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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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Latest News
Wyoming OKs construction of TerraPower’s Natrium plant
Progress continues for TerraPower’s Natrium plant, with the latest win coming in the form of a state permit for construction of nonnuclear portions of the advanced reactor.
A. Radkowsky, A. Dayan, A. Y. Temkin, L. Green
Nuclear Science and Engineering | Volume 75 | Number 3 | September 1980 | Pages 265-274
Technical Paper | doi.org/10.13182/NSE80-A19058
Articles are hosted by Taylor and Francis Online.
The optimum 235U enrichment of the uranium fuel for a once-through cycle for pressurized water reactors (PWRs) is ∼20%. Such an enrichment leads to a core design having the following major advantages in safety, economy, and uranium utilization over present standard designs. 1. There is a reduction in core volume by about a factor of 2, resulting in important savings in costs of core and pressure vessel. 2. Safety will be enhanced as a result of utilization of metallic fuel elements with much greater strength and a factor of 10 better heat conduction and less stored energy than standard ceramic fuel elements. The maximum temperature is 700°F below melting, as compared with 300°F for ceramic fuel. 3. Plutonium discharge is reduced by about a factor of 7. 4. Need for a soluble neutron-absorber control is eliminated. 5. While a detailed core design was beyond the scope of this work, a relatively simple fuel management scheme appears to be feasible which would reduce initial uranium ore requirements by ∼50% of that of standard PWRs and separative work by ∼35% reduce annual usage of uranium ore by ∼15% with a slight increase in separative work.