Article ID | Journal | Published Year | Pages | File Type |
---|---|---|---|---|
790959 | International Journal of Pressure Vessels and Piping | 2008 | 9 Pages |
Abstract
In this paper, we discuss an ASME-NH program that has been developed to overcome the complexity and costs arising from the real application of the ASME-NH rules by hand calculations for class 1 nuclear facility component design for elevated temperature operations. A computerized program is described for implementing all the assessment procedures such as the time-dependent primary stress limits, total accumulated creep-ratcheting strain limits, and the creep–fatigue damage limits by the elastic and inelastic analysis methods complying with the ASME-NH rules. As an example application, a preliminary structural integrity evaluation for a high-temperature reactor vessel design of a typical lead-cooled reactor is described.
Related Topics
Physical Sciences and Engineering
Engineering
Mechanical Engineering
Authors
Gyeong-Hoi Koo, Jae-Han Lee,