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Nonlinear Dynamic Analysis of Reactor Coolant Loop Coupled With Reactor Building

[+] Author Affiliations
Rodolfo L. M. Suanno, Carlos L. M. Prates, Maria Inês L. Prates de Lima, Tarcísio F. Cardoso

Eletrobrás Termonuclear S/A, Rio de Janeiro, RJ, Brazil

Paper No. PVP2006-ICPVT-11-93316, pp. 767-775; 9 pages
  • ASME 2006 Pressure Vessels and Piping/ICPVT-11 Conference
  • Volume 3: Design and Analysis
  • Vancouver, BC, Canada, July 23–27, 2006
  • Conference Sponsors: Pressure Vessels and Piping Division
  • ISBN: 0-7918-4754-3 | eISBN: 0-7918-3782-3
  • Copyright © 2006 by ASME


The steam generator (SG) snubber elimination process in a nuclear power plant requires a new evaluation of the structural behavior of the complete primary system components for licensing purposes. The forces and stresses have to be evaluated in all supports, piping, nozzles and internals of all components of the reactor coolant loop (RCL) for the required load cases, including dead weight, thermal conditions, seismic excitations and postulated piping ruptures. The SG snubber elimination intends to obtain a safer operating condition, avoiding problems with snubber maintenance, inspection and mal-function. The paper describes the methodology adopted for this type of analysis, where a very detailed modeling procedure is required, both for the primary loop itself, where nonlinearities are introduced to represent the supporting devices, as well as for the coupling with the reactor building structure. The piping and the components (Reactor vessel, SG and pumps) are modeled in order to represent their weight distribution, stiffness and supporting conditions in detail. The reactor building complete 3D-finite element model is reduced to a corresponding representative simple beam model in order to make the nonlinear dynamic analysis feasible. The seismic response spectra from both building models were compared at supporting points of the primary circuit in order to guarantee that the simple beam model represents the behavior of the refined building model in a correct way. The dynamic analysis is performed with seismic acceleration time histories applied at the foundation of the reactor building model and a direct integration method is used. The Rayleigh damping values as well as the effects in the results of refining the integration time steps are discussed. The impact forces due to postulated pipe ruptures are also evaluated as impact loads. The results of these analyses are displacements, accelerations and forces in all structural elements and their supports, as well as time histories and response spectra for the stress analysis of the component internal structures.

Copyright © 2006 by ASME



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