[en] This publication is the collection of the paper presented at the title meeting. The 18 of the presented papers are indexed individually. (J.P.N.)

[en] This paper describes a probabilistic fracture mechanics (PFM) analysis of aged nuclear reactor pressure vessel (RPV) material. New interpolation formulas are first derived for both embedded elliptical surface cracks and semi-elliptical surface cracks. To investigate effects of transition from embedded crack to surface crack in PFM analyses, one of PFM round-robin problems set by JSME-RC111 committee, i.e. 'aged RPV under normal and upset operating conditions' is solved, employing the interpolation formulas. (author)

[en] Taiwan BWR-6 Kuosheng Nuclear Power Plant Unit 1 implemented the inspection of the IGSCC-susceptible weldments of stainless steel piping in the reactor recirculation, reactor water clean-up, residual heat removal, core spray and feedwater systems. The purpose of this paper is to present the status of the fracture problems in the weldments. The crack growth analysis due to IGSCC and the standard weld overlay design based on the ASME Code Section XI and NUREG-0313 Rev.2 for the fracture weldments are discussed in detail. Then, the contingent programs including the inspection program, fracture evaluation, the standard weld overlay are completely established to prevent the pipe break during the reactor operation. (author)

[en] The three dimensional finite element analysis of the bolted joints with finite sliding deformable contact has been studied, and the helical and friction effect on the load distribution of each thread is analyzed. It shows the analytical analysis by Yamamoto's method reaches a lower value of load ratio than the finite element analysis at the first thread. The load distribution on each thread between axisymmetrical model and three dimensional model are provided. Hence, although increasing the coefficient of friction and decreasing of the lead angle may improve the load distribution slightly, for 1'-16UNF bolt joints, the error of load ratio at the first thread in axisymmetrical finite element model is 12% with respect to three dimensional analysis. (author)

[en] It is very important to be able to predict the remaining life of components and structures in a power plant, both for nuclear and fossil units. The information needed can be obtained from the controlled laboratory experiments and the plant operating data. On materials degradation, we have accumulated a large amount of data from both sources. However, it is essential to formulate the best methodology to utilize these information so that our needs can be met. In this paper, the methods currently used for remaining life prediction are discussed with typical results. Also discussed are the limitations and the benefits of different approaches along with suggestions for the future R and D directions in this area. (author)

[en] To estimate the structural integrity of the Light Water Reactor piping, combined loading consists of tensile load due to internal pressure and bending load under seismic condition should be considered as a basic loading mode. However, theoretical investigation on the methodology to evaluate ductile fracture behavior is not adequate to date. In this study, an approximate evaluation method for ductile fracture analysis of circumferentially through-wall-cracked pipe subjected to combined bending and tension was newly developed. This method can explicitly incorporate the contribution of both tension and bending. The effect of growing crack is also considered in the method. This method was then applied to the full-scale pipe fracture tests. Based on the comparison with experimental results as well as finite element calculations, it could be ascertained that the proposed method could well predict ductile fracture behavior under combined loading. The effect of combined loading on ductile fracture was sensitivity-studied using the proposed method. As a result, it was quantitatively found that the superposition of internal pressure reduced the maximum load of cracked pipe. (author)

[en] Fracture resistance (J-R) curves, which are used for elastic-plastic fracture mechanics analyses, are known to be dependent on the cyclic loading history. The objective of this paper is to investigate the effect of reverse cyclic loading on the J-R curves in C(T) specimens. Two parameters were observed to be effective on the J-R curves during the reverse cyclic loading. One was the minimum-to-maximum load ratio (R) and the other was the incremental plastic displacement (δ_cycle/δ_i), which is related to the amount of crack growth that occurs in a cycle. Fracture resistance test on C(T) specimens with varying the load ratio and the incremental plastic displacement were performed, and the test results showed that the J-R curves were decreased with decreasing the load ratio and decreasing the incremental plastic displacement. Direct Current Potential Drop (DCPD) method was used for the detection of crack initiation and crack growth in typical laboratory J-R tests. The values of J_i and δ_i were also obtained by using the DCPD method. (author)

[en] Taiwan PWR Nuclear Power Plant Units 1 and 2 implemented the measurements of the wall thinning of the carbon steel piping under the request of regulation authority to prevent the events due to the erosion/corrosion since 1989. At the first, the licensee established the comprehensive inspection program by itself. Over 2000 components were inspected per each unit and 300-500 pipe components were examined by ultrasonic testing per scheduled outage. The simple predictive methodology determined the operability of each individual piping component in the next fuel cycle. Based on the inspection results, the susceptible pipe components were established. The implementation of the effective correction management and improved inspection program can improve the safety, as well as the efficiency, of long-term reactor operations. (author)

[en] Since the suggestion of External Reactor Vessel Cooling (ERVC), the effects of malting and cooling on the response of structural integrity of the Reactor Pressure Vessel (RPV) under core melting accident conditions have been investigated. This paper describes the vessel response according to the ERVC condition and analysis method. The steady state and transient analysis for the temperature and stress field were performed using ABAQUS. Especially, transient analyses were studied for the variable boundary conditions. To obtain an analogy with real phenomena, the material properties were determined by combining and modifying the existing results considering phase transformation and temperature dependency. The results show that the vessel can be melted if there is no external cooling. Finally, the potential for vessel damage is discussed using the Larson-Miller curve and damage rule. (author)

[en] The core shroud replacement of a Boiling Water Reactor (BWR) is now in progress at Fukushima-Daiichi Unit no.3(1F3) of the Tokyo Electric Power Company (TEPCO) in Japan. The core shroud and other core internal components made of type 304 stainless steel (SS) are replaced with the ones made of low carbon type 316L SS to improve Intergranular Stress Corrosion Cracking (IGSCC) resistance. This project is the first application of the replacement procedure developed for the welded core shroud, and employs various advanced technologies. The outline of the core shroud replacement project and applied technologies are discussed in this paper. (author)