0
Journal of Pressure Vessel Technology | Volume 129 | Issue 1 | Research Paper
RESEARCH PAPERS

Fluid-Structure Interaction Effects Modeling for the Modal Analysis of a Nuclear Pressure Vessel

[+] Author and Article Information
Jean-François Sigrist

Service Technique et Scientifique,  DCN Propulsion, 44620 La Montagne, Francejean-francois.sigrist@dcn.fr

Daniel Broc

Service d’Etude Mécanique et Sismique,  CEA Saclay, 91191 Gif-Sur-Yvette, France

Christian Lainé

Service Technique et Scientifique,  DCN Propulsion, 44620 La Montagne, France

J. Pressure Vessel Technol 129(1), 1-6 (Mar 17, 2006) (6 pages) doi:10.1115/1.2389025 History: Received October 17, 2005; Revised March 17, 2006
Article
References
Figures
Tables
Citing Articles

The present paper deals with the modal analysis of a nuclear reactor with fluid-structure interaction effects. The proposed study aims at describing various fluid-structure interaction effects using several numerical approaches. The modeling lies on a classical finite element discretization of the coupled fluid-structure equation, enabling the description of added mass and added stiffness effects. A specific procedure is developed in order to model the presence of internal structures within the nuclear reactor, based on periodical homogenization techniques. The numerical model of the nuclear pressure vessel is developed in a finite element code in which the homogenization method is implemented. The proposed methodology enables a convenient analysis from the engineering point of view and gives an example of the fluid-structure interaction effects, which are expected on an industrial structure. The modal analysis of the nuclear pressure vessel is then performed and highlights of the relative importance of FSI effects for the industrial case are evaluated: the analysis shows that added mass effects and confinement effects are of paramount importance in comparison to added stiffness effects.
FIGURES IN THIS ARTICLE
<>
Copyright © 2007 by American Society of Mechanical Engineers
Your Session has timed out. Please sign back in to continue.

References

1
Makerle, J., 1999, “Fluid-Structure Interaction Problems, Finite Element Approach and Boundary Elements Approaches. A Bibliography,” Finite Elem. Anal. Design, 31 , pp. 231–240.
2
Morand, H. J.-P., and Ohayon, R., 1995, "Fluid Structure Interaction", Wiley, New York.
3
Sigrist, J. F., Broc, D., and Lainé, C., 2005. “A Homogenization Method for the Analysis of a Coupled Fluid-Structure Interaction Problem with Inner Solid Structures,” European Nuclear Conference , Versailles.
4
Axisa, F., 2001, "Interactions Fluide Structure", Hermès, London.
5
Rajakumar, C., and Rogers, C. G., 1991, “The Lanczos Algorithm Applied to Unsymmetric Generalized Eigenvalue Problem,” Int. J. Numer. Methods Eng.  [CrossRef], 32 , pp. 1009–1026.
6
Boujot, J., 1987, “Mathematical Formulation of Fluid-Structure Interaction Problems,” Model. Math. Anal. Numer., 21 , pp. 239–260.
7
Gibert, R. J., 1986, "Vibrations des structures. Interaction avec les fluides. Sources d’excitation aléatoires", Collection de la Direction des Etudes et Recherches d’Electricté de France, vol. 69 , Eyrolles.
8
Sigrist, J. F., Broc, D., and Lainé, C., 2005, “Seismic Analysis of a Nuclear Reactor with Fluid-Structure Interaction,” Pressure Vessel and Piping, Denver, CO.
9
Brochard, D., and Hammami, L., 1998, “Non-Linear Analysis of Nuclear Reactor Cores by an Homogenization Method,” Pressure Vessel and Piping, San Diego, CA.
10
Cheval, K., 2001, “Modélisation du comportement sismique de structures multitubulaires baignées par un fluide dense,” Ph.D. Thesis, Université d’Evry.
11
Planchard, J., and Ibnou-Zahir, M., 1983, “Natural Frequencies of a Tube Bundle in an Incompressible Fluid,” Comput. Methods Appl. Mech. Eng.  [CrossRef], 41 , pp. 47–68.
12
Conca, C., Planchard, J., Thomas, B., and Vanninathan, M., 1994, "Problèmes mathématiques en couplage fluide∕structure. Applications aux faisceaux tubulaires", Collection de la Direction des Etudes et Recherches d'Electricité de France, Vol. 85 , Eyrolles.
13
Broc, D., Queval, J. C., and Viallet, E., 2003, “Seismic Behaviour of PWR Reactor Cores: Fluid-Structure Effects,” 17th Conference on Structural Mechanic in Reactor Technology , Prague.
14
Broc, D., 2004, “Seismic Behaviour of PWR Reactor Cores: Whole Core Model with Fluid-Structure Interaction Effects,” Flow Induced Vibrations , Paris, July 6–9, Vol. 1 , pp. 283–288.
15
Verpeaux, P., 1989, “A Modern Approach of Computer Codes for Structural Analysis,” 10th Conference on Structural Mechanics in Reactor Technology , Anaheim, CA.
16
Cheng, F. Y., 2001, "Matrix Analysis of Structural Dynamics", Marcel Dekker, New York.

Figures

Grahic Jump Location
+Axisymmetric model of the nuclear pressure vessel
View Large  |  Save Figure  |  Download Slide (.ppt)
Axisymmetric model of the nuclear pressure vessel
Figure 1

Axisymmetric model of the nuclear pressure vessel

Grahic Jump Location
+Fluid problem mesh with and without inner structures modeling
View Large  |  Save Figure  |  Download Slide (.ppt)
Fluid problem mesh with and without inner structures modeling
Figure 2

Fluid problem mesh with and without inner structures modeling

Grahic Jump Location
+Elementary fluid cell mesh
View Large  |  Save Figure  |  Download Slide (.ppt)
Elementary fluid cell mesh
Figure 3

Elementary fluid cell mesh

Grahic Jump Location
+Axisymmetric finite element model of the nuclear pressure vessel
View Large  |  Save Figure  |  Download Slide (.ppt)
Axisymmetric finite element model of the nuclear pressure vessel
Figure 4

Axisymmetric finite element model of the nuclear pressure vessel

Grahic Jump Location
+Fluid and structure finites elements, coupling and interface elements
View Large  |  Save Figure  |  Download Slide (.ppt)
Fluid and structure finites elements, coupling and interface elements
Figure 5

Fluid and structure finites elements, coupling and interface elements

Grahic Jump Location
+Uncoupled eigenmode of the reactor (without fluid)
View Large  |  Save Figure  |  Download Slide (.ppt)
Uncoupled eigenmode of the reactor (without fluid)
Figure 6

Uncoupled eigenmode of the reactor (without fluid)

Grahic Jump Location
+Coupled eigenmodes of the reactor (with fluid, without inner pressure, without inner structures)
View Large  |  Save Figure  |  Download Slide (.ppt)
Coupled eigenmodes of the reactor (with fluid, without inner pressure, without inner structures)
Figure 7

Coupled eigenmodes of the reactor (with fluid, without inner pressure, without inner structures)

Grahic Jump Location
+Stiffness forces from correction terms Kπ and Kσ
View Large  |  Save Figure  |  Download Slide (.ppt)
Stiffness forces from correction terms Kπ and Kσ
Figure 8

Stiffness forces from correction terms Kπ and Kσ

Grahic Jump Location
+Cumulated modal masses for the coupled eigenmodes calculated with the homogenization technique
View Large  |  Save Figure  |  Download Slide (.ppt)
Cumulated modal masses for the coupled eigenmodes calculated with the homogenization technique
Figure 9

Cumulated modal masses for the coupled eigenmodes calculated with the homogenization technique

Tables

Errata

Discussions

Some tools below are only available to our subscribers or users with an online account.

Related Content

Customize your page view by dragging and repositioning the boxes below.

Related Journal Articles
Related eBook Content
Introduction
Heat Transfer & Hydraulic Resistance at Supercritical Pressures in Power Engineering Applications> Chapter 1
Lessons Learned: NRC Experience
Continuing and Changing Priorities of the ASME Boiler & Pressure Vessel Codes and Standards> Chapter 30
PSA Level 2 — NPP Ringhals 2 (PSAM-0156)
Proceedings of the Eighth International Conference on Probabilistic Safety Assessment & Management (PSAM)
Source Term Assessments in PSA Level 2 for the Outage Period (PSAM-0168)
Proceedings of the Eighth International Conference on Probabilistic Safety Assessment & Management (PSAM)
Evaluation of Test Interval Strategies with a Risk Monitor (PSAM-0166)
Proceedings of the Eighth International Conference on Probabilistic Safety Assessment & Management (PSAM)
Topic Collections

Sorry! You do not have access to this content. For assistance or to subscribe, please contact us:

  • TELEPHONE: 1-800-843-2763 (Toll-free in the USA)
  • EMAIL: asmedigitalcollection@asme.org
Sign In