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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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How to talk about nuclear
In your career as a professional in the nuclear community, chances are you will, at some point, be asked (or volunteer) to talk to at least one layperson about the technology you know and love. You might even be asked to present to a whole group of nonnuclear folks, perhaps as a pitch to some company tangential to your company’s business. So, without further ado, let me give you some pointers on the best way to approach this important and surprisingly complicated task.
C. A. Brandon, G. J. Kidd, Jr.
Nuclear Science and Engineering | Volume 32 | Number 1 | April 1968 | Pages 8-15
Technical Paper | doi.org/10.13182/NSE68-A18818
Articles are hosted by Taylor and Francis Online.
During a series of in-pile experiments designed to study irradiation effects on high-performance oxide fuel elements for advanced gas-cooled reactors, heat-transfer data were obtained from four specially instrumented fuel rods. An annular geometry was utilized with rods of 1.9- and 2.18-cm diam being contained in channels of 2.44- and 2.67-cm diam, respectively. The effects of wire-wrapped and machined square-thread surface roughness were measured and compared with the results obtained from a smooth rod. The fuel rods contained UO2 pellets of varying enrichment and were clad with type-304 stainless-steel tubing. The test parameters for the data reported are: 1) coolant flow rate from 45 to 150 kg/h of helium at 20 atm which corresponds to Reynolds numbers from 15 000 to 45 000; 2) cladding temperatures to 840°C; and 3) heat fluxes from 30 to 100 W/cm2. The smooth-rod data can be correlated with a standard deviation of ±10% by the expression Roughening the rods increased the heat transfer by approximately a factor of 2 with no significant difference between the wire-wrapped and machined roughnesses. The results are generally found to be in good agreement with the results of previous heat-transfer studies. Some consequences of using heat-transfer promoters in nuclear reactor fuel elements are discussed.
