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Division Spotlight
Fuel Cycle & Waste Management
Devoted to all aspects of the nuclear fuel cycle including waste management, worldwide. Division specific areas of interest and involvement include uranium conversion and enrichment; fuel fabrication, management (in-core and ex-core) and recycle; transportation; safeguards; high-level, low-level and mixed waste management and disposal; public policy and program management; decontamination and decommissioning environmental restoration; and excess weapons materials disposition.
Meeting Spotlight
International Conference on Mathematics and Computational Methods Applied to Nuclear Science and Engineering (M&C 2025)
April 27–30, 2025
Denver, CO|The Westin Denver Downtown
Standards Program
The Standards Committee is responsible for the development and maintenance of voluntary consensus standards that address the design, analysis, and operation of components, systems, and facilities related to the application of nuclear science and technology. Find out What’s New, check out the Standards Store, or Get Involved today!
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Latest News
TerraPower begins U.K. regulatory approval process
Seattle-based TerraPower signaled its interest this week in building its Natrium small modular reactor in the United Kingdom, the company announced.
TerraPower sent a letter to the U.K.’s Department for Energy Security and Net Zero, formally establishing its intention to enter the U.K. generic design assessment (GDA) process. This is TerraPower’s first step in deployment of its Natrium technology—a 345-MW sodium fast reactor coupled with a molten salt energy storage unit—on the international stage.
W. R. Martin, J. R. Weir
Nuclear Technology | Volume 1 | Number 2 | April 1965 | Pages 160-167
Technical Paper | doi.org/10.13182/NT65-A20485
Articles are hosted by Taylor and Francis Online.
The tensile properties of Hastelloy N have been determined after irradiation at 700° C to a dose level of 7 × 1020 n/cm2 (E > 1 MeV) and 9 × 1020 n/cm2 (thermal). The strength and ductility of the material were determined as functions of deformation temperature for the range 20 to 900°C. These properties were also examined as functions of strain rate within the limits of 0.002 and 0.2 in./min (0.005 and 0.5 cm/min) for deformation temperatures of 500, 600, 700, and 800°C., The stress-strain relationship is not affected by irradiation at 700°C. Ductility, as measured by the true uniform and fracture strains, is reduced for deformation temperatures of 500°C and above. The loss in ductility results in a reduction in the true tensile strength. This loss is more significant at test conditions resulting in intergranular failure, such as low strain rates at elevated temperature. Postirradiation annealing of the irradiated alloy does not result in improved ductility. These data are compatible with the experiments suggesting helium generation from the (n,α) reaction of boron as the cause of low ductility., The low ductility of irradiated alloys in general is described in terms of the present knowledge of intergranular fracture. Means of improving the ductility are discussed.