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Division Spotlight
Thermal Hydraulics
The division provides a forum for focused technical dialogue on thermal hydraulic technology in the nuclear industry. Specifically, this will include heat transfer and fluid mechanics involved in the utilization of nuclear energy. It is intended to attract the highest quality of theoretical and experimental work to ANS, including research on basic phenomena and application to nuclear system design.
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.
B. F. Gore, B. R. Leonard, Jr.
Nuclear Science and Engineering | Volume 53 | Number 3 | March 1974 | Pages 319-323
Technical Note | doi.org/10.13182/NSE74-A23356
Articles are hosted by Taylor and Francis Online.
Calculations have been performed which indicate the possibility of reducing below ten years the effective half-life for transmutation of massive loadings of 137Cs placed in the blanket of a controlled thermonuclear reactor (CTR). The calculations assume the cylindrical “standard blanket” geometry and neutron source (which yields a vacuum wall loading of 10 MW/m2 of 14-MeV neutrons). Significant thermal flux enhancement is obtained by (n,2n) reactions in a beryllium moderator. Gas production and induced radioactivity problems in the beryllium moderator are not much worse than in a graphite moderator. For an 80% target-zone loading of 137Cs, a transmutation rate of 290 kg per year per meter of CTR length is obtained. At this loading, the transmutation rate in roughly 1% of the length of a CTR blanket would balance the production rate in a fission reactor of the same power. Constraint of the CTR source strength to yield a wall loading of 1 MW/m2 would increase the effective half-life for 137Cs to more than 20 years.