A Script Generator API for the Full-Scale Three-Dimensional Vibration Simulation of an Entire Nuclear Power Plant within AEGIS

2009 ◽  
Author(s):  
G. Kim ◽  
Y. Suzuki ◽  
N. Teshima ◽  
A. Nishida ◽  
T. Yamada ◽  
...  
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Three Dimensional ◽  
Full Scale ◽  
Vibration Simulation

Study on 3-D Flow Field of Concrete Volute Conduit of Circulating Water Pump in Nuclear Power Plant

2011 ◽  
Vol 99-100 ◽  
pp. 350-353
Author(s):  
Xiao Bing Sun ◽  
Xu Bin Qiao
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Stokes Equations ◽  
Three Dimensional ◽  
Gradual Change ◽  
Water Pump ◽  
Pump System ◽  
Circulating Water ◽  
Operating Modes

As the largest unit capacity of nuclear power plant at present, the flow conduit of circulating water pump in EPR1750 nuclear power plant is a volute conduit, which is a cast-in-situ conceret structure with complexly gradual change cavity. Therefore, the hydraulic efficiency of circulating water pump is not only related with the design of pump leaves, but also closely related to the design of volute and the complicated spatial type of intake and outtake conduits. With the pump leaves and the intake and outtake conduits of conceret volute as the research model, based on computational fluid dynamics (CFD)and the three dimensional Reynolds averaged Navier-Stokes equations, an analytic model suitable for computation is established to simulate the three-dimensional steady flow in the whole pumping system under different operating modes. By use of the commercial fluid-computation softer ANSYS, the distribution of basic physic quantities in the fluid field inside the pump and the conduits is obtained. The analysis and prediction of the performance of pump system are made, and the spatial type design of intake and outtake conduits is evaluated. The calculation results can be referenced to improve the design of pump systems in the similar projects.

scholarly journals Particle-laden flow around an obstacle in a square pipe: experiments and modeling

2020 ◽  
Vol 21 (5) ◽  
pp. 517
Author(s):  
Ouardia Ait Oucheggou ◽  
Véronique Pointeau ◽  
Guillaume Ricciardi ◽  
Élisabeth Guazzelli ◽  
Laurence Bergougnoux
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Steam Generator ◽  
Nuclear Power ◽  
Particle Deposition ◽  
Three Dimensional ◽  
Vertical Tube ◽  
Nuclear Industry ◽  
Solid Particles ◽  
Complex Nature

Particle trapping and deposition around an obstacle occur in many natural and industrial situations and in particular in the nuclear industry. In the steam generator of a nuclear power plant, the progressive obstruction of the flow due to particle deposition reduces the efficiency and can induce tube cracking leading to breaking and damage. The steam generator then loses its role as a safety barrier of the nuclear power plant. From a fundamental standpoint, dilute and concentrated particulate flows have received a growing attention in the last decade. In this study, we investigate the transport of solid particles around obstacles in a confined flow. Experiments were performed in a simplified configuration by considering a laminar flow in a vertical tube. An obstacle was inserted at the middle height of the tube and neutrally-buoyant particles were injected at different locations along the tube. We have investigated first the trajectories of individual particles using particle tracking (PT). Then, the particle trajectories were modeled by using the Boussinesq-Basset-Oseen equation with a flow velocity field either measured using particle image velocimetry (PIV) or calculated by the Code_Saturne software in order to account for the three-dimensional (3D) character of the obstacle wake. This paper presents a comparison between the experimental observations and the predictions of the modeling for an obstacle consisting of a rectangular step at a Reynolds number of ≈100 and evidences the importance of accounting for the 3D complex nature of the flow.

Three-Dimensional Profilometry Utilization on VVER Type Reactors

2020 ◽  
Vol 7 (2) ◽  
Author(s):  
Jaroslav Brom ◽  
Jan Patera ◽  
Pavel Zahrádka
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Three Dimensional ◽  
Laser Scanner ◽  
Primary Circuit ◽  
3D Profilometry ◽  
Circuit Components ◽  
3D Profile ◽  
Reactor Pressure

Abstract The paper is focused on the application of three-dimensional (3D) profilometry on water–water energetic reactor (VVER) type nuclear power plant (NPP) equipment. This method is becoming increasingly used in the power industry and replaces conventional methods such as micrometer measurements. One of the greatest benefits is the accurate recording of the 3D profile of the measured surface and the possibility of its comparison with the production documentation or with the results from previous measurements. Centrum Výkumu Řež, s.r.o. (Research Center Řež) uses 3D laser scanner with a measuring arm. This method was, for example, successfully used for reactor pressure vessel (RPV), steam generator (SG), and bolts. The results are used by the NPP operator for the lifetime management of the primary circuit components.

scholarly journals Full-Scale 3D Vibration Simulator of an Entire Nuclear Power Plant on Simple Orchestration Application Framework

2010 ◽  
pp. 93-106
Author(s):  
Guehee Kim ◽  
Kohei Nakajima ◽  
Takayuki Tatekawa ◽  
Naoya Teshima ◽  
Yoshio Suzuki ◽  
...  
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Full Scale ◽  
Application Framework

Development of Experimental and Computational Procedures for Nuclear Power Plant Component Testing Under Flooding Conditions

2017 ◽  
Author(s):  
Antonio Tahhan ◽  
Cody Muchmore ◽  
Larinda Nichols ◽  
Alison Wells ◽  
Gregory Roberts ◽  
...  
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Full Scale ◽  
Nuclear Industry ◽  
Small Scale ◽  
Wave Impact ◽  
Plant Component ◽  
Component Testing ◽  
Water Rise

Idaho State University (ISU), with support from Idaho National Laboratory, is actively engaged in enhancing nuclear power plant risk modeling. The ISU team is significantly increasing the understanding of non-containment, nuclear power plant component performance under flooding conditions. The work involves experimentation activities and development of mathematical models, using data from component flooding experiments. The research consists in developing experimentation procedures that comprised small scale component testing, followed by simple and then complex full scale component testing. The research is taking place in the Component Flooding Evaluation Laboratory (CFEL). Tests in CFEL will include water rise, spray, and wave impact experiments on passive and active components. Initial development work focused on small scale components, radios and simulated doors, that served as a low-risk and low-cost proof-of-concept options. Following these tests, full-scale component tests were performed in the Portal Evaluation Tank (PET). The PET is a semi-cylindrical 7500-1 capacity steel tank, with an opening to the environment of 2.4 m. × 2.4 m. The opening allows installation of doors, feedthroughs, pipes, or other components. The first set of experiments with the PET were conducted in 2016 using hollow doors subjected to a water rise scenario. Data collected during the door tests is being analyzed using Bayesian regression methods to determine the parameters of influence and inform future experiments. A practical method of simulating full scale wave impacts on components and structures is also being researched to further enhance CFEL capabilities. Early on, the team determined full scale wave impacts could not be simulated using traditional wave flumes or pools; therefore, closed conduit flow is under consideration. Computational fluid dynamics software is being used to simulate fluid velocities associated with tsunami waves of heights up to 6-m, and to design a wave impact simulation device capable of accurately recreating a near vertical wave section with variable height and fluid velocity. The component flooding simulation activities associated with this project involve use of smoothed particle dynamics codes. These particle-based simulation methods do not require a mesh to be applied to the fluid, which allows for more natural flows to be simulated. Finally, CFEL can be described as a pioneering element, comprised of several ongoing research and experimental projects, that are vital to the development of risk analysis methods for the nuclear industry.

VVER-440 Steam Generator’s Two-Phase Flow Analysis

2016 ◽  
Vol 821 ◽  
pp. 57-62
Author(s):  
Lukas Joch ◽  
Roman Krautschneider
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Steam Generator ◽  
Nuclear Power ◽  
Flow Analysis ◽  
Three Dimensional ◽  
Reactor Power ◽  
Physical Models ◽  
Two Phase ◽  
Ansys Fluent

The subject of this report is creation of three-dimensional thermal hydraulic model of horizontal steam generator for Dukovany nuclear power plant. A procedure is presented for simulation and analysis of secondary side of PGV-440 steam generator for nominal and increased reactor power. A two-fluid approach is applied for modeling physical processes inside the steam generator. Physical models were implemented in ANSYS Fluent CFD environment using User Defined Functions (UDFs). Results from this thermal hydraulic numerical model can be used for various other subsequent nuclear power plant operations and safety analysis.

A Calibration of a Feedwater Flowmeter Installed to Model Loop of a Nuclear Power Plant Using High Reynolds Number Facility

2008 ◽  
Author(s):  
Noriyuki Furuichi ◽  
Yoshiya Terao ◽  
Masaki Takamoto
Keyword(s):  
Reynolds Number ◽  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
High Reynolds Number ◽  
Tube Bundle ◽  
Three Dimensional ◽  
Calibration Result ◽  
Calibration Facility ◽  
High Reynolds

A calibration result of ultrasonic flowmeters using in a feedwater flowrate in a nuclear power plant, is described under a variety of upstream conditions using the new high Reynolds number calibration facility. The pipe layouts are classified to five type three-dimensional one with two or three elbows. The flow conditioners are tube bundle type and Mitsubishi type. Pipe Reynolds number is up to 1.6×107. The large effect of the flow conditioner and pipe layout is observed for cramp-on type. For multi-path type, individuality is observed.

Development of Three-Dimensional Seismic Isolation Technology for Next Generation Nuclear Power Plant Applications

2005 ◽  
Author(s):  
K. Takahashi ◽  
K. Inoue ◽  
M. Morishita ◽  
T. Fujita
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Seismic Isolation ◽  
Base Isolation ◽  
Three Dimensional ◽  
Scale Model ◽  
Isolation System ◽  
The Real ◽  
Reactor Building

Seismic isolation technology plays an important role in the area of architect engineering, especially in Japan where earthquake comes so often. This technology also makes the nuclear power plant rationalized. The horizontal base isolation with laminated rubber bearings has already been proven its effectiveness. These days, seismic isolation technology is expected to mitigate even the vertical load, which affects the structural design of primary components. Seismic isolation system has possibility to improve the economical situation for the nuclear power plant. From these points of view, a research project has been proceeded to realize practical three dimensional seismic isolation systems from 2000 to 2005 under the sponsorship of the Ministry of Economy, Trade and Industry of the Japanese government. The isolation system is developed for the supposed “Fast Breeder Reactor (abbreviated FBR)” of the next generation. Two types of seismic isolation systems are developed in the R&D project. One is a three-dimensional base isolation for a reactor building (abbreviated 3D SIS) and the other is a vertical isolation for main components with horizontal base isolation of the reactor building (abbreviated V. +2D SIS). At first step of the R&D, requirements and targets of development for the seismic isolation system were identified. Seismic condition for R&D was discussed based on the real seismic response. Vertical natural frequency and damping ratio required to the system were introduced from the response to the seismic movement. As for 3D SIS, several system concepts were proposed to satisfy the requirements and targets. Through discussions and tests on performance, reliability, applicability, maintainability, “Rolling seal type air spring system with hydraulic anti-rocking devices” was decided to be developed. Verification shaking tests with the 1/7 scale model of the system and analysis for applicability to the real plant are conducted. The result shows that the system is able to support the reactor building, to suppress the rocking motion and to mitigate the vertical seismic load. As for V.+2D SIS, coned disk spring device was selected at the beginning of R&D. Performance tests of the elements, which include common deck movement, were conducted and the system applicability to the plant is confirmed. Verification tests were conducted with 1/8 scale model of the system and the result proves the applicability to the real plant.

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