancement of supply chain finance. Through the analysis of core enterprise credit risks in supply chain finance, by means of a ‘fuzzy analytical hierarchy process’ (FAHP), the paper constructs a supply chain financial credit risk evaluation system, making quantitative measurements and evaluation of core enterprise credit risk. This enables enterprises to take measures to control credit risk, thereby promoting the healthy development of supply chain finance. The examination of core enterprise supply chains suggests that a unified information file should be collected based on the core enterprise, including the operating conditions, asset status, industry status, credit record, effective information to the database, collecting related data upstream and downstream of the archives around the core enterprise, developing a data information system, electronic data information, and updating the database accurately using the latest information that might be available. Moreover, supply chain finance and modern information technology should be integrated to establish the sharing of information resources and realize the exchange of information flows, capital flows, and logistics between banks. This should reduce a variety of risks and improve the efficiency and effectiveness of supply chain finance.

scholarly journals Research on Influence of Different Simulation Methods of Bypass Flow in Thermal Hydraulic Analysis on Temperature Distribution in HTR-10

2020 ◽  
Vol 2020 ◽  
pp. 1-8
Author(s):  
Shiyan Sun ◽  
Youjie Zhang ◽  
Yanhua Zheng
Keyword(s):  
Temperature Distribution ◽  
Reactor Core ◽  
Flow Distribution ◽  
Maximum Temperature ◽  
Flow Ratio ◽  
Hydraulic Analysis ◽  
Bypass Flow ◽  
Pebble Bed ◽  
The Core ◽  
Thermal Hydraulic Analysis

In pebble-bed high temperature gas-cooled reactor, gaps widely exist between graphite blocks and carbon bricks in the reactor core vessel. The bypass helium flowing through the gaps affects the flow distribution of the core and weakens the effective cooling of the core by helium, which in turn affects the temperature distribution and the safety features of the reactor. In this paper, the thermal hydraulic analysis models of HTR-10 with bypass flow channels simulated at different positions are designed based on the flow distribution scheme of the original core models and combined with the actual position of the core bypass flow. The results show that the bypass coolant flowing through the reflectors enhances the heat transfer of the nearby components efficiently. The temperature of the side reflectors and the carbon bricks is much lower with more side bypass coolant. The temperature distribution of the central region in the pebble bed is affected by the bypass flow positions slightly, while that of the peripheral area is affected significantly. The maximum temperature of the helium, the surface, and center of the fuel elements rises as the bypass flow ratio becomes larger, while the temperature difference between them almost keeps constant. When the flow ratio of each part keeps constant, the maximum temperature almost does not change with different bypass flow positions.

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