抄録
Low cycle fatigue behavior of low alloy steels ASTM A508 C1.3(JIS SFVQ1A) and ASTM A533B C1.1(JIS SQV2A) for nuclear reactor pressure vessels was investigated in high temperature pressurized water simulating BWR coolant environments. Total strain range, strain rate and dissolved oxygen concentration were varied from 0.5 to 2.2 %, 0.1 to 0.001 %/s and 10 to 8 000 ppb, respectively. Fatigue tests in ambient air and 561 K air were also conducted for comparison.
It was found that fatigue lives in high temperature water were shorter than those in ambient air. However, the reduction of fatigue life decreased with decreasing total strain range and rather longer fatigue lives than those in ambient air were observed at lower total strain range. A533B material showed the distinct strain rate dependence of fatigue life compared with A508 material, while they showed the similar dependence on dissolved oxygen concentration. It was found that fatigue cracks initiated at corrosion pits generated by dissolution of MnS inclusions and the low cycle fatigue behavior dependent on sulfur content of the material.
It can be concluded that the materials tested possess safety margins in reactor coolant environments by judging from the fact that all the present data fell on a region above the design fatigue curves in the ASME Code Sec. III.
