It has been clarified that the fatigue life is reduced in the fatigue test of high-temperature and high-pressure water that simulates PWR reactor coolant environment compared to that in the tmosphere, and temperature, strain rate，etc. affect the decrease of fatigue life. In this study, the influence of crack growth behavior on fatigue life of SUS316 in simulated PWR reactor coolant environment of different temperature was investigated. Fatigue tests were conducted under different temperatures (325℃ and 200℃...
英字タイトル:
Influence of Temperature on the Fatigue Crack Initiation and Growth in Reactor Coolant Environment

Abstract: After Fukushima-Daiichi NPP accident, improvement of nuclear safety is highly requested in order to prevent re-occurrence of severe accident. New sc me of fatigue evaluation to confirm system safety in NPPs is enhanced to be established based on Defense-in-Depth concept. In step 1, grand design of fatigue evaluation is reviewed on the viewpoint of fatigue management in long-term safe operation of NPPs, which is required after Fukushima-Daiichi NPP accident. Thi study focuses on the direction to r...
英字タイトル:
Reconstruction of Grand Design for Fatigue Evaluation (Step-2) -Fatigue failure analysis of Japanese NPPs-

Abstract: In this research, we investigate the issues of fatigue we have now and review Grand Design of fatigue evaluation on the basis of new role of fatigue anagement in plant long term operating scheme that is required after Fukushima daiichi NPP accident. ...
英字タイトル:
Reconstruction of Grand Design for Fatigue Evaluation

Fatigue life of nuclear facilities tends to be decreased by the influence of reactor coolant, which is called environmental effect. The effect accelerates crack growth rate but the influence for cra ck initiation is not clarified. This study intends to discuss the environmental effect in crack initiation. The crack length and the number of cracks are measured from the investigation of fatigue t est specimens in reactor coolant and air. The behavior of crack initiation is revealed from the measurement of nu...
英字タイトル:
Evaluations of environmental effect on micro crack initiation and propagation by surface observations of fatigue specimens

1.はじめに 現在、原子力発電所においては経年劣化による設備の 機能喪失を防ぐため保全活動が行われている。プラント 機器の保全において疲労の累積を予測し損傷の防止を図 ることは事故防止につながる重要な活動である。疲労累 積評価は単軸試験で得たデータを基にして作成された設 計疲労曲線を用いて行われている。設計疲労曲線による 評価はもともと機器の設計に用いる目的で作られている ため、運転中の機器の実際の疲労累積状態や余寿命の評 価が難しいという課題がある。 そこで、実機の疲労累積状態を仮想的なき裂の長さに 置......
英字タイトル:
Evaluation of Micro Fatigue Crack Growth under Equi-biaxial Stress by Membranous Pressure Fatigue Test

Cumulative Usage Factor (CUF) is currently used for fatigue design and management to prevent fatigue failure. Fatigue damage is evaluated by CUF but CUF doesn’t show the real extent of fatigue damage and cannot be used for probabilistic index for risk assessment. Therefore, in order to clarify relation between CUF and damage state, the introduction of postulated crack growth curve (P-curve) of stainless steel was examined by Kamaya [1]. This curve was defined by single crack growth curve made from fati...
英字タイトル:
Study on Fatigue Crack Growth Prediction Model of Carbon Steel based on Surface Observation of Fatigue Test Specimen

After Fukushima Daiichi accident, nuclear power plants are required to ensure the system safety with long te rm operation. Present design code requires the assessment of CUF (cumulated usage factor) to prevent the occ rence of fatigue damage. However, it is difficult to introduce the concept of CUF into the system safety eva luation which needs probabilistic risk assessment. For the probabilistic evaluation of fatigue failure, thi s study intends to show the prediction of micro crack growth by simulation. ...
英字タイトル:
Approach to establish micro crack growth prediction model based on surface observations of fatigue test specimen

Abstract: The pipe rupture accident by FAC (Flow Accelerated Corrosion) at Mihama Unit3 of The Kansai Electric Power Company made the Japanese utilities recognize the significance of pipe wall thinning in plant life cy cle management. In Japanese PWR, BWR, and thermal power plant, the pipe wall thinning is managed under individual codes. This paper studies the differences of wall thinning management in the codes and their reasons cl early. Quantitative analysis is performed based on the data in the report ...
英字タイトル:
Analysis of comparison between the codes of nuclear and thermal power plantfor improving pipe wall thinning management

Abstract: The pipe rupture accident by FAC (Flow Accelerated Corrosion) at Mihama Unit3 of The Kansai Electric Power Company made the Japanese utilities recognize the significance of pipe wall thinning in plant life cy cle management. In Japanese PWR, BWR, and thermal power plant, the pipe wall thinning is managed under individual codes. This paper studies the differences of wall thinning management in the codes and their reasons cl early. Quantitative analysis is performed based on the data in the report ...

Abstract: The pipe rupture accident by FAC (Flow Accelerated Corrosion) at Mihama Unit3 of The Kansai Electric Power Company made the Japanese utilities recognize the significance of pipe wall thinning in plant life cy cle management. In Japanese PWR, BWR, and thermal power plant, the pipe wall thinning is managed under individual codes. This paper studies the differences of wall thinning management in the codes and their reasons cl early. Quantitative analysis is performed based on the data in the report ...
英字タイトル:
Analysis of comparison between the codes of nuclear and thermal power plantfor improving pipe wall thinning management