Finite Element Based Stress Analysis of BWR Internals Exposed to Accident Loads


Finite Element Based Stress Analysis of BWR Internals Exposed to Accident Loads

Altstadt, E.; Weiß, F.-P.; Werner, M.; Willschütz, H.-G.

During a hypothetical accident the reactor pressure vessel internals of boiling water reactors can be exposed to considerable loads resulting from temperature gradients and pressure waves. The finite element (FE) analysis is an efficient tool to evaluate the consequences of those loads by computing the maximum mechanical stresses in the components. 3 dimensional FE models were developed for the core shroud, the upper and the lower core supporting structure, the steam separator pipes and the feed water distributor. The models of core shroud, upper core support structure and lower core support structure were coupled by means of the substructure technique. All FE models can be used for thermal and for structural mechanical analyses. As an example the FE analysis for the case of a station black-out scenario (loss of power supply for the main circulating pumps) with subsequent emergency core cooling is demonstrated. The transient temperature distributions within the core shroud and within the steam dryer pipes as well were calculated based on the fluid temperatures and the heat transfer coefficients provided by thermo-hydraulic codes. At the maximum temperature gradients in the core shroud, the mechanical stress distribution was computed in a static analysis with the actual temperature field being the load. It could be shown that the maximum resulting material stresses do not exceed the permissible thresholds fixed in the appropriate regulations. Another scenario which was investigated is the break of a feed water line leading to a non-symmetric subpressure wave within the reactor pressure vessel. The dynamic structural response of the core shroud was assessed in a tranisient analysis. Even for this load case the maximum resulting stresses remain within the allowed limits at any time.

  • Lecture (Conference)
    Jahrestagung Kerntechnik‘98, München, 26.-28. Mai 1998, Tagungsbericht S. 721-724
  • Contribution to proceedings
    Jahrestagung Kerntechnik‘98, München, 26.-28. Mai 1998, Tagungsbericht S. 721-724

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Publ.-Id: 1065