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Division Spotlight
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.
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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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.
Wayne K. Lehto
Nuclear Science and Engineering | Volume 39 | Number 3 | March 1970 | Pages 361-367
Technical Paper | doi.org/10.13182/NSE70-A19996
Articles are hosted by Taylor and Francis Online.
The fission cross-section ratios of 239Pu to 235U and 233U to 235U have been measured from 0.24 to 24 keV using parallel-plate, back-to-back fission detectors and a slowing down lead spectrometer as a neutron source. Relative fission rates obtained from the experiments were converted to true fission cross-section ratios by detector intercalibration in a thermal-neutron flux. In the case of the Pu/U ratio, these results extend the data to below several keV where measurements have not previously been made and the 233U/235U results verify previous measurements made in this energy region.
