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Division Spotlight
Accelerator Applications
The division was organized to promote the advancement of knowledge of the use of particle accelerator technologies for nuclear and other applications. It focuses on production of neutrons and other particles, utilization of these particles for scientific or industrial purposes, such as the production or destruction of radionuclides significant to energy, medicine, defense or other endeavors, as well as imaging and diagnostics.
Meeting Spotlight
Utility Working Conference and Vendor Technology Expo (UWC 2024)
August 4–7, 2024
Marco Island, FL|JW Marriott Marco Island
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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BWXT will scout potential TRISO fuel production sites in Wyoming
BWX Technologies Inc. announced today that its Advanced Technologies subsidiary has signed a cooperation agreement with the state of Wyoming to evaluate locations and requirements for siting a potential new TRISO nuclear fuel fabrication facility in the state.
J.-J. Huet, V. Leroy+
Nuclear Technology | Volume 24 | Number 2 | November 1974 | Pages 216-224
Technical Paper | Material | doi.org/10.13182/NT74-A31476
Articles are hosted by Taylor and Francis Online.
Dispersion-strengthened ferritic steels are being studied for possible use as canning material for sodium-cooled fast reactors. The basic alloy chosen contains nominally Fe—13% Cr—1.5% Mo— 3.5% Ti to which 2% TiO2 or 1% Y2O3 is added by a powder metallurgy technique. At 700°C, the alloys studied can favorably be compared in stress rupture tests (up to 12 000 h) to the best austenitic steels. Corrosion tests in dynamic sodium at 700°C showed that after 4 000 h the affected zones remained narrow and had no significant influence on the mechanical resistance at high temperature. Neutron irradiation of these alloys demonstrated their remarkable resistance to embrittlement in mechanical tests at 700°C. Comparison with other alloys showed that they had the highest elongation to rupture after irradiation. Simulation tests by 1-MeV electrons gave almost zero swelling in the temperature range of 475 to 700°C. The combined properties of dispersion-strengthened ferritic alloys make them excellent candidates not only for canning material but also for shroud tubes for fast-reactor fuel elements.