Irradiation Effects on Reactor Structural Materials.
Quarterly progress rept. 1 Aug-31 Oct 69,
NAVAL RESEARCH LAB WASHINGTON D C
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The report includes 1 initial comparisons between dynamic tear test and Charpy V-notch impact data for irradiated steels, including the PM-2A vessel steel, 2 the recovery of ductility by annealing heat treatment of steels irradiated to different ratios of thermal to fast neutrons, 3 the unirradiated properties of special A533-B steel heat procured for low enbrittlement sensitivity, 4 data describing the thermal stability of a potential advanced reactor structural alloy, 5Ni-Cr-Mo-V steel, and 5 initial strength and ductility data on selected austenitic stainless steels, 304, 304L, 316, and 316L, after irradiation in the EBR-II reactor to fluences between 0.4 and 9.0 x 10 to the 20th power nsq cm 1 MeV at temperatures ranging from 700F 371C to 830F 443C. Author
- Metallurgy and Metallography
- Radioactivity, Radioactive Wastes and Fission Products
- Fission Reactor Materials