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Education, Training & Workforce Development
The Education, Training & Workforce Development Division provides communication among the academic, industrial, and governmental communities through the exchange of views and information on matters related to education, training and workforce development in nuclear and radiological science, engineering, and technology. Industry leaders, education and training professionals, and interested students work together through Society-sponsored meetings and publications, to enrich their professional development, to educate the general public, and to advance nuclear and radiological science and engineering.
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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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High-temperature plumbing and advanced reactors
The use of nuclear fission power and its role in impacting climate change is hotly debated. Fission advocates argue that short-term solutions would involve the rapid deployment of Gen III+ nuclear reactors, like Vogtle-3 and -4, while long-term climate change impact would rely on the creation and implementation of Gen IV reactors, “inherently safe” reactors that use passive laws of physics and chemistry rather than active controls such as valves and pumps to operate safely. While Gen IV reactors vary in many ways, one thing unites nearly all of them: the use of exotic, high-temperature coolants. These fluids, like molten salts and liquid metals, can enable reactor engineers to design much safer nuclear reactors—ultimately because the boiling point of each fluid is extremely high. Fluids that remain liquid over large temperature ranges can provide good heat transfer through many demanding conditions, all with minimal pressurization. Although the most apparent use for these fluids is advanced fission power, they have the potential to be applied to other power generation sources such as fusion, thermal storage, solar, or high-temperature process heat.1–3
J. K. Vaurio, C. Mueller
Nuclear Science and Engineering | Volume 65 | Number 2 | February 1978 | Pages 401-413
Technical Paper | doi.org/10.13182/NSE78-A27167
Articles are hosted by Taylor and Francis Online.
Response surface techniques are presented for obtaining the probability distributions of selected consequences of a liquid-metal fast breeder reactor hypothetical core disruptive accident. The uncertainties of the consequences are considered as a variability of the system and model input parameters used in the accident analysis. Probability distributions are assigned to the input parameters, and parameter values are systematically chosen from these distributions. These input parameters are then used in deterministic consequence analyses that are performed by fast-running analogs of the comprehensive mechanistic accident analysis codes. The results of these deterministic consequence analyses are used to generate the coefficients for response surface functions that approximate the consequences in terms of the selected input parameters. These approximating functions are then used to generate the probability distributions of the consequences with random sampling being used to obtain values for the accident parameters from their distributions. Two different schemes are presented for selecting the knot-point values of the input parameters. The first generates a single second-order polynomial for the entire parameter space; the second generates separate polynomials for specified regions of the parameter space. A technique to handle nonindependent or correlated input parameters is presented. Finally, the calculation of conditional distributions of the consequences and the use of these distributions to define importance distributions of the input parameters are presented. The use of these procedures is illustrated by applications to a postulated loss-of-flow transient with failure to scram in a Clinch River Breeder-type reactor.