Fatigue crack growth in 316 type stainless steel at temperatures and displacement damage rates representative for the NET first wall loadingFunded under: FP2-FUSION 10C
Fatigue crack growth initiating from pre-existing 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 thermo-mechanical 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 approximately 1.0 E-7 dpa/s 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.
Bibliographic Reference: Paper presented: International Conference on Fusion Reactor Materials (ICFRM-5), Oak Ridge National Laboratory, Clearwater (US), Nov. 17-22, 1991
Availability: Available from (1) as Paper EN 36641 ORA
Record Number: 199210374 / Last updated on: 1994-12-02
Original language: en
Available languages: en