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LWR pressure vessel failure assessment in case of loca using advanced thermohydraulic and fracture mechanics methods

 
: Schmitt, W.; Nagel, C.; Hertlein, R.; Blauel, J.G.

INSTIN '94. Proceedings of the Saclay International Seminar on Structural Integrity
1994
pp.31-39
Saclay International Seminar on Structural Integrity <1994, F-Gif-sur-Yvette>
English
Conference Paper
Fraunhofer IWM ()
fracture mechanics; reactor pressure vessel safety analysis; subclad cracks; warm prestress effect

Abstract
Whereas earlier analyses of the integrity of the reactor pressure vessel of NPP Stade (KKS) under emergency core cooling conditions have been based on the assumption of an abrupt change from service to cooling medium temperature applied to the complete inner surface of the vessel [1 ] it became obvious through international experience and own investigations that strip and strand cooling could induce more severe conditions. Especially, if the same maximum temperature drop as for the axi-symmetric case was assumed to cool only a limited strip below the inlet nozzle the analyses yielded higher stresses in the vessel wall and more unfavourable conditions in terms of crack loading and crack resistance for exclusion of crack initiation and limitation of crack growth. These findings have caused enormous efforts in experimental and theoreticallnumerical investigations to improve modelling of the complex thermohydraulic processes during loss of coolant accidents (LOCA) and by that to gain more realistic input data for an updated fracture mechanics safety analysis. The results to be presented for RPV-KKS also take into account the effect of the austenitic cladding and include a reevaluation and application of the warm prestress effect.

: http://publica.fraunhofer.de/documents/PX-22565.html