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Nuclear Installations Safety
Devoted specifically to the safety of nuclear installations and the health and safety of the public, this division seeks a better understanding of the role of safety in the design, construction and operation of nuclear installation facilities. The division also promotes engineering and scientific technology advancement associated with the safety of such facilities.
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Conference on Nuclear Training and Education: A Biennial International Forum (CONTE 2025)
February 3–6, 2025
Amelia Island, FL|Omni Amelia Island Resort
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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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Latest News
NRC issues subsequent license renewal to Monticello plant
The Nuclear Regulatory Commission has renewed for a second time the operating license for Unit 1 of Minnesota’s Monticello nuclear power plant.
Robert P. Martin
Nuclear Technology | Volume 175 | Number 3 | September 2011 | Pages 652-662
Technical Paper | NURETH-13 Special / Thermal Hydraulics | doi.org/10.13182/NT175-652
Articles are hosted by Taylor and Francis Online.
This paper describes a general methodology for quantifying the importance of specific phenomenological elements to analysis measures evaluated from nonparametric best-estimate plus uncertainty evaluation methodologies. The principal objective of an importance analysis is to reveal those uncertainty contributors having the greatest influence on key analysis measures. This characterization supports the credibility of the uncertainty analysis, the applicability of the analytical tools, and even the generic evaluation methodology through the validation of the engineering judgments that guided the evaluation methodology development. A demonstration of the importance analysis is provided using data from a sample problem considered in the development of AREVA's realistic large-break loss-of-coolant (LOCA) methodology. The results are presented against the original large-break LOCA phenomena identification and ranking table developed by the technical program group responsible for authoring the code scaling, applicability, and uncertainty methodology.