Safety research of light water reactor pressure vessels in the Federal Republic of Germany (FRG) is presently focused on validation programs based on fracture mechanics concepts which study irradiation embrittlement, pressurized thermal shock, and cyclic crack growth under operating conditions. Crack initiation values derived from the JR-curve could be confirmed as reliable material property to enable transferability of results from small specimen testing to complex structures.

For validation of irradiation surveillance practice, trepans were taken from a forged shell course of a commercial reactor pressure vessel wall to compare material degradation of the wall with that of the surveillance specimens and to establish properties through the thickness of the wall. The experimentally determined amount of reduction in toughness of the vessel wall depends strongly on the orientation of the specimens with respect to the main forging direction of the hot formed material. The degradation of the material with respect to transition temperature shift and upper shelf energy drop in Charpy impact tests could not be predicted conservatively on the basis of the existing trend curves in the U.S. Code of Federal Register. On the other hand, the gradient of embrittlement measured in short distances from the inner to the outer surface of the vessel wall is much less and therefore less important than assessed in the proposed revision of the U.S. Code.

Archive material representing the shell course from which the trepans were taken was irradiated in test reactors in the United States and the United Kingdom. The irradiated archive material shows a lower degradation than the vessel wall of the commercial reactor.