Session: SE-06-02 The T.H. Liu Memorial Symposium on the Seismic Analysis and Design of Piping Systems-2

Paper Number: 123304

123304 - Design Margins of Fatigue Life of Carbon Steel Elbows and Tees Subjected to Reversing Dynamic Loads 

Fatigue damage is a type of cumulative damage observed in nuclear piping systems. The fatigue damage is caused by various reversing dynamic loads including seismic loads. When the reversing dynamic loads produce plastic strain in each cycle, the loading cycles cause severe damage in the component, and thus the number of cycles to failure is relatively low. This type of failure due to the severe reversing load is called low-cycle fatigue. Past research revealed that some uncertainty is generally inevitable in predicting the number of cycles to low-cycle fatigue. Hence, it is of great importance to understand the mechanism of low-cycle fatigue and to ensure an ample margin in fatigue life evaluation in piping systems.

In this study, a series of shaking-table tests was carried out to study the mechanism of low-cycle fatigue and the design margin of pipe fittings. Three types of specimens were chosen as typical pipe fittings: elbows, equal tees, and reducing tees. All the specimens are made of JIS G3456 STPT370 seamless carbon steel pipe, equivalent to ASTM A106 Grade A. The amplitudes of acceleration, displacement, and strain of each specimen were measured under the sinusoidal excitation of uniform frequency. We investigated the relation between input and output response of the observed data, and then discussed the margin of pipe fittings based on the seismic design of nuclear piping systems.

Presenting Author: Kisaburo Azuma Nuclear Regulation Authority

Presenting Author Biography: Dr. Kisaburo Azuma is a researcher working for the Nuclear Regulation Authority, Japan. He specializes in fracture mechanics, and has expertise in finite element analyses. He has been published in the Journal of Pressure Vessel Technology, Transactions of the ASME and awarded the Outstanding Conference Paper Award from the ASME PVP2018 (Prague) for the “Code and Standard (C&S)” track. He holds a doctoral degree from the University of Tokyo, where he completed a doctoral program in the field of nuclear engineering. He has been a member of the Working Group on Pipe Flaw Evaluation for the ASME Boiler and Pressure Vessel Code, Section XI.

Authors:

Kisaburo Azuma Nuclear Regulation Authority
Keita Fujiwara Nuclear Regulation Authority
Satoru Kai IHI
Akihito Otani IHI
Osamu Furuya Tokyo Denki University