Abstract
This work is concerned with assessing the lifetime integrity of VVER (water-water power reactor) reactor pressure vessel. Results based on the Soviet Code are compared with those of the ASME Code, Section III and XI. Similarities and discrepancies are illustrated and discussed in connection with numerical results of a typical RPV (reactor pressure vessel) under operational conditions. Involved are data for fracture toughness, crack initiation and crack arrest conditions. Considerations are also given for possible adoption of the ASME Code for VVER reactors.
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