Session: CS-07-04 Recent Developments in ASME Codes and Standards - 4 (with CS-08)

Paper Number: 176221

176221 - Fatigue Crack Growth Rate Thresholds of Reactor Pressure Vessel Steel in Water Environment for the Asme Code Section Xi 

Abstract: 

 The reference fatigue crack growth rates (da/dN) for ferritic steels in water environments are provided in Appendix Y of the ASME B& PV Code Section XI. The fatigue crack growth rate is expressed as a power law relationship with the range of applied stress intensity factor range (ΔK). The existing da/dN relations were determined using a large quantity of experimental data that were primarily obtained in a relatively high ΔK regime, since the intent was to model the power law regime and avoid the differing behavior of da/dN in the near-threshold regime. Consequently, few data were available to support a fatigue crack growth rate threshold (ΔK_th) specific to water environments, and ASME Code Section XI currently treats ΔK_th for water environments as equivalent to the ΔK_th obtained for data in air environment. The fatigue test explicitly for obtaining ΔK_th is not straightforward, considering time requirements as well as the need to keep consistent environmental conditions. We have conducted a literature survey and collected experimental fatigue crack growth data on near-threshold fatigue crack growth rates. Also, we gathered ΔK_th for various stress ratios (R) at both room and elevated temperatures in water environments.

 In this paper, we propose a fatigue crack growth rate threshold (ΔK_th) relation specifically for ferritic steels in water environments based on the data from an extensive literature survey. The proposed ΔK_th is expressed as a function of the stress ratio, and the results are compared with the reference fatigue crack growth curves in the current edition of the ASME Code Section XI.

Presenting Author: Yoosung Ha Japan Atomic Energy Agency

Presenting Author Biography: Dr. Yoosung Ha is an assistant principal researcher at Japan Atomic Energy Agency (JAEA).She has been working in the Ageing Management Research Group, focusing on the degradation behavior of structural materials in nuclear environments. She has also been working on the flaw evaluation of structural materials, taking operational conditions into account.

Authors:

Yoosung Ha Japan Atomic Energy Agency
Yoshihito Yamaguchi Japan Atomic Energy Agency
Kunio Hasegawa Japan Atomic Energy Agency
Martin Negyesi Centre for Advanced and Innovative Technologies - VSB-Technical University of Ostrava
Kevin Fuhr Dominion Engineering. Inc.