Development of Elevated Temperature Design Criteria for Nuclear Components
Our engineering staff has assisted in the development of elevated temperature structural design criteria which have become part of the ASME Boiler and Pressure Vessel Code for Nuclear Power Plant Components. This work included development of criteria and methods for fatigue evaluation, ductile failure mode analysis, the inelastic behavior of complex components, and design for cyclic service at elevated temperature. Some of these methods use the results of purely elastic analyses to perform simplified inelastic analyses of structures in elevated temperature service.
Design criteria for acceptance of cracks in structural components subjected to cyclic loading at elevated temperature are included. This work was sponsored by Oak Ridge National Laboratories, Westinghouse Electric, Rockwell International, General Electric, Novatome and EDF in France, Interatom in West Germany, and ISES in Japan. It included strain limits and the “O’Donnell-Porowski Bounds” for creep ratcheting and creep fatigue used to design high temperature reactors worldwide.
We perform elastic, elastic-plastic and elevated temperature analysis on equipment
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– Finite Element Analysis of a Vessel Processing Nuclear Waste
– Finite Element Analysis & ASME Code Section VIII Division 2 Calculations on Feedwater Heaters
– Design & Analysis
– Nuclear Power Plant Consulting
– Publications – Heat Exchangers, Pressure Vessels, Welds and Other Applications, Fatigue, Elevated Temperature
– Links to Engineering Resources
– Our Engineering Team
– Background of the ASME Code