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Fuel Cycle & Waste Management
Devoted to all aspects of the nuclear fuel cycle including waste management, worldwide. Division specific areas of interest and involvement include uranium conversion and enrichment; fuel fabrication, management (in-core and ex-core) and recycle; transportation; safeguards; high-level, low-level and mixed waste management and disposal; public policy and program management; decontamination and decommissioning environmental restoration; and excess weapons materials disposition.
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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.
M. Salvatores
Nuclear Science and Engineering | Volume 50 | Number 4 | April 1973 | Pages 345-353
Technical Paper | doi.org/10.13182/NSE73-A26569
Articles are hosted by Taylor and Francis Online.
The benchmark experiments, performed in a plutonium-fueled critical assembly (ZPR-6 Assembly 7) in support of the fast reactor demonstration plant program, were used, in conjunction with a statistical method of correlation, as a source of suggestions for possible modifications to multigroup neutron cross-section sets based on ENDF/B files. Version 1 of these files was a starting point in the correlation procedure. The general trends of the results (lower 238U capture cross section and 235U fission cross section, higher 239Pu capture to fission ratio, with respect to ENDF/B Version 1 data) are discussed, and the suggested multigroup cross-section adjustments are compared with recent evaluations. Calculations of keff, and of the ratios of 238U and 235U fission rate and of 238U capture rate to 239Pu. fission rate, using the proposed adjusted set, give results in good agreement with the experiments.
