Initialising ...
Initialising ...
Initialising ...
Initialising ...
Initialising ...
Initialising ...
Initialising ...
Wakai, Takashi; Machida, Hideo*; Arakawa, Manabu*
Nihon Kikai Gakkai 2018-Nendo Nenji Taikai Koen Rombunshu (DVD-ROM), 5 Pages, 2018/09
According to the fitness for service code of Sodium-Cooled fast Reactor (SFR), the volumetric tests as in-service inspection can be replaced with continuous leak monitoring, where the Leak Before Break (LBB) is demonstrated, because the primary stress caused by internal pressure is not significant in SFR components. Basically, if the detectable crack length and the penetrated crack length are sufficiently smaller than the unstable critical crack length, it can be concluded that LBB is successfully demonstrated. The authors had already proposed a simplified method to calculate the penetrated crack length both of the circumferential and axial cracks in the pipe as a function of pipe geometry, fatigue crack growth characteristics and loading conditions. However, some problems in the method have been pointed out in the process of the reviewing by the JSME code committee. This study describes an improved method to calculate the penetrated crack length.
Hasegawa, Kunio; Li, Y.; Bezensek, B.*; Hoang, P. H.*; Rathbun, H. J.*
Journal of Pressure Vessel Technology, 138(1), p.011101_1 - 011101_8, 2016/02
Times Cited Count：1 Percentile：8.17(Engineering, Mechanical)Piping components in power plants may experience combined bending and torsion moments during operation. There is a lack of guidance for pipe evaluation for pipes with local wall thinning under the combined bending and torsion moments. ASME Code Section XI Working Group is currently developing fully plastic bending pipe evaluation procedures for pressurized piping components containing local wall thinning subjected to combined torsion and bending moments. Using elastic fully plastic finite element analyses, plastic collapse bending moments under torsion were obtained for 114.3 mm to 609.6 mm diameter pipes with various local wall thinning flaw sizes. The objective of this paper is to introduce an equivalent moment, which combines torsion and bending moments by a vector summation, and to establish the applicable range of wall thinning lengths, angles and depths, where the equivalent moments are equal to pure bending moments.
Hasegawa, Kunio; Li, Y.; Saito, Koichi*
Journal of Pressure Vessel Technology, 137(4), p.041101_1 - 041101_7, 2015/08
Times Cited Count：3 Percentile：16.43(Engineering, Mechanical)If a subsurface flaw is located near a component surface, the subsurface flaw is transformed to a surface flaw in accordance with a flaw-to-surface proximity rule. The re-characterization process from subsurface to surface flaw is adopted in all fitness-for-service (FFS) codes. However, the concrete criteria of the re-characterizations are different among the FFS codes. Cyclic tensile experiment was conducted on a carbon steel flat plate with a subsurface flaw at ambient temperature. The objective of the paper is to compare the experiment and calculation of fatigue crack growth behavior for a subsurface flaw and the transformed surface flaw, and to describe the validity of the flaw-to-surface proximity rule defined by ASME Code Section XI, JSME S NA1 Code and other codes.
Takaya, Shigeru
no journal, ,
Effective and efficient ISI is crucial for safety, stable operation, and economic efficiency of nuclear power plants. ISI rules need to be developed rationally by considering features of the reactor type and design of an individual nuclear power plant. SFRs have several desirable features while traditional ISI methods are not as easily performed as in light water reactors. ASME Boiler and Pressure Vessel Code, Section XI, Division 3 had provided ISI rules for liquid-metal cooled reactors. However, several parts of the Code were left as being in the course of preparation, thus it was practically difficult to apply. To address this situation, a new Code Case, N-875, was developed based on the SBC concept in cooperation between ASME and JSME. One of key features of the Case is the logic flow to determine ISI requirements by using reliability targets for components. Recently, a new Section XI, Division 2, Reliability and Integrity Management (RIM), has been developed for advanced nuclear reactors. RIM shares some key concepts with SBC, and the Case has been partially incorporated. JSME has also developed a new fitness-for-service code for SFRs by referring the Case, and it will be published soon.
