Numerical investigation of residual heat removal pumps based on fluid-structure interaction in 1000 MW nuclear power plants

2014 ◽  
Author(s):  
Banglun Zhou ◽  
Yin Luo ◽  
Jianping Yuan ◽  
Ji Pei ◽  
Jiaxing Lu ◽  
...  
Keyword(s):  
Numerical Investigation ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Fluid Structure Interaction ◽  
Heat Removal ◽  
Fluid Structure ◽  
Structure Interaction ◽  
Residual Heat

Tackling FSI Simulation for FIV Problems in Tube Bundle Systems With POD Approach

2011 ◽  
Author(s):  
Marie Pomarede ◽  
Aziz Hamdouni ◽  
Erwan Liberge ◽  
Elisabeth Longatte ◽  
Jean-Franc¸ois Sigrist
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Tube Bundle ◽  
Fluid Structure Interaction ◽  
Flowing Fluid ◽  
Fluid Structure ◽  
Reduced Order ◽  
Structure Interaction ◽  
Flow Past ◽  
Steam Boilers

Tube bundles in steam boilers of nuclear power plants and nuclear on-board stokehold are known to be exposed to high levels of vibrations under flowing fluid. This coupled fluid-structure problem is still a challenge for engineers, first because of the difficulty to fully understand it, second because of the complexity for setting it up numerically. Although numerical techniques could help the understanding of such a mechanism, a complete simulation of a fluid past a whole elastically mounted tube bundle is currently out of reach for engineering purposes. To get round this problem, the use of a reduced-order model has been proposed with the introduction of the widely used Proper Orthogonal Decomposition (POD) method for a flow past a fixed structure [M. Pomare`de, E. Liberge, A. Hamdouni, E.Longatte, & J.F. Sigrist - Simulation of a fluid flow using a reduced-order modelling by POD approach applied to academic cases; PVP2010, July 18–22, Seattle]. Interesting results have been obtained for the reconstruction of the flow. Here a first step is to propose to consider the case of a flow past a fixed tube bundle configuration in order to check the good reconstruction of the flow. Then, an original approach proposed by Liberge (E. Liberge; POD-Galerking Reduction Models for Fluid-Structure Interaction Problems, PhD Thesis, Universite´ de La Rochelle, 2008) is applied to take into account the fluid-structure interaction characteristic; the so-called “multiphase” approach. This technique allows applying the POD method to a configuration of a flow past an elastically mounted structure. First results on a single circular cylinder and on a tube bundle configuration are encouraging and let us hope that parametric studies or prediction calculations could be set up with such an approach in a future work.

scholarly journals Investigation of dynamic stress of rotor in residual heat removal pumps based on fluid–structure interaction

2016 ◽  
Vol 8 (9) ◽  
pp. 168781401666574 ◽  
Author(s):  
Banglun Zhou ◽  
Jianping Yuan ◽  
Yanxia Fu ◽  
Feng Hong ◽  
Jiaxing Lu
Keyword(s):  
Dynamic Stress ◽  
Fluid Structure Interaction ◽  
Heat Removal ◽  
Fluid Structure ◽  
Structure Interaction ◽  
Residual Heat

An Overview of Numerical Method for Fluid-Structure Interaction

2014 ◽  
Author(s):  
J.-H. Jeong ◽  
M. Kim ◽  
P. Hughes
Keyword(s):  
Numerical Simulation ◽  
Numerical Methods ◽  
Nuclear Power ◽  
Power Plants ◽  
Fluid Structure Interaction ◽  
Phase Flow ◽  
Two Phase ◽  
Fluid Structure ◽  
Structure Interaction ◽  
Deformable Structure

Fluid-structure interaction (FSI) is the interaction of some movable or deformable structure with an internal or surrounding fluid flow. Therefore, fluid-structure interaction problems are too complex to solve analytically and so they have to be analysed by means of experiments or numerical simulation. This paper provides an overview of numerical methods for fluid-structure interaction evaluation in an draft IAEA technical guideline: large eddy simulation (LES), direct numerical simulation (DNS), Lattice-Boltzmann method (LBM), finite element method (FEM) and computational fluid dynamics (CFD) method. In addition to providing general applications of numerical methods for fluid-structure interaction evaluation, the paper also describes some cases applied for problems associated with single-phase flow and two-phase flow in nuclear power plants.

Pipe Rupture Analysis Considering Fluid-Structure Interaction

2011 ◽  
Author(s):  
Yukari Hamamoto ◽  
Makoto Toyoda
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Thrust Force ◽  
Fluid Structure Interaction ◽  
Low Carbon ◽  
Analysis Model ◽  
Fluid Structure ◽  
Structure Interaction ◽  
Fluid Analysis ◽  
Autogenous Control

Global warming is caused by the emission of greenhouse gases, like CO2. Nuclear energy is one of the main sources of low-carbon energy. In the events of serious accidents, a nuclear power plant may emit radioactivity that is harmful to human health. Nuclear power should be used after enough evidence of its safety is provided. Measures for safety of nuclear power plants, such as autogenous control and LBB, have been developed. Moreover, there is requirement with respect to the design, safety, equipments components and systems of nuclear plant. For example, it is necessary to place components that restrain pipe whip behavior, and to design peripheral equipments that may be affected by high-pressured fluid in pipe rupture accidents [1], [2]. In the case of pipe rupture that occurs to structures such as nuclear plants and steam generators, a pipe deforms releasing its inner high-pressured fluid. In previous studies, the pipe whip behavior analyses have been performed by using blowdown thrust force that is estimated by fluid analysis. In this study, we simulate pipe whip behavior and reduction of blowdown thrust force by releasing inner fluid to the atmosphere. The analysis model is an elbow pipe and high-pressure fluid running inside. We considered fluid-structure interaction effect in the analysis because ovalization of the cross-section of the elbow part as well as a change of the elbow torus radius affects fluid flow blowing out from the ruptured part of the pipe.

Analysis on Fluid-Structure Interaction Dynamic Characteristics of Steam Generator Heat Exchanger Tubes

2011 ◽  
Vol 382 ◽  
pp. 52-55
Author(s):  
Li Na Zhang ◽  
Su Zhen Wang
Keyword(s):  
Steam Generator ◽  
Dynamic Characteristics ◽  
Nuclear Power ◽  
Power Plants ◽  
Fluid Structure Interaction ◽  
Finite Element Program ◽  
Fluid Structure ◽  
Structure Interaction ◽  
Interaction Dynamic ◽  
Steam Generator Tubes

The fluid-structure interaction (FSI) dynamic characteristics of steam generator tubes counting for much with safety of an operating nuclear power plants are investigated by analytical methods based on dynamics mechanics and FSI theories. By using the parametric design language APDL of finite element program ANSYS, intelligently dividing model, setting up material and geometric parameters, the models of tubes with internal and external fluid, the different factors influencing on fluid-structure interaction dynamic characteristics of steam generator tubes are investigated by numerical method.

scholarly journals Seismic safety evaluation during site selection for the nuclear power plants in Bangladesh

2018 ◽  
Vol 4 (4) ◽  
pp. 251-256 ◽  
Author(s):  
Sergey Shcheklein ◽  
Ismail Hossain ◽  
Mohammad Akbar ◽  
Vladimir Velkin
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Spent Nuclear Fuel ◽  
Heat Removal ◽  
Good Time ◽  
Diagnostics System ◽  
Seismic Impacts

Bangladesh lies in a tectonically active zone. Earlier geological studies show that Bangladesh and its adjoining areas are exposed to a threat of severe earthquakes. Earthquakes may have disastrous consequences for a densely populated country. This dictates the need for a detailed analysis of the situation prior to the construction of nuclear power plant as required by the IAEA standards. This study reveals the correlation between seismic acceleration and potential damage. Procedures are presented for investigating the seismic hazard within the future NPP construction area. It has been shown that the obtained values of the earthquake’s peak ground acceleration are at the level below the design basis earthquake (DBE) level and will not lead to nuclear power plant malfunctions. For the most severe among the recorded and closely located earthquake centers (Madhupur) the intensity of seismic impacts on the nuclear power plant site does not exceed eight points on the MSK-64 scale. The existing predictions as to the possibility of a super-earthquake with magnitude in excess of nine points on the Richter scale to take place on the territory of the country indicate the necessity to develop an additional efficient seismic diagnostics system and to switch nuclear power plants in good time to passive heat removal mode as stipulated by the WWER 3+ design. A conclusion is made that accounting for the predicted seismic impacts in excess of the historically recorded levels should be achieved by the establishment of an additional efficient seismic diagnostics system and by timely switching the nuclear power plants to passive heat removal mode with reliable isolation of the reactor core and spent nuclear fuel pools.

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