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Division Spotlight
Nuclear Criticality Safety
NCSD provides communication among nuclear criticality safety professionals through the development of standards, the evolution of training methods and materials, the presentation of technical data and procedures, and the creation of specialty publications. In these ways, the division furthers the exchange of technical information on nuclear criticality safety with the ultimate goal of promoting the safe handling of fissionable materials outside reactors.
Meeting Spotlight
Conference on Nuclear Training and Education: A Biennial International Forum (CONTE 2025)
February 3–6, 2025
Amelia Island, FL|Omni Amelia Island Resort
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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Article considers incorporation of AI into nuclear power plant operations
The potential application of artificial intelligence to the operation of nuclear power plants is explored in an article published in late December in the Washington Examiner. The article, written by energy and environment reporter Callie Patteson, presents the views of a number of experts, including Yavuz Arik, a strategic energy consultant.
Andrej Prosek, Borut Mavko
Nuclear Technology | Volume 126 | Number 2 | May 1999 | Pages 170-185
Technical Paper | Thermal Hydraulics | doi.org/10.13182/NT99-A2965
Articles are hosted by Taylor and Francis Online.
When best-estimate calculations are performed, the uncertainties need to be quantified. Worldwide, various methods have been proposed for this quantification. Rather than proposing a new uncertainty methodology, a contribution is made to the existing code scaling, applicability, and uncertainty (CSAU) method. A small-break loss-of-coolant accident with the break in the cold leg of a Westinghouse-type two-loop pressurized water reactor was selected for the analysis, and the CSAU methodology was used for uncertainty quantification. The uncertainty was quantified for the RELAP5/MOD3.2 thermal-hydraulic computer code. Some tools suggested by the uncertainty methodology based on accuracy extrapolation (UMAE) method were successfully applied to improve the CSAU methodology, particularly for nodalization qualification. A critical scenario with core uncovery was selected for the analysis, which showed that when uncertainty is added to the peak cladding temperature, the safety margin is sufficient. The tools developed by the UMAE method showed that the structure of the CSAU method is universal because it does not prescribe tools for the analysis.