Session: CS-23-01 Improvement of Flaw Assessment Procedures in Fitness-for-Service Codes

Paper Number: 122432

122432 - Allowable Stress and Allowable Flaw Sizes Estimated by Converted Tensile Properties From Hardness for Austenitic Stainless Steel Pipe 

 When a flaw is detected in a high-toughness ductile pipe of a power plant, stress applied at the flaw location of the pipe is compared with an allowable stress calculated by the combination of failure stress and safety factor. When the applied stress at the flaw location is less than the allowable stress, the plant can continue operation for a certain evaluation period. The failure stress is estimated by flow stress, and the safety factor is defined by fitness-for-service codes. The flow stress is normally given by the average of yield stress and ultimate tensile strength of the pipe material and the values are tabulated in fitness-for-service codes. On the other hand, many fitness-for-service codes and technical reports recently adopted yield stress and ultimate tensile strength converted from the hardness value using the correlation between the hardness and tensile properties. It is convenient and useful method to obtain the flow stress, because it is not necessary to extract specimens from actual plant pipe. This paper is to introduce the equations on conversion of Vickers hardness to tensile properties provided by fitness-for-service codes and technical reports. In addition, allowable bending stress for circumferential flaw is derived from the flow stress which is estimated using yield stress and ultimate tensile strength converted from the Vickers hardness. Then, that allowable stress is compared with the allowable stress obtained by actual measured flow stress, calculated by experimental yield stress and ultimate tensile strength, for small diameter austenitic stainless steel pipe subjected to bending moment.

Presenting Author: Yoosung Ha Japan Atomic Energy Agency

Presenting Author Biography: Working as a researcher in JAEA.
Ph. D graduated Kyoto University in Japan.
The main research topics are the mechanical test and evaluation, irradiation embrittlement evaluation of nuclear materials.

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 – VŠB-Technical University of Ostrava