No abstract

In order to predict long-term behavior of a high-creep resistant low-carbon Type 316L stainless steel under low-cycle fatigue with long hold times, a series of tests of fatigue relaxation was undertaken at 550, 600, 650, and 700°C for medium strain ranges (Δεt = 0.7, 1.2, and 1.6 percent). Hold times up to 5 h were introduced at the maximum tensile strain. It has been shown that a reduction of fatigue life occurred, generally associated with intergranular cracking when hold times increased. A maximal effect was observed at 600°C. Different methods for extrapolating results for very long hold times, such as those encountered on fast breeder reactor components (∼1000 h) were proposed. These methods were based on a time-temperature equivalence comparable to those used for extrapolating creep rupture data. A correlation between reduction of fatigue life with the amount of stress relaxation during hold times was also used. Predictions by these methods are compared with ASME N47 fatigue design curves.

No abstract

Research on irradiation embrittlement in modern pressure vessel steels was initiated at the French Atomic Energy Commission ten years ago. During this period many steels and welds were irradiated in the Triton and Siloe reactors; research was undertaken both independently and within the framework of research programs coordinated by the International Atomic Energy Agency. Mainly through tests of tensile and Charpy impact properties, the effects of chemical composition were studied, confirming the influence of copper and phosphorus and uncovering the surprisingly detrimental effect of nickel content. At the same time, new fracture mechanics testing procedures for irradiated materials were developed. Today the major part of these programs is complete and future activities will probably be reduced on pressure vessel steels. The main topics for forthcoming studies are the effect of flux or irradiation rate on embrittlement and elastic-plastic fracture mechanics. A review of past activities and a preview of forthcoming research is given. During the period covered by these studies, many pressurized water reactors were built in France that are now starting to operate, with each reactor containing irradiation surveillance capsules. This important program, which is described in the second part of the paper, will lead in the next ten years to an increased knowledge of radiation embrittlement on a more statistical basis, taking into account results obtained on a reference plate.

No abstract