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PWR Fuel Assembly Modal Testing and Analysis

[+] Author Affiliations
Bruno Collard, Stéphane Pisapia, Frédéric Witters

CEA, St Paul lez Durance, France

Sergio Bellizzi

CNRS, Marseille Cedex, France

Paper No. PVP2003-2084, pp. 147-152; 6 pages
  • ASME 2003 Pressure Vessels and Piping Conference
  • Flow-Induced Vibration
  • Cleveland, Ohio, USA, July 20–24, 2003
  • Conference Sponsors: Pressure Vessels and Piping Division
  • ISBN: 0-7918-4156-1
  • Copyright © 2003 by ASME


Pressurized Water Reactor (PWR) seismic or Lost Of Coolant Accident (LOCA) loads could result in impacts between nuclear fuel assemblies or between fuel assemblies and the core baffles. Forces generated during these shocks are often the basis for the determination of the maximum loads and of the spacer grid and fuel rod design. The knowledge of the fuel assembly kinematics is essential to compute these maximum loads, and this requires experimental tests. Our study aims at characterizing the behavior of a full-scale fuel assembly subjected to various excitations. The effect of the assembly environment (air, still water and water under flow) is studied. The French Nuclear Reactor Directorate experimental facility HERMES T allows hydraulic and mechanical testing of full-scale fuel assemblies. It is designed for flow rate up to 1200 m3/h and temperature up to 170°C. Specific excitation devices allow mechanical tests with amplitudes of motion up to 20 mm. Laser vibrometry, displacement transducers and tracking camera apparatus measure the fuel assembly displacement. To identify this Multi Degree Of Freedom (MDOF) system (assembly or assembly + fluid), two dependent problems have to be addressed: the linear or non-linear model selection, and the estimation of the corresponding parameters. Under different environments and excitation types, it is shown that the mechanical system is strongly non-linear. The damping term, essentially fluid, increases with flow rate and with motion amplitude, while the stiffness decreases with amplitude. The main results, the measuring and identification methods and the extrapolation to the reactor thermohydraulic conditions are presented and discussed.

Copyright © 2003 by ASME



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