| ||Format||Pages||Price|| |
|PDF (176K)||14||$25||  ADD TO CART|
|Complete Source PDF (13M)||759||$66||  ADD TO CART|
Cite this document
Present irradiation surveillance programs are based mainly on Charpy V-notch impact bend specimens. The nil-ductility transition (NDT) temperature shift as a relative measure of embrittlement is determined from the transition temperature curves. Under the assumption of identical shifts of the Charpy V-notch transition curves and the pressure vessel material fracture behavior, as expressed by the KIR = f(T − RTNDT) curve, the pressure-temperature limits for the reactor pressure vessel are derived.
The adequacy of Charpy V-notch impact bend specimens for the aforementioned purpose has been questioned. Since 1973 we in Switzerland have looked into these problems and have tried to overcome them by introducing some modifications in the surveillance program as recommended by the ASTM Practice for Conducting Surveillance Tests for Light Water-Cooled Nuclear Power Reactor Vessels (E 185-73). Our concept is oriented toward a more direct assessment of the fracture toughness of the irradiated material. This assessment can be achieved by using additional fracture mechanics specimens in the surveillance program.
This extended concept has been applied to two nuclear power plants in Switzerland.
Problems inherent in the use of Charpy V-notch impact bend specimens are discussed and our approach to overcome them presented. The surveillance programs for a pressurized water reactor and boiling water reactor, according to our concept, are described.
thermal reactors (nuclear), pressure vessels, irradiation effects, embrittlement, impact tests, evaluation
Swiss Nuclear Safety Division, Wuerenlingen,
Professor, Institute for Research and Technology at the Technical University of Viennax,