Dynamic fracture analysis of thermal shock loading in nuclear pressure vessels

The propagation and eventual arrest of a preexisting flaw in a reactor vessel subjected to thermal shock is of prime importance in the analysis of nuclear reactor safety. Four thermal shock experiments have been performed at Oak Ridge National Laboratories on thick walled cylinders (outside diameter 0.5 m, wall thickness 146 mm). However, observations on the smaller vessels cannot be related to large reactor pressure vessels (5 m OD, wall thickness 0.2 m) unless an analysis is available to connect the two sets of conditions. Although several fracture analyses are available, their critical evaluations have been confined mainly to run-arrest events in small laboratory test pieces. The two-dimensional, dynamic, linear elastic, finite difference code described in this paper is designed to analyze run-arrest events in thick-walled cylinders. Preliminary results obtained with this code were compared with observations obtained during the fourth ORNL thermal shock experiment (TSE-4) in which the open-ended cylinder was preheated to 2910C and then cooled by pumping a fluid at -250C through its interior cavity. In this experiment, designed to simulate conditions similar to those encountered by a PWR (pressurized water reactor) vessel during a hypothetical loss-of-coolant accident, the crack driving force is provided by an adverse temperature gradient across the cylinder's wall. Experimentally, it was observed that the originally 11 mm deep, long axial crack started propagating 150 seconds after the beginning of fluid injection into the cylinder and arrested after having propagated in a single jump to a depth of 23 mm. (orig.)