Koichi Kashima

Low cycle fatigue and ductile fracture for Japanese carbon steel piping under dynamic loadings

Abstract Dynamic fracture behavior of circumferentially cracked pipe is important to evaluate the structural integrity of nuclear piping from the viewpoint of the LBB concept under seismic conditions. Fracture tests have been conducted for Japanese carbon steel (STS410) circumferentially through-wall cracked pipes that are subjected to monotonic or cyclic bending loads at room temperature. In the monotonic-loading tests, the maximum load to failure increases slightly with increasing loading rate. The failure cycles can be expressed simply by ratio of the load amplitude to the plastic collapse load. Fracture analysis has been also conducted to model the pipe tests. A new equation for calculating ΔJ for a circumferentially through-wall cracked pipe subjected to bending has been proposed. The failure cycles under cyclic loads are satisfactorily evaluated using an elastic-plastic fracture mechanics parameter ΔJ .

Elastic-plastic analysis on stable crack growth for center cracked plate: a benchmark study

Abstract For the evaluation of the accuracy of the finite element method in the elastic-plastic stable crack growth analysis, a benchmark test using a center cracked specimen of Type 304 stainless steel is performed. Six groups in the Japan Society of Mechanical Engineers participate in this project and provide six numerical solutions for the given problem using different solution methods and computer codes assuming the plane stress condition. The criterion for the crack extension used here is the experimental relationship between the gauge displacement and the crack extension amount at the specimen surface. The results of the numerical analyses and the experiment are compared with each other with regard to the global deformation behavior and the J -integral value before the crack initiation as well as during the crack extension. Discussion is made about the sources of scattering between the numerical solutions and the experiment.

Overview of JSME flaw evaluation code for nuclear power plants

A Japanese flaw evaluation code for nuclear power plant components has been developed at the Japan Society of Mechanical Engineers (JSME). The code prescribes methods for the evaluation of flaws, which are detected during inservice inspection for pressure vessels and pipes in nuclear power plants. This paper describes the basic flow chart, methods of evaluation and allowable flaw sizes for acceptance standards and criteria, including comparisons with the ASME Code Section XI.

Leak-before-break verification test and evaluations of crack growth and fracture criterion for carbon steel piping

Abstract A proving test on the integrity of carbon steel piping in light water reactors (LWRs) was planned by the Nuclear Power Engineering Test Center (NUPEC) as a four-year verification test program; it was completed at the end of March 1989. The objective of this proving test was to demonstrate the validity of the Leak-Before-Break (LBB) concept for high quality carbon steel piping under actual plant conditions. This paper briefly describes the results of material property tests, fracture behavior tests, LBB verification tests, numerical analyses of pipe fracture behavior and evaluation of flaw growth and fracture criterion. From these results, LBB has been verified and a fracture criterion has been developed for carbon steel piping in LWRs.

Ductile Fracture Behaviour of Class 2 and 3 LWR Piping and Its Implications for Flaw Evaluation Criteria

To achieve a rational maintenance program for aged Light Water Reactor components, it is important to establish and to improve the flaw evaluation criteria. The current flaw evaluation criteria such as ASME Boiler and Pressure Vessel Code Section XI are focused on Class 1 piping which usually shows relatively higher toughness. On the other hand, flaw evaluation criteria suitable for Class 2, 3 piping with moderate-toughness are also required because some Class 2, 3 piping systems are as important to plant safety analysis as Class 1 piping. In this study, both analytical and experimental studies were conducted to provide the evaluation method of fracture loads for acceptance criteria for Class 2, 3 piping. Pipe fracture tests by four-point bending were conducted on circumferentially cracked carbon steel pipes with moderate-toughness. The Net-Section Collapse criterion overpredicted experimental maximum loads for through-wall-cracked pipes, which suggested the necessity of Z-factor. Three-dimensional finite element analysis and simplified analysis based on the reference stress method were conducted to complement the limited pipe fracture tests. It was ascertained that the reference stress method always gave moderately conservative fracture loads compared with the finite element analysis and pipe fracture tests as well. Z-factor for Class 2, 3 piping was then derived and formulated using the reference stress method. Z for Class 2, 3 piping was affected by radius-to-thickness ratio, and was higher than Z for Class 1 piping in the present codes.

Simplified stable crack growth analyses of welded CT specimens—comparison study of GEEPRI, reference stress and R6 methods

Abstract This paper describes some simplified stable crack growth analyses of two kinds of inhomogeneous CT specimens. The one is machined from a submerged are welded plate of a nuclear pressure vessel A533B Class 1 steel, while the other is machined from an electron-beam welded plate of the A533B Class 1 steel and a high strength HT80 steel. In both specimens, initial cracks are placed to be normal to the fusion line. The ratio of yield stresses of the weld metal and the base metal of the A533B Class 1 steel is about 1·15, while that of the HT80 and the A533B Class 1 steels is about 1·4. The generation phase crack growth analyses using the GE EPRI and the reference stress methods are performed, calculating an applied load (P) and the J-value, while the application phase analyses of analyses using the R6 method are performed to calculate the maximum value of the applied load (Pmax). Finally, some modification procedures of the three simplified estimation schemes are discussed in order to apply them to inhomogeneous material regimes.

Research activities on fracture mechanics for nuclear piping in Japan

Abstract This paper presents an overview of the piping studies, such as the studies on ductile fracture of piping and the development of fracture analysis methods, that have been or are being conducted in Japan.

Flaw Evaluation Procedures and Acceptance Criteria for Nuclear Piping in ASME Code Section XI

Section XI of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code provides rules and requirements for maintaining pressure boundary integrity of components, piping, and equipment during the life of a nuclear power plant. Evaluation procedures and acceptance criteria for the evaluation of flaws in nuclear piping in Section XI of the ASME Code were first published in 1983 and have been under revision for the past several years. This paper provides an overview of the procedures and acceptance criteria for pipe flaw evaluation in Section XI. Both planar and nonplanar flaws are addressed by Section XI. The evaluation procedures and acceptance criteria cover: failure by plastic collapse as characterized by limit load analysis; fracture due to ductile tearing prior to attainment of limit load, as characterized by elastic-plastic fracture mechanics (EPFM) analysis; and brittle fracture as characterized by linear elastic fracture mechanics (LEFM) analysis. A major revision to the evaluation procedures and acceptance criteria was published in the 2002 Addenda to Section XI. Evaluation procedures and acceptance criteria in the 2001 Edition, as well as the revisions in the 2002 Addenda, are described in this paper. Code Cases that address evaluation of wall thinning in piping systems, as well as temporary acceptance of flaws in moderate energy piping systems, are also described.Copyright

Flaw evaluation of Japanese carbon steel piping by load curve approach

Abstract New simplified flaw evaluation method, ‘load curve approach’ was developed to evaluate the fracture load of circumferentially surface-cracked pipe. This approach has the same functions with the current two-criteria approach. Fracture stress and fracture criteria are easily estimated by two load curves based on elastic–plastic fracture mechanics and plastic collapse. Fracture analysis was conducted for Japanese carbon steel piping using this approach. The approach showed the dependency of flaw geometry and pipe diameter on pipe fracture. Z -factors were calculated from this approach and compared with Z -factors by ASME Boiler & Pressure Vessel Code, Section XI (ASME-XI) and Japanese Code. It is shown that Z -factors by the load curve approach can improve the conservativeness in the estimation of pipe fracture load.

A research program for dynamic fracture evaluation of Japanese carbon steel pipes

A research program was developed to investigate the dynamic load effect on fracture behavior of Japanese carbon steel STS410 pipe. The program comprises material tests, pipe fracture tests and development of estimation scheme. Material property tests showed that the flow stress was nearly constant or slightly increased with strain rate. Pipe tests showed that fracture load was nearly predicted by the net-section collapse criterion for both quasi-static and dynamic loading. Significant dynamic effect was not observed for STS410 carbon steel piping. Crack growth was well formulated by using J-integral parameter for low cycle fatigue with large scale yielding. Combining the crack growth behavior and unstable fracture criterion, an estimation scheme was newly developed and validated for constant amplitude cyclic loading conditions.