Abstract
Extending the safety analysis of a nuclear reactor pressure vessel beyond the requirements of the regulations which were first laid down in the ASME-Code, the behavior of two crack sizes (1/4-t axial, 3/4-t circumferential, a/2c = 1/6) in the upper shelf region is analyzed to demonstrate the capability of a reactor pressure vessel for Leak-Before-Break. The postulated load is a monotonic increase of pressure, ending when the crack penetrates through the wall. The conditions for re-initiation and ductile extension of the conservatively postulated largest possible axial and circumferential crack geometries resulting from initiation and arrest during a thermal shock are evaluated on the basis of the J-integral concept. Three-dimensional finite element models simulating large amounts of crack growth are employed as well as analytical approximations. The results demonstrate that very high internal pressure far beyond the capacity of the plant and also far beyond the strength of the other co mponents of the primary circuit would be required to initiate and grow the cracks, and that Leak-Before-Break behavior is confirmed.