Published: Jan 1983
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Reactor pressure vessels of nuclear installations are subjected to operating parameters that lead to a change in material and flaw state. The safety margin of both the material for older vessels and the optimized material has to be quantified with reference to extremely long-term effects. Research programs in West Germany are focused on assessing criteria for the applicability of small-scale specimen testing to component behavior under all realistic and postulated loading conditions. The problem will be treated theoretically and experimentally for a wide range of materials, including the worst-case material state at the end of life.
light water reactors, pressure vessel steels, irradiation, fracture mechanics, pressurized thermal shock, large scale specimens, intermediate size vessels, welded joints
Professor and Director, Staatliche Materialprüfungsanstalt, University of Stuttgart, Stuttgart,
Paper ID: STP31845S