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- W2088916903 abstract "Abstract Fatigue crack growth initiating from preexisting mechanical flaws is one of the major parameters to be considered when evaluating the endurance life of tokamak reactor components. The anticipated pulsed operation of Next Step systems will impose cyclic thermomechanical loads on structural materials associated with radiation damage which can limit the lifetime of the first wall. Fatigue crack growth under cyclic tensile stress was measured on AISI 316 stainless steel under proton irradiation at a displacement damage rate of ≈10−7 dpa s−1 and at 373, 473, and 573 K. The influence of proton irradiation appears to be more significant at the primary stage of crack propagation, causing only a slight decrease of the fatigue crack growth rate in the secondary stage. Irradiation tends also to prolong fatigue life. The results suggest that under irradiation at low fluences and low temperatures, the effect is due to hardening by defect agglomerates produced by displacements." @default.
- W2088916903 created "2016-06-24" @default.
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- W2088916903 date "1992-09-01" @default.
- W2088916903 modified "2023-10-16" @default.
- W2088916903 title "Fatigue crack growth in 316 type stainless steel at temperatures and displacement damage rates representative for the first wall loading" @default.
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- W2088916903 doi "https://doi.org/10.1016/0022-3115(92)90707-r" @default.
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