ANS is committed to advancing, fostering, and promoting the development and application of nuclear sciences and technologies to benefit society.
Explore the many uses for nuclear science and its impact on energy, the environment, healthcare, food, and more.
Division Spotlight
Mathematics & Computation
Division members promote the advancement of mathematical and computational methods for solving problems arising in all disciplines encompassed by the Society. They place particular emphasis on numerical techniques for efficient computer applications to aid in the dissemination, integration, and proper use of computer codes, including preparation of computational benchmark and development of standards for computing practices, and to encourage the development on new computer codes and broaden their use.
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!
Latest Magazine Issues
Jul 2024
Jan 2024
Latest Journal Issues
Nuclear Science and Engineering
August 2024
Nuclear Technology
Fusion Science and Technology
Latest News
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.
R. O. Meyer, B. J. Buescher
Nuclear Technology | Volume 14 | Number 2 | May 1972 | Pages 153-156
Technical Paper | Fuel Cycle | doi.org/10.13182/NT72-A31130
Articles are hosted by Taylor and Francis Online.
Approximate solutions in analytic form are presented for the heat-flow equation that yields temperature as a function of radial position in mixed-oxide fuel irradiated in a fast-neutron flux. These equations are applied to fuel with density that varies due to restructuring. In addition, the necessary thermal conductivity data are discussed. Results, which do not require the use of a computer, are compared with a computer-generated numerical solution and the agreement is shown to be within a few percent.