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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.
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2025 ANS Annual Conference
June 15–18, 2025
Chicago, IL|Chicago Marriott Downtown
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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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Smarter waste strategies: Helping deliver on the promise of advanced nuclear
At COP28, held in Dubai in 2023, a clear consensus emerged: Nuclear energy must be a cornerstone of the global clean energy transition. With electricity demand projected to soar as we decarbonize not just power but also industry, transport, and heat, the case for new nuclear is compelling. More than 20 countries committed to tripling global nuclear capacity by 2050. In the United States alone, the Department of Energy forecasts that the country’s current nuclear capacity could more than triple, adding 200 GW of new nuclear to the existing 95 GW by mid-century.
Robert E. Einziger, James A. Cook
Nuclear Technology | Volume 69 | Number 1 | April 1985 | Pages 55-71
Technical Paper | Nuclear Fuel | doi.org/10.13182/NT85-A33595
Articles are hosted by Taylor and Francis Online.
A whole-rod test was conducted at 229 °C to investigate the long-term stability of light water reactor spent fuel rods with artificially induced defects stored in inert and unlimited air atmospheres. Both boiling water reactor (BWR) and pressurized water reactor (PWR) rods were tested. After 2235 h, visual observations, diametral measurements, and radiographic smears were used to assess the degree of cladding deformation and particulate release. The same examinations plus metallography and x-ray analysis were conducted after 5962 h. Neither of the breached rods tested in inert atmosphere, nor the breached PWR rod tested in unlimited air, showed any measurable change from the pretest condition. The upper defect on the BWR rod tested in unlimited air had a 12.7-mm split after 2235 h and had 10% cladding deformation. The crack grew to 63.5 mm after 5962 h. X-ray analysis indicated that the UO2 had oxidized to U3O8 at this defect. The difference in the behavior of the upper and lower defects is attributed to the accessibility of the air to the fuel due to the positioning of the defect with relation to the pellet/pellet interface.
